ML062900401

Safety Analysis Report for Constant Pressure Power Uprate.

ML062900401
Person / Time
Site:	Susquehanna
Issue date:	10/11/2006
From:	Susquehanna
To:	Office of Nuclear Reactor Regulation
References
PLA-6076
Download: ML062900401 (359)

Text

{{#Wiki_filter:Attachment 6 to PLA-6076 Power Uprate Safety Analysis Report PUSAR Non-Proprietary Attachment 6 to PLA-6076 Power Uprate Safety Analysis Report PUSAR Non-Proprietary Version PPL Susquehanna LLC Two North Ninth Street Allentown, PA Susquehanna Steam Electric Station Units 1 and 2 Safety Analysis Report for Constant Pressure Power Uprate October 2006 Non-Proprietary Version Safety Analysis Report for Susquehanna Steam Electric Station Units 1 and 2 Constant Pressure Power Uprate Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.i Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.ii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.iii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.iv Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.V Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.vi Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.vii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Page GLOSSARY xvii EXECUTIVE

SUMMARY

xxvii 1 INTRODUCTION 1-1 1.1 REPORT APPROACH 1-1 1.1.1 Generic Assessments 1-2 1.1.2 Plant-Specific Evaluation 1-3 1.1.3 Report Generation and Review Process 1-3 1.2 PURPOSE AND APPROACH 1-7 1.2.1 Uprate Analysis Basis 1-7 1.2.2 Computer Codes 1-8 1.2.3 Approach 1-12 1.2.4 Concurrent Changes Unrelated to CPPU 1-14 1.3 CPPU PLANT OPERATING CONDITIONS 1-14 1.3.1 Reactor Heat Balance 1-14 1.3.2 Reactor Performance Improvement Features 1-15 1.4

SUMMARY

AND CONCLUSIONS 1-15 2 REACTOR CORE AND FUEL PERFORMANCE 2-1 2.1 FUEL DESIGN AND OPERATION 2-1 2.2 THERMAL LIMITS ASSESSMENT 2-1 2.2.1 Safety Limit MCPR 2-2 2.2.2 MCPR Operating Limit 2-2 2.2.3 MAPLHGR and Maximum LHGR Operating Limits 2-2 2.3 REACTIVITY CHARACTERISTICS 2-3 2.4 STABILITY 2-3 2.5 REACTIVITY CONTROL 2-4 2.5.1 Control Rod Scram 2-5 2.5.2 Control Rod Drive Positioning and Cooling 2-5 2.5.3 Control Rod Drive Integrity Assessment 2-7 3 REACTOR COOLANT AND CONNECTED SYSTEMS 3-1 3.1 NUCLEAR SYSTEM PRESSURE RELIEF/OVERPRESSURE 3-1 PROTECTION

3.2 REACTOR

VESSEL 3-3 3.2.1 Fracture Toughness 3-3 viii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Page 3.2.2 Reactor Vessel Structural Evaluation 3-3 3.3 REACTOR INTERNALS 3-7 3.3.1 Reactor Internal Pressure Differences 3-7 3.3.2 Reactor Internals Structural Evaluation 3-7 3.3.3 Steam Dryer/Separator Performance 3-13 3.4 FLOW INDUCED VIBRATION 3-13 3.4.1 FIV Influence on Piping 3-14 3.4.2 FIV Influence on Reactor Internal Components 3-15 3.5 PIPING EVALUATION 3-17 3.5.1 Reactor Coolant Pressure Boundary Piping 3-17 3.5.2 Balance-Of-Plant Piping 3-23 3.6 REACTOR RECIRCULATION SYSTEM 3-28 3.7 MAIN STEAM LINE FLOW RESTRICTORS 3-29 3.8 MAIN STEAM ISOLATION VALVES 3-31 3.9 REACTOR CORE ISOLATION COOLING/ISOLATION 3-32 CONDENSER 3.10 RESIDUAL HEAT REMOVAL SYSTEM 3-34 3.10.1 Shutdown Cooling Mode 3-36 3.10.2 Steam Condensing Mode 3-36 3.11 REACTOR WATER CLEANUP SYSTEM 3-36 4 ENGINEERED SAFETY FEATURES 4-1 4.1 CONTAINMENT SYSTEM PERFORMANCE 4-1 4.1.1 Containment Pressure and Temperature Response 4-2 4.1.2 Containment Dynamic Loads 4-4 4.1.3 Containment Isolation 4-7 4.1.4 Generic Letter 89-10 Program 4-7 4.1.5 Generic Letter 89-16 4-8 4.1.6 Generic Letter 96-06 4-8 4.2 EMERGENCY CORE COOLING SYSTEMS 4-8 4.2.1 High Pressure Coolant Injection 4-8 4.2.2 High Pressure Core Spray 4-9 4.2.3 Core Spray or Low Pressure Core Spray 4-9 4.2.4 Low Pressure Coolant Injection 4-10 4.2.5 Automatic Depressurization System 4-11 4.2.6 ECCS Net Positive Suction Head 4-12 4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE 4-12 4.4 MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM 4-13 4.5 STANDBY GAS TREATMENT SYSTEM 4-14 ix Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Page 4.6 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL 4-17 SYSTEM 4.7 POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM 4-17 5 INSTRUMENTATION AND CONTROL 5-1 5.1 NSSS MONITORING AND CONTROL 5-1 5.1.1 Neutron Monitoring System 5-1 5.1.2 Rod Worth Minimizer/Rod Control and Information System 5-4.5.2 BOP MONITORING AND CONTROL' 5-4 5.2.1 Pressure Control System 5-5 5.2.2 Turbine Steam Bypass System 5-6 5.2.3 Feedwater Control System 5-7 5.2.4 Leak Detection System 5-7 5.3 TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS 5-9 5.3.1 Main Steam Line High Flow Isolation 5-11 5.3.2 Turbine First-Stage Pressure Scram and Recirculation Pump Trip 5-12 Bypass 5.3.3 APRM Flow Biased Scram 5-12 5.3.4 Rod Worth Minimizer/RCIS Rod Pattern Controller Low Power 5-13 Setpoint 5.3.5 Rod Block Monitor 5-14 5.3.6 RCIS Rod Withdrawal Limiter High Power Setpoint 5-15 5.3.7 APRM Setdown in Startup Mode 5-15 5.4 CHANGES TO INSTRUMENTATION AND CONTROLS 5-15 6 ELECTRICAL POWER AND AUXILIARY SYSTEMS 6-1 6.1 AC POWER 6-1 6.1.1 Onsite Power (degraded voltage) 6-1 6.1.2 Offsite Power (normal operation) 6-2 6.2 DC POWER 6-3 6.3 FUEL POOL 6-3 6.3.1 Fuel Pool Cooling and Clean Up System and RHRFPC Cooling Mode 6-4 6.3.2 Crud Activity and Corrosion Products 6-5 6.3.3 Radiation Levels 6-6 6.3.4 Fuel Racks 6-6 6.4 WATER SYSTEMS 6-6 6.4.1 Service Water Systems 6-7 6.4.2 Main Condenser/Circulating Water/Normal Heat Sink Performance 6-10 6.4.3 Reactor Building Closed Cooling Water System 6-11 6.4.4 Turbine Building Closed Cooling Water System 6-11 X Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Page 6.4.5 Ultimate Heat Sink 6-12 6.5 STANDBY LIQUID CONTROL SYSTEM 6-13 6.6 POWER DEPENDENT HVAC 6-15 6.7 FIRE PROTECTION 6-16 6.7.1 10 CFR 50 Appendix R Fire Event 6-17 6.8 OTHER SYSTEMS AFFECTED BY POWER UPRATE 6-19 7 POWER CONVERSION SYSTEMS 7-1 7.1 TURBINE-GENERATOR 7-1 7.2 CONDENSER AND STEAM JET AIR EJECTORS 7-3 7.3 TURBINE STEAM BYPASS 7-5 7.4 FEEDWATER AND CONDENSATE SYSTEMS 7-5 7.4.1 Normal Operation 7-6 7.4.2 Transient Operation 7-7 7.4.3 Condensate Demineralizers 7-7 8 RADWASTE AND RADIATION SOURCES 8-1 8.1 LIQUID AND SOLID WASTE MANAGEMENT 8-1 8.2 GASEOUS WASTE MANAGEMENT 8-2 8.3 RADIATION SOURCES IN THE REACTOR CORE 8-3 8.4 RADIATION SOURCES IN REACTOR COOLANT. 8-4 8.4.1 Coolant Activation Products 8-5 8.4.2 Activated Corrosion Products and Fission Products 8-5 8.5 RADIATION LEVELS 8-6 8.6 NORMAL OPERATION OFF-SITE DOSES 8-8 9 REACTOR SAFETY PERFORMANCE EVALUATIONS 9-1 9.1 ANTICIPATED OPERATIONAL OCCURRENCES 9-1 9.1.1 Fuel Thermal Margin Events 9-2 9.1.2 Power and Flow Dependent Limits 9-2 9.1.3 Loss of Water Level Events 9-3 9.2 DESIGN BASIS ACCIDENTS 9-4 9.3 SPECIAL EVENTS 9-4 9.3.1 Anticipated Transient Without Scram 9-5 9.3.2 Station Blackout 9-6 9.3.3 ATWS with Core Instability 9-6 10 OTHER EVALUATIONS 10-1 10.1 HIGH ENERGY LINE BREAK 10-1 10.1.1 Steam Line Breaks 10-1 10.1.2 Liquid Line Breaks 10-2 xi Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Page 10.2 MODERATE ENERGY LINE BREAK 10-3 10.3 ENVIRONMENTAL QUALIFICATION 10-3 10.3.1 Electrical Equipment 10-4 10.3.2 Mechanical Equipment With Non-Metallic Components 10-5 10.3.3 Mechanical Component Design Qualification 10-5 10.4 TESTING 10-5 10.5 INDIVIDUAL PLANT EVALUATION 10-7 10.5.1 Initiating Events 10-12 10.5.2 Component Reliability 10-13 10.5.3 Operator Response 10-14 10.5.4 Success Criteria 10-14 10.5.5 External Events 10-21 10.5.6 Shutdown Risk 10-24 10.5.7 PRA Quality 10-27 10.5.8 Integrated Risk Impact 10-29 10.6 OPERATOR TRAINING AND HUMAN FACTORS 10-30 10.7 PLANT LIFE 10-31 10.8 NRC AND INDUSTRY COMMUNICATIONS 10-34 10.9 EMERGENCY AND ABNORMAL OPERATING PROCEDURES 10-34 10.10 NON-TURBINE RELATED MISSILES 10-35 10.10.1 Internally Generated Missiles (Outside Primary Containment) 10-35 10.10.2 Internally Generated Missiles (Inside Primary Containment) 10-36 10.10.3 High Energy Line Breaks (Inside and outside Primary Containment) 10-36 11 REFERENCES 11-1 Tables Table 1-1 Computer Codes Used for CPPU 1-16 Table 1-2 Current and CPPU Plant Operating Conditions 1-20 Table 3-1 P + Q CUFs of Limiting Components 3-38 Table 3-2 RIPDs for Normal Conditions 3-39 Table 3-3 RIPDs for Upset Conditions 3-40 Table 3-4 RIPDs for Emergency Conditions 3-41 Table 3-5 RIPDs for Faulted Conditions 3-42 Table 3-6 Main Steam ASME Class I Piping 3-43 Table 3-7 Feedwater ASME Class I Piping 3-44 Table 3-8 Summary of Governing Stresses for RPV Intemals 3-45 xii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Table 3-9 Table 3-10 Table 3-11 Table 3-12 Table 3-13 Table 3-14 BOP Piping -FW, Condensate, Extraction Steam, FW Heater Drains &Vents Main Steam including Turbine By-Pass and Reactor Feed Pump Turbine Main Steam & HP Steam to RFPT Unit 1 Main Steam and Steam Seal Pipe Supports Unit 2 Main Steam and Steam Seal Pipe Supports Units I & 2 High Pressure Steam To Reactor Feed Pump Turbines Table 4-1 Containment Performance Results Table 4-2 ECCS Performance Results Table 4-3 SGTS Iodine Removal Capacity Table 5-1 SSES Analytical Limits For Setpoints Table 5-2 Changes to Instrumentation and Controls Table 6-1 SSES CPPU Electrical Characteristics Table 6-2 SSES Offsite Electric Power System Table 6-3 UHS Performance Parameter Comparison Table 6-4 Spent Fuel Pool Parameters for CPPU -Batch Offload with Cross-Tied Pools Plus 8.5 MBTU/Hr Background Heat Table 6-5 Spent Fuel Pool Parameters for CPPU -Full-Core Off-Load (EHL)with Cross-Tied Pools Plus 8.5 MBTU/hr Background Heat Table 6-6 SSES Appendix R Fire Event Evaluation Results Table 6-7 Basis for Classification of No Significant Effect Table 8.1 Post Accident Vital Occupancy / Mission Dose Summary Table 9-1 Parameters Used for ATRIUM-10 CPPU Transient Analyses Table 9-2 ATRIUM-1 0 Transient Analysis Results Table 9-3 Key Inputs for ATWS Analysis Table 9-4 Results of ATWS Analysis Table 10-1 High Energy Line Break Table 10-2 Environmental Qualification for CPPU Table 10-3 Results of UNIT I PRA Sensitivity Cases Table 10-4 Results of UNIT 2 PRA Sensitivity Cases Table 10-5 Summary of Changes in Post-Initiator HEPs Due to CPPU Table 10-6 Summary of Changes in Post-Initiator Dependent HEPs Due to CPPU Table 10-7 Increase in SORV Probability Table 10-8 Release Category Changes Due to CPPU Conditions Table 10-9 Estimate of Impact on SSES Fire CDF Due to CPPU Table 10-10 Definition of Importance Levels Page 3-47 3-47 3-47 3-48 3-49 3-50 4-18 4-19 4-20 5-16 5-17 6-20 6-20 6-21 6-22 6-23 6-24 6-25 8-11 9-8 9-9 9-10 9-10 10-36 10-37 10-38 10-39 10-41 10-45 10-45 10-46 10-48 10-48 xiii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Open 'B' Peer Review Certification Resolution Comparison of CLTP CDF vs. CPPU CDF by Initiator Comparison of CLTP CDF vs. CPPU CDF by Sequence FAC Parameter Comparison for CPPU Table 10-11 Table 10-12 Table 10-13 Table 10-14 Page.10-49 10-51 10-52 10-54 Figures Figure 1-1 Power/Flow Operating Map for CPPU Figure 1-2 CPPU Heat Balance -Nominal (@ 100% Power and 100% Core Flow)Figure 1-3 CPPU Heat Balance -Nominal (@ 102% Power and 108% Core Flow)Figure 1-4 CPPU Heat Balance -Overpressure Protection Analysis (@ 102%Power and 108% Core Flow)Figure 3-1 MSIV Closure With Flux Scram at 102%P/108%F -Power, Heat Flux, and Flows Figure 3-2 MSIV Closure With Flux Scram at 102%P/108%F -Downcomer Water Level Figure 3-3 MSIV Closure With Flux Scram at 102%P/108%F -Pressures Figure 3-4 MSIV Closure With Flux Scram at 102%P/108%F -MSRV Position Figure 3-5 MSIV Closure With Flux Scram at 102%P/108%F -MSRV Flow Figure 3-6 MSIV Closure With Flux Scram at 102%P/108%F -Reactivities (Ak/k)Figure 3-7 MSIV Closure With Flux Scram at 102%P/108%F -Core Inlet Enthalpy Figure 4-1 Time-Integrated Containment Hydrogen Generation at CPPU Figure 4-2 Controlled and Uncontrolled Oxygen Concentrations Figure 4-3 Controlled and Uncontrolled Hydrogen Concentrations 1-21 1-22 1-23 1-24 3-51 3-51 3-52 3-52 3-53 3-53 3-54 4-22 4-23 4-24 9-11 9-11 9-12 9-12 9-13 9-13 9-14 Figure 9-1 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F -Power, Heat Flux, and Flows Figure 9-2 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 1 00%P/99%F -Downcomer Water Level Figure 9-3 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at I 00%P/99%F -Pressures Figure 9-4 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 1 00%P/99%F -MSRV Position Figure 9-5 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F -MSRV Flows Figure 9-6 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at I 00%P/99%F -Reactivities (A k/k)Figure 9-7 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at I 00%P/99%F -Core Inlet Enthalpy xiv Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Figure 9-8 Figure 9-9 Figure 9-10 Figure 9-11-Figure 9-12 Figure 9-13 Figure 9-14 Figure 9-15 Figure 9-16 Figure 9-17 Figure 9-18 Figure 9-19 Figure 9-20 Figure 9-21 Figure 9-22 Figure 9-23 Figure 9-24 Figure 9-25 Figure 9-26 Figure 9-27 Figure 9-28 Figure 9-29 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F -Power, Heat Flux, and Flows Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 1 00%P/1 08%F -Downcomer Water Levels Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F -Pressures Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F -MSRV Position Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at I 00%P/1 08%F -MSRV Flows Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 1 00%P/1 08%F -Reactivities (Ak/k)Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at I 00%P/1 08%F -Core Inlet Enthalpy Feedwater Controller Failure Maximum Demand at 100%P/108%F -Power, Heat Flux, and Flows Feedwater Controller Failure Maximum Demand at 1 00%P/1 08%F -Downcomer Water Level Feedwater Controller Failure Maximum Demand at 1 00%P/1 08%F -Pressures Feedwater Controller Failure Maximum Demand at I 00%P/1 08%F -MSRV Position Feedwater Controller Failure Maximum Demand at 1 00%P/1 08%F -MSRV Flows Feedwater Controller Failure Maximum Demand at I 00%P/1 08%F -Reactivities (Ak/k)Feedwater Controller Failure Maximum Demand at I 00%P/l 08%F -Core Inlet Enthalpy Feedwater Controller Failure Maximum Demand with TBV-OOS at I 00%P/1 08%F -Power, Heat Flux, and Flows Feedwater Controller Failure Maximum Demand with TBV-OOS at 1 00%P/l 08%F -Downcomer Water Level Feedwater Controller Failure Maximum Demand with TBV-OOS at 100%P/108%F -Pressures Feedwater Controller Failure Maximum Demand with TBV-OOS at I 00%P/1 08%F -MSRV Position Feedwater Controller Failure Maximum Demand with TBV-OOS at 100%P/108%F -MSRV Flows Feedwater Controller Failure Maximum Demand with TBV-OOS at 100%P/108%F -Reactivities (Ak/k)Feedwater Controller Failure Maximum Demand with TBV-OOS at 1 00%P/1 08%F -Core Inlet Enthalpy Key Parameters for Limiting Full-Power Page 9-14 9-15 9-15 9-16 9-16 9-17 9-17 9-18 9-18 9-19 9-19 9-20 9-20 9-21 9-21 9-22 9-22 9-23 9-23 9-24 9-24 9-25 XV Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate,October 2006-Non-Proprietary Version -Table Of Contents Pressure Regulator Downscale Failure Event (PR-OOS)Appendix Activity (Curies) per Full Core of ATRIUM- 10 Fuel Activity (Curies) per Single Assembly of ATRIUM-1 0 Fuel Page A-1 A-2 A-2 A-26 xvi Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Glossary of Terms 7 .... .... Y ": ... ...... ... ....... : ....... ..... .... .. ..7 : ,Term Definition A CPR Delta CPR AP Differential pressure AC Alternating current ACP Activated Corrosion Products ADHR Alternate Decay Heat Removal ADS Automatic Depressurization System AL Analytical Limit ALARA As Low As Reasonable Achievable ANS American Nuclear Society ANSI American National Standards Institute AOO Anticipated operational occurrence (moderate frequency transient event)AOPs Abnormal Operating Procedures AOV Air-operated valve AP/JR Annulus pressurization/jet reaction APRM Average Power Range Monitor ARI Alternate Rod Insertion ASDC Alternate Shutdown Cooling ASME American Society of Mechanical Engineers ASQL Approved Supplier Quality List AST Alternate Source Term ATWS Anticipated Transient Without Scram AV Allowable Value AVZ Above vessel Zero BHP Brake horse power BIT Boron injection initiation temperature BOL Beginning of life BOP Balance-of-plant BWR Boiling Water Reactor BWROG BWR Owners Group BWRVIP BWR Vessel and Internals Project xvii Term BYPOOS CDF CFD CFR CGCS CHF CLTP CLTR CO COLR CPPU CRD CRDA CRDH CREOASS CRGT CRHE CRWE CS CSC CSS CST CUF DBA DC DHR DIR DLO DOR DRF Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Glossary of Terms Definition Turbine Bypass Out of Service Core damage frequency Condensate filter demineralizer Code of Federal Regulations Combustible Gas Control System Critical Heat Flux Current Licensed Thermal Power Constant Pressure Power Uprate Licensing Topical Report Condensation oscillation Core Operating Limits Report Constant Pressure Power Uprate Control Rod Drive Control Rod Drop Accident Control Rod Drive Housing Control Room Emergency Outside Air Supply System Control Rod Guide Tube Control Room Habitability Envelope Control Rod Withdrawal Error Core Spray Containment Spray Cooling Core Support Structure Condensate storage tank.Cumulative usage factors Design basis accident Direct current Decay heat removal Design Input Request Dual (recirculation) loop operation Division of Responsibility Design Record File xviii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -TJerm DTR ECCS EFPY EHA EHL EOC EOP EPRI EQ ESF ESW F&Os FAC FANP FEA FFWTF FHA FIV FLIM FPC FPCC FPCCS FSAR FTR FW FWCF FWHO(FWLB GE GL Glossary of Terms Definition Draft Task Report Emergency Core Cooling System Effective full power years Equipment Handling Accident Emergency heat load)End of cycle Emergency Operating Procedure(s) Electric Power Research Institute Environmental qualification Engineered Safety Feature Emergency Service Water Facts and Observations Flow Accelerated Corrosion Framatome ANP Inc.finite element analyses.Final Feedwater Temperature Reduction Fuel Handling Accident Flow induced vibration Failure likelihood index methodology Fuel Pool Cooling Fuel Pool Cooling and Cleanup Fuel Pool Cooling and Cleanup System Final Safety Analysis Report Final Task Report Feedwater Feedwater Control Failure)S Feedwater heater out of service FW line break General Electric Company Generic Letter xix Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Glossary of Terms fTerm Definition

• GNF Global Nuclear Fuel LLC GPM Gallons per minute HCR Human cognitive reliability HCTL Heat Capacity Temperature Limit HELB High Energy Line Break HEP Human error probability HEPA High Efficiency Particulate Air Hga Inches of mercury absolute HP High Pressure HPCI High Pressure Coolant Injection HPSP High Power Setpoint HPT High-pressure turbine Hr hour HRA Human Reliability Analysis HTSP High Trip Setpoint HVAC Heating Ventilating and Air Conditioning HX Heat exchanger IASCC Irradiation-assisted stress corrosion cracking ICA Interim Corrective Action ICF Increased Core Flow IEBs Inspection and Enforcement Bulletins IEEE Institute of Electrical and Electronics Engineers IGSCC Inter-granular stress corrosion cracking IHSI Induction Heat Stress Improvement ILBA Instrument Line Break Accident INPO Institute of Nuclear Power Operations IORV Inadvertent Opening of Relief Valve IPE Individual Plant Examination IPEEE Individual Plant Examination of External Events IPSP Intermediate Power Setpoint XX Term IRM ISI ISLOCA ISP ITSP LBLOCA Lbm LCS LDS LERF LFWH LHGR LOCA LOFW LOOP LP LPCI LPCS LPRM LPSP LRNB LTSP MAAP MAPLHGR MBTU MCPR MCPRf MCPRp MELB MELLLA Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Glossary of Terms Definition Intermediate Range Monitor In-Service Inspection Interfacing System LOCA Integrated surveillance program Intermediate Trip'Setpoint Large break LOCA Pounds of mass Leakage Control System Leak Detection System Large early release frequency Loss of Feedwater Heating Linear Heat Generation Rate, Loss-Of-Coolant Accident Loss of feedwater Loss of Offsite Power Low Pressure Low Pressure Coolant Injection Low Pressure Core Spray Local Power Range Monitor Low Power Setpoint Load Rejection No Bypass Low Trip Setpoint Modular accident analysis program Maximum Average Planar Linear Heat Generation Rate Millions of BTUs Minimum Critical Power Ratio Flow-dependent MCPR Limits Power-dependent MCPR Limits Moderate Energy Line Break Maximum Extended Load Line Limit Analysis xxi Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -[,Term,, MeV Mlbm MOV MS MSIP MSIV MSIVC MSIVF MSL MSLB MSLBA MSRV MSVV MVA MVAr MWe MWR MWt NA NDE NEI NPSH NRC NSSS NUREG OLMCPR OLTP OOS OPRM P/T Glossary of Terms Definition Million Electron Volts Millions of pounds mass Motor-operated valve Main steam Mechanical Stress Improvement Process Main Steam Isolation Valve Main Steam Isolation Valve Closure MSIV closure with scram on high flux Main steam line Main Steam Line Break Main Steam Line Break Accident Main steam relief valve Main steam valve vault Million Volt Amps Megavar Megawatts-electric Metal-water reaction Megawatt-thermal Not Applicable Non-Destructive Examination Nuclear Energy Institute Net positive suction head Nuclear Regulatory Commission Nuclear steam supply system NRC technical report designation (Nuclear Regulatory Commission)

Operating Limit Minimum Critical Power Ratio Original Licensed Thermal Power Out-of-service Oscillation Power Range Monitor Pressure and Temperature xxii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Term P 2 5 PCS PCT PICSY PPL PRA PR PRFO PRNMS PSA PSF Psi Psia Psid Psig PSL PWP RB RBCCW RBM RCIC RCIS RCPB RCW RE RFP RFPT RG RHR RHRFPC Glossary of Terms Definition 25% of CPPU Rated Thermal Power Pressure Control System Peak cladding temperature Plant Integrated Computer System PPL Susquehanna LLC Probabilistic Risk Assessment Pressure Regulator Pressure Regulator Failed Open Power Range Neutron Monitor System Probabilistic Safety Analysis Performance-shaping factor Pounds per square inch Pounds per square inch -absolute Pounds per square inch -differential Pounds per square inch -gauge Pressure Suppression Limit Project Work Plan Reactor Building Reactor Building Closed Cooling Water Rod Block Monitor Reactor Core Isolation Cooling Rod Control and Information System Reactor Coolant Pressure Boundary Raw Cooling Water Responsible Engineer Reactor Feed Pump Reactor Feed Pump Turbine Regulatory Guide Residual Heat Removal Residual Heat Removal Fuel Pool Cooling xxiii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Glossary of Terms[Term Definition RHRSW Residual Heat Removal Service Water RIPD Reactor internal pressure difference(s) RLAR Reload Licensing Analysis Report RPS" Reactor Protection System RPT Recirculation Pump Trip RPTOOS Recirculation Pump Trip Out of Service RPV Reactor Pressure Vessel RRS Reactor Recirculation System RSLB Recirculation system line break RTNDT Reference temperature of nil-ductility transition RTP Rated Thermal Power RWCU Reactor Water Cleanup RWE Rod withdrawal error RWM Rod Worth Minimizer Salt CPPU alternating stress intensity SAMG Severe Accident Management Guidelines SAR Safety Analysis Report SBLOCA Small Break LOCA SBO Station blackout SDC Shutdown Cooling SER Safety Evaluation Report SF-LPCI Single failure low-pressure coolant injection valve.SFP Spent Fuel Pool SFWPT Single Feedwater Pump Trip SGTS Standby Gas Treatment System SILs Service Information Letters SJAE Steam Jet Air Ejectors SLC Standby Liquid Control SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio xxiv Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Glossary of TermsDefinition SLO Single-loop operation Sm Code allowable stress limit SMA Seismic margins assessment SMT Scale model testing SORV Stuck-open relief valve SPC Suppression Pool Cooling SPCB Framatome's Critical Power Correlation SPDS Safety Parameter Display System SRM Source Range Monitor SRV Safety relief valve(s)SRVDL Safety relief valve discharge line SSC Systems Structures and Components SSES Susquehanna Steam Electric Plant SSP Supplemental surveillance capsule program SV Safety .Valve SW Service Water System TAF Top of active fuel TBCCW Turbine Building Closed Cooling Water TEDE Total Effective Dose Equivalent T-G Turbine-generator TIP Traversing In-core Probe TLO Two Loop Operation TOP Axial power shape -top-peaked. TS Technical Specifications TSD Task Scoping Document TSSS Technical Specification Scram Speed TSV Turbine Stop Valve TTNB Turbine Trip No Bypass Tw Time available UHS Ultimate heat sink XXV Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Glossary of Terms Termp Definition USE Upper shelf energy xxvi Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -EXECUTIVE

SUMMARY

This report summarizes the results of all significant safety evaluations performed that justify uprating the licensed thermal power at Susquehanna Steam Electric Station (SSES) Units I and 2. The requested license power level is an increase to 120% Original Licensed Thermal Power (OLTP), 3952 MWt, from the Current Licensed Thermal Power (CLTP) of 3489 MWt.GE has previously developed and implemented Extended Power Uprate using, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32424P-A, February 1999 (ELTRI) and "Generic Evaluations for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32523P-A, February 2000 (ELTR2). Based on the Extended Power Uprate experience, GE has developed an approach to uprate reactor power that maintains the current plant maximum normal operating reactor dome pressure. This approach is referred to as Constant Pressure Power Uprate (CPPU) and is contained in NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," referred to as the CLTR.Some topics in this CPPU report are directly fuel dependent because the fuel type affects the resulting evaluation or the consequences of transients or accidents. Because SSES contains only Framatome ATRIUM-I10 fuel, this CPPU report does not reference the CLTR as the basis for areas involving reactor systems and fuel issues, consistent with the NRC's Conditions and Limitations on the use of the CLTR. The fuel dependent evaluations are performed by PPL or Framatome using approved codes and methods. The ATWS analysis, which is a fuel dependent evaluation, is performed by GE.For the fuel independent evaluations, this report provides a systematic application of the CLTR approach to SSES, including the performance of plant specific engineering assessments and confirmation of the applicability of the CLTR generic assessments required to support a CPPU.It is not the intent of. this report to explicitly address all the details of the analyses and evaluations described herein. For example, only previously NRC-approved or industry-accepted methods were used for the analyses of accidents and transients, as referred to in the CLTR.Therefore, the safety analysis methods have been previously addressed, and thus, are not explicitly addressed in this report. Also, event and analysis descriptions that are already provided in other licensing reports or the Final Safety Analysis Report (FSAR) are not repeated within this report. This report summarizes the significant evaluations needed to support a licensing amendment to allow for uprated power operation. Uprating the power level of nuclear power plants can be done safely within plant-specific limits and is a cost-effective way to increase installed electrical generating capacity. Many light water reactors have already been uprated worldwide, including many BWR plants.An increase in the electrical output of a BWR plant is accomplished primarily by generating and supplying higher steam flow to the turbine-generator. A higher steam flow is achieved by increasing the reactor power along specified control rod and core flow lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are xxvii Safety Analysis Report for Susquehanna Units I .and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -recalibrated. Plant procedures are revised and power ascension testing is performed. Modifications to some non-safety power generation equipment will be implemented, as needed.Detailed evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, environmental issues, and design basis accidents were performed. This report demonstrates that SSES can safely operate at the requested CPPU level. However, non-safety power generation modifications must be implemented in order to obtain the electrical power output associated with the uprate power. Until these modifications are completed, the non-safety balance of plant equipment may limit the electrical power output, which in turn may limit the operating thermal power level to less than the rated thermal power (RTP) level. These modifications have been evaluated-and they do not constitute a material alteration to the plant, as discussed in 10 CFR 50.92.The evaluations and reviews were conducted in accordance with the CLTR, or the PPL and Framatome codes and methods as approved by the NRC. The results of these evaluations and reviews are presented in the succeeding sections of this report." All safety aspects of SSES that are affected by the increase in thermal power were evaluated;

• Evaluations were performed using NRC-approved or industry-accepted analysis methods;" No changes, which require compliance with more recent industry codes and standards, are being requested;" The FSAR will be updated for the CPPU related changes, after CPPU is implemented, per the requirements in 10 CFR 50.71(e);" No new design functions that require modifications are necessary for safety related systems, and any modification to power generation equipment will be implemented per 10 CFR 50.59;" Systems and components affected by CPPU were reviewed to ensure there is no significant challenge to any safety system;" Compliance with current SSES environmental regulations were reviewed;* Potentially affected commitments to the NRC have been reviewed; and* Planned changes not yet implemented have also been reviewed for the effects of CPPU.The SSES licensing requirements have been reviewed, and it is concluded that this CPPU can be accommodated (1) without a significant increase in the probability or consequences of an accident previously evaluated, (2) without creating the possibility of a new or different kind of accident from any accident previously evaluated, and (3) without exceeding any existing regulatory limits applicable to the plant, which might cause a significant reduction in a margin of safety. Therefore, the requested CPPU does not involve a significant hazards consideration.

The environmental assessment accomplished for SSES demonstrates that while the proposed increase in capacity results in minor increases in the environmental effects, the CPPU can be accommodated without changes to existing permit limitations. The environmental assessment considered plant effects such as increased temperature of the circulating water leaving the plant, xxviii Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -alternative power sources, plant modifications required to implement CPPU, spent fuel storage, and low level radioactive waste.xxix Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -1. INTRODUCTION

1.1 REPORT

APPROACH Uprating the power level of nuclear power plants can be done safely within certain plant-specific limits. Most GE BWR plants have the capability and margins for an uprate of 5 to 20%without major NSSS hardware modifications. Many light water reactors have already been uprated worldwide. Over a thousand MWe have already been added by uprate in the United States. Several BWR plants are among those that have already been uprated. This evaluation justifies a CPPU to 3952 MWt (120% OLTP) which corresponds to 113% of CLTP.This report follows the generic format and content for CPPU licensing reports, as described in the CLTR (Reference 1).GE has previously developed and implemented Extended Power Uprate using Reference 2,"Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32424P-A, February 1999 (ELTRI) and Reference 3, "Generic Evaluations for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32523P-A, February 2000 (ELTR2). Based on the Extended Power Uprate experience, GE has developed an approach to uprate reactor power that maintains the current plant maximum normal operating reactor dome pressure. This approach is referred to as CPPU and is contained in Reference 1, NEDC-33004P-A, Revision I and "Constant Pressure Power Uprate," referred to as the CLTR.Some topics in this report are fuel dependent because the fuel type affects the resulting evaluation or the consequences Of transients or accidents. Because SSES contains only Framatome (FANP) ATRIUM-10 fuel, this CPPU report does not reference the CLTR as the basis for areas involving reactor systems and fuel issues, consistent with the NRC's Conditions and Limitations on the use of the CLTR. *Most fuel dependent evaluations have been performed by PPL Susquehanna or Framatome using approved codes and methods. The ATWS analysis is a fuel dependent evaluation that was performed by GE.For the non-fuel dependent evaluations, this report provides a systematic application of the CLTR approach to SSES, including the performance of plant specific engineering assessments and confirmation of the applicability of the CLTR generic assessments required to support a CPPU. The format and content for CPPU licensing reports, as described in the CLTR, is used for the non-fuel dependent evaluations. For non-fuel dependent evaluation, this report provides a similar approach to evaluations in the fuel dependent subject areas and utilizing plant specific engineering assessments to support the CPPU. The overall scope of evaluations and the level of detail provided in this report are generally consistent with that provided in previous BWR power uprate requests. Methods and criteria for acceptability used in the fuel dependent evaluations are based on methods and criteria that are specifically applicable to ATRIUM-10 fuel using approved methods and codes utilized by the fuel vendor (Framatome). It is not the intent of this report to explicitly address all the details of the analyses and evaluations described herein. For example, only previously NRC approved or industry accepted methods were used for the analyses of accidents and transients, as 1-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -referred to in the LTRs. Therefore, the safety analysis methods have been previously addressed, and thus, are not explicitly addressed in this report. Also, event and analysis descriptions that are already provided in other licensing reports or the Final Safety Analysis Report (FSAR) are not repeated within this report. This report summarizes the significant evaluations needed to support a licensing amendment to permit uprated power operation.

1.1.1 Generic

Assessments Many of the component, system, and performance evaluations contained within this report have been generically evaluated in the CLTR, and found to be acceptable. The plant-specific applicability of these generic assessments is identified and confirmed in the applicable sections of this report. Generic assessments are those safety evaluations that can be dispositioned for a group or all BWR plants by:* A bounding analysis for the limiting conditions," Demonstrating that there is a negligible effect due to CPPU, or* Demonstrating that the required plant cycle specific reload analyses are sufficient and appropriate for establishing the CPPU licensing basis.Bounding analyses may be based on either a demonstration that previous pressure increase power uprate assessments provided in Reference 2 or 3 (ELTRI and ELTR2, respectively) are bounding or on specific generic studies provided in the CLTR. For these bounding analyses, the current CPPU experience is provided in the CLTR along with the basis and results of the assessment. For those CPPU assessments having a negligible effect, the current CPPU experience plus a phenomenological discussion of the basis for the assessment is provided in the CLTR. For assessments that are fuel design dependent, methodologies that are applicable to the ATRIUM-10 fuel design, were utilized.Some of the evaluations affected by CPPU are fuel operating cycle (reload) dependent. Reload dependent evaluations require that the reload, fuel design, core loading pattern, and operational plan be established so that analyses can be performed to establish core operating limits. The reload analysis demonstrates that the core design for CPPU meets the applicable requirements of TS 5.6.5, COLR.The reload fuel design and core loading pattern dependent plant evaluations for CPPU operation will be performed with the reload analysis as part of the standard reload licensing process. PPL Susquehanna cannot implement a power uprate unless the appropriate reload core analysis is performed and all applicable requirements of TS 5.6.5 are satisfied. Based on current requirements, the reload analysis results are documented in the Reload Licensing Analysis Report (RLAR), and the applicable core operating limits are documented in the plant specific Core Operating Limits Report (COLR).1.1.2 Plant-Specific Evaluation Plant-specific evaluations are assessments of the principal evaluations that are not addressed by the generic assessments described in Section 1.1.1. The effect of CPPU on the plant-specific evaluations and the methods used for their performance are provided in this report. Where 1-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -applicable, the assessment methodology is referenced. If a specific computer code is used, the name of this computer code is provided in the subsection. If the computer code is identified in Reference 1, 2, or 3, these documents may be referenced rather than the original report. Table 1 -iprovides a summary of the computer codes used.The plant-specific evaluations performed and reported in this document use plant-specific values to model the actual plant systems, transient response, and operating conditions. These plant-specific analyses are performed using a equilibrium cycle using ATRIUM 10 fuel representative of SSES design for operation at 120% of OLTP for a cycle length of 24 months.1.1.3 Report Generation and Review Process 1.1.3.1 GE Scope This PUSAR represents several years of project planning activities, engineering analysis, technical verification, and technical customer review. The final stages of the PUSAR preparation include PUSAR integration, additional customer review, on-site and offsite review committee review, and submittal to NRC. The PPL Susquehanna CPPU project relied on the generic power uprate licensing topical reports (References 1, 2, and 3) submitted to and approved by NRC.The project begins with the respective GE and PPL Susquehanna Project Managers creating a Project Work Plan (PWP). This PWP, developed in accordance with GE engineering

• procedures, was used to define the plant specific work scope, inputs and outputs required for Project activities.

A Division of Responsibility (DOR) between PPL Susquehanna and GE was used to further develop the work scope and assign responsible engineers (REs) from each organization. A Task Scoping Document (TSD) applicable for each GE task was created, reviewed, and approved by PPL Susquehanna. Each GE task RE submitted a Design Input Request (DIR) to the PPL Susquehanna task RE interface to define the correct plant information for use in the GE task analysis and evaluation. The information requested by the DIR was supplied by the PPL Susquehanna RE. The DIR then received an independent review by a PPL Susquehanna engineer and the DIR was approved by PPL Susquehanna engineering supervision. Additional DIRs were submitted as the project continued. Each subsequent DIR and each revision to a DIR is subjected to the same prepare, review and approval process. A plant specific PUSAR "shell" was created that contains the appropriate depth of information (but not the specifics) expected in the final PUSAR.Pertinent GE task information is captured in an individual Design Record File (DRF) maintained by GE with oversight of the respective engineering manager. The respective engineering manager oversees the preparation and the quality of the DRF. Each DRF contains the Quality Assurance records applicable to the task, including evidence of design verification. A Draft Task Report (DTR) was created for every GE task; the DTR includes a description of the analysis performed, inputs, methods, results obtained, and includes input to the applicable PUSAR section(s). The DTR is design verified, in accordance with the GE Quality Assurance Program, by a GE technical verifier and a GE Regulatory Services verifier, with oversight by the 1-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -responsible GE technical manager and GE Project Manager. The DTR was transmitted by the GE Project Manager to PPL Susquehanna and reviewed by the PPL Susquehanna RE and other PPL Susquehanna engineers, and subject matter experts, as appropriate. The PPL Susquehanna task RE then compiled the comments and provided an integrated set of comments to GE.Subsequent comments are resolved between the GE and the PPL Susquehanna RE and a Final Task Report (FTR) is developed. The FTR is again design verified in accordance with the GE Quality Assurance Program, by a GE technical verifier and a GE Regulatory Services verifier, with oversight by the responsible GE technical manager and GE Project Manager. The GE Project Manager transmits the FTR to the customer. Subsequently PPL Susquehanna formally accepted the FTR. The PPL Susquehanna RE then compiles the DIR, TSD and FTR into a PPL Susquehanna calculation in accordance with NEPM-QA-0221 (Preparation of Engineering Calculations) and enters the calculation into the PPL Susquehanna records management system as a quality record.For the SSES CPPU, PPL Susquehanna personnel:

1. Conducted multidisciplinary technical reviews of GE evaluation reports (DTRs and FTRs);2. Provided technical review results, in the form of detailed comments, to GE performers;

3. Participated in discussions with GE REs to address and resolve comments; and 4. Controlled the application of the off-site services process to GE.5. Compiled and entered the task documents as a PPL Susquehanna quality record (calculation).

The Regulatory Services RE integrated the individual PUSAR sections creating a Draft PUSAR that was design verified, in accordance with the GE Quality Assurance Program, by another GE Regulatory Services engineer, with oversight by the GE Regulatory Services Manager and the GE Project Manager. The GE Project Manager transmitted the verified Draft PUSAR to PPL Susquehanna where it received another complete review by PPL Susquehanna's technical personnel, project staff, and Licensing staff.PPL Susquehanna personnel generated questions and comments, which were responded to by GE's technical and Regulatory Services personnel. A Technical assessment of GE's work was performed during a design assessment review conducted at the GE offices in Wilmington, NC on February 1 through February 3, 2006. The scope of these assessments included work performed by GE Nuclear Energy, and Global Nuclear Fuels (GNF) in support of the PPL Susquehanna CPPU project. Participating in those activities were representatives of PPL Susquehanna mechanical/structural, nuclear, and reactor engineering disciplines as well as a representative of Performance Improvement and Oversight. The PPL Susquehanna team reviewed design inputs, analysis methodologies, and results in the GE Design Record Files. The reviews included discussion with GE technical task performers to obtain a thorough understanding of GE analysis methods and a review of completed DRFs.1.1.3.2 Framatome (FANP) Scope The CPPU project began with the respective FANP and PPL Susquehanna Project Managers creating a "Contract" developed in accordance with FANP and PPL Susquehanna procedures, 1-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -which was used to define the specific workscope, inputs and outputs required for various project activities. Also, the "Contract" divided various project activities into "tasks" and assigns specific schedules for each of the tasks. The next step was accomplished by each FANP task responsible engineer by submitting a Design Input Request (DIR) to the PPL Susquehanna task interface to define the correct plant information for use in each specific task analysis and evaluation. The information requested by the DIR was supplied by the PPL Susquehanna RE.The DIR then received an independent review by a PPL Susquehanna engineer and the DIR was approved by PPL Susquehanna engineering supervision. Additional DIRs were submitted as the project continued. Each subsequent DIR and each revision to a DIR is subjected to the same preparation, review and approval process.Analyses supporting various tasks were performed and documented in calculation notebooks, in accordance with the FANP Quality Assurance Program, by a qualified engineer, with oversight from the responsible engineer for each task. The relevant results of the analyses supporting various tasks were submitted to PPL Susquehanna as a draft report. Subsequent comments are resolved between the FANP and the PPL Susquehanna responsible engineers and a final task report was developed and submitted to PPL Susquehanna. The PPL Susquehanna RE then compiles the DIR and FTR into a PPL Susquehanna calculation in accordance with NEPM-QA-0221 (Preparation of Engineering Calculations) and enters the calculation into the PPL Susquehanna records management system as a quality record.Subsequently, the PUSAR input was structured and developed to meet the format requirements of the plant and fuel safety analysis report templates provided by PPL Susquehanna which are based on the format of Reference

1. A Draft PUSAR input, reviewed in accordance with the FANP Quality Assurance Program, was provided to PPL Susquehanna.

PPL Susquehanna personnel generated questions and comments, which were responded to by FANP's technical and regulatory personnel. After all questions/comments were resolved the final PUSAR input was submitted to PPL Susquehanna. A technical assessment of FANP's work was performed during a design assessment review conducted at the Framatome offices in Richland, Washington on January 9 through January 13, 2006. The scope of these assessments included work performed by Framatome in support of the PPL Susquehanna CPPU project. Participating in those activities was the PPL Susquehanna lead nuclear engineer. The PPL Susquehanna lead nuclear engineer reviewed design inputs, analysis methodologies, and results in the FANP Design Record Files. The reviews included discussion with FANP technical task performers to obtain a thorough understanding of FANP analysis methods and a review of completed DRFs.1.1.3.3 PPL Susquehanna Scope As noted in Section 1.1.3.1 above a Division of Responsibility (DOR) between PPL Susquehanna and GE was used to further develop the work scope and assign responsible engineers (REs) from each organization. Tasks assigned to PPL Susquehanna responsible engineers were performed under the PPL Susquehanna 10 CFR 50, Appendix B Quality Assurance Program as specified in engineering procedure NEPM-QA-0001 (Nuclear Design 1-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Engineering Procedure Program). The PPL Susquehanna assigned tasks were performed internally by PPL Susquehanna engineers or contracted out to engineering consulting firms on the PPL Susquehanna approved supplier quality list (ASQL). Contractors were chosen based on areas of critical expertise and previous experience with CPPU projects. Where applicable, the contractors applied a CFR 50, Appendix B Quality Assurance Program. All contractor 10CFR 50, Appendix B Quality Assurance Programs were reviewed and accepted by PPL Susquehanna prior to obtaining CPPU analysis results.All PPL Susquehanna internal tasks were prepared in accordance with calculation procedure NEPM-QA-0221 (Preparation of Engineering Calculations) and reviewed and approved per procedure NEPM-QA-0241 (Review Design Verification & Approval of Design Documents). In addition to the engineering review required by NEPM-QA-0241 (Review Design Verification & Approval of Design Documents), a broad based inter departmental review was performed. Upon approval of PPL Susquehanna internally prepared tasks, the calculations are retained as quality records in the PPL Susquehanna nuclear records management system.For PPL Susquehanna responsible tasks which were subcontracted to engineering consulting firms, a work scope was developed by PPL Susquehanna and the engineering consulting firm.This work scope then formed the mutually agreed upon basis for the CPPU task. The engineering consultant then forwarded a list of design inputs to PPL Susquehanna. The design inputs were then collected reviewed and forwarded to the engineering consultant. Draft task reports were created which included a description of the analysis performed, inputs, methods, results obtained, and also included input to the applicable PUSAR section(s). The draft task report was reviewed by the PPL Susquehanna RE and other PPL Susquehanna engineers, and subject matter experts, as appropriate. An integrated set of comments on the draft task reports were forwarded for comment resolution and incorporation into the final task report. The engineering consultant then transmits the final task report along with the applicable PUSAR section(s) to PPL Susquehanna. Upon final acceptance of the engineering consultants work, the PPL Susquehanna RE then compiles the engineering contractor documentation into a PPL Susquehanna calculation in accordance with NEPM-QA-0221 (Preparation of Engineering Calculations) and enters the calculation into the PPL Susquehanna records management system as a quality record.1.1.3.4 Overall PPL Susquehanna Project Control Although the support of several vendors was utilized in performing various technical activities, PPL Susquehanna maintained overall control of the work activities and retains responsibility for the quality of the project. PPL Susquehanna asserted its control, as well as oversight function, through various means including control of the scope of individual task activities, control, and approval of appropriate design input information, concurrence with evaluation approaches and final acceptance and approval authority for the work activities performed. Task activity results were provided by the various vendors and received multi-disciplinary technical reviews by PPL Susquehanna. Comments and questions regarding the results and approaches were provided by PPL Susquehanna to the task vendors. Appropriate resolution of 1-6 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -the comments and questions were required after which the results were provided to PPL Susquehanna. Upon formal acceptance by PPL Susquehanna of the task results, the reports including appropriate supporting documentation were processed through the PPL Susquehanna calculation system in accordance with NEPM-QA-0241 (Review Design Verification & Approval of Design Documents) for retention as quality records.For tasks performed directly by PPL Susquehanna, the activities were controlled in accordance with PPL Susquehanna's 10 CFR 50, Appendix B Quality Assurance Program and the results are documented and retained as PPL Susquehanna Engineering Calculations in accordance with NEPM-QA-0221 (Preparation of Engineering Calculations), and NEPM-QA-0241 (Review Design Verification & Approval of Design Documents).

1.2 PURPOSE

AND APPROACH An increase in electrical output of a BWR is accomplished primarily by generation and supply of higher steam flow to the turbine generator. Most BWRs, as originally licensed, have an as-designed equipment and system capability to accommodate steam flow rates at least 5% above the original rating. In addition, continuing improvements in the analytical techniques (computer codes) based on several decades of BWR safety technology, plant performance feedback, operating experience, and improved fuel and core designs have resulted in a significant increase in the design and operating margin between the calculated safety analyses results and the current plant licensing limits. The available margins in calculated results, combined with the as-designed excess equipment, system, and component capabilities have: 1) allowed many BWRs to increase their thermal power ratings by 5% without any Nuclear Steam Supply System (NSSS) hardware modification; and 2) provide for power increases up to 20% with some non-safety hardware modifications. These power increases involve no significant increase in the hazards presented by the plants as approved by the NRC at the original license stage.The method for achieving higher power is to extend the power/flow map (Figure 1-1) along the Maximum Extended Load Line Limit Analysis (MELLLA). However, there is no increase in the maximum normal operating reactor vessel dome pressure or the maximum licensed core flow over their pre-CPPU values. CPPU operation does not involve increasing the maximum normal operating reactor vessel dome pressure, because the plant, after modifications to non-safety power generation equipment, has sufficient pressure control and turbine flow capabilities to control the inlet pressure conditions at the turbine.1.2.1 Uprate Analysis Basis PPL Susquehanna has performed 2 licensed power uprates, and the ARTS/MELLLA operating domain expansion has been submitted to the NRC for approval. The first licensed uprate, termed a Stretch Uprate, increased the licensed thermal power by approximately 4.5%(References 4 and 5). The second licensed uprate of 1.4% was a result of improved instrumentation allowing a reduction in the uncertainty in thermal power, termed an Appendix K 1-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -uprate (Reference 6). The SSES License Amendment for the ARTS/MELLLA operating domain expansion (Reference

7) is expected to be approved by the NRC by November 23, 2006.The key thermal power levels are as follows: " The Original Licensed Thermal Power (OLTP) is 3293 MWt." The Stretch Uprate Licensed Thermal Power is 3441 MWt." The CLTP is the Appendix K Uprate Power, which is 3489 MWt.The CPPU RTP level included in this evaluation is 3952 MWt, which is 120% of the OLTP.Plant specific CPPU parameters are listed in Table 1-2. The power level used in the safety analyses is stated in each section.Fuel dependent analyses presented in this report include the effects of the 2 percent power uncertainty factor discussed in Regulatory Guide 1.49. Most of the analyses were performed at 100% power level and the impact of the two percent power uncertainty factor is accounted for either statistically or through the inherent conservatism of the methodology.

For three of the analyses, (ASME over-pressurization, loss of feedwater flow, and LOCA-ECCS analyses) the effects of the 2 percent power uncertainty factor are not directly included in the methodology used for the analyses; therefore, these three analyses were performed at 102% of CPPU rated power to account for the 2 percent power uncertainty factor.The non-fuel dependent analysis power levels include the Regulatory Guide (RG) 1.49 two percent power uncertainty factor unless a smaller value is specifically justified or the uncertainty is accounted for in the analysis methods.RG 1.49 does not apply to some events that have been historically analyzed from nominal initial conditions, such as the Anticipated Transient Without Scram (ATWS), Station Blackout (SBO)events and, containment new loads evaluations.

1.2.2 Computer

Codes 1.2.2.1 Non-Fuel Dependent Analyses NRC approved or industry accepted computer codes and calculational techniques are used to demonstrate compliance with the applicable regulatory acceptance criteria. The application of these codes to the CPPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The limitations on use of these codes and methods as defined in the NRC staff position letter reprinted in ELTR2 were followed for this CPPU analysis. Any exceptions to the use of the code or conditions of the applicable SER are noted in Table 1-1. The application of the computer codes in Table 1-1 is consistent with the current SSES licensing basis except where noted in this report.1.2.2.2 Fuel Dependent Safety Analyses The computer codes and calculation techniques used in the CPPU analyses have been approved by the U.S. Nuclear Regulatory Commission (NRC) or are accepted throughout the nuclear 1-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -industry. The computer codes used in the ATRIUM-10 CPPU evaluations are listed in Table 1-1. Framatome reload licensing analyses are performed to ensure that fuel design and.operating limits are satisfied for the limiting assembly in the core. The analyses are performed in accordance with the NRC limitations and restrictions included in the applicable SERs. Any exceptions are noted in Table 1-1. The applicability of FANP's methodology and computer codes to CPPU conditions is discussed in the following. The first step in determining the applicability of current licensing methods to CPPU conditions was a review of FANP current BWR topical reports to identify SER restrictions on the BWR methodology. This review identified that there are no SER restrictions on power level or on the parameters most impacted by the increased power level: steam flow, feedwater flow, jet pump M-ratio, and core average void fraction.The second step consists of an evaluation of the differences between current core and reactor conditions and those experienced under CPPU conditions to determine any challenges to the validity of the models. When the reactor power is increased, the resultant impact on operating margin is mitigated to a large extent by a decrease in the limiting assembly radial power factor.This decrease in the limiting assembly radial power factor is necessary because the thermal operating limits (MCPR, MAPLHGR and LHGR) are fairly insensitive to the increase in core thermal power. From this fundamental constraint the following observations may be made about the CPPU operating conditions:

1. The reduction in the hot assembly radial peaking factor leads to a more uniform radial power distribution and consequently a more uniform core flow distribution.

The net result is less flow starvation of the hottest assemblies.

2. With the .flatter radial power distribution, more assemblies and fuel rods are near the thermal limits.3. From a system perspective, there will be higher steam flow and feedwater flow rates.4. With an increase in the average assembly power in the reactor the core pressure drop will increase slightly resulting in a decrease in the jet pump M-ratio for a given core flow rate.5. Core average void fraction will increase.Based on these fundamental characteristics of power uprate, each of the major analysis domains (thermal-hydraulics, core neutronics, transient analysis, LOCA and stability) can be assessed to determine any challenges to CPPU application.

1.2.2.2.1 Thermal-hydraulics Framatome ANP assembly thermal-hydraulic methods are qualified and validated against full-scale heated bundle tests in the KATHY test facility in Karlstein, Germany. The KATHY tests are used to characterize the assembly two-phase pressure drop and CHF performance. This allows the hydraulic models to be verified for Framatome ANP fuel designs over a wide range 1-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -of hydraulic conditions prototypic of reactor conditions. The wide range of test conditions bounds both uprated and non-uprated assembly conditions. In addition, the key physical phenomena (e.g. heat flux, fluid quality and assembly flows) for uprated conditions are within the scope of current reactor experience. This similarity of assembly conditions is further enforced in Framatome analysis methodologies by the imposition of SPCB CHF correlation limits and therefore both current designs and uprated designs must remain within the same parameter space. Because the bundle operating conditions for CPPU are within the envelope of hydraulic test data used for model qualification and operating experience, the hydraulic models used to compute the core flow distribution and local void content remain valid for CPPU conditions. 1.2.2.2.2 Core Neutronics The Framatome neutronic methodologies are characterized by technically rigorous treatment of phenomena and are very well benchmarked (>100 cycles of operation plus gamma scan data for ATRIUM-10). Susquehanna CPPU reactor operating conditions were compared against data from recent operating experience. It was shown that the neutronic codes and methods are applicable to a wide range of operating conditions which includes Susquehanna CPPU conditions. The increased steam flow from power uprate comes from increased power in normally lower power assemblies in the core running at higher power levels. High powered assemblies in uprated cores will be subject to the same LHGR, MAPLHGR, MCPR, and cold shutdown margin limits and restrictions as high powered assemblies in non-uprated cores.Again, this similarity of operating conditions between current and uprate conditions assures that the neutronic methods used to compute the nodal reactivity and power distributions remain valid. Furthermore, the neutronic characteristics computed by the steady-state simulator and used in safety analysis remain valid.1.2.2.2.3 Transient Analysis The phenomena of primary interest for limiting transients in BWRs are void fraction/quality relationships, determination of CHF, pressure drop, reactivity feedbacks and heat transfer correlations. One fundamental validation of the core hydraulic solution is separate effects testing against Karlstein transient CHF measurements. The transient benchmark to time of dryout for prototypic LRNB and Pump Trip transients encompass the transient integration of the continuity equations (including the void-quality closure relation), heat transfer and determination of CHF. Typical benchmarks to Karlstein illustrate that the transient hydraulic solution and application of SPCB result in conservative predictions of the time of dryout.Outside of the core, the system simulation relies primarily on solutions of the basic conservation equations and equations of state. While there are changes to the feedwater flow rate and jet pump M-ratio associated with power uprate, the most significant change is the steam flow rate and the associated impact on the steam line dynamics for pressurization events. The models associated with predicting the pressure wave, are general and have no limitation within the range of variation associated with power uprate.The reactivity feedbacks are validated by a variety of means including initial qualification of advanced fuel design lattice calculations to Monte Carlo results as required by SER restrictions, 1-10 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -steady-state monitoring of reactor operation (power distributions and eigenvalue) and the Peach Bottom 2 turbine trip benchmarks that exhibited a minimum of 5% conservatism in the calculation of integral power.From these qualifications the transient methods remain valid for CPPU.1.2.2.2.4 LOCA Analysis LOCA results are strongly dependent on local (bundle) power and are weakly dependent on core average power. As discussed in previous sections, maximum bundle power is not significantly changed due to power uprate. The parameters associated with power uprate that may impact LOCA results are increased core average initial stored energy, decreased initial coolant inventory, and increased decay heat.BWR LOCA analyses are not sensitive to initial stored energy. During the blowdown phase the heat transfer remains high and the stored energy is removed prior to the start of the heatup phase. The impact of initial inventory differences due to higher core power is analogous to the impact associated with the different size breaks that are analyzed in a break spectrum.Therefore, these aspects do not change the capability of the codes to model LOCA.Consequently, the impact of power uprate on LOCA analysis is primarily associated with the increase in decay heat levels in the core. Decay heat is conservatively modeled using industry standards applied as specified by regulatory requirements. The models used for decay heat calculations are valid for CPPU.From the above discussion and the observation that nodal thermal-hydraulic conditions during CPPU are expected to be within the current operating experience, the LOCA methods remain applicable for CPPU conditions. 1.2.2.2.5 Stability Analysis The flatter radial power profile induced by the power uprate will tend to decrease the first azimuthal eigenvalue separation and result in slightly higher regional decay ratios. These effects are computed by STAIF as it directly computes the channel, global and regional decay ratio and does not rely on a correlation to protect the regional mode.STAIF has been benchmarked against full assembly tests to validate the channel hydraulics from a decay ratio of approximately 0.4 to limit cycles. These benchmarks include prototypic ATRIUM-10 assemblies. From a reactor perspective, STAIF is benchmarked to both global and regional reactor data as late as 1998, and therefore includes current reactor cycle and fuel design elements. This strong benchmarking qualification and the direct computation of the regional mode assure that the impact of the "flatter" core design for power uprate will be reflected in the stability analysis.I-12 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -1.2.2.2.6 Special Events The Appendix R analysis is performed using the approved LOCA analysis codes. Similar to LOCA events, the impact of power uprate on Appendix R analysis is primarily with the increase in decay heat in the core. Decay heat is conservatively modeled using industry standards applied as specified by regulatory requirements. Use of the Appendix K heat transfer correlations and logic is conservative for Appendix R calculations. The review concluded that there are no SER restrictions on FANP methodology that are impacted by CPPU. Because the CPPU core and assembly conditions for the Susquehanna units are equivalent to core and assembly conditions of other plants for which the methodology was benchmarked the FANP methodology (including uncertainties) remains applicable for CPPU conditions.

1.2.3 Approach

The planned approach to achieving the higher power level consists of the change to the SSES licensing and design basis to increase the licensed power level to 3952 MWt, consistent with the approach outlined in the CLTR, except as specifically noted in this report. Consistent with the CLTR, the following plant-specific exclusions are exercised:

• No increase in maximum normal operating reactor dome pressure* No increase to maximum licensed core flow No increase to MELLLA upper boundary requested in reference 7[[The plant-specific evaluations are based on a review of plant design and operating data, as applicable, to confirm excess design capabilities; and, if necessary, identify required modifications associated with CPPU. For specified topics, generic analyses and evaluations in the CLTR or the ELTRs as applicable, demonstrate plant operability and safety. The dispositions are based on a 20% of OLTP increase, which is the requested power increase for SSES. For this increase in power, the conclusions of system/component acceptability stated in the CLTR or the ELTRs are bounding and have been confirmed for SSES. The scope and depth of the evaluation results provided herein are established based on the approach in the CLTR or ELTRs and unique features of the plant. The results of these evaluations are presented in the following sections: (a) Reactor Core and Fuel Performance:

A CPPU equilibrium ATRIUM-10 fuel cycle operating at uprated power conditions was used as the basis for the CPPU analyses. Thermal limits, reactivity characteristics, and stability were evaluated for the equilibrium cycle design and will continue to be evaluated each cycle.(b) Reactor Coolant System and Connected Systems: Evaluations of the NSSS components and systems have been performed at CPPU conditions. These evaluations confirm the acceptability 1-12 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -of the effects of the higher power and the changes in process variables (i.e., increased steam and feedwater flows). Safety-related equipment performance is the primary focus in this report, but key aspects of reactor operational capability are also included. Reactor overpressure protection and reactor internal pressure differences were evaluated for CPPU operation with ATRIUM-10 fuel.(c) Engineered Safety Feature Systems: The effects of CPPU power operation on the Containment, Emergency Core Cooling System (ECCS), Standby Gas Treatment System and other Engineered Safety Features have been evaluated for key events. The evaluations include the containment responses during limiting Anticipated Operational Occurrences (AOOs), special events, ECCS-Loss-Of-Coolant Accident (LOCA), and safety relief valve (SRV) containment dynamic loads. The effects of CPPU operation with ATRIUM-10 fuel on the engineered safety features have been evaluated for key events. The evaluations show that the appropriate acceptance criteria are met with ATRIUM-10 fuel.d) Control and Instrumentation: The control and instrumentation signal ranges and analytical limits (ALs) for setpoints have been evaluated to establish the effects of the changes in various process parameters such as power, neutron flux, steam flow and feedwater (FW) flow.As required, setpoint evaluations have been performed to determine the need for any Technical Specification setpoint changes for various functions (e.g., main steam line high flow isolation setpoints). Local power range monitor calibration and rod block monitor setpoints were evaluated for CPPU operation with ATRIUM-10 fuel.(e) Electrical Power and Auxiliary Systems: Evaluations have been performed to establish the operational capability of the plant electrical power and distribution systems and auxiliary systems to ensure that they are capable of supporting safe plant operation at the CPPU power level.(f) Power Conversion Systems: Evaluations have been performed to establish the operational capability of various non-safety balance-of-plant (BOP) systems and components to ensure that they are capable of delivering the increased power output, and/or the modifications necessary to obtain full CPPU power.(g) Radwaste Systems and Radiation Sources: The liquid and gaseous waste management systems have been evaluated at limiting conditions for CPPU to show that applicable release limits continue to be met during operation at higher power. The radiological consequences have been evaluated for CPPU to show that applicable regulations have been met for the CPPU power conditions. This evaluation includes the effect of higher power level on source terms, on-site doses and off-site doses, during normal operation.(h) Reactor Safety Performance Evaluations: The limiting Final Safety Analysis Report (FSAR) analyses for design basis events have been addressed as part of the CPPU evaluation. Limiting accidents, AOOs, and special events have been analyzed or generically dispositioned and show continued compliance with regulatory requirements. Analyses for CPPU conditions with ATRIUM-10 fuel were performed for the potentially limiting design basis events using limiting conditions for CPPU. The results of the ATRIUM 10 limiting accident and transient analyses show compliance with regulatory requirements. 1-13 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -(i) Additional Aspects of CPPU: High-energy line break (HELB) and environmental qualification evaluations have been performed at bounding conditions for CPPU to show the-continued operability of plant equipment under CPPU conditions. The effects of CPPU on the SSES Probabilistic Risk Assessment (PRA) have been analyzed to demonstrate that there are no new vulnerabilities to severe accidents.

1.2.4 Concurrent

Changes Unrelated to CPPU Consistent with the conditions and limitations on the use of the CLTR, PPL Susquehanna is not requesting a concurrent review of any changes listed among the restrictions applicable to the CLTR.1.3 CPPU PLANT OPERATING CONDITIONS

1.3.1 Reactor

Heat Balance The operating pressure, the total core flow, and the coolant thermodynamic state characterize the thermal hydraulic performance of a BWR reactor core. The CPPU values of these parameters are used to establish the steady state operating conditions and as initial and boundary conditions for the required safety analyses. The CPPU values for these parameters are determined by performing heat (energy) balance calculations for the reactor system at CPPU conditions. The reactor heat balance relates the thermal-hydraulic parameters to the plant steam and FW flow conditions for the selected core thermal power level and operating pressure. Operational parameters from actual plant operation are considered (e.g., steam line pressure drop) when determining the expected CPPU conditions. The thermal-hydraulic parameters define the conditions for evaluating the operation of the plant at CPPU conditions. The thermal-hydraulic parameters obtained for the CPPU conditions also define the steady state operating conditions for equipment evaluations. Heat balances at appropriately selected conditions define the initial and boundary conditions for plant safety analyses.Figure 1-2 shows the CPPU heat balance at 100% of CPPU RTP and 100% rated core flow.Figure 1-3 shows the CPPU heat balance at 102% of CPPU RTP and 108% core flow. Figure 1-4 shows the CPPU heat balance used in the Overpressure Protection Analysis at 102% of CPPU RTP and 108% core flow.Table 1-2 provides a summary of the reactor thermal-hydraulic parameters for the current rated and CPPU conditions. At CPPU conditions, the maximum nominal operating reactor vessel dome pressure is maintained at the current value, which minimizes the need for plant and licensing changes. With the increased steam flow and associated non-safety BOP modifications, the current dome pressure provides sufficient operating turbine inlet pressure to assure good pressure control characteristics. 1-14 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -1.3.2 Reactor Performance Improvement Features The reactor performance improvement features and the equipment allowed to be out-of-service (OOS) are listed in Table 1-2. The use of these performance improvement features and allowing for equipment OOS are permitted during CPPU operation. When limiting, the input parameters related to the performance improvement features or the OOS have been considered in the safety analyses for CPPU, and as applicable, will be also be included in the CPPU reload core analyses. Where appropriate, the evaluations that are dependent upon cycle length are performed for CPPU assuming a 24-month fuel cycle length.1.4

SUMMARY

AND CONCLUSIONS This evaluation has covered a CPPU to 120% of OLTP. The strategy for achieving this higher power is to extend the MELLLA power/flow map region along the upper boundary extension. The SSES licensing requirements have been reviewed to demonstrate how this uprate can be accommodated without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any existing regulatory limits or design allowable limits applicable to the plant which might cause a reduction in a margin of safety. The CPPU described herein involves no significant hazard consideration. 1-15 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 1-1 Computer Codes for CPPU*Task Computer Version or NRC Comments Code 'Revision Approved .________ .....______*____ Nominal Reactor Heat ISCOR 09 Y (2) NEDE-2401 IP Rev. 0 SER Balance Reactor core and CASMO-4 UFEBO1A Y EMF-2158(P)(A) Rev. 0 fuel performance MICROBURN-B2 UMAR05 Y EMF-2158(P)(A) Rev. 0 Safety limit MCPR SAFLIM2 UMAR05 Y ANF-524(P)(A) Rev. 2 Reactor Internal ISCOR 09 Y(2) NEDE-2401 IP Rev. 0 SER Pressure Differences LAMB 07 (3) NEDE-20566P-A TRACG 02 (4) NEDE-32176P, Rev. 2 NEDC-32177P, Rev. 2 NRC TAC No. M90270 RCPB Piping MEI01 N9 (13) Piping Stress and Analysis Program RELAP5 MOD 3.2 (9) Piping Fluid Transient Loads BOP Piping MEI01 N9 (13) Piping Stress and Analysis Program RELAP5 MOD 3.2 (9) Transient Analysis Program Transient analysis MICROBURN-B2 UMAR05 Y EMF-2158(P)(A) Rev. 0 (except Single HTBAL UMAY94 (II)Feedwater Pump trip) XCOBRA UMAR05 Y (15) XN-NF-80-19(P)(A) Vol. 3 Rev. 2 COTRANSA2 UJUL04 Y (16) ANF-913(P)(A) Vol. 1 Rev. I XCOBRA-T UAPR05 Y (16) XN-NF-84-105(P)(A) Vol. 1 RODEX2 UAPR02 Y XN-NF-81-58(P)(A) Rev. 2 Transient analysis PANACEA II Y (5) NEDE-30130-P-A (Single Feedwater ISCOR 09 Y(2) NEDE-2401 IP Rev. 0 SER Pump Trip ODYN 10 Y NEDO-24154-A SAFER 04 (6) NEDC-32424P-A, NEDC-32523P-A, (17), (18), (19)Reactor core stability STAIF UOCT04 Y EMF-CC-074(P)(A) Vol. 4 Rev. 0 RAMONA5-FA UFEB05 (12) BAW-10255(P) Rev. I Anticipated Transient ODYN 10 Y NEDE-24154P-A Supp. I, Vol. 4 Without Scram STEMP 04 (7)PANACEA 11 Y (5) NEDE-30130-P-A TASC 03A Y NEDC-32084P-A Rev. 2 ISCOR 09 Y (2) NEDE-24011-P Rev. 0 SER 1-16 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 1-1 Computer Codes for CPPU*:Task Computer Version or NRC Comments, Code:,, Revision Approved .___" ____ "______ "____._...... Containment System SHEX 05 Y(8)Response M3CPT 05 Y NEDO-10320, Apr. 1971 LAMB 08 (3) NEDE-20566P-A, Sept. 1986 PICSM 01 Y NUREG-0808, August 1981 Annulus Pressurization LAMB 08 (3) NEDE-20566P-A, Sept. 1986-Mass and Energy ISCOR 09 Y(2) NEDE-2401I P Rev. 0 SER Releases Leak Detection System COTTAP 2 0, 1, 2, & 3 13 Compartment Pressurization and COTTAP 4 0 13 Temperature Code Appendix R Fire RELAX UFEB05 (10) EMF-2361(P)(A), Rev.0 Protection HUXY UJAN01 (10) XN-CC-33(P)(A), Rev.1 RODEX2 UAPR02 (10) XN-NF-81-58(P)(A), Rev.2 MAAP Rev. 4 N Version 4.0.5 Reactor Recirculation BILBO 04V NA(I) NEDE-23504, February 1977 System LOCA-ECCS RELAX UFEB05 Y EMF-2361(P)(A) Rev. 0 HUXY UJAN01 Y XN-CC-33(P)(A) Rev. I RODEX2 UAPR02 Y XN-NF-81-58(P)(A) Rev. 2 Ultimate Heat Sink UHSPPL 003-93200 N(14)SPRAY-LP 002-92125 N(14)SPRAY-H 001-86260 N(14)WIND 1.00 N(14)DRIFTLOS 1.00 N(14)HELB COTTAP 4 0 13 Compartment Pressurization and Subcompartment Temperature Code Evaluation Fission Product SAS2H/ORIGEN-S S N Isotope Generation and Depletion Inventory Code Station Blackout MAAP Rev. 4 N Version 4.0.5 Turbine Generator PDMISSILE 1.0 N Verified vendor method PDBURST 1.1 N Verified vendor method Plant Life CHECKWORKS 2.1 N Industry Standard*

• The application of these codes to the CPPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The application of the codes also complies with the SERs for the EPU programs where applicable.(I) Not a safety analysis code that requires NRC approval.

The code application is reviewed and approved by GENE for "Level-2" application and is part of GENE's standard design process. The application of this code has been used in previous power uprate submittals. (2) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-2401 IP Rev. 0 by the May 12, 1978 letter from D.G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of 1-17 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 1-1 Computer Codes for CPPU*ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.(3) The LAMB code is approved for use in ECCS-LOCA applications (NEDE-20566P-A and NEDO-20566A), but no approving SER exists for the use of LAMB in the evaluation of reactor internal pressure differences or containment system response. The use of LAMB for these applications is consistent with the model description of NEDE-20566P-A. (4) NRC has reviewed and accepted the TRACG application for the flow-induced loads on the core shroud as stated in NRC SER TAC No. M90270.(5) The physics code PANACEA provides inputs to the transient code ODYN. The use of PANAC Version I I in this application was initiated following MFN-035-99, S. Richards (NRC) to G. Watford (GE), Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, "GESTAR 11" -Implementing Improved GE Steady State Methods (TA C No. MA 6481), November 10, 1999.(6) The ECCS-LOCA codes are not explicitly approved for Transient or Appendix R usage. The staff concluded that SAFER is qualified as a code for best estimate modeling of loss-of-coolant accidents and loss of inventory events via the approval letter and evaluation for NEDE-23785P, Revision' I, Volume II. (Letter, C.O. Thomas (See NRC) to J.F. Quirk (GE), "Review of NEDE-23785-1 (P), "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volumes I and II", August 29, 1983.) In addition, the use of SAFER in the analysis of long term Loss-of-Feedwater (LOFW) events is specified in the approved LTRs for power uprate: "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32424P-A, February 1999 and "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32523P-A, February 2000. The Appendix R events are similar to the loss of FW and small break LOCA events.(7) The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool heatup. The use of STEMP was noted in NEDE-24222, "Assessment of BWR Mitigation of ATWS, Volume I& II (NUREG-0460 Alternate No. 3) December 1, 1979." The code has been used in ATWS applications since that time. There is no formal NRC review and approval of STEMP or the ATWS topical report.(8) The application of the methodology in the SHEX code to the containment response is approved by NRC in the letter to G. L. Sozzi (GE) from A. Thadani (NRC), "Use of the SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.(9) The RELAP/MOD

3.2 computer

code is described in NUREG/CR-5535 and is an industry accepted thermal-hydraulic code used to simulate single and multiphase transients in nuclear reactor systems..(10) ll (11) HTBAL is not explicitly approved by the NRC but it is a stand-alone version of the heat balance routine included in the NRC-approved MICROBURN-B2 code documented in EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors:. Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999.(12) RAMONA5-FA is currently being used for determining OPRM setpoints at each reload.(13) No NRC Safety Evaluation Report exists but this program was used from the original plant design (OLTP)piping analyses through today and the NRC is aware of its existence and use.(14) These existing codes, as described in the SSES FSAR, have been used historically for the SSES UHS analysis.(15) The approval of XCOBRA is included in the approval of the THERMEX methodology in XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.1-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 1-1 Computer Codes for CPPU*(16) The list of events for which COTRANSA2 and XCOBRA-T can be used was expanded in the clarification acceptance in Letter, S. Richards (NRC) to J.F. Mallay (FANP), "Siemens Power Corporation RE: Request for Concurrence on Safety Evaluation Report Clarifications (TAC No. MA6160)," May 31, 2000.(17) Letter, J.F. Klapproth (GE) to USNRC, Transmittal of GE Proprietary Report NEDC-32950P "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model," dated January 2000 by letter dated January 27, 2000.(18) Letter, S.A. Richards (NRC) to J.F. Klapproth, "General Electric Nuclear Energy (GENE) Topical Reports GENE (NEDC)-32950P and GENE (NEDC)-32084P Acceptability Review," May 24,2000.(19) "SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-Jet Pump Plants," NEDE-30996P-A, General Electric Company, October 1987.1-19 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 1-2 Current and CPPU Plant Operating Conditions Current Licensed Parameter 1aue. CPPU Value i Valuel Thermal Power (MWt) 3489 3952 Vessel Steam Flow (Mlb/hr) 2. 14.437 16.532 Full Power Core Flow Range Mlb/hr 81.9 to 108.0 99.0 to 108.0% Rated 81.9 to 108.0 99.0 to 108.0 Maximum Nominal Dome Pressure (psia) 1050 1050 Maximum Nominal Dome Temperature ('F) 550.5 550.5 Pressure at upstream side of turbine stop valve (TSV) 994 977 (psia)Full Power Feedwater Flow (Mlb/hr) 14.405 16.500 Temperature (OF) 391.0 399.3 Core Inlet Enthalpy (Btu/lb) 3. 524.8 523.1 Notes: I.2.3.Based on current reactor heat balance.At normal FW heating.At 100% core flow conditions. Currently licensed performance improvement features and/or equipment OOS that are included in CPPU evaluations: (a) Single Loop Operation (SLO)(b) Increased Core Flow (ICF) (108%= 108 Mlbm/hr)(c) Recirculation Pump Trip Out of Service (RPTOOS)(d) Turbine Bypass Out of Service (BYPOOS)(e) 24 Month Cycle (1) 3% SRV Setpoint tolerance (g) Pressure Regulator Out of Service.(PR-OOS) 1-20 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Core Flow (Mlbfhr)0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 120 110 100 90 90 70 60 50 40 0.9 l100 EPU Power 3952 MWt l loot CLT

• 3489 Mt loot core Plow 1 00 r4 A: 48.3t Power/ 30.0 Flow B: 57.9% Power/ 41.7% Flow C: B8.3% Power/ 81.9% Flow D: 100.0% Power/ 99.01 Flow E: 100.01 Power/100.0%

Flow F: 100.0t Power/108.0t Flow G: 88.3% Power/108.0t Flow H: 19.2t Power/108.0% Flow 1: 19.2% Power/100.0% Flow J: 19.2% Power/ 41.1% Flow... ... .. .. .. ........ .. .. ... ..Naa CA........Natural; Circulation -.---..C 0.9 30 20 20--_ _____F_0 I0 20 30 40 50 60 Core Flow (%)70 80 90 100 110 120 Figure 1-1 Power/Flow Operating Map for CPPU 1-21 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Legend#= Flow, Ibm/hr H = Enthalpy, Btu/ibm F = Temperature, *F M = Moisture, %P = Pressure, psia*Conditions at upstream side of TSV Core Thermal Power Pump Heating Cleanup Losses Other System Losses Turbine Cycle Use 3952.0 7.5-4.5-2.8 3952.2 MWt Figure 1-2 CPPU Heat Balance -Nominal (@ 100% Power and 100% Core Flow)1-22 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Legend# = Flow, lbm/hr H = Enthalpy, Btu/Ibm F = Temperature, *F M = Moisture, %P = Pressure, psia*Conditions at upstream side of TSV Core Thermal Power Pump Heating Cleanup Losses Other System Losses Turbine Cycle Use 4031.0 9.4-4.5-2.8 4033.1 MWt Figure 1-3 CPPU Heat Balance -Nominal (@ 102% Power and 108% Core Flow)1-23 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Flow = 146.300 Ibm/hr AEnthalpy = -106.0 Btu/lbm Core Thermal Power 4031.0 Pump Heating 9.5 Cleanup Losses -4.5 Other System Losses -2.8 Turbine Cycle Use 4033.2 MMt Figure 1-4 CPPU Heat Balance -Overpressure Protection Analysis (@ 102% Power and 108% Core Flow)1-24 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -2. REACTOR CORE AND FUEL PERFORMANCE This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapter 4, applicable to the SSES CPPU. The CPPU evaluations assumed an equilibrium core of FANP's ATRIUM-10 fuel type. The reload process will result in a plant-specific RLAR and COLR.2.1 FUEL DESIGN AND OPERATION Reactor core and fuel operation at CPPU conditions with ATRIUM-10 fuel is described below.The average bundle power for CPPU conditions will be 5.17 MWt/bundle. For each cycle of operation the fuel lattice and core design will be modified to meet the operating requirements for that cycle. These changes might include enrichment and burnable absorber distributions, and reload batch size. The core loading is designed such that the power distributions throughout the cycle provide margin to the following operating limits.MCPR -Ensures an acceptable low probability of fuel cladding failure resulting from the fuel experiencing boiling transition.

• LHGR -Ensures that fuel mechanical design bases are met.* MAPLHGR -Ensures that peak cladding temperature (PCT) and metal-water reaction (MWR) criteria for the limiting LOCA are met.The evaluations performed to assess operation at CPPU conditions with ATRIUM-10 fuel assume an equilibrium core of ATRIUM-10 fuel. Fuel and core design limits are met by the planned enrichment, burnable absorber, and control rod positions in the reload core design. The methods used to perform the ATRIUM-10 fuel and core design analyses in this report have been approved by the NRC and show that the equilibrium cycle core of ATRIUM-10 meets the specified CPPU operating requirements while remaining within the operating limits. The ATRIUM-10 reload core designs for operation at CPPU conditions will take into account the operating limits discussed above (MCPR, LHGR, and MAPLHGR) to ensure acceptable design margins exist between the licensing limits and their corresponding operating values. The NRC approved exposure limits are not exceeded in the equilibrium core design used in the CPPU evaluations.

The percent power level above which fuel thermal margin monitoring is required will change with CPPU. The original plant operating licenses set this monitoring threshold at a typical value of 25% of RTP.For SSES, the fuel thermal monitoring threshold is established at 23% of CPPU RTP. A change in the fuel thermal monitoring threshold also requires a corresponding change to the Technical Specification reactor core safety limit for reduced pressure or low core flow.2.2 THERMAL LIMITS ASSESSMENT Assurance that regulatory limits are not exceeded during AOOs and postulated accidents is accomplished by applying operating limits on the fuel. This section discusses the impact the 2-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -transition to uprated power conditions has on thermal limits. The evaluations were performed using an ATRIUM-10 CPPU equilibrium core. Consistent with the current practice, cycle-specific thermal limits are established or confirmed each reload based on the cycle-specific core configuration.

2.2.1 Safety

Limit MCPR The SLMCPR can be affected by the transition to a higher power due to changes in the power and flow distributions. In addition, differences in plant-related uncertainties will impact the SLMCPR results. The SLMCPR analysis reflects the actual core loading and is performed for each reload core. An analysis 'of the SLMCPR for the CPPU equilibrium core of ATRIUM-1 0 was performed. The FANP standard channel bow data base was used in the analyses.The results of the analysis of the SLMCPR for the CPPU equilibrium core support an SLMCPR of 1.07 for Susquehanna CPPU/MELLLA conditions. The final SLMCPR is determined by the reload analysis.2.2.2 MCPR Operating Limit The OLMCPR is determined each cycle based on the results of the reload transient analyses.The OLMCPR for a given core design depends on the critical power performance of the fuel design used and the operating conditions of the core at any given time during the operation (initial core power and core flow conditions, equipment out of service, etc.). For the CPPU equilibrium core, and rated CPPU conditions, the OLMCPR for CPPU operation is shown in Table 9-2.The OLMCPR is established to protect the sum of the change in MCPR (ACPR) for the limiting AOO and the SLMCPR. The impact that the uprated power has on AOO events at CPPU conditions is addressed in Section 9.1.2.2.3 MAPLHGR and Maximum LHGR Operating Limits LOCA-ECCS analyses are performed to demonstrate that the MAPLHGR limits provide the necessary protection. With FANP methods, MAPLHGR operating limits are established for a fuel type (e.g. ATRIUM-10 fuel) at a given plant. Analyses are performed each reload cycle to ensure that established MAPLHGR limits are applicable to the new fuel assembly design. The results presented in Section 4.3 show that the ATRIUM-10 MAPLHGR limits meet the regulatory limits. The ATRIUM-10 MAPLHGR limits are the same for CLTP and CPPU conditions. The LHGR limits ensure that the plant does not exceed the thermal-mechanical design limits of the fuel. LHGR limits are fuel type dependent and apply regardless of power level, and thus are not affected by CPPU. To support operation at off-rated conditions, power- and flow-dependent multipliers are applied to the LHGR limits to ensure that the fuel meets the thermal-mechanical limits during AOO events. While the LHGR limits for ATRIUM-10 fuel are not cycle-specific, the power- and flow-dependent LHGR multipliers are established each cycle becasue they are 2-2 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -affected by the core response during a transient. The LHGR operating limits and the power- and flow-dependent multipliers are addressed in Section 9.1.2.3 REACTIVITY CHARACTERISTICS Reload core design analyses are performed on a cycle-specific basis to ensure that required reactivity margins are maintained. Current technical specification requirements for cold shutdown margin are maintained at CPPU conditions with ATRIUM-10 fuel by appropriate design of the enrichment and burnable neutron absorber content of the fuel lattices and by judicious placement of fresh and irradiated assemblies in the core. Operation with ATRIUM-10 fuel at CPPU conditions does not change cold shutdown margin requirements. Current TS reactivity control requirements at the most reactive conditions of the core are met and confirmed using cycle-specific analyses.Cycle length and hot excess reactivity are maintained by appropriate selection of initial enrichment, fresh batch size, and burnable neutron absorber design. Sufficient design flexibility exists with the ATRIUM-10 fuel to accommodate operation at CPPU conditions while maintaining adequate power distribution control.The Susquehanna fuel storage calculations conservatively assume that the maximum reactivity lattice extends over the entire length of the fuel assembly; therefore, the CPPU assemblies remain bounded by the current fuel storage criticality analyses.2.4 STABILITY Susquehanna has installed a power range neutron monitoring system with oscillation power range monitors to implement the BWROG Long-Term Stability Solution Option-Ill. This system is designed to provide for an automatic scram for the reactor when power oscillations above the system setpoint are detected.The Option III automatic scram is provided by the Oscillation Power Range Monitor (OPRM).The generic analyses for the Option III hot channel oscillation magnitude and the OPRM hardware were designed to be independent of core power. 1The Option III trip is armed only when plant operation is within the Option III trip-enabled region. The Option III trip-enabled region is defined as the region on the power/flow map with power > 30% of OLTP and core flow < 60% of rated core flow (expanded as needed). For CPPU, the Option III trip-enabled region is rescaled to maintain the same absolute power/flow 2-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -region boundaries. Because the rated core flow is not changed, the 60% core flow boundary is not rescaled. The 30% of OLTP boundary changes by the following equation: CPPU Region Boundary = 30% OLTP * (100% + CPPU (% OLTP))Thus, for a 120% of OLTP CPPU: CPPU Region boundary = 30% OLTP * (100% + 120%) = 25% CPPU When the OPRM system is inoperable, the plant may use an alternate stability detect and suppress method. Current practice with the Option-III system is to use the stability.ICAs as the backup method. The ICAs include specific requirements for operator action as well as restrictions on operation in certain regions of the power/flow map. These ICA regions are validated on a cycle-specific basis using FANP's STAIF methodology (Reference

8) and expanded as necessary.

The OPRM system utilizes setpoints to ensure the SLMCPR is not exceeded during a postulated power oscillation. The OPRM setpoint is evaluated for each reload design (including transition cycles) on a cycle-specific basis.2.5 REACTIVITY CONTROL The CRD system is used to control core reactivity by positioning neutron absorbing control rods within the reactor and to scram the reactor by rapidly inserting withdrawn control rods into the core. No change is made to the control rods due to the CPPU. The effect on the nuclear characteristics of the fuel is discussed in Section 2.3. The topics addressed in this evaluation are: Topic CLTRDisposition SSES Result 2.5.1 Scram Time Response [[2.5.2 CRD Positioning 2.5.2 CRD Cooling 2.5.3 CRD Integrity 2-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -2.5.1 Control Rod Scram1For pre-BWR/6 plants, the scram times are decreased by the transient pressure response, [[3] At normal operating conditions, the CRD Hydraulic Control Unit accumulator supplies the initial scram pressure and, as the scram continues, the reactor becomes the primary source of pressure to complete the scram. 2.5.2 Control Rod Drive Positioning and Cooling Er* Reactor dome pressure is unchanged for CPPU. 2-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -, and the automatic operation of the system flow control valve maintains the required drive water pressure and cooling water flow rate. Therefore, the CRD positioning and cooling functions are not affected. The CRD cooling and normal CRD positioning functions are operational considerations, not safety-related functions, and are not affected by CPPU operating conditions. 2-6 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Plant operating data has confirmed that the CRD system flow control valve operating position has sufficient operating margin. [[2.5.3 Control Rod Drive Integrity Assessment E[*1-]]The postulated abnormal operating condition for the CRD design assumes a failure of the CRD system pressure-regulating valve that applies the maximum pump discharge pressure to the CRD mechanism internal components. This postulated abnormal pressure bounds the ASME reactor overpressure limit. [[]] Other mechanical loadings are addressed in Section 3.3.2 of this report.Er 2-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -3. REACTOR COOLANT AND CONNECTED SYSTEMS This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapter 5, and to a very limited extent Chapter 3, that applies to CPPU.3.1 NUCLEAR SYSTEM PRESSURE RELIEF/OVERPRESSURE PROTECTION The nuclear system pressure relief system topics addressed'in this evaluation are as follows: Topic " CLTR Disposition SSES Result, Overpressure capacity Flow-induced vibration SSES plant procedure addresses these characteristics. [[]]I]]The nuclear system pressure relief system prevents over pressurization of the nuclear system during AOOs, the plant ASME Upset overpressure protection event, and postulated ATWS events. The plant SRVs along with other functions provide this protection.' An evaluation was performed in order to confirm the adequacy of the pressure relief system for CPPU conditions. The adequacy of the pressure relief system is also demonstrated by the overpressure protection evaluation performed for each reload core and by the ATWS evaluation performed for CPPU (Section 9.3.1).For SSES, no SRV setpoint increase is needed because there is no change in the dome pressure or simmer margin. Therefore, there is no effect on valve functionality (opening/closing). Over pressurization analyses performed at CPPU conditions for an ATRIUM-10 core demonstrate the adequacy of the pressure relief system. The COTRANSA2 transient code (Reference

31) was used for this analysis.

Compliance with the ASME pressure vessel code criteria is demonstrated for each reload core.The transition to uprated power conditions does not affect the main steam relief valves (MSRV)pressure setpoints; therefore, there is no effect on the MSRV functionality (opening/closing). 3-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, Ociober 2006-Non-Proprietary Version -The design pressure of the reactor vessel is not affected by the transition to uprated power and remains 1250 psig. Per the ASME code, the acceptance limit for pressurization events is 110%of the design pressure, or 1375 psig, for the reactor vessel. Over pressurization analyses using a equilibrium core of ATRIUM 10 were performed for the main steam isolation valve (MSIV)closure and turbine trip with turbine bypass failure events. The events were analyzed at 102%of CPPU rated thermal power (RTP) and an initial dome pressure of 1050 psig (1064.7 psia).The MSRV setpoints presented in Table 9-1 were used with 2 MSRV (with the lowest setpoint)assumed out of service. No credit was taken for the MSIV or turbine stop valve position scram.The results show that the MSIV closure with scram on high flux (MSIVF) is the limiting overpressure event. The calculated peak vessel pressure at the bottom of the vessel is 1328 psig.The corresponding calculated peak dome pressure is 1298 psig. The results remain below the 1375 psig ASME peak vessel limit and the Technical Specification 1325 psig Safety Limit. The results of the limiting ATRIUM-10 overpressure analysis are presented in Figure 3-1 through Figure 3-7.The SRVs are dual function Crosby 6R1 0 HP 65 BP direct acting valves., Each of these valves provides a pneumatically assisted relief function, as well as a single stage direct acting spring (safety) function.SRV tolerance is independent of CPPU. The appropriate CPPU evaluations (discussed above)are performed using the existing SRV safety setpoint tolerance AL of+/- 3% as a basis. The SRV Technical Specification safety function setpoint tolerance of 3% is unchanged for CPPU.Actual historical in-service surveillance test results of SRV performance are monitored for compliance to the Technical Specification requirements. To-date, of 261 "as found" SRV lift setpoint verification tests performed from 1985 to 2005, only 2 SRV tests were found to exceed the current +3% setpoint tolerance, and 13 tests were found to lift below the -3% setpoint tolerance. Thus, the in-service surveillance testing of the plant's SRVs have not shown a significant propensity for high setpoint drift greater than 3%.In general, higher main steam flows may increase incidents of SRV leakage due to FIV. The vibration frequency, extent, and magnitude depend upon plant-specific parameters, valve locations, valve design, and piping support arrangements. However, Strouhal number calculations, which have been confirmed via plant-specific 1/5th scale model testing (SMT), indicate that over the entire increased power range, up to and exceeding the CPPU levels, FIV is not expected to result in unacceptable excitation of the SRV standpipes, and will not result in unexpected SRV leakage. Nonetheless, SSES currently has procedures which address a leaking SRV.Increased main steam line (MSL) flow may affect flow-induced -vibration of the piping and SRVs during normal operation. However, FIV on the Crosby direct acting SRV design is unlikely to result in an inadvertent SRV opening or a "stuck open" SRV, as this valve design has not exhibited a propensity to unexpectedly open during normal'plant operation. Nonetheless, SSES off-normal operating procedures address such an event. Further, the consequences of a stuck open SRV have been previously considered in the plant-specific safety analyses and have been demonstrated to be non-limiting. 3-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The flow-induced vibration of the piping is addressed by vibration testing during initial plant operation at the higher steam flow rates (see Sections 3.4.1 and 10.4). This testing includes the direct vibration monitoring of an SRV. [[3.2 REACTOR VESSEL The RPV structure and support components form a pressure boundary to contain the reactor coolant and moderator, and form a boundary against leakage of radioactive materials into the drywell. The RPV also provides structural support for the reactor core and internals. The topics addressed in this evaluation are: Topic CLTRDisposition SSES Result 3.2.1 Fracture Toughness [[3.2.2 Reactor Vessel Structural Evaluation (Components not significantly affected)3.2.2 Reactor Vessel Structural Evaluation (Affected components)

3.2.1 Fracture

Toughness In accordance with the provisions of I OCFR50.90, PPL Susquehanna submitted a request for amendment to Technical Specification 3.4.10 "RCS Pressure and Temperature (P/T) Limits" for the Susquehanna SES Units I & 2 in October of 2005 (Reference 9). The Pressure/Temperature curves presented in that submittal account for the CPPU operating conditions up to 3952 MWt.Subsequently, in March 2006 (Reference

52) the NRC approved license amendments that changed the Units I and 2 Technical Specification 3.4.10 and bound the CPPU operating conditions.

The maximum normal operating dome pressure for CPPU is unchanged from that for original power operation. Therefore, the hydrostatic and leakage test pressures are acceptable for the CPPU. Because the vessel is still in compliance with the regulatory requirements, operation with CPPU does not have an adverse effect on the reactor vessel fracture toughness.

3.2.2 Reactor

Vessel Structural Evaluation [[l 3-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version 4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -]]Certain reactor vessel components are generically dispositioned without detailed structural analysis. For components with no increase in flow, temperature, reactor internal pressure difference (RIPD), or other mechanical loads, no further evaluation is required. In addition, previous GE BWR power uprate experience has demonstrated that, using the evaluation process documented in Appendix I of ELTR1, numerous components are below the [[ ]] CUF criteria, thus requiring no further evaluation. Because the Recirculation Outlet Nozzle (N-1), Recirculation Inlet Nozzle (N-2), Core Spray Nozzle (Nozzle/Shell Junction)(N-5);Top Head Spray and Spare Nozzles (N-6), Vent Nozzle (N-7), Jet Pump Instrumentation Nozzle (N-8), CRD-HSR Nozzle (N-9); Instrumentation Nozzles (N-11, N-12, N 16), Seal Leak Detection Nozzle (N-13), Drain Nozzle (N 15), CRD Penetration Stub Tube, Shroud Support, Refueling Bellows Support, and In-Core Housing Penetration components have fatigue usages less than or equal to [[ ]] or were originally exempt from fatigue requirements or were qualified by comparison to another component that has been evaluated, these components are confirmed to be consistent with the generic disposition provided in the CLTR, thus requiring no further evaluation. The IRM/SRM/LPRM and Dry Tube are considered life-limited components, which are replaced as required. These original components were previously evaluated using design OLTP conditions, which bound the CPPU operating conditions. The effect of CPPU was evaluated to ensure that the reactor vessel components continue to comply with the existing structural requirements of the ASME Boiler and Pressure Vessel Code. For the components under consideration, the 1968 Code with addenda to and including Summer 1970, which is the Code of construction, was used as the governing Code with the following exceptions: " The 1971 Edition of the ASME Boiler and Pressure Vessel Code formally incorporates the simplified elastic-plastic analysis procedures, which were commonly used in conjunction with the 1968 Code." Code Cases related to the materials of construction of the vessels include 1141-1, 1332-5, 1401, 1420, 1441-1, and 1492.However, if a component's design has been modified and/or re-evaluated, the governing Code for that component was the Code used in the stress analysis of the modified component. The following components were modified and/or re-evaluated since the original construction: 3-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Feedwater Nozzle (N-4): This nozzle was modified *and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda through Summer 1976.CRD-HSR Nozzle Cap: The CRD-HSR nozzle was modified (capped) and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section Ill, 1974 Edition with Addenda through Winter 1975." Recirculation Inlet Nozzle (N-2): The Recirculation Inlet nozzle was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section Ill, 1974 Edition with Addenda through Summer 1976." In-Core Housing Penetration: The In-Core Housing Penetration was analyzed and the governing Code for the evaluation is the ASME Boiler and Pressure Vessel Code, Section III,.1968 Edition with Addenda through Summer 1970." IRM/SRM/LPRM and Dry Tube: These components were evaluated using the 1971 Edition with Addenda through Summer 1973.Typically, new stresses are determined by scaling the "original" stresses based on the CPPU conditions (pressure, temperature, and flow). The analyses were performed for the design, the normal and upset, and the emergency and faulted conditions. There are no changes in annulus pressurization, jet reaction, pipe restraint or fuel lift loads.3.2.2.1 Design Conditions Because there are no changes in the design conditions due to CPPU, the design stresses are unchanged and the Code requirements are met.3.2.2.2 Normal and Upset Conditions The reactor coolant temperature and flows (except core flow) at CPPU conditions are only slightly changed from those at current rated conditions. Evaluations were performed at conditions that bound the slight change in operating conditions. The evaluations are mainly reconciliation of the stresses and usage factors to reflect CPPU conditions. A primary plus secondary stress analysis was performed showing CPPU stresses still meet the requirements of the ASME Code, Section III, Subsection NB for all components. Lastly, the fatigue usage was evaluated for the limiting location of components with a usage factor greater than 0.5. The fatigue analysis results for the limiting components are provided in Table 3-1. The analysis results for CPPU show that all components meet their ASME Code requirements. 3.2.2.3 Emergency and Faulted Conditions The stresses due to Emergency and Faulted conditions are based on loads such as peak dome pressure, which are unchanged. These loads remain unchanged and bound the CPPU values.Therefore, Code requirements are met for all RPV components. 3-6 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -3.3 REACTOR INTERNALS The reactor internals include core support structure (CSS) and non-core support structure (non-CSS) components. The topics considered in this section are: Topic CLTR Disposition SSES Result: 3.3.1 Reactor Internals Pressure Differences

3.3.2 Reactor

Internals Structural Evaluation

3.3.3 Steam

Dryer Separator Performance

3.3.1 Reactor

Internal Pressure Differences The increase in core average power alone would result in higher core loads and reactor internals pressure differences (RIPDs) due to the higher core exit steam quality. The maximum acoustic and flow-induced loads, following a postulated recirculation system line break (RSLB), were shown to be unaffected by the CPPU.The RIPDs are calculated for Normal (steady-state operation), Upset, Emergency, and Faulted conditions for all major reactor internal components. [[Tables 3-2 through 3-5 compare results for the various loading conditions between original analysis results and operation with CPPU for the vessel internal that are affected by the changed RPIDs. All RIPDs, with the exception of fuel channel wall RIPDs, are bounded by the GE14 fuel analysis. The fuel channel wall RIPDs presented in Tables 3-2 through 3-5 are calculated for the Atrium 10 fuel pressure drop. The Atrium 10 fuel channel RIPDs are acceptable for all CPPU conditions with exceptions for the 80-mil fuel channels. The 80-mil fuel channels will be discharged prior to the implementation of the full CPPU level. In addition, the 80-mil fuel channels will be operated in core locations with lower radial power peaking during the last cycle at partial CPPU levels such that the resulting differential pressures satisfy the pressure load limit.3.3.2 Reactor Internals Structural Evaluation The RPV internals consist of the Core Support Structure (CSS) components and non-Core Support Structure components. The RPV internals (except the Control Rod Drive Mechanism) are not certified to the ASME code; however, the requirements of the ASME Code are used as guidelines in their design basis analysis. The evaluation/stress reconciliation in support of the CPPU is performed consistent with the design basis analysis of the components. The RPV internal components evaluated in this section are: 3-7 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Core Support Structure Components

• Shroud* Shroud Support* Core Plate* Top Guide* Control Rod Drive Housing* Control Rod Guide Tube" Orificed Fuel Support" Fuel Channel Non-Core Support Structure Components" Steam Dryer* FW Sparger* Jet Pump Assembly* Core SprayLine and Sparger" Access Hole Cover* Shroud Head and Steam Separator Assembly (including Shroud Head Bolts)" In-core Housing and Guide Tube* Vessel Head Cooling Spray Nozzle* Jet Pump Instrument Penetration Seal* Core Differential Pressure and Liquid Control Line* Control Rod Drive Mechanism The original configurations of the internal components are considered in the CPPU evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation.

The effects on the loads as a result of the thermal-hydraulic changes due to CPPU were evaluated for the reactor internals. All applicable Normal, Upset, Emergency, and Faulted service condition loads were considered consistent with the existing design basis analysis.These loads include the dead weight, RIPDs, seismic loads, hydrodynamic loads (SRV and LOCA loads), annulus pressurization/jet reaction (AP/JR) loads, flow induced and acoustic loads, fuel lift loads and thermal loads. As a result of CPPU, the RIPDs increased for some components. The dead weight and seismic loads remain unchanged due to CPPU conditions. The effect of CPPU on the thermal loads for the reactor internals is insignificant. The SRV, LOCA, AP/JR and fuel lift loads in the CPPU conditions remain bounded by the original design basis loads. The acoustic load in the annulus as a result of RSLB has increased for CPPU due to the use of the Method of Characteristics, and this increase has been considered in the evaluations of the components affected by the acoustic load. The flow-induced loads are bounded by the acoustic loads.A qualitative or quantitative assessment is performed for the RPV internals, consistent with the existing design basis and the changes in the loads for CPPU. If the loads do not increase due to CPPU, the existing analysis results are assumed valid and bounding for CPPU conditions. No further analysis/qualification is required or performed. 3-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version - The combined stresses and fatigue usage factors obtained are compared to the ASME Code allowable limits for the various service conditions, as applicable, consistent with the existing design basis analysis. Conservative assessment is the initial approach, however, if required, excessive conservatism is removed from the assessment and/or the design basis analysis, as appropriate and justifiable. Table 3-8 presents the governing stresses for the various reactor internal components of SSES as affected by CPPU. All stresses are within allowable limits, and the RPV internal components are demonstrated to be structurally adequate for operation in the CPPU condition. The following reactor vessel internals are evaluated for the effects of changes in loads due to CPPU.a) Shroud: [[]] Based on the reconciliation of the stress calculations for the shroud with the increased RIPD loads and acoustic loads, it is shown that the shroud stresses are within the ASME Code allowable limits. Therefore, the shroud is structurally qualified for CPPU conditions. b) Shroud Support: [[]] Based on the reconciliation of the calculations for shroud support, it is shown that the shroud stresses are within the ASME Code allowable limits. Therefore, the shroud support is structurally qualified for the CPPU conditions. c) Core Plate: 3-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version - These parameters were Therefore, the core plate is found to remain within ASME Code allowable limits.structurally qualified for CPPU conditions. d) Top Guide:[[]] The assessment of the top guide for the CPPU conditions showed that all stresses remain within allowable limits. Therefore, the Top Guide is structurally qualified for CPPU.e) Control Rod Drive Housing (CRDH): [[]] The qualitative assessment of the CRDH for the CPPU conditions showed that the stresses are within ASME Code allowable limits. Therefore, the CRDH is structurally qualified for CPPU conditions. f) Control Rod Guide Tube (CRGT): 1Therefore, the CRGT is structurally qualified for the CPPU conditions. g) Orificed Fuel Support: 3-10 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version - Based on these calculations, the Orificed Fuel Support remains structurally qualified for the CPPU conditions. h) Fuel Channel: The fuel channels are qualified for the CPPU conditions. i) Steam Dryer: Detailed finite element analysis of the Steam Dryer is performed separately. j) Feedwater Sparger: [[]] Therefore, the existing CLTP structural qualification for the Feedwater sparger remains valid for the CPPU conditions. k) Jet Pump Assembly: [[]]loads in the CPPU conditions shows that the stresses in the jet pump components are within ASME Code allowable limits, and that adequate margin exists in the CUF for the Jet Pump Riser Brace for the CPPU conditions.

1) Core Spray Line and Sparger: [[]] Therefore, the Core Spray Line and Sparger is structurally qualified for CPPU.3-11 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -m) Access Hole Cover: [[]] Based on qualitative assessment of the Access Hole Cover in the CPPU conditions, and considering bounding values of acoustic loads, it is shown that current evaluation remains valid for the CPPU. Therefore, the access hole cover is structurally qualified for the CPPU conditions.

n) Shroud Head and Steam Separator Assembly (including Shroud Head Bolts): [[]] Therefore, the Shroud Head and Steam Separator assembly is structurally qualified for CPPU.o) In-Core Housing and Guide Tube: [[]] Therefore, the In-Core Housing and Guide Tube is structurally qualified for CPPU.p) Vessel Head Cooling Spray Nozzle: [[Therefore, the Vessel Head Cooling Spray Nozzle is structurally qualified for CPPU.q) Jet Pump Instrument Penetration Seal: [[]] Therefore, the structural integrity of the jet pump instrument penetration seal is acceptable for CPPU.r) Core Differential Pressure and Liquid Control Line: 3-12 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version - Therefore, the Core Differential Pressure and Liquid Control Line is structurally qualified for CPPU.s) Control Rod Drive Mechanism: [[]] Therefore, the Control Rod Drive Mechanism is structurally qualified for CPPU.3.3.3 Steam Dryer/Separator Performance At SSES, the performance of the steam separators and dryer has been evaluated to ensure that the quality of the steam leaving the reactor pressure vessel continues to meet existing operational criteria at CPPU conditions. CPPU results in an increase in saturated steam generated in the reactor core. For constant core flow, this in turn results in an increase in the separator inlet quality and dryer face velocity and a decrease in the water level inside the dryer skirt. These factors, in addition to the radial power distribution affect the steam separator-dryer performance. The results of the evaluation demonstrate that the steam separator-dryer performance remains acceptable (e.g., moisture content <0.1 weight %) at CPPU conditions. 3.4 FLOW INDUCED ViBRATION The FIV evaluation addresses the influence of an increase in flow during CPPU on reactor coolant pressure boundary (RCPB) piping, RCPB piping components, and RPV internals. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result 3.4.1 Structural Evaluation of Recirculation Piping 3.4.1 Structural Evaluation of.MS and FW Piping 3.4.1 Safety-Related Thermowells and Probes 3.4.2 Structural Evaluation of core flow dependent RPV Internals 3.4.2 Structural Evaluation of other RPV Internals 3-13 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -3.4.1 FIV Influence on Piping[[I Key applicable structures include the Main Steam (MS) system piping and suspension, the FW system piping and suspension, and the RRS piping and suspension. In addition, branch lines attached to the MS system piping or FW system piping are considered. RRS drive flow is not significantly increased (< 2.3%) during CPPU operation. [[The MS piping and the FW piping have increased flow rates and flow velocities in order to accommodate CPPU. As a result, the MS and FW piping experience increased vibration levels, approximately proportional to the square of the change in flow velocities. The ASME Code (NB-3622.3) and nuclear regulatory guidelines require some vibration test data be taken and evaluated for these high-energy piping systems during initial operation at CPPU conditions. Vibration data for the MS and FW piping inside containment will be acquired using remote sensors, such as accelerometers and strain gages. A piping vibration startup test program, which meets the ASME code and regulatory requirements, is performed during the CPPU power ascension test program. See separate Attachment 8 for more information. and FIV testing of the MS and FW piping system is performed during the CPPU power ascension test program.3-14 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The safety-related thermowells and probes in the MS and FW piping systems were evaluated for the increased MS and FW flow as a result of CPPU as described in separate Attachment 9.3.4.2 FIV Influence on Reactor Internal Components [[The required RPV internals vibration assessment of the other RPV internals is described in the CLTR. CPPU operation increases the steam production in the core, resulting in an increase in the core pressure drop. There is only a slight increase (2.2%) in maximum drive flow at CPPU conditions for SSES as compared to CLTP. The increase in power may increase the level of reactor internals vibration. Analyses were performed to evaluate the effects of FIV on the reactor internals at CPPU conditions. This evaluation used a reactor operational power of 3952 MWt and 108% of rated core flow. This assessment was based on vibration data obtained during startup testing of the prototype plant (Browns Ferry 1). For components requiring an evaluation but not instrumented in the prototype plant, vibration data acquired during the startup testing from similar plants or acquired outside the RPV or analysis was used. The expected vibration levels for CPPU were estimated by extrapolating the vibration data recorded in the prototype plant or similar plants and on GE BWR operating experience. These expected vibration levels were then compared with the established vibration acceptance limits. The following reactor internal components were evaluated:

• Shroud* Shroud head and moisture separator* Jet pumps* FW sparger* In-core guide tubes (generic disposition)
• Control rod guide tubes (generic disposition)
• Steam dryer* Jet pump sensing lines 3-15 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -In addition all reactor internal components in the steam flow and feedwater flow paths were evaluated.

The results of the vibration evaluation show that continuous operation at a reactor power of 3952 MWt and 108% of rated core flow does not result in any detrimental effects on the safety related reactor internal components. During CPPU, the components in the upper zone of the reactor, such as the moisture separators and dryer, are mostly affected by the increased steam flow. Components in the core region and components such as the CS line are primarily affected by the core flow. Components in the annulus region such as the jet pump are primarily affected by the recirculation pump drive flow and core flow. For the CPPU conditions at SSES, there is no change in the maximum licensed core flow in comparison to the CLTP condition, resulting in negligible changes in FIV on the components in the annular and core regions. However, the steam separator and dryer are significantly affected by CPPU conditions. The steam, dryer and steam separators are non safety-related components. Recent uprate experience indicates that FIV at CPPU conditions may lead to high cycle fatigue failure of some dryer components. Failure of a dryer component does not represent a safety concern, but can result in a large economic effect.The generally accepted source of FIV affecting steam dryers is generated outside of the reactor pressure vessel. As such itdoes not affect the steam separators, due to the isolating effect of the steam dryer between the source and the separator. Extensive analysis has been performed and continues to be performed to determine FIV effects upon the steam dryer at SSES. A multi-faceted approach to identify original, current and CPPU loads upon the steam dryer is continuing. A separate attachment to the CPPU submittal describes, in detail, the methods and results of the ongoing analysis. Multiple finite element analyses (FEA) are included as part of the steam dryer analysis. Should these FEAs result in identification of excessive loads upon the steam dryer, appropriate compensatory measures will be taken, i.e. dryer modifications or steam system modifications. The calculations for CPPU conditions indicate that vibrations of all safety related reactor internal components are within the GE acceptance criteria. The analysis is conservative for the following reasons: 3-16 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The GE criteria of 10,000 psi peak stress intensity is less than the ASME Code criteria of 13,600 psi;The modes are absolute summed; and the maximum vibration amplitude in each mode is used in the absolute sum process, whereas in reality the peak vibration amplitudes are unlikely to occur at the same time.Based on the above, it is concluded that FIV effects remain within acceptable limits at CPPU conditions.

3.5 PIPING

EVALUATION

3.5.1 Reactor

Coolant Pressure Boundary Piping The RCPB piping systems evaluation consists of a number of safety related piping subsystems that move fluid through the reactor and other safety systems. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result Structural evaluation for unaffected safety related piping Structural evaluation for affected safety related piping The evaluation of the materials and inspection programs confirm that the CPPU conditions are consistent with the Generic qualifications of the RCPB piping materials presented in Section 3.6.1, "Intergranular Stress Corrosion Cracking (IGSCC) and Erosion/Corrosion" of NEDC-32523P-A, "Licensing Topical Report Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2). See Section 10.7.The ASME Section XI In-service Inspection (ISI) program for the RRS uses BWRVIP-75 as the inspection requirements. The RRS is considered to have two stress improvements -resistant material or a stress improvement (either IHSI or MSIP) and effective HWC. Only 2 flaws exist in the RRS. Weld overlays were installed over the flaws in 2004.The flow, pressure, temperature, and mechanical loading for most of the RCPB piping systems do not increase for CPPU. 1[[3-17 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -[1 3-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -t I.i i 3-19 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version 20 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Section 3.1 demonstrates that the RCPB piping remains below the ASME pressure limit during the most severe pressurization transient. The safety related thermowells and probes in the MS and FW piping systems were evaluated (see Section 3.4.1).]]Scaling Factors Simple, conservative scaling factors are used to evaluate the effects of CPPU pressure, temperature, and flow changes. Pipe stress increase is proportional to pressure increase, temperature increase (70'F is baseline), and flow increase. Pipe support and equipment nozzle load increases are proportional to temperature increase, and (flow increase)2.These scaling factors are well agreed upon through the nuclear industry and have been demonstrated to be applicable to SSES.Main Steam and Associated Piping System Evaluation The MS piping system and associated branch piping (inside containment) were evaluated for compliance with the ASME section III, 1971 Edition with Addenda through Winter 1972 Piping Code stress criteria, including the effects of CPPU on piping stresses, piping supports including 3-21 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -the associated building structure, piping interfaces with the RPV nozzles, penetrations, flanges, and valves.[[]] The increase in MS flow results in increased forces from the turbine stop valve closure transient. The turbine stop valve closure loads bound the MSIV closure loads because the MSIV closure time is significantly, longer than the stop valve closure time.Pipe Stresses, Main Steam Line Components, and RPV Nozzles A review of the increase in flow associated with CPPU indicates that piping load changes do not result in load limits being exceeded for the MS system and attached branch piping or for RPV nozzles. The original design analyses have sufficient margin between calculated stresses and ASME section III, 1971 Edition with Addenda through Winter 1972 Piping Code stress criteria allowable limits (See Table 3-6) to support operation at CPPU conditions. Similarly, the branch pipelines (SRVDL, Reactor Core Isolation Cooling (RCIC), High Pressure Coolant Injection (HPCI), RPV Vent, and MSIV Drain) connected to the MS headers are evaluated to determine the effect of the increased MS flow on the lines. This evaluation concluded that there is no effect on the existing MS branch line qualifications due to the increased flows resulting from CPPU. As with the MS piping, the pressures and temperatures for these branch pipelines do not change as a result of CPPU. No new postulated break locations were identified. Based on existing margins available for the MS piping, components, and RPV nozzles it was concluded that CPPU does not result in reactions that exceed the current design capacity.Pipe Supports The MS piping (inside containment) was evaluated for the effects of flow increase and vibration (see section 3.4) on the piping snubbers, hangers, struts, and pipe whip restraints. A review of the increase in MS flow associated with CPPU indicates that piping load changes do not result in any load limit being exceeded.Feedwater Evaluation The FW system (inside containment) was evaluated for compliance with the ASME section III, 1971 Edition with Addenda through Winter 1972 Piping Code stress criteria, and for the effects of vibration (see section 3.4) and thermal expansion displacements on the piping snubbers, hangers, and struts. Seismic, hydrodynamic, and SRV discharge inertia and building displacement loads are not affected by CPPU, thus, there is no effect on the analyses for these load cases. Piping interfaces with RPV nozzles, penetrations, flanges, and valves were also evaluated. Pipe Stresses, Feedwater Line Components, and RPV Nozzles 3-22 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -A review of the increases in pressure, temperature and flow associated with CPPU indicates that piping load changes do not result in load limits being exceeded for the FW piping system or for RPV nozzles. The original design analyses have sufficient design margin between calculated stresses and ASME section III, 1971 Edition with Addenda through Winter 1972 Piping Code allowable limits to justify operation at CPPU conditions. The design adequacy evaluation shows that the requirements of ASME section III, 1971 Edition with Addenda through Winter 1972 Piping Code requirements remain satisfied (See Table 3-7).Therefore, CPPU does not have an adverse effect on the FW piping design. No new postulated pipe break locations were identified.

• Pipe Supports The FW system was evaluated for the effects of vibration (see section 3.4) and thermal expansion displacements on the piping snubbers, hangers, and struts. A review of the increases in temperature associated with CPPU indicates that there are no piping load changes. Therefore, the existing analyses bound the CPPU conditions.

Other RCPB Piping Evaluation This section addresses the adequacy of the other RCPB piping designs, for operation at the CPPU conditions. The nominal operating pressure and temperature of the reactor are not changed by CPPU. Aside from MS and FW, no other system connected to the RCPB experiences a significant increased flow rate at CPPU conditions. Only minor changes to fluid conditions are experienced by these systems due to higher steam flow from the reactor and the subsequent change in fluid conditions within the reactor. Seismic, hydrodynamic, and SRV discharge inertia and building displacement loads are not affected by CPPU, thus, there is no effect on the analyses for these load cases. These effects have been evaluated for the RCPB portion of the RPV head, spray, and vent lines, RV/SRV discharge piping, and Reactor Water Cleanup piping, as required.These systems were evaluated for compliance with the ASME section III, 1971 Edition with Addenda through Winter 1972 Piping Code stress criteria. Because none of these piping systems experience any significant change in operating conditions, they are all acceptable as currently designed.3.5.2 Balance-Of-Plant Piping The Balance-of-Plant Piping (BOP) systems evaluation consists of a number of piping subsystems that move fluid through systems outside the RCPB piping. The topics addressed in this evaluation are: 3-23 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The effects of the CPPU conditions have been evaluated for the following piping systems: " MS (outside containment) including Turbine Bypass Piping" Extraction Steam, FW Heater Vents and Turbine Drains* FW and Condensate" RWCU -Outside Containment

• RHR- Outside Containment
• CS- Outside Containment" HPCI- Outside Containment" RCIC -Outside Containment
• SLCS- Outside Containment" Fuel Pool Cooling* Standby Gas Treatment* Service Water" Reactor Building Closed Cooling Water* Turbine Building Closed Cooling Water* Off Gas" Stator Cooling" Containment Attached Piping including ECCS Suction Strainers For some BOP piping systems, [[1][[3-24 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version 25 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Large bore and small bore ASME Class 2, and 3 and ANSI B3 1.1 piping and supports not addressed in Section 3.5.1 were evaluated for acceptability at CPPU conditions.

The evaluation of the BOP piping and supports was performed in a manner similar to the evaluation of RCPB piping systems and supports (Section 3.5.1), using applicable ASME Section III, Subsections NC/ND or B31.1 Power Piping Code equations. The original Codes of Record (as referenced in the appropriate calculations), Code allowables, and analytical techniques were used and no new assumptions were introduced. The Design Basis Accident (DBA)-LOCA dynamic loads, including the pool swell loads vent thrust loads, condensation oscillation (CO) loads and chugging loads were originally defined and evaluated. The loads used in the qualification of structures attached to the suppression pool, such as piping system, vent penetrations, and valves are based on these DBA-LOCA hydrodynamic loads. For CPPU conditions, the DBA-LOCA suppression pool response loads were re-evaluated (See Section 4.1) and found acceptable and there are no resulting effects on the containment attached structures. Feedwater/Condensate, Extraction Steam, Feedwater Vents and Drains and Containment Attached Piping.Pipe Stresses, Piping In-Line Components, and Loads on Equipment Nozzles Operation at the CPPU conditions increases stresses on piping and piping system components due to slightly higher operating temperatures, pressures, and flow rates internal to the pipes.The maximum stress levels were reviewed based on specific increases in temperature, pressure, and flow rate (see Table 3-9). For most of the piping systems the existing temperatures and pressures bound the CPPU values. For the few segments where CPPU pressure or temperature exceeds existing, the piping systems have been evaluated and found to meet the appropriate code criteria, based on the design margins between actual stresses and code limits in the original design. All piping is below the code allowables of the plant code of record, ASME BP&V Code-Section III, Division 1, 1971 Edition, through Winter 1972 Addenda for Class 2 and 3 piping and ANSI B31.1, 1973 Edition. No new postulated pipe break locations were identified. Pipe Supports Operation at the CPPU conditions slightly increases the pipe support loadings due to increases in the temperature of the affected piping systems (see Table 3-9).The pipe supports of the systems affected by CPPU loading increases (FW, Extraction Steam, Drains, and Vent systems) were reviewed to determine if there is sufficient margin to code acceptance criteria to accommodate the increased loadings. This review shows that there is adequate design margin between the original design stresses and code limits of the supports to accommodate the load increase. The original design analyses have sufficient design margin to justify operation at the CPPU conditions. 3-26 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Main Steam and Associated Piping System Evaluation (Outside containment) The MS piping system (outside containment) was evaluated for compliance with SSES criteria based on the codes of records stated above. Included in the evaluation were the effects of CPPU on piping stresses, piping supports and the associated building structure, turbine nozzles, and valves.Because the MS piping pressures and temperatures outside containment are not affected by CPPU, there was no effect on the analyses for these parameters. The increase in MS flow results in increased forces from the turbine stop valve closure transient that is evaluated using the same scale factor method as presented in section 3.5.1; the results are provided in Tables 3-10 through 3-14.The turbine stop valve closure loads bound the MSIV valve loads because the MSIV closure time is significantly longer than the stop valve closure time.Pipe Stresses and Piping In-Line Components A review of the increase in flow associated with CPPU indicates that piping stress changes do not result in limits being exceeded for the main steam piping system outside containment; however, the 8" Steam Seal Evaporator Piping (Unit 2 only) and the Reactor Feedwater Pump Turbine Piping (Units I & 2) do exceed code allowable stresses in localized areas (See Table 3-11). No new postulated pipe break locations were identified. The areas that exceed allowable stress values are now in the modification process.Pipe Supports and Loads on Equipment Nozzles The pipe supports and RFP turbine nozzles for the MS piping system outside containment were evaluated for the increased loading and movements associated with the turbine stop valve closure transient at CPPU conditions. This review shows that in most cases there is adequate design margin between the original design stresses and code limits of the supports and nozzles to accommodate the load increase. A limited number of pipe supports and RFPT nozzles (Units I &2) CPPU results exceed their allowables (see Tables 3-12, 3-13, and 3-14). The nozzles and supports that exceed allowable values are now in the modification process.Other BOP Piping System Evaluations For those BOP sytems listed, but not specifically addressed above, the effects of CPPU are either none or not significant. There are no effects on pipe stress, support loads, or other piping in-line components for the following systems.* RWCU -Outside Containment

• RHR- Outside Containment
• CS- Outside Containment
• HPCI- Outside Containment (Steam Segment)* RCIC -Outside Containment (Steam Segment)* SLCS- Outside Containment" CRD" Emergency Service Water 3-27 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -* Service Water* Reactor Building Closed Cooling Water* Turbine Building Closed Cooling Water" Fuel Pool Cooling and Cleanup" Circulating Water* Standby Gas Treatment* Stator Cooling* Gaseous Radwaste Recombiner Closed Cooling Water System" Off Gas* Containment Attached Piping including ECCS Suction Strainers 3.6 REACTOR RECIRCULATION SYSTEM The RRS evaluation for CPPU addressed the following topics: Topic. CLTR Disposition SSES Result System evaluation NPSH Flow mismatch Single loop operation 1'[[3-28 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The CPPU power condition is accomplished by operating along extensions of current rod lines on the power/flow map with no increase in the maximum core flow. The core reload analyses are performed with the most conservative allowable core flow. The evaluation of the RRS performance at CPPU power determines that adequate core flow can be maintained.

The cavitation protection interlock remains the same in terms of absolute flow rates. This interlock is based on subcooling in the external recirculation loop and thus is a function of absolute FW flow rate and FW temperature at less than full thermal power operating conditions. Therefore, the interlock is not changed by CPPU.[[3.7 MAIN STEAM LINE FLOW RESTRICTORS The MSL flow restrictor evaluation for CPPU at SSES addressed the following topics: Topic CLTR Disposition ,SSES Result Structural integrity ]3-29 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Maximum normal operating dome pressure at CLTP and CPPU is 1050 psia The increase in steam flow rate has no significant effect on flow restrictor erosion. There is no effect on the structural integrity of the MSL flow element (restrictor) due to the increased differential pressure because the restrictors were designed and analyzed for the choke flow condition. After a postulated steam line break outside containment, the fluid flow in the broken steam line increases until it is limited by the MSL flow restrictor. 3-30 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -II1 The SSES restrictors were originally analyzed for these flow conditions and therefore the restrictors remain within the acceptable calculated differential pressure drop and choke flow limits under CPPU conditions. 3.8 MAIN STEAM ISOLATION VALVES The MSIVs evaluation for CPPU at SSES addressed the following topics: Topic CLTR Disposition SSES Result Isolation performance Valve pressure drop The MSIVs are part of the RCPB, and perform the safety function of steam line isolation during certain abnormal events and accidents. The MSIVs must be able to close within a specified time range at all design and operating conditions. They are designed to satisfy leakage limits set forth in the plant Technical Specifications. The MSIVs have been evaluated, as discussed in Section 4.7 of Reference

3. The evaluation covers both the effects of the changes to the structural capability of the MSIV to meet pressure boundary requirements, and the potential effects of CPPU-related changes to the safety functions of the MSIVs. The generic evaluation from Reference 3 is based on (1) a 20%thermal power increase, (2) an increased operating dome pressure to 1095 psia, (3) a reactor temperature increase to 556°F, and (4) steam and feedwater increases of about 24%. The evaluation from Reference 3 is confirmed applicable to SSES. An increase in flow rate assists MSIV closure, which results in a slightly faster MSIV closure time. Therefore, the CPPU described herein is bounded by conclusions of the evaluation in Section 4.7 of Reference 3, and the MSIVs are acceptable for CPPU operation.

3-31 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -3.9 REACTOR CORE ISOLATION COOLING/ISOLATION CONDENSER The Isolation Condenser is not applicable to SSES.The RCIC system evaluation for CPPU at SSES addressed the following topics: Topic CLTR Disposition SSES Result System performance and hardware [[Net positive suction head Adequate core cooling for limiting LOFW events Inventory makeup -Operational Level I avoidanceS13-32 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The RCIC system is required to maintain sufficient water inventory in the reactor to permit adequate core cooling following a reactor vessel isolation event accompanied by loss of flow from the FW system. The system design injection rate must be sufficient for compliance with the system limiting criteria to maintain the reactor water level above top of active fuel (TAF) at the CPPU conditions. The RCIC system is designed to pump water into the reactor vessel over a wide range of operating pressures. As described in Section 9.1.3, this event is addressed on a plant-specific basis. The results of the SSES plant-specific evaluation indicate adequate water level margin above TAF at the CPPU conditions. Thus, the RCIC injection rate is adequate to meet this design basis event.An operational requirement is that the RCIC system can restore the reactor water level while avoiding Automatic Depressurization System (ADS) timer initiation and MSIV closure activation functions associated with the low-low-low reactor water level setpoint (Level 1).This requirement is intended to avoid unnecessary initiations of safety systems. The results of the SSES plant-specific evaluation indicates that the RCIC system is capable of maintaining the water level outside the shroud above nominal Level I setpoint throughout a limiting LOFW event at the CPPU conditions. Thus, the RCIC injection rate is adequate to meet the requirements for inventory makeup. (See Section 9.1.3.)3-33 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -For the CPPU, there is no change to the normal reactor operating pressure and the SRV set points remain the same. There is no change to the maximum specified reactor pressure for RCIC system operation, [[]] there are no physical changes to the pump suction configuration, and no changes to the system flow rate or minimum atmospheric pressure in the suppression chamber or condensate storage tank (CST). CPPU does not affect the capability to transfer the RCIC pump suction on high suppression pool level or low CST level from its normal alignment, the CST, to the suppression pool, and does not change the existing requirements for the transfer. For ATWS (Section 9.3.1) and fire protection (Section 6.7), operation of the RCIC system at suppression pool temperatures greater than the operational limit may be accomplished by using the dedicated CST volume as the source of water. Therefore, the specified operational temperature limit for the process water does not change with the CPPU. [[]] The effect of CPPU on the operation of the RCIC system during SBO events is discussed in Section 9.3.2.The reactor system response to an LOFW transient with RCIC is discussed in Section 9.1.3.3.10 RESIDUAL HEAT REMOVAL SYSTEM The RHR system evaluation for CPPU at SSES addressed the following topics: 3-34 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Topic CLTRDisposition SSES Result LPCI mode Suppression pool and containment spray cooling modes Shutdown cooling mode Steam condensing mode Fuel pool cooling assist The RHR system is designed to restore and maintain the reactor coolant inventory following a LOCA and remove reactor decay heat following reactor shutdown for normal, transient, and accident conditions. The CPPU effect on the RHR system is a result of the higher decay heat in the core corresponding to the uprated power and the increased amount of reactor heat discharged into the containment during a LOCA. For SSES, the RHR system is designed to operate in the LPCI mode, Shutdown Cooling (SDC) mode, Suppression Pool Cooling (SPC) mode, Containment Spray Cooling (CSC) mode, and Fuel Pool Cooling (FPC) (Supplemental Spent Fuel Pool Cooling) assist.The LPCI mode, as it relates to the LOCA response, is discussed in Section 4.2.4.The SPC mode is manually initiated following isolation transients and a postulated LOCA to maintain the containment pressure and suppression pool temperature within design limits.The CSC mode reduces drywell pressure, drywell temperature, and suppression chamber pressure following an accident. The adequacy of these operating modes is demonstrated by the containment analysis (Section 4.1).The higher suppression pool temperature and containment pressure during a postulated LOCA (Section 4.1) do not affect hardware capabilities of the RHR equipment to perform the LPCI, SPC, and CSC functions. The FPC Assist (Supplemental Spent Fuel Pool Cooling) mode, using existing RHR heat removal capacity, provides supplemental fuel pool cooling capability in the event that the fuel pool heat load exceeds the heat removal capability of the Fuel Pool Cooling and Cleanup System (FPCCS). The adequacy of fuel pool cooling, including use of the Supplemental Spent Fuel Pool Cooling mode, is addressed in Section 6.3.1.The effects of CPPU on the remaining modes are discussed in the following subsections. 3-35 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -3.10.1 Shutdown Cooling Mode[[1].The SDC analysis for the CPPU determined that the time needed for cooling the reactor to 200'F with a single heat exchanger loop meets the requirement for shutdown within 24 hours.The FSAR Section 15.E.2.9 indicates that cold shutdown can be reached in a shorter time even considering the availability of only one RHR heat exchanger. For the CPPU, an alternate shutdown cooling analysis based on the FSAR criteria was performed. The results of this analysis show that the reactor can be cooled to 200TF in less than 34 hours.3.10.2 Steam Condensing Mode This mode is not installed at SSES.3.11 REACTOR WATER CLEANUP SYSTEM The RWCU system evaluation for CPPU at SSES addressed the following topics: Topic CLTR.Dispositionf SSES Result System performance Containment isolation RWCU system operation at the CPPU RTP level slightly decreases the temperature within the RWCU system (-2°F). This system is designed to remove solid and dissolved impurities from recirculated reactor coolant, thereby reducing the concentration of radioactive and corrosive species in the reactor coolant. The system is capable of performing this function at the CPPU RTP level.The CPPU review included evaluation of water chemistry, heat exchanger performance, pump performance, flow control valve capability and filter / demineralizer performance. All aspects 3-36 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -of performance were found to be within the design of RWCU at the analyzed flowrate (146,300 Ibm/hr). The RWCU analysis concludes that:* There is negligible heat load effect due to the decrease in the inlet temperature;

• A small increase in filter / demineralizer backwash frequency occurs, but this is within the capacity of the Radwaste system;* The slight changes in operating system conditions result from a decrease in inlet temperature and increase in FW system operating pressure;0 The RWCU filter / demineralizer control valve operates in a slightly more open position to compensate for the increased FW pressure;* No changes to instrumentation are required; setpoint changes are not expected due to the negligible system process parameter changes.Based on operating experience, the FW iron input to the reactor increases as a result of the increased FW flow. This input increases the calculated reactor water iron concentration from 11.55 ppb to 13.24 ppb for Unit I and 7.50 ppb to 8.6 ppb for Unit 2. However, this change is considered insignificant, and does not affect RWCU.The effects of CPPU on the RWCU system functional capability have been reviewed, and the system can perform adequately during CPPU with the original RWCU system flow. This RWCU system flow results in a slight increase in the calculated reactor water conductivity (from 0.107 [tS/cm to 0.115 [tS/cm for Unit I and 0.13 pS/cm to 0.141 pS/cm for Unit 2)because of the increase in FW flow. The present reactor water conductivity limits are unchanged for CPPU and the actual conductivity remains within these limits.The increase in FW line pressure has a slight effect on the system operating conditions.

The effect of this increase is included in Section 4.1 containment isolation assessment. 3-37 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-1 P + Q Stresses & CUFs of Limiting Components t P + Q Stress (ksi)CUF Allowable Allowable Nozzle# : 'Component CLTP: CPPU (ASNE Code CLTP a CPPU (ASME Limit= Code.. ...: " " I * " ' i i : :: " ... :iL im it) .N-3 Steam Outlet Nozzle 30.98 34.42 40.05[1' 0.841 0.841 1.000 N4 Feedwater Nozzle Safe End 2 1 7./ 73 53.1 0.816 0.816 [4 1.000 18.8 18.8 Nozzle Shell Junction 71.5 71.5 80.0 0.815 0.815[4] 1.000 N-5 Core Spray Nozzle Safe End'2 1 100.82/ 100.82/6.8 684369.9 0.615 0.615 1.000 68.483 68.483 N-10 Core AP and Liquid 77.35 / 77.35 /Control Nozzle[2) 69.5 69.5 69.9 0.823 0.823 1.000 Stabilizer Bracket 76.48 76.48 80.1 0.625 0.625 1.000 Support Skirt 56.2 56.2 65.1 Unit 1 0.913 0.913 1.000 Unit 2 0.888 0.888 1.000 Main Closure Region Stud 62.175 62.175 79.39 0.890 0.890 1.000 Head Flanget[5 70.530 70.530 80.1 0.920 0.920 1.000 1. Only the limiting components with CUF values > 0.5 are provided.2. P+Q exceeds 3 Si therefore, an elastic-plastic analysis was performed; this component is acceptable per Code requirements. The second value represents P+Q -Thermal Bending.3. The stresses are at the nozzle/shell junction. Only thermal stresses are listed, which meet the requirement of 1.5 Sm,.4. The CUF values for the safe end and nozzle/shell junction include system cycling only. Rapid cycling effects are managed using the inspection requirement defined in BWROG Letter BWROG-00068 (Reference 55),"Alternative BWR Feedwater Nozzle Inspection Requirements, GE-NE-523-A71-0594-A Revision 1, May 2000," as approved by the NRC, consistent with a plant specific fracture mechanics evaluation.

5. The head flange bounds the vessel shell flange.3-38 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-2 RIPDs for Normal Conditions (psid)* Parameter CLTP' CLTP, CPPU: Fuel Type: GE61 GEM4 GE14 Shroud Support Ring and Lower Shroud 25.77 25.18 26.99 Core Plate and Guide Tube 18.90 18.97 19.85 Upper Shroud 6.86 6.48 7.41 Shroud Head 6.95 7.09 8.06 Shroud Head to Water Level (Irreversible

7) 9.87 9.80 10.89 Shroud Head to Water Level (Elevation 2.) 0.92 0.90 0.81 Top Guide 0.94 0.89 0.89 Steam Dryer 0.34 0.34 0.42 Fuel Channel Wall 11.93 12.0'- 13.2 3.Notes: 1.. 108% core flow.2. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

3. Results are for Atrium 10 fuel.3-39 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-3 RIPDs for Upset Conditions (psid)Parameter CLTP', CLTP ' CPPU 1 FuelType:

GE6 GEM4 GEI4 Shroud Support Ring and Lower Shroud 28.17 27.58 29.39 Core Plate and Guide Tube 21.30 21.37 22.25 Upper Shroud 10.29 9.72 11.12 Shroud Head 10.43 10.63 12.10 Shroud Head to Water Level (Irreversible 2.) 14.81 14.70 <16.3 Shroud Head to Water Level (Elevation 2.) 1.38 1.35 1.22 Top Guide <1.2 <0.94 <0.94 Steam Dryer 0.51 0.54 0.64 Fuel Channel Wall 14.83 <13.2' 15.3 3.Notes: 1. 108% core flow.2. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

3. Results are for Atrium 10 fuel.3-40 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-4 RIPDs for Emergency Conditions (psid)Parameter "CLTPI CLTPI CPPUI Fuel Type: GE6 GE14 GEi4 Shroud Support Ring and Lower Shroud 30 N/C 32.5 Core Plate and Guide Tube 20.5 N/C 22.5 Upper Shroud 10.7 N/C 15 Shroud Head 11.3 N/C 14.6 Shroud Head to Water Level (Irreversible 2.) 13.6 N/C 16.3 Shroud Head to Water Level (Elevation 2.) 1.5 N/C 1.5 Top Guide 0.34 N/C 0.94 Steam Dryer 4.7 N/C 5.1 Fuel Channel Wall 13.1 N/C N/C 3 Notes: I. 108% core flow.2. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

3. Emergency condition fuel channel RPIDs are bounded by faulted values.3-41 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-5 RIPDs for Faulted Conditions (psid)Parameter." CLTP' CLTP 1 CPPU'FuelType:

GE6 GEl4 GE14 Shroud Support Ring and Lower Shroud 45 44 44 Core Plate and Guide Tube 23 22.5 23.5 Upper Shroud 27.2 27 27 Shroud Head 27.4 27 27 Shroud Head to Water Level (Irreversible 2.) 28.5 28.5 28 Shroud Head to Water Level (Elevation 2.) 2.4 2.4 2.4 Top Guide 2.1 2.2 2.2 Steam Dryer 3. 6.3 N/C 5.8 Fuel Channel Wall 15.3 15.54 16.04.Notes: 1. The pressure differential value is the maximum value of both high power and interlock condition. The high power and interlock condition are analyzed at 108% core flow and 110% core flow, respectively.

2. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud to the exit of the separators.

3. These pressure drops are for an MSLB outside primary containment.

The steam dryer pressure drop is greatest for the high flow, low power condition (interlock point). The interlock condition has not changed with the CPPU.4. Results are for Atrium 10 fuel.3-42 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-64 Main Steam ASME Class 1 Piping ASME Class 1 Pipe Stresses and Cumulative Usage Factor (CUF)UASME ASNE CLTP CPPU CPPU/A'EQUATION 3 CONDITION (PSI) (PSI) (PSI)ABLE ALLOWABLE" , : .. .... ..: (P S I) ' * : : .. .: .: : 9A DESIGN 25,785 25,931 26,550 0.977 9B NORMAL/ 25,352 25,498 UPSET 31,860 .0.800 9C EMERGENCY 21,380 23,087 39,825 0.580 9D FAULTED 52,131 52,781 53,100 0.994 10 71,411 71,660 53,100 1.350'12 32,004 32,295 53,100 0.608 13 49,491 50,460 53,100 0.950 CUMULATIVE FATIGUE USAGE FACTOR 2 CHECK 0.630 0.654 1.0 0.654 Notes 1.2.3.4.This value is acceptable because equations 12 and 13 are within allowable. These values are dimensionless. ASME B&PV Code, Section 1II, Subsection NB 3650 Results based on detailed review of Unit 2, Main Steam Line C and RCIC branch piping. Other SSES Unit I and Unit 2 Main Steam Lines were reviewed and determined that the above results are representative of those lines.3-43 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-74 Feedwater ASME Class 1 Piping ASME Class 1 Pipe Stresses and Cumulative Usage Factor (CUF)ASME ASMIE EQUATION' 3 ODTO CLTP: CPPU. : CPPU/CONDITION ALLOWABLE ALLOWABLE 2 EQUATION 3.. ... ., ,,(PSI). ..:(PSI) :..i (PSI) .__ __... .. __9A DESIGN 26,393 24,770 .26,550 0.933 9B NORMAL/ 24,275 *25,271 31,860 0.793 UPSET 9C EMERGENCY 25,232 25,232 39,825 0.634 9D FAULTED 47,952 48,388 53,100 0.911 10 104,952 105,576 53,100 1.9921 12 38,648 38,648 53,100 0.728 13 50,822 52,425 53,100 0.987 CUMULATIVE FATIGUE 0.912 0.972 1.0 0.972 USAGE FACTOR 2 CHECK Notes 1.2.3.4.This value is acceptable because equations 12 and 13 are within allowable. These values are dimensionless. ASME B&PV Code, Section III, Subsection NB 3650 Results based on detailed review of Unit 2, Feedwater Loop B piping. Other SSES Unit I and Unit 2 Feedwater Lines were reviewed and determined that the above results are representative of those lines.3-44 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-8 Summary of Governing Stresses for RPV Internals. CLTP CPPU No Component See Stress CLTP Stress Stress CPPU Stress Note, Condition Category/ C/Critial Condition* Ct.goy ..Crtil Allowable C o diio /C itca C o diio 4 .... .i. triacalLi i-..Critical Critical Limit aarameter-, rm .... ter ParameterPr

• .... .. .. : ' : Param eter .. ." : , a a e e I Shroud I Upset Pm+Pb 11.48 ksi Faulted Pm+Pb 37.37 ksi 42.9 ksi 2 Shroud Support 2 Faulted Pm+Pb 15.70 ksi Upset Pm 18.31 ksi 23.3 ksi Critical Collapse 3 Core Plate 3 Upset Moment 1.04E6 in-lb Upset Pressure 24.80 psi 27.0 psi 4 Top Guide 4 Faulted Pm+Pb < 27.56 ksi Upset Shear 8.76 ksi 10.14 ksi 5 CRD Housing -Upset Pm 12.525 ksi Upset Pm 15.71 ksi 16.6 ksi 6 Control Rod Guide Tube -Upset Collapse 0.33 Upset Collapse 0.361 0.45 7 Orificed Fuel Support 5 Upset Load 8.46 kips Upset Pm+Pb 13.87 ksi 15.59 ksi 8 Fuel Channel Qualified 9 Steam Dryer Detailed finite element analysis of the Ste3am Dryer is performed separately.

10 Feedwater Sparger Upset CUF 0.88 Upset CUF Bounded by 1.00 CLTP Bolt Bolt Bounded by 11 Jet Pump Assembly -Faulted Poltaulted Bolt Bone 22.80 kips 17.5ekpsoauled Preload CLTP 12 Core Spray Line -Upset Pm+Pb 17.7 ksi Upset Pm+Pb Unchanged 21.45 ksi 13 Core Spray Sparger -Upset Pm+Pb 6.56 ksi Upset Pm+Pb Unchanged 21.45 ksi 14 Access Hole Cover Faulted Pm+Pb <46.4 ksi Faulted Pm+Pb Unchanged 49.4 ksi 15 Shroud Head and Steam 6 Faulted Pm+Pb 66.9 ksi Faulted Pm+Qm 30.0 ksi 61.02 ksi Separators Assembly 16 In-Core Housing and Upset Pm 6.0 ksi Upset Pm, Unchanged 16.6 ksi Guide Tube 3-45 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-8__________Summary of Governing Stresses for RPV Internals. CLTP_ CPPU See -Stress -I Stress-No Component Note,' "I CLTP" Stress , Cate gory/ CPPU Stress AlloWable Condition Cae.. /Critical Condition I Criti.a /Critical* ' I _ ...Critical -Limit Critical Parameter P parameter Li mit_ I -. .." ..... .Param eter -___ ... .. ... -Param eter .._. .... .._17 Vessel Head Cooling Upset Pm+Pb+ Qm 27.8 ksi Upset Pm+Pb+ Qm Unchanged 50.0 ksi Spray Nozzle 18 Jet Pump Instrument Faulted Pm+Pb 22.2 ksi Faulted Pm+Pb Unchanged 25.4 ksi Penetration Seal.Core Differential Pressure Vortex. Vortex 19 and Liquid Control Line Normal Shedding 28.4 Hz Normal Shedding Unchanged 48 Hz Frequency, of Frequency or 20 Control Rod Drive Qualified Qualified Mechanism Limiting service conditions, stress categories and critical parameters are listed for CPPU conditions (with least margins) and reported above. Due to changes in loads and conservatism in the calculations, the critical service conditions and stress categories have changed for some components for CPPU with respect to CLTP. Explanations are provided below for the cases where such changes have occurred.I. Shroud: The CPPU acoustic loads in the horizontal and vertical directions have increased with respect to those used in the CLTP calculations. Hence the critical service condition has changed from Upset to Faulted condition.

2. Shroud Support: The cumulative effect of the increase in Normal and Upset RIPD loads on core plate, top guide and shroud head has caused the increase in primary membrane stress for the shroud support. The primary membrane stress has become the limiting stress.3. Core Plate: Elastic-plastic analysis was required to qualify the core plate. This approach is based on the use of collapse pressure (psi).4. Top Guide: Conservative loads were considered in the CPPU calculations.

As a result, the shear stress has become the limiting stress.5. Orificed Fuel Support: Based on the calculations using the CPPU RIPDs, (Pm+Pb) is the critical stress category in the CPPU.6. Shroud Head and Steam Separators Assembly: In order to reduce conservatism, credit was taken for the seismic pins to resist lateral loads, which results in the elimination of bending stress in the shroud head bolts. Thus, the primary plus secondary membrane stress has become the limiting stress.3-46 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-9 BOP Piping FW, Condensate, Extraction Steam, FW Heater Drains & Vents Maximum pipe stress increase: Temperature expansion 3%Pressure *4%-20%Fluid Transients 0 Maximum pipe support loading increase 3%(due to thermal expansion loading)Table 3-10 Main Steam including Turbine By-Pass and Reactor Feed Pump Turbine Maximum pipe stress increase: Temperature expansion No change Pressure No change Fluid Transients 15%-20%Maximum pipe support loading increase CPPU (due to thermal expansion No change loading)CPPU (due to fluid transient loading) 40%- 45%Table 3-11 Units 1 & 2 -Main Steam & HP Steam to RFPT Maximum ANSI B31.1 Pipe Stresses for ANSI Equation 9B, Normal/Upset Condition CLTP CPPU AN ICPPU/Unit 'Line 7"ALOABLE.(PSI) (PSI) (PS)N ALLOWABLE' I HP Steam to RFPT 17,752 21,302 18,000 1.18 2 Steam Seal Evaporator 17,180 19,791 18,000 1.10 2 HP Steam to RFPT 16,687 19,380 18,000 1.08 Note: 1. The pipe sections that exceed the allowable stress values are in the modification process.3-47 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-12 Unit 1 -Main Steam and Steam Seal Pipe Supports Supports Whose CPPU Values Exceed The Minimum Support Allowable Oper. CPPU/PIPE Cond. CLTP. CPPU Allowable Allowable i, Pipe Support Type (1) SIZE: (2): (3): (3) (3) .(4)DBB-104-H20 GUIDE 24 N/U 68289 81101 72113 1.12 DBB-104-H07 SNB 24 E/F 47950 63851 49101 1.30 MSL-100-HOI SNB 28 E/F 25873 30086 28874 1.04 MSL-100-H03 SNB 28 N/U 14952 20933 19565 1.07 MSL-100-H02 SNB 28 N/U 33985 47579 41530 1.15 DBB-104-H18 SNB 24 N/U 118414 165780 119953 1.38 DBB-103-H07 SNB 24 N/U 40861 57205 41229 1.39 MSL-100-H1O SNB 28 N/U 27849 36343 28796 1.26 DBB-102-H07 SNB 24 N/U 38842 54379 38842 1.40 MSL-100-H11 SNB 28 N/U 43904 61466 50007 1.23 DBB-102-H17 SNB 24 N/U 118641 159767 119946 1.33 DBB-101--H08 SNB 24 N/U 36008 50411 36008 1.40 EBD-113-H04 SNB 8 F 4049 4373 4094 1.07 Notes: 1. SNB = Snubber 2. N/U = Normal/Upset, E = Emergency, F = Faulted. Only the worst case is shown in the table.3. Values are Pounds (lbs).4. Values are dimensionless. These supports are in the modification process because they exceed their allowable. 3-48 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-13 Unit 2 -Main Steam and Steam Seal Pipe Supports Whose CPPU Values Exceed The Minimum Support Allowable Oper CPPU/,: :,:PpeSup~t: Tpei) PIPE:: CLTP. 'CPPU--ieType (1)pe. Cond

• Allowable

'Allowable SIZE : (3)(2)3) ..DBB-201-H24 SNB 24 N/U. 117,520 164,482 120,000 1.37 DBB-202-H04 SNB 24 E 26,082 29,184 26,082 1.12 DBB-203-H21 SNB 24 N/U 117,855 159,813 119,034 1.34 DBB-205-H12 ANC 24 N/U 141,664 161,049 156,312 1.03 EBD-213-H05 SNB 8 E 3,427 3,881 3,530 1.10 EBD-213-H1O ANC 8 F 3,786 5,447 4,271 1.28 EBD-213-H10 (3) ANC 8 F 18,144 23,558 20,466 1.15 MSL-200-HO1 SNB 28 N/U 19,519 23,622 20,885 1.13 MSL-200-H03 SNB 28 N/U 14,145 19,824 14,286 1.39 MSL-200-HO9A SNB 28 E 52,804 59,705 55,972 1.07 MSL-200-HO9B SNB 28 E 16,297 18,314 17,438 1.05 MSL-200-H1IA SNB 28 N/U 41,185 57,617 47,363 1.22 MSL-200-H1IB SNB 28 N/U 50,990 71,281 50,000 .1.43 Notes: 1. ANC = Anchor[Only the highest of the 3 forces and the highest of the 3 moments are shown in the table] SNB = Snubber, 2. N/U = Normal/Upset, E= Emergency, F=Faulted. Only the worst case is shown in the table.3. Values are Pounds (Ibs) except for the moment value at the anchor that is ft-lbs.4. Values are dimensionless. These supports are in the modification process because they exceed their allowable. 3-49 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 3-14 Units I & 2 -High Pressure Steam To Reactor Feed Pump Turbines Nozzles Where CPPU Pipe Loads Exceed The Allowable FEEDPUMP UI TURBINE HP PIPE Oper. CPPU Allowable Aob UN T UBN Pi Location .Con&.i CLTP: !: !.. Allowable UT STEAM1% SIZE NOZZLES (1) (RFPT I S- 105A Force (2) N/U 2,306 2,862 6,000 0.477 Moment (3) 4 N/U 16,235 20,820 36,000 0.578 Interaction (4) N/U N/A N/A 1.0 1.055 RFPT IS-105B Force (2) N/U 1,976 2,406 6,000 0.401 Moment (3) 4 N/U 19,162 23,032 36,000 0.640 Interaction (4) N/U N/A N/A 1.0 1.041 RFPT 1S-105C Force (2) 4 N/U 2,090 2,224 6,000 0.371 Moment (3) N/U 22,750 24,381 36,000 0.677 Interaction (4) N/U N/A N/A 1.0 1.048 2 RFPT2S-105A Force (2) N/U 1,362 1,851 3,000 0.617 Moment (3) 4 N/U 10,359 14,359 18,000 0.798 Interaction (4) N/U N/A N/A 1.0 1. 15 2 RFPT2S-105B Force (2) N/U 1,120 1,483 3,000 0A94 Moment (3) 4 N/U 11,148 14,547 18,000 0.808 Interaction (4) N/J N/A N/A 1.0 1.303 Notes: 1. N/U = Normal/Upset is the controlling case; the Emergency and Faulted cases will have the same pipe loads with higher allowables thus smaller interaction values.2. Values are Pounds (lbs)3. Values are fl-lbs.4. Values are dimensionless. Interaction value is the sum of the force + moment values.These nozzles and the attached piping are in the modification process because the applied forces and the moments exceed the nozzle allowable. 3-50 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Figure 3-1 MSIV Closure With Flux Scram at 102%P/108%F -Power, Heat Flux, and Flows 300.0 Relative Core Power Relative Heat Flux A. Relrvo...Co t.(ý ,% .Fo ....Realaqtive_ Core Flow Relative Steam FlowRelative Feed Flow ,~o' ,/-F- _ : o100.0 -----, '-------w. yV-¶00.0.0 ' 1:0 o .30 4.0 5!0 8.0 7..o 8.Time (seconds)SQ EPU ASME-MSIV at P102F-&O.._X19403.7 12/19/06 15.41:42 N V-21684 Figure 3-2 MSIV Closure With Flux Scram at 102%P/108%F -Downcomer Water Level 400 r 30.0* 2&.0.-20.0.15.0...... ... ..10.04 * .....0 1.0 2.0 3.0 4:0 6.0 Time (seconds)0.0 7.0 6:0 SO EPU ASME.MSIV at P1O2..F1O8-...X19403.7 12/19/05 1W.41,2 ICMO24. Me 0-0164 3-51 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Figure 3-3 MSIV Closure With Flux Scram at 102%P/108%F -Pressures.ýSO EPU ASME-MSIV at P102-F108__Xt9403.7 12/19/06 1&4142 N4V-2245. JOB 0.21"4 Figure 3-4 MSIV Closure With Flux Scram at 102%P/108%F -MSRV Position 1.2-.2 a 0 1.0-76 .4-(a 4)Low Setpoint MeAdium Setpoint I: 'I.)IIi ..0 1.0 .e2.0 3 s0 4.0 Time (seconds)6.0 7.0 6o SQ EPUI ASME..MSIV at P102-.F1O&...XI9403.7 12/19/03 M&4142 MV)-26245. 0M -21684 3-52 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Figure 3-5 MSIV Closure With Flux Scram at 102%P/108%F -MSRV Fl\ow 2000.0 3 1500.0-a a V 1000.0-V)500.0-.0 Low Setpoint 1.9 A .1.0 2.0 3.0 4.0 5,0 6.0 Time (seconds)7.0 8.0 SO EPU ASMEMSIV ot P102_F1O8__X19403.7 12/19/05 1M41:42 NM-25245, Figure 3-6 MSIV Closure With Flux Scram at 102%P/108%F -Reactivities (Ak/k)115 .~.0 .- --------.05.-.0*-.15,\ -Scram Reactivity.Mod Rectivity TotalReectivity -A°5 .D 1m0 2s0 3n0 4.0 Time (seconds)5.0 s.0 7:0 8.0.SQ EPU ASME-MSIV at P102_F108__XI94O3.7 12/10/05 1541-412 105-29245. JO D-21654 3-53 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Figure 3-7 MSIV Closure With Flux Scram at 102%P/108%F -Core Inlet Enthalpy 328.0 -527.5-526.0 0 0 ofl.u +.0 1.0 2.0 M.0 4.0 Time (seconds)5.0 0.0 7.0 8:0 SO EPU ASME..MSIV at P102-.F1O8--.X19403.7 12/19/05 16.41-.42 NQS-20245. JOB D-21584 3-54 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -4. ENGINEERED SAFETY FEATURES This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapter 6, which applies to CPPU. RG 1.70, Chapter 6 states, "engineered safety features are provided to mitigate the consequence of postulated accidents," and "are those (features) that are commonly used to limit the consequences of postulated accidents." NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 6.1.1; subsection I states, "Engineered safety features (ESF) are provided in nuclear plants to mitigate the consequences of design basis or loss-of-coolant accidents." The SSES plant.features evaluated within this .section are designed to (directly) mitigate the consequences of postulated accidents, and thus, are classified in the plant FSAR as engineered safety features, consistent with RG 1.70 and NUREG-0800.

4.1 CONTAINMENT

SYSTEM PERFORMANCE This section addresses the effect of the CPPU on various aspects of the SSES containment system performance. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result 4.1.1 Pool temperature response 4.1.1 Wetwell pressure 4.1.1 Drywell temperature

4.1.1 Drywell

pressure 4.1.2 Containment dynamic loads 4.1.3 Containment isolation 4.1.4 Motor-operated valves 4.1.5 Hardened wetwell vent system 4.1.6 Equipment operability ]]The FSAR provides the containment responses to various postulated accidents that validate the design basis for the containment. CPPU operation changes some of the conditions for the containment analyses. For example, the short-term DBA-LOCA containment response during the blowdown is governed by the blowdown flow rate. This blowdown flow rate is dependent on the reactor initial thermal-hydraulic conditions, such as vessel dome pressure and the mass and energy of the vessel fluid inventory, which change slightly with CPPU. Also, the long-term heatup of the suppression pool following a LOCA or a transient is governed by the ability of the RHR to remove decay heat. Because the decay heat depends on the initial reactor power level, 4-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -the long-term containment response is affected by CPPU. The containment response was reanalyzed to demonstrate the plant's capability to operate with a rated power increase to 3952 MWt.The analyses of containment pressure and temperature responses, as described in Section 4.1.1, were performed in accordance with RG 1.49 and ELTRI using GE codes and models. The M3CPT code was used to model the short-term containment pressure and temperature response.The modeling used in the M3CPT analyses is described in References 10 and 11. References 10 and 11 describe the basic containment analytical models used in GE codes. Reference 16 describes the more detailed RPV model (LAMB) used for determining the vessel break flow in the containment analyses for CPPU.The LAMB code models the recirculation loop as a separate pressure node. It also allows for inclusion of flashing in the pipe and vessel during the blowdown and flow choking at the jet pump nozzles when the conditions warrant. The use of the LAMB blowdown flow in M3CPT was identified in ELTRI by reference to the LAMB code qualification in Reference 16.The PICSM code was used to model the wetwell pressurization during the pool swell phase of the DBA-LOCA. The PICSM model for pool swell is described in Reference

13. The NRC has approved the use of the PICSM code to model the pool swell phenomenon in Reference

14. Use of the PICSM is in accordance with the acceptance criteria of Reference 14.The SHEX code was used to model the long-term containment pressure and temperature response.

The key models in SHEX are based on models described in Reference

11. The GE containment analysis methodologies have been applied to all BWR power uprate projects performed by GE and accepted by the NRC. The SHEX code was used in the current licensing basis analyses for long-term containment pressure and temperature response and therefore no confirmatory calculation is necessary.

4.1.1 Containment

Pressure and Temperature Response Short-term and long-term containment analysis results are reported in the FSAR. The short-term analysis is directed primarily at determining the containment pressure response during the initial blowdown of the reactor vessel inventory to the containment following a large break inside the drywell. The long-term analysis is directed primarily at the pool temperature response, considering the decay heat addition to the suppression pool. Peak values of the containment pressure and temperature responses to the DBA-LOCA are given in Table 4-1. The effect of CPPU on the events, which yield the limiting containment pressure, and temperature response, is provided below.4.1.1.1 Long-Term Suppression Pool Temperature Response (a) Bulk Pool Temperature The long-term bulk pool temperature response for CPPU is evaluated for the limiting DBA-LOCA in Section 6.2 of the FSAR and the limiting Alternate Shutdown activity in Section 15.2 of the FSAR.4-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The analysis of the DBA-LOCA was performed at 102% of CPPU RTP. The calculated peak values for LOCA bulk pool temperature for the current FSAR case and the CPPU RTP case are compared in Table 4-1. The CPPU analyses were performed using a decay heat table based on ANS/ANSI 5.1-1979 with 2-sigma adders, which is more realistic than what was used for the FSAR Section 6.2 analysis, and includes passive heat sinks in containment. The resulting calculated peak bulk suppression pool temperature is 211.2 0 F.The highest bulk pool temperature response from a non-LOCA event results from an Alternate Shutdown Cooling event. This event was also analyzed at 102% of CPPU RTP. The limiting alternate shutdown activity assumes reactor isolation with availability of one RHR heat exchanger. The resulting peak bulk pool temperature at 102% of CPPU RTP is 211.6 0 F.(b) Local Pool Temperature with SRV Discharge The local pool temperature limit for SRV discharge is specified in NUREG-0783, because of concerns resulting from unstable condensation observed at high pool temperatures in plants without quenchers. The peak local suppression pool temperature at SSES has been evaluated for CPPU and remains below the NUREG-0783 limit. Therefore, the peak local suppression pool temperature at SSES is acceptable for CPPU conditions. 4.1.1.2 Gas Temperature Response The drywell design temperature (340'F) has been determined based on a bounding analysis of the superheated gas temperature, which can be reached with a blowdown of steam to the drywell during a LOCA. Because the vessel dome pressure assumed for the FSAR containment analysis is unchanged by CPPU the initial break flow rate for this event is unchanged. Therefore, there is no effect on the short-term peak drywell temperature. The wetwell gas space peak temperature response was calculated assuming thermal equilibrium between the pool and wetwell gas space. Table 4-1 shows the bulk pool temperature would be 211.2'F for CPPU. Therefore, the wetwell gas space temperature would also be 211.2'F for CPPU. This temperature remains below the design temperature of 340'F.4.1.1.3 Short-Term Containment Pressure Response Short-term containment response analyses were performed for the limiting DBA LOCA, which assumes a double-ended guillotine break of a recirculation suction line to demonstrate that CPPU does not result in exceeding the containment design limits. The short-term analysis covers the blowdown period during which the maximum drywell pressure, wetwell pressure and differential pressure between the drywell and wetwell occur. These analyses were performed at 102% of CPPU RTP level. The results of these short-term analyses are summarized in Table 4-1. Table 4-1 also includes comparisons of the pressure values calculated for CPPU to the design pressures and to pressure values from previous calculations based on the current power. The maximum calculated containment pressure for CPPU remains within the design value, and thus, is acceptable. The peak calculated drywell-to-wetwell pressure also remains within its design value.4-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -4.1.2 Containment Dynamic Loads 4.1.2.1 Loss-of-Coolant Accident Loads The LOCA containment dynamic loads analysis for CPPU is primarily based on the short-term MSLB and RSLB LOCA analyses. These analyses were performed as described in Section 4.1.1.3 except that break flows for the RSLB were also calculated using a more detailed RPV model (Reference 16). The NRC in Reference 11 approved use of this model for the CPPU containment evaluations. A description of the LOCA loads and the methodologies for specifying the loads are found in the SSES DAR (Reference 23). The LOCA hydrodynamic load specifications used in the original plant design, augmented by additional calculations for the previous SSES uprates were reviewed to determine if operation at CPPU conditions will have an effect on the LOCA hydrodynamic loads.4.1.2.1.1 Submerged Boundary Loads During Vent Clearing The vent clearing phenomenon following a LOCA results from the clearing of water from the main vent downcomers due to drywell pressurization. As a result of this phenomenon, pressure loads are produced on the containment basemat and the submerged wetwell walls. The NRC acceptance criteria specify a 24 psi overpressure on the containment based the correlation presented in Reference 14, page 2-3. An evaluation of the specified load concludes that the 24 psi overpressure is unchanged for CPPU.4.1.2.1.2 Pool Swell Loads Pool swell loads, which occur during vent clearing from a LOCA, are the result of the rise of the suppression pool surface, which increases the pressure in the wetwell air space. Pool swell loads act on the wetweli boundary and impacts both the drag and fall back loads on the wetwell components located within the pool swell zone. Pool swell loads are a function of the initial drywell pressurization rate during a LOCA. The results of the CPPU pool swell analysis are bounded by the current analysis.4.1.2.1.3 LOCA Steam Condensation Pool Boundary Loads After the initial pool-swell transient resulting from a postulated LOCA in a BWR, steam with decreasing amounts of air is vented from the drywell into the wetwell of the pressure suppression system. The purpose of this venting is to condense the steam in the wetwell pool and limit the pressure buildup in the containment. During such steam venting, condensation-driven oscillations have been observed in related experiments. Two types of condensation-driven oscillations occur. The first type, called condensation oscillation, occurs during the earlier portion of the blowdown of a large break LOCA or MSLB, when the steam mass flux and air content in the steam are high. CO is characterized by approximately sinusoidal pressure oscillations in the drywell and wetwell system. These CO are 4-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -followed by the second type of condensation-driven oscillations called chugging. The pressure oscillations during chugging are associated with the rapid collapse of the steam bubble at the vent exit and typically exhibit a pressure spike, followed by a damped ringout which has predominant frequency components at the vent and pool natural frequency. The load definition for SSES is based on the full scale LOCA steam condensation tests conducted by Kraftwerk Union at their GKM II-M test facility. Reference 23, Section 9.0 provides a description of the test matrix and results.4.1.2.1.3.1 Condensation oscillation (CO) Submerged Boundary Loads CO loads result from oscillation of the steam-water interface that forms at the vent exit during a LOCA during the period of high vent water vapor mass flow rate. This occurs after pool swell.The CO loads include loads on submerged boundaries and submerged structures. Generally, the CO load increases with higher suppression pool temperature and/or higher vent mass flow rate.A comparison of the break flow (and hence vent flow) for the CPPU conditions with the flow used to document the adequacy of the CO loads for stretch uprate, indicates very similar flow, and both are enveloped by the vent flow calculated for the GKM II-M test. Hence, the CO load developed for the GKM-II-M test data remains bounding for SSES at CPPU conditions. 4.1.2.1.3.2 Chugging Submerged Boundary Loads Chugging occurs when the steam (water vapor) mass flux through the vents during a LOCA is not high enough to maintain a steady steam/water interface at the vent exit. Chugging loads result from a collapse of steam (water vapor) bubbles that form at the vent exit, and they include loads on the suppression pool boundary and submerged structures. Because the vessel pressure is unchanged for CPPU at full power, the MSL break flow and consequently the steam vent flow will be the same for the operation in the CPPU domain.Because the chug amplitude is proportional to the vent steam mass flux, which is not changed for CPPU, the design pressure traces selected from the smaller steam break tests are conservative for CPPU domain. Therefore, the chugging load developed for the GKM-1I-M test data remains bounding for SSES at CPPU conditions. 4.1.2.1.4 Downcomer Lateral Tip Loads Chugging produces a lateral load at the downcomer exit. The chugging loads are dominated by the MSLB and smaller steam breaks. As concluded in Section 4.1.2.1.3.2, the current chugging loads bound the CPPU operating conditions; therefore, the current downcomer lateral tip loads remain bounding for SSES at CPPU conditions. 4-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -4.1.2.1.5 LOCA Submerged Structure Loads 4.1.2.1.5.1 Downcomer Jet Loads The clearing of the downcomers following the design basis LOCA produces a water jet load on submerged structures located beneath the downcomers. Reference 24 describes this phenomenon as a water discharge in the form of a narrow jet whose transverse dimension remains approximately the size of the exit diameter. A review of the original downcomer jet load evaluation found that no submerged structure were located underneath the downcomers. 4.1.2.1.5.2 LOCA Air Bubble Loads The maximum air-bubble pressure occurs at the maximum pool swell height as stated in Section 2.1.2.5 of Reference

14. The maximum air bubble pressures at CPPU conditions are less than the current calculated pressures.

Therefore, CPPU does not affect the LOCA air bubble loads.4.1.2.1.5.3 CO and Chugging Submerged Structure Drag Load The original CO and chugging submerged structure drag load used the same acoustic model of the SSES suppression pool and design pressure traces for sourcing as the CO and chugging submerged boundary load methodology. However, the pressures are calculated at the submerged structure surfaces, instead of containment boundary. As described in sections 4.1.2.1.3.1 and 4.1.2.1.3.1 of this report, the current CO and chugging loads are conservative for CPPU. Thus, the acoustic sources derived from these loads are also conservative for CPPU.4.1.2.2 Safety Relief Valve Loads The SRV dynamic loads which need to be evaluated for CPPU. are: loads on the quenchers, quencher supports, and SRV discharge lines, loads on the submerged boundary of the suppression pool, and loads on the submerged structures within the suppression pool.The parameters which affect the SRV loads, the vessel pressure, the SRV opening and closing setpoints, the submergence, the line air volume, and the ADS setpoints, are not changed for CPPU.The SRV pressure traces used in Reference 23 to define the SRV loads described below were observed during subsequent SRV actuations and include the trace that gives rise to the highest peak overpressure (POP) for all tests. A pressure multiplier of 1.5 was applied to all traces leading to a maximum POP value of 1.2 bar. In Reference 14, the NRC found acceptable the use of this multiplier for subsequent actuation peak pressure loads. Therefore, CPPU does not affect the SRV loads or loads definition. 4.1.2.3 Subcompartment Pressurization The mass and energy releases that affect annulus pressurization loads on the biological shield wall caused by a postulated recirculation suction line break (RSLB), FW line break (FWLB) or main steam line break (MSLB) were evaluated at CPPU conditions. The methods used to 4-6 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -perform the evaluations are consistent with those used in the existing SSES analysis of record, including the evaluations performed for ARTS/MELLLA. For the RSLB and MSLB cases, the current licensing basis bound the mass and energy release rates. The FWLB mass and energy release rates exceed the CLTP licensing basis values. The FWLB mass and energy release rates are bounded, however, by the original design basis values.4.1.3 Containment Isolation The system designs for containment isolation are not affected by CPPU. The capabilities of isolation actuation devices to perform during normal operations and under post-accident conditions have been determined to be acceptable. Therefore, the SSES containment isolation capabilities are not adversely affected by the CPPU.4.1.4 Generic Letter 89-10 Program Motor-Operated Valves The motor-operated valve (MOV) process parameters (temperature, pressure, and flow) were reviewed, and increases in design differential pressure due to operation at CPPU conditions have been identified for some MOVs. Based on a margin review, it is concluded that Generic Letter (GL) 89-10 MOVs are capable of performing their design basis safety functions at CPPU conditions. Operation at the CPPU conditions increases post-accident room temperatures where some MOVs are located, potentially reducing the actuator output torque. The SSES GL 89-10 MOVs have been reviewed and they remain capable of performing their design basis functions. MOVs used as containment or HELB isolation valves have been reviewed for the effects of operations at CPPU conditions, including thermal binding and pressure locking (GL 95-07).The operability of motor operated valves is documented as part of the SSES GL 89-10 program.Air-Operated Valves The air-operated valve (AOV) parameters (temperature, pressure, and flow) were reviewed, and increases in design differential pressure due to operation at CPPU conditions have been identified for some AOVs. Based on a review of the scenarios, all AOVs with active, safety related or safety significant functions are capable of performing their design basis safety functions at CPPU conditions. Operation at the increased CPPU conditions is within the pressure and temperature capability of the AOVs. AOVs used as containment isolation valves have been reviewed for the effects of operations at CPPU conditions, including thermal binding and pressure locking (GL 95-07).Therefore, the SSES AOVs remain capable of performing their design basis functions. 4-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -4.1.5 Generic Letter 89-16 Not Applicable to SSES.4.1.6 Generic Letter 96-06 The NRC staff acceptance of PPL's responses to Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions" is provided in Reference

17. These responses were reviewed for CPPU post accident conditions.

The issues identified in the GL 96-06 review were addressed through procedural changes and analysis of the potential for pipe penetration leakage, which are unaffected, by CPPU.Therefore, the accepted SSES response to Generic Letter 96-06 remains valid for CPPU.4.2 EMERGENCY CORE COOLING SYSTEMS Each ECCS is discussed in the following subsections. The effect on the functional capability of each system, due to CPPU is addressed. The assumption of constant pressure minimizes the effect of CPPU for ECCS evaluation. The topics addressed in this evaluation are: Topic CLTR Disposition i ,SSES.Result 4.2.1 High Pressure Coolant Injection 4.2.2 High Pressure Core Spray 4.2.3 Core Spray 4.2.4 Low Pressure Coolant Injection 4.2.5 Automatic Depressurization 4.2.6 ECCS Net Positive Suction Head 4.2.1 High Pressure Coolant Injection Er 4-8 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The HPCI system is designed to pump water into the reactor vessel over a wide range of operating pressures. The primary purpose of the HPCI is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the reactor vessel. In this event, the HPCI system maintains reactor water level and helps depressurize the reactor vessel. The adequacy of the HPCI system is demonstrated in Section 4.3.]] For CPPU, there is no change to the maximum nominal reactor operating pressure and the SRV setpoints remain the same.Er The NPSH available for the HPCI pump does not change because there are no physical changes to the pump suction configuration, and no changes to the system flow rate or minimum atmospheric pressure in the suppression chamber or CST. The specified operational temperature limit for the process water does not change with the CPPU. The NPSH required by the HPCI pump does not change because there is no change to the maximum rated pump speed or the required pump flow rate.E[1]4.2.2 High Pressure Core Spray The High Pressure Core Spray system is not applicable to SSES.4.2.3 Core Spray or Low Pressure Core Spray The Low Pressure Core Spray (LPCS) system is not applicable to SSES.[[4-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -0]The Core Spray (CS) system is automatically initiated in the event of a LOCA. When operating in conjunction with other ECCS, the CS system is required to provide adequate core cooling for all LOCA events. There is no change in the reactor pressures at which the CS is required.The CS system sprays water into'the reactor vessel after it is depressurized. The primary purpose of the CS system is to provide reactor vessel coolant inventory makeup for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. It also provides long-term core cooling in the event of a LOCA. The CS system meets all applicable safety criteria for the CPPU. The adequacy of the CS system performance is demonstrated by the margins discussed in Section 4.3.4.2.4 Low Pressure Coolant Injection[[4-10 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The LPCI mode of the RHR system is automatically initiated in the event of a LOCA. The primary purpose of the LPCI mode is to help maintain reactor vessel coolant inventory for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. The LPCI operating requirements are not affected by CPPU. The adequacy of this system is demonstrated by the margins discussed in Section 4.3.4.2.5 Automatic Depressurization System The ADS uses SRVs to reduce the reactor pressure following a small break LOCA when it is assumed that the high-pressure systems have failed. This allows the CS and LPCI to inject coolant into the reactor vessel. The adequacy of this system is demonstrated by the margins discussed in Section 4.3. The ADS initiation logic and valve control is not affected by CPPU conditions. The CPPU does not change the conditions at which the ADS must function.I]]4-11 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -4.2.6 ECCS Net Positive Suction Head CPPU increases the reactor decay heat, which increases the heat input to the suppression pool.This increased heat input increases the peak suppression pool water temperature, which may affect RHR, CS and HPCI pump operation. As discussed in Section 4.1.1.1.a, the calculated peak suppression pool temperature for the most limiting case, ASDC, is 211.61F, which exceeds the current peak pool temperature of 208'F. Peak suppression pool temperatures for an ATWS, LOCA, Appendix R, and SBO event are bounded by the most limiting case.The NPSH requirements for the RHR and CS pumps were analyzed at the maximum calculated pump flow, which exceeds the design basis pump flows, and a peak suppression pool temperature of 220'F. The analysis assumed that the containment pressure will equal the vapor pressure of the suppression pool water to ensure that credit is not taken for containment pressurization during the transient, which is consistent with (Reference 15), (NUREG-0800, SRP 6.2.2, Containment Heat Removal Systems). The inputs in the ECCS NPSH calculations for friction loss, static head, strainer debris loading, and NPSH required have not been changed and are not affected by CPPU. Using these conditions, the available NPSH exceeds the required NPSH for the RHR and CS pumps.The NPSH requirements for the HPCI pump are based on design basis peak suppression pool temperature of 140'F, 0 psig containment pressure, as required by RG 1.1, and design basis pump flow. Under CPPU conditions, the peak suppression pool temperature for operation of the HPCI pump does not exceed 140 'F (including ATWS, Appendix R and SBO events). The design requirements of the Condensate Storage Tank (CST) are not affected by CPPU;therefore, the NPSH requirements for the HPCI pump from the CST are not affected by CPPU.SSES strainer design requirements are pressure drop, flow, and debris loading based on worst case short term and long term ECCS operation following a postulated LOCA. All debris sources in the drywell and suppression chamber are available for transport to the suction strainers, which is more conservative than the guidance provided in Reference 18 (NEDO-32686 Rev. 0, Utility Resolution Guidance for ECCS Suction Strainer Blockage), and Reference 19 (NUREG/CR-6224, Parametric Study of the Potential for BWR ECCS Strainer Blockage due to LOCA Generated Debris). Therefore, the suction strainer debris loading analysis remains bounding for CPPU.4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE The Susquehanna ECCS performance evaluation for ATRIUM-10 fuel is summarized in this section. A Susquehanna break spectrum analysis for ATRIUM-10 fuel was performed using FANP methodology documented in References 20 and 21. Based on the results of the break spectrum analysis, the limiting break characteristics for ATRIUM- 10 fuel at Susquehanna are as follows: 0 Break size/geometry -Double-ended guillotine / 1.0 discharge coefficient (1.0 DEG).0 Break location -recirculation suction line.0 Single failure -Low-pressure coolant injection valve (SF-LPCI).

• Axial power shape -top-peaked (TOP).4-12 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The ATRIUM- 10 break spectrum analyses assumed an initial core power of 102% of the CPPU rated thermal power (RTP) value of 3952 MWt, providing a licensing basis power of 4031*MWt. The 2% increase reflects the maximum uncertainty in monitoring reactor power, as per NRC requirements (Regulatory Guide 1.49, as discussed in Section 1.2.1). LOCA Break spectrum analyses support operation over a range of core flow rates from 99 to 108 Mlb/hr within CPPU/MELLLA domain.The licensing basis peak clad temperature (PCT) for ATRIUM-I10 fuel at CPPU RTP is 18447F, which is well within the Appendix K 2200'F limit. This PCT result is not directly comparable to the PCT for the pre-CPPU condition due to changes in the analysis assumptions.

An overly conservative assumption used :in the pre-CPPU analysis was removed for the CPPU analysis.The pre-CPPU PCT for the limiting CPPU break conditions is 1945'F. In order to provide a comparison between pre-CPPU and CPPU PCTs, the limiting CPPU case was rerun using the same pre-CPPU analysis assumptions. The limiting CPPU case yields a PCT of 1914'F when the pre-CPPU assumptions are used. The CPPU maximum local cladding oxidation is 0.80%(< 17%), and the total hydrogen generation is < 1% of the total possible. Compliance with these three criteria (PCT, cladding oxidation and hydrogen generation) ensures that a coolable geometry is maintained. The local fuel conditions for CPPU are not significantly changed because the hot bundle operation is still constrained by the same MAPLHGR limit; therefore, long term cooling is also supported. Table 4-2 presents the limiting LOCA analysis results for ATRIUM-10 fuel. The MAPLHGR limit analyses show that the limiting exposure for Susquehanna ATRIUM-10 fuel is beginning of life (BOL).Calculations were also performed to support single-loop operation (SLO) with ATRIUM-I10 fuel with a multiplier applied to reduce the MAPLHGR limit. The SLO analyses show that the PCT is bounded by the Two Loop Operation (TLO) PCT and is well below the Appendix K limit of 2200 0 F.4.4 MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM The SSES topic addressed in this evaluation is: Topic. CPPU Effect SSES Result Iodine Intake The heat sources for the Main Control Room Atmosphere Control System are from equipment, ambient outside air temperature and the personnel in the Control Room. Heat loads from these sources do not change for CPPU. The SSES Control Room Habitability Envelope (CRHE)includes all areas of the Control Structure between plan floor elevations 697'-0" and 783'-0," Stairwells 120 and 221, vestibule C-131 on elevation 676'-0" and the passenger elevator. The Control Room Emergency Outside Air Supply System (CREOASS) processes outside air needed to provide ventilation and pressurization of the CRHE during accident conditions. The CREOASS unit is automatically started and the CRHE is isolated upon receipt of a DBA 4-13 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -initiation signal or high radiation signal in the CREOASS intake duct. When the CRHE is isolated, a fixed flow rate of outside air is filtered through CREOASS filter banks which includes a heating coil, roughing filter, upstream HEPA filter, charcoal filter bed, and downstream HEPA filter.The radiological effect of CPPU on CRHE and CREOASS is due to an increase in the core iodine activity released during the DBAs. The effect of CPPU in combination with a 24 month fuel cycle and the AST methodology on the post-LOCA iodine loading of the CREOASS charcoal filter was evaluated. Per Regulatory Guide 1.183, Table 1, a total of 30% of the core iodine is released to the drywell. All iodines (particulates, elemental, organic and stable) in the drywell are conservatively assumed to be released to the environment without crediting radiological decay or removal by SGTS charcoal filtration system, or holdup and plate out of iodine on the main steam piping/condenser surface. Dispersion of the released radioiodine is conservatively based on the limiting 0-2 hour X/Q value for the entire 30-day event. As a result of CPPU, the iodine loading on the CREOASS filters remains a small fraction of the allowable limit of 2.5 mg of total iodine (radioactive plus stable) per gram of activated carbon, identified in RG 1.52. Therefore, CREOASS iodine filter efficiency is not affected by CPPU.SSES has submitted a Licensing Amendment Request (Reference

22) describing the full implementation of the AST methodology based on CPPU power level of 4032 MWt (102% of 3952) that complies fully with RG. 1.183. Per above reference, it is determined that Control Room doses are less than the RG 1.183 criteria for all DBAs at CPPU power level.4.5 STANDBY GAS TREATMENT SYSTEM The Standby Gas Treatment System (SGTS) is one of the Fission Product Control Systems and Structures which provides for fission products control during DBA conditions.

Other systems and structures which provide this function include Primary Containment, Secondary Containment, and the Reactor Building (RB) Recirculation System. These systems and structures are determined to be acceptable for CPPU operation. Primary Containment System is discussed in section 4.1. The secondary containment isolation system is not significantly affected by CPPU and is addressed in section 6.8. The RB recirculation system fan flow capacity, or boundary parameters or any of the system component parameters, or system start signals do not change as a result of CPPU Conditions. Therefore, RB recirculation system is unaffected by CPPU.Following a LOCA, the fission products released from the reactor are contained within the primary containment. Fission product leakage from the primary containment is into the Secondary Containment (Reactor Building) where it is processed via the SGTS prior to being released to environment. During LOCA conditions, the Reactor. Building Recirculation system mixes the leakage from the primary containment with the secondary containment volume so that the concentration of fission products is diluted. SGTS filters the air through the SGTS charcoal and HEPA filters prior to exhaust to the outside atmosphere and maintains a negative pressure inside 4-14 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -secondary containment with respect to outside atmosphere. This minimizes, the potential for uncontrolled release during SGTS operation. The assumptions regarding leakage and exhaust paths from the primary and secondary containments and other sources are described in detail in a separate request for license amendment submitted in support of the full implementation of the Alternative Source Term (AST)methodology for SSES (Reference 22). This submittal is based on the CPPU power level of 4032 MWt (102% of 3952) and complies with RG 1.183.Details of the radiological transport models used in the AST submittal are provided in Reference 22 and include changes from previous assumptions (i.e., pre-AST submittal assumptions) with regard to the fission product control systems and structures. These revised assumptions included values used for primary containment leakage to the secondary containment, primary containment bypass leakage directly to the environment, MSIV leakage via the condenser, secondary containment leakage and drawdown time and SGTS flow. In addition, no credit is assumed for suppression pool scrubbing. Evaluations performed for this CPPU evaluation are consistent with those used in the AST submittal. The Standby Gas Treatment System The Standby Gas Treatment System along with the Reactor Building Recirculation System is designed to maintain secondary containment at a negative pressure and to filter the exhaust air for removal of fission products potentially present during abnormal conditions. By limiting the release of airborne particulates and halogens, the SGTS limits off-site dose following a postulated DBA.The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result Flow capacity Iodine removal capability ]I 4-15 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The SSES SGTS design flow capacity is adequate to maintain the secondary containment at the required negative pressure to minimize the potential for ex-filtration of air from the reactor building. [[]] The total (radioactive plus stable) post LOCA iodine loading on the charcoal adsorbers increases proportionally with the increase in core iodine inventory, which is proportional to core thermal power. Sufficient charcoal mass is present so that the post-LOCA iodine loading on the charcoal remains below the guidance provided by RG 1.52.While decay heat from fission products accumulated within the system filters and charcoal adsorbers increases in proportion to the increase in thermal power, the cooling air flow required to maintain components below operating temperature limits is well* below the cooling flow capability of the system.4-16 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -support of the above conclusions, [[ ]] has been performed in the CLTR to evaluate systems that implement Alternate Source Term (AST) in accordance with RG 1.183.1The results of the bounding AST analysis show that the maximum charcoal loading based on only 50 pounds of charcoal per adsorber train, is approximately 0.26 mg of total iodine per gram of charcoal. This is well below the 2.5 mg/gm maximum value of RG 1.52. The maximum component temperature is approximately 168°F with normal flow conditions and 500'F under a failed fan with minimum cooling flow, which is well below the 625°F charcoal ignition temperature. The SSES SGTS charcoal filter loading is confirmed to be substantially lower than 0.26 mg of total iodine per gram of charcoal. The SGTS filter train design includes the capability to operate in a cooling mode using outside air and a fire suppression deluge water system. It also includes a cross connection between the two trains to provide a cooling air flow path if one of the SGTS fans fails. The maximum component temperatures for the SGTS design are confirmed to be substantially lower than the temperatures identified in the paragraph above for the bounding GE CLTR AST analysis.Er 1]]Based on the discussion above, the SSES fission product control systems and structures are adequate for CPPU conditions. 4.6 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM SSES does not use a Main Steam Isolation Valve Leakage Control System (MSIV-LCS). 4.7 POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM The Combustible Gas Control System is designed to maintain the post-LOCA concentration of oxygen or hydrogen in the containment atmosphere below the lower flammability limit. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result System initiation time Recombiner operating temperature Nitrogen makeup 4-17 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Following a LOCA, the Combustible Gas Control System (CGCS) is designed to maintain the hydrogen concentration of the drywvell and wetwell atmospheres below the lower flammability.limit for hydrogen. The post-LOCA production of hydrogen and oxygen by radiolysis (Figure 4-1) increases proportionally with power level. The increase in radiolysis due to CPPU has an impact on the time available to start the recombiners before reaching the Regulatory Guide 1.7 hydrogen concentration limit of 4% (by volume). Based on starting the recombiners at a 3.5% hydrogen concentration and accounting for a four hour warm-up time, the required start time for the drywell recombiner decreases from 24.5 hours to 19.5 hours and the wetwell recombiner start time decreases from 38 hours to 11.1 hours as shown in figure 4.3. Figure 4.2 shows controlled and uncontrolled oxygen concentration. These changes in CGCS initiation times are primarily a result of increased generation of radiolytic hydrogen and oxygen under CPPU conditions. Because several of the analysis bases and assumptions for the CPPU evaluation are changed from those in the current licensing basis evaluation (e.g. degree of initial mixing between drywell and wetwell atmospheres; initial gas volume in containment; amount of oxygen generation), the magnitude and direction of the evaluated change in recombiner start times may not be a true indication of the impact of increased power level alone. However, the ability of the system to maintain hydrogen and oxygen below flammability limits during the event is not impacted, and remains primarily dependent on plant operator response in accordance with plant operating procedures and accident guidelines. While the reduction in time allowable for starting the drywell and wetwell recombiners* is substantial (from 24.5 hours to 19.5 hours and from 38 hours to 11.1 hours respectively), the ability of the operators to respond to this event is not adversely affected. The existing emergency procedures specify starting the hydrogen recombiners before the hydrogen concentration reaches 2%, which occurs approximately 2 to 3 hours after event initiation, well before the 11.1 and 19.5 hour start times required to maintain hydrogen below the flammability limit.CPPU has no effect on component maximum operating temperatures.. The operating temperature of the recombiners is dependent on the containment oxygen and hydrogen concentration when the recombiners are in operation. Because the oxygen and hydrogen concentrations are maintained below the Regulatory Guide 1.7 flammability limits throughout the event, the maximum operating temperature is not affected by CPPU.]4-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 4-1 Containment Performance Results for DBA-LOCA Parameter CLTP from FSAR CLTPAwith CPPU CPPU 2 Limit Method" Peak Drywell Pressure 44.6 47.9 48.6 53 (psig)Peak Drywell 320' 3373 3373 340 Temperature (fF)Peak Bulk Pool 203 192 211.2 220 Temperature (°F)Peak Wetwell Pressure 35.3 36.7 36.5 53 (psig)Peak Drywell-to-27.0 25.9 25.6 28 Wetwell (Down)Differential Pressure (psid)Notes: 1.2.3.Using the CPPU analysis method with CLTP assumptions. Includes the impact of conservative assumptions with regard to initial containment conditions. These peak drywell temperatures reported are for a large, double-ended guillotine break of a main steam line.4-19 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 4-2 ECCS Performance Results CLTP" CPPU 10 CFR 50.46 Limit Method EXEM BWR-2000 EXEM BWR-2000 Power 106% OLTP 120% OLTP -PCT (OF) N/A < 1844 < 2200< 1945' < 19142 Cladding Oxidation (% Original Clad <.2.0 < 1.0 < 17 Thickness) Hydrogen Generation, Core wide Metal- < 1.0 < 1.0 < 1.0 Water Reaction (%)Coolable Geometry OK OK Meet 1, 2 and 3 above Core Long-Term Cooling OK OK Core flooded to TAF or Core flooded to jet pump.suction elevation and at least one core spray system is operating at rated flow Notes: (1) Pre-CPPU PCT for the limiting CPPU break conditions (2) CPPU PCT with pre-CPPU assumptions 4-20 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 4-3 SGTS Iodine Removal Capacity]]4-21 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -.0 0 C C, C 0)0 V 0)400 Time After LOCA (hours)800 Figure 4-1 Time-Integrated Containment Hydrogen Generation at CPPU 4-22 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -lul 1/1'1 1//7 0 5_____Wethwell -Uncotrolled __v __ _D-Lrywell* -ontrolled I Wetwell -Controlled4 0 0.1 I 10 Time After LOCA (hours)100 1000 Figure 4-2 Controlled and Uncontrolled Oxygen Concentrations 4-23 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -1 0 Wetwell -Uncontrolled ._ Drywell 5 _- --e-Uncontrolledi __ T t Dywell -Controlled etwell -Controlled 0 I1 1 0.1 1 10 100 1000 Time After LOCA (hours)Figure 4-3 Controlled and Uncontrolled Hydrogen Concentrations 4-24 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -5. INSTRUMENTATION AND CONTROL This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapter 7, as it applies to CPPU. The principal instrumentation affected by CPPU is addressed in the following. 5.1 NSSS MONITORING AND CONTROL The instruments and controls used to monitor and directly interact with or control reactor parameters are usually withinithe NSSS. Changes in process variables and their effects on instrument performance and setpoints were evaluated for CPPU operation to determine any related changes. Process variable changes are implemented through changes in normal plant operating procedures. Technical Specifications address instrument AVs and/or setpoints for those NSSS sensed variables that initiate protective actions. The effect of CPPU on Technical Specifications is addressed in Section 5.3. The topics addressed in this evaluation are: Topic CLTR Disposition ' SSESResult 5.1.1.1 Average Power Range Monitors, R Intermediate Range Monitors, and Source Range Monitors 5.1.1.2 Local Power Range Monitors 5.1.1.3 Rod Block Monitor 5.1.2 Rod Worth Minimizer/Rod Control Information System[o 5.1.1 Neutron Monitoring System CPPU affects the performance of the Neutron Monitoring System. These performance effects are associated with the APRMs, Intermediate Range Monitors (IRMs), Local Power Range Monitors (LPRMs), RBM, and Rod Worth Minimizer (RWM).5.1.1.1 Average Power Range Monitors, Intermediate Range Monitors and Source Range Monitors 5-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The increase in power level due to CPPU increases the average flux in the core and at the in-core detectors. The APRM power signals are calibrated to read 100% at the new licensed power (i.e., CPPU RTP). CPPU has little effect on the IRM overlap with the SRMs and the APRMs.Using normal plant surveillance procedures, the IRMs may be adjusted, as required, so that overlap with the SRMs and APRMs remains adequate.The SRM, IRM, and APRM Systems installed at SSES are in accordance with the requirements established by the GE design specifications. [[5.1.1.2 Local Power Range Monitors[[4- i At CPPU RTP, the average flux experienced by the detectors increases due to the average power increase in the core. The maximum flux experienced by an LPRM remains approximately the same because the peak bundle powers does not appreciably increase. Due to the increase in neutron flux experienced by the LPRMs and traversing incore probes (TIPs), the neutronic life of the LPRM detectors may be reduced and radiation levels of the TIPs may be increased. LPRMs are designed as replaceable components. The LPRM accuracy at the increased flux is within specified limits, and LPRM lifetime is an operational consideration that is handled by routine replacement. TIPs are stored in shielded rooms. A small increase in radiation levels is accommodated by the radiation protection program for normal plant operation. 5-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The LPRMs and TIPs installed at SSES are in accordance with the requirements established by the GE design specifications. [[]]The increase in reactor power from CPPU will require more frequent LPRM calibrations to be performed because the current LPRM calibration surveillance interval of 1000 MWD/MT is in units of core exposure. This will cause an undue increase in the wear and tear on the TIP system and the time that the TIP related primary containment isolation valves are open. An increase in the LPRM calibration interval from 1000 MWD/MT to 2000 MWD/MT is proposed to alleviate this condition. Increasing the LPRM calibration interval to 2000 MWD/MT is acceptable. because of improvements in ýfuel analytical bases, core monitoring processes, and nuclear instrumentation. LPRMs are calibrated periodically because of changes in the instrument, sensitivity due to depletion of the fissile detection material. Calibration data is obtained from the TIP system, using the movable neutron detectors to measure the in-core flux distribution for comparison with the LPRM readings.The APRM and RBM systems are the only nuclear instrumentation systems, which use LPRM readings. SSES uses improved LPRM chambers which are NA 300 series. The LPRM calibration interval extension has no significant effect on the APRM or RBM accuracy as the APRM calibration is performed independently from the LPRM calibration. The APRMs are required to be within the calibration criteria on a weekly basis. This calibration is not affected by the extension of the LPRM calibration interval. The ARTS based RBM system is automatically normalized to a reference value which is independent of the APRM when a control blade is selected.The justification to increase the surveillance interval is based on the assumptions used in the safety limit analysis methodology (Reference 25). The safety limit analysis uses an assembly power uncertainty value which includes a component of LPRM depletion uncertainty. This uncertainty component is due to the sensitivity changes during irradiation. The uncertainty component is primarily a function of thermal fluence since the last calibration. The safety limit analysis is performed using an uncertainty based upon the calibration interval of 2500 effective full power hours according to the original power densities so this is equivalent to approximately 2500 MWD/MT. The proposed value of 2000 MWD/MT is chosen to allow for a 25%surveillance extension which is supported by the value used in the safety limit determination. In addition, SSES uses the Framatome POWERPLEX' core monitoring system which utilizes nodal diffusion theory, coupled with plant data and the improved flux instrumentation. The POWERPLEX system is based on accepted BWR calculation methods used to monitor on-line core performance and thus allows accurate assessment of thermal limits.'POWERPLEX is a trademark of Framatome ANP registered in the Unites States and various other countries. 5-3 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -5.1.1.3 Rod Block Monitor[[1]The increase in power level at the same APRM reference level results in increased flux at the LPRMs that are used as inputs to the RBM. The RBM instrumentation is normalized to a reference value. Because the APRM has been rescaled, there is only a small effect on the RBM performance due to the LPRM performance at the higher average local flux. The change in performance does not have a significant effect on the overall RBM performance. The RBMs installed at SSES are in accordance with the requirements established by the GE design specifications. 15.1.2 Rod Worth Minimizer/Rod Control and Information System The Rod Control and Information System (RCIS) is not applicable to SSES.The RWM is a normal operating system that does not perform a safety related function. The function of the RWM is to support the operator by enforcing rod patterns until reactor power has reached appropriate levels. [[]] The TS applicability for the RWM for CPPU is conservatively maintained at the same RTP value as the CLTP value which results in the RWM enforcing control rod patterns over a greater range. The power-dependent instrument setpoints for the RWM are included in the plant Technical Specifications (see Section 5.3.4).5.2 BOP MONITORING AND CONTROL Operation of the plant at CPPU has minimal effect on the BOP system instrumentation and control devices. Based on CPPU operating conditions for the power conversion and auxiliary systems, most process control valves and instrumentation have sufficient range/adjustment capability for use at the CPPU conditions. However, some (non-safety) modifications may be needed to the power conversion systems to obtain CPPU RTP. No safety-related BOP system 5-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -setpoint change is required as a result of the CPPU, with the exception of MSL high flow discussed in Section 5.3.1. The topics considered in this section are: Topic CLTR Disposition SSES Result 5.2.1 Pressure Control System 5.2.2 Turbine Steam Bypass System (Normal Operation)

5.2.2 Turbine

Steam Bypass System (Safety Analysis)5.2.3 Feedwater Control System (Normal Operation)

5.2.3 Feedwater

Control System (Safety Analysis)5.2.4 Leak Detection System 5.2.1 Pressure Control System[[E 5-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The PCS is a normal operating system to provide fast and stable responses to system disturbances related to steam pressure and flow changes to control reactor pressure within its normal operating range. This system does not perform a safety function. Pressure control operational testing is included in the CPPU implementation plan as described in Section 10.4 to ensure adequate turbine control valve pressure control and flow margin is available. S 15.2.2 Turbine Steam Bypass System (Normal Operation) + 4 II]S The Turbine Steam Bypass System is a normal operating system that is used to bypass excessive steam flow. The Turbine Steam Bypass System at SSES is [[Turbine Steam Bypass System (Safety Analysis)The Turbine Steam Bypass System is a normal operating system that is used to bypass excessive steam flow. [[]] The bypass flow capacity is included in some AOO evaluations (Section 9.1).These evaluations demonstrate the adequacy of the bypass system. Cycle specific reload analysis takes credit for the availability of the bypass system in the reload analysis to establish the core operating limits.5-6 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -5.2.3 Feedwater Control System[[The Feedwater Control System is a normal operating system to control and maintain the reactor vessel water level. CPPU results in an increase in FW flow. FW control operational testing is included in the CPPU implementation plan as described in Section 10.4 to ensure that the FW response is acceptable. Failure of this system is evaluated in the reload analysis for each reload core with the FW controller failure-maximum demand event. A LOFW event can be caused by downscale failure of the controls. The LOFW is discussed in Section 9.1.3.[[t 5.2.4 Leak Detection System[[I 5-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -4 The only effect on the LDS due to CPPU is slight increase in the FW temperature and steam flow. [[]] The increased FW temperature results in a small increase in Main Steam tunnel temperature (< 1'F). 5-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version - MSL high flow is discussed in section 5.3.1.[[5.3 TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS Technical Specifications instrument AVs and/or setpoints are those sensed variables, which initiate protective actions and are generally associated with the safety analysis. Technical Specification AVs are highly dependent on the results of the safety analysis. The safety analysis generally establishes the ALs. The determination of the Technical Specification AVs and other instrument setpoints includes consideration of measurement uncertainties and is derived from the ALs. The settings are selected with sufficient margin to minimize inadvertent initiation of the protective action, while assuring that adequate operating margin is maintained between the system settings and the actual limits. There is typically substantial margin in the safety analysis process that should be considered in establishing the setpoint process used to establish the Technical Specification AVs and other setpoints. Increases in the core thermal power and steam flow affect some instrument setpoints. These setpoints are adjusted to maintain comparable differences between system settings and actual limits, and reviewed to ensure that adequate operational flexibility and necessary safety functions are maintained at the CPPU RTP level. Where the power increase results in new instruments being employed, an appropriate setpoint calculation is performed and'Technical Specification changes are implemented, as required. If there is no change in the instrument equipment, the simplified process outlined in the CLTR may be used to determine the instrument AV and setpoint.ER]] The justification for implementing this simplified process for the individual Technical Specification setpoints is provided for each instrument below.[R 5-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -1]]These restrictions are satisfied for SSES, except for the Technical Specification setpoints for Main Steam Line High Flow Isolation, APRM Setdown in Startup Mode, and Turbine First Stage Pressure Scram Bypass.The setpoint methodology of Reference 26 was applied to the Technical Specification setpoints for Main Steam Line High Flow Isolation and APRM Setdown in Startup Mode. All Technical Specification instruments were evaluated for effects from CPPU. This evaluation included a review of environmental (i.e., radiation and temperature) effects, process (i.e., measured parameter) effects and analytical (i.e., AL and margins) effects on the subject instruments. SSES applied their existing setpoint methodology to the Technical Specification instrument setpoints for Turbine First Stage Pressure Scram Bypass.Table.5-1 summarizes the current and CPPU ALs for SSES.The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result 5.3.1 Main Steam Line High Flow Isolation -Setpoint Calculation Methodology 5.3.1 Main Steam Line High Flow Isolation -Setpoint Value 5.3.2 Turbine First-Stage Pressure Scram Bypass -Setpoint Calculation Methodology

5.3.2 Turbine

First-Stage Pressure Scram Bypass -Setpoint Value 5.3.3 APRM Flow-Biased Scram -Setpoint Calculation Methodology 5.3.3 APRM Flow-Biased Scram -Setpoint Value 5.3.4 Rod Worth Minimizer/ RCIS Rod Pattern Controller Low Power Setpoint -Setpoint Calculation Methodology 5-10 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Topic CLTR Disposition SSES Result 5.3.4 Rod Worth Minimizer! RCIS Rod Pattern Controller Low Power Setpoint -Setpoint Value 5.3.5 Rod Block Monitor 5.3.6 RCIS Rod Withdrawal Limiter High Power Setpoint -Setpoint Calculation Methodology 5.3.6 RCIS Rod Withdrawal Limiter High Power Setpoint -Setpoint Value 5.3.7 APRM Setdown in Startup Mode -Setpoint -Setpoint Calculation Methodology 5.3.7 APRM Setdown in Startup Mode -Setpoint Value 1]5.3.1 Main steam Line High Flow Isolation The MSL high flow isolation setpoint is used to initiate the isolation of the MSIV primary containment isolation valves. The only safety analysis event that credits this trip is the MSLBA.For this accident, there are diverse trips from high area temperature and high area differential temperature. For SSES, there is sufficient margin to choke flow. A previous license amendment (Reference

6) reduced the margin for this function because 100% RTP steam flow was increased but the setpoints for this function were not revised. To restore the margin lost as a result of the previous uprate, the AL for CPPU has been increased from 138 percent to 140 percent of rated steam flow in each MSL.For SSES, the AL of 138% of steam flow is changed to 140% and new instrumentation is required (the existing instrumentation does not have the required upper range limit to re-span the instrument loops to accommodate the new setpoint).

Therefore, a new setpoint was calculated using the methodology as noted in Section 5.3. Therefore, a Technical Specification change is required because the setpoint function is listed in units of psid versus percent rated steam flow.5-11 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -5.3.2 Turbine First Stage Pressure Scram and Recirculation Pump Trip Bypass CPPU results in an increased power level and the high-pressure turbine (HPT) modifications result in a change to the relationship of turbine first-stage pressure, to reactor power level. The turbine first-stage pressure scram bypass signal is used to reduce scrams and recirculation pump trips at low power levels where the turbine steam bypass system is effective for turbine trips and generator load rejections. In the safety analysis, this function only applies to events at low power levels that result in a turbine trip or a load rejection. The scram bypass applicability threshold for CPPU is reduced such that the applicability threshold occurs at a slightly lower point in terms of absolute power (MWt) than for the current value, establishing a slightly more conservative relationship to the transient analysis basis and scram avoidance range of the bypass valves.Because the HPT will be modified to support achieving the CPPU RTP level, a new applicability threshold (in psig) corresponding to the slightly more conservative absolute power than the current design limit was established. Therefore, an instrument setting was established using existing PPL Susquehanna setpoint methodology as noted in Section 5.3, and the Technical Specification applicability for this function in percent RTP has been changed. The instrument settings (in psig) will be revised prior to CPPU implementation. To assure that the new setting is appropriate, CPPU plant ascension startup test or normal plant surveillance will be used to validate that the actual plant scram bypass functions are cleared consistent with the safety analysis.5.3.3 APRM Flow Biased Scram[[5-12 " Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -0 Er]] The clamped setpoint retains its value (in percent power).The NTSPs were adjusted by the same difference as the changes in the ALs. This allows the current license basis to be maintained through the application of the same uncertainties in the same manner as previous setpoint evaluation. 5.3.4 Rod Worth Minimizer/RCIS Rod Pattern Controller Low Power Setpoint[[1-0 5-13 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The RCIS Rod Pattern Controller is not applicable to SSES.The Rod Worth Minimizer LPSP is used to bypass the rod pattern constraints established for the control rod drop accident (CRDA) at greater than a pre-established low power level. The consequences of the CRDA are acceptable above 10% CLTP, and the rod pattern constraints are no longer necessary. The measurement parameter is main steam flowrate.The LPSP in the plant Technical Specifications is conservatively kept at the same value in terms of percent power and no Technical Specification change is required. This approach does not affect the limitations on the sequence of control rod movement to the absolute core power level for the LPSP associated with the requirements of the CRDA. To assure that the new setting is appropriate, CPPU plant ascension startup test or normal plant surveillance will be used to validate that the enable interlock is established consistent with the safety analysis.5.3.5 Rod Block Monitor[[l 0 The severity of rod withdrawal error (RWE) during power operation event is dependent upon the RBM rod block setpoint. This setpoint is only applicable to the control rod withdrawal error. The power-dependent RBM enable setpoint is maintained at the same percent power for CPPU, and no Technical Specification change is required. The cycle specific reload analysis establishes the rod block setpoints. The trip setpoints (corresponding to the various power-dependent setpoint levels) are evaluated as part of the cycle specific reload analysis.5-14 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -5.3.6 RCIS Rod Withdrawal Limiter High Power Setpoint The RCIS Rod Withdrawal Limiter High Power Setpoint is not applicable to SSES.5.3.7 APRM Setdown in Startup Mode The value for the Technical Specification safety limit for reduced pressure or low core flow condition is established to satisfy the fuel thermal limits monitoring requirements described in Section 2.1.]] The current design limit is reduced as required by the reduction in the threshold for thermal limits monitoring requirements. A new nominal trip setpoint and AV are calculated using GE setpoint methodology as noted in Section 5.3.5.4 CHANGES TO INSTRUMENTATION AND CONTROLS In the CLTR SER, the staff requested that the plant-specific submittal address all CPPU-related changes to instrumentation and controls, such as scaling changes, changes to upgrade obsolescent instruments, and changes to the control philosophy. Table 5-2 provides this information. No obsolescent instrument changes are required as a result of CPPU and there are no changes to instrument philosophy as a result of CPPU.5-15 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 5-1 SSES Analytical Limits For Setpoints AnalyticalLimits Parameter Current CPPU APRM Calibration Basis (MWt) 3489 3952 APRM High Flux Flow Bias (Scram)(') ALs TLO (% RTP) 0.62Wd + 67.0 0.55Wd + 63.5 SLO (% RTP) 0.62Wd + 61.6 0.55Wd + 58.7 Clamped TLO (% RTP) 118 118(2)Clamped SLO (% RTP) 118 118(2)APRM Flow Biased Rod Block ALs.TLO (% RTP) 0.62Wd + 62.5 0.55Wd + 59.0 SLO (% RTP) 0.62Wd + 57.1 0.55Wd + 54.2 Clamped TLO (% RTP) 113.5 113.5 Clamped SLO (% RTP) 113.5 113.5 APRM Setdown in Startup Mode (% RTP)Scram DL 25 23 Rod Block Monitor Power Setpoints* Low Power Setpoint (LPSP) No Change(3)* Intermediate Power Setpoint (IPSP) No Change(3)* High Power Setpoint (HPSP) No Change(3)Trip Setpoints (%RTP):-Low Trip Setpoint (LTSP) No Change(3)* Intermediate Trip Setpoint (LTSP) No Change(3)-High Trip Setpoint (HTSP) No Change(3)Rod Worth Minimizer LPSP (% RTP) No Change Main Steam Line High Flow Isolation (% rated steam 138 140 flow)Turbine First-Stage Pressure Scram Bypass (% RTP) 30 26 Notes: 1.2.3.No credit is taken in any safety analysis for the flow referenced setpoints. The cycle specific rcload analysis is used to determine any change in the rod block trip setpoints. 5-16 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 5-2 Changes to Instrumentation and Controls Parameter Change MSL High Flow Replace pressure switches to encompass new steam flow values MSL High Flow Setpoint changes for new setpoints for MSIV isolation at new steam flows Feedwater Flow Re-span or replace transmitters, indicators and associated loop instruments to encompass new range Condensate Flow Re-span or replace transmitters, indicators and associated loop instruments to encompass new range EHC Pressure Sensing Install Steam Line resonance Cards to dampen 3 rd harmonic frequency EHC Turbine Control Valve Digital Position Modify Turbine Control Valve Digital Position cards for new steam flow conditions. EHC Power load Imbalance Recalibrate for new operating conditions. Condensate Demineralizer Discharge Header Re-span transmitters to encompass new range Temperature RFP Hydrogen Injection Flow Re-span transmitter to encompass new range APRM flow biased Simulated Thermal Power Revise setpoints to CPPU values for both two loop and single Scram Setpoints loop operation. APRM flow biased Simulated Thermal Power Revise setpoints to CPPU values for both two loop and single Rod Block Setpoints loop operation. APRM Neutron Flux Upscale Setdown Scram Revise setpoints to CPPU values APRM Neutron Flux Upscale Setdown Rod Block Revise setpoints to CPPU values RFP Hydrogen Injection High Flow Change alarm setpoint.RWM LPSP Revise setpoints to reflect increased steam flow at 10% RTP Off Gas Recombiner Oxygen Injection Flow Re-span transmitter to encompass new range Condensate Pump Suction Oxygen Injection Flow Re-span transmitter to encompass new range Reactor Recirculation Runback Limiter No. 2 Revise logic to remove the confirmatory low reactor water level trip signals from logic that initiates runback of reactor recirculation system upon detection of trip of FW pump and/or condensate pump. Change condensate pump trip input signal to pump breaker position and increase recirculation system runback rate.Standby Liquid Storage Tank Low Level Change alarm setpoint Standby Liquid Storage Tank High Level Change alarm setpoint Standby Liquid Control System Logic Revise System Logic to allow for single pump initation Feedwater Dilution Steam Flow Re-span transmitter to encompass new range Main Steam Line Flow Re-span transmitters, indicators and associated loop instruments to encompass new range 5-17 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 5-2 Changes to Instrumentation and Controls eter ...Change:ii:: .:- :Parame Main Steam Line Differential Pressure Change alarm setpoint RHR Pump Logic (Appendix R) Logic change to eliminate fire induced failure mechanisms. Steam Flow Recorder Re-span to encompass new range Feedwater Flow Recorder Re-span to encompass new range.Turbine I" Stage pressure Recalibrate for revised scram bypass value.Reactor Feed Pump Turbine Speed Control Modify to support higher pump operational speeds including utilization ofFeed Pump Turbine Digital Speed Control.Reactor Feed Pump Low Suction Pressure Trip Revise setpoints to CPPU values and increase time delay stagger.Reactor Feed Pump seal water Injection Increase the setpoint of the temperature controller. Temperature Unit 2 Overall Differential Protection Relay Revise setpoint to accommodate CPPU condition. 5-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -6. ELECTRICAL POWER AND AUXILIARY SYSTEMS This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapters 8 and 9, that applies to CPPU.6.1 AC POWER The SSES AC power supply includes both off-site and on-site power. The on-site power distribution system consists of transformers, buses, and switchgear. AC power to the distribution system is provided from the transmission system or from onsite Diesel Generators. Plant electrical characteristics are given in Table 6-1. The topics addresses in this evaluation: Topic CLTR Disposition SSES Results AC Power (degraded voltage)AC Power (normal operation) Emergency Diesel Generator Fuel Oil Storage and Transfer 6.1.1 On-Site Power (degraded voltage)The on-site power distribution system loads were reviewed under normal and emergency operating scenarios for CPPU conditions. Loads were computed based on equipment nameplate data or BHP. These loads were used as inputs for computation of load, voltage drop, and short-circuit current values. Operation at the CPPU RTP level is achieved for normal and emergency conditions by operating equipment within the nameplate rating running kW and BHP.There is no significant change in electrical demand load associated with the power generation system; therefore the existing load flow and short circuit calculations verify the adequacy of the on-site AC system for the proposed changes. The existing protective relay settings are adequate to accommodate the increased load on the 13.8kV power system. Selective coordination is maintained between the Reactor Recirculation M-G Set and condensate pump motor feeder breakers and the 13.8kV switchgear main feeder breaker.Station loads under emergency operation/distribution conditions (Emergency Diesel Generators) are based on equipment nameplate data except for the ECCS pumps where a conservative high flow BHP is used. CPPU conditions are achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the required pump motors;therefore, under emergency conditions the electrical supply and distribution components are adequate.6-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -]] The systems have sufficient capacity to support all required loads to achieve and maintain safe shutdown and to operate the ECCS equipment following accidents and transients. The current emergency diesel generator fuel oil storage volume, as required by the plant technical specification, is based on the continuous full load diesel rating, and not DBA loads.Thus the Emergency Diesel Fuel Oil Storage and Transfer System is not affected by CPPU.A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criteria 17 (10 CFR 50, Appendix A). The analysis establishes grid voltage schedules, generator reactive power limits and reduced generation limits that are required under certain pre-event outages. At full MWe, the reactive limits of the generator are -210/+425 MVAR; however, the grid stability analysis shows that more limiting reactive limits apply under certain system configurations. The leading power factor is restricted by the under excited reactive ampere limiter.6.1.2 AC Power (normal operation) The existing off-site and on-site electrical equipment was determined to be adequate for normal operation with the CPPU electrical output as shown in Table 6-2.The review concluded the following: " The Isolated Phase Bus Duct is rated at 35,000 Amperes and supports the Generator output at the CPPU conditions;" The BHP of the recirculation M-G set motors increases 6.43% for CPPU but remains within its nameplate capability; The electrical demand load associated with power generation system motors for the condensate pump will increase for CPPU but will remain within their nameplate rating.The system pumps experience increased flow demand at CPPU conditions; The existing Unit I Main Power Transformers were determined to be adequate for operation with CPPU related electrical output of the generator; The Unit 2 Main Transformers ratings are upgraded to meet the requirements for CPPU operation and the tap setting is changed;The 230kV & 500kV switchyard components including circuit breakers, disconnect switches, and current transformers are suitable to meet CPPU continuous current and short circuit requirements after replacement of the 230kV synchronizing breaker;The protective relaying for the main generator, transformer, and switchyard is adequate for the CPPU generator output, except for the Unit 2 overall differential protection relay which requires a setpoint change;* Capacitor banks are installed in the 230kV and 500kV switchyard for MVAR support; and* The main generator rating is increased to 1354 MVA for CPPU.6-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -6.2 DC POWER The Susquehanna DC power distribution system provides control and motive power for various systems/components within the plant. The topics addressed in this evaluation are: ToPic CLTR Disposition SSESResults DC power requirements ]The direct current (DC) loading requirements in FSAR were reviewed, and no reactor power-dependent loads were identified. The DC power distribution system provides control and motive power for various system/components within the plant. In normal and emergency operating conditions, loads are computed based on equipment nameplate ratings. These loads are used as inputs for the computation of load, voltage drop, and short circuit current values.Operation at the CPPU conditions does not increase any load or revise any component operating duty cycle; therefore, the DC power distribution system remains adequate.6.3 FUEL POOL The SSES fuel pool systems consist of storage pools, fuel racks, and the Fuel Pool Cooling and Cleanup System (FPCCS). The objective of the fuel pool systems is to provide specially-designed underwater storage space for the spent fuel assemblies. The objective of the fuel pool systems is to remove the decay heat from the fuel assemblies and maintain the fuel pool water within specified chemistry and temperature limits. The effects of CPPU on the SSES fuel pool are addressed in the following evaluation: Topic CLTR Disposition SSES Result 6.3.1 Fuel Pool Cooling (normal core offload)6.3.1 Fuel Pool Cooling (full core offload)6.3.2 Crud Activity and Corrosion Products 6.3.3 Radiation Levels 6.3.4 Fuel Racks Each unit at SSES has a spent fuel pool, with high-density fuel storage racks with a capacity for 2840 fuel assemblies and a FPCCS. Each FPCCS removes decay heat from spent fuel stored in the fuel pool and maintains the specified water quality. This capacity provides space for scheduled batch offloads plus one complete core load of fuel before the pool is filled. Each fuel pool also has ten storage locations for control rods, control rod guide tubes, defective fuel 6-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -containers, or sipping containers. Both SSES units share a common cask storage pit that is accessed through a transfer canal from each unit's spent fuel pool. When filled, the cask storage pit connects both fuel pools resulting in shared decay heat removal systems. This is the normal operating configuration and-results in redundant cooling systems and level and temperature instrumentation. The FPCCS consists of three pumps, three heat exchangers, and one filter demineralizer per unit. Normal refueling outage lineup for the FPCCS has the common Filter/Demineralizer aligned to the outage unit such that the outage unit has (3) heat exchangers, (2)pumps, and (2) demineralizers in service. Increased heat removal capability is provided by scheduling refueling offloads during periods of reduced river water temperature. During refueling conditions, the three FPCCS heat exchangers in the outage unit are typically supplied with river water (also referred to as Supplemental Decay Heat Removal) with an assumed.maximum river temperature limit of 75°F. The three FPCCS heat exchangers in the non-outage unit are supplied with service water (SW) with a maximum design temperature of 95°F. The FPCCS is not designed to Seismic Category I requirements and is not supplied with Class 1E power.In addition to cooling provided by the FPCCS, the RHR system can be aligned to provide decay heat removal in the RHRFPC mode. The RHRFPC can provide cooling to the SFP from a safety-related, Seismic Category I system. The RHRFPC Cooling Mode can remove the maximum abnormal heat load and the ESW system can provide make up water to keep the fuel covered following a postulated loss of I fuel pool cooling, preventing the fuel pools from boiling. Initiation of the RHRFPC system requires operation of several manual and motor operated valves (which are accessible following a seismic/LOOP event) to establish a flow path.Proper operation of active components in the RHRFPC mode is confirmed on a periodic basis in accordance with plant procedures. This manual evolution is conservatively assumed to be completed within 25 hours after any loss of all FPCCS.6.3.1 Fuel Pool Cooling and Clean Up System and RHRFPC Cooling Mode The SSES SFP bulk water temperature is maintained below the licensing limits of 125°F and greater than 25 hour time to boil when a Seismic Category I, Class IE cooling system is not assisting in the fuel pool cooling. The spent fuel pools are normally connected in a cross tied configuration during dual unit operation and refueling outages. When the spent fuel pools are not in a cross tied configuration, additional restrictions on heat loads are required by the Technical Requirements Manual. These restrictions are not changed by the CPPU and remain valid. This configuration assures that the time to boil following a loss of normal fuel pool cooling is a minimum of 25 hours. During a refueling outage, when the spent fuel pools are interconnected with the reactor vessel via the reactor cavity, the time to boil can be less than the 25 hours criteria due to the combined decay heat load in the spent fuel pools and the reactor vessel. However, one loop of RHR in normal shutdown cooling is required to be in service prior to interconnecting the spent fuel pools with the reactor cavity. The SFP bulk water temperature is maintained below the administrative limit of 115'F during outage conditions to provide margin below the FSAR limit of 125°F.As stated in Reference 56, a loss of FPCCS fuel pool cooling is only postulated for a seismic/LOOP event. If a seismic/LOOP event occurred during dual unit power operation with 6-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -cross-tied pools, manual alignment of one loop of RHRFPC will prevent SFP boiling. A single failure of one RHR loop in one of the units would still allow a sufficient number of RHR loops to cool both reactors and the cross-tied spent fuel pools. The unit with the failed RHR loop would utilize the normal shutdown cooling mode to cool the reactor. The other unit, which has two RHR loops available, would operate one loop in RHRFPC to cool the spent fuel pools and the other loop in alternate shutdown cooling for long-term decay heat removal from reactor.Since the normal shutdown cooling mode of RHR and RHRFPC mode share a common suction header, concurrent operation of both RHR loops in these modes can not be performed on a given unit. However, alternate shutdown cooling mode and RHRFPC can be performed concurrently with two RHR loops because they do not share a common suction path. The CPPU SFP heat load is greater than the CLTP heat load. The CPPU heat loads at the limiting full core offload condition and the normal batch offload are calculated based on a 24-month fuel cycle, ANSI/ANS 5.1-1979 decay heat standard, and the spent fuel pools are full of fuel assemblies which have been operated at CPPU conditions. CPPU does not affect the heat removal capability of the FPCCS or the spent fuel pool cooling mode of the RHR system. CPPU results in slightly higher core decay heat loads during refueling. A full core offload in one pool with the other pool full of spent fuel is defined as an EHL (emergency heat load). A normal batch offload of 342 assemblies was assumed for CPPU A normal batch offload was analyzed (Table 6-4) for the SFP FSAR and Administrative temperature limits. The EHL case (Table 6-5) is based on the event occurring 6 days after the opposite units batch refueling has ended. An EHL condition will require both trains of FPCCS in operation and one train of RHR in the outage unit aligned for RHR shutdown cooling (SDC).The Condensate Transfer System is the normal SFP makeup source for the Unit 1 FPCCS and the Demineralized Water Transfer System is the normal makeup source for the Unit 2 FPCCS.These systems are not affected by CPPU and remain adequate for CPPU conditions. The ESW System is the emergency makeup source for both units FPCCS. The ESW system is a Seismic Category I system and can provide makeup to the spent fuel pool if required after a seismic event if normal makeup capability is lost. Each ESW loop has the capacity to provide 35 GPM of makeup to each fuel pool (70 GPM total), for up to 30 days. At CPPU conditions, this system maintains the capability to compensate for water lost due to boil-off and evaporation to maintain at least 23 feet of water above the fuel at all times. Existing plant instrumentation and procedures provide adequate indications and direction for monitoring and controlling SFP temperature and level during normal batch offloads and the limiting full core offload. Operating procedures provide mitigation strategies including placing additional cooling in service, stopping fuel movement, and initiating fuel pool make-up if necessary. The symptom based entry conditions and mitigation strategies for these procedures do not require changes as a result of CPPU.6.3.2 Crud Activity and Corrosion Products The total crud in the SFP is not expected to increase for CPPU. Any increase would be insignificant and SFP water quality would be maintained by the FPCCS.6-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -6.3.3 Radiation Levels The normal radiation levels around the SFP may increase slightly primarily during fuel handling operation. As stated in Section 8.5, radiation levels in those areas of the plant, which are directly affected by the reactor core and spent fuel increase by the percentage increase in the average power density of the fuel bundles. Therefore, it is conservative to estimate that radiation dose rates increase 14% for CPPU.The design of spent fuel pools is typically very conservative from the perspective of radiation exposure such that changes in the fuel inventory / bundle surface dose rate of 14% results in a small change in operating dose rates. The CPPU evaluation concluded that the SFP. area remains within the zone limit of 2.5 mR/hr. Therefore, the spent fuel pool area dose rates for CPPU will remain within acceptable limits.6.3.4 Fuel Racks The increased decay heat from the CPPU will result in a higher heat load in the fuel pool during long-term storage. The fuel racks are designed for higher temperatures (212'F) than the licensing limit of 125'F. There is no effect on the design of the SSES fuel racks because the original fuel pool design temperature is not exceeded.[[6.4 WATER SYSTEMS The SSES water systems are designed to provide a reliable supply of cooling water for normal operation and DBA conditions. The topics addressed in this evaluation are: Topic CLTR Dis osition SSES Result Water systems performance (normal operation) 6-6 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Water systems performance (safety related)Suppression pool cooling (RHR service operation) Ultimate heat sink 6.4.1 Service Water Systems The safety related service water systems include the Emergency Service Water System (ESW), the Residual Heat Removal Service Water (RHRSW) system, and the Ultimate Heat Sink (UHS). The Service Water System (SW) is non-safety related.6.4.1.1 Safety-Related Loads The safety-related service water systems are designed to provide a reliable supply of cooling water during and following a DBA for essential equipment and systems.6.4.1.1.1 Emergency Service Water System The Emergency Service Water (ESW) system is a safety related system designed to remove heat from HVAC coolers, Diesel Generator coolers, Emergency Core Cooling (ECCS) components, and other equipment required to operate under normal and accident conditions, including Loss of Offsite Power (LOOP) and Loss of Coolant Accident (LOCA) conditions. The system draws water from the Ultimate Heat Sink (UHS) spray pond, pumps it to the various heat exchangers, and returns it to the spray pond by way of a spray network which dissipates the heat to the atmosphere. The safety-related performance of the ESW system during and following the most limiting design basis event, the LOCA, for the following equipment and systems is not dependent on RTP: " Diesel Generator Coolers" RHR and CS Pump Room Coolers* HPCI and RCIC Pump Room Coolers* RHR Pump Motor Coolers* Control Structure Chiller Condenser* Unit 2 Direct Expansion Unit (Emergency Switchgear Room Cooling)ESW also can be used to cool the Turbine Building Closed Cooling Water heat exchangers and the Reactor Building Closed Cooling Water heat exchangers. These are non-safety systems that can be used in the event of LOOP or loss of the non-safety service water system. These systems are only supplied with cooling from ESW if their use will not jeopardize cooling of ESW safety-related loads required for safe shutdown.6-7 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -As demonstrated by the CPPU evaluation of the UHS, Section 6.4.5, the design temperature of the ESW will remain at 97°F following CPPU implementation. The effects of the CPPU on the safety-related ESW heat exchangers design process heat loads have been evaluated and shown to be bounded by the CLTP design. The process heat loads for safety-related ESW users are not impacted by CPPU because they are based on bounding design conditions that envelope CPPU conditions. Similarly, the ESW design minimum flow rates, velocities, and tube plugging limits have been determined based on bounding design conditions and therefore not impacted by the CPPU.As discussed in Section 6.4.5, the large spray arrays of the UHS spray pond will be modified by capping approximately 44 of the array's 356 spray nozzles to improve spray efficiencies and maintain the 97°F design temperature. The ESW system is planned to be field flow tested as part of post-modification testing prior to CPPU implementation to confirm that the system can perform its intended function with the additional pump backpressure and that minimum flow requirements will be met for worst case system alignments. As discussed in Section 6.7.1, a modification is being performed that will allow ESW cooling to be provided to the I C/2C and I D/2D RHR pumps from both loops of ESW in order to ensure that adequate cooling exists for all possible fire scenarios. The CPPU does not modify the ESW system design temperature or pressure, therefore, the system's piping and components can meet all their safety and design objectives -following CPPU implementation. As discussed in Section 6.3 the ESW system provides makeup water to the Spent Fuel Pool (SFP)in the event of a complete loss of SFP cooling capability. Design calculations demonstrate that the system can provide sufficient ESW makeup water to support the required CPPU boil-off rate identified in Section 6.3.CPPU implementation will not adversely affect the SSES Generic Letter 89-13 program with regard to ESW system heat exchanger performance. The ESW design minimum flow rates and design temperatures are not changed by CPPU implementation. In addition, heat exchanger tube plugging margins have been maintained. Therefore, no Generic Letter 89-13 program changes are required as a result of CPPU implementation. 6.4.1.1.2 Residual Heat Removal Service Water System The RHR Service Water (RHRSW) system is a safety related system designed to remove reactor core decay heat during normal or emergency conditions. The systemj draws water from the Ultimate Heat Sink (UHS) spray pond, pumps it to the RHR heat exchangers and returns it to the spray pond by way of a spray network, which dissipates the heat to the atmosphere. The system can also be used to cool the RHR heat exchangers in the Fuel Pool Cooling (FPC) Assist Mode.The containment cooling analysis in Section 4.1.1 shows that the CPPU post-LOCA RHR system heat load increases due to an increase in the maximum suppression pool temperature following a LOCA. The post-LOCA containment and suppression pool responses have been calculated based on an energy balance between the post-LOCA heat loads and the existing heat removal capacity of the RHR and RHRSW systems. As discussed in Sections 3.5.2 and 4.1.1, the existing suppression pool structure and associated equipment have been reviewed for acceptability based on this 6-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -increased post-LOCA suppression pool temperature. Therefore, the containment cooling analysis and equipment review demonstrate that the suppression pool temperature can be maintained within acceptable limits in the post-accident condition at CPPU based on the existing capability of the RHRSW system. The RHRSW system has sufficient capacity at CPPU to supply adequate cooling in the RHR Fuel Pool Cooling Assist Mode. In addition, the RHRSW system has sufficient capacity to serve as a standby coolant supply for long term core and containment cooling as required for CPPU conditions. The RHRSW system flow rate is not changed.As demonstrated by the CPPU evaluation of the UHS, Section 6.4.5, the design temperature of the RHRSW will remain at 97'F following CPPU implementation. The effects of the CPPU on the RHR heat exchanger's design process heat loads have been evaluated (see Sections 6.4.5, 4.1.1, and 6.3.1) and have been shown to maintain the plant design limits for reactor and containment cooling, RHR Fuel Pool Cooling Assist Mode, and the Ultimate Heat Sink (UHS) using existing RHRSW design minimum flow rates and maximum tube plugging limits.As discussed in Section 6.4.5, the large spray arrays of the UHS spray pond will be modified by capping approximately 44 of the array's 356 spray nozzles to improve spray efficiencies and maintain the 97°F design temperature. The RHRSW system is planned to be field flow tested as part of post-modification testing prior to CPPU implementation to confirm that the system minimum flow requirements will be met under the worst case system alignments. The CPPU does not modify the RHRSW. system design temperature or pressure, therefore, the system's piping and components can meet all their safety and design objectives following CPPU implementation. CPPU implementation will not adversely affect the SSES Generic Letter 89-13 program with regard to RHRSW heat exchanger performance. The RHRSW design minimum flow rates and design temperatures are not changed by CPPU implementation. In addition, heat exchanger tube plugging margins have been maintained. Therefore, no Generic Letter 89-13 program changes are required as a result of CPPU implementation. 6.4.1.2 Service Water System (Non-safety) The Service Water (SW) system is designed to remove heat from heat exchangers located in the Control Structure, the Turbine Building, the Reactor Building, and the Radwaste Building and to transfer this heat to the cooling towers where it is dissipated to the atmosphere. The SW system recirculates cold water from the cooling tower outlet and returns the heated water to the cooling tower via the Circulating Water discharge header. The SW system is designed to operate during normal plant operation and during plant shutdowns with offsite power available. The system has no safety related function.As a result of CPPU, the heat load on the SW system increases < 3.6%. The existing design service water temperature assumed for heat exchanger capacity calculations varies from 92°F to 95'F. The design temperature for SW system piping, piping supports and components is 95'F. For components cooled by SW, heat loads at CPPU conditions remain below the heat exchanger design heat load capacity with a SW supply temperature of 92°F.With the implementation of CPPU, the increase in heat load on the cooling towers will result in slightly higher SW supply temperatures than currently being experienced. These short-lived elevated SW supply temperatures above 92°F, which occur occasionally under current CLTP 6-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -operating conditions, may require mitigating measures or temporary core thermal power reductions because heat loads on a limited number of heat exchangers may exceed design capacities. The need for mitigating measures or temporary core thermal power reductions are identified by existing high temperature alarms in process systems cooled by SW and are controlled by existing plant procedure. The maximum expected SW supply temperature is less than 95'F.The SW system flow rate at CPPU conditions remains below the system design flow rate and pump NPSH available exceeds NPSH required at CPPU conditions. 6.4.2 Main Condenser/Circulating Water/Normal Heat Sink Performance The main condenser, circulating water, and normal heat sink (cooling tower) systems are designed to remove the heat rejected to the condenser and thereby maintain the condenser backpressure as recommended by the turbine vendor. Maintaining adequately low condenser pressure assures the efficient operation of the turbine-generator and minimizes vibratory stress on the turbine last stage blades.CPPU operation increases the heat rejected to the condenser and, therefore, reduces the difference between the operating pressure and the recommended maximum condenser pressure.The maximum condenser operating pressure has been established to limit the condensate temperature entering the condensate demineralizers. If condenser pressures approach this maximum limit, a reactor thermal power reduction would be required thereby reducing the heat rejected to the condenser and maintaining condenser pressure and condensate demineralizer inlet temperatures within established limits.The main condenser, circulating water, and heat sink systems are not being modified for CPPU operation. The thermal performance of these systems was evaluated for CPPU. This evaluation was based on the 100% CPPU power design duty over the actual range of circulating water inlet temperatures, and confirms that the condenser, circulating water system, and cooling tower (heat sink) are adequate for CPPU operation. The evaluation of the circulating water system at CPPU conditions indicates sufficient system capacity to ensure that the plant maintains adequate condenser backpressure while meeting all environmental permit conditions related to the plant cooling towers. The effect of CPPU on the flooding analyses is addressed in Section 10.1.2.Current condensate temperature limitations may require load reductions at the upper range of the anticipated circulating water inlet temperatures. 6-10 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -6.4.2.1 Discharge Limits The safety related and non-safety related cooling water loads potentially affected by CPPU are addressed in the following subsections. The only potential environmental concern is the effect on thermal discharges. SSES uses cooling-tower-based heat dissipation system to remove waste heat from the condensers. A concrete spray pond is the station's ultimate heat sink for the ESW system and RHRSW System. There is a small increase in the amount of water withdrawn from the Susquehanna River for increased Cooling Tower losses including tower blowdown, and in the amount of waste heat discharged to the river. The SSES NPDES permit contains no discharge temperature limits, per se, but discharges must adhere to state water quality standards for waters designated "WWF" (warm water fishes). These water quality standards specify different maximum allowable water. temperatures at different times of the year. The environmental effects of the CPPU are controlled such that none of the water quality standards are exceeded for the CPPU. Additional discussion on discharge limits is provided in the Supplemental Environmental Report provided separately.

6.4.3 Reactor

Building Closed Cooling Water System The Reactor Building Closed Cooling Water (RBCCW) System transfers heat from systems/components located in the Reactor and Radwaste Buildings during normal operation. In the event of a Loss of Offsite Power (LOOP), the RBCCW may be used to provide cooling water to the primary containment cooling system in the Drywell Cooling mode of operation. Cooling water to the RBCCW Heat Exchangers is provided by Service Water (SW) during normal operation and by Emergency Service Water (ESW) during a LOOP. The system has no safety related function, but provides a buffer between potentially contaminated systems/components and the SW/ESW.The total heat load on the RBCCW system slightly increases (< 0.1%) as a result of CPPU. The RBCCW heat loads are mainly dependent on the reactor vessel temperature and/or flow rates in the systems cooled by the RBCCW. The change in vessel temperature is minimal, and does not result in any significant increase in drywell cooling loads. The flow rates in systems cooled by the RBCCW do not change significantly due to CPPU (e.g., Reactor Recirculation and Reactor Water Cleanup pump cooling) and, therefore, are minimally affected by CPPU. The operation of the remaining equipment cooled by RBCCW (e.g., sample coolers and drain sump coolers) is not power-dependent and is not affected by CPPU.The RBCCW system contains sufficient redundancy in pumps and heat exchangers to assure that adequate heat removal capability is available. Sufficient heat removal capacity is available to accommodate the small increase in heat load due to CPPU.6.4.4 Turbine Building Closed Cooling Water System The Turbine Building Closed Cooling Water (TBCCW) system is designed to cool auxiliary plant equipment located in the Turbine Building during normal operation and may be used in the event of a Loss of Offsite Power (LOOP). Cooling water to the TBCCW Heat Exchanger is provided by Service Water (SW) during normal operation and by Emergency Service Water (ESW) during a LOOP or a loss of SW. The system has no safety related function.6-11 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The only heat load increase on the Turbine Building Closed-Cooling Water (TBCCW) system due to CPPU is from the condensate pump motor bearing coolers. As a result of CPPU, the overall TBCCW system heat load increase during normal operation is < 3.1%. The TBCCW system has sufficient heat removal capacity available to accommodate this minor increase in heat load because the system is thermally oversized. No increase in TBCCW system flow rate is required due to CPPU. The TBCCW system contains sufficient redundancy in pumps and heat exchangers to assure that adequate heat removal capability is available at CPPU conditions.

6.4.5 Ultimate

Heat Sink As described in FSAR section 9.2.7, the SSES UHS is an 8-acre, 25 million gallon concrete lined spray pond. As a result of operation at the CPPU RTP level, the post-LOCA ultimate heat sink heat load increases, primarily due to higher reactor decay heat.A review was performed to evaluate the increased UHS heat load for the CPPU. The review concludes that the existing UHS system, modified as noted below, has sufficient capacity to maintain a maximum design temperature of 97'F following a design basis LOCA in one unit and concurrent safe shutdown of the other unit. The results of the evaluation are provided in Table 6-3. The current Technical Specifications for UHS temperature limits are adequate, due to conservatism in the original design.The UHS spray system consists of 2 divisions, each with a large and small spray array. To increase the heat dissipation performance of the UHS for CPPU operations, each of the large spray arrays are being modified by capping approximately 44 of the array's 356 spray nozzles. This modification effectively increases the spray nozzle pressure which in turn increases spray height and thus, the array's, cooling effectiveness. The increase in spray nozzle pressure increases the pond's evaporation and drift losses. The post-accident analysis of the UHS under assumed maximum water loss conditions was evaluated for CPPU conditions. This analysis confirmed that the UHS, modified with fewer large spray array nozzles, has adequate inventory to provide cooling for 30-days without makeup. The UHS provides makeup water to both SFPs in the event of a complete loss of SFP cooling capability. The UHS maximum water loss analysis also demonstrates that the UHS can provide sufficient makeup water to meet the CPPU SFP boil-off rate identified in Section 6.3.The ESW and RHRSW systems (Sections 6.4.1.1.1 and 6.4.1.1.2) are planned to be field flow tested as part of post-modification testing prior to CPPU implementation to ensure that these systems perform their intended design function with the additional pump backpressure associated with this modification. Spray array flow rate will be measured as part of this testing. Measurement of the large spray array flow rate will allow the average nozzle flow rate to be determined. The average nozzle flow rate, in turn, will allow confirmation that analytical assumptions for nozzle pressure and spray height are conservative based on known correlations between flow and spray height.Each division of the UHS spray arrays are bypassed during cold weather operations via a bypass line. Remote isolation of the bypass line is achieved by a motor operated valve. Currently, the worst case single failure for the UHS System is the failure of a bypass line motor operated isolation valve to close when required. In the event of a failure of the bypass line's motor operated valve to 6-12 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -close, the RHRSW and ESW system heat loads on the affected loop cannot be adequately cooled due to the large amount of water left unsprayed (i.e., heat addition to the UHS) via the bypass line.In this condition, operator action is currently required to transfer the major heat loads to the unaffected UHS spray division to complete the plant shutdown. To improve system reliability and performance, an additional manually operated valve is being installed in each spray array bypass line. This valve provides redundant capability to isolate the bypass line in the event of a failure of the motor operated bypass line isolation valve to close and establish spray cooling operations on the affected loop. Analyses have been performed that demonstrate that the UHS design temperature (i.e., water supplied to the RHRSW and ESW systems) is not exceeded in the event of a dual unit shutdown as a result of a LOCA/LOOP coincident with a single failure. In the event of a failure of the motor operated bypass valve to close, the analysis credits operator action to manually isolate the affected line within 3 hours of the event.The failure of a large spray array motor operated isolation valve to open was also evaluated. For this case, heated water is transferred to the UHS unsprayed through the bypass line of the affected spray division for 3 hours. After 3 hours, operators are required to reduce RHRSW and ESW flow to the affected loop, establish spray cooling through the small spray array in that loop and close the associated bypass valve. The analysis for this scenario demonstrates that the UHS design temperature is not exceeded. A new Technical Specification Surveillance Requirement is proposed to be added to TS 3.7.1, RHRSW and UHS, to periodically verify the operability of the motor operated small spray array isolation valves. A new Technical Specification Surveillance Requirement is proposed to be added to TS 3.7.1, RHRSW and UHS, to periodically verify operability of the new manual isolation valve.The UHS has also been analyzed assuming the failure of an entire division of sprays. This analysis demonstrates the UHS is capable of maintaining a maximum design temperature of 97°F following a design basis LOCA in one unit and concurrent safe shutdown of the other unit utilizing a single division of sprays. For CPPU, the UHS analysis decay heat inputs have been updated from ASB 9-2 (BTP in SRP 9.2.5) to the ANSI/ANS-5.1-1979 Standard.The UHS analyses are performed using computer models which determine spray cooling thermal performance, drift loss and system response. These models were not modified for CPPU. The UHS analysis is supported by testing results that demonstrate that thermal performance methodologies are conservative. This testing was performed as part of the original plant licensing phase. The computer models, methodology and testing are described in FSAR 9.2.7.6.5 STANDBY LIQUID CONTROL SYSTEM PPL has requested approval of license amendments to the Standby Liquid Control (SLC) System Technical Specifications in Reference

51. The proposed changes modify the SLC System to operate with one pump, versus the current two pump operation, and would also employ the use of an enriched boron solution.

The CPPU SLC System discussion presented below assumes this SLC System operating configuration. The Standby Liquid Control System (SLCS) is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that some or all of the control rods cannot be inserted. This manually operated system pumps an enriched sodium pentaborate solution into the vessel, to provide neutron absorption and achieve a subcritical reactor 6-13 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -condition. SLCS is designed to inject over a wide range of reactor operating pressures. The following topics are addressed in this evaluation: Topic ,CLTRDispositi SSES Result Core shutdown margin System performance and hardware Suppression pool temperature following limiting ATWS events The ability of the SLCS boron solution to achieve and maintain safe shutdown is not a direct function of core thermal power. SLCS shutdown capability (in terms of the required reactor boron concentration) is reevaluated for each fuel reload. Reload core design analyses are performed on a cycle-specific basis to ensure that required reactivity margins are maintained. Current technical specification requirements for cold shutdown margin are maintained at CPPU conditions with ATRIUM-10 fuel. The ATRIUM-10 fuel product line designs are used for CPPU. The total weight of natural boron required for cold shutdown does not increase for CPPU. The equivalent of 660 ppm of natural boron does not change for CPPU.The CPPU ATWS analysis verifies that the SLCS boron injection rate will produce acceptable suppression pool temperature results. For the CPPU analysis, the SLCS neutron absorber solution enrichment of 88 atom-percent and the minimum solution concentration Of 7 weight-percent are used. In addition, a single pump SLCS flow rate of 40 gpm is used. With these assumptions and inputs, an acceptable suppression pool temperature response is calculated. The suppression pool temperature is evaluated in section 9.3.1.The SLCS continues to meet the requirements contained in 10 CFR 50.62(c)(4) for SLCS injection capability for ATWS events. The combination of the neutron absorber boron enrichment of 88 atom-percent, minimum solution concentration of 7 weight-percent, and minimum SLCS pump flow rate of 40 gpm exceeds the equivalency in control capacity to 86 gallons per minute of 13 weight-percent sodium pentaborate solution for a 251-inch inside diameter reactor vessel contained in 10 CFR 50.62(c)(4). With the minimum required concentration of sodium pentaborate versus volume, as required in Technical Specification 3.1.7, it is satisfactorily demonstrated that the SLCS neutron absorber solution contains sufficient boron to shutdown the reactor to cold shutdown conditions for CPPU, including the 25% margin for leakage and imperfect mixing.6-14 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Based on the results of the plant specific ATWS analysis, the maximum reactor lower plenum pressure following the limiting ATWS event reaches 1220 psia during the time the SLCS is analyzed to be in operation. With single pump operation, the CPPU SLCS pump discharge pressure is conservatively calculated to be 1250 psig. Therefore, the SLCS pump discharge relief valve margin is 250 psig. This value is above the minimum value needed to assure that the relief valves remain closed during SLCS injection. Therefore, there is adequate margin to prevent the SLCS relief valve from lifting during SLCS operation to meet the guidelines published in NRC Information Notice 2001-13. In the event that the SLCS is initiated before the time that the reactor pressure recovers from the first transient peak, resulting in opening of the SLCS pump relief valves, the reactor pressure must reduce sufficiently to ensure SLCS pump relief valve closure. The evaluation compared the, reactor vessel lower plenum pressure prior to the time when rated SLCS injection is assumed in the ATWS analysis against the SLCS pump relief valve reseat pressure. Consideration was also given to system flow head losses for full injection, elevation head differences, and cyclic pressure pulsations due to the positive displacement pump operation. Analysis results indicate that the reactor pressure reduces sufficiently from the first transient peak to allow the SLCS pump relief valves to close.The SLCS ATWS performance is evaluated in Section 9.3.1 for an equilibrium core ATRIUM-10 design for a 24-month CPPU cycle. The evaluation shows that CPPU has no adverse effect on the ability of the SLCS to mitigate an ATWS.6.6 POWER DEPENDENT HVAC The heating, ventilation and air conditioning (HVAC) systems consist mainly of heating, cooling, supply, exhaust and recirculation units in the primary containment, control structure (including the control room), reactor building (including spent fuel pool area and the ECCS pump rooms), turbine building, radwaste building, diesel generator buildings, ESSW pumphouse, and circulating water pumphouse. Also included are the associated chilled water systems which are the control structure chilled water, reactor building chilled water, radwaste building chilled water and turbine building chilled water systems. CPPU results in slightly higher process temperatures and a small increase in the heat load due -to higher electrical currents in some motors and cables. The topics addressed in the evaluation areas follows: Topic CLTR Disposition SSES Result'Power Dependent HVAC Performance The affected areas are the primary containment; steam tunnel in the reactor building; and the FW heater bay, condenser areas, condensate pump room and the steam driven FW pump areas in the turbine building. The HVAC systems which provide cooling to these areas and their associated chilled water systems are affected. Other areas of the reactor building and turbine building are unaffected by the CPPU because the process temperatures remain relatively constant. The control structure building, radwaste building, diesel generator buildings, ESSW pumphouse, and circulating water pumphouse are unaffected by CPPU. Because the temperature 6-15 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -of the areas served by the control structure chilled water and radwaste building chilled water systems are unaffected by CPPU, these systems are not impacted by CPPU.During normal plant operation, the increased heat load due to CPPU results in less than 0.5°F increase in the area temperature for the primary containment, and less than 1.0°F for the reactor building main steam tunnels. The design of these HVAC systems is adequate to handle the increase in heat load. CPPU results in an increase of less than 20 tons (refrigeration) in the cooling requirement for reactor building chilled water system for the recirculation pump motor coolers and this system is adequately designed to handle this increase in heat* load. The increased heat load due to CPPU during normal plant operation results in less than 3°F increase in the condenser bay, condensate pump room, FW heater bay, and the FW pump areas of the turbine building. The design of the HVAC systems serving these areas is adequate to handle the increases in heat load. Further, these changes result in an increase of less than 15 tons (refrigeration) for the turbine building chilled water system and this system is adequately designed to handle this increase.The primary containment cooling system is designed to provide containment mixing during LOCA and this function is unaffected by CPPU. The main steam tunnel cooling systems, the turbine building HVAC systems, reactor building chilled water and turbine building chilled water systems are non-safety systems. They are not credited during post accident conditions. Based on a review of the design basis calculations and the evaluations performed, the design of the HVAC systems is adequate for CPPU.6.7 FIRE PROTECTION This section addresses the effect of CPPU on the fire protection program, fire suppression and detection systems, and reactor and containment system responses to postulated 10 CFR 50 Appendix R fire events. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result Fire suppression and detection systems'Operator response time Peak cladding temperature Vessel water level Suppression pool temperature [[]] As part of the plant modification process, any changes in physical plant configuration or combustible loading as a result of modifications to implement the CPPU are being evaluated in accordance with the plant modification and fire protection programs. The Appendix R Safe Shutdown Analysis does not rely on any repair activities to 6-16 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -achieve or maintain safe shutdown. The CPPU does not affect any administrative controls, fire brigade training, or fire protection responsibilities of plant personnel. The reactor and containment response, to the postulated 10 CFR 50 Appendix R fire event at CPPU conditions is evaluated in Section 6.7.1. The results show that the peak fuel cladding temperature, reactor pressure, and containment pressures and temperatures are below the acceptance limits and demonstrate that there is sufficient time available for the operators to perform the necessary actions in accordance with plant procedures to achieve and maintain cold shutdown conditions. The CPPU results in an increase in decay heat and, consequently, an increased heat addition to the Suppression Pool. Additional Suppression Pool cooling capability is required in order to maintain the suppression pool temperature within limits. This capability, is planned by protecting RHR iC/lD/2C/2D pumps that are not currently protected from the effects of a fire.Providing protection for these RHR pumps from the effects of a fire ensures their availability to provide additional suppression pool cooling using plant operating procedures. For limited fire scenarios, manual operator actions, similar to established existing operator actions, are necessary to utilize the additional Suppression Pool cooling capability. The additional cooling capability provided by protecting the RHR pumps results in the Suppression Pool temperature remaining within limits at CPPU conditions. Based on the above evaluation and the evaluation in Section 6.7.1, the fire protection program, fire suppression and detection systems, reactor and containment systems response to postulated 10 CFR 50 Appendix R fire events are acceptable at CPPU.6.7.1 10 CFR 50 Appendix R Fire Event SSES utilizes three paths for shutdown in the event of a fire. Paths I (Division 1 Redundant Safe Shutdown Path) and 3 (Division 2 Redundant Safe Shutdown Path) allow the operators to take action immediately using available equipment from the Control Room. Path 2 (Alternative Shutdown Path for a Control Room fire) has a 20 minute delay due to the transfer of control to the RSP.Path 2 is the most limiting because of (1) the limited amount of shutdown equipment at the RSP and (2) the time delay in transferring control of the reactor due to transit time and activation of the RSP.The RSP has three SRVs available to depressurize the vessel and the available RHR pump can be operated in the LPCI, SPC and ASDC mode. Twenty minutes (20) has been demonstrated by operators to be sufficient time to relocate to the RSP and begin injecting into the vessel using RCIC.A plant-specific evaluation was performed to demonstrate safe shutdown capability in compliance with the requirements of 10 CFR 50 Appendix R assuming CPPU conditions. A limiting Appendix R fire event, which bounds the current analysis, was analyzed for CPPU.The fuel heat up analysis was performed using the RELAX, HUXY, and RODEX2 analysis 6-17 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -models. The containment analysis was performed using MAAP. These evaluations determined the effect of CPPU on fuel cladding integrity, reactor vessel integrity, and containment integrity as a result of the fire event.This plant specific analysis for CPPU bounds all safe shutdown paths using the minimum set of equipment and the time delay associated with Path 2. The analysis was performed with the following assumptions:

• One SRV opens at the start of the event due to spurious operation as a result of the fire and remains open for the entire event.* RCIC injection to the RPV, both automatically and manually, is assumed to fail to make this a bounding analysis.A Loss of Offsite Power (LOOP) occurs at the start of the event.These assumptions provide a bounding value for PCT. Specifically, this analysis provides the longest delay in providing cooling to the reactor resulting in highest PCT. Two cases were analyzed for CPPU.Case 1 :Reactor is operating at CPPU conditions.

At t = 0, the reactor is scrammed with no makeup from RCIC or HPCI. The operator initiates vessel depressurization using two SRVs at 25 minutes into the event. One LPCI pump is operated to reflood the vessel.Case 2:Reactor is operating at CPPU conditions. At t = 0, the reactor is scrammed with no makeup from RCIC or HPCI. The operator initiates vessel depressurization using three SRVs at 25 minutes into the event. One LPCI pump is operated to reflood the vessel.The bounding PCT case for SSES is Case 1. The indicated water level will drop below TAF approximately 21 minutes into the transient. Water level will be restored above TAF approximately 35 minutes into transient. Normal water level will be restored approximately 41 minutes into the event. Case I was also analyzed at CLTP conditions to provide a comparison of CPPU results.The results of the Appendix R evaluation for CPPU provided in Table 6-6 demonstrate that the fuel cladding integrity, reactor vessel and piping integrity and containment integrity are maintained and that sufficient time is available for the operator to perform the necessary actions.The Net Positive Suction Head for systems using the Suppression Pool as a water source are adequate. (See Section 4.2.6 for a more detailed discussion.) Because the CPPU PCT is well below the allowable limit in the worst case scenario, the increase in decay heat will not contribute any increase in radiological release due to a fire. The current deviations for the short term core uncovery during depressurization remains necessary for CPPU. CPPU does not affect any other deviations described in Reference

27. The aforementioned RHR pump protection modification allows the operators increased flexibility to initiate suppression pool cooling operations early in the event on both units. CPPU does not affect systems required to achieve and maintain safe shutdown conditions from either the main control room or the remote shutdown panel.6-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -6.8 OTHER SYSTEMS AFFECTED BY POWER UPRATE This section addresses the effect of CPPU on systems report. The topics addressed in this evaluation are: not addressed in other sections of this Topic 1: CLTR Disposition SSES Result Other systems Based on experience and previous NRC reviews, all systems that are significantly affected by CPPU are addressed in this report. Other systems not addressed by this report are not significantly affected by CPPU.[[6-19 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-1 CPPU Electrical Characteristics Parameter.

CPPU Generator Output (MWe) 1300 Rated Voltage (kV) 24 Power Factor 0.94 Guaranteed Generator Output (MVA) 1354*Current Output (kA) 32.572 Isolated Phase Bus Duct Rating (kA) 35 Main Transformers Rating (MVA) Unit 1 1500 CPPU Transformer Output (MVA) Unit 1 1303 Main Transformers Rating (MVA) Unit 2 1350*CPPU Transformer Output (MVA) Unit 1 1306* These ratings reflect the upgraded equipment. Table 6-2 Offsite Electric Power System Component Rating CPPU Output, Generator (MVA) 1354 1354 Isolated Phase Bus Duct (kA) 35 34.29 Main Transformers (MVA) Unit 1 1500 1312 Main Transformers (MVA) Unit 2 1350 1310 Auxiliary Transformer (MVA) Unit 1 61.6 42 Auxiliary Transformer (MVA) Unit 2 61.6 44 Switchyard (limiting) (MVA) Unit 1 1593.5 1593.5 Switchyard (limiting) (MVA) Unit 2 2814.6 2814.6* These ratings reflect the upgraded equipment (after generator rewind, Unit 2 transformer upgrade and 230kV breaker replacement). 6-20 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6.3 UHS Performance Parameter Comparison Parameter, Limit CLTP CPPU'Worst Case Peak UHS 97 97 96 Temperature ('F)Maximum Water Loss Case -30 N/A 669 >6692 day Water Elevation (fl)Notes: 1.2.CPPU values reflect effects of modifications. Post-accident water elevation controlled via operating procedure guidance on use of sprays and/or spray bypass cooling.6-21 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-4 Spent Fuel Pool Parameters for CPPU Batch Offload with Cross-Tied Pools Plus 8.5 MBTU/Hr Background Heat* * ~~~~~Allowed " .. ..".Allomed. Heat Maximum Time to Boil SFP- -~Cooling,-.o Document or .......... Drn_ ... .... _ .. .Postulated Operating Heat Removal SFP Water after loss or.eros ... heat .. ... Excangers. .~.....w e mp....... .Water. Temp.Commitment -(or) (%BUh).T-Tm" FPCCS (hours) (MBTU/hr) T~emp (",Caaiy ep___________ 0 F) ___________ (MBTUI/ir) '(OF) (hrs), SSES FSAR 144 28.8 No RHR 6 FPCCS HX 125 3@9534.24 118.6 28.94+3@75 I FPCCS SSES FSAR 144 28.8 HX + No 5 FPCCS HX 125 2@95 29.96 123.4 27.42 RHR +'3@75 Assist 5 FPCCS HX+ One 3@95 N/A due to SSSjI FPCCS +2One 69.6 34.6 RHRHX in 115 +2@75 45.77 108.3 RHRSDC ADMIN HX RHRSDC RHR @ in-service 88 Mode 5 FPCCS HX SSES I FPCCS + One 5@95 N/A due to ADMIN 69.6' 34.6 HX RHRHX in 115 RHR @ 40.1 111.7 RHRSDC RHRSDC 88 in-service Mode Notes: 1 2 3 Batch offload = 348 assemblies, 7.5 assemblies moved per hour = 46.4 hrs + 24 hrs to begin fuel movement after reactor shutdown = 69.6 hrs One RHRHX is available when pools are cross-tied (either in shutdown cooling with the reactor cavity flooded or in the FPC Cooling Mode).Time to boil is based on the maximum calculated SFP temperature. This is conservative because no credit is taken for heat losses to the environment through evaporation or surrounding structures. Therefore, actual temperatures would be lower and time to boil would be longer. If boiling where to occur, the 60 gpm needed for make-up is within the ESW system available capacity.6-22 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-5 Spent Fuel Pool Parameters for CPPU Full-Core Off-Load (EHL) with Cross-Tied Pools Plus 8.5 MBTU/hr Background Heat Assumed nAllowed eat : .Maxmum Time tOsBO Document o'r ...Decay To.a I: .Postulated Operating Heat SF Commitmeit.. Periods.' Dec- ht , Failure Exchangers Water Co.igi Rmpacityl chngr Water Temp. Cp ae fe oso---(hours) (MBTU/hr) t ~ ~ Temp. FCC 6 FPCCS HX + 3@95°F SSES FSAR 250' 62.32 1 RHRHX I RHRHX in 125 +3@75 0 F 70.6 120.5 N/A RHRSDC @_ _@88OF..1 EHL = Full Core offload 764 assemblies / 7.5 assemblies /hr = 102 hrs + 24 hrs prep time = 126 hours. The fuel movement to the SFP is delayed an additional 124 hours to result in a full core offload to the SFP in 250 hours (10.4 days) after reactor shutdown.2 Decay heat from a partial core at CPPU, 16.4 days after reactor shutdown, is projected to be 14.7 MBTU/hr. Decay heat from a full core at CPPU, 10.4 days after reactor shutdown, is projected to be 39.1 MBTU/hr. The background heat from spent fuel resident in both pools is approximately 8.5 MBTU/hr. Total EHL in the cross-tied SFPs = 14.7 + 39.1 + 8.5 = 62.3 MBTU/hr.3 One RHRHX will always be available when pools are cross-tied (either in shutdown cooling with the reactor cavity flooded or in the FPC Cooling Mode).6-23 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-6 Appendix R Fire Event Evaluation Results Parameter CLTP CPPU " 'App. R111ý.Criteria Cladding Heat up (PCT) (°F) 904 ' 1191 < 1500 6 Primary System Pressure (psig)" 1050 1050 < 1375 2 Primary Containment Pressure (psig) 4.3 5.1 < 53 5 Suppression Pool Bulk Temperature ('F) 4 181 191 <230 3.Notes: 1. PCT data from Framatome CPPU analysis at CLTP conditions.

2. SSES maximum allowable reactor vessel pressure.3. The bulk temperature of the Suppression Pool cannot exceed the operational range for SPOTMOS. NRC IN 84-09, Lessons Learned from NRC Inspection of Fire Protection Safe Shutdown Systems, requires the availability of Suppression Pool temperature monitoring instruments for BWRs.4. Worst case suppression pool temperature with additional RHR pumping capability.

5. Containment structural design limit..6. This value satisfies the Appendix R Performance Goal of no fuel cladding damage for an extended duration of core uncovery (NUREG-0562 and EPGs/SAGs).

Based on NUREG-0562, higher values may be justified for limited duration core uncovery.6-24 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect rr S 6-25 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect 0 6-26 Safety Analysis Report for Susquehanna Units I and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect 6-27 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect o 0 6-28 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect-0 6-29 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect 0 6-30 Safety Analysis Report for Susquehanna Units I and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect 6-31 Safety Analysis Report for Susquehanna Units I and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect 6-32 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect 6-33 Safety Analysis Report for Susquehanna Units 1 and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 6-7 Basis for Classification of No Significant Effect The SSES convention for unit-specific systems begins with a "1" for Unit 1 and "2" for Unit 2. Additionally, unit-common systems begin with a "0." However, unit-specific and unit-common designators are not identified for simplicity. 6-34 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -This page intentionally left blank.6-35 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -7. POWER CONVERSION SYSTEMS This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapter 10 that applies to CPPU.7.1 TURBINE-GENERATOR The turbine-generator converts the thermal energy in the steam into electrical energy. The topics addressed in this evaluation are: Topic .!CLTR:Disposition SSES Result Turbine-generator performance Turbine-generator missile avoidance ]The turbine and generator was originally designed with a maximum flow-passing capability and generator output in excess of rated conditions to ensure that the original rated steam-passing capability and generator output is achieved. This excess design capacity ensures that the turbine and generator meet rated conditions for continuous operating capability with allowances for variations in flow coefficients from expected values, manufacturing tolerances, and other variables that may adversely affect the flow-passing capability of the units. The difference in the steam-passing capability between the design condition and the rated condition is called the flow margin.The turbine-generator was originally designed with a flow margin of 5%; however, for operation prior to CPPU the turbine was redesigned and currently operates with a flow margin of approximately 3%. The 3% flow margin will be retained for CPPU and revisions will be made to the Electro Hydraulic Control system to support the increased steam flow. The current rated throttle steam flow is 14.39 Mlbm/hr at a throttle pressure of 997 psia. The generator is currently rated at 1298 MVA at a power factor of 0.90.At the CPPU RTP and a reactor dome pressure of 1050 psia, the turbine operates at an increased rated throttle steam flow of 16.49 Mlbm/hr and a throttle pressure of 976.1 psia. To maintain control capability GE uses a minimum target value of approximately 3% throttle flow ratio, with controllability confirmed by unit testing as described in Section 10.4. For operation at CPPU, the high pressure turbine will be modified to include a design with a new inner cylinder, two new blade carriers, a new rotor, and new blades for at least the minimum target throttle flow margin, to increase its flow passing capability. PPL Susquehanna will be installing a new liquid cooled stator winding, a core end ventilation baffle and hydrogen coolers that have been designed specifically for a 1354 MVA rating. The stator core can handle the new rating. The generator rotor torsional stress was evaluated for fatigue life expenditure, which indicates performing an NDE inspection of the coupling is 7-1 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -appropriate. Prior to CPPU the isolated phase duct cooling system capacity was increased over the original design and will support CPPU heat loads.The expected environmental changes such as diurnal heating and cooling effects changing cycle efficiency will periodically require management of reactor power to remain within the generator rating. The required variations in reactor power do not approach the magnitude of changes periodically required for surveillance testing and rod pattern alignments and other occasional events requiring de-rating, such as equipment out of service for maintenance. The turbine gland sealing system is not being modified for the CPPU. For CLTP the. total sealing steam flow was reduced from the OLTP design. For CPPU the total sealing steam flow is further reduced. The evaluation of the turbine gland seal system, taking into account the modification of the main turbine to accept the increased steam flow at CPPU operating conditions, demonstrated that the system is capable of adequately performing its design function without modification. No increase in capacity or changes in any control settings are required for the CPPU.Prior to CPPU, in 2003 for Unit 2 and 2004 for Unit 1, the main turbine internals were changed from a GE monoblock design for both the HP and LP rotors to a Siemens monoblock for the HP rotor and to a Siemens shrunk on wheel design for the LP rotors. Per the CLTR, Section 7.1, a separate rotor missile analysis is not required for plants with integral wheels. At the time of the turbine modification, the turbine missile licensing basis was changed to the CLTP turbine missile licensing basis which is the methodology specified in NUREG-1048 Supplement 6 Appendix U dated July 1986 (Reference 53).The CLTP turbine missile licensing basis uses: P1 -The probability that a main turbine missile will be generated P2 -The probability that a missile will strike a barrier that houses a critical plant component P3 -The probability that a missile will breach the barrier and damage a critical plant component Appendix U states, in part, "In view of operating experience and NRC safety objectives, the NRC staff has shifted emphasis in the reviews of the turbine missile issue from the strike and damage probability (P2 x P3) to the missile generation probability (P 1)." For CPPU the existing 12 stage HP monoblock rotor is being replaced by an 11 stage HP monoblock rotor while the existing LP rotors are being retained. The missile analysis for this replacement is supported by the Siemens Technical Report CT-27332, Revision 2 which was approved by the NRC (Reference 28). This methodology is the same as the CLTP turbine missile licensing basis with slight revision only.Reference 28 refers to a prior Safety Evaluation (Reference 54), where confirmation of certain input variables is required for plant specific applications. PPL has confirmed that the eight parameters listed in section 4.0 of Reference 54 are the same as used in the SSES specific 7-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -missile analysis. PPL has also confirmed, by reviewing material certificates for the six LP rotors and discs, that the plant specific parameters listed in section 3.2.2 (of Reference

54) are within in design range of these parameters.

For both the CLTP and CPPU analysis the turbine vendor determined the P1 values using the NRC approved methodology. The NRC specified value of 1E-02 for an unfavorably oriented turbine is used for P2xP3. Results of the revised missile analysis indicate that the missile probabilities for P 1 are virtually unchanged from CLTP to CPPU and remain below the NRC specified limit of IE-05 per year for an unfavorably oriented unit. The CPPU analysis is based on up to 100,000 operating hours (approximately 12 years) between disc inspections. Since the CLTP inspection frequency of 10 years is not being changed, the actual probabilities are less.The current licensing basis for the Reactor Feed Pump Turbines (RFPT) does not include a missile analysis. Further, the RFPT's both have integral turbine wheels and have a favorable orientation (i.e., the axis is perpendicular to the reactor). Therefore no missile analysis is required for the RFPT's.For CPPU the main turbine over speed calculation evaluates the entrapped steam energy contained within the turbine and the associated piping, after the stop valves close, against the inertia of the rotor train and the capability of over speeding. The entrapped energy has been increased slightly for the CPPU conditions. The main turbine has previously been replaced with a more massive design which reduced the acceleration rate of the turbine by 35% from with the original design. The design and vendor tested overspeed is 125% of rated speed. The 125% is conservative since the turbine missile analysis uses 120% as the design overspeed. For CPPU, the acceleration rate of the turbine increases slightly over the acceleration rate of the current turbine, but is reduced by 30% less than the original turbine design. The current overspeed trip settings are not changed. As such, the corresponding over speed will also be less and the existing protection will provide increased margin when compared to the original design. The primary speed control is the turbine EHC control system which sends a signal to close the turbine control valves when the speed exceeds 100%. The normal backup is a mechanical overspeed trip with a setting of 110%. The emergency backup is an electrical overspeed trip with a setting of 112%. For CLTP the overspeed trip system is tested in accordance with Technical Requirements Manual section 3.3.7 "Main Turbine Overspeed Protection System" and this testing will continue with CPPU. Overspeed testing involves cycling the turbine control valves, cycling the turbine combined intercept valves (CIV), cycling the turbine stop valves, a channel calibration of the instrumentation, and periodic disassembly of the control, CIV and stop valves.7.2 CONDENSER AND STEAM JET AIR EJECTORS The condenser converts the steam discharged from the turbine to water to provide a source for the Condensate and FW systems. The SJAEs remove non-condensable gases from the condenser to improve thermal performance. The topics addressed in this evaluation are: 7-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -., :.Topic .CLTR'Disposition SSES Result , Condenser and SJAE The condenser and SJAE functions are required for normal plant operation and are not safety-related.The condensers were evaluated for performance at CPPU conditions based on a maximum cold water temperature of 92.8°F and current circulating water system flow. Additional analysis at CPPU conditions also determined the condenser back pressure would be below, the 5.5" Hga design limit assuming cleanliness levels of 85% along with condenser tube plugging of 1.0% at wet bulb temperature and relative humidity < 78.0°F and 65%.Condenser hotwell capacities and level instrumentation are adequate for CPPU conditions. The evaluation considered heater drain and extraction steam holdup times in the condenser hotwell.The condenser hotwell inventory is adequate to provide more than a 2-minute holdup time for CPPU flow conditions. Periodic eddy current testing and water chemistry monitoring are performed which provide monitoring of the effect of CPPU RTP operation on the condenser tubes. The design of the condenser air removal system is not adversely affected by CPPU and no modification to the system is required.The following aspects of the condenser air removal system were evaluated for this determination:

• Non-condensable gas flow capacity of the SJAE system;* Capability of the SJAEs to operate satisfactorily with available dilution / motive steam flow;SJAEs and inter-condensers' performance at the higher expected non-condensable flow and condenser pressure conditions for CPPU, considering water vapor carryover and the maximum expected condensate temperature and flow rate; and Mechanical vacuum (hogging) pump (low capacity, less than '5%) capability to remove required non-condensable gases from the condenser at CPPU conditions.

The physical size of the primary condenser and evacuation time are the main factors in establishing the capabilities of the vacuum pumps. These parameters do not change. Because flow rates do not change, there is no change to the holdup time in the pump discharge line routed to the turbine building vent stack. The capacity of the SJAEs is adequate because they were originally designed for operation at flows greater than those required at CPPU conditions. 7-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -7.3 TURBINE STEAM BYPASS The Turbine Steam Bypass System provides a means of accommodating excess steam generated during normal plant maneuvers and transients. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result:'Turbine steam bypass (normal [[operation) Turbine steam bypass (safety analysis)1]][[The turbine bypass valves were initially rated for a total steam flow capacity of not less than 25% of the original rated reactor steam flow, or -3.54 Mlbm/hr. Each of five bypass valves is designed to pass a steam flow of -733,400 Ibm / hr and does not change at CPPU RTP. At CPPU conditions, rated reactor steam flow is 16.532 Mlbm/hr, resulting in a bypass capacity of 22.18% of CPPU rated steam flow. The bypass capacity at SSES remains adequate for normal operational flexibility at CPPU RTP.The Turbine Steam Bypass System is a normal operating system that is used to bypass excessive steam flow. The absolute flow capacity (mass flow rate) of the bypass system is unchanged. The bypass flow capacity is included in some AOO evaluations (Section 9.1). These evaluations demonstrate the adequacy of the bypass system. Cycle specific reload analysis takes credit for the availability of the bypass system in the reload analysis to establish the core operating limits.7.4 FEEDWATER AND CONDENSATE SYSTEMS The FW and Condensate Systems provide the source of makeup water to the reactor to support normal plant operation. The topics addressed in this evaluation are: Topic. CLTR Disposition SSES Result.FW and Condensate Systems The FW and Condensate Systems do not perform a system level safety-related function,. and are designed to provide a reliable supply of FW at the temperature, pressure, quality, and flow rate as required by the reactor. However, their performance has a major effect on plant availability and capability to operate at the CPPU conditions. For CPPU, the FW and Condensate Systems will meet their performance requirements with the following modifications to non-safety related equipment: 7-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -For uprate to the intermediate power level of 107% of CLTP the following modifications are planned: 1. Replace the condensate pump impellers with higher head pump impellers 2. Modify the condensate minimum flow valves to support condensate pump. higher minimum flow requirements.

3. Modify FW heaters #1, #2, and #4 as necessary to support higher flows and heat loads.4. Replace FW heater #3 Emergency Dump Valves with higher capacity valves.5. Re-rate FW heater #5 shells to a higher pressure.6. Upgrade FW heater #5 safety relief valves to support a higher setpoint.7. Replace reactor feed pump suction flanges with piping to support a higher condensate pressure.8. Modify crossaround relief valve internals to accommodate a higher relieving pressure.9. Remove the confirmatory low reactor water level trip signals from the initiation logic that initiates a runback of the reactor recirculation system upon detection of a trip of a FW pump and/or condensate pump. (Current logic requires both a pump trip signal AND a confirmatory reactor low water level signal to initiate the runback. After the logic modification, only the pump trip signal will be required to initiate the runback).Also changes condensate pump trip input signal to pump breaker position and increase recirculation system runback rate.10. Modify the Reactor Feed Pump seal water injection system to increase the setpoint of the temperature controller due to a higher seal water temperature and raise the elevation of the drain line vent piping.11. Modify the FW suction pressure trip setpoint and time delay to accommodate CPPU operating conditions.

For Uprate to the final CPPU power level, the following modification are planned: 12. Modify the FW pump turbines to accommodate'the Reactor Feed Pump's higher loadings and rotating speeds.13. Upgrade RFPT turbine speed controls and overspeed trip to digital controls.7.4.1 Normal Operation System operating flows at CPPU increase to approximately 114% of rated flow at the CLTP.The FW and Condensate Systems will be modified to assure acceptable performance with the new system operating conditions. An evaluation of the FW heaters shows that FW heaters #1, #2 and #4 need to be modified in order to be acceptable for the higher FW heater flows, temperatures, and pressures for the CPPU. The evaluation identified cascading drain flow flashing concems in FW heater #i and tube vibration concerns for increased drain flows in FW heaters #2 and #4. FW heater #3 was scheduled for replacement prior to CPPU while FW heater #5 will need to be recertified prior to implementation of CPPU in order to accommodate a higher CPPU extraction pressure. The performance of the FW heaters will be monitored during the CPPU power ascension program.7-6 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -During the operating cycle after the first ; 7% up-rate cycle, the reactor feed pump turbines will experience higher loadings, but the higher loading is within the existing capability of this equipment. For the full 14% up-rate, the reactor feed pumpturbines will be modified for the higher loadings and rotating speeds needed to drive the reactor feed pumps (RFPs). Prior to the first uprate the Reactor Feed Pump seal water injection system will be modified to ensure proper operation at CPPU condensate and feedwater pressures and temperatures..

7.4.2 Transient

Operation To assess FW transients, the FW system was evaluated and determined to have an 11.8% margin at CPPU FW flow. For system operation with all system pumps available, the predicted operating parameters were acceptable and within the component capabilities. The system capacity was evaluated with the modifications and changes listed above for the FW and Condensate Systems to assess the capability to supply the transient flow requirements. Analysis of the single FW pump trip for CPPU conditions indicates that minimum reactor vessel water level will approach the low reactor water level SCRAM setpoint. When instrument accuracies and modeling uncertainties are applied to the analysis results, the ability to avoid a SCRAM upon a single FW pump trip cannot be assured. Section 10.4.7 of the FSAR currently indicates that a single FW pump trip will not result in a reactor SCRAM. Section 10.4.7 of the FSAR is being changed to reflect the results under CPPU condition. The analysis does demonstrate that adequate FW flow is provided to the reactor vessel post SCRAM and that the loss of a single FW pump is bounded by the loss of total FW flow transient. Analysis of a single condensate pump trip at CPPU conditions was also performed. Upon a trip of a single condensate pump the RFP trip system receives an initiation signal resulting from RFP low suction pressure. At the time of the condensate pump trip, a reactor recirculation system runback to the #2 limiter (48% recirculation pump speed) is initiated and reactor power begins to decrease. The low suction pressure trip has varying time delays for the three RFPs. Should one RFP trip before reactor water level is recovered, FW flow will decrease and restore RFP suction pressure to clear the trip signal to the other two RFPs. The recirculation runback will reduce reactor power level to a point where the remaining condensate and FW pumps will provide adequate flow to the reactor vessel to maintain water level, or the reactor will scram on low water level. Section 10.4.7 of the FSAR currently states that a single condensate pump trip will not result in a reactor SCRAM. Section 10.4.7 of the FSAR will be revised to reflect the results under CPPU conditions. The analysis does demonstrate, as stated above, that adequate FW flow is provided to the reactor vessel post SCRAM and that the loss of a single condensate pump is bounded by the loss of total FW flow transient.

7.4.3 Condensate

Demineralizers The effect of CPPU on the condensate filters and demineralizers was reviewed. The systems will be modified by adding a condensate filter and a condensate demineralizer to accommodate the full uprate. At CPPU conditions, the additions will result in equivalent vessel hydraulic flux design comparable to current CLTP values.7-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -During operation at the first plateau (approximately 107% of CLTP), the existing filters, and demineralizers, which will not yet have been modified, will experience higher loadings, resulting in reduced run times and potential reduction of effluent quality.The chemistry guidelines/commitments to EPRI, INPO, and BWRVIP, required to protect reactor vessel and fuel integrity, will continue to be met.7-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -8. RADWASTE AND RADIATION SOURCES This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapters 11 and 12 that applies to CPPU.8.1 LIQUID AND SOLID WASTE MANAGEMENT The Liquid and Solid Radwaste System collects, monitors, processes, stores and returns processed radioactive waste to the plant for reuse or for discharge. The topics considered in this section are: Topic CLTR Disposition SSES Result Coolant fission and corrosion product levels Waste Volumes The single largest source of liquid and wet solid waste is from the backwash of condensate filters and spent resin from the condensate demineralizers. Additional sources of liquid and wet solid waste are from the RWCU backwash. Increased flow at CPPU results in the need for an additional condensate filter and condensate demineralizer on each Unit to maintain the vessel hydraulic flux. Operation at CPPU conditions results in a greater backwash volume and in more solid waste (filter crud and spent resin) generated. The increase in filter backwash and waste solids generated does not affect plant safety. The RWCU filter-demineralizers will also require more frequent backwashes due to the slightly higher levels of impurities as a result of the increased FW flow, however, the increased liquid and solid waste generated are a small fraction of that generated by the condensate vessels.Liquid radwaste is collected in building sumps from a variety of sources including floor and equipment drains. The building sumps are independent and contain their own oil water separators. Power uprate does not affect system operation or equipment performance. Therefore, this subsystem is not expected to experience a significant increase in the total volume of liquid and solid waste due to operation at the uprated condition. Based on previous uprate experience, for power uprates of 5 to 20 percent, it is expected that the quantity (activity) of Activated Corrosion Products (ACP) will increase by I to 4 percent in liquid wastes and by 5 to 13 percent in processed solid wastes. The total volume of liquid and solid processed waste typically increases less than the percentage increase in feedwater flow because, while there is an increase in processed waste due to more frequent backwashes of the condensate filters, more frequent backwashes of the RWCU filter-demineralizers and more frequent spent resin waste from the condensate demineralizers, other process streams, which do not change due to CPPU, also contribute to the total annual waste quantity. A SSES CPPU evaluation indicates an increase of approximately 1% in liquid waste, an increase of approximately 11% in wet solid waste and an increase of approximately 2% in dry solid waste.The total liquid and solid increases are well within the Radwaste System capacity. Therefore, 8-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -CPPU does not have an adverse effect on the processing of liquid and solid radwaste, and there are no significant environmental effects.8.2 GASEOUS WASTE MANAGEMENT The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result Offsite release rate Recombiner performance [II The primary function of the Gaseous Waste Management (Offgas) System is to process and control the release of gaseous radioactive effluents to the site environs so that the total radiation exposure of persons in offsite areas is within the guideline values of 10 CFR 50, Appendix I.8-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The Offgas System interfaces with the condenser air removal system; gland seal exhaust and mechanical vacuum pump operation exhaust; and building ventilation system exhausts. Plant procedures exist to test for air infiltration (e.g., condenser) and repair as needed to maintain the Offgas System functional. The radiological release rate is administratively controlled to remain within existing site release rate limits, and is a function of fuel cladding performance, main condenser air inleakage, charcoal adsorber inlet dew point, and charcoal adsorber temperature. [[The administrative controls mentioned above to maintain the offgas radiological release rate below limits include power reduction or shutdown, reducing main condenser air inleakage (increasing charcoal adsorber holdup time), and, if necessary, local power suppression (inserting control rods near any fuel leaker). In addition, decreasing adsorber temperature (increasing dynamic adsorption coefficients and holdup times) can be effective in dealing with slow increases in offgas release rate. SSES has Technical Specifications requirements and administrative controls to limit fission gas releases to the environment. Plant procedures for reducing power, identifying and suppressing power near leaking fuel, and repairing condenser air inleakage exist and have been used at SSES to maintain the offgas limits. These procedures are not affected by CPPU.]] In addition, HWC operation, when used will cause a reduction in core radiolysis.. The combination of the HWC injected hydrogen plus the reduced radiolysis is expected to produce a lower net hydrogen flow to the Offgas System. The evaluation of the Offgas System and those systems connected to it concludes that sufficient capacity exists without modification. to process expected offgas for CPPU. Therefore, the gaseous radwaste system at SSES is confirmed to be consistent with GE design specifications for radiolytic flow rate [[1]8.3 RADIATION SOURCES IN THE REACTOR CORE During power operation, radiation sources in the core are directly proportional to the fission rate. These sources include radiation directly produced in the fission process, from decay of accumulated fission products and by secondary neutron reactions as a result of fission.The topics addressed in this evaluation are: 8-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Topic CLTR Disposition' SSES Result Post operational radiation sources for radiological and shielding analysis The post-power production radiation sources in the core are primarily the result of the decay of accumulated fission products. Two forms of post-production radiation source data can be used to characterize the core fission product source term. The first of these is the core gamma-ray source. The second set of post-power production source data consists of nuclide activity inventories for the accumulated fission products in the fuel.Fission product inventory data is used for post-accident dose consequence analyses and SFP evaluations. These analyses are performed compliant with regulatory guidance that applies varying release and transport assumptions to differing fission product families. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops "equilibrium" activities in the fuel (core average exposure of 39 GWD/MTU). Most radiologically significant fission product isotopes reach equilibrium within a 60-day period.Er For CPPU, fission product inventories for the uprated power level of 4032 (102% RTP) MWt were generated utilizing the SAS2H/ORIGEN-S code system. These are provided in Appendix Tables A-I and A-2 in terms of Curies at various times after shutdown for a full Susquehanna core of Fiamatome's ATRIUM-10 fuel and for an ATRIUM-10 fuel assembly, respectively. Basic input parameters utilized in the ORIGEN-S fuel burnup analyses include:* ATRIUM 10 fuel design* Enrichment < 4.5 %* Core average exposure < 39 GWd/MTU* Single assembly exposure 58 GWd/MTU* Fuel assembly weight: 181 kg U* Assembly average power < 5.28 MWt The SSES plant-specific core radiation source terms are employed in the determination of radiological dose consequences for the Loss of Coolant Accident, Fuel, and Equipment Handling Accidentand Control Rod Drop Accident discussed in Section 9.2.8.4 RADIATION SOURCES IN REACTOR COOLANT Radiation sources in the reactor coolant at SSES include activation products, activated corrosion products and fission products. The topics addressed in this evaluation are: 8-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Topilc CLTR Disposition, SEResul 8.4.1 Coolant Activation Products 8.4.2 Activated Corrosion Products and Fission Products 8.4.1 Coolant Activation Products During reactor operation, the coolant passing through the core region becomes radioactive as a result of nuclear reactions. The coolant activation, especially N-16 activity, is the dominant source in the turbine building and in the lower regions of the drywell. The activation of the water in the core region is in approximate proportion to the increase in thermal power. The activation products in the steam from the proposed CPPU are bounded by the existing design basis concentration. The margin in the SSES plant design basis for reactor coolant activation concentrations exceeds potential increases due to CPPU. Therefore, no change is required in the activation design basis reactor coolant concentrations for the CPPU.8.4.2 Activated Corrosion Products and Fission Products The reactor coolant contains activated corrosion products, which are the result of metallic materials entering the water and being activated in the reactor region. Under the CPPU conditions, the FW flow increases with power and the activation rate in the reactor region increases with power. Although some increase in the activated corrosion product production may result with CPPU, the magnitude is expected to be negligible. Calculated corrosion product activity concentrations based on an equilibrium ANS 18.1-1999 analysis are bounded by the design basis. The ANS 18.1-1999 Standard method is an equilibrium analysis for determining coolant concentration which is proportional to power, inversely proportional to total water mass, and to a lesser extent inversely proportional to steam flow. As discussed in Section 3.3.3 the levels of moisture carry over expected at CPPU steaming rates are unchanged (<0.1 weight %)and are not expected to result in any added buildup or dose rate consequence in the balance of plant.Fission products in the reactor coolant are separable into the products in the steam and the products in the reactor water. The activity in the steam consists of noble gases released from the core plus carryover activity from the reactor water. This activity is the noble gas offgas that is included in the plant design. The calculated offgas rates for CPPU after thirty minutes decay are well below the original design basis of 0.10 curies/sec. Therefore, no change is required in the design basis for offgas activity and the Technical Specification limit for offgas activity does not change for the CCPU.The fission product activity in the reactor water, like the activity in the steam, is the result of minute releases from the fuel rods. Predicted concentration of fission product activity levels in the reactor water were calculated for CPPU conditions using ANSI 18.1 methods. The resultant 8-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -concentrations are bounded by the CLTP design basis values. The Technical Specification limit for reactor water concentrations does not change for the CPPU.8.5 RADIATION LEVELS For CPPU at SSES, normal operation radiation levels increase by approximately the percentage increase in power level. For conservatism, many aspects of the plant were originally designed for higher-than-expected radiation sources. Thus, the increase in radiation levels does not affect radiation zoning or shielding in the various areas of the plant because it is offset by conservatism in the original design, source terms used and analytical techniques. The topics addressed in this evaluation are: opic CLTR Disposition SSES Result Normal operational radiation levels Post-operation radiation levels Post-accident radiation levels .]The normal operation CLTP radiation levels will increase with the implementation of CPPU in approximate proportion to the change in radiation sources as described in Sections 8.3 and 8.4.These changes in source terms are summarized as follows: Core radiation sources increase in approximate proportion to the increase in power level.CLTP source terms are based upon plant operation with fuel defects resulting in an offgas release rate of 100,000 gCi/sec after 30 minutes delay. The following coolant source term changes will result from CPPU: " Nitrogen-I 6 dose rates from main steam lines and related equipment may increase up to 20% due the combined effects of increased activation rates, (< 14%) due to the higher power level and reduced transit decay times (< 6%), from the increased steam flow rate." Non-coolant activation (corrosion) products increase in proportion to the thermal power increase.* Reactor water and fission product offgas release rate will increase approximately proportional to the power increase.An evaluation of these changes demonstrated that CPPU activity levels are bounded by the CLTP design basis concentrations. SSES radiation shielding was originally designed with the conservative radiation source terms noted above and conservative analytical techniques. An assessment of the impact of CPPU on radiation shielding was performed based upon current operating experience and projected source term changes. With CPPU, no required change to the existing plant radiation zoning was 8-6 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -identified, although localized hot spots 2 were identified in areas outside feedwater heater rooms and near drywell penetrations. Therefore, any increase in radiation levels resulting from CPPU is not expected to significantly affect the established radiation shield design and designated radiation zoning. Radiation surveys of selected areas are to be conducted as part of the CPPU startup and test plan and will identify areas that may require changes in radiation shielding or zone designation. Individual worker exposures are maintained within acceptable limits by station personnel using work management practices effectively, controlling access to radiation areas, and using the ALARA (As Low as Reasonably Achievable) program. Procedural and work package controls are used to compensate for any increased radiation levels and radiation worker training programs stress the importance of ALARA concepts. SSES stresses the need for ALARA continuous improvement and worker dose reduction. Changes to shielding and temporary/permanent plant changes will be considered as necessary to ensure compliance with regulatory limits and implementation of the ALARA concept. Current projections are that few if any plant changes will be required.Shutdown radiation levels and resultant personnel exposures are strongly dependent on the generation and distribution of activated corrosion products, (i.e. crud). This material is released into the condensate and feedwater from metal surfaces and is transported to the reactor coolant system where it is deposited, activated, and subsequently released. The radioactive crud is then transported throughout the coolant piping system. CPPU is expected to increase the shutdown radiation levels 1SSES implements various strategies to control shutdown dose rates. These include: 0 Maintaining near constant feedwater hydrogen injection rate with minimal cycling,* Use of depleted zinc oxide injection to minimize reactor water Co-60 levels,* Installation of permanent local radiation shielding.

• Use of soft shutdown or similar techniques to reduce the likelihood of crud bursts at times of unit shutdown.Other efforts will also be used to maintain doses ALARA during outage conditions.

Examples include revisions in the way work is to be performed and to use concepts such as "time, distance, and shielding" to reduce doses below those otherwise projected. Post accident radiation levels and doses at CLTP were evaluated using the TID-14844 (Reference

29) source term methodology.

This methodology assumes a DBA LOCA and an instantaneous activity release from the core of 100% noble gases, 50% of halogens and 1% other solids 2 localized higher radiation areas of limited extent 8-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Post accident radiation levels used in assessing the impact of CPPU conditions (i.e. core inventory changes) were based upon TID-14844 source term methods for assessing the impact to Environmental Qualification and to the NUREG-0737, Item II.B.2 Vital Areas not considered in the Alternate Source Term (AST) license amendment request (Reference 22). This submittal showed that the original TID source term is bounding for dose consequences and result in more limiting radiation levels. To account for CPPU changes in power level and to bracket core inventory source term changes a conservative scaling factor of 1.5 was applied to the vital area CLTP doses. Recently designated vital areas are assessed based upon consideration of CPPU contained sources.A specific evaluation of the impact of CPPU on NUREG-0737 post accident vital areas and mission doses was performed and is summarized in Table 8-1.8.6 NORMAL OPERATION OFF-SITE DOSES The primary sources of normal operation offsite doses are from liquid and gaseous radioactive effluents, gamma radiation shine from the plant turbines, and from transport and storage of radioactive materials. Of these dose pathways, CPPU will have a minor impact on the offsite doses resulting from plant effluents and radiation shine from steam turbines and associated piping as quantified below. The offsite dose from the transport and storage of radioactive materials are not affected by CPPU due to conservatisms in the original design of facilities and in the analytical techniques used to estimate dose impacts. The topics addressed are:.7 Topic CLTR Disposition SSES Result Plant emissions [[Plant skyshine from the turbine The CPPU increase in normal operation activity levels in the reactor coolant is approximately proportional to the percentage increase in core thermal power. The Technical Specifications and Offsite Dose Calculation Manual implement the ALARA guidelines of 10 CFR 50, Appendix I and govern the amount of radioactive waste effluents released to the environment. The Radioactive Effluent Control Program and Technical Requirements Manual define allowable releases and ensure that CPPU will not involve unacceptable offsite dose associated with gaseous (noble gases, airborne particulates, iodine, tritium) and liquid effluents. For CPPU, an analysis was performed to estimate offsite doses resulting from expected gaseous effluents. The following tabulation shows the doses from gaseous effluents are within 10 CFR 50 Appendix I ALARA design objectives and are a small fraction of 10 CFR 20 annual dose limits of 100 mrem.8-8 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -C :'.,cPPU Doses From Gaseous Effluent Releases R icvE~Ef a SeIOCFR50 Appendix I Expected Annual d.:. a ioactlve Eflunneusal ie : ,:... ... : ...*:::_,: :: .: ... _,. ...D esign o~jecuives. D, ..Boundary/Critical Location. Design Objectives.Dos (per Unit) (per Unit)Gamma Air Dose (mrad) _<10 3.0 Beta Air Dose (mrad) 520 5.1 Total Body of an Individual (mrem) "_5 1.9 Skin of an Individual (mrem) <15 4.6 Radioiodines and Particulates Organ Dose to Infant Thyroid (mrem) _<15 14.5 (Dairy pathway)Organ Dose to Child Thyroid (mrem) _<15 11.9 (Garden pathway)For liquid effluent releases, the maximum individual doses are estimated to be 0.34 mrem/year/site to the total body of a child and 7.7 mrem/year/site to the bone of a child. These doses comply with the Appendix I ALARA design objectives of 6 and 20 mrem/year/site to the total body and any organ, respectively, and are a small fraction of the 10 CFR 20 annual dose limit for unrestricted access of 100 mrem.With CPPU, it is estimated that radiation shine from the turbine building may increase up to.20%. This increase results from a combined effect of the power related increase in the production of N-16 activity coupled with the higher steam velocity which results in less radioactive decay in transit to the turbine. For CPPU, the highest estimated dose to a critical offsite location due to radiation shine from turbine building components is approximately 4 mrem per year.Any increase in offsite radiation exposure would be measured in the site area thermo-luminescent dosimeter stations. Implementation of Hydrogen water chemistry caused N-16 steam activity to increase by approximately 500%. Data from the environmental stations which evaluated the implementation of Hydrogen water chemistry have not shown any discemable increases in radiation at offsite locations (Reference 30). As such the potential 20% increase associated with CPPU is not expected to have any effect on offsite doses to the public.Of the total annual public dose at SSES, the major source of offsite dose results from the transport and storage of radioactive materials. CPPU will have essentially no impact on this dose contributor. Currently, radioactive materials are transported to and stored in the Low Level Radwaste Storage Facility, Interim Spent Fuel Storage Installation, Sealand Containers, Dry Active Waste Reduction System Facility, and Condensate Storage Tanks. CLTP offsite doses estimates for operation of these facilities are based on conservative material quantities and source terms that are not directly related or proportional to power level. Consequently the CLTP dose contribution to 8-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -offsite dose from the transport and storage of radioactive materials will be essentially unchanged for CPPU. This pathway is estimated to provide a cumulative exposure of 12.2 mrem at the limiting dose receptor location subject to the limits of 40 CFR 190 (25 mrem per year from effluents and external shine). For CPPU, the total offsite exposure to this location from all sources is estimated to be 13.6 mrem/year and is in compliance with the requirements of 40 CFR 190.8-10 .Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 8-1 Post Accident Vital Occupancy / Mission Dose Summary 3 For NURGE-0737 ll.B.2, Design Review of Plant Shielding Location I Function CLTP Dose CPPU Dose I I (REM) I (REM)__"________________________ Vital Occupancy Areas ___________________ ___________.. ..Main Control Room 1.2 Post accident control and monitoring of 0.24 0.11 critical safety related and supporting balance of plant equipment. Technical Support Center 1,2 Provide technical direction and analytical 1.6 0.11 support for control room operators during accident conditions. Operations Support Center4 Primary location for command and control of 0.15 0.23 (located in South post accident operational activities and Administration Building) staging for vital missions.Alternate Operations Alternate location for operations command 1.6 0.80 Support Center in Control and control of post accident operations Structure , 2 Security Control Center 4'.5 Location for overall command and control of 0.10 < 0.15 security related activities. Alternate Security Control Alternate location for security command and 0.10 0.15 Center 4 (located in North control.Gate House)Security Staging Areas 6 Undisclosed locations used for various NA6 < 3.04,7 security functions. Vital Missions Post Accident Sampling Obtain post accident coolant samples and Station Mission Doses perform sampling missions as required.Whole Body 0.29 NA'Extremity (Hand/Leg) 0.3 / 0.59 Radiation Chemistry Perform analyses of post accident samples 0.1 0.003 10 Laboratory Maintenance of Spent Fuel Provide adequate cooling of SFP.Pool (SFP) Cooling- Mission (I) to control ESW makeup flow 1.41 12.13 2.2612.13 Emergency Service Water (ESW) Valve Actuation" Mission (2) to tie-in ESW system to SFP 4.80'4 NA 1 1 Post Accident Vent Provide post accident sampling and Sampling System 1 2 monitoring of plant vent releases.Whole Body 0.001 0.002 Extremity (Hand/Leg) 0.002 0.004 Table Notes CLTP doses are calculated with TID 14844 Source Terms for the DBA LOCA.2 CPPU doses are calculated based on the Alternate Source Term for the DBA LOCA.3 Doses are based on post LOCA contained sources.4 Doses conservatively based on one year occupancy. 8-11 .Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 8-1 Post Accident Vital Occupancy / Mission Dose Summary 3 For NURGE-0737 11.13.2, Design Review of Plant Shielding Table Notes (continued) 5 Security Control Center dose is based on the North Gate House dose. The North Gate House is located closer to the reactor building contained radiation sources.6 Representative heightened security locations not previously addressed. 7 Dose is based on a conservatively assumed minimum distance 50 ft from the contained radiation sources in the reactor building.8 PASS has been deleted from Technical Specifications pursuant to approved license amendment and the requirement to have and maintain the system eliminated. 9 Sample station ingress, prepare sample, collect sample, and egress for a total duration of 130 minutes.10 Dose is based on AST contained source term at 2 hours post LOCA for 30 minute occupancy. I I For CLTP two separate operator access missions are required to provide ESW makeup to the spent fuel pool under LOCA conditions. One mission is to Elevation 749' of the reactor building to control ESW System makeup flow. The second mission is to Elevation 670' of the reactor building to tie-in the ESW system to the spent fuel pool. With CPPU a modification is made which eliminates the requirement for the ESW tie in mission.12 Mission involves ingress to reactor building at el. 749', check skimmer surge tank Water level at control panel in corridor outside Room 1-514, perform operator actions inside Room 1-514 to control flow, and egress for a total duration of approximately 8 minutes.13 The total mission dose was calculated at 24 hours post-LOCA. 14 The total mission dose was calculated at 40 hours post-LOCA. 8-12 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -9. REACTOR SAFETY PERFORMANCE EVALUATIONS This section primarily focuses on the information requested in RG 1.70 (Reference 49), Chapter 15, which applies to CPPU.9.1 Anticipated Operational Occurrences The anticipated operational occurrences, also referred to as AOOs or transients, include fuel thermal margin and loss of water level events. The overpressure protection analysis events are addressed in Section 3.1 of this report.AOOs are the result of single equipment failures or single operator errors that can be reasonably expected to occur during any mode of operation. The events are categorized based on the potential initiating cause of the threat to the fuel and reactor system.A disposition of events analysis was performed in support of transitioning to CPPU conditions. The purpose of the disposition of events evaluation was to identify the potentially limiting events that needed to be analyzed to ensure that appropriate thermal limits are established for the transition to CPPU conditions of SSES Units I and 2 using FANP methodology and ATRIUM-10 fuel. The disposition of events was primarily based upon a review of the SSES Final Safety Analysis Report (FSAR) Chapter. 15 events..Analyses for the potentially limiting transient events have been performed to assess the impact of operation at CPPU conditions. The analyses are based on an equilibrium core of ATRIUM-10 fuel operating at full CPPU thermal power and are discussed below. The transient analysis results show that the ATRIUM-10 fuel has the capability to meet the transient analyses licensing criteria at CPPU conditions. Table 9-1 presents key reactor operating parameters used in the transient analyses.Pressurization transients (LRNB, TTNB, FWCF, PRFDS, and MSIVC) were analyzed using the approved transient analysis methodology documented in References 31 and 32. The fast recirculation flow increase and inadvertent HPCI actuation events were also analyzed using the transient methodology. The loss of feedwater flow event was analyzed using COTRANSA2, the system transient analysis code (Reference 31). The other non-pressurization events used the MICROBURN-B2 methodology (CRWE, LFWH) or the XCOBRA methodology (slow recirculation flow increase). The SLMCPR of 1.07 (Section 2.2.1) presented in Table 9-1 was used to calculate OLMCPR values.For AOO's, the 4 lowest setpoint MSRVs (two MSRV-OOS for ASME over pressurization event) were assumed to be out-of-service, and a +3% MSRV setpoint tolerance was applied.The results of transient analyses at full CPPU RTP with Technical Specification Scram Speed (TSSS) are presented in Table 9-2. Figure 9-1 through Figure 9-29 present the'response of several system parameters for the limiting pressurization transients. FANP evaluated the planned change to reduce the percent of rated power at which thermal limit monitoring is required. The evaluation was performed to support the beginning of the thermal limit monitoring at 23% of 3,952 MWt (CPPU rated power) for the ATRIUM-10 fuel. The 9-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -evaluation addressed the fuel cladding integrity at low power and low flow conditions for the ATRIUM-10 fuel. In addition to consideration of critical power bound, physical characteristics of LHGR and Average Planar Linear Heat Generation Rate (APLHGR) were evaluated. The conclusions of this analysis are that the thermal power limit of 23% of CPPU rated thermal power for reactor pressures less than 785 psig is justified for the ATRIUM-I10 fuel design. This supports the small power increase from 872 MWt (25% of 3,489 MWt) to 909 MWt (23% of 3,952 MWt). As the MCPR at the original power condition (current licensed thermal power of 3,489 MWt) is viewed as having substantial margin, the MCPR values at the uprated condition are only slightly lower and would also be viewed as having substantial margin.; The small increase in the threshold power for monitoring operating limits continues to ensure adequate margin to the LHGR and APLHGR limits. These conclusions are cycle independent. The application of power- and flow-dependent limits at off-rated power and flow conditions is not significantly affected by the transition to CPPU/MELLLA conditions. The actual limits will be established using cycle-specific analysis results to ensure that adequate protection is provided.9.1.1 Fuel Thermal Margin Events Transition to CPPU conditions does not have a significant impact on the relative severity of the limiting AOO events. While the higher initial power can impact the magnitude of the event result, the potentially limiting events remain the same. This is consistent with transient analyses results for different power levels. The potentially limiting events are consistent with those identified in the SSES FSAR.The rated CPPU/MELLLA power OLMCPR necessary to support operation of the CPPU equilibrium core with TSSS is 1.34 for all cycle exposures.

9.1.2 Power

and Flow Dependent Limits A flow-dependent multiplier is applied to the LHGR thermal limits when the plant is operating at less than 100% core flow. Flow-dependent MCPR limits (MCPRf) are also established. The flow-dependent limits and multipliers are based on the results of the slow recirculation flow increase analysis. The flow-dependent limits and multipliers are established or confirmed each cycle and are based on a conservative flow run-up path.The LHGR thermal limits are also modified by a power-dependent multiplier when the unit is operating at less than 100% power.Power-dependent MCPR (MCPRp) limits were not established at this time because the transient analyses were performed only at rated CPPU conditions. The power-dependent limits based on transient analyses at off-rated conditions are expected to be similar with the current operating licensed power MCPRp limits. The CPPU off-rated power transient analyses will be performed as part of cycle specific CPPU analyses.9-2 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -9.1.3 Loss of Water Level Events 9.1.3.1 Loss of Feedwater Flow During a LOFW event, the water level decreases due to the mismatch between steam flow and feedwater flow. Makeup water is needed to maintain adequate core cooling and keep the water level above the top of active fuel. A LOFW analysis was performed for a full core of ATRIUM-10 fuel at CPPU conditions assuming that the RCIC system is the only system available to restore the reactor water level. Results of the LOFW analysis show that the minimum water level inside the shroud is approximately 90 inches above the top of active fuel (TAF), thereby ensuring adequate cooling of the core.In addition to the requirement that the reactor water level remain above TAF, an operational requirement is applied that the water level remains above the low-low-low water level setpoint (Level 1). This ensures that the ADS timer and MSIV closure trip are not activated. While this requirement is not a safety function, meeting it avoids unnecessary actuation of the safety systems. The LOFW analysis results show that the water level remains well above the Level 1 setpoint throughout the transient. 9.1.3.2 Loss of Single Feedwater Pump The loss of one FW pump event only addresses operational considerations to avoid reactor scram on low reactor water level (Level 3). This requirement is intended to avoid unnecessary reactor shutdowns. The loss of one FW pump event was evaluated, for SSES, at 3952 MWt and with core flows ranging from 99 to 108% rated. While the most limiting condition is that evaluated at 108%flow, both 99 and 108% cases are expected to result in a Level 3 scram. While low level scram is not avoided, the level will recover rapidly after the scram.9-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -9.2 DESIGN BASIS ACCIDENTS This section addresses the radiological consequences of Design Basis Accidents (DBAs) for SSES. The topics addressed in this evaluation are: Loss of Coolant Accident Inside Containment (LOCA)Main Steam Line Break Outside Containment (MSLB)Fuel and Equipment Handling Accident (FHA/EHA)Control Rod Drop Accident (CRDA)Recirculation Pump Seizure, and Instrument Line Break Accident (ILBA)The impact of the proposed CPPU on the radiological consequences of the major DBAs (i.e.LOCA, MSLB, FHA/EHA, and CRDA) has been evaluated by PPL Susquehanna in a separate license amendment request, Reference 22, which proposed a full-scope implementation of an Alternative Source Term (AST) that complies with the guidance given in RG 1.183 and USNRC Standard Review Plan 15.01. The calculated AST dose analyses show that the dose criteria of 10 CFR 50.67 are met for CPPU conditions. The consequences of the Recirculation Pump Seizure accident have been evaluated for CPPU conditions with the results of this event determined to be non-limiting. The ILBA analysis was evaluated for CPPU event remain bounded by the MSLB.9.3 SPECIAL EVENTS This section considers two special events: evaluation are: conditions. The radiological consequences of this ATWS and SBO. The topics addressed in this Topic CLTR Disposition SSES Result 9.3.1 ATWS (Overpressure) -Event Selection 9.3.1 ATWS (Overpressure) -Limiting Events 9.3.1 ATWS (Suppression Pool Temperature) -Event Selection 9.3.1 ATWS (Suppression Pool Temperature) -Limiting Events 9.3.1 ATWS (Peak Cladding Temperature)

9.3.2 Station

Blackout 9.3.3 ATWS with Core Instability 9-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -9.3.1 Anticipated Transient Without Scram PPL has requested approval of amendments to the Standby Liquid Control (SLC) System Technical Specifications in Reference

51. The proposed changes would modify the SLC System to operate with one pump, versus the current two pump operation, and would also employ the use of enriched boron. The CPPU ATWS analysis presented below assumes this SLC System configuration.

The overpressure evaluation includes consideration of the most limiting RPV overpressure cases: (1) MSIVC and (2) PRFO. The vessel overpressure response for these two cases bounds the IORV and the LOOP cases.For SSES, the LOOP does not result in a reduction in the RHR pool cooling capability relative to the MSIVC and PRFO cases. With the same RHR pool cooling capability, the containment response for the MSIVC and PRFO cases bound the LOOP case.SSES meets the ATWS mitigation requirements defined in 10 CFR 50.62: 1. Installation of an Alternate Rod Insertion (ARI) system;1. Boron injection equivalent to 86 gpm; and 2. Installation of automatic Recirculation Pump Trip (RPT) logic (i.e., ATWS-RPT). In addition, plant-specific ATWS analysis is performed to ensure that the following ATWS acceptance criteria are met: 1. Peak vessel bottom pressure less than ASME Service Level C limit of 1500 psig;2. Peak suppression pool temperature less than 220 'F (Suppression Pool temperature design limit), and 3. Peak containment pressure less than 53 psig (Containment pressure design limit).The ATWS analysis is performed for CPPU RTP to demonstrate the effect of the CPPU on the ATWS acceptance criteria' There are no changes to the assumed operator actions for the CPPU ATWS analysis and there is no change to the required hot shutdown boron weight. The required sodium pentaborate decahydrate solution concentration, minimum boron *enrichment, and number of required SLCS pumps provide acceptable results at CPPU conditions. The key inputs to the ATWS analysis are provided in Table 9-3. The results of the analysis are provided in Table 9-4.The results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the SSES response to an ATWS event at CPPU is acceptable. Coolable core geometry is assured by meeting the 2200*F PCT and the 17% local cladding oxidation acceptance criteria of 10 CFR 50.46. The highest calculated PCT for CPPU is 1434'F, which is significantly below 1500 'F. Consequently, the fuel cladding oxidation is insignificant and less than the 17% local limit.9-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -9.3.2 Station Blackout Station Blackout was re-evaluated using the MAAP computer code at CPPU power levels and the guidelines provided in NUMARC 87-00. The pre-CPPU SBO evaluation was performed using the BWRSAR computer code. For benchmarking purposes, the event was also evaluated using the BWRSAR code at CPPU conditions. Both codes produced similar results for CPPU.Future analyses will utilize the MAAP code for SBO evaluations. The CPPU analysis assumed the following;

1. RCIC is the only system available to provide makeup to the reactor vessel.2. HPCI is assumed to be unavailable.

3. The CST is the water source for vessel makeup.4. Recirculation pump seal leakage is 100 gpm.5. Suppression pool cooling is not available until after 4 hours.The plant responses to and the coping capabilities for the SBO event are affected slightly by operation at CPPU RTP, due to the higher initial power and the increased decay heat. Decay heat was calculated based on operation at 100% rated power for 100 days prior to the SBO.There are no changes to the systems and equipment used to respond to an SBO and the coping time (4 hours) remains unchanged.

Areas containing equipment necessary to cope with a station blackout event were evaluated for the effect of loss of ventilation due to a SBO. The evaluation shows that equipment operability is assured due to conservatism in the existing design and qualification bases. The battery capacity remains adequate to support RCIC operation after CPPU. Adequate compressed gas capacity exists to support SRV actuations. The current CST reserve (135,000 gallons) for RCIC use ensures that adequate water volume is available to remove decay heat, depressurize the reactor, and maintain reactor water level between Level 4 and Level 7 (approximately 132,000 gallons required for the four hour coping time at CPPU). Peak containment pressures and temperatures remain within design basis.Adequate NPSH margin exists for the RCIC pump during the event and the RHR pumps at the end of the event.Based on the above evaluation, SSES continues to meet the requirements of 10CFR50.63 at CPPU conditions. 9.3.3 ATWS with Core Instability The ATWS with core instability event occurs at natural circulation following an RPT.Therefore, it is initiated at approximately the same power level as a result of CPPU operation because the MELLLA upper boundary is not increased. The core design necessary to achieve CPPU operations may affect the susceptibility to coupled thermal-hydraulic/neutronic core oscillations at the natural circulation condition, but would not significantly affect the event progression. Several factors affect the response of an ATWS instability event, including operating power and flow conditions and core design. The limiting ATWS core instability evaluation presented in 9-6 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -References 33 and 34 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum flow point. [[CPPU allows plants to increase their operating thermal power but does not allow an increase in control rod line. [[]J The conclusion of Reference 34 and the associated NRC SER that the analyzed operator actions effectively mitigate an ATWS instability event are applicable to the operating conditions expected for CPPU at SSES.Impact of ATRIUM-1O Fuel The NRC has concluded based on the BWROG submittals and their own work (Technical Evaluation Report for Reference

33) that the presence of large amplitude power oscillations does not alter the consequences of an ATWS event significantly enough to warrant a change in the current ATWS Rule. Furthermore, even though a large uncertainty is associated with the calculated power oscillation amplitude, the conclusion would remain valid even if the amplitude was in error by an order of magnitude.

These conclusions are generic for GE fuel designs, core loading patterns, step-through sequences and all plants.When one considers the wide range of initial stability margins for this generic range of core and plant conditions, the change in fuel design to ATRIUM-1OTM fuel is small. It is true that there will be some changes in parameters important to determining the reactor stability [[]] Even these larger variations in initial decay ratio are small compared with the potential uncertainties in the models for the range of conditions anticipated for the events.Based on these observations, the impact of operation with ATRIUM- 10 is small compared to the large uncertainties associated with this event analysis.9-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 9-1 Parameters Used for ATRIUM-10 CPPU Transient Analyses Rated thermal power, MWt 3952 Analysis power, % rated 100*Analysis dome pressure, psia 1050 Analysis turbine pressure, psia 977 Rated vessel steam flow, Mlbmlhr 16.532 Analysis steam flow, % rated 100 Rated core flow, Mlbm/hr 100 Rated power core flow range, % rated 99-108 Analysis core flow, Mlbm/hr 99-108 Normal feedwater temperature, OF 399.3 Steam bypass capacity, % rated steam flow 22.21 Reactor low-level water level 3 scram (in, AVZ)-Loss of feedwater flow 535.5-Loss of one feedwater pump 540.5 Safety limit MCPR 1.07 Number of MSRVs assumed in analysis Low bank 2+Mid-bank 6 High bank 8 MSRV setpoint, psig Low bank 1175+3%Mid-bank 1195+3%High bank 1205 +3%* The only events analyzed at a different power level (102%) are the loss of feedwater flow, ASME Over pressurization and LOCA." The value given is for 5 turbine bypass valves (TBV). A conservative value of 21.4% of rated steam flow was used in the transient analyses. AOO analyses crediting the turbine bypass assumed only 4 TBVs in operation (where conservative to do so). N/A for events analyzed without turbine bypass.4 lowest setpoint (2 from low bank and 2 from mid-bank) MSRVs are assumed OOS for the transients analyzed (except for ASME analysis which considered 2 MSRV-OOS). 9-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 9-2 ATRIUM-10 Transient Analysis Results*Peak Neutron Flux%of rated: CPPU Peak Heat: Flux,% of rated CPPU ACPR MCPR Operating* Limit Event Generator load rejection/turbine trip 292 122 0.27 1.34 with turbine bypass failure Generator load rejection/turbine trip with turbine bypass failure and recirculation 426 133 0.36 1.43 pump trip failure (EOC-RPT-OOS) Feedwater controller failure max 221 119 0.27 1.34 demand Feedwater controller failure maximum demand with turbine bypass failure 268 123 0.31 1.38 (TBV-OOS)Pressure regulator downscale failure .147t 118t 0.29t 1.36t Loss of feedwater heating + + 0.18 1.25 Inadvertent HPCI actuation 112 112 0.18 1.25 Rod withdrawal error NA 1.32§Rod withdrawal error with turbine bypass +

• NA +.48§failure (TBV-OOS)Slow recirculation increase NA NA NA MCPRr Fast recirculation increase 188 92 0.13 1.20 Generator load rejection/turbine trip 235 116 0.22 1.29 with turbine bypass MSIV closure -all valves 145 101 0.11 1.18 MSIV closure -1 valve 120 109 0.12 1.19 Loss of feedwater flow 102 102 NA NA* All analyses performed with TSSS insertion times.This event is only possible if the backup pressure regulator is out of service.+ Peak neutron flux and peak heat flux are not reported for the slow transients.

§ With rod block monitor (RBM) setpoint of 111%.9-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 9-3 Key Inputs for ATWS Analysis Input Variable CPPU Reactor power (MWt) 3952 Reactor dome pressure (psia) 1050 SRV capacity (Mlbm/hr@ 1175 psig and 3% 14.143 accumulation) High pressure ATWS-RPT (psig) 1170 Number of SRVs Out-of-service (OOS) 0 Table 9-4 Results of ATWS Analysis'Parameter CPPU Acceptance Criteria Peak vessel bottom pressure (psig) 1336 <1500 Peak cladding temperature (TF) 1434 52200 Peak suppression pool temperature ('F) 206 :220 Peak containment pressure (psig) 16.1 :_53 Note: 1. Cladding oxidation remains below the 10 CFR 50.46 limits.9-10 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -0 0.SQ EPU LRNBTTNB at PlOOF99Ix18440.4 12/15/05 0&43:31 NOS-2577. J I-279M2 Figure 9-1 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F Power, Heat Flux, and Flows Q.0 35.0-F-A! 30.0 o 25.0-Z 0 20.0-15.0.0 1.0 -20.0 4.0 0.0 Time (seconds)SQ EPU LRNBTTNB at P100-799.x18440.4 12/15/05 0043.3) 90S-2577.

• JO D-27XM2 Figure 9-2 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F Downcomer Water Level 9-11 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -a.Time (seconds)SO EPU LRNBTTNB at P100_F99__X18440.4 12/15/05 0&4"31 .NOS-25779, J- 0-27M8 Figure 9-3 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F Pressures 1.2o Low Setpoint__AMedium Setpoint..h...S pot. ..I I A I".*~~s0 0&43 ,O-59 h. \ ,?I .- I S 0; I S Figur 94*0 ~~I jI .o .,.oI:.'a I I *I/ 0-3,1 / -NT9,Je D27e Fimue (9-ods Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F MSRV Position 9-12 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -1500.0-0.> 10oo0o-'B-x 0 Low Setpoint Medium Se!paint Hi~hSetpoint 4***. .~.I ,~f-'-----

'I I" .2 II .-.I,.'I I: I'I: I: i~ I.Ii Ii I 2. 2o Time (seconds)4.0 o.0 GA SO EPU LRNBTTNB at P100_F99_.X18440.4 12,/15,/0 & 04".1 NOS-257T9. 0-2792 Figure 9-5 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F MSRV Flows-..................... .------------------------ -.10, Scram Reactivity -.2- Doper Rectivity -X Total Reactivi ..-.25'. 10 2.0 20 4.0 00 Time (seconds)SQ EPU LRNBTTNB at PlOO-F99..X18440.4 12/15/05 0042.I NOS-25775. "8 I-.2792 Figure 9-6 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F Reactivities (Ak/k)9-13 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -524.0-C.).0 1.0 20 0o Time (seconds)4.0 5.0 6.0 SO EPU LRNBTTNB at P100_F99__X18440.4 12/15/05 0&-4"31 Nl-25T9. ,0D27-2M2 Figure 9-7 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure at 100%P/99%F Core Inlet Enthalpy*0 0 01 SQ EPU LRNBTTNB.NORPT at P1OOi-FO8--X19403.7 01/31/se 0:02.40 NoS-M0116. MBe I-2"?7 Figure 9-8 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F Power, Heat Flux, and Flows 9-14 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -40.-N C W 25.0-_J C 15,0-(OO 10.0..0 -1.0 A. 20 4.0 A. 6.0 Time (seconds)SO EPU LRNB_TTNBNORPT at P1OOFX108__X19403.7 01/31/06 NOS-miole. J06 0-2"7 Figure 9-9 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F Downcomer Water Level 1350.0.1300.0.1250.0, 0.Steam Dame Pressure 1050.0.0 1.0 2.0 30 40 5.0 6.0 Time (seconds)SO EPU LRNB_TTNNOIRPT at PI00_F108__X19403.7 01/31/06 M060540 HoQS-301tS. do -I-2975W Figure 9-10 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F Pressures 9-15 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -C.0 0 CL.0.4>..2-V)Low Setpoint Me .!ýdium ~IPipp'in H ~l~Af. 'v.r~ %, a MIE%I.%%% \.0 Time (seconds)4.0 5.0 6.0 SQ EPU LRNB_TTNBNORPT at PIOO _.F108__X19403.7 01/31/06 060240 HOS-el 16. JD8 ID-227M8 Figure 9-11 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F MSRV Position 2500.0 Low Setpoint Medium Setpoint____.jh..Se~tp.°.n~t~ 2000.0- `v,. -------------,, 4, 1soo "N o M1'Time (seconds)SQ EPU LRNB.TTNBNORPT at 01/31/06 OOZZ40 100-201 6. JOB M-29758 P100_F108__X19403.7 Figure 9-12 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F MSRV Flows 9-16 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Z, SQ EPU LRNa..TTNE..NoRPT at 01/31/06 0002.4 mr'.mm16 "0 0-275 P100_F108__X19403.7 Figure 9-13 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F Reactivities (Ak/k)527.0-5260 t5 524.0-.11..o 1:0 0o io Time (seconds)4.0 5.0 6.o SO EPU LRNBTTNB-NORPT at P100-F108_-X19403.7 01/31/06 0"92:40 0n"..2O11.M " 0,-207M5 Figure 9-14 Generator Load Rejection/Turbine Trip With Turbine Bypass Failure and EOC-RPT-OOS at 100%P/108%F Core Inlet Enthalpy 9-17 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -0 U 0.4 Time (seconds)SO EPU FWCF at PIOO_F1O8_-X15750.O 12/15/05 0&48=2 Mr)-25780. JOB -M025 Figure 9-15 Feedwater Controller Failure Maximum Demand at 100%P/108%F Power, Heat Flux, and Flows 70.0-0 N 00.0-6)30.0.0 0.0 10.0 45.0 20.0 25.0 30.0 Time (seconds)SO EPU FWCF at PIOOFI08--X15750.O 12/15/05 04M22 NO-25780. = 2-50020 Figure 9-16 Feedwater Controller Failure Maximum Demand'at 100%P/108%F Downcomer Water Level 9-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -f.2 Time (seconds)SQ EPU FWCF ot P100_F1O8__X15750.O 12/15/08 0&4&22 NV)--25780M " D0-0=25 Figure 9-17 Feedwater Controller Failure Maximum Demand at 100%P/108%F Pressures 0.2.2 -a.5 A Low Setpoint Aediu Fe~pIRT!% , I N N 0Tim.0 (10.0 ds Time (seconds)2&0 25.0 36.0 SQ EPU FWCF at P100_F108__X15750.O 12/1S/05 0&.4&U M-257M0. X 0,00525 Figure 9-18 Feedwater Controller Failure Maximum Demand at 100%P/108%F MSRV Position 9-19 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -A600.0*400.0-200.0-V)Time (seconds)SO EPU FWCF at P100-FIOS.X15750.O 12/15/05 0&41.22 k -25780. " 000525 Figure 9-19 Feedwater Controller Failure Maximum Demand at 100%P/108%F MSRV Flows bs0-.12-.2---------------------- Scram Reactivity Moderator Reactivit------ ....... ....Eemctvity --25 -0 550 16.0 15.0 Time (seconds)20.0 2!.O 30.0 SQ EPU FWCF at P100O-F08--X15750.O 12/15/05 0&4&22 NOI-257M0. J- -OM025 Figure 9-20 Feedwater Controller Failure Maximum Demand at 100%P/108%F Reactivities (Ak/k)9-20 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -522.0-In520.0 0 OItU4.&0 10.0 1s.0 Time (seconds)20L0 2250 SO EPU FWCF at P100-F108-X15750.O 12/15/05 0&4=22 00-2527A0 s 6n Figure 9-21 Feedwater Controller Failure Maximum Demand at 100%P/108%F Core Inlet Enthalpy 300.0 200.0........ ------- --0I t,, ', Relative Core Flow Relative Steam Flow Relative Feed Flow-100,0.0 5.0 10.0 1.0 20.0 2i.0 =0.0 Time (seconds)SO EPU FWCFNOBYP at P1OOF108..X15750.O 02/09/06 12:12M0 6 -3004. J Figure 9-22 Feedwater Controller Failure Maximum Demand With TBV-OOS at 100%P/108%F Power, Heat Flux, and Flows 9-21 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -70.0 B1 40.0-~0 30.0 ,.0 0-0 10.0 10 200 2A0 Time (seconds)SO EPU FWCF-NOBYP at P100F108_.X15750.O 02/09/06 11121M NV-3N04. J 0-19010 Figure 9-23 Feedwater Controller Failure Maximum Demand with TBV-OOS at 100%P/108%F Downcomer Water Level 130W 1300.0 -1250.0-1200.0-01500-1t0.0 -Steam Dome Pressure Lower Plenumn Pressure I\I I 13ARJ.U4 5.0 10.0 15.0 Time (seconds)20L0 2&.0 SQ EPU FWCF-NOBYP at P100FO8F10-X15750.O 02/09/06 1:1204611 J00 C-19010 Figure 9-24 Feedwater Controller Failure Maximum-Demand with TBV-OOS at 100%P/108%F Pressures 9-22 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -1.2 0 0 10 0.L.2 Low Setpoint__+/-ýI~~ 9L§q I2 ANii lI v"-"I 5.0 10O I0o Time (seconds)26.o 2&.0 36.o SQ EPU FWCFJ'l0BYP at 02/00/06 MU1M0 NOS-3004. XB 15-19010 P100-F.108_X15750.0 Figure 9-25 Feedwater Controller Failure Maximum Demand with TBV-OOS at 100%P/108%F MSRV Position Low Setpoint__Aediwnf ýeRij.... fiah tpint1ý .1200.0.".5 0W.Z, 400.0 V)%%"?..,,.0 5T0 16.0 1 i.0 Time (seconds)2.0 25.0 30.0 SQ EPU FWCFNOBYP at P100.-F1O8X15750.0 02/09/00 12:12M0 NV-30004. D0 I-19010 Figure 9-26 Feedwater Controller Failure Maximum Demand with TBV-OOS at 100%P/108%F MSRV Flows 9-23 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -0 SQ EPU FWCF-NOBYP at P10O_F1O8_X15750.O 02/09/06 12:1+/-06 N-306 " 1>I0WI0 Figure 9-27 Feedwater Controller Failure Maximum Demand With TBV-OOS at 100%P/108%F Reactivities (Ak/k)524.0 522.0, 02 51&0.-516..0. 1520.0.Time (seconds)SO EPU FWCFNOBYP at P1OO_F1O8__X15750.O 02/00/06 12A12Z6N q-094 v~ D-1"10t Figure 9-28ýFeedwater Controller Failure Maximum Demand With TBV-OOS at 100%P/108%F F Core Inlet Enthalpy 9-24 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -4)0 0 6.0 &0 Time (seconds)SQ EPU PRFDS at P100_F108__X19403.7 08/15/05 170= NQS-54M, X6 D-258 Figure 9-29 Key Parameters for Limiting Full-Power Pressure Regulator Downscale Failure Event.(PR-OOS) 9-25 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -10. OTHER EVALUATIONS 10.1 HIGH ENERGY LINE BREAK High-energy line breaks (HELBs) are evaluated for their effects on equipment qualification. The topics addressed in this evaluation are: i Topi Cc cLTR Disposition SSES Result 10.1.1 Steam lines 10.1.2 Liquid lines Valve closure times do not change and there is no change to any flow restricting orifice. The results of the evaluation of HELBs are provided in Table 10-1.10.1.1 Steam Line Breaks[[The steam line HELBs in the licensing basis were evaluated for CPPU.Main Steam Line Breaks CPPU has no effect on MSLBs because steam conditions at the postulated break locations are unchanged. Valve closure times do not change and there is no change to any flow restricting orifice. CPPU has no effect on the steam pressure or enthalpy at the postulated break locations. Therefore, CPPU has no effect on the mass and energy releases from an HELB in an MSL.HPCI Steam Line Breaks Because there is no increase in the reactor dome pressure relative to the CLTP analysis, the mass flow rates for HPCI steam line breaks do not increase. Valve closure times do not change.Therefore the CLTP analysis of the HPCI steam line breaks is bounding for CPPU conditions. 10-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -RCIC Steam Line Breaks Because there is no increase in the reactor dome pressure relative to the CLTP analysis, the mass flow rates for RCIC steam line breaks do not increase. Valve closure times do not change.Therefore, the CLTP analysis of the RCIC steam line breaks is bounding for CPPU conditions. 1]]10.1.2 Liquid Line Breaks Operation at CPPU conditions requires an increase in the MS and FW flows, which results in a slight increase in downcomer sub cooling. This increase in sub cooling may lead to increased* break flow rates for liquid line breaks. Only the mass and energy releases for HELBs in the RWCU and FW systems may be affected by CPPU and were re-evaluated at CPPU conditions. RWCU System Line Breaks The CPPU mass and energy releases for RWCU line breaks are bounded by the conditions at ARTS/MELLLA. The reactor building pressure, temperature and relative humidity profiles at CPPU conditions were evaluated for the effect on equipment qualification as discussed in Section 10.3.Feedwater System Line Break The environmental conditions that result based on the CPPU mass and energy releases for the Feedwater line breaks are bounded by the Main Steam conditions at ARTS/MELLLA. The reactor building pressure, temperature and relative humidity profiles at CPPU conditions were evaluated for the effect on equipment qualification as discussed in Section 10.3.Thus safety related structures, systems, and components affected by the RWCU and Feedwater liquid line breaks are able to perform their safety related functions as intended at CPPU condition Pipe Whip and Jet Impingement Pipe Whip and Jet Impingement Pipe whip and jet impingement loads resulting from the high-energy pipe breaks are directly*proportional to system pressure. With the exception of the Feedwater system, CPPU conditions do not result in an increase in the pressure considered in the high-energy piping evaluations and thus there is no increased pipe whip or jet impingement loads on HELB targets or pipe whip restraints. The loads, due to the increased pressure in the Feedwater System, are acceptable because they remain below the allowable used in the existing qualifications of pipe whip restraints and jet impingement targets. Additionally, a review of pipe stress calculations determined that the increases in the Feedwater system temperatures and pressures associated with CPPU conditions will not result in pipe stress levels above the thresholds required for postulating HELBs, except at locations already evaluated for breaks (see Section 3.5). As a result, CPPU 10-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -conditions do not result in new HELB locations, nor affect existing HELB evaluations of pipe whip restraints and jet targets.Internal Flooding from Feedwater Line Break This section addresses the evaluation of the effects ofjet impingement from postulated breaks in high energy lines addressed in FSAR, Section 3.6.Internal flooding due to a High Energy Liquid Line Break, outside containment, is postulated only inside the reactor and turbine buildings. The only significant flooding in the reactor building is a FW break in the main steam tunnel where the maximum flood depth is dependent only upon the hydrostatic head at which the existing floor-level pressure relief panels release into the turbine building tunnel. The turbine building basement elevation is designed to contain the full inventory of the hotwell which is not changing due to CPPU. The FW system hardware changes and pressure increases due to CPPU do not affect the existing flooding analyses.Because the existing pressure relief panels are not changing, the existing maximum FW break flooding analysis is valid for the CPPU conditions. 10.2 MODERATE ENERGY LINE BREAK Moderate energy line breaks (MELBs), cracks, are evaluated for their effects on equipment qualification. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result Flooding Environmental Qualification ]1 Flooding due to MELBs cracks, are dependent on one, or more, of the following: system/vessel/tank inventories, floor-level pressure relief panels (in the Reactor Building Main Steam Tunnels), crack flow rates [proportional to (operating pressure)" 2], and operator actions.CPPU does not significantly change normal operating pressures and [[]] pressure relief panels, and operator actions. Therefore, the existing MELB internal flooding analyses are valid for the CPPU conditions. 10.3 ENVIRONMENTAL QUALIFICATION Safety related components are required to be qualified for the environment in which they are required to operate. The topics addressed in this evaluation are: 10-3 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Topic CLTR Disposition SSES Result Electrical Equipment Mechanical Equipment'With Non-Metallic Components Mechanical Component Design Qualification .10.3.1 Electrical Equipment The safety-related electrical equipment was reviewed consistent with the requirements of IOCFR50.49 to ensure the existing qualification remains adequate for the normal and accident conditions expected in the installed locations as a result of CPPU. 10CFR50.49 acceptance criteria including pressure, temperature, humidity, and radiation requirements were used to make this determination. Table 10-2 provides a listing of the parameters of interest associated with environmental qualification and the expected effect as a result of CPPU.Inside Primary Containment Environmental qualification of safety-related electrical equipment located inside primary containment is based on the environment expected to exist during DBA conditions and the resultant temperature, pressure, humidity, and radiation consequences. The qualification also includes the effects of normal plant operation up to the time of the accident.The normal maximum ambient temperature, pressure, and humidity conditions inside primary containment are not increased as a result of CPPU. The post-accident humidity conditions are also unaffected by CPPU. The normal and post-accident radiation values and the post-accident peak temperature and pressure inside primary containment increase for CPPU.The CPPU peak temperature is bounded by the EQ temperature profile used for DBA qualification of safety-related electrical equipment. The increased drywell peak pressure that results from CPPU is bounded by the existing qualification levels of the drywell EQ equipment. The radiation levels under normal plant conditions increase in proportion to the increase in reactor thermal power as a result of CPPU. The accident radiation levels increase in the primary containment as a result of CPPU. The increased radiation levels as a result of CPPU have reduced the qualified life of some solenoid valves. Certain additional equipment (i.e., power and instrument cable) may require radiation dose reduction analysis to demonstrate qualification for the radiation level associated with the CPPU condition. 10-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Outside Primary Containment Environmental qualification of safety-related electrical equipment installed outside primary containment is based on normal operating conditions and the effects of DBAs which occur inside the primary or secondary containments. These accidents include MSLB or LOCA inside primary containment and HELB or CRDA in secondary containment. The qualification is based upon the most limiting accident for the room under analysis.The normal temperature, pressure, and humidity conditions in secondary containment do not change as a result of CPPU. The accident peak temperature, pressure, and humidity conditions in secondary containment do not change as a result of CPPU. The HELB and MELB flooding levels in secondary containment are also unaffected by CPPU effects.Normal operating radiation levels may increase up to 20% with CPPU in some areas. This increase is less than the design basis normal radiation levels used for environmental qualification. As such, the normal operating radiation levels used for qualification are unchanged. Accident radiation levels increase in secondary containment and the control structure in the vicinity of the SGTS filters (control room and relay rooms are not affected). The increased radiation levels as a result of CPPU have reduced the qualified life of some EQ equipment. Certain additional equipment (i.e., relays, solenoid valves, flow switches, fire detection controls, pressure switches, hydrogen analyzer cell o-rings, heater control equipment, terminal blocks, and RHR pump motor oil) initially do not meet qualification requirements based on CPPU conditions. However, the affected EQ equipment is expected to remain qualified with the application of radiation dose reduction analysis.10.3.2 Mechanical Equipment With Non-Metallic Components The accident temperatures, pressure, and radiation level, and the normal radiation level increase due to CPPU as discussed in Section 10.3.1. The design control program ensures that non-metallic components (e.g., seals, gaskets, lubricants, diaphragms) are specified and procured for the environment in which they are intended to function.10.3.3 Mechanical Component Design Qualification The mechanical design of safety related equipment/components (e.g., pumps, heat exchangers) in certain systems is affected by operation at CPPU due to increased temperatures, and in some cases, flows, and pressures. The CUFs of mechanical components are evaluated in Section 3.The effects of increased fluid induced loads on safety-related components are described in Sections 3 and 4.1. Increased nozzle loads and component support loads due to the revised operating conditions were evaluated within the piping assessments in Section 3.10.4 TESTING Testing is required for the initial power ascension following the implementation of CPPU. The topics addressed in this section are: 10-5 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Topic CLTR Disposition SSES Result Plant/Component Testing Large Transient Testing Based on the analyses and GE BWR experience with uprated plants, a been established for the initial power ascension steps of CPPU. These the normal Technical Specification testing requirements, are as follows: standard set of tests has tests, which supplement" Testing will be done in accordance with the Technical Specifications Surveillance Requirements on instrumentation that is re-calibrated for CPPU conditions. Overlap between the IRM and APRM will be assured." Steady-state data will be taken at points from 90% up to the 100% of the pre-CPPU RTP, so that system performance parameters can be projected for CPPU power before the pre-CPPU RTP is exceeded.CPPU power increases above the 100% pre-CPPU RTP will be made along an established flow control/rod line in increments of equal to or less than 5% power. Steady-state operating data, including fuel thermal margin, will be taken and evaluated at each step. Routine measurements of reactor and system pressures, flows, and vibration will be evaluated from 10-6 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -each measurement point, prior to the next power increment. Radiation measurements will be made at selected power levels to ensure the protection of personnel." Control system tests will be performed for the reactor FW/reactor water level controls, pressure controls, and recirculation flow controls, if applicable. These operational tests will be made at the appropriate plant conditions for that test at each of the power increments, to show acceptable adjustments and operational capability." Testing will be done to confirm the power level near the turbine first-stage scram bypass setpoint.The same performance criteria will be used as in the original power ascension tests, unless they have been replaced by updated criteria since the initial test program. [[[[I SSES does not plan to perform large transient testing as part of CPPU implementation. The justification for not performing large transient testing is provided as a stand-alone attachment (Attachment

8) to the CPPU LAR.Further, the important nuclear characteristics required for transient analysis are confirmed by the steady state physics testing. Transient mitigation capability is demonstrated by other .tests required by the Technical Specifications.

In addition, the limiting transient analyses are included as part of the reload licensing analysis.10.5 INDIVIDUAL PLANT EVALUATION Probabilistic risk assessments (PRAs) are performed to evaluate the risk of plant operation. The topics considered in this section are: 10-7 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Topic CLTR Disposition SSES Result 10.5.1 Initiating Events 10.5.2 Component Reliability 10.5.3 Operator Response 10.5.4 Success Criteria 10.5.5 External Events 10.5.6 Shutdown Risk 10.5.7 PRA Quality 10.5.8 Integrated Risk Impact Background The SSES PRA is a state-of-the-technology tool developed consistent with current PRA methods and approaches. The SSES PRA model is developed and quantified using the EPRI CAFTA suite of codes.The SSES PRA is derived based on realistic assessments of system capability over the 24 hour mission time of the PRA analysis. Therefore, PRA success criteria may be different than the design basis assumptions used for licensing SSES. This risk assessment examines the risk profile changes from this realistic perspective to identify changes in the risk profile on a best estimate basis that may result from postulated accidents, including severe accidents. Scope of CPPU Risk Evaluation The scope of the risk assessment for the CPPU addresses the following plant risk contributors:

• Level I Internal Events At-Power Core Damage Frequency (CDF)* Level 2 Internal Events At-Power Large Early Release Frequency (LERF)* External Events At-Power* Seismic Events* Internal Fires* Other External Events* Shutdown Assessment Risk impacts due to internal events are assessed using the "AUG06preEPU-C" and "AUG06EPU" versions of the integrated Level I and Level 2 PRA models. External events are evaluated using the analyses of the SSES Individual Plant Examination of External Events (IPEEE) Submittal.

The impacts on shutdown risk contributions are evaluated based on more generic insights and assumptions obtained from a review of other industry BWR shutdown PRA results.10-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -All commitments resulting from the SSES Individual Plant Examination (IPE) and the IPEEE Programs have been resolved.The identification of PRA elements evaluated in the CPPU risk assessment was derived from the NEI PRA Peer Review Guidelines, the ASME PRA Standard,- and the NRC Review Standard for Extended Power Uprates. Each of these major risk elements was examined:* Initiating Events* Systemic/Functional Success Criteria 9 Accident Sequence Modeling* System Modeling* Failure Data* Human Reliability Analysis* Structural Evaluations

• Quantification
• Containment Response (Level 2)In addition, shutdown risk and external events were also investigated.

Overview of CPPU Modifications The following are modifications to the plant and operations for the CPPU.Hardware Modifications The hardware changes required to support the CPPU were reviewed and determined not to result in new accident types or a measurable increase in the frequency of challenges to plant response. This assessment is based on review of the plant hardware modifications and engineering judgment based on knowledge of the PRA models. The majority of the changes are characterized by either:* Replacement of components with enhanced like components" Upgrade of existing components Extensive changes to plant equipment have been shown by operating experience to result in an increase in system unavailability or failure rate during the initial testing and break-in period. It can be expected that there will be some short-term increase in such events but the frequency and duration of such events cannot be predicted. Nevertheless, it is expected that a steady state condition equivalent to (or potentially better than) current plant performance would result.Procedural Modifications CPPU Implementation -This "modification" controls the procedural requirements for ascending to a new CPPU power level and assures that all required documentation is updated. This modification will control all applicable changes to implement the uprate, but does not represent the procedural changes that are required to operate during and after the uprate.In order to ensure the plant is operated safely, adjustments to the Emergency Operating Procedures/Severe Accident Management Guidelines (EOPs/SAMGs) will be made consistent with CPPU operating conditions. In almost all respects, the EOPs/SAMGs are expected to remain 10-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -unchanged because they are symptom-based; however, certain parameter thresholds and curves are dependent upon power and decay heat levels and will require slight modifications. EOP variables that play a role in the PRA and which may require adjustment for the CPPU include:* Heat Capacity Temperature Limit (HCTL)o Pressure Suppression Limit (PSL)]These variables may require adjustment to reflect the change in power level, but will not be adjusted in a manner that involves a change in accident mitigation philosophy. The [[ ]]relate to long-term scenarios, any changes in the scenario timings associated with CPPU changes to these curves will be minor and would not significantly impact the human error probabilities in the PRA. However, the change in timing values has been explicitly addressed in the event tree evaluations and the human reliability analysis for CPPU conditions in the AUG06EPU PRA model.Changes to the EOPs/SAMGs as a result of the CPPU were not available prior to completion of the PRA evaluation. It is assumed that the procedural changes (e.g., modification to HCTL curve) have a minor impact on the PRA results. No changes to the PRA are identified as a result of potential EOP/SAMG procedural changes.Setpoint Changes Setpoint changes for the CPPU that have been identified include:* RFP Turbine High Speed Stop Setpoint* FW Low Suction Pressure Trip Setpoint* Rod Worth Minimizer (RWM)Low Power Alarm Point/Low Power Setpoint* Reactor Protection System (RPS) First Stage Pressure SCRAM Bypass* Delete Low RPV Water Level Confirmatory Signal for #2 Recirculation Runback* Reactor Recirculation Runback Limiter #2 Ramp Rate Change* Moisture Separator Cross Around Relief Valve Setpoint Increase (requires some hardware changes).* Computer Point and BOP Calculation Changes.* EHC Upgrades for CPPU* Power Dependent Condenser Pressure Alarm Power Input Rescaling* APRM-biased ARTS/MELLLA Setpoint Changes* MSIV High Flow Isolation Setpoint* Standby Liquid Control Storage Tank Hi/Low Level Signal* Instrumentation and Computer Software Changes for CPPU Implementation Other minor setpoint changes may be made to various systems for operational margin purposes.Such minor setpoint changes have no direct quantifiable impact on the plant risk.The RPV operating pressure and the operating temperature are not being changed as part of the CPPU. In addition, changes to the following setpoints are not anticipated for the CPPU: " RPT/ATWS high dome pressure* RPV level trips/actuations.

• SV/SRV setpoints 10-10 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Most of the planned setpoint changes listed above will not result in any quantifiable impact to the PRA. Key setpoints that play a role in the PRA are planned to remain unchanged, such as:* Main Steam SRV opening and closing setpoints* RPV pressure setpoint (e.g., ATWS RPT high pressure setpoint)Analyses performed by the NSSS vendor showed that the above existing current licensed thermal power (CLTP) setpoints remain adequate for CPPU conditions, which results in no required changes to the PRA model.Plant Operating Conditions The key plant operational modifications to be made in support of the CPPU are: Increase in the current licensed thermal power from 3489 to 3952 MWt (general change, not identified in the Modification List)Corresponding increase in the FW/Condensate flow and steam flow rates (general change, not identified in the Modification List)* Operation with ARTS/MELLLA" Hydrogen Water Chemistry Flow Increase" Standby Liquid Control (SLC) Boron Enrichment and Single Pump Operation RPV pressure will remain unchanged for the CPPU, and the maximum core flow will also remain unchanged.

The Feedwater/Condensate flow rates will be increased to support the CPPU.Despite the increase in flow, it is anticipated that the long term initiating event frequency will not change.The MELLLA refers to a region on the power/flow map where the plant will be licensed to operate at a higher rod line without increasing recirculation flow. The MELLLA curve is planned to be implemented for the CPPU power level to allow operational flexibility by restoring acceptable core flow range. It is not expected that operation with the revised power/flow map will significantly impact transient initiating event frequencies. In any event, no significant numerical difference in the PRA transient initiating event frequencies due to the MELLLA curve can be reasonably quantified at this time. However, sensitivity cases that increase transient frequencies are quantified in this risk assessment. 10-11 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -10.5.1 Initiating Events The evaluation of the plant and procedural changes indicates no new initiators or increased frequencies of existing initiators are anticipated to result from the CPPU.The SSES PRA initiating events can be categorized into the following: " Transients

• Loss of Offsite Power (LOOP)" Loss of Coolant Accidents (LOCAs)" Support System Failures" Internal Floods Additionally, external event initiators are also discussed for completeness.

Transients The evaluation of the CPPU plant and procedural changes do not result in any new transient initiators, nor is there anticipated any direct impact on transient initiator frequencies due to the CPPU (i.e., no changes are being made for the CPPU to the number of normally operating pumps and equipment in BOP systems).However, sensitivity quantification was performed that increases the Non-isolation transient initiator frequency to bound the various changes to the BOP side of the plant (e.g., main turbine modifications). Additionally, because it may not be possible for the units to remain online following the trip of a single main feed pump under CPPU conditions, this was also taken into account in the sensitivity case for the non-isolation transient scenario (refer to Tables 10-3 and 10-4).LOOP No change in the Loss of Offsite Power initiating event frequency is expected. Analysis indicated that the existing Off-Site Power System electrical equipment was determined to be adequate for operation with the CPPU related electrical output. The isolated phase bus duct was modified to accommodate the additional power output by cooling modifications. The 230kV and 500kV switchyard components including circuit breakers, disconnect switches, and current transformers are suitable to meet CPPU continuous current and short circuit current requirements,, following the replacement of the 230kV synchronizing breaker. The main transformers for Unit 2 rating have been upgraded and meet the requirements for CPPU operation. Based on this analysis, there is no significant impact on grid stability due to the CPPU.LOCAs No changes. to RPV operating pressure, inspection frequencies, or primary water chemistry are planned in support of the CPPU; as such, no impact on LOCA frequencies due to the CPPU can be postulated. However, acknowledging that increased flow rates of the CPPU can result in increased piping erosion/corrosion rates, a risk sensitivity case quantification is performed that increases the LOCA initiating event frequencies (Tables 10-3 and 10-4).10-12 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Support System Initiators No significant changes to support systems (e.g., AC, DC, Service Water, etc.) are planned in support of the CPPU; as such, no impact on support system initiating event frequencies due to the CPPU can be postulated. Internal Flood Initiators Because the methodology used in calculating the initiating event frequency for intemal flooding is based on the number of pipe segments found within a system and the fact that the geometry for the major piping is not changing, the internal flooding initiator frequencies remained the same.However, sensitivity cases were performed that included doubling the initiating event frequencies for main feedwater flooding to account for the effect of increased flow under CPPU conditions on piping erosion/corrosion rates. Because service water flow rates remain the same for CPPU, no sensitivity cases were necessary for flooding due to a service water rupture.External Event Initiators The frequency of external event initiators (e.g., seismic events, extreme winds, fires) is not linked to reactor power or operation; as such, no impact on external event initiator frequencies due to the CPPU can be postulated. 10.5.2 Component Reliability The majority of the hardware changes in support of the CPPU may be characterized as either:* Replacement of components with enhanced like components

• Upgrade of existing components Although equipment reliability as reflected in failure rates can be theoretically postulated to behave as a "bathtub" curve (i.e., the beginning and end of life phases being associated with higher failure rates than the steady-state period), no significant effect on the long-term average of initiating event frequencies, or equipment reliability during the 24 hour PRA mission time due to the replacement/modification of plant components is anticipated, nor is such a quantification supportable at this time. If any degradation were to occur as a result of CPPU implementation, existing plant monitoring programs would address any such issues. This assessment is consistent with [[ ]] on this issue (Reference 1),[[No planned operational modifications as part of the CPPU include operating equipment beyond design ratings. However, sensitivity cases that increase transient initiating event frequencies are quantified in this CPPU risk analysis to bound the various changes to the BOP side of the plant (refer to Tables 10-3 and 10-4 of this report).10-13 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -10.5.3 Operator Response The SSES risk profile, like other plants, is dependent on the operating crew actions for successful accident mitigation.

The success of these actions is in turn dependent on a number of.performance shaping factors. The performance shaping factor that is principally influenced by the power uprate is the time available within which to detect, diagnose, and perform required actions. The higher power level results in reduced times available for some actions.Modular Accident Analysis Program (MAAP) calculations for the CLTP and CPPU configurations were performed to determine changes in allowable operator action timings. All the post-initiator human error probabilities (HEPs) in the model were then re-calculated using the same HRA methods used in the SSES AUG06preEPU-C PRA. Refer to Table 10-5 for a summary of the changes in operator action timings and associated HEPs due to the CPPU. Table 10-6 includes the corresponding changes to the human reliability dependency analysis. Those HEPs that were not credited or that did not change in value, which amounted to approximately 100 independent and about 20 dependent events, were omitted from both tables.One additional action was added to the model related to manual operation of the new manual isolation valve in the spray pond return path that provides an alternate means of preventing the overheating of the ultimate heat sink given failure of the current spray pond bypass valve to close.To appropriately account for HEP dependencies, the same HEP and basic event that was already in the model for manual operation of valves in the same location at the same time was utilized for representation of the HEP associated with the new valve in the AUG06EPU PRA model.No significant changes are to be made to the Control Room for the CPPU that would impact the PRA human reliability analysis. Potential changes to be made to the Control Room displays for the CPPU are re-scaling certain indicators/recorders and/or replacement of certain indicators with digital units. None of these Control Room display changes will impact in any quantifiable way the human reliability analysis for the PRA.10.5.4 Success Criteria The success criteria for the AUG06preEPU-C PRA are derived based on realistic evaluations of system capability over the 24 hour mission time of the PRA analysis. These success criteria therefore may be different than the design basis assumptions used for licensing SSES. Risk profile changes caused by CPPU are examined from a realistic perspective to identify changes in the risk profile that may result from severe accidents on a best estimate basis. The following subsections discuss different aspects of the success criteria as used in the PRA. Both the CPPU task reports performed by General Electric and MAAP 4.0.5 runs performed for the CPPU risk assessment were used to assess impacts on success criteria.Timing Shorter times to boil down are likely on an absolute basis due to the increased power levels. The reduction in timings can impact the human error probability calculations, especially for short-term 10-14 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -operator actions. This has been directly factored into revised HEP values for CPPU conditions (See HRA discussion in Section 10.5.3).RPV Inventory Makeup Requirements The PRA success criteria for RPV makeup remain the same for the post-uprate configuration. Both high pressure, e.g., FW, high pressure coolant injection (HPCI), and reactor core isolation cooling (RCIC), and low pressure, e.g., low pressure coolant injection (LPCI), core spray (CS), and condensate, injection systems have more than adequate flow margin for the post-uprate configuration. RPV injection systems that were considered marginal, and therefore not credited, in the pre-uprate configuration, e.g., control rod drive (CRD) injection as an independent RPV makeup source during the initial stages of an accident are still deemed marginal and are not adequate in the post-uprate configuration. CRD remains a viable RPV makeup source at high and low pressures in the post-CPPU configuration following initial operation of another injection system, i.e., as a late injection source, for certain accidents. All success criteria have been verified with MAAP4.0.5 runs for both pre-CPPU and CPPU conditions. Heat Load to the Pool Energy to be absorbed by the pool during an isolation event or RPV depressurization increases for the CPPU case relative to the original license basis power level. For non-ATWS scenarios, the RHR heat exchangers, the main condenser, and the containment vent all have capacities that exceed the increase in heat load due to extended power uprating. The heat removal capability margins are sufficiently large such that the changes in power level associated with CPPU do not affect the success criteria for these systems. By design, the main condenser and RHR SPC systems are sufficiently sized for containment heat removal at the CPPU condition. With respect to containment venting in long term loss of decay heat removal scenarios, a MAAP run showed that the emergency containment vent is clearly sufficient for the CPPU conditions. No changes to the above decay heat removal (DHR) systems to augment their capabilities for the CPPU configuration are necessary or planned.Blowdown Loads Dynamic loads would increase slightly because of the increased stored thermal energy. This change would not quantitatively influence the PRA results.RPV Overpressure Margin The RPV dome operating pressure will not be increased as a result of the power uprate. However, the RPV pressure following a failure to SCRAM is expected to increase slightly, but the number of relief valves expected to lift is the same as would have been expected under CLTP conditions. For turbine trip events (i.e., non-isolation, non-ATWS), the SRVs likely will not be challenged for overpressure control. The number of SRVs assumed required for RPV overpressure protection in the AUG06PreEPU-C PRA for non-ATWS isolation scenarios is two, i.e., 15 out of 16 SRVs would have to fail to open mechanically to result in over pressurization and failure of the reactor coolant 10-15 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -system (RCS). Because common cause failure data is not available for group sizes larger than 8, only the global, i.e., all SRVs fail due to common cause, is utilized in the AUG06preEPU-C PRA model. Therefore, even if one were to conservatively assume that two more SRVs would be required for the CPPU power level, there would be no change to the common cause failure contribution and any increase in the independent failure contributions to risk (not modeled) would be extremely negligible. For ATWS scenarios, the NSSS vendor, General Electric, confirms that the condition of having four SRVs out of service is acceptable during an isolation ATWS to prevent over pressurizing the RPV for the CPPU case. This is consistent with the AUG06preEPU-C model such that no change to the common cause failure contribution for*ATWS scenarios is required.The AUG06preEPU-C SSES PRA does not require any SRVs for initial RPV overpressure control for LOCA initiators. This success criterion also remains unchanged for the CPPU.As such, no model changes to the SSES PRA regarding this function are required for this CPPU risk assessment. SRV Actuations The SRV setpoints have not been changed as a result of the CPPU. Given the power increase of the CPPU, one may postulate that the probability of a stuck-open relief valve (SORV) given a transient initiator would increase due to an increase in the number of SRV cycles.The AUG06preEPU-C PRA base SORV probability may be modified using different approaches to consider the effect of a postulated increase in valve cycles. The following three approaches are considered:

1. The upper bound approach would be to increase the SORV probability by a factor equal to the increase in reactor power (i.e., a factor of 1.13 in the case of the CPPU. This approach assumes that the SORV probability is linearly related to the number of SRV cycles, and that the number of cycles is linearly related to the reactor power increase.2. A less conservative approach to the upper bound approach would be to assume that the SORV probability is linearly related to the number of SRV cycles, but the number of cycles is not necessarily directly related to the reactor power increase.

In this case the postulated increase in SRV cycles due to the CPPU would be determined by thermal hydraulic calculations. (e.g., MAAP runs).3. The lower bound approach would be to assume that the SORV probability is dominated by the initial cycle and that subsequent cycles have a much lower failure rate. In this approach the base SORV probability could be assumed to be insignificantly changed by a postulated increase in the number of SRV cycles.Approach #1 is used :to define the SSES AUG06EPU PRA SORV probability as shown in Table 10-7. The SORV probability basic event in the SSES PRA is increased 13.3% for the CPPU base case risk evaluation. 10-16 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -RPV Emergency Depressurization The current AUG06preEPU-C PRA requires three SRVs for RPV emergency depressurization. MAAP cases performed in support of this CPPU risk assessment show that this success criterion remains unchanged by the CPPU. Therefore, the AUG06EPU PRA success criterion of 3 SRVs is maintained in this analysis.Structural Evaluations This assessment did not identify issues associated with postulated impacts from the CPPU on the PRA modeling of structural (e.g., piping, vessel, containment) capacities. This is consistent with]] (Reference 3): Success Criteria Summary The Integrated Level I and Level 2 PRA developed success criteria for each branch of the event trees. The PRA success criteria are affected by the increased boil off rate, the increased heat load to the suppression pool, and the increase in containment pressure and temperatures. MAAP runs demonstrate the significant margins associated with the installed systems. However, MAAP runs did indicate the impact of CPPU on timing. The impact of these timing changes is then reflected in the available LOOP recovery times for each sequence and in the Human Error probabilities developed for the AUG06EPU PRA model. It can be noted that all emergency diesel generator failures, whether run or start failures, are assumed to occur at the start of the event, i.e., at t= 0. Therefore, the available time for LOOP recovery has a first order impact on the PRA model results. The changes to the available times for LOOP recovery due to CPPU conditions have been directly factored into the AUG06EPU PRA model.Besides the timing issues described above, no other changes in the modeled success criteria have been identified for the Level 1 or Level 2 PRA.Accident Sequence Modeling The CPPU does not change the plant configuration or operation in a manner such that new accident sequences or changes to existing accident scenario progressions result. A slight exception is the reduction in available accident progression timing for some scenarios and the associated impact on operator action HEPs (this aspect is addressed in the Human Reliability Analysis section).This assessment is [[ on this issue (Reference 3): 10-17 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -[[demonstrate that the accident progression is basically unchanged by the uprate." Level 2 PRA Analysis Fission product inventory in the reactor core is higher as a result of the increase in power due to the CPPU. The increase in fission product inventory results in an increase in the total radioactivity available for release given a severe accident. The total activity available for release is approximately 13% higher. However, this does not impact the definition or quantification of the LERF risk measure used in Regulatory Guide 1.174, and as the basis for this risk assessment. The SSES PRA release categories are defined based on the percentage (as a function of EOC inventories) of CsI released to the environment, this is consistent with most industry PRAs.Given the minor change in Level 1 results, minor changes in the Level 2 release frequencies can be anticipated. Such changes are directly attributable to the changes described previously and the minor changes in short term accident sequence timing and the impact on HEPs. The structure of the accident sequence modeling in the Level 2 PRA is not impacted by the CPPU. However, there were some changes in the release categorization in some cases where a pre-CPPU Late Release is now classified as an Intermediate Release or an Intermediate Release is now classified as and Early release. These changes in release category assignments are based on MAAP4.0.5 calculations for pre-CPPU and CPPU conditions that resulted in reductions in the calculated times between the declaration of a General Emergency and the time of first fission product release to the environment. Radionuclide Release (Level 2 PRA)The Level 2 PRA calculates the containment response under postulated severe accident conditions and provides an assessment of the containment adequacy. In the process of modeling severe accidents (i.e., the MAAP code), the complex plant structure has been reduced to a simplified mathematical model that uses basic thermal hydraulic principles and experimentally derived correlations to calculate the radionuclide release timing and magnitude. Changes in plant response due to CPPU represent relatively small changes to the overall challenge to containment under severe accident conditions. Approximately 200 Level 1 and Level 2 MAAP runs were performed in support of the CPPU risk assessment. The Level 2 MAAP runs were focused on the assessment of any significant changes in release categories. No changes to the SSES PRA Level 2 accident progression logic modeling or release magnitude assignment were judged necessary for CPPU. However, as stated previously, there were some sequences that resulted in reassignment of the release timing category due to the changes in the time between General Emergency declaration and initial fission product release based on the increased decay heat levels associated with CPPU conditions. These changes are delineated below in the Release Magnitude and Timing section.The following aspects of the Level 2 analysis are briefly discussed:

• Level I input* Accident Progression 10-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -* Human Reliability Analysis* Success Criteria Containment Capability
• Radionuclide Release Magnitude and Timing Level I Input The front-end evaluation (Level 1) involves the assessment of those scenarios that could lead to core damage. The subsequent treatment of mitigating actions and the inter-relationship with the containment after core damage (Level 2) is then treated in subsequent nodes within the integrated Event Trees.In the SSES Level 1 PRA, accident sequences are postulated that lead to core damage and potentially challenge containment.

The SSES Level 1 PRA has identified discrete accident sequences that contribute to the core damage frequency and represent the spectrum of possible challenges to containment. The Level 1 core damage sequences are also directly propagated through the Level 2 portion of the integrated event trees. Therefore, changes to the Level 1 PRA modeling directly impact the Level 2 PRA results. However, the percentage increase in total CDF due to the CPPU is not a direct translation to the percentage increase in total LERF. For example, a change to long-term core damage accidents would not impact the LERF results because these accidents do not result in Level 2 LERF sequences. Therefore, the Level 2 at-power internal events PRA model is also re-quantified as part of this CPPU risk assessment. Accident Progression The CPPU does not change the plant configuration and operation in a manner that produces new accident sequences or changes accident sequence progression phenomenon. This is particularly true in the case of the Level 2 post-core damage accident progression phenomena. The changes in decay heat levels has either a minor or no impact on Level 2 PRA safety functions, such as containment isolation, ex-vessel debris coolability and challenges to the ultimate containment strength. No Level 2 safety function success criteria would be changed due to the CPPU (although the timing requirements may be shifted somewhat). Regarding energetic phenomena occurring at or near the time of core slump or RPV breach, such accident progression scenarios are appropriately modeled in the SSES Level 2 PRA as leading directly to High magnitude releases. This is a reasonable and standard PRA industry approach.- This approach would not be changed due to the CPPU.Therefore, no changes are made as part of this assessment to the Level 2 models (either in structure or basic event phenomenon probabilities) with respect to accident progression modeling.10-19 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Human Reliability Analysis Because the SSES PRA employs a fully integrated Level I and Level 2 PRA model, changes to HEP values (refer to Section 10.5.3) have a direct effect on both the Level 1 and Level 2 results. In other words, changing HEP can affect the outcome of core damage, which then provides the input to the sequences responsible for calculating release categories. It should be noted, however, that in both the pre-CPPU and CPPU PRA models, local valve manipulations following core damage are not credited.Success Criteria No changes in success criteria have been identified with regard to the Level 2 containment evaluation. The slight changes in accident progression timing and decay heat load has a minor or negligible impact on Level 2 PRA safety functions, such as containment isolation, ex-vessel debris coolability and challenges to the ultimate containment strength. Therefore, no changes to Level 2 modeling with respect to success criteria are made as part of this analysis.Containment Capability The containment capacity with respect to severe accidents is analyzed in the PRA using plant specific structural analyses as well as information from industry studies and experiments. The minor changes to the plant from the CPPU have no impact on the definition of these containment loading profiles or the likelihood of containment isolation failure. The slightly higher decay heat levels associated with the CPPU will result in minor reductions in times to reach loading challenges; however, the time frames are long (many hours) and the accident timing reductions of 10-15% due to the CPPU will typically have a small impact on the Level 2 results. However, the timing changes are accounted for in the model with the revised LOOP recovery times that are available for each sequence.For example, MAAP evaluations were performed in support of this analysis that showed that the time to reach the containment venting pressure (i.e., 65 psig) for a loss of all decay heat removal scenario is 28 hours for the pre-CPPU condition, and this time drops to 23.2 hours for the CPPU condition. This timing difference has minimal impact on the Level 2 results, but is used to define the release categories and the LOOP recovery times that are available for each sequence.Release Mag-nitude and Timing The following issues can substantially increase or. decrease the ability to retain fission products or mitigate their release:* Radionuclide removal processes* Containment failure modes* Phenomenology

• Accident sequence timings Each of these issues is considered and analyzed in the SSES Level 2 PRA.10-20 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The SSES Level 2 PRA release categorization scheme uses both release magnitude and timing.Release categories were assigned to the SSES AUG06preEPU-C PRA based on results of.representative MAAP runs for many accident scenarios, and based on judgment and standard industry approaches for selected scenarios.

The SSES release magnitude classification is based on the percentage (as a function of the initial EOC inventory in the core) of CsI released to the environment; this approach is consistent with the majority of US BWR PRAs and standard industry techniques (Reference 35 and 36). Changing the release magnitude categories assigned to individual accident sequences in the SSES Level 2 PRA is not necessary, which was confirmed by MAAP runs. However, there were a few cases that were assigned to different release timing categories. These changes are summarized in Table 10-8.Level 2 Impact Summary Based on the above discussion, the impact of the CPPU on the SSES Level 2 PRA results, independent of the Level I analysis, is judged to be small. The change in the Level 2 is due primarily to changes in the Level 1 accident sequences propagated through to the Level 2 quantification. That is, an increase in a Level 1 accident sequence gave rise to a proportional increase in the Level 2 result that was associated with that core damage state, i.e., the Level 2 results are coupled to the Level I results.10.5.5 External Events Internal Fires The plant risk due to internal fires was evaluated in 1994 as part of the SSES IPEEE Submittal (Reference 37). The results were amended based on the NRC audit of the IPEEE. PPL Susquehanna document PLA-4983 (Reference

38) summarizes the results of the audit on the fire analysis in addition to the updated conclusions of the seismic analysis.The SSES fire analysis was performed using the methodology prescribed in the PRA Procedures Guide (Reference 39), which produced results similar to those yielded by the internal events analysis.

While the fire analysis did yield a CDF, the intent of the analysis was to identify the most risk significant fire areas in the plant using a screening process and by calculating conservative core damage frequencies for fire scenarios. As such, the accident sequence frequencies calculated for the SSES fire PRA are not a best estimate calculation of plant fire risk and are not acceptable for integration with the best estimate internal events PRA results for comparison with Regulatory Guide 1.174 acceptance guidelines. The screening attributes of the fire PRA are summarized below.Attributes of Fire PRA Fire PRAs are useful tools to identify design or procedural items that could be clear areas of focus for improving the safety of the plant. Fire PRAs use a structure and quantification technique similar to that used in the internal events PRA.Historically, because less attention has been paid to fire PRAs, conservative modeling is common in a number of areas of the fire analysis to provide a "bounding" methodology for fires. This 10-21 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -concept is contrary to the base internal events PRA that has had more analytical development and is judged to be closer to a realistic assessment (i.e., not conservative) of the plant.There are a number of fire PRA topics involving technical inputs, data, and modeling that prevent the effective comparison of the calculated core damage frequency figure of merit between the internal events PRA and the fire PRA. These areas are identified as follows: Initiating Events: The frequency of fires and their severity are generally conservatively overestimated. A revised NRC fire events database indicates the trend toward both lower frequency and less severe fires. This trend reflects the improved housekeeping, reduction in transient fire hazards, and other improved fire protection steps at nuclear utilities. The database used in the fire assessment used significantly older data that is not judged applicable. In addition, it reflects conservative judgments regarding fire severity.System Response: Fire protection measures such a sprinklers, CO 2 , fire brigades may be given minimal (conservative) credit in their ability to limit the spread of a fire. Therefore, the severity of the fire and its impact on requirements is exacerbated. In addition, cable routings are typically characterized conservatively because of the lack of data regarding the routing of cables or the lack of the analytic modeling to represent the different routings. This leads to limited credit for balance of plant systems that are extremely important in CDF mitigation. Sequences: Sequences may subsume a number of fire scenarios to reduce the analytic burden.The subsuming of initiators and sequences is done to envelope those sequences included. This causes additional conservatism. Fire Modeling: Fire damage and fire propagation are conservatively characterized. Fire modeling presents bounding approaches regarding the fire immediate effects (e.g., all cables in a tray are always failed for a cable tray fire) and fire propagation. HRA: There is little industry experience with crew-actions under conditions of the types of fires modeled in fire PRAs. This has typically led to conservative characterization of crew actions in fire PRAs. Because the CDF is strongly correlated with crew actions, this conservatism may have a profound influence on the calculated fire PRA results.Quality of Model: The peer review process for fire PRAs is less well developed than for internal events PRAs. For example, no industry standard, such as NEI 00-02, exists for the structured peer review of a fire PRA. This may lead to less assurance of the realism of the model.10-22 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The fire PRA is subject to more modeling uncertainty than the internal events PRA evaluations. While the fire PRA is generally self-consistent within its calculational framework, the fire PRA calculated quantitative risk metric does not compare well with internal events PRAs because of the number of conservatisms that have been included in the fire PRA process. Therefore, the use of the fire PRA figure of merit as a reflection of CDF may be inappropriate. Any use of fire PRA results and insights should properly reflect consideration of the fact that the "state of the technology" in fire PRAs is less evolved than the internal events PRA.Relative modeling uncertainty is expected to narrow substantially in the future as more experience is gained in the development and implementation of methods and techniques for modeling fire accident progression and the underlying data.CPPU Impact on Fire Risk A qualitative impact on the fire risk profile due to the CPPU is estimated here based on review of the SSES IPEEE fire PRA results. This estimate is performed as follows: " The CDFs from the quantified Fire Zones of the fire analysis are used to represent the fire risk profile (based on the results of the IPEEE audit (Reference 40)." The quantitative results of these fire scenarios are limited to fire induced transients. Fire induced LOCAs and ATWS events were considered to be sufficiently unlikely so as not to be considered in the IPEEE. Therefore, the CDF for each Fire Zone is assumed to be impacted by CPPU implementation in a manner similar to the non-isolation, isolation and LOOP initiator contributions to CDF from the internal events model. The percent increase for these initiators is 29% between pre-CPPU and CPPU conditions (see Table 10-12), which is considered to be a bounding estimate of the impact on CDF due to fire events under CPPU conditions." The results of the CPPU base case quantification (in terms of percentage CDF increase as a function of initiating event type) are applied to the Fire Zones.The fire impact calculation estimate is summarized in Table 10-9, which shows that the fire PRA CDF would increase by 9.9E-09 due to the CPPU.Seismic Risk The seismic risk analysis was performed as part of the IPEEE (Reference 37). SSES performed a seismic margins assessment (SMA) following the guidance of EPRI NP-6041 (Reference 41).The SMA is a deterministic evaluation process that does not calculate risk on a probabilistic basis. No core damage frequency sequences were quantified as part of the seismic risk evaluation.

a. The final results of the seismic analysis are documented in the response to audit issues on the IPEEE submittal (Reference 40). While many of the same results were provided in the IPEEE submittal, the SSES Response to Audit Issues provides a more complete description of the actions taken by PPL Susquehanna to close out the seismic related issues at the site.10-23 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Based on the efforts to correct the seismic issues that were identified as part of the IPEEE program and the ongoing process to monitor seismic issues at the plant, no additional measures are considered to be required based on the implementation of CPPU. The CPPU has little or no impact on the seismic qualifications of the systems, structures and components (SSCs).Specifically, the power uprate results in additional thermal energy stored in the RPV, but the additional blowdown loads on the RPV and containment given a coincident seismic event are judged not to alter the results of the SMA.The decrease in time available for operator actions, and the associated increases in calculated HEPs, is judged to have a non-significant impact on seismic-induced risk. Industry BWR seismic PRAs have typically shown (e.g., Peach Bottom NUREG-1 150 study, as detailed in NUREG/CR-4550 (Reference 42); Limerick Generating Station Severe Accident Risk Assessment (Reference 43); and NUREG/CR-4448 (Reference

44) that seismic risk is overwhelmingly dominated by seismic induced equipment and structural failures.Based on the above discussion it is judged that the percentage increase in the seismic risk due to the CPPU is much less than that calculated for internal events.Other External Events Risk In addition to internal fires and seismic events, the SSES IPEEE Submittal analyzed a variety of other external hazards:* High Winds/Tornadoes" External Floods" Transportation and Nearby Facility Accidents The SSES IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, and nearby facility accidents was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards. Based upon this review, it was concluded that SSES meets the applicable NRC Standard Review Plan requirements and therefore has an acceptably low risk with respect to these hazards."Other" external event hazards, such as volcanic activity, were not analyzed in detail in the SSES IPEEE because these types of events were determined not to be applicable to SSES or because they were included in other analyses (IPE or Station Blackout analysis).

Note that the SSES IPEEE also analyzed internal flooding scenarios. The internal flooding scenarios are incorporated into the current PRA internal events model of record such that the impacts on internal flooding accident sequences are addressed quantitatively in this CPPU risk assessment as part of the internal events risk impact.10.5.6 Shutdown Risk The impact of the CPPU on shutdown risk is similar to the impact on the at-power Level 1 PRA.Based on the insights of the at-power PRA impact assessment, the areas of review appropriate to shutdown risk are the following:

• Initiating Events* Success Criteria 10-24 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -* Human Reliability Analysis The following qualitative discussion applies to the shutdown conditions of Hot Shutdown (Mode 3), Cold Shutdown (Mode 4), and Refueling (Mode 5). The CPPU risk impact during the transitional periods such as at-power (Mode 1) to Hot Shutdown and Startup (Mode 2) to at-power is judged to be subsumed by the at-power Level 1 PRA. This is consistent with the U.S.PRA industry, and with NRC Regulatory Guide 1.174 which states that not all aspects of risk need to be addressed for every application.

While higher conditional risk states may be postulated during these transition periods, the short time frames involved produce an insignificant impact on the long-term annualized plant risk profile.Shutdown Initiating Events Shutdown initiating events include the following major categories:

• Loss of RCS Inventory-Inadvertent Drain down-LOCAs* Loss of Decay Heat Removal (includes LOOP)No new initiating events or increased potential for initiating events during shutdown (e.g., loss of DHR train) can be postulated due to the CPPU.Shutdown Success Criteria The impact of the CPPU on the success criteria during shutdown is similar to the Level I PRA.The increased power level decreases the time to boildown.

However, because the reactor is already shutdown, the boildown times are much longer compared to the at-power PRA. The time to uncover the core with the existing power level (CLTP) is estimated to be 9.1 hours (8.0 hours for the CPPU) at one day into the outage with the RPV level at the flange. The time exceeds 24 hours after one day into the outage when the water level is flooded up into the refueling cavity.The increased decay heat loads associated with the CPPU impacts the time when low capacity DHR systems can be considered successful alternate DHR systems. The CPPU condition delays the time after shutdown when low capacity DHR systems may be used as an alternative to Shutdown Cooling (SDC). However, this reduction in time for alternate decay heat removal system success minimally impacts shutdown risk.Other success criteria are marginally impacted by the CPPU. The CPPU has a minor impact on shutdown RPV inventory makeup during loss of decay heat removal scenarios in shutdown because of the low decay heat level. The heat load to the suppression pool during loss of decay heat removal scenarios in shutdown (i.e., during shutdown phases with the RPV intact) is also lower because of the low decay heat level such that the margins for suppression pool cooling capacity are adequate for the CPPU condition. The CPPU impact on the success criteria for blowdown loads, RPV overpressure margin, and SRV actuation is estimated to be negligible because of the low RPV pressure and low decay heat level during shutdown.10-25 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Shutdown HRA Impact The primary impact of the CPPU on risk during shutdown operations is the decrease in allowable operator action times in responding to off-normal events. The reduction in times to core damage (i.e., CLTP case compared to CPPU case) is on the order of 10%. Such small changesin already lengthy allowable operator response times result in negligible changes (<<1%) in calculated human error probabilities. The allowable operator action times to respond to loss of heat removal scenarios during shutdown operations are many hours long. Very early in an outage, the times are approximately 5-10 hours;later in an outage the times are dozens of hours. As a result of CPPU conditions, time to core damage due to boiloff is estimated to be reduced from 10.3 hours to 9.2 hours within 1 day after shutdown, which would not result in a significant increase in human error probabilities for most operator actions using current human reliability analysis methods. The allowable timing reductions for times later in the outage would result in indiscernible changes in HEPs using current human reliability analysis methods.Shutdown Risk Summary Based on a review of the potential impacts on initiating events, success criteria, and HRA, the CPPU is assessed to have a non-significant impact (delta CDF of roughly.one percent) on shutdown risk.This assessment is consistent with [[ ]] on this issue (Reference 1):[[I SSES Outage Risk Management Process The plant uses a computerized risk monitor and site -specific management guidelines as tools for controlling outage risk. The impact of the outage activities upon key safety functions is assessed as follows:* Identify key safety functions affected by the SSC planned for removal from service." Consider the degree to which removing the SSC from service will impact the key safety functions." Consider degree of redundancy, duration of out-of-service condition, and appropriate compensatory measures, contingencies, or protective actions that could be taken if appropriate for the activity under consideration. The Key Safety Function Matrices were developed consistent with guidance provided by NUMARC 91-06. The shutdown key safety functions are achieved by using systems or combinations of systems. The scope of the SSCs to be addressed by the assessment for shutdown conditions are those SSCs necessary to support the following shutdown key safety functions (from Section 4 of NUMARC 91-06):* Decay heat removal capability 10-26 . Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -* Inventory Control" Power Availability

• Reactivity control" Containment (primary/secondary)

Managing the risk involves invoking some or all of the following elements:.

• Pre-job briefs of operating and maintenance crews* System engineering oversight* Management oversight* Outage management approval of the proposed activity* Pre-staged parts and materials* Walk down of tag outs and maintenance activity prior to conducting the maintenance
• Mockup training" Reduce OOS time through overtime or additional shift coverage." Contingency plans for returning equipment to service in a timely manner if needed." Compensatory measures to minimize initiators and/or mitigate the consequences.
• Reschedule or minimize work on functionally related equipment.
• Proceduralize other success paths of the safety function affected.10.5.7 PRA Quality The quality of the SSES AUG06preEPU-C PRA model that is used as the starting point in performing the risk assessment for the CPPU is manifested by the following:
• Sufficient scope and level of detail in PRA" Active maintenance of the PRA models and inputs 0 Comprehensive Critical Reviews Scope and Level of Detail The SSES AUG06preEPU-C PRA is of sufficient quality and scope to measure the potential changes in plant risk related to CPPU implementation.

The SSES PRA modeling is highly detailed, including a wide variety of initiating events (e.g., transients, internal floods, LOCAs inside and outside containment, support system failure initiators), modeled systems, extensive level of detail, operator actions, and common cause events.The External Events models were developed with a goal of identifying potential plant vulnerabilities with respect to a specific set of initiating events caused by events outside of the normal power generating activities of the plant. The level of detail of these models varies by initiator type from the very detailed to screening approaches. Consistent with the original intent of the IPEEE process, these models can be used to determine if CPPU will impact the major contributors to risk from external events. The IPEEE was reviewed by the NRC and updated in order to resolve some of the questions developed as part of that review. No changes have been made to the IPEEE since that time.10-27 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -SSES does not maintain a shutdown PRA model. However, insights from other available industry studies were utilized to allow for quantitative comparisons of the likelihood of boiling and fuel damage scenarios based on equipment availability, reliability, and decay heat levels.The magnitude of the changes to shutdown risk resulting from CPPU was estimated by examining how the corresponding increased heat load and equipment changes would impact the risk profile at SSES. Therefore, the impact on shutdown risk based on CPPU conditions is based on more generic shutdown insights and assumptions obtained from a review of other industry BWR shutdown PRA results.Maintenance of Model, Inputs, Documentation The Level 1 and Level 2 SSES PRA analyses were originally submitted to the NRC in December 1991 as the SSES IPE Submittal. PPL Susquehanna's IPE received an NRC safety evaluation report (SER) in 1998. Since the time the IPE was submitted, there have been several extensive revisions produced prior to the Boiling Water Reactor Owners Group (BWROG) Peer Review in 2003. The model that underwent the Peer Review was not an upgrade to the IPE, but a new model based on thermal hydraulic calculations for the current fuel type and current rated power.New event trees were developed based on the calculated accident progression and current emergency operating procedures (EOPs). Subsequent to the BWROG Peer Review, the SSES PRA model was updated to address the comments generated from that review.Critical Reviews The significant, recent reviews of the SSES PRA model include the NRC activities related to the development of the SSES IPE SER and the 2003 BWROG Peer Review. The major findings of these reviews are summarized below.PPL Susquehanna's IPE was submitted to the NRC and received an SER on August 11, 1998.There were three weaknesses identified in the SER, which were related to the following issues: 9 The evaluation of sequences with containment failure prior to core damage ended with the assumption of core damage and did not analyze the consequences of these sequences,* The impact on conditional containment failure probability of some severe accident phenomena and resulting containment failure modes appeared to have been understated,* The treatment of Interfacing System LOCA (ISLOCA) was not as robust as required.These issues were addressed and corresponding changes were incorporated into the PRA prior to the BWROG Peer Review, which was performed in 2003.The consensus of the Peer Review team, as stated in the exit meeting, was that the SSES PRA was "top quartile" in the industry. The BWROG peer review provided PPL Susquehanna with Level B, C, D and S Facts and Observations (F&Os) as defined in Table 10-10. PPL Susquehanna did not receive any Level A F&Os.PPL incorporated approximately half of the B level F&Os and some of the C Level F&Os into the FEB05RA model. PPL performed a self-assessment using the guidance included in RG 1.200 (Reference

12) that supplements NEI 00-02 (Reference 35). This review indicated the necessity to address the remaining

'B' open items. Other identified 'Gaps' to Capability Category II of the 10-28 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -ASME PRA Standard (Reference 48 were judged to not have an impact on the CPPU evaluation. The remaining open B level comments were reviewed to determine if any outstanding F&Os had the potential to significantly impact the CPPU results. The result of the review is Summarized in Table 10-11. All issues that were identified as important for resolution in the model prior to performing the CPPU application were resolved in the AUG06preEPU-C and AUG06EPU models. The remaining items and Gaps would not have a significant impact on the CPPU application and were therefore deferred until the next update. Sensitivity studies were performed where warranted, however, as described in Section 10.5-2 of this report.Summary In summary, it is found thatý the integrated Level I and Level 2 PRA model provides the necessary scope and level of detail to allow the calculation of CDF and radioactive release frequency changes due to the CPPU. The External Events models will allow for a review of the largest contributors to External Events risk and how they might be impacted by CPPU. The information from generic shutdown PRA results will provide the capability to determine the magnitude of the changes to plant shutdown risk that would occur based on CPPU implementation. 10.5.8 Integrated Risk Impact Based on the model impact discussed previously, the CPPU is estimated to increase the unit I SSES internal events PRA CDF from the base value of 1.65E-06/yr to 1.71 E-06/yr, an increase of 6.OE-08/yr (4.0%). For unit 2, the CPPU is estimated to increase the internal events PRA CDF from the base value of 1.63E-06 to 1.70E-06, an increase of 7.OE-08/yr (4.0%). The composition and comparative distribution of the CPPU results remain unchanged with respect to the base SSES PRA. Table 10-12 shows quantitative CDF comparisons categorized by initiating event and Table 10-13 compares various accident sequences. The at-power internal events LERF increased from the base value of 1.75E-07/yr to 1.76E-07/yr, an increase of 1.OE-09/yr (0.6%) for both units 1 and 2.Shutdown risk and external event risk was also evaluated and determined to be impacted to a similar or lesser degree than the internal events.Quantitative Sensitivity Cases In addition to the base (best estimate) CPPU quantification, five quantitative sensitivity cases were performed. These cases are summarized in Tables 10-3 and 10-4). These cases address the key modeling assumptions and areas of uncertainty used in this risk evaluation. The sensitivity cases are reasonable in nature and focus on aspects that would increase the calculated risk. As can be seen from Tables 10-3 and 10-4, these quantitative sensitivities do not alter the conclusion that the CPPU has a small impact on the plant risk profile.10-29 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Conclusions The key result of the SSES CPPU risk evaluation is the following: Small risk increases were calculated for both CDF and LERF. The risk increase is associated with reduced times available for certain operator actions, LOOP recovery times, and the assumed increase in the SORV probability. Using the NRC guidelines established in Regulatory Guide 1.174 and the calculated results from the Level 1 PRA, the best estimate for the CDF risk increase due to the CPPU (6E-08/yr for Unit 1 and 7E-08 for Unit 2) is in the lower left comer of Region III (i.e., very small risk changes). The best estimate for the LERF increase (I.OE-09/yr for both units 1 and 2) is also in the lower left comer of the Region III range of RG 1.174.The CPPU is assessed to result in a very small impact on the plant risk profile and is, therefore, acceptable from a risk evaluation perspective. 10.6 OPERATOR TRAINING AND HUMAN FACTORS Some additional training is required to enable plant operation at the CPPU RTP level. The topics addressed in this section are: Topic CLTR Disposition SSES Result Operator training and human factors ]For CPPU conditions, operator responses to transients, accidents, and special events are minimally affected. Most abnormal events result in automatic plant shutdown (scram). Some abnormal events result in automatic RCPB pressure relief, ADS actuation and/or automatic ECCS actuation (for low water level events). All events result in safety-related systems, structures, and components (SSCs) remaining within their design allowable values. CPPU does not change any of the automatic safety functions, except for those setpoint changes identified in Section 5.3. After the applicable automatic responses have initiated, the subsequent operator actions for plant safety (e.g., maintaining safe shutdown, core cooling, and containment cooling)do have some changes for CPPU. For example, a modification for the Ultimate Heat Sink (Spray Pond) will install manual valves that would be closed in the event of a failure of the motor operated bypass header valves, to ensure dissipation of decay heat to the atmosphere. Also, Appendix "R" modifications are being designed to increase the operational options and equipment capabilities for containment cooling.Susquehanna Unit 1 and Unit 2 are of similar design, but are operated, maintained, and modified on an individual unit basis. As such, the units will differ in rated power, steam, and feedwater flow rates until both units have completed the full CPPU uprating (a period of 3 years). The units are uprated in 2 phases of'-7% each, associated with 2 operating cycles. The Operating Procedures, Off-Normal Procedures, and Emergency Operating Procedures are unit specific and reflect the parameters and limits associated with the individual unit and phase of uprate.10-30 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -The analog and digital inputs for the Plant Integrated Computer System (PICSY) and Safety Parameter Display System (SPDS) are reviewed to determine the effect of CPPU. This includes required changes to monitored points, calculations, alert and trip setpoints. Changes in Emergency Operating Procedure (EOP) curves and limits require an update of the SPDS.Modifications required for CPPU are evaluated for effect on the PICSY and SPDS computer and any required changes (including any new monitoring points) are addressed as a part of the modification. Any changes required to the PICSY and SPDS computer are completed prior to CPPU implementation. Physical equipment changes required by the CPPU are implemented by a modification package.The modification package initiates the physical equipment changes, procedure changes, and required training.Following a review of the CPPU modifications and identified key procedure changes, the Systematic Approach to Training determines the scope of operator training presented on modifications and parameter changes for CPPU.Training on design changes completes during the last training cycle (training week) prior to an upcoming refueling outage. Larger scope modifications are presented over several training cycles to ensure thorough knowledge and comprehension of the changes by the operators. Lesson plans are developed, and operator classroom and simulator training is performed prior to restart of the unit from the outage implementing the CPPU modifications. This training covers plant modifications, procedure changes, startup test procedures, and other aspects of CPPU including changes to parameters, setpoints, scales, systems, PICSY and SPDS displays. Existing lesson plans are revised to reflect changes as a result of the CPPU.Operator training for CPPU conditions is performed on the simulator prior to operating the unit at CPPU conditions. The training cycle prior to CPPU implementation completes the recommended operator classroom and simulator training for CPPU implementation. This training includes the normal operating procedure actions required to achieve the CPPU RTP level, power ascension testing being performed, and comparisons of plant conditions between the current RTP level and the CPPU RTP level. Data obtained during startup testing is incorporated into training material as needed.Installation of the CPPU changes in the simulator is accomplished on a demonstration load for training purposes. Final changes are incorporated after actual installation in the plant. The simulator changes include hardware changes for new or modified control room instrumentation and controls, software updates for modeling changes due to CPPU (i.e., reactor feed pump and condensate pump performance upgrades, HP turbine modifications), setpoint changes, and re-tuning of the core physics model for cycle specific data.Operating data is collected during CPPU implementation and start-up testing. This data is compared to simulator data as required by ANSI/ANS 3.5-1985, Section 5.4.1. Simulator acceptance testing is conducted to benchmark the simulator performance based on design and engineering analysis data as required in ANSI/ANS 3.5-1985.10-31 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -10.7 PLANT LIFE The plant life evaluation identifies degradation mechanisms influenced by increases in fluence and flow. The topics addressed in this evaluation are: Topic CLTR Disposition SSES Result Irradiated Assisted Stress Corrosion Cracking Flow Accelerated Corrosion SSES has a procedurally controlled program for the augmented nondestructive examination (NDE) of selected RPV internal components in order to ensure their continued structural integrity. The inspection techniques utilized are primarily for the detection and characterization of service-induced, surface-connected planar discontinuities, such as intergranular stress corrosion cracking (IGSCC) and irradiation-assisted stress corrosion cracking (IASCC), in welds and in the adjacent base material. SSES belongs to the BWR Vessel and Internals Project (BWRVIP) organization and implementation of the procedurally controlled program is consistent with the BWRVIP issued documents. The inspection strategies recommended by the BWRVIP consider the effects of fluence on applicable components and are based on component configuration and field experience. Components selected for inspection include those that are identified as susceptible to in-service degradation and augmented examination is conducted for verification of structural integrity. These components have been identified through the review of NRC Inspection and Enforcement Bulletins (IEBs), BWRVIP documents, and recommendations provided by General Electric Service Information Letters (GE SILs). The inspection program provides performance frequency for NDE and associated acceptance criteria. Components inspected include the following:

• Core spray piping* Core spray spargers* Core shroud and core shroud support* Jet pumps and associated components
• Top guide* Lower plenum* Vessel ID attachment welds* Instrumentation penetrations
• Steam dryer* Feedwater spargers Continued implementation of the current procedure program assures the prompt identification of any degradation of reactor vessel internal components experienced during CPPU operating 10-32 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -conditions.

To mitigate the potential for IGSCC and IASCC, SSES utilizes hydrogen water chemistry. Reactor vessel water chemistry conditions are also maintained consistent with the Electric Power Research Institute (EPRI) and established industry guidelines, except where technical justification in accordance with BWRVIP-1 30 have been documented. The service life of most equipment is not affected by CPPU. [[]] The current inspection strategy for the reactor internal components is expected to be adequate to manage any potential effects of CPPU.The procedurally controlled piping flow-accelerated corrosion (FAC) program uses selective component inspections to provide a measure of confidence in the condition of systems susceptible to FAC. These selective inspections are the basis for qualifying un-inspected components for further service. This approach is based upon program guidelines developed by the Electric Power Research Institute (EPRI) in NSAC-202L,R2 "Recommendations for an effective Flow Accelerated Corrosion Program". The criteria for selecting components for inspection after the CPPU will be the same as used for CLTP. In addition to this long-term monitoring program, selected piping replacements have been performed to maintain suitable design margins. Where possible, FAC resistant replacement materials are used to mitigate future occurrences of FAC.A CHECWORKSTM FAC model (in accordance with the CHECWORKS T m FAC users guide and EPRI modeling guidelines) has been developed for SSES to predict the FAC wear rate (single and two-phase fluids) and the remaining service life for each piping component. As a minimum, the controlled CHECWORKSTM FAC model is updated after each refueling outage.The FAC models are also used to identify FAC examination locations for the outage examination list and uses empirical data input to the model.Variables that influence FAC include:* Moisture content* Water chemistry* Temperature

• Oxygen* Flow path geometry and velocity* Material composition SSES has predicted CPPU system operating conditions that will be used as inputs to the CHECWORKSTM FAC model. Implementation of CPPU will affect moisture' content, temperature, oxygen, and flow velocity.

For most systems the moisture content and oxygen will change but remain within the CHECWORKSTNI FAC model parameter bounds. Selected portions of some system piping are predicted to increase a maximum of I 0"F. Depending on operating power levels, flow velocities in main steam, feedwater, turbine cross around piping and Moisture Separator drains will increase approximately 15%. For drains off the Main steam piping the flow and velocities will not increase because the source pressure is the same for both 10-33 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -CPPU and CLTP. Four out of five lines in the extraction steam piping system is constructed of FAC resistant materials. The SSES CHECWORKSTM FAC model is capable of accepting these CPPU related parameter changes. Based on experience at pre CPPU operating conditions and previous FAC modeling results, it is anticipated that the CPPU operating conditions may result in the need for additional FAC monitoring points. The CHECWORKSTM FAC modeling techniques allow for the identification of additional monitoring points required for CPPU. The SSES FAC program targets FAC susceptible piping, including small pipes and drains, and includes the installation of FAC resistant material.Table 10-14 compares key parameter values (CLTP and CPPU) affecting FAC.The increased Main Steam and FW flow rates at CPPU conditions do not significantly affect the potential for FAC in these systems. Therefore, the SSES program for FAC is adequate to manage any potential effects of the CPPU on NSSS, turbine-generator (T-G), and BOP components. The reactor internals inspection and FAC programs do not significantly change for CPPU. In addition, the Maintenance Rule provides oversight for the other mechanical and electrical components, important to plant safety, to guard against age-related degradation. 10.8 NRC AND INDUSTRY COMMUNICATIONS NRC and industry communications could affect the plant design and safety analyses. However, as stated in Section 6.8, the systems significantly affected by CPPU are addressed in this report.In addition, the plant safety analyses affected by CPPU are addressed in this report. As a result, evaluations of plant design and safety analyses affected by the communications in place are inherently included in the plant-specific CPPU assessments. Therefore, it is not necessary to review prior dispositions of NRC and industry communications and no additional information is required in this area.10.9 EMERGENCY AND ABNORMAL OPERATING PROCEDURES Emergency and abnormal operating procedures can be affected by CPPU. Some of the EOPs variables and limit curves depend upon the value of rated reactor power., Some Abnormal Operating Procedures (AOPs) may be affected by plant modifications to support the higher power level. The topics addressed in this section are: Topic CLTR Disposition SSES Result Emergency operating procedures [[Abnormal operating procedures ]EOPs include variables and limit curves, defining conditions where operator actions are indicated. Some of these variables and limit curves depend upon the RTP value. Changing some of the variables and limit curves requires modifying the values in the EOPs and updating the 10-34 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -SSES support documentation. EOP curves and limits may also be included in the SPDS and will be updated accordingly. The charts and tables used by the operators to perform the EOP's are reviewed for any required changes prior to each core reload. The EOPs will be reviewed for any changes required to implement CPPU. The operators will receive training on these procedures as described in Section 10.6.AOPs include event based operator actions. Some of these operator actions may be influenced by plant modifications required to support the increase in RTP. Changing some of the operator actions may require modifications to the AOPs and updating the SSES support documentation. The plant AOPs will be reviewed for any effects of CPPU and only minor changes to the event-based actions are expected. Some of the setpoints used in the AOPs will change due to the CPPU. The operators will receive training on these procedures as described in Section 10.6.10.10 NON TURBINE RELATED MISSILES The topics addressed in this section are: Topic' CLTRDisposition SSES Result Non-Turbine Related Missiles 10.10.1 Internally Generated Missiles (Outside Primary Containment) The current licensing basis for internally generated missiles outside primary containment is described in FSAR section 3.5.1.1. This includes both rotating component failure missiles and pressurized component failure missiles.FSAR section 3.5.1.1.1 discusses rotating component failure missiles and concludes that "large, massive rotating components, such as the various ECCS pumps and motors, fans and compressors outside the primary containment, do not have sufficient energy to move the masses of their rotating parts through the housings in which they are contained." For CPPU there are no changes to these components. Hence, the current licensing basis evaluation described in FSAR section 3.5.1.1.1 is not affected by CPPU.FSAR section 3.5.1.1.2 discusses pressurized component failure missiles. This includes a) high energy piping, b) valve bonnets, c) valve stems, d) temperature detectors, e) nuts and bolts, f)blind flanges and g) safety relief valves and Main Steam Isolation Valve Accumulators. High energy piping effects due to CPPU are evaluated in PUSAR sections 10.1 and 10.10.3. For CPPU there are no changes to components "b" through "g." Hence, the current licensing basis evaluation described in FSAR section 3.5.1.1.2 is not affected by CPPU.10-35 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -10.10.2 Internally Generated Missiles (Inside Primary Containment).The current licensing basis for internally generated missiles inside containment is described in FSAR section 3.5.1.2. This includes rotating component failure missiles, pressured component failure missiles and gravitationally generated missiles.FSAR section 3.5.1.2.1 describes rotating component failure missiles, except for main turbine missiles which are described in FSAR section 3.5.1.3. Main turbine missile effect is discussed in PUSAR section 7.1. The significant pieces of rotating equipment in the primary containment are the recirculating pumps and motors. For CPPU the recirculating pumps and motors are not changed. The FSAR section concludes that other rotating components inside the containment such as fans and chillers do not have sufficient energy to move the masses of their rotating parts through the housings in which they are contained. Hence, the current licensing basis evaluation described in FSAR section 3.5.1.2.1 is not affected CPPU.FSAR section 3.5.1.2.2 describes pressurized component failure missiles and FSAR section 3.5.1.2.3 describes gravitationally generated missiles. Because there are no changes for CPPU to pressured component evaluated in these two sections, there are no changes that would affect either pressurized component failure missiles or gravitationally generated missiles. Hence, the current licensing-basis evaluations described in FSAR sections 3.5.1.2.2 and 3.5.1.2.3 respectively is not affected by CPPU.10.10.3 High Energy Line Breaks (Inside and outside Primary Containment) Increased flow rates do not affect the SSES High Energy Line Break (HELB) pipe whip and jet impingement CLTP analyses since these loadings are not dependent on the initial (pre-break) flow rate. The HELB loadings are a function of the initial internal pressure and enthalpy.Systems such as Main Steam, RCIC steam and HPCI steam see no increase in pressure or decrease in enthalpy (i.e., more subcooling); therefore they have no change to their CLTP HELB Pipe Whip and Jet Impingement Loadings. The systems with either an internal pressure increase or decrease in enthalpy or both (Feedwater and RWCU) are evaluated in the PUSAR section 10.1.2.10.10.3.1 HELB in Turbine Building CPPU does increase the flow rates, temperatures and pressures in the Condensate, Feedwater and Extraction Steam piping in the turbine building. However, CPPU has no effect since these lines are postulated to break at every fitting in the turbine building and no new break locations result.10.10.3.2 Fast Acting Valves Several transient effects from Fast-Acting Valves were reviewed as a part of the HELB review for CPPU. They are: The fast closure of Turbine Stop Valves at the higher CPPU flow rates has increased loading effects on the Main Steam and associated piping and are discussed in PUSAR section 3.5.2.10-36 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -However, no new break locations were postulated. This loading is designed to limits for an"upset" that preclude the generation of a missile.The MSIV closure following a postulated break downstream of either the inboard or the outboard MSIVs is a fast-acting valve transient. The CPPU effect is bounded by comparison to the CPPU Turbine Stop Valve Closure (TSVC) analyses since the TSV are much faster closing valves than the MSIVs.The rapid feedwater check valve closures following a postulated pipe break upstream of the three in-line check valves in each of the two main feedwater supply headers is another fast-acting valve transient. This loading was evaluated and is not affected by the increased flow, temperature and* pressure conditions prior to the postulated break.Other existing fast-acting valve transients, on various piping systems, have been reviewed for CPPU evaluations and determined not to be affected by the increased flow, temperature or pressure conditions. 10-37 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-1 High Energy Line Break Increase (Change) Due to CPPU Break Location Mass Release: Pressure Temperature Main Steam Line Break in Steam Tunnel No change No change No change Feedwater Line Breaks in Steam Tunnel (1) No change No change RCIC Steam Line Breaks in Reactor No change No change No change Building HPCI Steam Line Breaks in Reactor Building No change No change No change RWCU Breaks in Reactor Building No change (2) No change No change Notes: (1) FW line blowdown mass and energy release is bounded by the Main Steam Line Break in the Steam Tunnel.(2) RWCU mass and energy releases are bounded by the initial conditions evaluated at ARTS/MELLLA conditions. 10-38 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-2 Environmental Qualification for CPPU EQ Parameter CPPU Effect Inside Primary Containment Normal Temperature, Pressure and Humidity No Change Normal Radiation Levels <14% Increase Post-accident Peak Temperature 16.3°F Increase Post-accident Peak Pressure 4.0 psig Increase Post-accident Peak Humidity No Change Post-accident Radiation 513.8 % Increase (Wetwell)_514.7% Increase (Drywell)Outside Primary Containment Normal Temperature, Pressure and Humidity No Change Normal radiation Levels <90% Increase HELB Flooding Level No Change MELB Flooding Level No Change Post-accident Temperature, Pressure and No Change Humidity20.5% Increase (in Control Structure near Post-accident Radiation SGTS filters)118% Increase (Reactor Bldg.)10-39 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-3 Results of UNIT. 1 PRA Sensitivity Cases FRevised. CPPU_Calculated Calculated .Calculated Calculated Calculated Calculated BaseCLTPTP/BasCaseT# Post-Initiator HEPs BasesR using sing Csingusing CPPU using CPPU using CPPU Base CLTP using cPU values Tiig sn PU values uigCP Timings Timings Timings Timings Timings Timings Base CLTP Increased Increased Increased Increased Increased Base CLTP Increased Probabilities values 13% 13% 13% 13% 13% values 13%Turbine Trip w/Bypass Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP Turbine 1.194itBasenCLTPoBaseIC9T 1.194 (%INONISO, with units of (0.894) value value value value value I/yr) _ _ _ ___ ___MSIV Closure Initiator Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP 0.286 0.286 (%I ISO, with units of 1/yr) (0.136) value value value value value LOCA Initiators and internal Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP flooding due to feedwater values values values values value value Loss of Service Water (with N/A N/A N/A N/A N/A N/A 5E-3 5E-3 units of 1/yr)Loss of Instrument Air (with N/A. N/A N/A N/A N/A N/A 5E-3 5E-3 units of 1/yr)Unit I CDF (l/yr): 1.65E-6 1.71E-6 1.75E-6 1.76E-6 1.79E-6 1.88E-6 1.66E-6 1.72E-6 Unit I delta CDF: -6.0E-8 1.70E-7 1.1E-7 1.4E-7 2.3E-7 -6.0E-8 Unit I LERF (1/yr): 1.75E-7 1.76E-7 1.77E-7 1.77E-7 1.95E-7 1.96E-7 1.75E-7 1.76E-7.Unit I delta LERF: I .OE-9 2.0E-9 2.OE-9 2.0E-8 2.1 E-8 1.013E-9 Note: The notes for both Table 10-3 and Table 10-4 are located after Table 10-4.10-40 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-4 Results of UNIT 2 PRA Sensitivity Cases Revised CLTP.' CPPU Case # 5 Parameter:? ',, CLTP , .CPU "Case #1. *Case #2 case #3 Case #41 w/IA&SW l:.I/AS & SW: Base CLTP Calculated Calculated Calculated Calculated Calculated Base CLTP Calculated Post-Initiator HEPs values using CPPU using CPPU using CPPU using CPPU using CPPU values using CPPU Timings Timings Timings Timings Timings Timings SORV Probabilities Base CLTP Increased Increased Increased Increased Increased 13% Base CLTP Increased 13%values 13% 13% 13% 13% .. values Turbine Trip w/Bypass Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP (%2NONISO, with units of (0894) value 1.194 1.194 values values 1./yr) value value value MSIV Closure Initiator Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP Base CLTP (%21SO, with units of 1/yr) (0.136) value value 0.286 value 0.286 values values CLT Bse LT Bae LTP Bae CTPBase CLTP Base CLTP LOCA Initiators and internal Base CLTP Base CLTPIncreased 2x Increased 2x values values flooding due to feedwater values values values values Loss of Service Water (with N/A N/A N/A N/A N/A N/A 51-3 51-3 units of I/yr Loss of Service Water (with N/A units of 1/yr N/A N/A N/A N/A N/A 5E-3 5E-3 Unit 2 CDF (1/yr): 1.63E-6 1.70E-6 1.73E-6 1.75E-6 1.78E-6 1.86E-6 1.64E-6 1.71ZE-6 Unit 2 delta CDF: -7.0E-8 1.01E-7 1.2E-7 1.5E-7 2.3E-7 7.OE-8 Unit 2 LERF (1/yr): 1.75E-7 1.76E-7 1.77E-7 1.77E-7 1.95E-7 1.96E-7 1.75E-7 1.76E-7 Unit 2 delta LERF: 1.01E-9 2.0E-9 2.OE-9 2.0E-8 2.1E-8 L.OE-9 Notes to Tables 10-3 and 10-4: Sensitivity Case #1: This sensitivity increases the Non-isolation transient initiator frequency to bound the various changes to the BOP side of the plant (e.g., main turbine modifications). The revision to the Non-isolation frequency using an approach that assumes an additional turbine trip and additional loss of feedwater event is experienced in the first year following start-up in the CPPU condition and an additional

0.5 event

for both turbine trip and loss of feedwater in the second year. The change in the long-term average of the Non-isolation initiating event is calculated as follows for this sensitivity case:* Base long-term Non-isolation frequency is 0.894/yr for both unit I and unit 2* 10 years is used as the "long-term" data period 10-41 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-4 Results of UNIT 2 PRA Sensitivity Cases* End of 10 years does not reach the end-of-life portion of the bathtub curve* Revised Turbine Trip w/Bypass frequency for this sensitivity case is calculated as: (10 x 0.994)+ 1.5 + 1.5 = 1.194/yr 10 All other parameters are maintained the same as the CPPU base case. The results were obtained by re-quantifying the model.Sensitivity Case #2: This sensitivity case conservatively assumes that the potential impact on transient initiator frequencies is manifested in the MSIV Closure initiator frequency and not the Turbine Trip initiator. The base Isolation initiator frequency of 1.36E&l/yr for both unit I and unit 2 is revised in this sensitivity case in the same manner as that discussed in Sensitivity Case #1: (lOx 1.36E-1)+ 1.0 +0.5 = 0.286/yr 10 All other parameters are maintained the same as the CPPU base case. The results were obtained by re-quantifying the model.Sensitivity

1. 3: The CPPU base quantification does not modify the DBA LOCA frequency.

Acknowledging that the increased flow rates of the CPPU can result in increased piping erosion/corrosion rates, this sensitivity case conservatively doubles the LOCA initiator frequencies for the small, medium and large LOCA categories. The internal flooding initiating event frequencies for feedwater were also doubled due to increased flow in this system as a result of CPPU. All other parameters are maintained the same as the CPPU base case. The results were obtained by re-quantifying the model.Sensitivity

1. 4: This sensitivity case combines the changes of Sensitivity Cases 1, 2, and 3. All other parameters are maintained the same as the CPPU basse case. The results were obtained by re-quantifying the model.Sensitivity
2. 5: This sensitivity case takes into account two new initiators that were not previously accounted for in the CLTP model, i.e., loss of service water and loss of instrument air. Both of these initiating event frequencies were estimated to be approximately 5E-3/yr. All other parameters are maintained the same as the CPPU base case.10-42 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-5 Summary of Changes in Post-Initiator HEPs Due to CPPU CLTP CPPU I,."Action Description.,..

PRA BE ID Failure Failure " CPPU Basis -... ..._____ ___ __ __ ___ ___ Probability Probability .__________________"___ .....OPERATOR FAILS TO TRANSFER 137-N-N-CST_18-0 1.50E-2 2.30E-2 Available time of 13 minutes reduced to I I minutes for CPPU WATER TO CST WITHIN 18 MINUTES conditions. OPERATOR FAILS TO CONTROL 150- 1.00E-02 1.50E-02 Available time of 15 minutes reduced to 13 minutes for CPPU REACTOR WATER LEVEL 152RXLEVELCTRL-C conditions. OPERATOR FAILS TO START CS 151-N-N-F005_TR 1.30E-02 1.70E-02 Available time of 25 minutes reduced to 22 minutes for CPPU GIVEN AUTOSTART FAILURE -(TR- 0 conditions. 3 SEQUENCES) OPERATOR FAILS TO START CS 151-N-N-F005_TR 3.OOE-02 3.90E-02 Available time of 17 minutes reduced to 15 minutes for CPPU GIVEN AUTOSTART FAILURE -(TR- 0 conditions. 5 SEQUENCES) OPERATOR FAILS TO CONTROL RPV 152-N-N-RPVLVL_5-( 2.50E-01 3.OOE-01 Available time of 3 minutes reduced to 2.5minutes for CPPU LEVEL WITHIN 5 MINUTES conditions. OPERATOR FAILS TO CONTROL RPV 152-N-N-RPVLVL 20 6.1OE-03 1.OOE-02 Available time of 18 minutes reduced to 15 minutes for CPPU LEVEL WITHIN 20 MINUTES 0 conditions. OPERATOR FAILS TO INITIATE SLCS 153-N-N-SLCS5-O 7.80E-02 1.03E-01 Available time of 3.7 minutes reduced to 3.2 minutes for IN 4.7 MINUTES IN ATWS CPPU conditions. OPERATOR FAILS TO INITIATE SLCS 153-N-N-SLCS7-O 3.30E-02 4.50E-02 Available time of 6 minutes reduced to 5 minutes for CPPU IN 7 MINUTES IN ATWS conditions. OPERATOR FAILS TO INITIATE SLCS 153-N-N-SLCS20-O 2.60E-03 3.50E-03 Available time of 19 minutes reduced to 16 minutes for CPPU IN 20 MINUTES IN ATWS conditions. OPERATOR FAILS TO INITIATE SLCS 153-N-N-SLCS8-O 2.50E-02 3.30E-02 Available time of 7 minutes reduced to 6 minutes for CPPU IN 8 MINUTES IN ATWS conditions. OPERATOR FAILS TO INITIATE SLCS 153-N-N-SLCS12-O 1.20E-02 1.60E-02 Available time of I I minutes reduced to 9.5 minutes for IN 12 MINUTES IN ATWS CPPU conditions.. OPERATOR FAILS TO INITIATE MRI 156-N-N-MRI12-0 6.10E-02 8.30E-02 Available time of 7 minutes reduced to 6 minutes for CPPU WITHIN 12 MINUTES AFTER ATWS conditions. OPERATOR FAILS TO INITIATE 183-MAND-AT-O 6.40E-02 8.30E-02 Available time of 4 minutes reduced to 3.5 minutes for CPPU MANUAL DEPRESSURIZATION conditions.(ATWS)10-43 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-5 Summary of Changes in Post-Initiator HEPs Due to CPPU C L T P ...P... .... .Action Description PRA BE ID Failure Failure -.---CPPU Basis_____________________________ ______________ --Probability- -Probability.__________________________ OPERATOR FAILS TO INITIATE 183-MAND-NA-O 4.70E-04 7.60E-04 Available time of 25 minutes reduced to 22 minutes for CPPU MANUAL DEPRESSURIZATION (Non- conditions. ATWS, TRANSIENTS and Small Steam LOCA)OPERATOR FAILS TO INITIATE 183-MAND-SO-O 1.20E-02 1.60E-02 Available time of 9 minutes reduced to 8 minutes for CPPU MANUAL DEPRESSURIZATION (Non- conditions. ATWS, Small Liquid LOCA and SORV)OPERATOR FAILS TO INHIBIT ADS 183-N-N- 3.60E-02 4.70E-02 Available time of 9 minutes reduced to 8 minutes for CPPU WITHIN 10 MINUTES DURING ATWS ADS INH 10-0 conditions. OPERATOR FAILS TO INHIBIT ADS 183-N-N-ADS INH. 7- 8.30E-02 1.20E-01 Available time of 6 minutes reduced to 5 minutes for CPPU WITHIN 7 MINUTES DURING ATWS 0 conditions. OPERATOR FAILS TO CONTROL ICLPIA-O 1.60E-01 2.30E-01 Available time of 3.7 minutes reduced to 3.3 minutes for LOW PRESSURE INJECTION DURING CPPU conditions. ATWS OPERATOR FAILS TO MANUALLY LOCAM-O 4.50E-02 5.60E-02 Available time of 14 minutes reduced to 12.5 minutes for START THE RHR OR CS PUMPS FOR CPPU conditions. A LOCA M OPERATOR FAILS TO MANUALLY LOCAS-O 2.70E-02 3.40E-02 Available time of 18 minutes reduced to 16 minutes for CPPU START THE RHR OR CS PUMPS FOR conditions. ALOCA S OPERATOR FAILS TO MANUALLY TRANS-O 3.40E-02 4.50E-02 Available time of 16 minutes reduced to 14 minutes for CPPU START THE RHR OR CS PUMPS FOR conditions. A TRANSIENT OPERATOR FAILS TO TRANSFER 237-N-N-CST_18-0 1.50E-2 2.30E-2 Available time of 13 minutes reduced to II minutes for CPPU WATER TO CST WITHIN 18 MINUTES conditions. OPERATOR FAILS TO CONTROL 250- 1.00E-02 1.50E-02 Available time of 15 minutes reduced to 13 minutes for CPPU REACTOR WATER LEVEL 252RXLEVELCTRL-( conditions. OPERATOR FAILS TO START CS 251-N-N-F005_TR 1.30E-02 1.70E-02 Available time of 25 minutes reduced to 22 minutes for CPPU GIVEN AUTOSTART FAILURE -(TR- 0 conditions. 3 SEQUENCES) 10-44 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-5 Summary of Changes in Post-Initiator HEPs Due to CPPU-- ...... .......,.... ... ................ * .. ... ...., ..---_ i ....- .' " " CLTP CPPU Action Description -PRA BE ID Failure Failure CPPU Basis-,-____________________________-Probability-Probability:-......... ...-OPERATOR FAILS TO START CS 251-N-N-F005_TR 3.OOE-02 3.90E-02 Available time of 17 minutes reduced to 15 minutes for CPPU GIVEN AUTOSTART FAILURE -(TR- 0 conditions. 5 SEQUENCES) OPERATOR FAILS TO CONTROL RPV 252-N-N-RPVLVL_5-( 2.50E-01 3.OOE-01 Available time of 3 minutes reduced to 2.5 minutes for CPPU LEVEL WITHIN 5 MINUTES conditions. OPERATOR FAILS TO CONTROL RPV 252-N-N-RPVLVL_20 6.1 OE-03 1.00E-02 Available time of 18 minutes reduced to 15 minutes for CPPU LEVEL WITHIN 20 MINUTES 0 conditions. OPERATOR FAILS TO INITIATE SLCS 253-N-N-SLCS5-O 7.80E-02 1.03E-01 Available time of 3.7 minutes reduced to 3.2 minutes for IN 4.7 MINUTES IN ATWS ..CPPU conditions. OPERATOR FAILS TO INITIATE SLCS 253-N-N-SLCS7-O 3.30E-02 4.50E-02 Available time of 6 minutes reduced to 5 minutes for CPPU IN 7 MINUTES IN ATWS conditions. OPERATOR FAILS TO INITIATE SLCS 253-N-N-SLCS20-O 2.60E-03 3.50E-03 Available time of 19 minutes reduced to 16 minutes for CPPU IN 20 MINUTES IN ATWS conditions. OPERATOR FAILS TO INITIATE SLCS 253-N-N-SLCS8-O 2.50E-02 3.30E-02 Available time of 7 minutes reduced to 6 minutes for CPPU IN 8 MINUTES IN ATWS conditions. OPERATOR FAILS TO INITIATE SLCS 253-N-N-SLCSI2-O 1.20E-02 1.60E-02 Available time of I I minutes reduced to 9.5 minutes for IN 12 MINUTES IN ATWS CPPU conditions. OPERATOR FAILS TO INITIATE MRI 256-N-N-MRI12-O 6.1OE-02 8.30E-02 Available time of 7 minutes reduced to 6 minutes for CPPU WITHIN 12 MINUTES AFTER ATWS conditions. OPERATOR FAILS TO INITIATE 283-MAND-AT-O 6.40E-02 8.30E-02 Available time of 4 minutes reduced to 3.5 minutes for CPPU MANUAL DEPRESSURIZATION conditions.(ATWS)OPERATOR FAILS TO INITIATE 283-MAND-NA-O 4.70E-04 7.60E-04 Available time of 25 minutes reduced to 22 minutes for CPPU MANUAL DEPRESSURIZATION (Non- conditions. ATWS, TRANSIENTS and Small Steam LOCA)OPERATOR FAILS TO INITIATE 283-MAND-SO-O 1.20E-02 1.60E-02 Available time of 9 minutes reduced to 8 minutes for CPPU MANUAL DEPRESSURIZATION (Non- conditions. ATWS, Small Liquid LOCA and SORV)OPERATOR FAILS TO INHIBIT ADS 283-N-N- 3.60E-02 4.70E-02 Available time of 9 minutes reduced to 8 minutes for CPPU WITHIN 10 MINUTES DURING ATWS ADS INH 10-0 conditions. 10-45 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-5 Summary of Changes in Post-Initiator HEPs Due to CPPU CLTP_ CP Action Desc-iptn ..PRA BE ID '-'-Failure Failu.r.e_. " .. CPPU Basis__________________________________'Probability Prob-ability ___________________________ OPERATOR FAILS TO INHIBIT ADS 283-N-N-ADSINH_7-8.30E-02 1.20E-01 Available time of 6 minutes reduced to 5 minutes for CPPU WITHIN 7 MINUTES DURING ATWS 0 conditions. OPERATOR FAILS TO CONTROL 2CLPIA-O 1.60E-01 2.30E-01 Available time of 3.7 minutes reduced to 3.3 minutes for LOW PRESSURE INJECTION DURING CPPU conditions. ATWS 10-46 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-6 Summary of Changes in Post-Initiator Dependent HEPs Due to CPPU CLTP Failure' 'Independent HEP Names, pr-CPPU CP aiue ActionDesription

PRABED ..CTPrbability .Values andaAssumed DependenceP

Probability~

JOINT HEP OPERATOR FAILS TO ALIGN Z-EARLY-RXLC-O 1.21E-03 150-152RXLEVELCTRL-O 1.00E-02 1.82E-03 FIRE MAIN OR RHRSW AND FAILS TO 013-N-N-EARLY-O 7.50E-02 CONTROL RX WATER LEVEL Low Dependence JOINT HEP OPERATOR FAILS TO Z-IACIG-RXLC-O 3.33E-02 125-N-N-FXTIACIG-O 2.20E-01 3.43E-02 CROSSTIE IA & CIG AND FAILS TO 150-152RXLEVELCTRL-O 1.001E-02 CONTROL RX WATER LEVEL Moderate Dependence JOINT HEP OPERATOR FAILS TO Z-RXLC-CVLOC-O 5.48E-04. 150-152RXLEVELCTRL-O 1.00E-02 8.23E-04 CONTROL RX WATER LEVEL AND 159-CNTVNTLOCAL-O 5.1E-03 VENT CONTAINMENT LOCALLY Low Dependence Table 10-7 Increase in SORV Probability ,BEID Pre'CPPU" CPPU Probability Probability-NNREE-P%.5E37.6-183-N-N-RESET-PE 6.85E-3 7.76E-3 10-47 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-8 Release Category Changes Due to CPPU Conditions Sequence,, Pre-CPPU Release " CPPU Release D e si g n~a t o r S:~ * + ' : q u e n c e D e s c r ip tio n C a te go r y C a te go r y'D esignato r! C I .... .. .. .g y .. ..TR-I-009 Loss of Injection after successful venting L/I L/E through the wetwell airspace.TR-1-010 Loss of Injection after successful venting, but H/I H/E suppression pool bypass occurs.TR-1-01 I Loss of Injection after successful venting H/I H/E through the drywell airspace.TR-1-012 Loss of Injection after successful venting, but H/I H/E energetic phenomena fails containment at time of vessel failure.TR-2-015 Loss of high pressure injection with failure to L/L L/I depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup OK after containment fails in wetwell below the waterline. TR-2-016 Loss of high pressure injection with failure to M/L M/I depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup fails after containment fails in wetwell below the waterline. TR-2-017 Loss of high pressure injection with failure to LL/L LL/I depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup OK after containment fails in wetwell airspace.TR-2-018 Loss of high pressure injection with failure to L/L L/I depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup OK after containment fails in wetwell airspace, but pool is bypassed.TR-2-019 Loss of high pressure injection with failure to LL/L LL/I depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup fails after containment fails in wetwell airspace.TR-2-020 Loss of high pressure injection with failure to M/L M/I depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup fails after containment fails in wetwell airspace, but pool is bypassed.TR-2-021 Loss of high pressure injection with failure to depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup OK after containment fails in drywell.UL L/I i .10-48 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-8 Release Category Changes Due to CPPU Conditions' Sequence Seence Descip Pre-CPPU Release CPPU Release' ' ..: : : ..r.Sequence D escription .-.* .-" 1 * : Designator q Category, Category TR-2-022 Loss of high pressure injection with failure to M/L M/I depressurize (core damage at 40 min).Suppression pool cooling fails. Makeup fails after containment fails in drywell.TR-5-009 SORV with suppression pool cooling and M/L M/I containment venting failed. Loss of injection after containment fails in the wetwell below the waterline. TR-5-010 SORV with suppression pool cooling and L/L L/I containment venting failed. Loss of injection after containment fails in the wetwell airspace.TR-5-011 SORV with suppression pool cooling and M/L M/I containment venting failed. Loss of injection after containment fails in the wetwell airspace, but pool is bypassed.TR-5-012 SORV with suppression pool cooling and M/L M/I containment venting failed. Loss of injection after containment fails in the drywell region.TR-5-013 SORV with suppression pool cooling and H/L H/I containment ventingfailed. Loss of injection after containment fails. Energetic phenomenon at time of vessel failure leads to high release.TR-5-034 SORV with loss of injection and failure of M/L M/I drywell sprays. Containment over temperature failure occurs.I.. The first character(s) describes the magnitude of Csl as a percentage of core inventory released to the environment: H> 10%, M > 1%, L>0.1%,andLL<0.1% The second character describes time of the release where time zero represents declaration of a General Emergency: E = 0 to 6 hours, I = 6 to 24 hours, and L > 24 hours 10-49 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-9 Estimate of Impact on Fire CDF Due to CPPU Increasen m P C Fire Fire ZnEie Lot 'CDF per 15 CDF per cal-, PostCPPUFire

oreh yone cqupment Lost, CDF from.Cmonmcycle

.yr Contribution 1-2B Div I and 11 ESW 2.1E-09 1.6E-09 4.6E-10 2.1E-09 0-28B-II Battery Charger Area, Channels 1.3E-09 9.9E-10 2.9E-10 1.3E-09 A and B DC 0-27C Upper Cable Spreading Room. 3.5E-10 2.7E-10 7.8E-1 1 3.5E-10 Channels A and B DC Power 0-25E Lower Cable Spreading Room. 3.3E-09 2.5E-09 7.3E-10 3.2E-09 HPCI and Div I RHR* Various 15 Fire Zones Evaluated that 3.3E-08 2.5E-08 7.3E-09 3.2E-08 Lacked Defense in Depth 0-26H Panel 1C601 -Auto Initiation 5.1E-09 3.9E-09 1.IE-09 5.0E-09 of Emergency Core Cooling System (ECCS)Total 4.5E-08 3.4E-08 9.9E-09 4.4E-08 (1) Assuming an availability factor of 0.95.(2) Assuming a 29% increase (approximated by the CPPU impact for non-isolation, isolation and LOOP events in the internal events evaluation) in the CDF impact.Table 10-10 Definition of Importance Levels Importance Definition" L~elI: ~ i :: :: -i :Definition, :..: .:. :. :: Level A. Extremely important and necessary to address to assure the technical adequacy of the PRA or the quality of the PRA or the quality of the PRA update process.(Contingent Item for Certification). B. Important and necessary to address, but may be deferred until the next PRA update. (Contingent Item for Certification). C. Marginal importance, but considered desirable to maintain maximum flexibility in PRA Applications and consistency in the Industry.. D. Editorial or Minor Technical Items left to the discretion of the host utility.S. Considered a major strength of the PRA.10-50 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-11 Open 'B' Peer Review Certification Resolution Prior to Issuance of the AUG06preEPU-C PRA Model S F&O : .& :..i.. -, ...:. ' :. '.- .. :.. ..Disposition for CPPU :r ,Description Prior PPL Susquehanna Disposition dpplica.ion Identifer 'li : ... -.:: , .:: : .A p cation AS-7-4 Conservative RPV Frequency documented in the Initiating Because this directly influences Rupture Frequency Events Notebook. the LERF frequency, the initiating event (IE) frequency value was adjusted to 1.01E-8 consistent with many other industry BWRs. (Now closed).HR-4-1 Missing Pre- Adding more pre-initiators is not expected Acceptable to proceed as is.Initiator HEPs to affect the insights presently realized.IE-13-1 LOOP Frequency Future update to consider using Incorporated new data directly Pedigree INEEL/EXT-04-02326 LOOP frequency into the AUG06preEPU-C and and recovery data. AUG06EPU models. (Now closed).IE-5-2 Reconsider IE Future update to consider LOSW. Others Acceptable to proceed as is.exclusion for loss of adequately addressed. Consider sensitivity studies service water (refer to Table 10-3).(LOSW), etc.IE-6-1 Consider including Future update to consider LOIA. Acceptable to proceed as is.loss of instrument Consider sensitivity studies air (LOIA) (refer to Table 10-3).IE-7-1 Consider including Future update to consider BOC. Included in updated models break outside because it will influence LERF.containment (BOC) (Now closed).IE-7-2 Consider including Future update to consider LOIA and BOC. See resolution above for IE-6-1 LOIA and BOC and IE-7-1.IE-13-2 Compare IE Results indicate reasonableness of chosen Values are reasonable based on frequencies with values, comparison with other similar other similar sites. sites. (Now closed).L2-5-1 Reconsider timing Being evaluated as part of the Level 2 Addressed by updated detailed of containment update. Level 2 analysis included in the overtemperature AUG06preEPU-C and failure (COTF) AUG06EPU models. (Now scenarios closed).L2-8-2 Adjust CI node Being evaluated as part of the Level 2 Addressed by updated detailed placement in event update. Level 2 analysis included in the trees AUG06preEPU-C andAUG06EPU models. (Now closed).10-51 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-11 Open 'B' Peer Review Certification Resolution Prior to Issuance of the AUG06preEPU-C PRA Model F&O ...:uDescrip tion Prior PPL Susquehanna Disposition Disposition for CPPU Identirier

Application:

L2-1 0-1 Reconsider COTF Being evaluated as part of the Level 2 Addressed by updated detailed w/o drywell sprays update. Level 2 analysis included in the AUG06preEPU-C and AUG06EPU models. (Now closed).L2-15-1 Refine ATWS CF Being evaluated as part of the Level 2 Addressed by updated detailed assumptions update. Level 2 analysis included in the AUG06preEPU-C and AUG06EPU models. (Now closed).L2-22-3 Conservative LERF Being evaluated as part of the Level 2 Addressed by updated detailed Timing update. Level 2 analysis included in the AUG06preEPU-C and AUG06EPU models. (Now closed).MU-1 Formalize PRA Although overall PRA update procedure Continue existing calculation Model Update would be beneficial, the current model is review and approval processes. Process documented and controlled under PPL No impact on CPPU results.Susquehanna QA procedures. QU-19-1 Formalize PRA Although overall PRA model assembly Continue existing calculation Model Assembly procedure would be beneficial, the current review and approval processes. Process model is documented and controlled under No impact on CPPU results.PPL Susquehanna QA procedures. ST-5-2 Reconsider COTF Being evaluated as part of the Level 2 Addressed by updated detailed Assumptions update. Level 2 analysis included in the AUG06preEPU-C and AUG06EPU models. (Now closed).ST-5-3 Refine ATWS CF Being evaluated as part of the Level 2 Addressed by updated detailed assumptions update. Level 2 analysis included in the AUG06preEPU-C and AUG06EPU models. (Now_closed).SY-4-1 Complete System 10 remaining system notebooks to be Deferred. No significant model Notebooks completed. impacts are foreseen from the remaining low risk significant systems.SY-8-2 Missing Pre- Adding more pre-initiators is not expected Acceptable to proceed as is.Initiator HEPs to affect the insights presently realized.10-52 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-12 Comparison of CLTP CDF vs. CPPU CDF by Initiator-Uiiit-1 Unit iM7%ý-Ui2-Ui2'. .nt2--P--c~u -- pU:ý: Unit 1 -1 nitvTP Unit 2UntRliv%i.::CPP.!ii--2il i:!. CPPU_ ( ::.Untii.. Relative %n:Vaiue.r- .-..-,.Va~ _lue %inc-rese. -Relatie- ValueP PValue. 2---------by -Initiator---------- by Initiator Ices (1/yr) -(1/yr)- Increase ::. (I/yr) Inyr -Loss of Offsite Power (%LOOP-FLAG) 1.20E-06 1.25E-06 4.0% 73.8% 1.19E-06 1.24E-06 4.1% 73.9%Reactor Trip without MSIV Closure (%1(2)NONISO) 8.56E-08 9.51E-08 11.0% 14.4% 8.46E-08 9.39E-08 11.1% 14.3%Inadvertent MSIV Isolation (%1(2)ISO) 3.84E-08 4.37E-08 13.8% 8.1% 3.83E-08 4.36E-08 13.8% 8.1%Flood in Room 11-500 (%FLD- 1 (2)-749FLOODSW) 3.33E-9 4.56E-9 37.0% 1.9% 3.32E-9 4.56E-09 37.2% 1.9%Small Liquid LOCA (%1(2)LOCA-SM-LQD) 4.28E-08 4.32E-08 0.8% 0.5% 4.27E-08. 4.32E-08 1.3% 0.8%Steam Relief Valve Inadvertently Opens (%I(2)IORV) 1.22E-09 1.52E-09 24.4% 0.5% 1.22E-09 1.52E-09 24.4% 0.5%Other Initiators Contributing <0.5%to ACDF 2.75E-07 2.76E-07 0.2% 0.7% 2.70E-07 2.70E-07 0.1% 0.5%TOTALS: 1 1.65E-06 1.71E-06 4.0% 100%* 1.63E-06 1.70E-06 4.0% 100%* Total may not be exactly 100% due to round off error.10-53 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-13 Comparison of CLTP CDF vs. CPPU CDF by Sequence"- -= Unit -CLTP -Unit 1 CPPU Unit I Relative<- -Unit 2 CLTP Unit 2 CPPU Unit 2 Relative SSequence ............. De -I Value Value ofCDF ValUe Value %of CDF Designator'. l --. ,: In res ..of .ncreas... ... ... .... .. ._..... .. ..:_ .... ... ..---.. ... (l/yr) .. (l/y r) _ _. ( / r .... ... (1/ r) ......._ ___RCVSEQI(2)TR Loss of extended high pressure 9.01E-07 9.28E-07 40.8% 9.02E-07 9.28E-07 .39.8%OOICD makeup and depressurization (core damage at 5.9 hrs).RCVSEQI(2)TR Loss of injection after successful 3.17E-07 3.40E-07 34.8% 3.25E-07 3.49E-07 36.8%005CD containment venting (core damage at 25.6 hours).RCVSEQ1(2)TR Isolation ATWS with loss of high 1.30E-08 2.11 E-08 12.2% 1.29E-08 2.11 E-08 12.6%036CD pressure makeup, success of SLC and depressurization, and with failure to control level with low pressure makeup.RCVSEQI(2)TR Isolation ATWS with success of 1.50E8 1.76E-08 3.9% 1.49E-08 1.76E-08 4.1%030CD high pressure makeup and the failure of SLC and MRI.RCVSEQI(2)TR Isolation ATWS with loss of high 4.83E-09 7.04E-09 3.3% 4.83E-09 7.04E-09 3.4%038CD pressure makeup, success of SLC, and failure to depressurize. RCVSEQI(2)TR Loss of LP makeup following 4.59E-08 4.70E-08 1.7% 4.31E-08 4.41E-08 1.5%038CD failure of HP makeup and success of depressurization. RCVSEQI(2)TR Loss of LP makeup following 2.05E-08 2.13E-08 1.2% 5.61E-10 5.83E-10 0.0%023CD successful RPV depressurization (core damage at 5.8 hours).10-54 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-13 Comparison of CLTP CDF vs. CPPU CDF by Sequence-U- -- Unit 1 CLTP. Unit I cPPU Unit C Relaive Unit 2 CLTP :Unit 2 CPPU :Unit-2Relative Sequence.- Description Value Value -% ofCDF = =Value,,,. vale of CDF-l-t ign io"_-ýz.- I:ý-ý ::::::::::::_ý

ý_-ý Des.- eri :i -on, -1 .::.t.......

.. :: l y ) ..... .: ( /r..... c e s : C D , .i: : -: %ly ) (l y -: :- .: I cr se .(lyr.~ (fy)Increase (/)(Iy)Increase Other Other Sequences that Contribute 3.30E-07 3.31 E-07 2.0% 3127E-07 3.28E-07 1.7%<1% to ACDF TOTALS: 1.63E-06 1.71E-06 100*% 1.63E-06 1.70E-06 100%* Totals may not be exactly 100% due to round off.10-55 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table 10-14 FAC Parameter Comparison for CPPU CLTP RTP Range of CPPU RTP Range of ParmeerValues Values Main Steam Flow (Ibm/hr) 14,436,937 16,537,562 Main Steam Quality (%) 99.60 99.47 Main Steam Velocity (ft/sec) 138.2-187.4 158.9-219.4 (14-17%)Feedwater piping Temperature (CF) 390.4 399.2 Feedwater Flow (Ibm/sec) 14,404,936 16,505,562 10-56 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -11. REFERENCES .1. GE Nuclear Energy, "Constant Pressure Power Uprate," Licensing Topical Report NEDC-33004P-A, Revision 4, Class III (Proprietary), July 2003.2. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Licensing Topical Reports NEDC-32424P-A, Class III (Proprietary), February 1999.3. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," Licensing Topical Reports NEDC-32523P-A, Class III, February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999.4. Power Uprate with Increased Core Flow, Susquehanna Steam Electric Station, Unit 1 TAC No. M90075, Amendment 143 to License No. NPF-14, February 22, 1995 5. Power Uprate with Increased Core Flow, Susquehanna Steam Electric Station, Unit 2 (PLA-4055) (TAC No. M8831 1), Amendment 101 to License No. NPF-22, April 11, 1994Susquehanna Steam Electric Station, Proposed Amendment No. 235 to License NPF-14 and Proposed Amendment No. 200 to License NPF-22: Power Uprate, PLA-5212, October 15, 1992.6. Susquehanna Steam Electric Station, Units I And 2 -Issuance Of Amendment Re: 1.4 Percent Power Uprate (TAC Nos. MB0444 and MB0445) July 6, 2001.7. Susquehanna Steam Electric Station Proposed License Amendment Numbers 279 For Unit 1 Operating License No. NPF-14 and 248 For Unit 2 Operating License No. NPF-22 ARTS/MELLLA Implementation PLA-593 1, November 18, 2005.8. EMF-CC-074(P)(A) Volume 4 Revision 0, BWR Stability Analysis -Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000 9. PLA-5933, PPL Letter To NRC, "Proposed Amendment No. 280 To Unit 1 Facility Operating License NPF-14 And Proposed Amendment No. 249 To Unit 2 Facility Operating License NPF-22: Revise Technical Specification 3.4.10 "RCS Pressure And Temperature (P/T) Limits", McKinney, Britt T. To U.S. Nuclear Regulatory Commission, 10/5/2005 10. GE Nuclear Energy, "The GE Pressure Suppression Containment System Analytical Model," NEDM-10320, March 1971.11. GE Nuclear Energy, "The General Electric Mark III Pressure Suppression Containment System Analytical Model," NEDO-20533, June 1974.12. U. S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, February 2004.13. GE Nuclear Energy, "Mark II Pressure Suppression Containment Systems: An Analytical Model of the Pool Swell Phenomenon," NEDO-21544, December 1976.11-1 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -11. REFERENCES

14. NRC NUREG-0808, "Mark II Containment Program Load Evaluation and Acceptance Criteria," August 1981 15. NUREG-0800, Revision 4, dated October 1985, SRP 6.2.2, "Containment Heat Removal Systems 16. GE Nuclear Energy, "General Electric Model for LOCA Analysis In Accordance With 10 CFR 50 Appendix K," NEDE-20566-P-A, September 1986.17. Susquehanna Steam Electric Station Unit 1 & 2 Generic Letter 96-06 Assurance Of Equipment Operability

& Containment Integrity During Design Basis Accidents (TAC NOS. MB96875 & MB96876), Letter from Guzman to Shiver Dated 8/12/2003 18. NEDO-32686 Rev. 0, Utility Resolution Guidance for ECCS Suction Strainer Blockage 19. NUREG/CR-6224, Parametric Study of the Potential for BWR ECCS Strainer Blockage due to LOCA Generated Debris 20. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP, May 2001 21. XN-CC-33(P)(A) Revision 1, HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option Users Manual, Exxon Nuclear Company, November 1975 22. PLA-5963, "SSES Proposed Amendment No.281 to License NPF-14 and Proposed Amendment to No. 251 to License NPF-22: Application for License Amendment and Related Technical Specification Changes to implement Full-Scope Alternative Source Term in Accordance with 10 CFR 50.67," dated October 13, 2005.23. Pennsylvania Power & Light Company, "Susquehanna Steam Electric Station Units I &2 Design Assessment Report (DAR), Rev. 9" 24. USNRC NUREG-0487, "Mark II Containment Program Load Evaluation and Acceptance Criteria," Supplement 1, September 1980.25. ANF-524(P)(A) Revision 2 and Supplements I and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.26. GE Nuclear Energy, "General Electric Instrument Setpoint Methodology," NEDC-31336P-A, Class III (Proprietary), September 1996.27. Fire Protection Review Report, Rev. 16 28. TP-04124-NP-A, Missile Probability Analysis For The Siemens 13.9M 2 Retrofit Design Of Low-Pressure Turbine By Siemens AG, June 7, 2000 29. U. S. Atomic Energy Commission Technical Information Document TID-14844,"Calculation of Distance Factors for Power and Test Reactors," dated March 23, 1962.30. Susquehanna Steam Electric Station Units I & 2, Environmental Radiological Monitoring Annual Report, 1997 -2000 11-2 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -11. REFERENCES

31. ANF-913(P)(A)

Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990 32. XN-NF-80-19(P)(A) Volume 3 Revision 2, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description, Exxon Nuclear Company, January 1987.33. GE Nuclear Energy, "ATWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability," NEDO-32047-A.

34. GE Nuclear Energy, ,'Mitigation of BWR Core Thermal-Hydraulic Instabilities in ATWS," NEDO-32164.

35. NEI, PRA Peer Review Guidelines, NEI 00-02, Rev. A3, 3/20/2000.

36. U.S. Nuclear Regulatory, Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Parts 2-5, Vol. 2, NUREG-1560, December 1997 37. PPL (Pennsylvania Power and Light), Susquehanna Steam Electric Station Individual Plant Examination for External Events, June 1994.38. PPL (Pennsylvania Power and Light), Susquehanna Steam Electric Station Response to Audit Issues on IPEEE Submittal Units I and 2, PLA-4983, October 1998 39. Hickman, J. W., et al., PRA Procedures Guide, NUREG/CR-2300, January 1983.40. PPL (Pennsylvania Power and Light), Susquehanna Steam Electric Station Response to Audit Issues on IPEEE Submittal Units 1 and 2, PLA-4983, October 1998.41. EPRI (Electric Power Research Institute), A Methodology for Assessment of Nuclear Power Plant Seismic Margin, EPRI NP-6041 Revision 1, August 1991.42. Sandia National Laboratories, Analysis of Core Damage Frequency:

Peach Bottom, Unit 2 External Events, NUREG/CR-4550, Vol. 4, Rev. 1, Part 3, December 1990.43. Philadelphia Electric Company, Limerick Generating Station Severe Accident Risk Assessment, April 1983.44. Sandia National Laboratories, Shutdown Decay Heat Removal Analysis, GE BWR3/Mark I Case Study, NUREG/CR-4448, December 1986.45. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2, Siemens Power Corporation, October 1999 46. U. S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, February 2004.47. U. S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, February 2004.11-3 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -11. REFERENCES

48. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-S-2002, April 5, 2002, and Addenda to ASME RA-S-2002, ASME RA-Sa-2003, December 5, 2003'49. U. S. Nuclear Regulatory Commission, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition) Regulatory Guide 1.70 (Rev. 2)September, 1975.50. Letter, S. Richards (NRC) to J.F. Mallay (FANP), "Siemens Power Corporation RE: Request for Concurrence on Safety Evaluation Report Clarifications (TAC No.MA6160)," May 31, 2000.51. PLA-6049, PPL Letter to NRC, "Susquehanna Steam Electric Station Proposed License Amendment No. 289 to Unit 1 Operating License NPF-14 and Proposed License Amendment No. 257 to Unit 2 Operating License NPF-22, Standby Liquid Control System," McKinney, B.T. to US Nuclear Regulatory Commission, dated April 28, 2006 52. Susquehanna Steam Electric Station, Units I and 2 -Issuance of Amendment RE: Revision to Technical Specification 3.4.10, Reactor Coolant System Pressure and Temperature Limits (TAC Nos. MC8646 and MC8647), dated March 30, 2006.53 NUREG-I048M, Safety Evaluation Report Related to the Operation of Hope Creek Generating Station, Supplement 6, Appendix U, July 1986.54 Berkow (NRC) to Dembkowski (Siemens Westinghouse Power Corporation);

Safety Evaluation For Acceptance Of Referencing The Siemens Westinghouse Topical Report,"Missile Analysis Methodology For General Electric (GE) Nuclear Steam Turbine Rotors By The Siemens Westinghouse Power Corporation (SWPC)" (TAC NO.MB5679); dated July 22, 2003.55 W. Glenn Warren (BWROG) to NRC, BWROG Letter BWROG-00068, "Alternative BWR Feedwater Nozzle Inspection Requirments, GE-NE-523-A71-0594A, Revision 1, May, 2000," dated June 7, 2000.56 Letter, John F. Stolz (NRC) to Robert G. Byram (PPL), "Susquehanna Steam Electric Station, Units I and 2, Safety Evaluation Regarding Spent Fuel Pool Cooling Issues (TAC No. M85337)," June 19, 1995.11-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Appendix ATRIUM 10 Fuel Activity Tables Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel A-1 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements-- :: " Decay time FollowingBurnup to 39 GWd/MTU .._-- --_ __-_... .Nuclide T=0 1 see 30 min 1 hr 8 hr 1.0 d 4.0d 30.0d .90.0 d 180.0 d 1 yr i 3 yr H 3 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.58E+04 7.57E+04 7.55E+04 7.47E+04 7.37E1404 7.17E+04 6.40E404 C 15 1.951+01 1.47E+01 0.00E+00 O.OOE+00 0.00+E00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 O.00E+00 N 16 1.732+02 1.57E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 .O.E+00 0.002+00 0.00E+00 Ne 23 3.46E201 3.39E+01 9.78E-14. 2.79E-28 0.OOE+00 0.OOE+00 0.00E+00 0.001E00 0.002+00 O.OOE+00 0.00E+00 0.00E+00 Na 24 2.731+02 2.73E+02 2.66E202 2.60E+02 1.86E+02 8.793+01 2.91E400 4.46E-13 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Na 25 2.09E+01 2.06E+01 1.70E-08 1.37E-17 0.00E+00 0.00E+00 0.002+00 0.0013+00 0.00 0.01E+00 0.OOE+00 0.00E+00 Mg 27 1.052+03 1.05E+03 1.16E+02 1.29E+01 5.61E-13 O.OOE+00 0.OOE+00 0.00E+00 0.00E+00 .003E+00 0.00E+00 0.OOE+00 Al 28 1.90E+04 1.89E+04 1.77E+00 1.65E-04 4.35E-07 2.56E-07 .2.35E-08 2.45E-17 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 1128 1.19E+06 1.18E+06 5.17E+05 2.25E+05 1.96E+00 5.29E-12 0.00E+00 0.00E+00. 0.002+00 0.0E+00 0.00+E00. 0.00E+00 Al 29 5.46E+01 5.45E+01 2.30E+00 9.70E-02 5.41E-21 0.002+00 0.00+E00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 Si 31 4.08E+02 4.08E+02 3.58E+02 3.13E+02 4.92E+01 7.16E-01 3.88E-09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ti 51 5.86E+01 5.84E+01 1.58E+00 4.28E-02 0.001+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Cr51 5.65E+06 5.65E+06 5.65E+06 5.65E406 5.61E406 5.522406 5.1 IE+06 2.67E+06 5.95E+05 6.26E+04 6.08E+02 7.04E-06 V 52 8.56E+04 8.48E+04 3.34E-+02 1.30E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 O.00E+00 0.00E+00 0.00E-00 V 53 3.541+02 3.52E+02 8.71E-04 2.14E-09 O.00E+00 0.00E+00 0.00E+00 0.00E400 0.OOE+00 0.OOE+00 0.OOE+00 0.001+00 Mn 54 4.27E+05 4.271+05 4.27E+05 4.27E+05 4.27E+05 4.26E+05 4.231+05 4.00E+05 3.50E+05 2.87E+05 1.90E+05 3.75E+04 Cr 55 8.632+04 8.56E+04 2.25E+02 5.88E-01 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 O.OOE+00 Fe 55 1.902+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.90E+06 1.86E+06 1.79E+06 1.68E+06 1.48E+06 8.86E+05 Mn 56 1.21 E+07 1.21 E+07 1.05E+07 9.17E+06 1.40E+06 1.90E+04 7.47E-05 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Mn 57 1.361+03 1.351+03 8.33E-04 5.07E-10 0.00E+00 .0E 0.00+00 0.0E00 0.002+00 0.00E+00 0.00E+00 0.OOE+00 Co 58 5.91E+05 5.91E+05 5.91E+05 5.91E+05 5.90E+05 5.86E+05 5.68E+05 4.41E+05 2.45E+05 1.02E+05 1.67E+04 1.32E+01 Fe 59 1.28E+05 1.28E405 1.282+05 1.28E+05 1.28E+05 1.27E+05 1.21E+05 8.02E+04 3.161+04 7.79E+03 4.35E+02 4.97E-03 Ni 59 2.582+02 2.582+02 2.58E+02 2.58E+02 2.582+02 2.582+02 2.582+02 2.582+02 2.58E+02. 2.58E+02 2.5823+02 2.58E+02 Co 60 3.191+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.19E+05 3.192+05. 3.162+05 3.092+05 2.992+05 2.802+05 2.151+05 Co 60m 5.272+05 5.262+05 7.24E+04 9.93E+03 8.332-09 0.00+E00 0.002+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.00E+00 Co 61 7.11 E+03 7.11E+03 5.76E+03 4.67E+03 2.472+02 2.97E-01 '2.17E-14 0.00E+00 0.002-+00 0.00+E00 0.00E+00 0.002+00 Co 62 5.662+01 5.622+01 5.4 1E-05 5.14E-I I 0.00E+00 0.00E400 0.00+E00 0.002+00 0.002+00 0.002+00 0.00E+00 0.00E+00 Ni 63 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.55E+04 3.552+04 3.55E+04 3.55E+04 3.54E+04 3.54E+04 3.532+04 3A482+04 A-2 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements-. ... -,-,Decay time Following Burnup to 39 GWdIMTU* ___ ___Nuclide: :T0 --sec- 30ri : 1 hrT 8 I.0d ..4.0d 30.0d0 90.Od -80.0 d yr :3 3 yr Cu 64 2.65E+02 2.65E+02 2.58E+02 2.5 1E+02 1.72E+02 7.17E+O1 1.4 1E+00 2.28E-15 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Ni 65 5.57E+04 5.57E+04 4.86E+04 4.23E+04 6.17E+03 7.57E401 1.89E-07 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Cu 66 3.48E+02 3.48E402 6.30E+00 4.91E-01 3.58E-01 2.93E-01 1.17E-01 4.26E-05 4.92E-13 6.10E-25 O.OOE+00 O.OOE+00 Cu 73 2.80E+03 2.48E+03 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Zn 73 4.88E+03 4.82E+03 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+O0 O.OOE+00 O.OOE+00 Ga 73 4.97E403 4.97E+03 4.64E+03 4.32E+03 1.59E+03 1.63E+02 5.65E-03 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Ge 73m 4.91 E+03 4.91 E+03 4.58E+03 4.26E+03 1.57E+03 1.60E402 5.58E-03 O.OOE+00 O.OOE+O0 O.OOE+00 O.OOE+00 O.OOE+00 Cu 74 3.48E+03 1.32E+03 O.OOE+00 O.OOEs00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Zn 74 I.I8E+04 1.I8E+04 2.70E-02 6.10E-08 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Ga 74 3.73E403 3.73E+03 3.45E+02 2.66E+01 7.15E-15 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Cu 75 4.40E+03 2.10E+03 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Zn 75 2.75E+04 2.59E+04 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Ga 75 3.37E+04 3.36E+04 1.81E+00 9.02E-05 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Ge 75 3.39E+04 3.39E+04 2.71E+04 2.11 E+04 6.26E+02 2.02E-01 3.96E- 17 O.OOE+00 O.OOE+00 O.OOE+O0 O.OOE+O0 O.OOE+00 Cu 76 3.56E+03 2.58E+02 O.OOE+00 O.OOE÷00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+O0 O.OOE+00 O.OOE+00 Zn 76 6.04E+04 5.34E+04 O.OOE+00 O.OOE+00 O.OOE÷00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Ga 76 9.02E+04 8.94E+04 2.48E-12 5.23E-29 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 As 76 2.56E+03 2.56E+03 2.53E+03 2.50E+03 2.08E+03 1.36E+03 2.05E+02 1.49E-05 5.05E-22 O.OOE+O0 O.OOE+00 O.OOE+00 Cu 77 2.08E+03 2.14E+02 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE400 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 Zn 77 7.37E+04 5.30E+04 0.OOE+00 O.OOE+00 .OOE+00 .OOE-f O.OOE+00 .OOE+00 .OOE+00 .OOE-f00 O.00E+00 0.OOE+00 Ga 77 1.88E+05 1.82E+05 0.00E+00 0.00E400 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Ge 77 7.31 E404 7.31E+04 7.10E+04 6.88E+04 4.48E+04 1.68E+04 2.02E+02 4.83E-15 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Ge 77m 1.93E+05 1.93E+05 1.49E-05 8.48E-16 0.OOE+00 O.00E400 0.OOE+00 0.00E+00 O.OOE+00 ..OOE+00 O.OOE+00 s.OOE400 As 77 2.26E+05 2.26E+05 2.25E+05 2.23E+05 2.04E+05 1.60E+05 4.61E+04 6.72E-01 4.61E-12 8.71 E-29 0.00E+00 0.00E+00 Zn 78 9.47E+04 5.92E+04 0.00E+00 0.00E+00 0.OOE+00 0.00E400 O.OOE+00 O.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 Ga 78 4.25E+05 3.80E+05 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Ge 78 7.22E+05 7.22E+05 5.70E+05 4.50E+05 1.65E+04 8.56E+00 1.43E-14 0.00E+00 0.00E+00 0.OOE+00 0.00E400 0.00E+00 As 78 7.32E+05 7.32E+05 7.13E+05 6.70E+05 8.25E+04 1.25E402 1.39E-12 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 A-3 Safety Analysis Report for Susquehanna Units 1 and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-i Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU ________ ________Nuclide. T -=.0 -I sec:-.--: .30 min-:- I hr = 8hr--- ..Od: 4.0 30.0.d '90.0 d 180.0 d .1 yr 3 yr Zn 79 4.45E404 2.22E404 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Ga 79 4.19E+05 3.39E+05 0.00E+00 0.00E400 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Ge 79 1.24E+06 1.21E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 " 0.00E+00 As 79 1.34E+06 1.34E+06 1.38E+05 1.38E+04 1.28E-10 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 Se 79m 1.34E+06 1.34E+06 2.37E+05 2A IE+04 2.24E-10 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 O0.OOE+00 0.01+00 .003E+00 Zn 80 1.67E+04 4.611E+03 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00.Ga 80 3.39E+05 2.26E+05 0.OOE+00 0.OOE+00 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 Ge80 3.03E+06 2.96E+06 1.31E-12 0.00E+00 0.OOE+00 0.001E+00 0.00E400 0.OOE+00 0.00E+00 0.00E+00 0.001E+00 0.OOE+00 As 80 3.53E+06 3.5 11E+06 2.711E-12 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Zn 81 3.77E+03 1.31E÷+01 0.001E+00 0.00E+00 O.OOE+00 0.OOE400 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Ga 81 2.13E+05 1.211E305 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.00E3+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 GeS8 3.211E+06 2.94E406 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 I0.00E+00 0.00E+00 0.OOE+00 As 81 5.12E+06 5.07E+06 3.26E-10 1.73E-26 0.00E+00 O.00E+00 O.OOE+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 81 5.39E+06 5.39E+06 2.02E+06 8.02E+05 1.71E+03 1.54E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 81m 3.87E+05 3.87E+05 2.70E+05 1.88E+05 1.161E+03 1.04E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Ga 82 1.34E+05 4.20E+04 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ge 82 3.08E+06 2.66E+06 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 As 82 5.03E+06 4.96E+06 0.00E+00 0.00E+00 O.OOE+00 0.00E400 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 82m 1.96E+06 1.86E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.001E+00 0.00E+00 0.00E+00 0.00E+00 Br82 3.811E+05 3.81E+05 3.78E405 3.75E+05 3.27E+05 2.38E+05 5.811E+04 2.77E-01 1.46E-13 0.00E+00 0.00E+00 0.00E+00 Br 82m 3.29E+05 3.29E+05 1. 11E+04 3.72E1+02 8.86E-19 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 Ga 83 1.60E+04 1.70E+03 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 O.00E+00 0.00E+00 Ge 83 1.38E+06 9.63E1+05 0.OOE+00 O.001E00 O.OOE+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 83 8.02E+06 7.64E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.001E+00 0.00E+00 0.00E+00 Se 83 6.10E+06 6.101E+06 2.41E+06 9.47E+05 2.03E+00 2.23E-13 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 83m 6.45E+06 6.43E+06 1.46E-01 2.70E-09 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br 83 1.29E+07 1.29E+07 1.18E+07 1.04E+07 1.40E+06 1.38E+04 1.28E-05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 83m 1.311E+07 1.31E+07 1.29E+07 1.26E+07 3.64E+06 5.27E+04 5.39E-05 0.00E+00 0.001E+00 0.001E+00 0.00E+00 0.00E+00 A-4 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_______ Decay time oll owing Burnup to 3 ~IT Nuclide eT=0- Sec 30mml.7: 1 hr:.. ghr...: 1.0 d 4.0 d 30.Od.::.:.: .90.0 d 180.0d 1 yr 3yr Ga 84 1.28E+05 1.09E+02 0.00+E00 0.00E400 0.00+E00 0.00E400 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 Ge 84 8.86E+05 5.02E+05 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+0O 0.00E+00. 0.OOE+00 0.OOE+00 0.00E+00 O.00E+00 As 84 5.9 1E+06 5.29E+06 0.00+E00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Se 84 2.34E+07 2.33E+07 3.55E+04 5.34E401 0.OOE+00 0.00E400 0.OOE+00 0.000E+0 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 Br 84 2.40E+07 2AOE+07 1.38E+07 7.20E+06 7.62E+02 6.22E-07 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Br84m 6.43E+05 6.42E+05 2.01E+04 6.27E+02 5.32E-19 0.00E400 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 Ge 85 1,53E+05 9.47E+03 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 As 85 3.40E+06 2.43E406 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Se 85 1.10E+07 1.08E+07 9.02E-1I 7.15E-28 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 Se 85m 9.70E+06 9.32E+06 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Br85 2,67E+07 2.67E+07 2.17E+04 1.54E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 85 1.48E+06 1.48E+06 1.48E+06 1.48E406 1.48E+06 1.48E+06 1.48E+06 1.47E+06 I1.46E+06 1.44E+06 1.39E+06 1.22E+06 Kr 85m 2.68E+07 2.68E+07 2.51E+07 2.32E+07 7.87E+06 6.62E-05 9.63E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Ge 86 3.20E+04 1.92E+03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 As 86 1,89E+06 8.86E+05 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Se 86 2.64E407 2.52E+07 O.OOE+00 O.OOE+00 O.00E+00 O.OOE400 O.OOE+00 O.00E400 0.OOE+00 O.OOE+00 O.00E-O.00E-00 Br 86 3.25E+07 3.25E+07 6.28E-03 9.17E-13 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Br 86m 6.21E+06 5.33E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00+E00 Rb86 2.17E+05 2.17E+05 2.16E+05 2.16E+05 2.14E405 2.09E+05 1.86E+05 7.10+E04 7.62E+03 2.68E+02 2.73E-01 4.31E-13 Rb 86m 1.79E+04 1.76E+04 2.36E-05 3.09E-14 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 0.00+E00 O.00E+00 Ge 87 2.70E+05 1.51 E+03 0.00E+00 0.OOE+00 0.002E00 0.00+E00 0.00E+00 0.002+00 0.000000 0 0.00E+00 0.OOE+00 As 87 1.08E+06 1.25E+05 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Se 87 1.54E+07 1.36E+07 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 Br 87 4.24E+07 4.20E+07 8.25E-03 1.52E-12 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.00+E00 O.00E+00 Kr87 5.37E+07 5.37E+07 4.13E+07 3.15E+07 6.94E405 1.13E402 1.03E-15 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 As 88 4.79E+05 2.78E+03 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 Se 88 8. 817E+06 5.18E+06 0.00E+00 0.00E-00 0.00E+00 0.002E+00 0.00E200 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Br 88 4.09E+07 3.95E+07 0.00+E00 O.00+E00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 A-5 Sarety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-i Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements-Decay timei olowing Burnup to 39.... ..d.T ........... Nuclide T=: .1sec .30min I. lhr h0d' 4.0d. :30.0d' 90.0d' 180.0d j lyr 3yr.Kr 88 7.45E+07 7.45E+07 6.60E+07 5.84E+07 1.05E+07 2.12E+05 4.93E-03 0.00+E00 0.00+E00 0.00E+00 0.OOE+00 0.00+E00 Rb 88 7.64E+07 7.64E+07 7.16E+07 6.46E+07 1.18E+07 2.38E+05 5.5 1E-03 0.00+E00 0.00E+00 0.OOE+00 0.OOE+00 0.00+E00 Zr 89 5.52E+04 5.52E+04 5.50E+04 5.48E+04 5.15E+04 4.47E+04 2.37E+04 9.55E+01 2.84E-04 1.46E-12 1.74E-29 0.00+E00 As 89 8.56E+03 2.77E+01 0.00E+00 0.00+E00 0.00+E00 0.00E400 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 89 2.86E+06 5.25E+05 0.00+E00 0.00E400 0.OOE+00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Br 89 2.83E+07 2.44E+07 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.00E400 0.00E+00 0.00E+00 0.00E400 0.OOE+00 0.002E00 Kr 89 9.24E-07 9.24E+07 1.32E+05 1.86E+02 0.00E+00 0.00+E00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Rb 89 9.93E+07 9.93E+07 3.16E+07 8.02E+06 3.87E-02 3.76E-21 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 Sr89 1.03E+08 1.03E+08 1.03E+08 1.03E+08 1.03E+08 1.02E+08 9.78E+07 6.85E+07 3.01E+07 8.79E+06 6.91E+05 3.09E401 Y 89m 1.68E+05 1.63E+05 6.43E+04 6.40E+04 6.07E+04 5.39E+04 3.25E+04 6.46E+03 2.80E+03 8.18E+02 6.43E+01 2.87E-03.Se 90 6.30E+05 1.24E+05 0.00+E00 O.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Br 90 1.54E407 1.08E+07 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 90 9.93E+07 9.70E+07 1.73E-09 2.92E-26 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Rb 90 9.17E+07 9.17E+07 4.70E+04 1.23E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 Rb 90m 2.91E+07 2.90E+07 2.44E+05 1.94E+03 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00+E00 0.00+E00 Sr 90 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.31E+07 1.30E+07 1.30E+07 1.29E+07 1.27E+07 1.21E+07 Y 90 1.36E+07 1.36E+07 1.36E+07 1.36E+07 1.35E+07 1.34E+07 1.32E+07 1.30E+07 1.30E+07 1.29E+07 1.27E+07 1.21E+07 Zr 90m 8.86E+03 3.77E+03 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.002E00 0.OOE+00 0.OOE+00 Se 91 6.3 1E+04 4.82E+03 0.00E+00 O.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Br 91 5.03E+06 1.59E+06 0.OOE+00 O.00+E00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Kr 91 6.79E+07 6.28E+07 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Rb 91 1.21E+08 1.21E+08 7.04E-02 3.68E-1 I 0.00+E00 0.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Sr91 1.31E+08 1.31E+08 1.26E+08 1.21E+08 7.30E+07 2.28E+07 1.21E+05 2.23E-15 0.OOE+00 0.OOE+00 0.00+E00 0.OOE+00 Y 91 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.34E+08 1.34E408 1.29E+08 9.48E+07 4.65E+07 1.60E+07 1.78E+06 3.11 E+02 Y 91m 7.57E+07 7.57E+07 7.53E+07 7.40E+07 4.64E+07 1.44E+07 7.64E+04 IAIE-15 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Se 92 5.17E+03 8.33E+01 *0.00E+00 O.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Br 92 8.56E+05 1.28E+05 0.OOE+00 0.00E400 0.OOE+00 0.00+E00 0.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Kr 92 3.61E+07 2A9E+07 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 A-6 Sarety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements.: -Decay time Following Burnup to 39 GWd/MTU ".Nuclide .T 0;,.. I sec 30mini -1 hr- 8 hr: 1.0d 4.0 d 30.0 d 90.0 d 180.0d .d yr : 3 yr Rb 92 1.08E+08 9.63E+07 0.00E+00 0.00+E00 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Sr 92 1.39E+08 1.39E+08 1.22E+08 1.08E+08 1.80E+07 2.99E405 3.01E-03 0.00E+00. 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 Y 92 1 1.40E+08 1.40E+08 1.39E+08 1.37E+08 6.53E+07 4.42E+06 4.04E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 Zr93 2.97E+01 2.97E+01 2.97E+01 2.97E+0I 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 2.97E+01 Br 93 2.60E+05 5.06E+03 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 Kr 93 1.22E+07 7.14E+06' 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 Rb 93 8.94E+07 8.02E+07 0.OOE+00 0.00E+00 O.00+E00 0.00E400 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Sr 93 1.57E408 1.57E+08 9.63E+06 5.84E+05 5.42E-12 0.OOE+00 O.OOE+00 .OOE+00 0.000+00 .OOE+00 0.OOE+0O 0.00E+00 Y 93 1.07E+08 1.07E+08 1.04E+08 1.01E+08 6.23E+07 2.08E+07 1.48E+05 3.74E-14 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 Br 94 1.23E+04 2.33E+01 0.00E+00 0.OOE+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 O.OOE+00 Kr 94 5.64E+06 2.07E+05 0.OOE+00 O.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 94 4.63E+07 3.61 E+07 0.00E+00 O.00+E00 0.00E+O0 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 Sr 94 1.57E+08 1.56E+08 9.93E+00 6.13E-07 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 94 1.70E+08 1.70E+08 5.95E+07 1.96E+07 3.39E+00 1.19E-15 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 Kr 95 5.39E+05 2.21 E+05 0.002E00 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 95 2.25E+07 3.91E+06 O.OOE+00 O.OOE+00 O.00+E00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Sr 95 1.41E+08 1.38E+08* 3.67E-14 0.OOE+00 O.00+E00 0.00+E00 0.00E+00 0.00E+00. 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Y 95 1.76E+08 1.76E+08 2.52E+07 3.48E+06 3.16E-06 0.00+E00 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 O.00+E00 Zr95 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.91E+08 1.89E+08 1.84E+08 1.39E408 7.23E+07 2.73E+07 3.67E+06 1.35E+03 Nb95 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1.92E+08 1,92E408 1.92E+08 1.79E+08 1.21E+08 5.36E+07 7.92E+06 2.98E+03 Nb 95m 2.13E+06 2.13E+06 2.13E+06 2.13E+06 2.13E+06 2.12E+06 2.10E+06 1.63E+06 8.50E+05 3.21E+05 4.32E+04 1.59E+01 Kr 96 9.17E+04 8.56E+03 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 .0.00E+00 Rb 96 5.62E+06 1.87E+05 0.OOE+00 0.OOE+00 0.00+E00 0.OOE+00 O.00+E00 O.00E+00 0.OOE+00 O.00E+00 O.00+E00 O.OOE+00 Sr 96 1.03E+08 5.43E+07 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 O.00E+00 0.00E+00 Y 96 1.70E+08 1.60E+08 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00+E00 Nb 96 3.12E+05 3.12E+05 3.07E+05 3.02E+05 2.45E+05 1.53E+05 1.80E+04 1.63E-04 4.44E-23 0.00E+00 0.00E+00 0.00E+00 Rb 97 1.54E+06 2.70E+04 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 Sr 97 5.23E+07 1.02E+07 0.00+E00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 " 0.00E+00 0.OOE200 A-7 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-i Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements... Decay time Following Burnuo to 39 GWd/MTU Nuclide- sec- [30mn-.ihr-8hr-." 1.0d=-- d---- 180.0 d I yr,-., yr Y 97 1.40E+08 I.18E+08 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Zr 97 1.90E408 1.90E+08 1.86E+08 1.82E+08 1.36E+08 7.09E+07 3.70E+06 2.84E-05 0.00E+00 O.00E+00 0.00E+00 O.00E+00 Nb97 1.91E+08 1.91E+08 1.90E+08 1.88E+08 1.46E+08 7.13E+07 3.72E+06 3.06E-05 0.00E+00 0.00E÷00 0.00E+00 0.00E+00 Nb 97m 1.80E+08 1.80E+08 1.76E+08 1.73E+08 1.30E+08 6.73E+07 3.51 E+06 2.70E-05 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Rb 98 1.54E+05 3.64E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr 98 2.13E+07 7.33E+06 0.00E+00 0.00E+00 0.00E+00 O.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 98 1.04E+08 4.30E+07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 Zr 98 1.84E+08 1.82E+08 4.22E-1 0 9.24E-28 0.00E+00 0.00OE+00 ..0+E+00 0.OOE+00 0..0E+00 0..0E+00 0.00E+00 0.00E+00 Nb 98 1.86E+08 1.86E+08 4.65E-10 1.01E-27 0.00E+00 O.00E+00 *0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Nb 98m 1.47E+06 1.47E+06 9.78E+05 6.52E+05 2.24E+03 5.20E-03 0.00E+00 0.OOE+00 0.001E+00 0.OOE+00 0.00E+00. 0.00E+00 Rb 99 4.46E+03 3.44E-02 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 0.00E+00 Sr 99 8.17E+06 6.33E+05 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Y 99 6.59E+07 4.21 E+07 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 0.00E+00 Zr 99 1.82E+08 1.45E+08 0.00E+00 0.00E400 O.00E+00 0.00E+00 O.OOE+00 O.00E+00 O.OOE+00 O.00E+00 0.OOE+00 0.00E+00 Nb 99 1.18E+08 1.17E+08 O.00E+00 O.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 Nb 99m 8.10E+07 8.10E+07 2.76E+04 9.24E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E400 0.OOE+00 Mo99 2.02E+08 2.02E+08 2.01E+08 2.OOE+08 1.86E+08 1.57E+08 7.37E+07 1.05E+05 2.79E-02 3.84E-12 0.00E+00 .0.00E+00 TC 99 2.21E+03 2.21 E+03 2.211E+03 2.21E+03 2.21E+03 2.211E+03 2.21E+03 2.22E+03 2.22E+03 2.221E+03 2.221E+03 2.22E+03 Tc 99m 1.79E+08 1.79E408 1.79E+08 1.79E+08 1.73E+08 1.51E+08 7.14E+07 1.01E+05 2.70E-02 3.72E-12 O.00E+00 0.OOE+00 Sri00 1.08E+06 3.47E+04 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Y 100 2.13E+07 8.40E+06 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Zr1O0 1.79E+08 1.63E+08 0.OOE+00 0.00E+00 0.OOE+00 0.003E400 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 NbI00 1.94E+08 1.86E+08 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 NbI00m 1.60E+07 1.27E+07 0.00E+00 0.OOE+00 0.00E+00 0.00E3+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 TCe00 5.05E+07 4.83E+07 0.00E+00 0.00E+00 0.OOE+O0 0.00E+00 0.00E+00 0.00E+00 .001E+00 0.OOE+00 0.OOE+00 0.00E+00 SrOI 1.60E+05 4.45E+03 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 .001E+00 0.00E+00 YI01 1.05E+07 2.64E+06 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 ZrIO1 1.04E+08 7.52E+07 0.00E+00 O.00E+00 O.OOE+00 O.00E+00 O.00E+00 O.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 0.00E+00 A-8 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version _Table A-i Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements-Decay time. Following Burnup to 39 GWd/MTU ::______.. -______Nuclide- T- 0 Asec _-30 min I1 hr.- 8 hr-- --- 1.0 d 4.0 d 30.d .90.0d I:8.d Iyr 3r.Nbl01 1.74E+08 1.66E+08 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Mol01 1.83E+08 1.83E+08 4.45E+07 1.07E+07 2.35E-02 3.77E-22 0.00E+00 0.OOE+00. 0.00E+00 0.002+00 0.OOE+00 0.00E+00 TcIOM 1.83E+08 1.83E+08 1.06E+08 4.OOE+07 4.18E-01 1.18E-20 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Sr102 2.66E+04 2.37E+03 O.OOE+00 O.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 102 4.78E+06 2.21E+06 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 Zrl02 7.20E+07 5.73E+07 0.00E+00 O.00E+00 0.OOE+00 0.00+E00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Nb102 1.47E+08 1.12E+08 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 Mo102 1.74E+08 1.74E+08 2.77E+07 4.40E+06 2.85E-05 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Tcl02 1.74E+08 1.74E+08 2.80E+07 4.43E+06 2.87E-05 0.00E+00 0.00E+00 0.00E-f00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 TcI02m 1.77E+05 1.76E+05 1.49E+03 1.25E+01 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 0.00E+00 Y 103 1.41E+06 9.70E+04 O.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Zr103 2.67E+07 1.57E+07 0.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 Nbl03 1.09E+08 7.62E+07 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Mo103 1.69E+08 1.68E+08 1.61E+00 1.51E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Tcl03 1.71E+08 1.71E+08 8.10E+00 7.64E-08 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Rul03 1.72E+08 1.72E+08 1.72E+08 1.71E+08 1.70E+08 1.69E+08 1.60E408 1.01E+08 3.50E+07 7.15E+06 2.71E+05 6.80E-01 RhI03m 1.71E+08 1.71E+08 1.71E+09 1.71E+08 1.70E+08 1.68E+08 1.60E+08 1.01E÷f08 3.50E+07 7.14E+06 2.71E+05 6.79E-01 Y104 4.13E+04 1.86E+02 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 Zr104 8.02E+06 6.102+06 0.OOE+00 0.OOE+00 0.002E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 NblO4 5.13E+07 4.53E+07 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 MoI04 1.34E+08 1.34E+08 1.31E-01 I.21E-10 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 Tcl04 1.41E+08 1.41E+08 4.80E+07 1.54E+07 1.89E+00 3.06E-16 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Rh104 1.01E+08 9.931+07 7.30E+04 6.06E+02 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 RhI04m 7.38E+06 7.36E+06 6.13E+04 5.08E+02 0.OOE+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 O.00E+00 0.00E+00 0.OOE+00 Zr105 2.19E+06 5.35E+05 0.OOE+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00+E00 Nbl05 2.07E+07 1.66E+07 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 0.00+E00 Mo105 9.86E+07 9.70E+07 6.09E-08 3.62E-23 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Tc105 I.17E+08 1.17E+08 8.17E+06 5.28E+05 1.22E-I1 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 A-9 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time. Following Burnup to 39 GWdIMTU __________ Nuclide" T,0_. 1 sec.. 30min 1,hr .8hr.. 1.0d 4.0d 30.Od 90.Od 180.0d Ilyr I. 3yr Ru105 1.19E+08 1.19E+08 1.13E+08 1.05E+08 3.51E+07 2.89E+06 3.78E+01 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 Rhi05 1.11 E+08 1.11E408 1.12E+08 1.12E+08 1.05E+08 8.02E407 1.96E+07 9.55E+01. 5.26E-1 I 1.74E-29 0.00E+00 0.00E+00 RhI05m 3.38E+07 3.38E+07 3.22E+07 2.99E+07 1.00E+07 8.25E+05 1.08E+01 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Zrl06 1.12E+05 5.17E+04 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00+E00 Nbl06 4.20E+06 2.13E+06 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 O.00E+00 Mo106 5.62E+07 5.20E+07 0.00E400 0.00E+00 0.00+E00 O.OOE+00 0.00+E00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.OOE+00 Tcl06 8.48E+07 8.40E+07 9.09E-08 8.02E-23 0.OOE+00 0.00E4 00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Rul06 6.85E+07 6.85E+07 6.85E+07 6.85E+07 6.85E+07 6.84E+07 6.80E+07 6.48E+07 5.79E+07 4.90E+07 3.47E+07 8.86E+06 Rhl06 7.38E+07 7.37E+07 6.85E+07 6.85E+07 6.85E+07 6.84E+07 6.80E+07 6.48E+07 5.79E+07 4.90E+07 3.47E+07 8.86E+06 Rhl06m 2.39E+06 2.39E+06 2.04E+06 1.74E+06 1.85E+05 1.11E+03 1.1OE-07 0.OOE+00 0.OOE+00 0.OOE+00 0.00+E00 0.OOE+00 Nbl07 7.72E+05 3.11E+05 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 O.OOE+00 O.OOE0+00 0.0E+00 Mo107 2.33E+07 1.92E+07 0.00E+00 0.00+E00 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 Tcl07 5.83E+07 5.71 E+07 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 Rul07 6.97E+07 6.96E+07 2.98E+05 1.16E+03 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 Rhl07 6.98E+07 6.98E+07 3.29E+07 1.26E+07 1.88E+01 9.09E-13 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Pdl07m 5.43E+05 5.26E+05 0.00E+00 O.00E+00 0.OOE+00 0.00+E00 O.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.00+E00 Nbl08 2.41E+04 1.38E+03 0.00E+00 O.00E+00 0.OOE+00 0.00+E00 O.OOE+00 O.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Mo108 3.54E+06 2.23E+06 0.00+E00 O.00+E00 0.00+E00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 Tc108 2.07E+07 1.84E+07 O.OOE+00 O.00+E00 0.OOE+00 0.00+E00 O.OOE+00 .OOE+00 0.006+00 0.00E+00 0.OOE+00 0.OOE+00 Rul08 4.49E+07 4.48E+07 4.70E+05 4.86E+03 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 O.006+00 0.006+00 0.OOE+00 0.OOE+00 Rhl08 4.56E+07 4.55E+07 5.00+E05 5.17E+03 0.OOE+00 0.00E+00 0.00E+00 0.00E400 0.00+E00 0.00E+00 O.00E+00 0.00E+00 Rhl08m 6.83E+05 6.82E+05 2.14E+04 6.67E+02 5.65E-19 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 Mo109 3.69E+05 2.25E+05 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.OOE+00 O.OOE+00 Tcl09 6.89E+06 4.3 1E+06 O.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE400 0.00+E00 0.00E+00 O.00E+00 0.OOE+00 0.00E400 Rul09 3.00E+07 2.95E+07 1.OIE-08 3.28E-24 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Rhl09 3.48E+07 3A7E+07 1.80E+01 3.03E-06 0.OOE+00 0.00E400 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 O.OOE+00 Rhl09m 1.73E+07 1.73E+07 7.72E-04 1.12E-14 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Pd109 4.14E+07 4.14E+07 4.04E407 3.94E+07 2.77E+07 1.23E407 3.22E+05 6.26E-09 0.00E400 0.00E+00 0.00+E00 0.00E+00 A-10 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements________ -__ _ __ ......__... Decay time Following Burnup to 39 GWd/MTU Nuclide T=0".' 1 sec 30mifi 1 hr' 8h"i' 1.-0d d 4.0 d: 30.0d '_' 90.0dd 180.0d 1 yr- 3 yr Pd109m 2.221+05 2.22E+05 2.64E+03 3.13E+01 0.00E+00 0.002+00 0.001+00 0.002+00 0.0013400 0.00E+00 0.00E+00 0.001+00 AgI09m 4.14E+07 4.14E+07 4.04E+07 3.94E407 2.77E+07 1.23E+07 3.22E+05 7.79E-02 7.12E-02 6.221-02 4.71E-02 1.571-02 Mo110 3.80E404 2.96E+04 0.00E+00 0.00E+00 0.001+00 0.001+00 0.001300 0.001E+00 0.00E+00 0.00E+00 0.001+00 Tc 110 1. 11E+06 5.00E+05 0.001+00 0.00E+00 0.001+00 0.00E+00 0.00E400 0.001+00 0.002+00 0.002+00 0.OOE+00 0.00E+00 Ru110 9.78E+06 9.40E+06 0.OOE+00 0.002+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 O.00E+00 0.002+00 0.002+00 0.001+00 Rh! I0 1.56E+06 1.252+06 0.00E+00 0.002+00 0.OOE+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rh!l0mn 1.14E+07 1.13E+07 2.192-12 0.001+00 0.00E+00 0.001+00 0.00E+00 0.00E+00 0.00E+00 0.001+00 0.002+00 0.00E+00 AgI10 1.872+07 1.82E+07 6.23E+03 6.23E+03 6.23E+03 6.22E+03 6.171+03 5.74E+03 4.86E+03 3.782+03 2.26E+03 2.98E+02 Ag1lOm 4.58E+05 4.582+05 4.58E+05 4.582+05 4.58E+05 4.572+05 4.532+05 4.222405 3.572405 2.78E405 1.672+05 2.192+04 Mo0 11 3.11 E+03 7.01E+02 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.001E+00 0.002+00 0.002+00 0.002+00 Tc 11 2.11 E+05 1.49E+05 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.001+00 0.003+00 0.001+00 0.002+00 0.002+00 Rul!! 3.272+06 2.182+06 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.0023+00 Rhl!! 6.24E+06 6.032+06 0.002+00 0.002+00 0.00E+00 0.00E+00 0.0023400 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 Pdl I I 6.662+06 6.662+06 2.882+06 1.292+06 8.102+04 1.082+04 1.242+00 0.001+00 O.00E+00 0.002+00 0.00E+00 0.00E+00 Pd I im 2.832+05 2.832+05 2.65E+05 2.491+05 1.032+05 1.38E+04 1.572+00 0.002+00 0.002+00 0.002+00 0.002+00 0.00+E00 AgI!I 6.72E+06 6.72E+06 6.71E+06 6.70E+06 6.52E+06 6.13E+06 4.652+06 4.132+05 1.56E+03 3.592-01 1.18E-08 O.00E+00 AglIImn 6.71E+06 6.712+06 3.042+06 1.39E+06 1.01E+05 1.34E+04 1.54E+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.00E+00 CdI IIm 2.502+04 2.502+04 1.632+04 1.06E+04 2.662+01 3.01E-05 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002400 Tc 112 3.21 E+04 6.592+03 0.002-400 0.002+00 0.002+00 0.004E00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.00E+00 Ru! 12 1.052+06 8.71E+05 0.002+00 O.00E+00 0.002+00 0.001+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.00E+00 Rh! 12 2.472+06 1.902+06 0.002+00 O.00+E00 0.002+00 0.00+E00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 Pd! 12 3.02E+06 3.022+06 2.97E+06 2.92E+06 2.322+06 1.372+06 1.28E+05 1.5 1E-04 3.822-25 0.002+00 0.002+00 0.002+00 Ag1l2 3.032+06 3.03E+06 3.03E+06 3.022+06 2.642+06 1.601+06 1.51E+05 1.782-04 4.492-25 0.002+00 0.001+00 0.002+00 In B13m 3.472+05 3.472+05 3.472+05 3.47E+05 3.462+05 3.452+05 3.39E+05 2.902+05 2.02E+05 1.182+05 3.85E+04' 4.731+02 Sni 13 3.472405 3.472+05 3.4723+05 3.472+05 3.462+05 3.452+05 3.382+05 2.902+05 2.02E+05 1.182+05 3.84E+04 4.722+02 SnI13mn 1.132+05 1.132+05 4.272404 1.62E+04 2.002-02 6.262-16 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 TcI 13 6.772+03 2.342+03 0.00+E00 O.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.00E+00 Ru! 13 3.252+05 2.592405 0.002+00 0.003+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 A-I1 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_______________________Decay timeFollowing Burnup to 39,GWd~iMTU' _____Nuclide T=O" 1 sec 30rmin I hr 8r he 1.0d ... 4.0d 30.0d -90.0d 180.0d lyr : 3yr RhI 13 1.181+06 7.02E+05 0.00+E00 0.00+E00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.002+00 0.001+00 0.002+00 Pd1 13 1.742+06 1.73E-+06 2.64E+00 3.91E-06 0.002+00 0.003+00 0.00E000 0.00E100. 0.00E+00 0.00E+00 0.002+00 0.00E+00 AgI 13 1.69E+06 1.692+06 1.60E+06 1.50E+06 6.051+05 7.64E404 7.05E+00 0.001+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag 1t3m 3.38E+05 3.38E+05 1.86E+00 2.77E-06 0.00E3+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.001+00 0.00E+00 Cd 113m 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.33E+03 4.331-+03 4.33E-+03 4.322+03 4.29E+03 4.2313+03 4.131+03 3.74E+03 In 114 5.632+04 5.62E+04 3.54E+04 3.54E+04 3.53E+04 3.50E+04 3.35E+04 2.332+04 1.01E+04 2.86E+03 2.132+02 7.72E-03 In I 14m 3.71E+04 3.71E+04 3.71E+04 3.71E+04 3.692+04 3.6513+04 3.51E+04 2.44E+04 1.051E+04 2.98E+03 2.231E+02 8.10E-03 Ru1 14 1.081+05 9.93E+04 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.002+00 0.OOE+00 0.00E+00 .003E+00 0.00+E00 0.OOE+00 Rh 114 5.751405 4.17E+05 0.00E+00 0.OOE+00 0.0023+00 0.00+00 0.003+00 0.002+00 0.OOE+00 0.00E+00 100.003E+00 0.00E+00 PdI 14 1.32E+06 1.3 1E+06 2.75E+02 5.65E-02 0.001+00 0.00E+00 0.00E+00 0.00E+00 0.002300 0.00E+00 0.00E+00 0.001+00 AgI 14 1.37E+06 1.36E+06 2.832+02 5.83E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.002+00 Ru1 15 2.322+04 1.051+04 0.00E+00 O.OOE+00 0.002+00 0.001+00 0.OOE+00 0.0013+00 0.000+00 0.OO1+00 0.00E+00 0.00E+00 Rh1 15 2.301+05 2.13E+05 0.00E+00 0.001+00 0.00E+00 0.00E+00 O.003E+00 .003E+00 0.00E+00 .003E+00 0.00E+00 0.002+00 Pd 115 7.951+05 7.79E+05 4.742-09 2.58E-23 0.OOE+00 0.003+00 0.001+00 0.OOE+00 0.00E+00 .00.E+00 0.00E+00 0.00E+00 AgI 15 6.26E+05 6.26E+05 2.28E+05 8.10E÷04 3.852-02 1.37E-16 0.00E+00 0.0013+00 0.00E+00 0.002+00 0.002+00 0.00E+00 Agi 15m 2.62E+05 2.602+05 2.43E-09 1.32E-23 O.00E+00 0.00E+00 0.001+00 0.00E100 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Cdli 15 9.02E+05 9.02E+05 8.942+05 8.94E+05 8.171+05 6.62E+05 2.612+05 8.02E+01" 6.262-07 4.321-19 0.OOE+00 0.00E+00 Cdt 15m 4.20E+04 4.201+04 4.19E+04 4.192+04 4.18E+04 4.13E+04 3.95E+04 2.64E+04 1.04E+04 2.561+03 1.44E+02 1.68E-03 In115m 9.02E+05 9.02E+05 9.02E+05 9.021+05 8.63E+05 7.21E+05 2.842+05 9.02E+01 1.15E+00 2.83E-01 1.59E-02 1.86E-07 Rut 16 7.14E+03 4.741+03 0.00E+00 0.00E+00 0.001+00 0.00E+00 0.00E+00 0.001+00 0.00E+00 0.002+00 0.002+00 0.OOE+00 RhI 16 1.08E+05 5.51E+04 0.00E+00 .003E+00 0.00E+00 0.00E+00 .003E+00 0.001+00 0.002+00 0.00E+00 * .003E+00 0.00E+00 Pdt 16 8.40E+05 8.021+05 0.002+00 O.OOE+0O O.O0E+00 0.00E+00 0.00E+00 0.01E+O0 0.002+00 0.001+00 0.00E+00 .0.00E+00 AgI 16 9.63E+05 9.63E+05 4.42E+02 1.892-01 0.00E+00 0.001+00 0.002+00 0.0013+00 0.001+00 0.00E+00 0.00E+00 0.002+00 Ag! 16m 1.2 1E+05 1.141+05 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.001+00 0.00E+00 0.00E+00 0.003+00 0.00E+00 0.00E3+00 In116 1.61E+05 1.54E+05 0.001+00 0.001+00 0.002+00 0.001+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.OOE+00 0.002+00 In1 16mn 6.062+05 6.06E+05 4.132+05 2.81 E+05 1.301+03 5.99E-03 0.002+00 0.00E+00 0.0013+00 0.002+00 0.002+00 0.002+00 RhI 17 4.202+04 2.39E+04 0.001+00 0.00+E00 0.002+00 0.003+00 0.002+00 0.00E+00 0.002+00 0.00E+00 0.002+00 0.002+00 Pd1 17 5.691+05 4.99E+05 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.00+E00 0.002+00 0.00+E00 0.001+00 0.001+00 A-12 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU ________ ___._"__" Nuclide T=0 I 1sec 30rmin 1 hr 8hr- 1.0d 4.0d 30.0d 90.d 180.d lyr : 3yr 7 Ag! 17 5.071+05 5.05E+05 1.91E-02 6.85E-10 0.00E+00 0.001+00 0.OOE+00 0.002+00 0.00E+00 0.00E+00 0.001+00 0.OOE+00 AgI 17m 5.07E+05 4.78E+05 0.00E+00 0.00E+00 0.001+00 0.001+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.001+00 Cd1 17 8.86E+05 8.86E+05 7.79E+05 6.76E+05 9.63E+04 1.122+03 2.22E-06 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 Cd1 17m 2.031+05 2.03E+05 1.832+05 1.66E+05 3.901+04 1.44E+03 5.13E-04 0.003+00 0.001+00 0.00E+00 0.001+00 0.00+E00 In! 17 6.62r+05 6.62E+05 6.562+05 6.38E+05 2.02E+05 4.12E+03 6.63E-04 0.001+00 0.OOE+00 O.00E+00 0.OOE+00 .001E+00 Inll7m 8.10E+05 8.102+05 8.02E+05 7.79E+05 2.342+05 4.13E+03 2.74E-05 0.00E+00 0.00E+00 0.003+00 0.00E+00 0.00E+00 Sn! 17m 2.492+06 2.49E+06 2.49E+06 2.481+06 2.441+06 2.36E+06 2.03E+06 5.392+05 2.53E+04 2.58E+02 2.04E-02 1.38E-18 Rh! 18 9.93 E+03 1.1 8E+03 0.002+00 O.00E+00 0.00E+00 0.00+E00 0.001+00 0.00E+00 0.00E+00 0.001+00 0.00E+00 0.00E+00 Pd 18 2.73E+05 2.19E+05 0.001+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 AgI 18 4.682+05 4.3 1E+05 0.001+00 0.00E+00 0.00E+00 0.0013+00 0.O0E+00 0.OOE+00 0.001+00 0.00E+00 0.00E+00 0.00E+00 Ag! 18m 3.32E+05 2.70E+05 0.002+00 0.00E+00 0.002+00 0.00+E00 0.00E+00 0.00E+00 0.002+00 0.002+00 0.001+00 0.00E+00 Cdl 18 8.172+05 8.172+05 5.43E+05 3.59E+05 1.102+03 1.982-03 0.00E+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.002+00 In! 18 8.182+05 8.181+05 5.44E+05 3.601+05 1.102+03 1.981-03 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 Rh! 19 3.151+03 7.11 E+02 0.0013+00 0.000+00 0.00E+00 0.000+00 0.002+00 0.00E+00 0.00E+00 0.OO1+00 0.00E+00 0.001+00 Pd! 19 1.282+05 8.63E404 0.001+00 0.OOE+00 0.00E+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.00E+00 Ag) 19 5.102+05 3.95E+05 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.002+00 Cd1 19 5.65E+05 5.641405 2.51E+02 1.10E-01 0.00E+00 0.001300 0.002+00 0.00E+00 0.00E+00 0.OOE+00 0.002+00 0.002+00 Cd1 19m 2.802+05 2.782+05 2.222+01 1.73E-03 0.002+00 0.002+00 0.0013+00 0.001+00 0.002+00 0.00E+00 0.002+00 0.00E+00 In 119 3.642+05 3.642+05 5.971+03 1.73E+03 1.63E-04 1.441-20 0.00E+00 0.001+00 0.002+00 0.002+00 0.002+00 0.002+00 In! 19m 5.122+05 5.122+05 1.892+05 5.97E+04 5.65E-03 4.982-19 0.00E+00 0.00+E00 0.00+E00 0.002+00 0.002+00 0.002+00 SnI19m 2.462+06 2.462+06 2.46E+06 2.46E+06 2.462+06 2.451+06 2.432+06 2.292+06 1.99E+06 1.60E+06 1.041+06 1.84E+05 Pd20" 7.522+04 6.30E+04 0.00E+00 0.00E+00 0.001+00 0.0013+00 0.002+00 0.001+00 0.001+00 0.002+00 0.002+00 0.00E+00 Ag120 3.57E+05 2.28E+05 0.002+00 0.002+00 0.001+00 0.00E+00 0.002+00 0.0011+00 O.OOE+00 0.002+00 O.00E+00 0.002+00 CdI20 8.02E+05 7.952+05 1.76E-05 3.77E-16 0.002+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.00E+00 0.001+00 In120 8.172+05 8.102+05 1.88E-05 4.021-16 O.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 InI20m 1.442+04 1.422+04 2.70E-08 5.OOE-20 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 PdI2I 3.06E+04 1.052+04 0.00E+00 0.001+00 0.00E+00 0.003400 0.002+00 0.002+00 0.001+00 0.002+00 0.002+00 0.001+00 Ag121 2.52E+05 1.162+05 0.00E+00 0.OOE+00 0.00+E00 0.002E00 0.002+00 0.002+00 0.002+00 0.001+00 0.002+00 0.002+00 A-13 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements-, .. ecay,'time. Following Burnup to 39, GWdIMTU..Nuclide- T=0--- A-see-c-- min-' lhri:-- 8 hrf- A :.Od -4.0d-7-7 -30.0d'- -90.0d'-- 1A80.04d- .lyr .... :3yf : CdU21 7.95E+05 7.64E405 0.00E+00 0.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Inl21 7.72E+04 7.5 1E+04 5.40E+01 2.54E-01 0.OOE+00 0.00+E00 0.00E+00 0.00E400 .0.OOE+00 0.00+E00 0.OOE+00 0.00+E00 Inl21m 8.10E+05 8.10E+05 4.06E+03 1.90E401 O.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 Snl21 1.54E+06 1.54E+06 1.52E+06 1.51E+06 1.26E+06 8.35E+05 1.32E+05 4.17E+02 4.16E+02 4.16E+02 4.12E402 4.03E+02 Snl21m 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.92E+02 1.91E+02 1.91E+02 1.89E+02 1.85E+02 Pd122 9.93E+03 6.09E+03 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 O.00E+00 0.00+E00 0.OOE+00 0.00+E00 Ag122 1.30E+05 3.61E+04 0.00E+00 O.OOE+00 0.00E+00 O.00+E00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.00+E00 O.00+E00 Cd122 8.48E+05 7.55E+05 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 O.00E+00 In122 9.47E+05 8.94E+05 0.OOE+00 0.002E00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 In I22m 9.86E+04 9.24E+04 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 Sb122 2.13E+05 2.13E+05 2.12E+05 2.12E+05 1.96E+05 1.65E405 7.65E+04 9.66E+01 1.98E-05 1.83E-15 0.OOE+00 O.00E+00 Sb122m .1.28E+04 1.27E+04 9.16E+01 6.53E-01 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 O.00E+00 0.00+E00 0.00E+00 0.OOE+00 Te123m 9.70E+02 9.70E+02 9.70E+02 9.70E+02 9.70E+02 9.63E+02 9.47E+02 8.17E+02 5.76E+02 3.42E+02 .1.17E+02 1.70E+00 Ag123 5.14E+04 9.09E+03 0.00+E00 0.00E+00 0.OOE+00 0.00E400 0.OOE+00 0.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 0.00E+00 Cd123 5.5 1E+05 5.11E+05 0.00+E00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 O.OOE+00 In123 7.07E+05 6.75E+05 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 O.00+E00 In123m 1.93E+05 1.92E+05 1.03E-06 4.71E-18 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 O.OOE+00 Sn123 1.36E+05 1.36E+05 1.36E+05 1.36E+05 1.36E405 1.35E+05 1.33E+05 1.16E+05 8.40E+04 5.18E+04 1.92E+04 3.80E402 Sn123m 9.25E+05 9.25E+05 5.53E+05 3.29E+05 2.31E+02 1.42E-05 .0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag124 3.90E+04 2.76E+03 0.00+E00 0.00E+00 0.00E+00 0.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 Cd124 8.33E+05 3.91E+05 0.00E+00 0.OOE+00 0.00+E00 0.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 In124 1.57E+06 1.38E+06 0.00E+00 0.00+E00 0.00+E00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 O.OOE+00

• O.00+E00 Sb124 9.72E+04 9.72E+04 9.72E+04 9.72E+04 9.72E+04 9.64E+04 9.31E+04 6.89E+04 3.45E+04 1.22E+04 1.45E+03 3.22E-01 Sb124m 1.93E+03 1.91E+03 2.88E-03 4.28E-09 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 O.00+E00 0.00E+00 0.00E+00 Ag125 1.65E+04 2.08E+03 0.OOE+00 0.OOE+00 O.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 Cdl25 4.98E+05 3.20E+05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 O.00+E00 0.OOE+00 0.00E+00 In125 8.86E+05 7.27E+05 0.00+E00 0.00E+00 O.002E00 0.00E+00 O.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 In I25m 6.84E+05 6.52E+05 0.00+E00 O.00E+00 0.00+E00 O.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.002+00 A-14 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements:-Decay time Following Burnup to 39 GWd(MTU ___.____ ________ _________Nuclide: -T= 0 ." 1: sec._.:. :.30 rmin .,.hrrf ... 8 1.0 d 4.0 d ;30.0d -.::-. 90.0d::: -180.0d-.'

1 yr 3 yr Sn125 5.57E+05 5.57E+05 5.56E+05 5.55E+05 5.44E+05 5.18E+05 4.18E+05 6.45E+04 8.61E+02 1.33E+00 2.19E-06 0.00E+00 Sn125m 2.07E406 2.07E+06 2.35E+05 2.65E+04 1.39E-09 0.00E+00 0.00E+00 0.00E+00 0.OOE400 0.00E+00 0.00E+00 0.00E+00 Sb125 1.37E+06 1.37E406 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.37E+06 1.35E+06 1.29E+06 1.21E+06 1.07E+06 6.43E+05 Te125m 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.03E+05 3.04E+05 3.06E+05 3.04E+05 2.93E+05 2.61E+05 1.57E-05 Ag126 7.72E+03 5.97E+01 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cd126 6.10E+05 1.55E+05 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Tn 126 2.12E+06 1.31 E+06 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 0.00E+00 Sb126 4.64E+04 4.64E+04 4.64E+04 4.64E+04 4.56E+04 4.39E+04 3.71E+04 8.68E+03 3.15E+02 1.36E+01 1.16E+01 1.16E+01 Sb126m 5.45E+04 .5.45E+04 1.83E+04 6.18E+03 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 8.33E+01 Ag127 4.29E+03 8.17E+01 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Cd127 4.74E+05 1.41E+05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00!n127 1.93E+06 1.11 E+06 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 InI27m 1.93E+06 1.62E+06 0.OOE+00 O.OOE+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sn127 3.78E+06 3.78E+06 3.21E+06 2.72E+06 2.70E+05 1.38E+03 6.56E-08 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 Sn127m 5.10E+06 5.10E+06 3.35E+04 2.17E+02 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb127 9.40E+06 9.40E+06 9.40E+06 9.32E+06 8.94E+06 7.95E+06 4.61E+06 4.28E+04 8.71E-01 7.95E-08 2.60E-22 0.OOE+00 Tel27 9.32E+06 9.32E+06 9.32E+06 9.32E+06 9.24E+06 8.63E+06 5.78E+06 1.38E+06 9.09E+05 5.13E+05 1.58E+05 1.52E+03 Te127m 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.59E+06 1.57E+06 1.36E+06 9.32E+05 5.25E+05 1.61E+05 1.55E+03 Ag128 1.94E+03 1.22E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 O.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Cd128 4.12E+05 2.13E+05 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 In128 3.32E+06 1.69E+06 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Sn 128 1.5 11E+07 1.5 1 E+07 1.07E+07 7.50E+06 5.44E+04 7.01E-01 0.001E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Sb128 1.61E+06 1.611E+06 1.57E+06 1,53E+06 9.17E+05 2.67E+05 1.05E+03 1.50E-18 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Sb128m 1.60E+07 1.60E+07 1.26E+07 9.09E+06 6.61 E+04 8.48E-01 0.00E+00 0.OOE+00 0.OOE+00' 0.00E+00 0.OOE+00 0.OOE+00 Cd129 2.02E+05 1.98E+04 0.00E+00 0.00+E00 0.OOE+00 0.00E400 0.00E3+00 0.00E+00 0,00E+00 0.00E+00 0.00E+00 0.00+E00 In 129 3.811E+06 1.22E+06 0.00E+00 0.00E+00 0,O0E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.00E+00 Sn 129 1.35E+07 1.34E+07 8.9413+02 1.35E-01 1.03E-20 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Sn129m 1.29E+07 1.29E+07 5.80E+05 2.60E+04 3.51E-15 0.OOE+00 O.00E+00 0.00E+00 0,00E+00 0.00E+00 0.00E÷00 0.00E+00 A-15 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements--- .--- Decay time Following Burnup to 39 GWdIMTU. .......Nuclide- T=0.. 1 sec .30min. -.. 1 hr 8hr 1.0d 4.0dd -30.0d 90.0d 180.0d lyr : 3yr.Sb129 3.47E+07 3.47E+07 3.25E+07 3.OOE+07 9.93E+06 8.02E+05 9.47E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Te129 3.29E+07 3.29E+07 3.26E+07 3.19E+07 1.53E+07 5.10E406 3.95E+06 2.31E+06 6.71E+05 1.05E+05 2.29E+03 6.54E-04 Te129m 6.66E+06 6.66E+06 6.66E+06 6.66E+06 6.64E+06 6.56E+06 6.17E+06 3.61E+06 1.05E+06 1.63E+05 3.58E+03 1.02E-03 Xe129m 5.16E+03 5.16E+03 5.15E+03 5.14E+03 5.03E+03 4.78E+03 3.77E+03 4.97E+02 4.62E+00 4.15E-03 2.21E-09 0.OOE-00 Cdl30 7.41E+04 1.73E+04 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.002E00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 InI30 2.74E+06 3.38E+05 0.OOE+00 0.OOE+00 0.00+E00 0.00+E00 0.00E+00 0.OOE+00 0.OOE+00 0.00+E00 0.OOE+00 0.00+E00 Snl30 3.61E+07 3.59E+07 1.34E+05 5.02E+02 0.OOE+00 0.00E+00 0.OOE400 0.00E+00 0.00E+00 O.OOE+00

• 0.00+E00 0.OOE+00 Sb130 1.15E+07 1.15E+07 6.82E+06 4.03E+06 2.54E+03 1.22E-04 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Sb130m 4.82E+07 .4.81E+07 3.50E+06 1.35E+05 1.16E-15 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 1130 2.64E+06 2.64E+06 2.58E+06 2.51E+06 1.70E+06 6.91E+05 1.22E404 7.79E-12 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 1130m 1.41E+06 1.41E+06 1.40E+05 1.38E+04 1.24E-10 0.00E+00 O.OOE+00 O.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Cd131 1.15E+04 1.65E+01 0.OOE+00 0.OOE+00 O.OOE+00 0.00+E00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 In 131 .1 7E+06 9.02E+04 0.00+E00 0.00E400 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.002+00 0.00+E00 0.00E+00 Snl31 3.07E+07 3.02E+07 3.94E-07 4.99E-21 0.00E+00 0.004E00 0.00E+00 0.OOE+00 O.OOE+00 0.00+E00 0.00E+00 0.OOE+00 Sb131 8.40E+07 8.40E+07 3.45E+07 1.40E+07 4.45E+01 1.21E-1 I O.00E+00 0.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 Tel31 9.09E+07 9.09E+07 6.98E+07 4.42E+07 4.03E+06 2.79E+06 5.28E+05 2.90E-01 1.02E-15 0.OOE+00 0.OOE+00 0.00E+00 Tel31m 2.15E+07 2.15E+07 2.12E+07 2.11 E+07 1.80E+07 1.24E+07 2.35E+06 1.28E+00 4.56E-15 0.00E+00 0.OOE+00 0.OOE+00 1131 1.07E+08 1.07E+08 1.07E+08 1.07E+08 1.05E+08 1.00E+08 7.87E+07 8.40E+06 4.76E+04 2.03E+01 2.35E-06 0.00+E00 XeI31m 1.47E+06 1.47E+06 1.47E+06 1.47E+06 1.47E+06 1.45E+06 1.38E+06 5.05E+05 2.00E+04 1.11 E+02 2.29E-03 7.54E-22 In132 3.07E+05 7.38E+03 O.00E+00 0.OOE+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 Sn132 2.45E+07 2.41E+07 7.0 1E-07 1.97E-20 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Sb132 5.08E+07 5.07E+07 3.60E+05 2.54E+03 0.OOE+00 0.00+E00 0.OOE+00 0.000+00 0.OOE+00 0.000+00 .OOE+00 0.00E+00 Sb132m 4.8 1E+07 4.80E+07 3.32E+04 1.97E+01 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Te132 1.54E+08 1.54E+08 1.54E+08 1.53E+08 1.44E+08 1.25E+08 6.59E+07 2.61E+05 7.46E-01 3.61E-09 2.76E-26 0.OOE+00 1132 1.57E+08 1.57E+08 1.57E+08 1.56E+08 1.48E+08 1.28E+08 6.78E+07 2.69E+05 7.72E-01 3.71E-09 2.84E-26 0.00E+00 Cs132 4.35E+03 4.35E+03 4.35E+03 4.33E+03 4.20E+03 3.91E+03 2.83E+03 1.76E+02 2.87E-01 1.89E-05 4.67E-14 0.OOE+00 Tn133 9.63E+03 1.91E+01 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00+E00 Sn133 6.62E+06 4.09E+06 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00+E00 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 A-16 Safety Analysis Report for Susquehanna Units 1 and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU "___.:_.Nuclide.-

.-Tf0l 1 sec,_.-ý :30 min-v Il~hr:. 8 hr ----.-.1.0d,+ý- 4.0d-; :30.0 dý: -90.0d+e 180.0&=- 3 yr 3yr Sb133 6.96E+07 6.94E+07 1.70E+04 4.15E+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Tel33 1.20E+08 1.20E+08 3.66E+07 1.46E+07 5.53E+04 3.36E-01 O.OOE+00 O.OOE+00. O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Te133m 9.86E+07 9.86E+07 6.82E407 4.69E+07 2.44E+05 1.49E+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 1133 2.22E+08 2.22E+08 2.20E+08 2.18E+08 1.74E+08 1.02E+08 9.24E+06 8.63E-03 1.25E-23 O.OOE+00 O.OOE+00 O.OOE+00 1133m 1.70E407 1.64E+07 6.98E+06 4.80E+06 2.51E+04 1.52E-01 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Xe133 2.12E+08 2.12E+08 2.12E+08 2.12E+08 2.12E+08 2.06E+08 1.51E+08 4.99E+06 1.80E+03 1.22E-02 2.82E-13 0.OOE+00 Xe133m 6.99E+06 6.99E+06 6.99E+06 6.98E+06 6.86E+06 6.29E+06 3.OOE+06 8.48E+02 4.78E-06 2.03E-18 0.OOE+00 O.OOE+00 Snl34 1.10E+06 5.66E+05 0.OOE+00 0.O0E+00 0.OOE+00 O.OOE+00 0.OOE+O0 O.OOE+00 0.OOE+00 0.OOE+O0 0.OOE+00 O.OOE+00 Sb134 1.26E+07 5.92E+06 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 O.OOE+00 Sb134m 9.32E+06 8.71E+06 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Tel34 1.95E+08 1.95E+08 1.18E+08 7.2 1E+07 6.8 1E+04 8.33E-03 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 1134 2.45E+08 2.45E+08 2.15E+08 1.75E+08 1.53E+06 5.71E+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 1134m 2.17E+07 2.16E+07 7.72E+04 2.76E+02 0.OOE+00 0.OOE+00 0.OOE+O0 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 Xe134m 5.90E+06 9.93E+05 1.79E+03 6.36E+00 0.OOE+00 O.OOE+00 0.OOE+O0 O.OOE+00 0.OOE+00 0.OOE+O0 0.OOE+00 O.OOE+00 Cs134 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.30E+07 2.29E+07 2.24E+07 2.12E+07 1.95E+07 1.64E+07 8.40E+06 Cs134m 4.81E+06 4.81E+06 4.27E+06 3.79E+06 7.16E+05 1.59E+04 5.69E-04 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Sn135 9.40E+04 1.78E+04 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+0O 0.OOE+00 O.OOE+00 O.OOE+00 Sb135 5.84E+06 3.90E+06 0.OOE+00 O.OOE+0O 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 Te135 1.06E+08 1.02E+08 0.OOE+00 O.OOE+0O 0.OOE+00 O.OOE400 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 1135 2.11 E+08 2.11 E+08 2.OOE+08 1.90E+08 9.09E+07 1.68E+07 8.40E+03 0.OOE+00 0.OOE+00 0.OOE+00 O.OOEO00 O.OOE+00 Xe135 7.03E407 7.03E+07 7.56E+07 8.02E+07 1.01E+08 5.62E+07 4.OOE+05 1.18E-15 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 Xe135m 4.63E+07 4.63E+07 3.58E+07 3.18E+07 1.48E+07 2.74E406 1.38E+03 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Cs135m 4.65E+06 4.65E+06 3.14E+06 2.12E+06 8.71E+03 3.08E-02 0.OOE+00 O.OOE+00 O.00E400 O.OOE+00 O.OOE+00 0.OOE+00 Ba135m 4.19E+04 4.19E+04 4.13E+04 4.09E+04 3.45E+04 2.35E+04 4.12E+03 1.17E-03 9.17E-19 O.OOE+00 0.OOE+00 0.OOE+00 Snl36 8.10E+03 3.07E+03 O.OOE+00 0.O0E+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 Sb136 9.02E+05 3.89E+05 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+0O 0.OOE+00 Te136 4.70E+07 4.52E+07 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE400 0.OOE+00 O.OOE+O0 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 1136 9.63E+07 9.55E+07 3.46E+01 1.10E-05 0.00E+00 0.00E+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 A- 17 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 39 GWd/MTU __. _____ _ ..__"" Nuclide:: =T_=0.: I sec':_:::

30min-:::.

1.hr. 8 hr :, .0d A 4.0d ... .30.0d -90.0 d 1-80.0d ldýr :3yr 1136m 4.71E+07 4.65E+07 1.32E-04 3.67E-16 O.OOE+00 0.00+E00 O.OOE+00 0.00+E00 0.OOE+0O 0.OOE+00 O.00E+00 0.00+E00 Csl36 7.34E+06 7.34E+06 7.33E+06 7.33E+06 7.21E+06 6.96E+06 5.94E+06 1.51E+06 6.41E+04 5.60E+02 3.24E-02 6.31E-19 Ba136m 8.40E+05 8.25E+05 8.17E+05 8.17E+05 8.10E+05 7.79E+05 6.66E+05 1.70E+05 7.18E+03 6.27E+01 3.63E-03 7.07E-20 Sb137 8.40E+05 1.97E+05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 O.00+E00 Te137 1.57E+07 1.29E+07 O.OOE+00 0.00E+00 O.00+E00 0.00+E00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 O.00E+00 1137 1.03E+08 1.01E+08 0.OOE+00 O.OOE+00 0.OOE+0O0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Xe137 2.02E+08 2.02E+08 9.24E+05 3.97E+03 0.00E+00 0.OOE+0O 0.00E+00 0.OOE+00 O.OOE+00 O.OOE+O 0 O.OOE+00 O.00E+00 Cs137 1.73E407 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.73E+07 1.72E+07 1.70E+07 1.62E+07 Ba137m 1.65E+07 1.65E+07 1.64E+07 1.64E+07 1.64E+07 1.64E+07 1.64E+07 1.63E+07 1.63E+07 1.62E+07 1.60E+07 1.53E+07 Sb138 1.21E+04 2.22E+02 O.OOE+00 O.00+E00 0.00E+00 0.OOE+00 O.OOE+00 O.00+E00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 Te138 3.81E+06 2.32E+06 O.OOE+00 O.OOE+O0 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 0.00E+00 1138 5.23E+07 4.73E+07 0.OOE+00 O.OOE+00 0.00+E00 0.00E+00 O.OOE+00 0.00+E00 0.O0E+0O 0.00E+00 0.00E+00 0.00E+00 Xe138 1.89E+08 1.89E+08 4.32E+07 9.86E+06 1.03E-02 3.09E-23 0.00+E00 0.OOE+00 O.00+E00 0.00E+00 0.OOE+00 O.OOE+00 Cs138 2.05E+08 2.05E+08 1.51E+08 8.94E+07 1.15E+04 1.21E-05 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Cs138m 8.79E+06 8.71E+06 6.91E+03 5.43E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 O.OOE+00 0.OOE+00 Tel39 5.87E+05 1.77E+05 0.00E+00 O.00+E00 0.OOE+00 0.000+00 0.000+00 0.006+00 O.00+E00 O.OOE+00 0.OOE+00 0.OOE+00 1139 2.49E+07 1.85E+07 0.00E+00 O.00E+00 0.00+E00 0.006E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 Xe139 14AOE+08 1.38E+08 3.13E-06 6.85E-20 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 Csl39 1.90E+08 1.90E+08 2.13E+07 2.26E+06 5.20E-08 0.00+E00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 Ba139 1.95E+08 1.95E+08 1.70E+08 1.34E+08 4.32E+06 1.66E403 7.15E-13 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tel40 8.40E+04 3.87E+04 0.00+E00 O.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.004E00 0.00E+00 O.00E+00 0.00E+00 1140 6.58E+06 2.96E+06 0.00+E00 O.00E+00 0.00E+00 0.00+E00 0.00E+00 O.00+E00 0.00E+00 0.00+E00 O.00+E00 0.00E+00 Xel40 9.70E+07 9.24E+07 0.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Cs140 1.70E+08 1.70E+08 6.16E-01 1.91E-09 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 O.00E+00 0.00+E00 BaI40 1.96E+08 1.96E+08 1.96E+08 1.96E+08 1.93E+08 1.86E+08 1.57E+08 3.84E+07 1.47E+06 1.11E+04 4.69E-01 2.69E-18 LaI40 2.09E+08 2.09E+08 2.09E+08 2.09E+08 2.08E+08 2.03E+08 1.78E+08 4.42E+07 1.70E+06 1.28E+04 5.40E-01 3.106-18 Prl40 3.19E+03 3.19E+03 6.93E+00 1.50E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 O.00E+00 0.00E+00 1141 8.48E+05 1.87E+05 0.00+E00 0.00E+00 0.00E+00 0.006E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 O.00E+00 0.00+E00 A-18 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements i- Decay time. Following Burnup to 39 GWd/MTU-:- .______ ,_.___ _-Nuclide T= 0 Isec 30mn hr 8 hr .1.0 d 4.0d 30.0 d 90.0d 180.Od 1Yr., 3yr XeI41 3.71 E+07 2.49E+07 0.0013+00 0.0013+00 00 0.00E+00 0.0013+00 0.OOE+00 0.0013+00 0.00E400 0.0013+00 0.OOE+00 Cs141 1.30E+08 1.2813+08 2.511E-14 0.00E+00 0.00E+00 0.0013+00 0.OOE+00 0.00E+00. 0.0013+00 O.003E+00 0.00E+00 0.00E+00 Bal41 1.76E+08 1.76E+08 5.76E+07 1.84E+07 2.21E+00 3.34E-16 0.0013+00 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 La141 1.78E+08 1.78E+08 1.72E+08 1.60E+08 4.70E+07 2.77E+06 8.17E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 Cel41 1.80E+08 1.80E+08 1.80E+08 1.80E+08 1.79E+08 1.76E+08 1.66E+08 9.55E+07 2.64E+07 3.88E+06 7.46E+04 1.28E-02 1142 2.58E+05 8.1713+03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 0.0013+00 Xe142 1.46E+07 8.25E+06 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.0011+00 0.OOE+00 Cs142 7.54E+07 5.38E+07 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 Ba142 1.68E+08 1.6813+08 2.3713+07 3.32E+06 3.93E-06 0.0013+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 La142 1.74E+08 1.74E+08 1.54E+08 1.24E+08 5.10E+06 3.42E+03 1.82E-11 0.0013+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Pr142 7.33E+06 7.33E+06 7.2013+06 7.07E+06 5.49E+06 3.07E+06 2.26E+05 3.38E-05 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 1143 2.611E+03 4.611E+02 0.0013+00 0.00E+00 0.0013+00 0.0013+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 Xe143 2.1913+06 1.06E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs143 3.80E+07 2.63E+07 0.0013+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.0013+00 0.0013+00 0.0013+00 Ba143 1.4413+08 1.38E+08 0.0013+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.0013+00 La 143 1.65E+08 1.65E+08 3.86E+07 8.86E+06 1.0213-02 3.7 1E-23 0.00E+00 0.00E+00 0.0013+00 0.0013+00 0.00E+00 0.00E+00 Ce143 1.67E+08 1.67E+08. 1.6613+08 1.64E+08 1.4213+08 1.02E+08 2.2413+07 4.55E+01 3.3213-12 0.0013+00 0.OOE+00 0.00E+00 Pr143 1.61E+08 1.61E+08 1.61E+08 1.61E+08 1.61E+08 1.60E+08 1.4411+08 3.90E+07 1.82E+06 1.8313+04 1.411E+00 8.71E-17 Xe144 4.58E+05 2.44E+05 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Csl44 1.12E+07 5.79E+06 0.00E+00 0.001+00 0.00E+00 0.001+00 O.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 Ba144 1.1212+08 1.05E+08 0.00E+00 0.OOE+00 0.00E+00 0.0013+00 0.00E+00 0.OOE+00 0.00E+00 0.0013+00 0.00E+00 0.0013+00 La144 1.46E+08 1.4513+08 1.08E-05 6.01E-19 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cel44 1.51E+08 1.511E+08 1.51E+08 1.51E+08 1.511E+08 1.5113+08 1.50E+08 l.41E+08 1.21E+08 9.78E+07 6.22E+07 1.05E+07 Pr144 1.5213408 1.52E+08 1.51 E+08 1.51E+08 1.511E+08 1.51E+08 1.5013+08 1.411E+08 1.211E+08 9.7813+07 6.22E+07 1.0513+07 Pr144m 2.1213+06 2.12E+06 2.1213+06 2.12E+06 2.12E+06 2.111E+06 2.0913+06 1.9713+06 1.70E+06 1.3713406 8.7113+05 1.47E+05 Xe145 4.5713+04 2.111E+04 0.00E+00 O.OOE+00 0.00E+00 0.0013+00 0.00E+00 0.0013+00 0.0013+00 0.0013+00 0.0013+00 0.00E+00 Cs145 2.701+06 8.6313+05 0.00E+00 0.0013+00 0.00E+00 0.0013+00 0.00E+00 0.0013+00 0.0013+00 0.0013+00 0.00E+00 0.00E+00 Ba145 4.9713+07 4.2513+07 0.00E+00 0.00E+00 0.00E+00 0.001E+00 0.0013+00 0.0013+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 A-19 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements.- ... Decay time Following Burnup to 39 GWd/MTU Nuclide, T=0-- -see .- -30min .hr 8hr--- .OU d 4.Od ..... 30.0d d. 90.0d -180.0d lyr 3yr La145 1.02E+08 1.002+08 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.00E400 0.00+E00 0.00E+00 Ce145 1.13E+08 1.13E+08 1.3 1E+05 1.3 1E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Pr145 1.13E+08 1.13E+08 1.08E+08 1.02E408 4.53E+07 7.11E+06 1.70E+03 0.001+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 Xe146 4.03E+03 1.182+03 0.00E+00 0.OOE+00 0.00E300 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 Cs146 5.78E+05 7.79E+04 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 Ba146 2.54E+07 1.86E+07 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.00E400 0.00E+00 0.00E+00 La146 6.59E+07 6.132+07 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.002400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ce146 9.022+07 9.02E+07 1.952+07 4.19E+06 1.86E-03 7.87E-25 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 Pr146 9.09E+07 .9.09E+07 6.252+07 3.16E+07 2.15E+02 2.31E-10 0.00E+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.OOE+00 Cs147 1.28E+04 3.582+03 0.00E+00 0.00E+00 0.002+00 0.OOE+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.002+00 0.002+00 Ba147 4.26E+06 1.582+06 0.002+00 0.00E+00 0.002+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 La147 2.872+07 2.48E+07 0.00+E00 0.00E+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 Ce147 6.812+07 6.76E+07 1.75E-02 4.29E-12 0.00+E00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.00E+00 Pr147 7.17E+07 7.171+07 1.67E+07 3.61E+06 1.83E-03 1.032-24 0.00E+00 0.002+00 0.00E+00 0.001+00 0.002+00 0.00+E00 Nd147 7.242+07 7.242+07 7.232+07 7.22E407 7.092+07 6.801+07 5.62E+07 1.091+07 2.471+05 8.402402 7.022-03 6.59E-23 Pm147 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.51E+07 2.513+07 2.52E+07 2.522+07 2.432+07 2.282+07 1.991+07 1.181+07 Cs148 2.33E+03 7.952+01 0.00E+00 0.OOE+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.002+00 Ba148 8.332+05 2.662+05 0.00E+00. 0.OOE+00 0.002+00 0.002+00 0.00E+00 0.00E+00 0.002+00 0.002+O0 0.002+00 0.002+00 La148 9.09E+06 4.902+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.001+00 Ce148 4.822+07 4.772+07 1.022-02 2.15E-12 0.002+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.001+00 0.00+E00 0.002+00 Pr148 5.57E+07 5.562+07 9.40E+03 9.86E-01 0.002+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.002+00 0.002+00 Pm148 2.022+07 2.02E+07 2.022+07 2.01E+07 1.941+07 1.78E+07 1.21E+07 5.382+05 4.552+04 1.002+04 4.462+02 2.11E-03 Pm148m 3.882+06 3.882+06 3.882+06 3.87E+06 3.862406 3.81E406 3.63E+06 2.35E+06 8.56E+05 1.892+05 8.402+03 4.002-02 Ba149 9.172+04 3.382+04 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.002+00 0.00+E00 0.001+00 0.002+00 0.001+00 0.00E+00 La149 2.59E+06 1.962+06 0.002+00 0.00E+00 0.00+E00 0.002+00 0.002+00 0.00E+00 0.002+00 0.00E+00 0.002+00 0.002+00 Ce149 2.51 E+07 2.222+07 0.002+00 0.00E+00 0.00+E00 0.001+00 0.002+00 0.00E+00 0.00E+00 0.00+E00 0.002+00 0.00+E00 Pr149 3.852+07 3.84E+07 4.00E+03 4.03E-0l 0.002+00 0.00E+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 Nd149 4.162+07 4.16E+07 3.472+07 2.84E+07 1.70E+06. 2.752+03 7.492-10 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 A-20 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version =Table A-1 Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements------ --Decay time.Following Burnup to 39 GWd/MTU<: _ _.____- -__.__- .Nuclide: T=0 1 sec 30min I hr 8 hr,:- :1.0d 4.0d 30.0d 90.0d 180.0d 1 yr 3yr.Pm 149 6.46E+07 6.46E+07 6.44E+07 6.42E+07 5.94E+07 4.82E+07 1.89E+07 5.45E+03 3.72E-05 2.09E-17 0.OOE+00 0.00E+00 BaI50 8.25E+03 4.00E+03 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00. 0.00E+00 0.00E+00 0.OOE+00 0.001+00 Lal50 4.38E+05 1.44E+05 0.001+00 O.00E+00 0.OOE+00 0.00E+00 0.002+00 .003E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 CeIS0 1.14E+07 9.63E+06 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.001+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 PrIS0 2.43E+07 2.28E+07 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 O.00E+00 0.OOE+00 O.00E+00 0.OOE+00 O.OOE+00 Pm150 5.53E+05 5.53E+05 4.87E+05 4.27E+05 6.99E+04 1.12E+03 9.09E-06 0.00E+00 0.OOE+00 O.00E+00 0.OOE+00 0.00+E00 Lal51 7.27E+04 2.77E+04 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 O.00+E00 O.00+E00 0.00+E00 Ce151 3.28E+06 1.68E+06 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00E400 0.00E+00 O.OOE+00 O.00E+00 0.00E+00 0.OOE+00 Prl51 1.28E+07 1.25E+07 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 Nd5l1 2.15E+07 2.15E+07 4.10E+06 7.72E+05 5.29E-05 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Pml5! 2.17E+07 2.17E+07 2.15E+07 2.13E+07 1.80E+07 1.21E+07 2.10E+06 5.07E-01 2.73E-16 O.00+E00 0.00E+00 0.00E+00 SmlS1 6.80E+04 6.80E+04 6.80E+04 6.80E+04 6.81E+04 6.83E+04 6.87E+04 6.87E+04 6.86E+04 6.85E+04 6.82E+04 6.72E+04 CeI52 4.26E+05 3.89E+05 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pr152 4.35E+06 3.96E+06 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 0.OOE+00 Nd152 1.44E+07 1.442+07 2.332+06 3.76E+05 3.06E-06 0.00E+00 0.0013+00 0.002+00 0.000+00 O.OOE+00 0.002+00 0.00E+00 Pm152 1.50E+07 1.50E+07 3.60E+06 5.88E+05 4.77E-06 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 Pm152m 5.37E+05 5.36E+05 3.38E+04 2.13E-03 3.28E-14 0.00E+00 0.OOE+00 O.00+E00 O.00E+00 0.OOE+00 0.OOE+00 0.002E00 Eul52m 2.48E+04 2.48E+04 2.38E+04 2.30E+04 1.37E+04 4.15E+03 1.96E+01 1.371-19 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Ce153 1.47E+05 9.17E+04 0.OOE+00 0.OOE+00 0.00E+00 0.003+00 0.OOE+00 0.001+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 Pr153 1.82E+06 1 .57E+06 0.002+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 Nd153 8.25E+06 8.171+06 7.95E-02

• 7.39E-10 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Pm153 9.932+06 9.93E+06 2.59E+05 5.50E+03 0.00+E00 0.00E+00 O.OOE+00 0.002+00 0.00E+O0 0.002+00 0.00E+00 .0.00+E00 Sm153 5.31E+07 5.31 E+07 5.27E+07 5.23E+07 4.71E+07 3.71 E+07 1.26E+07 1.10E+03 4.73E-07 4.21E-21 0.00E+00 O.00+E00 Gd153 8.132+05 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.13E+05 8.05E+05 7.49E+05 6.30E+05 4.87E+05 2.86E+05 3.52E+04 Ce154 1.5 1E+04 1.06E404 0.00+E00 O.OOE400 0.002+00 0.00+E00 0.00E+00 0.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 PrI54 3.58E+05 1.92E405 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 O.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 Ndl54 4.19E+06 4.12E+06 1.20E-07 3.37E-21 O.00+E00 O.00+E00 0.00E+00 0.OOE+00 0.002+00 0.002+00 0.00E+00 0.00E+00 Pm15 4 5.25E+06 5.24E+06 4.452+01 2.49E-04 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 A-21 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements.......Decay time, olowing Burnupto39 GWdIMTU : 'Nuclide::

T=0 1 sec!-:;:-

30min 1 hr 8. hr, -1.0d 4.0 d *3.d..d 180.0 d;, Iyrý :3Yr'Pmi54m 1.06E+06 1.05E406 4.54E-+02 1.93E-01 0.0013+00 0.OOE+00 0.OOE+00 0.00E+00 0.0013+00 0.00E+00 0.O0E+00 0.OOE+00 Eu154 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06 1.06E+06.

1.0413+06 1.02E+06 9.77E+05 8.29E+05 Gd 155m 1.60E+04 2.96E-06 0.OOE+00 0.00E+00 0.00E+00 0.000+00 0..0E+00 0.OOE+00 0.003E+00 0.0013+00 0.0013+00 O.OOE+00 Pr155 7.72E404 4.19E+04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 00 0.00+0 0.0E+00 0.0013+00 0.00E+00 0.0013+00 Nd155 1.4713+06 1.41E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 Pm155 3.3213+06 3.29E+06 2.18E-05 1.12E-16 0.OOE+00 0.0013+00 0.00E+00 0.001E+00 .OOE+00 0.OOE+00 0.001E+00 0.0013+00 Sm155 4.1OE+06 4.10E+06 1.67E+06 6.58E+05 1.AE+00 1.5413-13 "O.00E+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 0.0013+00 Eu155 4.34E+05 4.34E+05 4.3413405 4.34E405 4.34E+05 4.34E+05 4.34E+05 4.29E+05 4.18E+05 4.04E+05 3.75E+05 2.78E+05 PrI56 1.12E+04 1.84E+03 0.OOE+00 0.OOE+00 0.OOE+00 0.0013+00 0.0013+00 0.00E+00 0.00E0O0 0.0013+00 0.00E+00 O.E+00 Nd156 5.38E+05 5.20E+05 0.001E+00 0.001+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Pm156 1.70E+06 1.64E+06 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.0013+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 SmI56 2.57E+06 2.5713+06 2.48E+06 2.38E+06 1.4213+06 4.38E+05 2.16E+03 2.25E-17 0.00E+00 0.00E+00 0.0013+00 0.00E+00 Eu156 2.74E+07 2.7413+07 2.74E+07 2.73E+07 2.70E+07 2.6213407 2.29E-+07 6.98E+06 4.511E+05 7.41E+03 1.58E+00 5.20E-15 Nd157 1.3613+05 1.02E+05 0.OOE+00 0.OOE+00 0.OO1+00 0.001E+00 0.00E+00 0.001+00 0.OOE+00 ..OOE+00 0.001E+00 0.00E+00 Pm157 7.87E+05 7.79E+05 1.10E-03 1.52E-12 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 .003E+00 0.OOE+00 0.00E+00 0.0013+00 Sm157 1.6213+06 1.62E+06 1.32E+05 1.00E+04 2.15E-12 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.OOE+0 0.OOE+00 0.0013+00 Eu157 2.63E+06 2.63E+06 2.5913+06 2.53E+06 1.83E+06 8.86E+05 3.31E+04 1.4013-08 0.001E+00 0.0013+00 0.0013+00 O.00E+00 Nd158 2.2313+04 1.73E+04 0.0013+00 0.OOE+00 0.OOE+00 0.00E+00 0.0013+00 0.0013+00 0.00E+00 0.00E+00 0.0013+00 0.0013+00 Pml58 2.1513+05 1.83E+05 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.001+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Sm158 8.48E+05 8.48E+05 1.9511+04 4.48E+02 0.0013400 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 0.0013+00 0.0013+00 Eu158 9.47E+05 9.47E+05 6.7413+05 4.30E1+05 7.57E+02 3.83E-04 0.00E+00 0.0013+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Nd159 1.9613+03 6.62E+02 0.00E+00 0.00E+00 0.00E+00 0.00E400 O.00E+00 0.0013+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm159 5.00E+04 3.99E+04 0.OOE+00 0.00E3+00 0.00E3+00 0.0013+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sm 159 3.52E+05 3.511E+05 1.60E+02 7.20E-02 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.0013+00 Eu159 4.811E+05 4.811E+05 1.72E+05 5.45E+04 5.64E-03 6.10E-19 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 Gd159 2.7413+07 2.74E+07 2.69E+07 2.65E+07 2.04E+07 1.12E+07 7.611E+05 5.7713-05 0.00E+00 0.00E+00 0.0013+00 0.0013+00 PmI60 5.54E+03 2.18E+03 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.0013+00 SmI60 1.05E+05 1.04E+05 3.611E-03 1.23E-10 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 A-22 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-i Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_____ "_ "Decay time Following Burnup to 39 GWd/MTU: Nuclide:*_T= 0 ., _sec 30 min.- .1-hr-m 8 hr,.-. 1.0 d 4.0 d -30.0 d '90.0 d .180.Od lyr : 3yr.Eu160 1.93E+05 1.92E+05 9.17E-03 3.12E-10 O.OOE+00 O.OOE-00 O.00E+00 0.OOE+00 O.OOE+O0 O.OOE+00 O.OOE+00 O.OOE+00 Th160 7.20E+06 7.20E+06 7.19E+06 7.19E+06 7.17E+06 7.13E+06 6.92E+06 5.40E+06 3.03E+06 1.28E+06 2.17E+05 1.97E+02 Sm161 2.28E+04 1.99E+04 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+O0 O.OOE+00 0.OOE+O0 O.OOE+00 0.OOE+00 O.OOE+00 Eu161 7.44E+04 7.35E+04 1.01E-08 1.30E-21 O.00E400 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 Gd161 3.27E+06 3.25E+06 1.12E+04 3.81E+01 0.OOE+00 O.OOE+-0 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Th161 4.33E+06 4.33E+06 4.32E+06 4.31E+06 4.19E+06 3.92E+06 2.90E+06 2.13E+05 5.14E+02 6.09E-02 5.04E-10 O.OOE+00 Sm162 3.19E+03 2.80E+03 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+O0 O.OOE+00 O.OOE+00 O.OOE+00 Eu 162 1.84E+04 1.83E+04 8.56E+00 3.92E-03 O.OOE+00 0.OOE+00 O.OOE+00 O.00E400 O.OOE+00 O.OOE+00 O.OOE+00 0.0EO00 Gd162 4.22E+04 4.21E+04 4.28E+03 3.60E+02 3.20E-13 0.OOE+00 0.00E400 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+0O0 O.OOE+00 Th162 4.29E+04 4.29E+04 1.34E+04 1.79E+03 4.06E-12 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+0O 0.OOE+00 0.00E+00 Eu163 3.67E+03 3.36E+03 O.OOE+00 O.OOE+O0 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Gd163 1.47E+04 1.46E+04 2.18E-02 3.12E-08 O.OOE+00 O.OOE-00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 Th163 1.68E+04 1.68E+04 6.24E+03 2.15E+03 7.05E-04 1.07E-18 O.OOE-O00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 Gd164 4.36E+03 4.36E+03 1.67E+03 6.41E+02 9.47E-04 4.45E-17 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Th164 6.03E+03 6.02E-403 1.94E+03 7.44E+02 1.10E-03 5.16E-17 0.OOE+00 O.O0E+00 0.OOE+00 O.OOE+0 0.OOE+00 O.OOE+00 Th165 2.05E+03 2.04E+03 1.33E-01 6.94E-06 0.OOE+00 0.OOE-I00 O.OOE+00 O.OOE+0O 0.OOE+0O 0.OOE+00 0.OOE+00 O.OOE+00 Dy165 5.38E+05 5.38E+05 4.66E+05 4.02E+05 5.03E+04 4.34E+02 2.25E-07 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Dy165m 3.46E+05 3.43E+05 3.05E-01 1.48E-05 0.OOE+00 0.OOE+00 0.OOE+0O 0.OOE÷00 0.OOE+00 0.OOE+00 O.0OE-I00 O.OOE+00 Dy166 2.09E+03 2.09E+03 2.09E+03 2.08E+03 1.96E+03 1.71E+03 9.24E+02 4.63E+00 2.26E-05 2.44E-13 1.74E-29 O.OOE+00 Ito166 8.58E+04 8.58E+04 8.41E+04 8.33E+04 6.99E+04 4.69E+04 8.43E+03 8.10E+00 3.95E-05 4.26E-13 1.74E-29 0.OOE+00 Er167m 1.14E+03 8.40E+02 O.OOE+00 O.OOE+00 O.OOE+00 0.00E400 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+O0 O.OOE+00 O.OOE+00 11f175 3.62E+03 3.62E+03 3.62E+03 3.62E+03 3.61E+03 3.58E+03 3.48E+03 2.69E+03 1.49E+03 6.10E+02 9.78E+01 7.05E-02 Lu176m 1.64E403 1.64E+03 1.50E+03 1.36E+03 3.58E+02 1.70E+01 1.86E-05 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Lu177 6.84E+02 6.84E+02 6.82E+02 6.81E+02 6.60E+02 6.17E+02 4.52E+02 3.12E+01 3.46E-01 1.93E-01 8.63E-02 3.72E-03 IMf178m 3.46E+02 2.90E+02 0.OOE+00 O.OOE+00 0.OOE+00 0.00E400 O.OOE+00 O.OE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 If179m 3.98E+05 3.84E+05 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE-00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Ilfl80m 1.12E+04 1.12E+04 1.05E+04 9.86E+03 4.10E+03 5.45E+02 6.24E-02 O.OOE+00 0.OOE+O0 O.OOE+00 0.OOE+00 0.OOE+00 Hf"81 2.38E+05 2.38E405 2.38E+05 2.38E+05 2.37E+05 2.35E+05 2.23E+05 1.46E+05 5.47E+04 1.25E404 6.07E+02 3.93E-03 A-23 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements:-!._'- ~ ~ ~ -'-_'. Dec --- 1-1-ay~timeFloigBrnup to 39GWdIMTUJ. .> .. ................. "" ,'... .- -.I.I. -....

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-.1 : 1. 'r ý .ý.Nuclide. T=0 I sec 30mmin 1- :hr 8hr 1.0d 4.0d -30.0 d :90.0dd , 180.0d 'lyr: 3yr'W181 1.42E+03 1.42E+03 1.42E+03 1.42E+03 1.42E+03 1.41E+03 1.39E+03 1.20E+03 8.48E+02 5.09E+02 1.76E+02 2.70E+00 Ta182 2.59E+04 2.59E+04 2.59E+04 2.59E+04 2.59E+04 2.57E+04 2.53E+04 2.16E+04. 1.51E+04 8.79E+03 2.87E+03 3.51E+01 Ta182m 4.18E+01 4.18E+01 1.12E+01 3.03E+00 3.13E-08 1.75E-26 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 TaI83 5.46E+04 5.46E+04 5.45C+04 5.43E+04 5.23E+04 4.77E+04 3.17E+04 9.32E+02 2.69E-01 1.32E-06 1.57E-17 O.OOE+00 W183m 6.40E+04 6.29E+04 5.45E+04 5.43E+04 5.23E+04 4.77E+04 3.17E+04 9.32E+02 2.69E-01 1.32E-06 1.57E-17 0.00E+00 W185 3.52E+04 3.52E+04 3.52E+04 3.52E+04 3.51E+04 3.49E404 3.39E+04 2.67E+04 1.54E+04 6.69E+03 1.21E+03 1.43E+00 Wl85m 7.55E+01 7.49E+01 2.96E-04 1.15E-09 0.00E+00 0.002+00E 0.OOE+00 0.OOE+00 0.00+E00 O.00+E00 0.002+00 0.002+00 Re186 2.58E+04 2.58E+04 2.57E+04 2.56E+04 2.43E+04 2.15E+04 1.24E+04 1.05E+02 1.73E-03 1.16E-10 1.99E-25 0.OOE+00 W187 4.42E+05 4.42E+05 4.36E+05 4.30E+05 3.51E+05 2.21E+05 2.74E+04 3.77E-04 2.75E-22 O.00+E00 0.00E+00 0.OOE+00 W188 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.96E+03 1.94E+03 1.89E+03 1.45E+03 8.02E+02 3.25E+02 5.12E+01 3.49E-02 ReI88 1.73E+05 1.73E+05 1.72E+05 1.69E+05 1.28E+05 6.75E+04 5.37E+03 1.47E+03 8.10E+02 3.29E+02 5.171+01 3.52E-02 ReI88m 1.68E+05 1.67E+05 5.49E+04 1.80E+04 2.86E-03 8.33E-19 0.002+00 0.00E+00 0.00+E00 0.00E+00 *0.00+00 0.002+00 Os191 7.72E+01 7.72E+01 7.64E+01 7.64E+01 7.63E+01 7.49E+01 6.59E+01 2.05E+01 1.38E+00 2.40E-02 5.75E-06 3.06E-20 Os191m 5.82E+01 5.82E+01 5.67E+01 5.52E+01 3.812+01 1.63E+01 3.62E-01 1.66E-15 0.00E+00 0.00E+00 O.00E+00 0.00E+00 Ir192 3.01E+01 3.01E+01 3.01E+01 3.01E+01 3.00+E01 2.98E401 2.90E+01 2.27E+01 1.29E+01 5.55E+00 9.78E-01 1.06E-03 Np236m 4.84E+02 4.84E+02 4.77E+02 4.70E+02 3.78E+02 2.31E+02 2.51E+O 1 1. 13 E-07 6.122E-27 0.00E+00 0.00E+00 0.00E+00 U237 1.002+08 1.00E+08* 1.00E+08 9.93E+07 9.70E+07 9.02E+07 6.64E+07 4.60E+06. 1.022E+04 4.48E+02 4.37E+02 3.97E+02 Pu237 6.7 1E+02 6.7 1E+02 6.70E+02 6.70E+02 6.67E+02 6.60E+02 6.30E+02 4.23E+02 1.69E+02 4.23E+01 2.47E+00 3.35E-05 Np238 4.70E+07 4.70E+07 4.67E+07 4.64E+07 4.22E+07 3.39E+07 1.27E+07 2.55E+03 7.44E+00 7.43E+00 7.41E+00 7.34E+00 Pu238 4.56E+05 4.56E+05 4.56E+05 4.56E+05 4.56E+05 4.57E+05 4.58E+05 4.63E+05 4.69E+05 4.75E+05 4.82E+05 4.81E+05 U239 2.12E+09 2.12E+09 8.79E+08 3.61E+08 1.47E+03 7.16E-10 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Np239 2.12E+09 2.12E+09 2.12E409 2.11E+09 1.93E+09 1.59E+09 6.59E+08 3.16E+05 3.13E+03 3.13E+03 3.13E+03 3.13E+03 Pu239 4.83E+04 4.83E+04 4.83E+04 4.83E+04 4.83E+04 4.84E+04 4.87E+04 4.88E+04 4.88E+04 4.88E+04 4.88E+04 4.88E+04 Np240 3.93E+06 3.93E+06 2.81E+06 2.01E+06 1.82E+04 3.90E-01 6.20E-16 7.40E-16 1.01E-15 1.41E-15 2.25E-15 5.53E-15 Pu240 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 I 7.79E+04 7.79E+04 7.79E+04 7.79E+04 7.79E+04 Pu241 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.92E+07 1.922+07 1.91E+07 1.89E+07 1.87E+07 1.83E+07 1.66E+07 Am241 2.54E+04 2.54E+04 2.54E+04 2.54E+04 2.54E+04 2.55E+04 2.57E+04 2.79E+04 3.29E+04 4.02E+04 5.5 1 E+04 1. 11E+05 Am242m 1.66E403 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.66E+03 1.65E+03 1.65E+03 1.652+03 1.63E+03 A-24 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-I Activity (Curies) per Full Core of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_____ -Decay. timee Following Burnu p to 39 GWd/MTU Nuclide T 0 1 isec .30 nmin I hri 8 hr A".0d 4.Od 30.d:, 90.Od3 180d yr 3yr.Am242 1.14E+07 1.14E+07 1.12E407 1.09E+07 8.10E+06 4.04E+06 1.81E+05 1.65E+03 1.64E+03 1.64E+03 1.64E+03 1.62E+03 Cm242 6.67E+06 6.67E+06 6.67E+06 6.67E+06 6.67E+06 6.66E+06 6.59E+06 5.91E+06 4.58E+06 3.12E+06 1.42E+06 6.47E+04 Pu243 4.19E+07 4.19E+07 3.90E+07 3.64E+07 1.37E+07 I.46E+06 6.17E+01 3.93E-05 3.93E-05 3.93E-05 3.93E-05 3.93E-05 Am243 3.12E+03 3.12E+03 3.12E+03 3.12E+03 3.12E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 3.13E+03 Cm243 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.87E+03 2.86E+03 2.84E+03 2.80E+03 2.67E+03 Am244 1.38E+07 1.38E+07 1.33E+07 1.29E+07 7.95E+06 2.66E+06 1.90E+04 4.79E-15 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Cm244 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.90E+05 3.89E+05 3.87E+05 3.83E+05 3.76E+05 3.48E+05 A-25 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements .______ _____ ______ Decay tieFollowing Burnup to 58 GWd/MTU Nuclide- :lO-- l I see in Z -30mm 1:hr -8 hr -1.0d-- d 4.0 d '30.0d: 90.0 d 180.0d 1yr: 3 yr H 3 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E+02 1.43E402 1.43E+02 1.41E+02 1.39E+02 1.35E+02 1.21E+02 C 15 2.97E-02 2.24E-02 0.00E+00 O.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.OOE+0 0.00E+00 0.00+E00 0.00E+00 0.00E+00 N 16 2.73E-0I 2.48E-01 0.OOE+00 O.OOE+00 O.00E+00 0.00E400 O.00E+00 O.00E+00 0.00E+00 O.00E+00 0.00E+00 O.00E+00 Ne 23 5.29E-02 5.19E-02 1.50E-16 4.1 1E-31 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00* Na 24 4.16E-01 4.16E-01 4.07E-01 3.97E-01 2.85E-01 1.34E-01 4.45E-03 6.81E-16 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Na 25 3.19E-02 3.16E-02 2.59E-I I 2.09E-20 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+O0 0.OOE+00 0.002E00 O.00E+00 Mg27 1.61E+00 1.6]E+00 1.79E-01 1.99E-02 8.63E-16 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 O.00E+00 O.00E+00 Al 28 3.16E+01 3.15E+01 2.95E-03 2.74E-07 8.78E-10 5.17E-10 4.75E-11 4.94E-20 0.OOE+00 0.OOE+00 O.00E+00 0.00E+00 1128 2.90E+03 2.90E+03 1.26E+03 5.49E+02 4.78E-03 1.29E-14 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Al 29 8.38E-02 8.37E-02 3.53E-03 1 .49E-04 8.30E-24 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Si 31 6.65E-01 6.65E-01 5.83E-01 5.11E-01 8.03E-02 1.17E-03 6.33E-12 0.00E+00 0.00E+00 O.OOE+00 O.00E400 0.OOE+00 Ti 51 1.07E-01 1.07E-01 2.90E-03 7.84E-05 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 O.00E+00 0.00E400 Cr51 9.20E+03 9.20E+03 9.19E+03 9.19E+03 9.12E+03 8.97E+03 8.32E+03 4.34E+03 9.68E+02 1.02E+02 9.90E-01 1.15E-08 V 52 2.05E+02 2.05E+02 8.02E-01 3.13E-03 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.00+E00 0.00E+00 V 53 5.34E-01 5.30E-01 1.32E-06 3.22E-12 0.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Mn 54 6.65E+02 6.65E+02 6.65E+02 6.65E+02 6.65E+02 6.64E+02 6.59E+02 6.22E+02 5.45E+02 4.46E+02 2.96E+02 5.84E+01 Cr 55 1.56E+02 1.55E+02 4.07E-01 1.06E-03 O.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00OOE+00 0.00E+00 Fe.55 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.65E+03 3.64E+03 3.58E+03 3.43E+03 3.22E+03 2.84E+03 1.71E+03 Mn 56 1.95E+04 1.95E+04 1.70E+04 1.49E+04 2.27E+03 3.07E+01 1.2 1E-07 0.00E+00 O.OOE00 0 O.OOE+00 O.OOE+00 O.OOE+00 Mn 57 2.35E+00 2.33E+00 1.44E-06 8.77E-13 *0.OOE-00 0.00E400 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 O.00E+00 O.00E+00 Co 58 8.15E+02 8.15E+02 8.14E+02 8.14E+02 8.12E+02 8.07E+02 7.83E+02. 6.07E+02 3.38E+02 1.40E+02 2.29E+01 1.82E-02 Fe 59 2.18E+02 2.18E+02 2.18E+02 2.18E+02 2.17E+02 2.15E+02 2.05E+02 1.37E+02 5.37E+01 1.32E+01 7.38E-01 8.44E-06 Ni 59 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 5.16E-01 Co 60 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.27E+02 6.26E+02 6.2 1E+02 6.07E402 5.88E+02 5.50E+02 4.23E+02 A-26 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWd/MTU ___ _-...___._ Nuclide- IAsec -30mmin -..- Ahr 8.hr 1.0 d 4.0d-"- '30.0-d- 90.0 d -180.0d.- I yr_'. .3 yr...Co 60m 8.40E+02 8.39E+02 I. 15E+02 1.58E+01 1.33E-1I 0.00E+00 0.00E+00 O.OOE+00 O0.00E+00 O.OOE+00.O.E+00 O.00E+00 Co 61 1.68E+01 1.68E+01 1.36E+01 1.10E+01 5.82E-01 7.01E-04 5.13E-17 0.00E+00 0.00E+00 O.ooE+00 O.00E+00 0.00E+00 Co 62 8.50E-02 8.43E-02 8.12E-08 7.71E-14 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Ni63 7.67E+01 7.67E+01 7.67E+01 7.67E+01 7.66E+01 7.66E+01 7.66E+01 7.66E+01 7.65E+01 7.64E+01 7.61E+01 7.51E+01 Cu 64 1.01E+00 1.01E+00 9.85E-01 9.58E-01 6.54E-01 2.73E-01 5.37E-03 8.69E-18 0.00E+00 0.00E+00 0.OOE+00 O.OOE+00 Ni 65 9.17E+01 9.17E+01 7.99E+01 6.97E+01 1.02E+01 1.25E-01 3.12E-10 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 Cu 66 9.48E-01 9.46E-01 1.69E-02 1.09E-03 7.50E-04 6.12E-04 2.45E-04 8.92E-08 1.03E-15 1.28E-27 O.OOE+00 0.00E+00 Cu 73 2.56E+00 2.56E+00 2.55E+00 2.53E+00 2.28E+00 1.79E+00 6.13E-01 5.60E-05 2.66E-14 2.77E-28 0.00E+00 0.00E+00 Zn 73 2.57E+00 2.57E+00 2.57E+00 2.57E400 2.52E+00 2.23E+00 8.70E-01 8.03E-05 3.82E-14 3.97E-28 0.00E+00 O.00+E00 Ga 73 3.65E400 3.23E+00 0.00+E00 O.00E+00 O.00E+00 0.OOE+00 0.OOE-00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Ge 73m 6.52E+00 6.43E+00 0.002E-00 O.OOE+00 0.00E+00 O.00E+00 0.OOE+00 O.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Cu 74 6.64E+00 6.64E+00 6.20E+00 5.77E+00 2.13E+00 2.17E-01 7.55E-06 0.00+E00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 Zn 74 6.58E+00 6.56E+00 6.12E+00 5.69E+00 2.10E+00 2.14E-01 7.46E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Ga 74 4.23E+00 1.60E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E400 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 Cu 75 1.49E401 1.48E+01 3.40E-05 7.68E-1 1 0.00E+00 0.OOE+00 O.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 Zn 75 4.80E+00 4.80E+00 4.42E-01 3.41E-02 9.17E-18 0.00+E00 0.00E+00 0.00+E00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 Ga 75 4.98E+00 2.38E+00 O.00+E00 O.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00+E00 0.00E+00 0.00E+00 0.00E+00 Ge 75 3.24E+01 3.05E+01 0.00E+00 O.00+E00 O.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00r+00 0.00E+00 0.00E+00 0.00E+00 Cu 76 4.04E+01 4.03E+01 2.17E-03. 1.08E-07 0.00E+00 0.00E+00 0.00E-00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zn 76. 4.07E+01 4.07E+01 3.26E+01 2.53E+01 7.53E-01 2.43E-04 4.75E-20 0.00E+00. 0.00E+00 0.00+E00 0.00+E00 0.00E+00 Ga 76 3.82E+00 2.78E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.002+00 As 76 6.54E+01 5.79E+01 0.OOE+00 O.00+E00 0.00E+00 0.00+E00 O.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cu 77 1.0 1E+02 1.00E+02 2.77E-15 6.85E-32 0.00+E00 0.00E+00 0.00E+00. 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 Zn 77 5.54E+00 5.54E+00 5.47E+00 5.40E+00 4.49E+00 2.95E+00 4.42E-01 3.23E-08 1.09E-24 0.00+E00 0.00E+00 0.00E+00 Ga77 8.17E+01 5.88E+01 0.00E+00 O.00E+00 0.002+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 A-27 Safety Analysis Report for Susquehanna Units I and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-I10 Fuel Sum of Actinides, Fission Products and Light Elements________ ________ ______ .Decay time Following Burnup to 58 GWd/MTU Nuclide T=- 0 1 sec 30mn 1 hr I .... 8hre ...... 1.0d 4.0d 30.0d 90.0d 180.0d 1 yr 3yr Ge 77 2.18E+02 2.10E+02 0.OOE+00 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+O0 0.00+E00 O.00E+00 0.00+E00 0.00E+00 0.00+E00 Ge 77m 8.69E+01 8.69E+01 8.44E+01 8.18E+01 5.32E+01 2.00E401 2.41E-01 5.74E-18 0.00E+00 O.OOE+00 0.00E+00 O.OOE+00 As 77 2.24E402 2.24E+02 1.72E-08 9.78E-19 0.00E+00 0.00+E00 0.00E+00 0.00E400 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Zn78 2.64E+02 2.64E+02 2.63E+02 2.61E+02 2.38E+02 1.87E+02 5.40E+01 7.87E-04 5.41E-15 9.13E-32 0.00E+00 0.00E+00 Ga 78 1.13E+02 7.04E+01 0.00E+06 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 Ge 78 5.20E+02 4.65E+02 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.00E400 0.00+E00 0.00+E00 0.OOE+00 0.OOE+00 As 78 9.12E+02 9.12E+02 7.20E+02 5.69E4 02 2.08E+01 1.08E-02 1.80E-17 0.00E+00 0.OOE+00 0.001+00 .0.OOE+00 0.00E+00 Zn 79 9.26E+02 9.26E+02 9.02E+02 8.48E+02 1.04E+02 1.57E-01 1.75E-15 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Ga 79 5.11E+01 2.55E+01 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Ge 79 4.54E+02 3.68E+02 0.00E+00 O.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 As 79 1.50E+03 1.46E+03 " 0.00+E00 O.002+00 0.OOE+00 0.000+00 0.002+00 0.OOE+00 0.002+00 0.000+00 O0.00E00 0.00E+00 Se 79m 1.65E+03 1.64E+03 1.69E+02 1.68E+01 1.56E-13 0.00+E00 0.002+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zn 80 1.64E+03 1.64E+03 2.90E+02 2.94E+01 2.73E-13 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.002+00 0.002+00 Ga 80 2.20E+01 6.07E+00 O.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 Ge 80 4.18E+02 2.79E+02 O.00E+00 O.00+E00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 O.00+E00 As 80 3.45E+03 3.38E+03 1.50E-15 O.00+E00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 Zn 81 4.13E+03 4.10E+03 3.10E-15 0.00E+00 O.00E+00 0.002E 00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Ga 81 5.16E+00 1.80E-02 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ge 81 2.60E+02 1.48E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 81. 3.66E+03 3.35E+03 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 81 6.07E+03 6.0 1E+03 3.85E-13 2.03E-29 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E-00 Se 81m 6.42E+03 6.42E+03 2.41E+03 9.63E+02 2.11E+00 1.89E-05 0.00E+00 0.00+E00 0.00E+00 0.002+00 0.002+00 0.002+00 Ga 82 4.76E+02 4.76E+02 3.32E+02 2.3 1E+02 1.43E+00 1.28E-05 0.00E+00. 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 Ge 82 1.33E+02 4.18E+01 0.00+E00 O.00E+00 0.00+E00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E-00 As 82 3.41E+03 2.94E+03 O.00E+00 O.00E+00 0.00E+00 0.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 A-28 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006.-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Nuclide -T=0 sec -30mlin- -1 hr-- 8 h -.0d -4.0 d 30.0-d--- -90.0d--- 180.0 d__- l-yi-I........ 3 yr...As 82m 5.62E+03 5.53E+03 0.00+E00 O.00E4-00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 Br 82 2.2 1E+03 2.10E+03 O.OOE+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 .0.00E+00 0.OOE+00 0.OOE+00 0.00+E00 Br82m 7.95E+02 7.95E+02 7.89E+02 7.81E+02 6.81E+02 4.97E+02 1.21E+02 5.78E-04 3.05E-16 0.00E+00 0.00+E00 0.00E+00 Ga 83 7.01E402 7.00+E02 2.36E+01 7.92E-01 1.88E-21 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 Ge 83 2.14E+01 2.27E+00 0.00E+00 O.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 As 83 1.63E403 1.13E+03 0.00E+00 O.00E+00 0.00E+00 0.00E4+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Se 83 8.83E+03 8.46E+03 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 Se 83m 7.02E+03 7.0 1E+03 2.77E+03 1.09E+03 2.34E-03 2.56E-16 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Br 83 7.15E+03 7.14E+03 1.62E-04 2.99E-12 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Kr 83m 1.45E+04 1.45E+04 1.32E+04 1.17E+04 1.57E+03 1.55E+01 1.44E-08 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 Ga 84 1.47E+04 1.47E+04 1.46E+04 1.42E+04 4.10E+03 5.93E+01 6.08E-08 0.00+E00 0.00+E00 0.00E+00 0.00E+00 O.00E+00 Ge 84 9.44E+01 8.10E-02 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 As 84 1.07E+03 6.05E+02 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 Se 84 6.69E+03 5.99E+03 0.00E+00 O.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Br 84 2.57E+04 2.56E+04 3.91E+01 5.88E-02 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.00+E00 Br 84m 2.64E+04 2.64E+04 1.53E+04 7.94E+03 8.39E-01 6.86E-10 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ge 85 7.41E+02 7.39E+02 2.32E+01 7.23E-01 6.13E-22 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E400 O.00E+00 As 85 1.90E+02 1.18E+01 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 85 3.67E+03 2.62E+03 0.00+E00 O.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E400 0.00+E00 0.00E+00 0.00E+00 0.00+E00 Se 85m 1.17E+04 1.14E404 9.55E-14 7.53E-31 0.00+E00 0.00+E00 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Br 85 1.02E+04 9.82E+03 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 O.00+E00 Kr 85 2.85E+04 2.85E+04 2.32E+01 1.65E-02 O.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 O.00E+00 Kr 85m 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.49E+03 2.48E+03 2.45E+03 2.41E+03 2.33E+03 2.05E+03 Ge 86 2.88E+04 2.88E+04 2.69E+04 2.49E+04 8.44E+03 7.10E+02 1.03E-02 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 As 86 4.042+01 2.42E+00 0.00+E00 0 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 A-29 Safety Analysis Report for Susquehanna Units I and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements.....Decay time Following Burnup to 58 GWdIMTU ______ __._:.___Nuclide::: .T= 0 .. sec:.. 30 min..- :1 hr.,' -8 hr: :1.0 d -4.0 dn. '30.0 d 90.0d7: 180.0 d- 1 yr 3 yr Se 86 1.95E+03 9.10E+02 0.00E+00 0.OOE+0O 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Br 86 2.73E+04 2.61E+04 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Br 86m 3.42E+04 3.411E+04 6.58E-06 9.57E-16 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 86 6.93E+03 5.94E+03 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Rb 86m 4.64E+02 4.64E+02 4.63E+02 4.63E+02 4.58EP02 4.47E+02 4.OOE+02 1.52E+02 1.63E+01 5.73E-0I 5.83E-04 9.22E-16 Ge 87 3.80E+01 3.76E+01 5.03E-08 6.58E-17 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 As 87 1.96E+02 1.10E+00 0.OOE+00 0.OOE+00 0.003E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 87 9.52E+02 1.06E+02 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Br 87 1.57E+04 1.39E+04 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.001E+00 0.OOE+00 0.00E+00 0.00E+00 Kr 87 4.46E+04 4.42E+04 8.66E-06 1.60E-15 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.003E400 0.00E+00 As 88 5.70E+04 5.70E+04' 4.38E+04 3.34E+04 7.35E+02 1.20E-01 1.10E-18 0.00E+00 0.OOE+00 0.OOE+00 0.001+00 0.OOE+00 Se 88 4.45E+00 4.45E+00 3.94E+00 3.48E+00 6.20E-01 1.20E-02 2.34E-10 0.00E+00 0.00E+00 0.OOE+00 0.001E+00 0.OOE+00 Br 88 3.78E+02 2.19E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 Kr 88 8.85E+03 5.59E+03 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 88 4.18E+04 4.04E+04 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zr 89 7.80E+04 7.80E+04 6.911E+04 6.11 1+04 1.111E+04 2.23E+02 5.16E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 As 89 8.02E+04 8.02E+04 7.50E+04 6.76E+04 1.24E+04 2.48E+02 5.76E-06 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 Se 89 8.711E01 1 8.711E+01 8.67E+01 8.64E+01 8.12E+01 7.05E+01 3.73E+01 1.50E-01 4.48E-07 2.30E-15 2.28E-32 0.00E+00 Br 89 1.15E+01 3.73E-02 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr 89. 3.10E+03 5.711E+02 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 89 2.98E+04 2.57E+04 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.001E+00 0.00E+00 0.00E+00 0.001E300 Sr 89 9.44E+04 9.42E+04 1.35E+02 1.91E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.001E+00 0.00E+00 0.00E+00 Y 89m 1.031E+05 1.03E+05 3.26E+04 8.30E+03 3.99E-05 3.88E-24 0.00E+00. 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Se 90 1.07E+05 1.07E+05 1.07E+05 1.07E+05 1.06E+05 1.05E+05 1.0 11E+05 7.07E+04 3.11 E+04 9.04E+03 7.13E+02 3.19E-02 Brg0 3.10E+02 3.011E+02 9.62E+01 9.58E401 9.06E+01 7.98E+01 4.64E+01 6.73E+00 2.89E+00 8.411E-01 6.63E-02 2.96E-06 A-30 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements______.. _.. _______ Decay time.Following Burnup to 58GWd/MTU __- -____" Nuclide:l, T-=0-., :Isec j,_ý 30min, -1 hr, 8hr -1.0d .. 4.0d_... 30.0d ,_ 90.0d .180.0d yr.. 3yr-Kr 90 7.19E402 1.41E+02 0.00+E00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00+E00 0.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 Rb 90 1.68E+04 1.18E+04 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 0.00+E00 0.OOE+00 0.00+E00 0.00+E00 Rb 90m 9.99E+04 9.80E+04 1.74E-12 2.95E-29 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 Sr 90 9.33E+04 9.32E+04 4.86E+01 1.32E-01 0.OOE+00 0.OOEs 00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 Y 90 3.15E+04 3.15E+04 2.64E+02 2.09E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.OOE+00 0.00E+00 Zr 90m 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.24E+04 2.23E+04 2.23E+04 2.21E+04 2.18E+04 2.08E+04 Se 91 2.36E+04 2.36t+04 2.36E+04 2.36E+04 2.35E+04 2.33E+04 2.28E+04 2.24E+04 2.23E+04 2.2 1E+04 2.19E+04 2.08E+04 Br 91 2.45E+01 1.04E+01 O.00E+00 O.00E+00 O.OOE+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Kr 91 7.99E+01 6.092+00 0.OOE+00 O.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 Rb 91 5.17E+03 1.64E+03 0.OOE+00 O.OOE+00 0.00+E00 0.00+E00 O.OOE+00 0.OOE+00 0.00E400 0.00E400 0.00E400 O.OOE+00 Sr 91 6.90E+04 6.39E+04 O.00E+00 O.00+E00 0.00E+00 O.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Y 91 1.29E+05 1.28E+05 7.43E-05 3.89E-14 0.OOE+00 0.00E+00 0.00E+00 0.002E00 0.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 Y 91M 1.4 1E+05 1.4 1E+05 1.36E+05 1.3 1E+05 7.86E+04 2.45E+04 1.30E+02 2.40E-18 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 Se 92 1.44E+05 1.44E+05 1.44E+05 1.44E+05 1.44E+05 1.43E405 1.39E+05 1.02E+05 5.00E+04 1.72E+04 1.92E+03 3.35E-01 Br92 8.15E+04 8.15E+04 8.10E+04 7.97E+04 4.99E+04 1.56E+04 8.23E+01 1.52E-18 0.00E+00 0.00+E00 0.00E+00 0.00E+00 Kr 92 7.00E+00 1.13E-01 0.00+E00 O.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Rb 92 9.37E+02 1.40E+02 0.00E+00 O.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Sr 92 3.772+04 2.60E+04 0.00E+00 O.00E+00 O.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 Y 92 1.16E+05 1.04E+05 0.00+E00 O.00+E00 0.00E+00 O.00O+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 Zr 93 1.55E+05 1.55E+05 1.36E+05 1.20E+05 2.00E+04 3.34E+02 3.36E-06 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 Br 93 1.56E+05 1.56E+05 1.55E+05 1.52E+05 7.28E+04 4.93E+03 4.51E-03 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00" Kr 93 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 6.27E-02 Rb 93 3.98E+02 7.73E+00 0.00E+00 O.00E+00 0.00E+00 0.002+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr 93 1.33E+04 7.76E+03 O.00E+00 O.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 93 1.02E+05 9.18E+04 0.002E-00 0.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 A-31 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements___... ..-.Decay tie lowingBurnup to 58 GWdIMTU._____________________ Nuclide T=0-:,, 1 sec 30minm Ihr *8hr. 1.0d '-4.0dd -30.0d 90.0d 180.0d lyr 3yr Br94 1.81E+05 1.81E+05 1.IIE+04 6.72E+02 6.251-15 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 Kr 94 1.23E+05 1.23E+05 1.20E+05 1.16E+05 7.19E+04 2.40E+04 1.71E+02 4.32E-17 0.00E+00 0.004E00 0.00E+00 0.00E+00 Rb 94 1.75E+01 3.32E-02 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 O.OOE+00 O.OOE400 0.00E+00 0.OOE+00 Sr 94 6.43E+03 2.36E+02 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Y 94 5.45E+04 4.25E+04 0.00E+00 0.004E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 Kr95 1.84E+05 1.83E+05 1.16E-02 7.20E-10 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 Rb 95 2.00E+05 2.00E+05 7.02E+04 2.31E+04 4.00E-03 1.41E-18 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 Sr 95 7.80E+02 3.20E+02 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 Y 95 2.69E+04 4.75E+03 0.OOE+00 0.00E+00 0.004E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.OOE+00 0.004E00 Zr 95 1.66E+05 1.62E+05 4.33E-17 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE400 0.OOE+00 0.00E+00 Nb 95 2.12E+05 2.12E+05 3.02E+04 4.17E+03 3.79E-09 0.00+E00 0.00+E00 0.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 0.00E+00 Nb 95m 2.33E+05 2.33E+05 2.33E+05 2.33E+05 2.32E+05 2.30E+05 2.23E+05 1.68E+05 8.79E+04 3.32E+04 4.46E+03 1.64E+00 Kr96 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.34E+05 2.33E+05 2.17E+05 1.47E+05 6.51E+04 9.62E+03 3.61E+00 Rb 96 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.59E+03 2.59E403 2.55E+03 1.98E+03 1.03E+03 3.90E+02 5.25E+01 1.93E-02 Sr 96 1.27E+02 1.18E+01 0.00+E00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00+E00 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 Y 96 6.76E+03 2.29E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 Nb 96 1.23E+05 6.48E+04 0.OOE+00 0.00E+00 O.OOE+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 Rb 97 2.08E+05 1.95E+05 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 Sr97 4.78E+02 4.78E+02 4.71E+02 4.64E+02 3.78E+02 2.35E+02 2.77E+01 2.50E-07 6.80E-26 0.00E+00 0.00E+00 0.00E+00 Y97 .1.61E+03 2.82E+01 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Zr 97 6.20E+04 1.20E+04 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.002+00 Nb 97 1.72E+05 1.45E+05 0.00E+00 O.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Nb 97m 2.43E+05 2.43E+05 2.38E+05 2.34E+05 1.75E+05 9.09E+04 4.74E+03. 3.63E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb 98 2.45E+05 2.45E+05 2.44E+05 2.41E+05 1.88E+05 9.13E+04 4.76E+03 3.92E-08 0.00E+00 0.00+E00 0.00+E00 0.00E+00 Sr 98 2.30E+05 2.30E+05 2.26E+05 2.22E+05 1.66E+05 8.62E+04 4.50E+03 3.46E-08 0.00+E00 0.00+E00 0.00E+00 0.00E+00 A-32 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements 11 7 ~ ecay- tii_ lwing Burnup toS G dIT_________ N6c6&d T 0 0 I sec 3Omi hil Ir 8hr .0&d 4.0'd' '0O 90.0-d l80.60d,_ I yr 3 yr Y 98 2.2 1E+02 5.22E-01 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 0.00E400 0.OOE+00 Zr 98 2.69E+04 9.23E+03 0.00E+00 O.0013+00 0.0E+00 0.OOE+00 0..0E+00 0.000+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Nb 98 1.28E+05 5.32E+04 O.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00OE+00 0.001+0 0.001E+00 0.OOE+00 0.OOE+00 Nb 98m 2.34E+05 2.3 1E+05 5.36E-13 1.19E-30 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Rb 99 2.37E+05 2.36E+05 5.91E-13 1.30E-30 O.OOE+00 O.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E400 0.OOE+00 0.OOE+00 Sr 99 2.12E+03 2.12E+03 1.4 11E+03 9.4 1 E+02 3.23E+00 7.5 1E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 Y 99 6.59E+00 5.07E-05 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 O.00E400 O.00E+00 0.00E+60 Zr 99 9.12E+03 7.03E+02 0.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.OOE+00 Nb 99 8.07E+04 5.14E+04 O.00E+00 .001E+00 0.001+00 0.OOE+00 O0.00E+00 0.00E+00 0.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 Nb 99m 2.34E+05 1.86E+05 0.00E+00 0.00E+00 0.003E400 0.00E-400 0.003E400 0.00E+00 O.OOE400 O.00E400 O.00E+00 0.001E+00 Mo 99 1.511E+05 1.50E+05 O.00E+00 O.00E+00 0.00E+00 0.00E+00 O.00E+00 O.00E+00 O.00E+00 O.00E+00 0.00E+00 0.OOE+00 Tc 99 1.04E+05 1.04E405 3.56E+01 1.199E-02 O.OOE400 O.00E400 0.00E+00 O.00E-400 0.00E+00 0.OOE+00 .003E+00 O.OOE+00 Tc 99m 2.611E+05 2.611E+05 2.59E+05 2.58E+05 2.40E+05 2.02E+05 9.50E+04 1.35E+02 3.60E-05 4.96E-15 0.00E+00 0.00E+00 Sr100 3.97E+00 3.97E+00 3.97E+00 3.97E+00 3.97E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 3.98E+00 Yl00 2.32E+05 2.32E+05 2.32E+05 2.31E+05 2.24E+05 1.95E+05 9.20E+04 1.30E+02 3.48E-05 4.80E-15 0.00E+00 0.OOE+00 ZrlO0 1.57E+03 5.03E+01 0.001E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 Nbl00 2.92E+04 1.15E+04 0.00E+00 .001E+00 O.00E+00 O.OOE400 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nb I 00m 2.25E+05 2.06E+05 0.00E+00 0.001+00 0.001+00 .OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 TcI00 2.48E+05 2.36E+05 0.00E+00 O.00E+00 0.00E+00 0:00E+00 0.00E+00 0.00E+00 0.00E3400 0.001+0 0.001E400 O.00E+00 SrI01 2.37E+04 1.87E+04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Y101 1.04E+05 9.93E+04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Zrl01 2.17E+02 6.05E+00 0.00E+00 O.00E+00 0.00E+00 0.00E4 00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nb101 1.25E+04 3.14E+03 0.00E+00 O.00E+00 0.00E+00 0.001E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Mol01 1.36E+05 9.83E+04 0.00E+00 0.00E+00 0.00E+00 0.001E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 Tc001 2.28E+05 2.18E+05 0.00E+00 O.00E400 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E3400 A-33 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_______i i :Decay time Following Burnup to 58,GWd/MTU __.____Nuclide T 0::, I sec 30min -1 hr 8 1.0 d 4.Od 30d 90.0 d 180.0 d Iyr 3yr Sr102 2.43E+05 2.43E+05 5.89E+04 1.42E+04 3.10E-05 4.99E-25 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 O.OOE+00 Y102 2.43E+05 2.43E+05 1.41Ef05 5.29E+04 5.53E-04 1.56E-23 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 ZrI02 3.20E+01 2.85E+00 0.OOE+00 O.00E+00 0.OOE+00 0.06E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nbl02 4.65E+03 2.15E403 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 MoI02 9.18E+04 7.29E+04 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 O.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 Tc102 1.98E+05 1.50E+05 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00+E00 O.00+E00 0.00E+00 O.00E+00 TcI02m 2.38E+05 2.38E+05 3.79E+04 6.01E403 3.89E-08 0.004E00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 0.00+E00 Y103 2.38E+05 2.38E+05 3.82E+04 6.06E+03 3.92E-08 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 Zr103 2.61E+02 2.60E+02 2.19E+00 1.84E-02 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE400 0.OOE+00 0.OOE+00 Nb103 1.30E+03 9.04E+01 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 MoI03 3.71E+04 2.19E+04 O.00E+00 O.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00+E00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 Tcl03 1.55E+05 1.08E+05 0.00E+00 O.002+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.00E+00 Ru103 2.43E+05 2.42E+05 2.33E-03 2.17E-1I 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 Rhl03m 2.47E+05 2.47E+05 1.17E-02 1.0-10 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E400 0.00+E00 O.00E+00 Y104 2.48E+05 2.48E+05 2.48E+05 2.48E+05 2.47E+05 2.44E+05 2.31E+05 1.46E+05 5.07E+04 1.03E+04 3.93E+02 9.83E-04 Zrl04 2.48E+05 2.48E+05 2.48E+05 2.48E+05 2.46E+05 2.43E+05 2.31E+05 1.46E+05 5.06E+04 1.03E+04 3.92E+02 9.82E-04 Nbl04 6.42E+01 2.90E-01 0.00E+00 O.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E400 Mo104 1.21E+04 9.22E+03 0.00E+00 O.00+E00 0.00E+00 0.002E00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 Tcl04 7.82E+04 6.91E+04 0.00E+00 O.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 Rh 104 2.04E+05 2.03E+05 1.98E-04 1.83E-13 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 RhI04m 2.15E+05 2.15E+05 7.28E+04 2.34E+04 '2.88E-03 4.65E-19 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 Zr105 1.98E+05 1.95E+05 1.43E+02 1.19E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 Nb105 1.45E+04 1.44E+04 1.20E+02 9.95E-01 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Mo105 2.53E+03 6.18E+02 0.00+E00 O.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 Tc005 3.24E+04 2.58E+04 0.00E+00 O.00E400 0.00E400 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.002+00 0.00E+00 A-34 Safety Analysis Report for Susquehanna Units 1 and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay. time Following Burnup to 58 G~Vd/MTU_____________ Nuclide T 0 1. sec-, 30min. -hr 8 hr 1.0 d 4.0d 30.0d 90.0d :180.0d lyr- 3yr Ru105 1.55E405 1.53E+05 9.58E-1I 5.70E-26 O.OOE+00 0.OOE400 0.OOE+0O 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Rh105 1.85E+05 1.85E+05 1.29E+04 8.34E+02 1.93E-14 0.00OE+00 ..OOE+0 O.OOE+00 O.00E-fOO0.OOE+00 O.OOE+00 0.00E+00 RhI05m 1.89E+05 1.89E+05 1.80E+05 1.66E+05 5.58E+04 4.59E+03 6.01E-02 0.00E+00 0.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 Zrl06 1.74E+05 1.74E+05 1.74E+05 1.74E405 1.64E+05 1.25E+05 3.07E+04 1.50E-0I 8.25E-14 2.28E-32 0.00E+00 0.00E+00 Nb 106 5.36E+04 5.36E+04 5.12E+04 4.74E+04 1.59E+04 1.3 1E+03 1.71E-02 0.00E400 0.OOE+00 0.02+00 0.00+E00 0.OOE+00 Mo106 1.87E+02 8.69E+01 0.OOE+00 O.00+E00 0.00E+00 0.00E+00 O.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tc006 7.32E+03 3.71E+03 0.OOE+00 0.OOE+00 O.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 O.OOE+00 Ru106 9.33E+04 8.63E+04 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 Rh106 1.40E+05 1.39E+05 1.51E-10 1.33E-25 O.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 Rh106m 1.26E+05 1.26E+05 1.26E+05 1.26E405 1.26E405 1.26E405 1.25E405 1.19E+05 1.06E405 8.99E+04 6.36E+04 1.63E+04 Nb107 1.36E+05 1.36E+05 1.26E+05 1.26E+05 1.26E+05 1.26E+05 1.25E+05 1.19E+05 1.06E+05 8.99E+04 6.36E+04 1.63E+04 MoWI7 4.69E+03 4.69E+03 4.OOE+03 3.41E+03 3.63E+02 2.17E+00 2.16E-10 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 Tc107 1.38E+03 5.56E+02 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 RuI07 4.02E+04 3.3 12E4 04 0.OOE+00 O.004E00 0.002400 O.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 RhI07 9.95E+04 9.75E+04 0.00+E00 O.OOE+00 0.OOE+00 0.OOE+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 0.OOE+00 Pdl07m 1.18E+05 1.18E+05 5.05E+02 1.97E+00 0.00E+00 0.00Es00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 NbIO8 1.18E+05 1.18E+05 5.56E+04 2.14E+04 3.19E-02 1.53E-15 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 0.00E+00 0.OOE+00 Mo108 1.78E+03 1.72E+03 0.00+E00 0.00E+00 0.OOE+00 0.004E00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 TC108 4.41E+01 2.53E+00 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.004E00 0.00E+00 0.00E+00 0.OOE+00 Ru108 6.47E+03 4.08E+03 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 Rh108 3.67E+04 3.27E+04 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 RhI08m 7.74E+04 7.73E+04 8.09E+02 8.37E+00 0.00E+00 0.002E00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 MoI09 7.85E+04 7.84E+04 8.63E+02 8.92E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 TC109 1.11E+03 1.11E+03 3.47E+01 1.08E+00 9.18E-22 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 O.00+E00 0.OOE+00 0.OOE+00 Rul09 6.76E+02 4.13E+02 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 A-35 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_____.. .. De.ay tieol.wn cru 'to 58 GWdIMTU ____ _____ ____Nuclidel T-.0:,--. -see-:.- :30 min- :1 hr 8 hr" I.0 d" -:4.0 d 30.0 d 90.0 d- i180.0 d I yr 3yr-Rh109 1.22E+04 7.63E+03 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.002E00 O.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 Rhl09m 5.11E+04 5.03E+04 1.71E-1 I 5.59E-27 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Pd109 5.88E+04 5.87E+04 3.06E-02 5.14E-09 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Pdl09m 2.94E+04 2.93E+04 1.31 E-06 1.89E-17 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 Agl09m 7.74E+04 7.74E+04 7.56E+04 7.37E+04 5.18E+04 2.30E+04 6.03E+02 1.17E-11 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Mo 110 5.25E+02 5.24E+02 6.24E+00 7.39E-02 0.OOE+00 0.OOE+00 O.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Tc1 10 7.74E+04 7.74E+04 7.57E+04 7.38E+04 5.18E+04 2.30E+04 6.03E+02 3.55E-04 3.24E-04 2.83E-04 2.15E-04 7.18&-05 Ru! I0 6.94E+01 5.40E+01 O.OOE+00 O.OOE000 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 O.00E+00 Rhl I0 2.02E+03 9.10E+02 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 RhIlOta 1.70E+04 1.63E+04 0.OOE+00 O.002E00 0.00E+00 0.OOE+00 0.00E-00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Agl 10 2.58E+03 2.07E+03 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Ag! lOi 1.96E+04 1.96E+04 3.79E-15 0.00+E00 0.00E+00 O.OOE00 0 O.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 Mol I I 4.66E+04 4.53E+04 1.74E+01 1.74E+01 1.74E+01 1.74E401 1.73E+01 1.60E+0I 1.36E+01 1.06E+01 6.33E+00 8.34E-01 Tcl I I 1.28E403 1.28E203 1.28E+03 1.28E+03 1.28E+03 1.28E+03 1.27E+03 1.18E+03 9.99E+02 7.78E+02 4.65E+02 6.13E+01 Rul I ! 5.45E+00 1.23E+00 0.00+E00 O.00+E00 0.00+E00 0.00+E00 O.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 Rh! I! 3.82E+02 2.70E+02 0.00E+00 O.00E+00 O.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 O.00+E00 Pdl I I 5.73E+03 3.82E+03 0.00E+00 O.00E400 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0o.00E+00 0.00E+00 0.00E+00 Pd! IIm 1.07E+04 1.03E+04 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 Ag! I ! 1.14E+04 1.14E+04 4.94E+03 2.21E+03 1.35E+02 1.80E+01 2.06E-03 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00+E00 Ag IIm 4.71E+02 4.71E+02 4.43E+02 4.16E402 1.72E+02 2.29E401 2.62E-03 O.00E+00 O.00E+00 0.00+E00 0.00E+00 O.00+E00 CdlI Im 1.16E+04 1.16E+04 1.16E+04 1.16E+04 1.13E+04 1.06E+04 8.01E+03 7.13E+02 2.68E+00 6.20E-04 2.032E-I1 0.00E+00 Tc 112 1.16E+04 1.16E+04 5.22E+03 2.38E+03 1.68E+02 2.24E+01 2.57E-03 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ru! 12 9.55E+0! 9.55E+01 6.23E+01 4.06E+01 1.02E-01 1.15E-07 O.00E+00 0.00E+00 O.00+E00 0.00+E00 0.00+E00 0.00E+00 Rh 112 5.59E+01 1.15E+01 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Pd112 1.80E+03 1.48E403 0.00+E00 O.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.00+E00 O.00E+00 0.00E+00 0.00+E00 A-36 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel.Sum of Actinides, Fission Products and Light Elements Deayti -Followiing Btirnup to 58 GWd/MTU ____Nuclide-.. -T= 0:_-_- :lsec-:iI-- '30 min-.i .1 hir ;8h.:.. :.1.0-d- .4.0d'::. 30.0 d: 90.0d- -180.0 d77 lyr. 3 yr Ag1 12 4.12E+03 3.20E+03 0.00E+00 O.00E400 0.00E+00 0.OO300 0.003E+O0 0.001E+00 O.OOE+00 0.00E+00 0.001E+00 0.00E+00 In 113m 5.01E+03 5.01E+03 4.93E+03 4.85E+03 3.85E+03 2.27E+03 2.12E+02 2.51E-07 6.34E-28 0.00E+00 0.00E+00 0.00E+00 Sn! 13 5.03E+03 5,03E+03 5.02E+03 5.011403 4.38E+03 2.67E+03 2.49E+02 2.95E-07 7.45E-28 0.00E+O0 0.00E+00 O.OOE+00 SnI13m 5.27E+02 5.27E+02 5.27E+02 5.27E+02 5.26E+02 5.241402 5.1413+02 4.40E+02 3.0613+02 1.78E+02 5.841+01 7.17E-01 TcS113 5.27E+02 5.27E+02 5.27E+02 5.27E+02 5.26E+02 5.23E+02 5.14E+02 4.40E+02 3.06E+02 1.78E+02 5.84E+01 7.17E-01 Ru! 13 !.70E+02 1.70E+02 6.45E+01 2.44E+01 3.02E-05 9.44E-19 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 O.OOE+00 Rh! 13 1.13E+01 3.88E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.0013E+00 O.OE+ O .OO E+0 .0O.OOE+O0 O.OOE+00 Pd! 13 5.43E+02 4.32E+02 0.00E+00 O.00E+00 O.OOE+00 0.00E+00 O.OOE+00 0.00E+00 O.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 Ag! 13 1.94E+03 1.16E+03 O.OOE+00 O.001EO0 O.OOE-.-0 O.0004.OOE+0 .OOE3+0 O .OOE+O00 0.OOE1+00 0.OOE+00 0.OOE+00 0.OOE-00 Ag 113m 2.82E+03 2.8 1E+03 4.27E-03 6.34E-09 .003E+00 0.00E+00 0.00E+00 O.OO13+00 0.001+00 0.003+00 0.0013+00 0.001+00 Cd! 13m 2.74E+03 2.74E+03 2.58E+03 2.42E+03 9.81E+02 1.24E+02 1.14E-02 O.00E+00 0.OOE+00 0.00E+00 O.OOE+00 0.00E+00 In! 14 5.48E+02 5.48E+02 3.01E-03 4.49E-09 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 In114m 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E+01 1.12E01+01 1.1 E+01 1.!11+01 1.07E+01 9.69E+00 Rull4 1.36E+02 1.35E+02 8.61E+0I 8.61E+01 8.57E+01 8.49E+01 8.14E+01 5.66E+01 2.44E+01 6.93E+00 5.18E-01 1.88E-05 Rh!14 9.00E+01 9.OOE+01 8.99E+01 8.99E+01 8.95E+01 8.87E+01 8.51E+01 5.91E+O1 2.55E101 7.24E+00 5.41E-0I 1.96E-05 Pd! 14 1.72E+02 !.58E+02 0.003E+00 O.OOE+00 O.0013E00 0.OOE+00 0.OOE1+00 0.OOE+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 Ag! 14 .9.0513E02 6.57E+02 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Rul 15 2.06E+03 2.05E+03 4.29E-01 8.80E-05 0.00E+00 0.00E+00 0.00E+00 0.60E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Rh! 15 2.13E+03 2.12E+03 4.42E-01 9.09E-05 0.00E+00 0.00E+00 O.OOE001 0.0OOE÷00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Pd 115 3.60E+01 1.63E+01 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Ag!15 3.54E+02 3.27E+02 0.0013+00 0.OOE+0 0.0013+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E200 0.00E+00 0.00E+00 Ag! 1Sm 1.201+03 1.19E+03 7.21E-12 3.93E-26 0.00E+00 0.00E+00 0.OOE+00 0.001+00 0.00E+00 0.00E+00 0.002+00 0.OOE+00 Cd! 15 9.53E+02 9.53E+02 3.48E+02 1.23E+02 5.86E-05 2.08E-19 0.00E+00. 0.OOE+00 0.002400 0.002+00 0.00E+00 0.00E+00 Cd1 15m 3.99E+02 3.96E+02 3.70E-12 2.02E-26 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.001+00 0.00E+00 0.00E+00 In 15m 1.43E+03 1.43E+03 1.42E+03 1.41E+03 1.29E+03 1.05E+03 4.13E+02 1.272-01 9.902-10 6.85E-22 0.00E+00 0.00E+00 A-37 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements________!' ______. : Decay time Following Burnup to 58 GWdlMTU _____ ____-... .Nuclide..

T=V ... I sec 30minm -hr.. 8hr. 1.0dd -4.0d .30.0d'. 90.0d 180.0d: lyr

• 3yr Ru! 16 7.03E+01 7.03E+01 7.03E+01
• 7.03E+01 7.00E+01 6.93E+01 6.61E+01 4.41E+01 1.74E401 4.29E+00 2.41E-01 2.82E-06 RhlI6 1.43E+03 1.43E+03 1.43E+03 !.43E+03 1.37E+03 1.14E+03 4.50E+02 1.43E-01 .1.92E-03 4.74E-04 2.66E-05 3.12E-10 Pd! 16 1.06E+01 7.04E +00 0.00E+00 0.00E+00 0.00 0E+00 0.00E+00 0.00E+00 0.00E+00 0.O0E+00 0.00E+00 0.00E+00 AgI 16 1.56E+02 7.95E+01 0.00E+00 0.OOE+00
• 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Ag 116m 1.20E+03 1.15E+03 O.00E400 O.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.00+E00 0.00E+00 O.OOE+00 0.00E+00 In 116 1.39E+03 1.39E+03 6.39E-01 2.72E-04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+/-00 0.00+E00 0.00E+00 In I6m i .82E+02 1.70E+02 0.00E+00 0.00 0E+00 0.00E+00 0.00E1300 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00.+00 RhI 17 2.83E+02 2.69E+02 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 Pd 117 i.06E+03 1.06E+03 7.25E+02 4.94E+02 2.28E+00 1.05E-05 0.00E+00 0.00E+00 0.001+00 0.00E+00 0.00E+00 0.00E+00 AgI 17 6.241+01 3.55E+01 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.00E+00 0.00E+00 0.00+E00 0.00E+00 Ag! 17m 8.40E+02 7.37E+02 0.001+00 0.00+E00 0.001+00 0.OOE+00 0.001+00 0.00E+00 0.001+00 0.00E+00 0.001+00 0.001+00 CdI 17 7.65E+02 7.61E+02 2.88E-05 1.03E-12 O.002E+00 0.00+00 0.00E+00 0.00E+00 0.000+00 0.00.E+00 0.00E+00 .003E+00 Cd I 17m 7.65E+02 7.19E+02 O.00E+00 0.001+00 0.001+00 0.001+00 0.001+00 0.0013+0 0.00E+00 0.001+00 0.003+00 O.00E+00 In 117 1.34E+03 1.34E+03 1.17E+03 1.02E+03 1.46E+02 1.69E+00 3.34E-09 0.001+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 In 117m 3.09E+02 3.09E+02 2.79E+02 2.52E+02 5.95E+01 2.191+00 7.81E-07 0.00E+00 0.001400 0.00E+00 0.OOE+00 0.00E+00 Sn1 17m 1.00E+03 .001E+03 9.94E+02 9.67E+02 3.06E+02 6.25E+00 1.0 1E-06 0.00E+00 0.00E+00 0.00E+00 0.001+00 0.OOE+00 Rh! 18 1.23E+03 1.23E+03 1.21E+03 1.18E+03 3.53E+02 6.24E+00 4.16E-08 0.OOE400 0.00E+00 O.00+E00 0.00E+00 O.00E+00 Pd1 18 3.801+03 3.80E+03 3.80E+03 3.79E+03 3.74E+03 3.61E+03 3.11 E+03 8.24E+02 3.87E+01 3.94E-01 3.13E-05 2.12E-21 Ag! 18 1.5 1E+0! 1.79E+00 0.00E+00 .003E+00 0.00E+00 0.00E400 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.00E100 0O.E+00 Ag! 18m 4.11 E+02 3.30E+02 0.00E+00 O.00E+00 0.002+00 0.00E+00 0.00+00 O 0.001+00 0.001+00 0.0013+00 0.00E+00 Cd1 18 6.89E+02 6.34E402 0.00E400 0.OOE+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 In! 18 4.89E+02 3.99E+02 0.001+00 0.002+00 0.002+00 0.002+00 0.OOE+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.00E+00 Rh119 1.21 E+03 1.21E403 8.02E+02 5.30E+02 1.63E+00 2.922-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 Pd! 19 1.211E303 1.21 E+03 8.04E+02 5.3 1E+02 1.63E+00 2.93E-06 0.002+00 0.002+00 0.00E+00 0.00E+00 0.002+00 0.001+00 Ag!19 4.45E+00 1.002+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 0.00+E00 0.002+00 0.001+00 0.00E+00 A-38 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements________ Decay time FollwingBurnup to 58 GWdIMTU ______Nuclide -ý=0 sec.- -30 min--- hr~ 8 hrý -~1.Od---

4.0,d -30.Od, 90.0 d , " 180.0Ad: ~ 3 y Cdl 19 1.70E+02 1.15E+02 0.00E+00 0.00+E00 0.OOE+00 0.003400 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.001+00 0.00E+00 Cd I19m 7.07E+02 5.47E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00 E+00 +00 0.OOE+00 0.001+00 0.00E+00 In 19 8.14E+02 8.13E+02 3.62E-01 1.59E-04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 In 1 19m 4.182+02 4.17E+02 3.31E-02. 2.59E-06 0.002+00 0.001400 0.00E+00 0.00E+00 0.0013+00 0.002+00 0.002+00 0.001+00 SnI 19m 5.47E+02 5.47E+02 8.63E+00 2.49E+00 2.35E-07 2.07E-23 0.00E+00 0.002+00 0.00E100 O.00E+00 0.OOE400 0.002+00 Pdl20 7.392+02 7.39E+02 2.74E+02 8.62E+01 8.16E-06 7.19E-22 0.00E+00 0.00E+00 0.002+00 0.00E+00 0.00E+00 0.002+00 Agl20 3.85E+03 3.85E+03 3.85E+03 3.85E+03 3.85E+03 3.84E+03 3.81E+03 3.58E+03 3.11E+03 2.52E+03 1.62E+03 2.88E+02 Cdl20 1.101+02 9.22E+01 0.002+00 0.OOE+00 0.00E+00 0.003+00 0.001+00 0.002+00 0.002E00 0.00E+00 O.00+E00 0.OOE+00 In 120 5.18E+02 3.31 E+02 0.00E+00 O.00E+00 0.002+00 0.00E+00 0.OOE+00 0.002+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 InI20m 1.171+03 1.16E+03 2.58E-08 5.52E-19 0.002+00 0.00E+00 0.002+00 0.00E+00 0.0023400 0.00E+00 0.00E+00 .003E+00 Pdl21 1.19E+03 1.19E403 2.75E-08 5.87E-19 0.00E+00 0.001+00 0.OOE+00 0.002+00 0.002+00 0.00E+00 0.OOE+00 0.002+00 Agl21 2.20E+01 2.17E+01 4.13E- I1 7.65E-23 0.OOE+00 0.00E+00 0.00E+00 0.OO00E+00 0.002+00 0.O0E+O0 0.00E+00 0.00E+00 Cdl21 4.52E+01 1.54E+01 0.00E+00 O0.OOE+00 0.002+00 0.001+00 0.000+00 0.002+00 0.OO1+00 0.00E+00 0.00E+00 O.0013+00 In 121 3.64E-+02 1.68E+02 0.00E3+00 0.OOE+00 0.OOE+00 0.002+00 0.002+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.00E+00 lnl21m 1.15E+03 1.1 IE+03 0.00E+00 O.OOE+00 0.00E+00 0.001+00 0.002+00 0.002+00 0.002+00 0.002+00 0.OOE+00 0.002+00 Sn121 1.162+02 1.132+02 7.85E-02 3.69E-04 O.00E+00 0.00E+00 0.002+00 .001E+00 0.002+00 0.00E+00 0.OOE+00 0.00E+00 Sn121m 1.18E+03 .1.18E+03 5.89E+00 2.77E-02 0.002+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00 0.00E+00 0.002+00 Pd122 2.31E+03 2.31E+03 2.28E+03 2.25E+03 1.89E+03 1.25E+03 1.992+02 8.962-01 8.932-01 8.91E-01 8.86E-01 8.632-01 Ag122 3.97E-01 3.97E-01 3.972-01 3.97E-01 3.97E-01 3.97E-01 3.97E-01 3.972-01 3.962-01 3.952-01 3.92E-01 3.831-01 Cd122 1.47E+01 8.992+00 0.002+00 0.002+00 0.00E+00 0.002+00 0.002+00 0.002+00. 0.002E00 0.00+E00 O.00E+00 0.002+00 In122 1.862+02 5.222+01 0.002+00 O.00E+00 0.002+00 0.002+00 0.002+00 0.002+00 0.0013400 0.002+00 0.002+00 0.0023400 lnl22m 1.232+03 1.092+03 0.002+00 O.00+E00 O.00E+00 0.002+00 0.002+00 0.002+00 0.00+E00 0.00E+00 0.00+E00 0.002+00 Sb122 1.38E+03 1.29E+03 0.002+00 0.002+00 0.00E+00 0.00E+00 0.002+00. 0.002+00 0.002+00 0.00E+00 0.00E+00 0.00E+00 Sb122m 1.492+02 1.40E+02 0.00+E00 O.00+E00 0.002+00 0.002+00 0.00E400 0.002+00 0.002+00 O.OOE+00 0.001+00 0.002+00 Te123m 5.062+02 5.062+02 5.02E+02 5.002+02 4.64E+02 3.91E+02 1.81E+02 2.28E-01 4.682-08 4.33E-18 0.00E+00 0.002+00 A-39 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_______ _______ ________ ,e~cay tm Following Burnup to58G dMU______________ Nuclide T--0 I. sec1 30min-- I hr. -.1 8 hr, 1.0d 4.d :30.0d d 90.0dd .180.0d .1 yr":", 3 yr Ag123 3.02E+01 3.02E+01 2.16E-01 1.55E-03 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Cd123 8.011E+00 8.01E400 8.00E+00 8.00E+00 8.00E+00 7.96E+00 7.82E+00 6.74E+00 4.76E+00 2.83E+00 9.65E-01 1.40E-02 In123 2.59E+00 2.56E-01 O.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+O0 0.003E+00 .00E+00 0n123m 7.58E+01 1.34E+01 0.00E+00. O.00E+00 0.00E+00 0.0013+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Sn123 8.031E+02 7.4513+02 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 SnI23m 1.0413+03 9.94E+02 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Ag124 2.84E+02 2.83E+02 1.51 E-09 6.92E-21 O.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE400 0.00E+00 0.OOE+00 0.00E+00 CdU24 1.97E+02 1.97E+02 1.97E+02 1.97E+02 1.96E402 1.95E+02 1.92E+02 1.67E+02 1.2 1E+02 7.48E+01 2.76E+01 5.49E-01 In124 1.34E+03 1.34E+03 8.03E+02 4.78E+02 3.351E-01 2.07E-08 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sb124 5.49E+01 3.90E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 .O.00E+00 Sb124m 1.171E+03 5.48E+02 O.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.003E400 0.003E400 Ag125 2.26E403 1.97E+03 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 O.00E00 0.003E+00 0.00E+00. 0.00E+00 0.00E+00.CdU25 2.39E+02 2.39E+02 2.39E402 2.39E+02 2.38E+02 2.36E+02 2.29E+02 1.69E+02 8.48E401 3.011E+OI 3.56E+00 7.92E-04 In125 4.11 E+00 4.08E+00 6.14E-06 9.111E-12 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 In125m 2.27E+01 2.86E+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sn125 6.8213+02 4.38E+02 0.OOE+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Sn125m. 1.27E+03 1.0413+03 O.00E+00 0.00E+00 0.0013+00 0.OOE+00 0.00E+00 0.001E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 Sb125 9.99E+02 9.53E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te125m 8.09E+02 8.09E+02 8.08E+02 8.07E+02 7.89E+02 7.52E+02 6.07E+02 9.36E+01 1.25E+00 1.94E-03 3.18E-09 0.0013+00 Ag126 3.08E+03 3.08E+03 3.49E+02 3.93E+01 2.06E-12 0.00E+00 O.00E+00 0.00E+00 0.0013+00 0.0013+00 0.00E+00 0.00E+00 Cd126 2.47E+03 2.47E+03 2.4713+03 2.47E+03 2.47E+03 2.47E403 2.47E403 2.43E+03 2.33E+03 2.19E+03 1.92E+03 1.161E+03 In126 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.65E+02 5.6513+02 5.65E402 5.65E+02 5.55E+02 5.30E+02 4.69E+02 2.83E+02 Sb126 1.131E+01 8.74E-02 O.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00. 0.0013+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 Sb126m 8.65E+02 2.20E+02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.0013+00 0.00E+00 0.00E+00 .001E+00 Ag127 3.13E+03 1.94E+03 0.0013+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.0013+00 A-40 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006.-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements______ --_--- Decay time Following-Burnup to 58 GWd/MTU -Nuclide- T = 0 sec -30min7 hr -T=-8 hr -- -1.0 d. 4.0 d -- 30.0 d 90.0 d:- 180.0 d :- 1 yr7ý_.- ,3 yr-Cd127 8.31E+O1 8.31E+01 8.31E+01 8.29E+01 8.16E+01 7.86E+01 6.65E+01 1.56E+01 5.67E-01 2.88E-02 2.53E-02 2.53E-02 In127 9.43E+01 9.42E+01 3.17E+01 1.07E+01 1.81E-01 1.81E-0I 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1.81E-01 1n127m 6.47E+00 1.23E-01 0.OOE+00 O.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Sn127 6.90E+02 2.05E+02 0.OOE+00 0.00+E00 0.00E+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.00E+00 0.OOE+00 Sn127m 2.82E+03 1.62E+03 O.OOE+00.002+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.002+00 O.OOE+00 0.00E+00 0.00+E00 Sb 127 2.82E+03 2.37E+03 0.OOE+00 O.OOE+00 0.00+E00 O.OOE+O0 O.OOE+00 0.OOE+00 0.00E+00 O.00+E00 0.00E+00 0.00+E00 Te127 5.62E+03 5.62E+03 4.77E+03 4.04E+03 4.01E+02 2.04E400 9.74E-11 0.00E+00 O.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 Te127m 7.52E+03 7.5 1E+03 4.93E+01 3.2 1E-0I 0.OOE+00 0.OOE+00 0.OE+00 O.OOE+00 0.00+E00 O.OOE+00 0.00+E00 0.OOE+00 Ag128 1.39E+04 1.39E+04 1.38E+04 1.38E+04 1.32E+04 1.17E+04 6.81E+03 6.31E+01 1.28E-03 1.18E-10 3.84E-25 O.00E+00 CdU28 1.38E+04 1.38E+04 1.38E+04 1.38E+04 1.36E+04 1.28E+04 8.54E+03 2.04E+03 1.35E+03 7.63E+02 2.35E+02 2.26E+00 In128 2.36E+03 2.36E+03 2.36E+03 2.36E403 2.36E+03 2.36E+03 2.34E+03 2.02E+03 1.38E+03 7.79E+02 2.40E+02 2.31E+00 Sn128 2.90E+00 1.83E-03 0.00E+00 O.00E+00 O.00+E00 0.00E+00 O.00E+00 0.00+E00 0.00E+00 O.00E+00 0.00+E00 0.00E+00 Sb128 5.82E402 3.01E+02 0.00+E00 O.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 O.OOE+00 0.00+E00 0.00E+00 Sb128m 4.42E+03 2.26E+03 0.00+E00 O.00E+00 0.00+E00 0.00E+00 O.00+E00 0.00+E00 0.00E+00 O.00E+00 0.00+E00 0.00E+00 Cdl29 2.08E+04 2.08E404 1.46E+04 1.03E+04 7.45E+01 9.60E-04 O.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 1n129 2.32E+03 2.32E403 2.26E+03 2.20E+03 1.31E+03 3.84E+02 1.51E+00 2.15E-21 0.OOE+00 0.00E+00 0.00E+00 O.00+E00 Sn129 2.22E+04 2.22E+04 1.73E+04 1.24E+04 9.04E+01 1.17E-03 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 Sn129m 2.94E+02 2.87E+01 0.06E+00 O.00+E00 0.00+E00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 Sb129 5.14E+03 1.65E+03 0.00E+00 O.00+E00 0.00E+00 0:00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 O.00E+00 Tel29 1.89E+04 1.88E+04 1.25E+00 1.80E-04 1.32E-23 O.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 Te129m 1.65E+04 1.65E+04 7.42E+02 3.33E+01 4.48E-18 0.00E+00 0.002+00 0.00E+00 0.OOE+00 0.00+E00 0.00+E00 0.00E+00 Xe129m 4.77E+04 4.77E+04 4.46E+04 4.12r+04 1.37E+04 1.10E+03 1.3 1E-02 0.00E+00 0.00E+00 O.00E+00 O.002+00 O.OOE+00 Cdl30 4.54E+04 4.54E+04 4.50E+04 4.39E+04 2.10E+04 7.03E+03 5.45E+03 3.19E+03 9.25E+02 1.45E+02 3.16E+00 9.02E-07 In130 9.19E+03 9.19E+03 9.19E403 9.19E+03 9.17E+03 9.05E+03 8.51E+03 4.98E+03 1.44E+03 2.25E+02 4.94E+00 1.41E-06 Sn130 2.23E+01 2.23E+01 2.22E+01 2.22E+01 2.17E+01 2.06E+01 1.63E+01 2.15E+00 2.00E-02 1.79E-05 9.55E.12 0.00E+00 A-41 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements_ý Decay time- Following Burnup to 58 GWd/MTU ____ _________ ____Nuclide" mT0 1sec 30min I hr.. 8r 1.0 d 4.0d 7 30.0d 90.0d 180.d lyr. 3yr.Sbl3O 1.05E+02 2.45E+01 0.00E+00 O.OOE+00 0.OOE+00 .003E+00 0.00E+00 0.OOE+00 0.OOE+0O 0.OOEO00 0.00E+00 0.OOE+00 Sbl30m 3.36E+03 4.20E+02 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.00E+00 1130 4.60E+04 4.59E+04 1.72E+02 6.42E-01 0.00E+00 0.OOE+00 0.00E+00 O.OOE+00 O.OOE÷00 0.00E+00 0.00E+00 0.00E+00 Wr30m 1.58E+04 1.58E+04 9.36E+03 5.53E+03 3.48E+00 1.68E-07 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Cdl31 6.18E+04 6.18E+04 4.48E+03 1.73E+02 1.49E-18 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 O.OOE+00 In131 6.42E+03 6.42E+03 6.27E+03 6.1OE+03 4.12E+03 1.68E+03 2.97E+01 1.89E-14 O.OOE+00 0.OOE00O .0.O1+00 .OOEE+00 Sn131 3.45E+03 3.44E+03 3.42E+02 3.39E+01 3.04E-13 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+0O 0.OOE+00 O.OOE+00 0.00E+00 Sbl31 1.69E+01 2.43E-02 0.00E+00 O.OOE+00 ..OOE+00 0.OOE+00 0.0E3+00 O.OOE+00 0.0E3+00 0.O13400 0.001E+00 0.OOE+00 TeO3M 1.56E+03 1.20E+02 O.00E+00 O0.OOE+00 0.0E3+00 O0 0.00E00 0.0OE+00 O.O0E+00 0.OOE+00 O.OOE+00 6.00E+00 Tel31m 4.04E+04 3.97E+04 5.19E-10 6.56E-24 0.00E+00 0.00E+00 0.OOE400 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 1131 1.08E+05 1.08E+05 4.42E+04 1.79E+04 5.70E-02 1.55E-14 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 Xe131m 1.18E+05 1.18E+05 9.07E+04 5.76E+04 5.70E+03 3.94E+03 7.46E+02 4.09E-04 1.45E-18 O.OOE+00 O.00E+00 0.OOE+00!n132 3.04E+04 3.04E404 3.01E+04 2.98E+04 2.53E+04 1.75E+04 3.32E+03 1.82E-03 6.45E-18 0.OOE+O0 O.E+O0 O.O0E+O0 Sn132 1.41E+05 1.41E+05 1.41E+05 1.41E+05 1.38E+05 1.32E405 1.04E405 I.11E+04 6.28E+01 2.68E-02 3.11E-09 O.OOE+00 Sb132 2.15E+03 2.15E+03 2.15E+03 2.15E+03 2.13E+03 2.1 1E+03 1.97E+03 7.02E+02 2.75E401 1.52E-01 3.14E-06 1.03E-24 Sb132m 4.10E+02 9.86E+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 O.OOE400 Te132 3.32E+04 3.26E404 9.48E-I0 2.67E-23 0.00E+O0 0.00E+00 0.00E+00 0.00E100 0.00E0 00 0.03 0.0E+00 0.OOE+00 1132 6.78E+04 6.76E+04 4.80E+02 3.39E+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Cs132 6.13E+04 6.12E+04 4.27E+01 2.54E-02 0.00E400 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.OOE+00 1n133 2.01E+05 2.01 E+05 2.01E+05 2.OOE+05 1.88E+05 1.6313+05 8.61E+04 3.41E+02. 9.75E-04 4.71E-12 3.61E-29 0.00E+00 Sn133 2.06E+05 2.06E+05 2.05E+05 2.05E+05 1.93E+05 1.68E+05 8.87E+04 3.51E+02 1.00E-03 4.86E-12 3.72E-29 .003E+00 Sb133 9.34E+00 9.34E+00 9.32E+00 9.30E+00 9.011E+00 8.39E+00 6.09E+00 3.77E-01 6.15E-04 4.05E-08 1.0013-16 0.00E+00 Te133 1.45E+01 2.86E-02 0.00E+00 O.00E+00 0.00E+00 0.0013+00 .001E+00. 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te133m 9.02E+03 5.57E+03 0.OOE+00 0.OOE+00 0.OOE+00 0.0013+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 1133 8.70E+04 8.66E+04 2.13E+01 5.18E-03 0.OOE+00 0.0011+00 0.OOE400 0.OOE+00 0.00E+O0 0.00E+00 0.001E+00 0.0011+00 A-42 Safety Analysis Report for Susquehanna Units 1 and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to-58 GWd/MTU .- ..-Nuclide T=O .1 see 30minm 1 hr 8hr.' 1.0.d. .r" 4.0d ,Ji. 730.0di.,i.' _90.0d.. .d 180.0d.I 1yr , 3yr 1133m 1.52E+05 1.52E+05 4.62E+04 1.84E+04 6.96E+01 4.23E-04 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 Xe133 1.24E+05 1.24E+05 8.59E+04 5.90E+04 3.08E+02 1.87E-03 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 Xe133m 2.85E+05 2.85E+05 2.83E+05 2.80E+05 2.24E+05 1.32E+05 1.19E+04 1.1 IE-05 1.60E-26 0.00E+00 0.OOE+00 0.00E+00 Sn134 2.30E+04 2.22E+04 8.78E+03 6.0313+03 3.15E+01 1.91 E-04 O.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Sb 134 2.74E+05 2.74E+05 2.74E+05 2.74E+05 2.73E+05 2.65E+05 1.95E+05 6.43E+03 2.3 1E+00 1.57E-05 3.63E-16 O.OOE+00* Sb134m 9.22E+03 9.22E+03 9.21E+03 9.20E+03 9.03E+03 8.25E+03 3.93E+03 1. 11E+00 6.25E-09 2.65E-21 0.OOE+00 0.00E+00 Te134 1.62E+03 8.2911+02 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 1134 1.55E+04 7.35E+03 0.00E400 O.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 1134m 1.25E+04 1.17E+04 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 .001E+00 0.00E+00 0.00E+00 0.OOE+00 O.00E+00 Xe134m 2.39E+05 2.39E+05 1.46E+05 8.85E+04 8.36E+01 1.02E-05 O.OOE+0 0 .OOE+00 0.OOE+00 0.001+00 0.00+00 .0.00E+00 Cs134 3.12E+05 3.12E+05 2.72E+05 2.20E+05 1.9013+03 7.08E-03 0.00E+00 0.001E00 0.0013400 0.OOE+00 0.OOE+00 0.OOE+00 Cs134m 3.1 IE+04 3.10E+04 1. 11E+02 3.96E-01 0.001E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.0013+00 0.00E+00 0.OOE+00 Sn135 9.61E+03 1.53E+03 2.56E400 9.13E-03 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.00E+00 0.00E+00 Sb135 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.58E+04 5.56E+04 5.43E+04 5.14E+04 4.j3E+04 3.99E+04 2.04E+04 Tel35 9.911E+03 9.911E+03 8.80E+03 7.81E+03 1.47E+03 3.27E+01 1.177E-06 0.OOE+00 0.OOE+00 0.00E+00 0.004E00 0.OOE+00 1135 1.39E+02 2.64E+01 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 O.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Xe135 7.90E+03 5.28E+03 O.OOE+00 0.OOE+00 0.OOE+00 0.000400 0.001+00 0.OOE+00 0.000+00 0.001+00 O.OOE+0 0 0.OOE+00 Xe135m 1.37E+05 1.32E+05 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 .001E+00 O.00E+00 O.OOE+O0 0.00E+00 O.00E+00 Cs135m 2.74E+05 2.74E+05 2.6013+05 2.46E405 1.18E+05 2.18E404 1.09E+01 0.00E+00 0.0013400 0.00E+00 0.00E+00 0.00E+00 Ba135m 7.50E+04 7.50E+04 8.26E+04 8.90E+04 1.22E+05 7.02E404 5.08E+02 1.50E-18 0.00E+00 0.00E+00 0.00E+00 0.00E3400 SnI36 6.25E+04 6.25E+04 4.70E+04 4.14E+04 1.92E+04 3.55E+03 1.78E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Sb136 1.3813+04 1.38E+04 9.3313+03 6.30E+03 2.60E+01 9.15E-05 O.00E+00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 Te136 1.82E+02 1.82E+02 1.80E+02 1.78E+02 1.50E+02 1.02E+02 1.79E+01 -5.10E-06 3.98E-21 0.00E+00 0.00E+00 0.00E+00 1136" 1.21 E+01 4.60E+00 O.OOE+00 O.0014 00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1136m 1.28E+03 5.511E+02 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.001E00 0.00E+00 A-43 Safety Analysis Report for Susquehanna Units 1 and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements........... .......-,.. Decay time FollowingBurnup to 58 GWd/MTU. _.__.- _______ _____.___ _________Nuclide T=0 1 see,,. 30mmn .1 hr ' ,8hr 1.0d 4.0dd .30.0-d 90.Od 180.0d .lyr: : 3yr.Cs136 5.96E+04 5.73E+04 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 O.00E+00 Ba136m 1.21E+05 1.20E+05 4.36E-02 1.38E-08 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 Sb137 5.99E+04 5.90E+04 1.68E-07 4.67E-19 0.00E+00 0.OOE+00 0.00E400 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 Te137 1.54E+04 1.54E+04 1.54E+04. 1.53E+04 1.51E+04 1.46E+04 1.25E+04 3.17E+03 1.34E+02 1.17E+00 6.79E-05 1.32E-21 1137 1.77E+03 1.73E+03 1.72E+03 1.72E+03 1.69E+03 1.63E+03 1.40E+03 3.55E+02 1.50E+01 1.31E-01 7.61E-06 1.48E-22 Xe137 7.20E+02 1.68E402 0.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Cs137 2.07E+04 1.70E+04 0.00E-00 0.00E+00 0.00E400 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 I0.00E+00 0.02+00 0.OOE+00 Ba137m 1.32E+05 1.29E+05 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 Sb138 2.61E+05 2.60E+05 1.19E+03 5.13E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.OOE+00 0.OOE+00 Te138 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.30E+04 3.28E+04 3.27E+04 3.23E+04 3.08E+04 1138 3.14E+04 3.14E+04 3.12E+04 3.12E+04 3.12E+04 3.12E+04 3.12E+04 3.11E+04 3.10E+04 3.08E+04 3.05E+04 2.91E+04 Xe138 1.83E+01 3.33E-01 0.00E+00 O0.OOE+00 0..0E+00 0.000+00 0.000+00 0..0E+00 0.000+00 0.00+00 0.00E+00 0.OOE+00 Cs138 5.41E+03 3.30E+03 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 Cs138m 6.81E+04 6.16E+04 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 O.OOE+00 Tel39 2.36E+05 2.35E+05 5.39E+04 1.23E+04 1.29E-05 3.85E-26 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 1139 2.58E+05 2.58E+05 1.90E+05 1.12E+05 1.44E+01 1.53E-08 0.00E+00 0.OOE+00 0.002+00 0.OOE+00 0.00E+00 0.00+E00 XeM39 1.24E+04 1.23E+04 9.77E+00 7.69E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs139 8.49E+02 2.56E+02 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.60E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 Ba139 2.92E+04 2.17E+04 O.00E+00 O.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 TeI40 1.69E+05 1.66E+05 3.79E-09 8.28E-23 0.002E00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 1140 2.38E+05 2.38E+05 2.67E+04 2.83E+03 6.51-E-1 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 Xe140 2.46E+05 2.46E+05 2.13E+05 1.69E+05 5.43E+03 2.09E+00 9.00E-16 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 Cs 140 1.18E+02 5.4 1E+01 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 O.OOE+00 0.00E+00 Ba140 8.51E+03 3.83E+03 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 Lal40 1.15E+05 1.10E+05 0.002+00 O.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 A-44 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements--Decay time Following Burnupto 58 GWdIMTU.________ _____ _________Nuclide T 0: i 30rmin 1hr f .ý8hr L, -:1.0 d 4.0d 30.0d 90.0d 180.0dd I yr. 3yr Pr140 2.13E+05 2.12E+05 7.64E-04 2.37E-12 O.OOE+00 0.OOE+00 0.OOE+00 O.O0E+00 O.OOE+00 O.OOE+00 O.00E+00 O.OOE+00 1141 2.47E+05 2.47E+05 2.46E+05 2.46E+05 2.42E+05 2.34E+05 1.99E+05 4.83E+04 1.85E+03 1.39E+01 5.91E-04 3.39E-21 Xel41 2.70E+05 2.70E+05 2.70E+05 2.70E+05 2.67E+05 2.60E+05 2.26E+05 5.57E+04 2.13E+03 1.60E+01 6.81E-04 3.91E-21 Csl41 7.32E+00 7.29E+00 1.59E-02 3.44E-05 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 Ba141 1. 1 9E+03 2.64E+02 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 La141 4.56E+04 3.06E+04 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.00E400 Ce141 1.6 1E+05 1.58E+05 3.12E-17 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+0 0.OOE+00 0.00E+00 1142 2.21E+05 2.21E+05 7.21E+04 2.31E+04 2.77E-03 4.18E-19 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 Xe142 2.24E+05 2.24E+05 2.16E+05 2.01E+05 5.90E+04 3.48E+03 1.03E-02 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Cs142 2.25E+05 2.25E+05 2.25E405 2.25E+05 2.24E+05 2.21E+05 2.08E+05 1.19E+05 3.31E+04 4.86E+03 9.35E+01 1.60E-05 Ba142 3.85E+02 1.22E+01 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 La142 1.86E+04 1.05E+04 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE÷00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 Pr142 9.28E+04 6.64E+04 O.OOE+00 O.OOE+00 0.OOE+00 O.00E400 O.OOE+00 O.OOE+00 O.0oE+00 O.00E+O0 O.OOE+00 0.OOE+00 1143 2.08E+05 2.08E+05 2.93E+04 4.11E+03 4.86E-09 O.OOE+00 O.OOE+00 O.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Xe143 2.17E+05 2.17E+05 1.90E+05 1.54E+05 6.33E+03 4.25E400 2.26E-14 O.OOE+00 O.OOE+00 O.00E+00 O.00E+00 O.OOE+00 Cs 143 1.72E+04 1.72E+04 1.69E+04 1.66E+04 1.29E+04 7.20E+03 5.29E+02 7.93E-08 O.OOE+00 0.OOE+00 O.00E+00 O.00E+00 Ba143 3.87E+00 6.85E-01 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 La143 2.95E+03 1.43E+03 O.00E+00 O.OOE+00 0.OOE+00 O.OOE+00 0.00E+00 0.OOE+00 O.00E400 0.OOE+00 0.00E+00 O.00E+00 Ce143 4.52E+04 3.13E+04 0.OOE+00 O.OOE+00 0.OOE+00 0OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 O.OOE+00 O.00E+00 Pr143 1.74E+05 1.67E+05 O.00E400 O.OOE+00 0.OOE+00 O.00E+00 O..0E+00 .OOE+00 O.OOE+00 O.00E+00 0.00E+00 O.OOE+00 Xe144 2.01E+05 2.01E+05 4.70E+04 1.08E+04 1.24E-05 4.5 1E-26 O.00E+00 0.00E+00 O.OE+00 O.00E+00 0.00E4-0 O.OOE+00 Cs144 2.04E+05 2.04E+05 2.03E+05 .2.01E+05 1.73E+05 1.24E+05 2.73E+04 5.55E-02 4.06E-15 O.OOE+00 0.OOE+00 0.OOE+00 Ba144 1.97E+05 1.97E+05 1.97E+05 1.97E+05 .1.96E+05 1.95E+05 1.76E+05 4.74E404 2.21E-E÷03 2.22E+01 1.72E-03 1.06E-19 La144 6.53E+02 3.47E+02 O.OOE+00 O.00+E00 O.00E+00 O.OOE+00 O.OOE+00 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 Ce144 1.42E+04 7.40E+03 O.OOE+00 O.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 A-45 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements______ ______ Decay timeFollowing Burnuptio' 58 GWdNITU: 2-Nuclide- T= I secT i.30 min,.-.. I hr 8 hr d 4.0.d 30.0od 90.0di& 180.0 d' -1 yr .3yr Pr144 1.3 1E+05 1.24E+05 0.00E+00 O.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0,00E+00 0.OOE+00 0.00E400 Pr144m 1.76E+05 1.75E+05 1.29E-08 7.19E-22 0.OOE+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 Xe145 1.89E+05 1.89E+05 1.89E+05 1.89E+05 1.89E+05 1.88E+05 1.87E405 1.76E+05 1.52E+05 1.22E+05 7.77E404 1.31E+04 Csl45 1.90E+05 1.90E+05 1.89E+05 1.89E+05 1.89E+05 1.88E+05 1.87E+05 1.76E+05 1.52E+05 1.22E+05 7.77E+04 1.31E+04 Ba145 2.65E+03 2.65E+03 2.64E+03 2.64E+03 2.64E+03 2.64E+03 2.62E+03 2.46E+03 2.12E+03 1.71E+03 1.09E+03 1.84E+02 La145 6.98E+01 3.23E+01 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 Ce145 3.74E+03 1.19E+03 0.60E+00 O.00E+00 0.00+E00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Pr145 6.02E404 5.15E+04 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 Xe146 1.24E+05 1.22E+05 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 Csl46 1.40E+05 1.40E+05 1.61E+02 1.61E-01 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Ba146 1.40E+05 1.40E+05 1.33E+05 1.26E+05 5.59E+04 8.75E+03 2.09E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 La146 6.05E+00 1.76E+00 0.00+E00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.OOE+00 O.00E+00 0.00E+00 Ce146 7.26E+02 9.81E+01 0.OOE+00 O.00+E00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 0.00+E00 0.00E400 0.00E+00 0.00E+00 Pr146 2.98E+04 2.18E+04 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Cs147 7.97E+04 7.40E+04 0.00E+00 O.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 O.00E+00 0.00+E00 Ba147 1.13E+05 1.13E+05 2.45E+04 5.26E+03 2.34E-06 9.87E-28 0.00E+00 0.00E+00 0.00E+00 0,00E+00 O.00+E00 0.00E+00 La147 1.14E+05 1.14E+05 7.85E+04 3.96E+04 2.70E-01 2.91E-13 0.OOE+00 0.004E00 0.00E+00 0.OOE+00 O.OOE+00 0.OOE+00 Ce147 1.91E+OI 5.35E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 Pr147 5.88E403 2.19E+03 0.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 Nd147 3.67E404 3.19E+04 0.00+E00 0.OOE+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 Pm147 8.662+04 8.59E+04 2.22E-05 5.45E-15 0.00+E00 0.OOE+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Cs148 9.15E+04 9.15E+04 2.13E+04 4.61E+03 2.33E-06 1.31E-27 0.00+E00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 Ba148 9.29E+04 9.29E+04 9.28E+04 9.27E+04 9.10E+04 8.73E+04 7.22E204 1.40E+04 3.17E+02 1.08E+00 9.01E-06 8.47E-26 La148 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.15E+04 3.16E+04 3.16E+04 3.05E+04 2.85E+04 2.50E+04 1.47E+04 Ce48 3.51E+00 1.19E-01 0.00E+00 0.OOE+00 0.00E+00 0.004E00 0.OOE400 0.00r+00 0.00E+00 0.OOE400 0.00+E00 0.00E+00 A-46 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel.Sum of Actinides, Fission Products and Light Elements Decay time FollowingBurnup to 58 -GWdIMTU_____ Nuclide: ;T- 0 sec- :30 min 1 hr 8 hr :1.0-d .4.0d .30.0-diI_` 90.0d= 1S0.0d 1yr.: 3yr.Pr148 1.21 E+03 3.85E+02 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E400 0.002E00 0.00E+00 0.00E+00 0.004E00 Pm148 1.20E+04 6.54E+03 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 0.OOE+00 0.00+E00 Pml48m 6.18E+04 6.11 E+04 1.31 E-05 2.75E-15 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 Ba149 7.23E+04 7.23E+04 1.22E401 1.28E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 La149 2.88E+04 2.88E+04 2.87E+04 2.87E+04 2.76E+04 2.54E+04 1.73E+04 7.46E+02 5.61E+01 1.23E+01 5.50E-01 2.60E-06 Ce149 4.78E+03 4.78E+03 4.78E+03 4.78E+03 4.75E+03 4.70E+03 4.47E+03 2.89E+03 1.05E+03 2.33E+02 1.04E+01 4.92E-05 Pr149 1.37E+02 5.04E+01 0.00+E00 O.OOE+00 0.00E+00 0.00E400 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Nd149 3.61E+03 2.73E+03 0.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm149 3.27E+04 2.90E+04 0.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 0.00E+00 0.OOE4-00 BaI50 5.15E404 5.14E404 5.34E+00 5.38E-04 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 La150 5.70E+04 5.70E+04 4.76E+04 3.90E+04 2.34E+03 3.77E+00 1.03E-12 0.00E+00 0.OOE+00 O.OOE+00 O.00E+00 0.OOE+00 CeI50 9.22E+04 9.22E+04 9.19E+04 9.16E+04 8.47E+04 6.88E+04 2.69E+04 7.77E+00 5.30E-08 2.99E-20 O.OOE+00 0.OOE+00 PrI50 1.25E+01 6.05E+00 0.00E+00 O.OOE+00 O.00E400 0.00E+00 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 Pm150 6.50E+02 2.13E+02 0.00+E00 O.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 O.00E+00 Lal51 1.59E+04 1.34E+04 0.00E+00 O.00-+00 0.OOE+00 0.004E00 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 Ce151 3.38E+04 3.18E+04 0.00+E00 O.OOE+00 0.000+00 0..0E+00 0.002+00 0.00E+00 0.OOE+00 0.00E+00 O.00E+00 0.00E+00 Pr1 51 1.00E+03 1.00E+03 8.81 E+02 7.74E+02 1.27E+02 2.02E+00 1.65E-08 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 O.OOE+00 Nd151 1.11 E+02 4.23E+01 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 O.00E+00 Pm 151 4.82E+03 2.47E+03 0.00E+00 O.00E+00 0.OOE+00 0.00+E00 0.00E+00 0.00+E00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 Sml51 1.84E+04 1.78E+04 0.00E+00 O.00E+00 0.OOE+00 0.00E400 0.002E00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Ce152 3.12E+04 3.12E+04 5.96E+03 1.12E+03 7.69E-08 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00+E00 Pr152 3.15E+04 3.15E+04 3.13E+04 3.10E+04 2.61E+04 1.77E+04 3.05E+03 7.37E-04 3.96E-19 0.00+E00 0.00E+00 0.00E+00 Ndl52 9.85E+01 9.85E+01 9.85E+01 9.86E+01 9.87E+01 9.90E+01 9.95E+01 9.96E+01 9.95E+01 9.93E+01 9.89E+01 9.74E+01 Pm152 6.58E+02 6.0 1E+02 0.00E+00 O.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00E+00 Pm152m 6.4E+03 5.88E+03 0.00E+00 O.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 0.00E+00 0.00+E00 A-47 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006.-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements______ Decay time. Following Burnup to 58 GWdIMTU.Nuclide T=Oj 0 I sec:--- 30 min 1:hr. 8 hr:. 1.0d 4.0 d -3 0.0 d 90.0 d 180.0 d 1 yr., "3yr Eu152m 2.11 E+04 2.11 E+04 3.42E+03 5.5 1E+02 4.47E-09 0.OOE400 0.OOE+00 0.OOE+00 O.0OE400 0.OOE+00 0.OOE+00 O.OOE+00 Ce153 2.19E+04 2.19E+04 5.26E+03 8.60E+02 6.99E-09 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Pr153' 8.20E+02 8.19E+02 5.17E+01 3.25E+00 5.OOE-17 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 Nd153 3.63E+01 3.63E+01 3.50E+01 3.37E+01 2.OOE+OI 6.10E+00 2.88E-02 2.01E-22 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Pm153 2.29E+02 1.43E+02 0.00E+00 O.OOE+00 0.OOE+00 0.00E400 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Sm153 2.78E+03 2.4 1E+03 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 Gd153 1.23E+04 1.22E+04 I.A8E-04 I.10E-12 0.OOE+00 0.OOE-f00 O.00E+00 0.OOE-00 O.OOE+00 0.00E+00 O.OOE+00 O.00E+00 Ce1 54 1.49E+04 1.49E+04 3.86E+02 8.20E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 Pr154 1.04E+05 1.04E+05 1.03E+05 1.02E+05 9.22E+04 7.26E+04 2.47E+04 2.15E+00 9.25E-10 8.24E-24 0.00E+00 0.OOE+00 Nd 154 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.05E+03 1.04E+03 9.67E+02 8.13E+02 6.28E+02 3.69E+02 4.54E+01 Pm154. 2.37E+01 1.68E+01 0.00E+00 O.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Pm154m 5.78E+02 3.10E+02 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 Eu154 6.56E403 6.45E+03 1.88E-10 5.28E-24 0.00E+O0 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+O0 Gd155m 8.20E+03 8.19E+03 6.95E-02 3.89E-07 0.00E+0O 0.OOE+0O 0.OOE+00 O.00E+00 0.00E400 O.OOE+00 0.00E+00 0.OOE+00 Pr155 1.64E+03 1.64E+03 7.02E-01 2.99E-04 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.00E+00 O.00E+00 Nd155 2.41E+03 2.41E+03 2.41E+03 2.41E+03 2.41 E+03 2.41E+03 2.41E+03 2.39E+03 2.36E+03 2.32E+03 2.22E+03 1.89E+03 Pm155 2.28E401 4.22E-09 0.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 O.00E+00 O.00E+00 0.00E400 0.OOE+00 O.00E+00 O.OOE+00 Sm155 1.27E+02 6.89E+01 0.00E+00 O.OOE+00 0.00E-fOO0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 Eu155 2.39E+03 2.30E+03 0.00E+00 0-00E+00 0..OOE+00 O.00E+00 O.OOE+00 0.OOE+00 O.00E+00 O.OOE+00 O.00E+00 0.00E+00 PrI56 5.3 1E+03 5.27E+03 3.50E-08 1.79E-19 O.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 O.00E+00 Nd156 6.80E+03 6.80E+03 2.77E+03 1.09E+03 2.33E-03 2.56E-16 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00 0.00ES00 PmI56 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.01E+03 1.00E+03 9.78E+02 9.42E+02 8.74E+02 6.50E+02 Sm156 1.91E+01 3.11E+00 O.00E+00 0..0E+00 O.OOE+00 0.0EE+00 0..0E+00 .OOE+00 0..0E+00 0..0E+00 0.00E+OO 0.OOE+00 Eu156 9.18E+02 8.86E+02 0.OE+0 b00E+00 0..0E+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 Nd157 2.85E÷03 2.75E+03 0.OOE+00 -O00E+00 0.00E+00 0.OOE+00 0.00E+00 0.OOE-400 0.00E+00 0.00E+00 0.OOE+00 0.00E- 00 A-48 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements......:r .:..., Decay time Following Burnup to 58GWd/MTUý. -_ _. ._.. ._-_.Nuclide T- O .- Alsec-: :.I, ý30mi-h.. 1 hr hr81 1.0 d 4.Od i30.0 d 90.0 d 180.0 d -lyr 3 yr.Pm157 4.23E+03 4.23E+03 4.07E+03 3.93E+03 2.34E+03 7.20E+02 3.56E+00 3.70E-20 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Sm157 7.32E+04 7.32E+04 7.31E+04 7.31E+04 7.22E+04 7.OOE÷04 6.1 IE+04 1.87E+04 1.21E+03 1.98E+01 4.21E-03 1.39E-17 Eu157 2.39E402 1.81E+02 0.OOE+00 O.00E400 0.OOE400 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Nd158 1.36E+03 1.35E+03 1.9 1E-06 2.64E-15 O.OOE+00 0.OOE÷00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE÷00 0.00E+00 Pm158 2.73E+03 2.73E+03 2.23E+02 1.69E+OI 3.65E-15 0.00E400 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Sm158 5.95E+03 5.95E+03 5.84E+03 5.71E403 4.15E+03 2.OOE+03 7.46E+01 3.15E-11 0.06E+00 0.OOE+00 0.OOE+00 0.OOE+00 Eu158 2.78E+00 4.48E-01 0.OOE400 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Nd159 3.95E+01 3.05E+O1 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE4 00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Pm 159 3.84E+02 3.26E+02 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Sm159 1.46E+03 1.45E+03 3.36E+01 7.70E-01 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 Eu159 1.62E+03 1.62E+03 1.15E+03 7.37E+02 1.30E+00 6.56E-07 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 O.OOE+00 Gd159 3.44E+00 1.17E+00 0.OOE+00 O.0OE+00 0.OOE÷00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E400 O.OOE+00 0.OOE+00 0.OOE+00 Pml60 9.18E+01 7.32E+01 0.00E+06 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Sm160 6.22E402 6.20E+02 2.82E-01 1.27E-04 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Eu160 8.36E+02 8.36E+02 3.OOE+02 9.50E+01 9.83E-06 1.06E-21 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 Th160 4.36E+04 4.36E+04 4.28E-04 4.21E+04 3.24E+04 1.78E+04 1.21E+03 9.18E-08 0.OOE+00 O.OOE+00 0.OOE400 0.OOE+00 Sml61 1.02E+01 4.01E+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE400 O.OOE+00 0.OOE+00 Eul61 1.88E+02 1.96E+02 6.45E-06 2.19E-13 0.OOE+00 0.OOE+00 0.OOEs00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Gd161 3.36E+02 3.33E+02 1.64E-05 5.57E-13 0.OOE+00 0.OOE-00 0.OOE+00 0.OOE+00 0.OOE÷00 0.OOE+00 O.OOE+00 0.OOE+00 Th161 1.64E+04 1.64E+04 1.64E+04 1.64E+04 1.63E+04 1.62E+04 1.58E+04 1.23E+04 6.91E+03 2.92E+03 4.95E+02 4.50E-01 Sm162 4.18E01 3.63E+01 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 O.00E-00 0.00E+00 0.OOE+00 0.OOE+00 Eu162 1.3 1E+02 1.29E+02 1.78E-1 I 2.29E-24 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.00E400 0.00E+00 O.OOE+00 O.OOE+00 Gd162 5.06E+03 5.04E+03 1.73E+01 5.90E-02 O.OOE+0O 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Th162 8.08E+03 8.08E+03 8.06E+03 8.05E+03 7.81E+03 7.30E-403 5.41E+03 3.97E+02 9.57E-01 1.13E-04 9AIE-13 0.OOE+00 EUI63 5.74E+00 5.04E+00 0.OOE+00 O.OOE+00 O.OOE+00 0.00E400 0.OOE+00 0.OOE÷00 0.00E+00 O.OOE+00 0.OOE+00 O.OOE+00 A-49 Safety Analysis Report for Susquehanna Units 1 and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements______ Decay. timeFollowing Bu rnup to 58 GWdIMTU:;lse'- 30mm ................ 4.0d---- 30.0d"- '90.0d --- 180.0d t. 3yr 3yr Gd163 3.19E+01 3.18E+01 1.48E-02 6.81E-06 0.OOE+00 0.002+00 0.OOE+00 0.000+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 Tb163 7.14E+01 7.13E+01 7.29E+00 6.13E-01 5.45E-16 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Gd164 7.25E+01 7.25E+01 2.27E+01 3.04E+00 6.89E-15 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 Tb164 6.44E+00 5.91E+00 0.00E+00 O.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 Tb165 2.48E+01 2.47E+01 3.67E-05 5.28E-I 1 6.00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Dy165 2.82E+01 2.82E+01 1.05E+01 3.61E+00 1.18E-06 1.79E-21 0.00E+00 0.00E+00 0.00E+00 0.00E+00 O.OOE+00 O.00E-*00 Dy165m 7.37E+00 7.36E+00 2.82E+00 1.08E+00 1.60E-06 7.51E-20 0.00E400 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+60 Dy166 1.01E+01 1.OIE+01 3.28E+00 1.26E+00 1.85E-06 8.72E-20 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00!!o166 3.40E+00 3.39E+00 2.21 E-04 1.16E-08 0.00E+00 0.00+E00 O.OOE+00 0.00E+00 0.00+E00 0.OOE+00 0.00E+00 0.OOE+00 Er167m 3.06E+03 3.06E+03 2.65E+03 2.29E+03 2.86E+02 2.47E+00 1.28E-09 0.OOE+00 O.00E+00 0.OOE+00 0.OOE+00 O.00E+00 i1t175 1.96E+03 1.94E+03 6.OOE-04 2.46E-08 0.00+E00 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 0.00+E00 0.OOE+00 LuI76m 1.45E+01 1.45E+01 1.45E+01 1.44E+01 1.36E+01 1.19E+01 6.43E+00 3.21E-02 1.57E-07 1.69E-15 6.85E-32 0.OOE+00 Lu177 7.57E402 7.57E+02 7.48E+02 7.38E+02 6.18E+02 4.13E+02 7.14E+01 5.01E-02 2.44E-07 2.63E-15 9.13E-32 0.00E+00 IM178m 1.81E+01 1.34E+01 0.00+E00 O.00+E00 0.00E+00 0.00E+00 0.00+E00 0.00E+00 0.00E+00 O.00+E00 0.00E+00 0.OOE+00 If179m 3.18E-01 3.18E-01 3.18E-01 3.17E-01 3.10E-0I 2.96E-01 2.37E-01 3.48E-02 4.17E-04 5.48E-07 6.40E-13 0.00E+00 III8gOm 5.36E-02 5.36E-02 5.36E-02 5.36E-02 5.35E-02 5.34E-02 5.25E-02 4.56E-02 3.30E-02 2.03E-02 7.49E-03 1.46E-04 ltfl81 2.86E+00 2.86E+00 2.86E+00 2.86E+00 2.85E+00 2.83E+00 2.75E+00 2.13E+00 1.17E+00 4.82E-01 7.70E-02 5.57E-05 W181 2.60E+00 2.60E+00 2.36E+00 2.14E+00 5.65E-01 2.67E-02 " 2.93E-08 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00+E00 Ta182 1.12E+00 1.12E+00 1.12E+00 1.11 E+00 1.08E+00 1.01E+00 7.40E-01 5.10E-02 6.23E-04 3.53E-04 1.59E-04 6.83E-06 Ta182m 3.02E-01 2.54E-01 0.00+E00 O.00E+00 0.00E+00 0.00E+00 0.00+E00 0.00+E00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 Ta 183 1.87E+02 1.81E+02 0.00E+00 0.00+E00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 0.00E+00 0.00+E00 0.00+E00 WI83m 2.30E-02 2.30E-02 2.22E-02 2.13E-02 1.25E-02 3.45E-03 8.16E-06 0.00+E00 0.OOE+00 0.00+E00 0.00E+00 0.00E+00 WI85 1.19E+01 1.19E+01 1.12E+01 1.05E+01 4.35E+00 5.80E-01 6.64E-05 0.00+E00 0.00E+00 0.002E00 0.00+E00 0.00E+00 W185m 4.48E+02 4.48E+02 4.48E402 4.47E+02 4.45E402 4.40E+02 4.19E402 2.74E+02 1.03E+02 2.36E+01 1.14E+00 7.39E-06 Re186 1.87E+00 1.87E+00 1.87E+00 1.87E+00 1.872400 1.86E+00 1.83E+00 1.58E+00 1.12E+00 6.68E-01 2.32E-01 3.55E-03 A-50 Safety Analysis Report for Susquehanna Units I and 2 -Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel.Sum of Actinides, Fission Products and Light Elements Decay time Following Burnup to 58 GWdIMTU _____:,_Nuclide- -T-=0-- -1 sec--__ -30min-- hr-- 8hr. 1.0d -4.0dd .30.0d 90.0 d,- 180.0dd 1 yr. 3yr W187 5.61E÷01 5.61E+O1 5.61E+O I 5.61E+01 5.60E+01 5.58E+01 5.48E+01 4.68E+01 3.26E+01 1.90E+01 6.21EE+00 7.60E-02 W188 1.06E-01 1.06E-01 2.86E-02 7.70E-03 7.97E-1 I 4.46E-29 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOEO00 O.OOE+00 ReI88 1.54E+02 1.54E+02 1.53E+02 1.53E+02 1.47E+02 1.34E+02 8.94E+01 2.62E+00 7.57E-04 3.72E-09 4.42E-20 O.OOE+00 Re188m 1.69E+02 1.67E+02 1.53E+02 1.53E+02 1.47E+02 1.34E+02 8.94E+01 2.62E+00 7.57E-04 3.72E-09 4.42E-20 O.OOEO00 Os191 5.96E+OI 5.96E+01 5.96E+Oi 5.96E+OI 5.94E+01 5.91E+OI 5.74E+01 4.52E+01 2.60E+01 1.13E+01 2.05EO00 2.42E-03 OsI91m 1.09E-01 1.08E-01 4.26E-07 1.66E-12 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Ir192 5.05E+01 5.05E+01 5.03E+01 5.01E+OI 4.75E+01 4.21E+OI 2.42E+01 2.05E-01 3.39E-06 2.27E-13 3.89E-28 O.OOE+00 Np236m 5.17E+02 5.17E+02 5.10E+02 5.02E+02 4.10E+02 2.58E+02 3.19E+01 4.41E-07 3.22E-25 O.OOE+00 O.OOE+00 O.OOE+00 U237 2.72E+00 2.72E+00 2.71E+00 2.71E+00 2.71E+00 2.69EO00 2.61E+00 2.01E+00 I.I IEE00 4.50E-01 7.09E-02 4.83E-05 Pu237 3.39E+02 3.39E+02 3.36E+02 3.31E+02 2.50E+02 1.31E+02 9.44E+00 2.03E+00 1.12E+00 4.55E-01 7.16E-02 4.88E-05 Np238 3.29E+02 3.29E+02 1.08E+02 3.52E+01 5.61E-06 1.63E-21 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Pu238 6.33E-01 6.33E-0I 6.33E-01 6.33E-01 6.29E-01 6.17E-01 5.43E-01 1.69E-01 1.13E-02 1.98E-04 4.74E-08 2.52E-22.U239 4.70E-01 4.70E-01 4.58E-01 4.46E-01 3.08E-01 1.32E-01 2.93E-03 1.34E-17 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Np239 3.14E-01 3.14E-01 3.14E-01 3.14E-01 3.13E-01 3.12E-01 3.03E-01 2.37E-01 1.35E-01 5.80E-02 1.02E-02 1.12E-05 Pu239 2.47E-02 2.47E-02 2.43E-02 2.39E-02 1.85E-02 1.04E-02 7.66E-04 1.85E-08 1.82E-08 1.77E-08 1.67E-08 1.32E-08 Np240 1.67E-05 1.67E+05 1.66E405 1.66E+05 1.61E+05 1.50E+05 I.i IE+05 7.65E+03 1.69E+01 7.36E-01 7.16E-01 6.50E-01 Pu240 .3.27E+06 3.27E+06 1.35E+06 5.56E+05 2.28E+00 1.10E-12 0.00+E00 0.OOE+00 0.OOE+00 0.OOE0+00 0.00E400 0.00E+00 Pu241 1.15E+00 1.15E+00 1.14E+00 1.12E+00 9.01E-01 5.51E-01 5.99E-02 2.69E-10 1.46E-29 0.00E+00 0.OOE+00 0.00E+00 Am241 2.52E+00 2.52E+00 2.52E+00 2.52E+00 2.51E+00 2.48E+00 2.37E+00 1.59E+00 6.33E-01 1.59E-01 9.28E-03 1.26E-07 Am242m 1. 11E+05 1.1 E+05 1.11 E+05 1.10E+05 9.99E+04 8.03E+04 3.01E+04 6.05E+O 0 1.35E-02 1.35E-02 1.35E-02 1.33E-02 Am242 1.38E+03 1.38E+03 1.38E+03 i.38E+03 1.38E+03 1.38E+03 1.39E+03 1.40E+03 1.41E+03 1.43E+03 1.45E+03 1.44E+03 Cm242 3.26E+06 3.26E+06 3.26E+06 3.24E+06 2.98E+06 2.45E+06 1.01E+06 4.94E+02 1.23E+01 1.23E+01 1.23E+01 1.23E+01 Pu243 6.13E+01 6.13E+01 6.13E+01 6.13E+01 6.14E+01 6.16E+01 6.19E+01. 6.22E+01 6.22E+01 6.22E+01 6.22E+01 6.22E+01 Am243 7.18E+03 7.18E+03 5.13E+03 3.67E+03 3.33E+01 7.13E-04 4.03E-17 4.55E-17 5.72E-17 7.48E-17 1.1 IE-16 2.54E-16 Cm243 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 1.32E+02 I.32E+02 1.32E+02 1.32E+02 1.33E+02 A- 51 Safety Analysis Report for Susquehanna Units 1 and 2-Constant Pressure Extended Power Uprate, October 2006-Non-Proprietary Version -Table A-2 Activity (Curies) per Single Assembly of ATRIUM-10 Fuel Sum of Actinides, Fission Products and Light Elements..... Decay time Following Burnup to 58GWd/MTU-,-.v Nuclide T_=0- -1 isec. 30miný hr :8 hi. '1.0 d, :4.0 d 30.0 d 90.0 d 180.0 d 1 yr. 3 yr Am244 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.14E+04 3.13E+04 3.10E+04 3.07E+04 2.99E+04 2.72E+04 Cm244 8.48E-01 8.48E-01 8A8E-01 8.48E-01 8.48E-01 8.48E-01 8.48E-01 8.48E-01 8.48E-01 8.48E-01 8.48E-01 8A.E-01 A-52

Attachment 7 to PLA-6076 List of Planned Modifications ATTACHMENT 7 PPL SUSQUEHANNA LLC SUSQUEHANNA STEAM ELECTRIC STATION (SSES)UNITS I AND 2 REQUEST FOR LICENSE AMENDMENT FOR EXTENDED POWER UPRATE OPERATION LIST OF PLANNED MODIFICATIONS The following is a list of planned modifications necessary to support Extended Power Uprate (EPU) for Susquehanna Steam Electric Station (SSES) Units 1 and 2.This list is organized to illustrate the expected implementation sequencing. Susquehanna Units 1 and 2 EPU Planned Modifications, Modification Description Pre-EPU Implementation Phase (U212RIO-2005 and U1.14RIO-2006) Vibration/Acoustic

• Install accelerometers on Main Monitoring Steam, Reactor Recirculation, RHR (Steam dryer monitoring and RWCU Lines for vibration instruments on Unit 1 Only) monitoring
• Install instrumentation on main steam lines for steam dryer acoustic wave monitoring Cross Around Relief Valve 0 Revise setpoint for EPU conditions Set Point Change 0 Revise design pressure of (Unit 1 Only) associated piping for EPU conditions
• Replace relief line expansion joints for EPU steam flow conditions Reactor Feed Pump Seal
• Revise Temperature Control Valve Water settings per vendor (Unit I Only) recommendation
• Revise drain line vent piping for increased drain flow I Susquehanna Units 1 and 2 EPU Planned Modifications Modification

'Description Power Range Neutron

• Replace existing GE analog system Monitoring System with GE digital NUMAC system (Unit I Only)[Provided for completeness only. NRC approval has been requested in a separate, prior submittal]

EPU Phase I Implementation (2007 Non-Outaqe, U213RIO-2007 and U115RIO-2008) I Ultimate Heat Sink

• Install a second isolation valve, (Non-Outage) manually operated, in each of two Spray Pond Spray Header Bypass Lines to reduce effects of a single bypass line isolation valve failure-to-close under accident conditions.
• Reduce number of large array nozzles to improve spray efficiency ESW to Fuel Pool Check
• Valve change to reduce mission dose Valve for post-LOCA manual action (Non-Outage)

ARTS/MELLLA

• Revise the APRM flow-biased scram (Unit I Non-Outage) and rod block trip setpoints[Provided for completeness only. NRC approval has been requested in a separate, prior submittal]

Appendix R RHR Pump a Logic change and raceway protection Logic Change (Non-Outage) to eliminate fire-induced failure mechanisms

• Provide cross-divisional cooling to RHR pump motor oil coolers Acid Injection (Non-Outage)
• Provide additional acid injection capability for the Cooling Tower basin Vibration/Acoustic 0 Install instrumentation on main steam Monitoring lines for steam dryer acoustic wave (Unit 2 Only) monitoring Cross Around Relief Valve 0 Revise setpoint for EPU conditions Set Point Change
• Revise design pressure of associated (Unit 2 Only) piping for EPU conditions
• Replace relief line expansion joints for EPU steam flow conditions 2

Susquehanna Units 1 and 2 EPU Planned.Modifications Modification I Description Reactor Feed Pump Seal

• Revise Temperature Control Valve Water settings per vendor recommendation (Unit 2 Only)
• Revise drain line vent piping for increased drain flow Power Range Neutron 0 Replace existing GE analog system Monitoring System with with GE digital NUMAC system ARTS/MELLLA
• Revise the APRM flow-biased scram (Unit 2 Only) and rod block trip setpoints[Provided for completeness only. NRC approval has been requested in a separate, prior submittal]

Neutron Monitoring System

• APRM Flow-biased SCRAM Settings
• APRM Flow-biased Rod Block o APRM Upscale Setdown SCRAM* APRM Upscale Setdown Rod Block EHC System
• Install accumulators on Turbine Control Valve EHC FAS lines* Install Steam Line Resonance Cards on pressure transmitter loops to dampen 3 rd harmonic frequency* Modify Turbine Control Valve Digital Positioning Cards* Recalibrate Power Load Unbalance circuit MSIV High Flow Isolation
• Revise setpoint for EPU conditions Setpoint (this will require new switches)Reactor Recirculation
• Logic change Runback Limiter #2 HP Turbine Instrument
• RPS SCRAM Bypass Change
• RWM Setpoints* RSCS Setpoints* Power dependent condenser high pressure alarm power signal Reactor Feed Pump Low 0 Revise setpoints for EPU Conditions Suction Pressure Instrument Calibration and
• Recalibrate instruments and revise Computer Software software for EPU conditions Changes Generator Rewind
• Increase main generator electrical rating to EPU conditions 3

Susquehanna Units I and.2 EPU Planned Modifications Modification ' Description High Pressure Turbine

• Replace High Pressure Turbine for increased steam flow at EPU conditions Condensate Pump Impellers 0 Replace Condensate Pump Impellers for increased Condensate flow at EPU conditions
• Replace minimum flow valve internals and controls to allow a larger minimum flow* Replace pump discharge valve motors to accommodate higher differential pressure#5 Feedwater Heaters 0 Increase design pressure and increase shell relief valve setpoints FW Heaters 0 Changes to manage velocity and tube vibration issues at EPU conditions Standby Liquid Control 0 Replace existing sodium pentaborate Boron Enrichment solution* Modify system logic to allow for single pump initiation

[Provided for completeness only. NRC approval has been requested in a separate submittal] Circulating Water Box Vents 0 Add automatic Circulating Water Box vent valves to prevent air binding of condenser tubes Hydrogen Water Chemistry 0 Increase hydrogen, oxygen and zinc injection flows due to increased Feedwater flow under EPU conditions EPU Implementation 0 Configuration modification for EPU implementation. No physical work involved Reactor Feed Pump Suction 0 Replace suction flanges, revise Piping piping design pressure, revise relief valve setpoints Main Steam, Feedwater, and 0 Revise piping supports as necessary Extraction Steam Piping for EPU conditions Supports Gaseous Radwaste 0 Revise drain piping for increased Recombiner Drain Piping EPU flow conditions

1. 3 FWH Emergency Dump 0 Replace valves for EPU conditions.

Valves 4 Susquehanna Units I and 2 EPU Planned Modifications Modification I Description Power

• Install new switchyard capacitor Distribution/Switchyard banks to meet PJM reactive power requirements for generators
• Replace Unit I Sync Breaker and associated controls with a breaker having a higher amperage rating* Uprate Unit 2 main transformers and change tap settings to meet EPU conditions Potential EQ Changes
• As required for EPU environmental conditions Potential Steam Dryer ' Reinforce steam dryer to mitigate Changes structural loads at EPU conditions EPU Phase II Implementation (U214RIO-2009 and U115RIO-2010)

FW Heaters

• Additional increase to #5 heater design pressure and shell relief valve setpoints Reactor Feed Pump 0 Replace Reactor Feed Pump Turbines Turbines due to higher turbine speeds required at EPU conditions 0 Upgrade turbine speed controls and overspeed trip to digital controls Condensate Demineralizer 0 Install an 8 th Condensate Demineralizer to maintain Condensate water quality under increased EPU flow conditions Condensate Filter o Install a 7 th Condensate Filter to maintain Condensate water quality under increased EPU flow conditions RWCU Filter/Demineralizers
• Internal changes to improve performance 5}}