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{{#Wiki_filter:ATTACHMENT 8 GEH Nuclear Energy Safety Analysis Report for Limerick Generating Station, Units 1 and 2 Thermal Power Optimization, NEDO-33484 (Non-Proprietary Version)  
{{#Wiki_filter:ATTACHMENT 8 GEH Nuclear Energy Safety Analysis Report for Limerick Generating Station, Units 1 and 2 Thermal Power Optimization, NEDO-33484 (Non-Proprietary Version)


HITACHI Non-Proprietary Information SAFETY ANALYSIS REPORT FOR LIMERICK GENERATING STATION UNITS 1 AND 2 THERMAL POWER OPTIMIZATION Copyright 2010 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved GE Hitachi Nuclear Energy NEDO-33484 Revision 0 Class I DRF 0000-0095-5957 March 2010 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION ii NON-PROPRIETARY INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33484P, Revision 0, from which the proprietary information has been removed. Portions of the document that have been removed are identified by white space within double square brackets, as shown here [[      ]].
HITACHI                                   GE Hitachi Nuclear Energy NEDO-33484 Revision 0 Class I DRF 0000-0095-5957 March 2010 Non-Proprietary Information SAFETY ANALYSIS REPORT FOR LIMERICK GENERATING STATION UNITS 1 AND 2 THERMAL POWER OPTIMIZATION Copyright 2010 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved


IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The design, engineering, and other information contained in this document is furnished for the purposes of supporting the Exelon Generation Company, LLC (Exelon) license amendment request for a thermal power uprate at Limerick Generating Station Units 1 and 2 to 3515 MWt in proceedings before the U.S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilit ies, and nothing contained in this document shall be construed as changing that contract. The use of this informa tion by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no  
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33484P, Revision 0, from which the proprietary information has been removed. Portions of the document that have been removed are identified by white space within double square brackets, as shown here ((      )).
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The design, engineering, and other information contained in this document is furnished for the purposes of supporting the Exelon Generation Company, LLC (Exelon) license amendment request for a thermal power uprate at Limerick Generating Station Units 1 and 2 to 3515 MWt in proceedings before the U.S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilities, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
No use of or right to copy any of this information contained in this document, other than by the NRC and its contractors in support of GEHs application, is authorized except by contract with GEH, as noted above. The information provided in this document is part of and dependent upon a larger set of knowledge, technology, and intellectual property rights pertaining to the design of standardized, nuclear powered, electric generating facilities. Without access and a GEH grant of rights to that larger set of knowledge, technology, and intellectual property rights, this document is not practically or rightfully usable by others, except by the NRC or through contractual agreements with Exelon, as set forth in the previous paragraph.
Copyright 2010 GE-Hitachi Nuclear Energy Americas LLC, All Rights Reserved ii


representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document. No use of or right to copy any of this information contained in this document, other than by the NRC and its contractors in support of GEH's app lication, is authorized except by contract with GEH, as noted above. The information provided in this document is part of and dependent upon a larger set of knowledge, technology, and intellectua l property rights pertai ning to the design of standardized, nuclear powered, elect ric generating facilities. Without access and a GEH grant of rights to that larger set of knowledge, technology, and intellectual property rights, this document is not practically or rightfully usable by othe rs, except by the NRC or through contractual agreements with Exelon, as set forth in the previous paragraph.
NEDO


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION ix LIST OF FIGURES Figure No.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.5.1 Reactor Coolant Pressure Boundary Piping....................................................... 3-11 3.5.2 Balance-of-Plant Piping Evaluation................................................................... 3-15 3.6 Reactor Recirculation System ..................................................................................... 3-16 3.7 Main Steam Line Flow Restrictors.............................................................................. 3-16 3.8 Main Steam Isolation Valves....................................................................................... 3-16 3.9 Reactor Core Isolation Cooling ................................................................................... 3-17 3.10 Residual Heat Removal System .................................................................................. 3-17 3.11 Reactor Water Cleanup System................................................................................... 3-17 4.0 Engineered Safety Features............................................................................................. 4-1 4.1 Containment System Performance ................................................................................ 4-1 4.1.1  Generic Letter 89-10 Program ............................................................................. 4-1 4.1.2  Generic Letter 95-07 Program ............................................................................. 4-2 4.1.3  Generic Letter 96-06 ............................................................................................ 4-2 4.1.4  Containment Coatings.......................................................................................... 4-2 4.2 Emergency Core Cooling Systems ................................................................................ 4-2 4.2.1  High Pressure Coolant Injection .......................................................................... 4-2 4.2.2  High Pressure Core Spray.................................................................................... 4-3 4.2.3  Core Spray ........................................................................................................... 4-3 4.2.4  Low Pressure Coolant Injection........................................................................... 4-3 4.2.5  Automatic Depressurization System.................................................................... 4-3 4.2.6  ECCS Net Positive Suction Head ........................................................................ 4-3 4.3 Emergency Core Cooling System Performance ............................................................ 4-4 4.4 Main Control Room Atmosphere Control System ........................................................ 4-4 4.5 Standby Gas Treatment System..................................................................................... 4-4 4.6 Main Steam Isolation Valve Leakage Alternate Drain Pathway................................... 4-4 4.7 Post-LOCA Combustible Gas Control System ............................................................. 4-5 5.0 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.......................................................................................... 5-2 5.2 BOP Monitoring and Control ........................................................................................ 5-2 5.2.1  Pressure Control System ...................................................................................... 5-2 5.2.2  EHC Turbine Control System.............................................................................. 5-3 5.2.3  Feedwater Control System................................................................................... 5-3 5.2.4  Leak Detection System ........................................................................................ 5-3 5.3 Technical Specification Instrument Setpoints ............................................................... 5-4 5.3.1  High-Pressure Scram ........................................................................................... 5-4 5.3.2  Hydraulic Pressure Scram.................................................................................... 5-4 5.3.3  High-Pressure Recirculation Pump Trip.............................................................. 5-5 5.3.4  Safety Relief Valve .............................................................................................. 5-5 5.3.5  Main Steam Line High Flow Isolation................................................................. 5-5 5.3.6  Fixed APRM Scram............................................................................................. 5-5 iv
Title 1-1 Power/Flow Map for the TPO (101.65% of CLTP) 1-2 Reactor Heat Balance - TPO Power (101.65% of CLTP), 100% Core Flow 2-1 Illustration of OPRM Trip-Enabled Region 2-2 BSP Regions for Nominal Feedwater Temperature Demonstration 2-3 BSP Regions for Minimum Feedwater Temperature Demonstration 9-1 TPO RTP MELLLA BO C MSIVC (Short Term) 9-2 TPO RTP MELLLA BOC MSIVC (Long Term - A) 9-3 TPO RTP MELLLA BOC MSIVC (Long Term - B) 9-4 TPO RTP MELLLA BOC MSIVC (Long Term - C) 9-5 TPO RTP MELLLA BO C PRFO (Short Term) 9-6 TPO RTP MELLLA BOC PRFO (Long Term - A) 9-7 TPO RTP MELLLA BOC PRFO (Long Term - B) 9-8 TPO RTP MELLLA BOC PRFO (Long Term - C) 9-9 TPO RTP MELLLA BOC LOOP (Short Term) 9-10 TPO RTP MELLLA BOC LOOP (Long Term - A) 9-11 TPO RTP MELLLA BOC LOOP (Long Term - B) 9-12 TPO RTP MELLLA BOC LOOP (Long Term - C) 9-13 TPO RTP MELLLA EOC MSIVC (Short Term) 9-14 TPO RTP MELLLA EOC MSIVC (Long Term - A) 9-15 TPO RTP MELLLA EOC MSIVC (Long Term - B) 9-16 TPO RTP MELLLA EOC MSIVC (Long Term - C) 9-17 TPO RTP MELLLA EOC PRFO (Short Term) 9-18 TPO RTP MELLLA EOC PRFO (Long Term - A) 9-19 TPO RTP MELLLA EOC PRFO (Long Term - B) 9-20 TPO RTP MELLLA EOC PRFO (Long Term - C) 9-21 TPO RTP MELLLA EOC LOOP (Short Term) 9-22 TPO RTP MELLLA EOC LOOP (Long Term - A)
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION x Figure No.
Title 9-23 TPO RTP MELLLA EOC LOOP (Long Term - B) 9-24 TPO RTP MELLLA EOC LOOP (Long Term - C)


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION xi ACRONYMS AND ABBREVIATIONS Term Definition ABA Amplitude Based Algorithm AC Alternating Current ADS Automatic Depressurization System AL Analytical Limit ALARA As Low As Reasonably Achievable AOO Anticipated Oper ational Occurrence AP Annulus Pressurization APRM  Average Power Range Monitor ART Adjusted Reference Temperature ARTS Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program ASME American Society of Mechanical Engineers ATWS Anticipated Transient Without Scram AV Allowable Value B&PV Boiler and Pressure Vessel BHP Brake Horsepower BOP Balance of Plant BSP Backup Stability Protection BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Project CD Condensate Demineralizer CFR Code of Federal Regulations CLTP Current Licensed Thermal Power CRD Control Rod Drive CRGT Control Rod Guide Tube CS Core Spray CSC Containment Spray Cooling CSS Core Support Structure CUF Cumulative Usage Factor DBA Design Basis Accident DC Direct Current ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EFPY Effective Full Power Years EHC Electro-Hydraulic Control ELTR1 NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION xii Term Definition ELTR2 NEDC-32523P-A, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate EOC End of Cycle EOOS Equipment Out-of-Service EOP Emergency Operating Procedure EPG Emergency Procedure Guidelines EPU Extended Power Uprate EQ Environmental Qualification Exelon Exelon Generation Company, LLC FAC Flow Accelerated Corrosion FFWTR Final Feedwater Temperature Reduction FIV Flow-Induced Vibration FPCC Fuel Pool Cooling And Cleanup FW Feedwater FWHOOS Feedwater Heater(s) Out-of-Service GDC General Design Criterion GE General Electric Company GEH GE Hitachi Nuclear Energy GL Generic Letter GRA Growth Rate Algorithm HELB High Energy Line Break HEPA High Efficiency Particulate Air HFCL High Flow Control Line HPCI High Pressure Coolant Injection HVAC Heating, Ventilation And Air Conditioning IASCC Irradiation Assisted Stress Corrosion Cracking ICF Increased Core Flow ICH&GT In-Core Housing and Guide Tube IPE Individual Plant Examination IRM Intermediate Range Monitor JR Jet Reaction ksi Kips Per Square Inch kV Kilovolt kW Kilowatt LBPCT Licensing Basis Peak Clad Temperature LCO Limiting Conditions For Operation LHGR Linear Heat Generation Rate Limerick Limerick Generating Station Units 1 and 2 LOCA Loss-of-Coolant-Accident NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION xiii Term Definition LOOP Loss of Offsite Power LPCI Low Pressure Coolant Injection LPRM Local Power Range Monitor LPSP Low Power Setpoint MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCC Motor Control Circuit/Center MCPR Minimum Critical Power Ratio MELB Moderate Energy Line Break MELLLA Maximum Extended Load Line Limit Analysis
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.3.7    APRM Flow-Biased Scram.................................................................................. 5-5 5.3.8    Rod Worth Minimizer Low Power Setpoint........................................................ 5-6 5.3.9    Rod Block Monitor .............................................................................................. 5-6 5.3.10 Flow-Biased Rod Block Monitor (%RTP) .......................................................... 5-6 5.3.11 Main Steam Line High Radiation Isolation ......................................................... 5-6 5.3.12 Low Steam Line Pressure MSIV Closure (RUN Mode) ..................................... 5-6 5.3.13 Reactor Water Level Instruments ........................................................................ 5-6 5.3.14 Main Steam Line Tunnel High Temperature Isolations ...................................... 5-7 5.3.15 Low Condenser Vacuum...................................................................................... 5-7 5.3.16 TSV Closure Scram, TCV Fast Closure Scram Bypasses ................................... 5-7 6.0   Electrical Power and Auxiliary Systems ........................................................................ 6-1 6.1 AC Power ...................................................................................................................... 6-1 6.1.1    Off-Site Power ..................................................................................................... 6-1 6.1.2    On-Site Power...................................................................................................... 6-2 6.2 DC Power ...................................................................................................................... 6-3 6.3 Fuel Pool........................................................................................................................ 6-3 6.3.1    Fuel Pool Cooling ................................................................................................ 6-3 6.3.2    Crud Activity and Corrosion Products................................................................. 6-3 6.3.3    Radiation Levels .................................................................................................. 6-4 6.3.4    Fuel Racks............................................................................................................ 6-4 6.4 Water Systems ............................................................................................................... 6-4 6.4.1    Service Water Systems ........................................................................................ 6-4 6.4.2    Main Condenser/Circulating Water/Normal Heat Sink Performance ................. 6-5 6.4.3    Reactor Enclosure Cooling Water System........................................................... 6-5 6.4.4    Turbine Enclosure Cooling Water System .......................................................... 6-6 6.4.5    Ultimate Heat Sink............................................................................................... 6-6 6.5 Standby Liquid Control System .................................................................................... 6-6 6.6 Power-dependent Heating, Ventilation and Air Conditioning ...................................... 6-7 6.7 Fire Protection ............................................................................................................... 6-7 6.7.1    10 CFR 50 Appendix R Fire Event...................................................................... 6-7 6.8 Systems Not Affected By TPO Uprate.......................................................................... 6-7 7.0 Power Conversion Systems.............................................................................................. 7-1 7.1 Turbine-Generator ......................................................................................................... 7-1 7.2 Condenser And Steam Jet Air Ejectors ......................................................................... 7-1 7.3 Turbine Steam Bypass................................................................................................... 7-2 7.4 Feedwater And Condensate Systems............................................................................. 7-2 7.4.1    Normal Operation ................................................................................................ 7-2 7.4.2    Transient Operation ............................................................................................. 7-3 7.4.3    Condensate Filters and Condensate Deep Bed Demineralizers........................... 7-3 8.0 Radwaste and Radiation Sources.................................................................................... 8-1 8.1  Liquid and Solid Waste Management ........................................................................... 8-1 8.2  Gaseous Waste Management......................................................................................... 8-1 v


MeV Million Electron Volts MFWT Minimum Feedwater Temperature Mlb Millions of Pounds MOV Motor Operated Valve MS Main Steam MSF Modified Shape Function MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSL Main Steam Line MSLB Main Steam Line Break MVA Million Volt Amps MWe Megawatt-Electric MWt Megawatt-Thermal NCL Natural Circulation Line NFWT Nominal Feedwater Temperature NPDES National Pollutant Discharge Elimination System NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NTSP Nominal Trip Setpoint NUREG Nuclear Regula tions (NRC Document) OLMCPR Operating Limit Minimum Critical Power Ratio OOS Out-of-Service P/F Power/Flow P-T Pressure-Temperature PBDA Period Based Detection Algorithm PCS Pressure Control System PCT Peak Clad Temperature PRA Probabilistic Risk Assessment NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION xiv Term Definition PRFO Pressure Regulator Failure Open - Maximum Steam Demand psi Pounds Per Square Inch psia Pounds Per Square Inch - Absolute psid Pounds Per Square Inch - Differential psig Pounds Per Square Inch - Gauge RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RG Regulatory Guide RHR Residual Heat Removal RIPD Reactor Internal Pressure Difference RIS Regulatory Issue Summary RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RTNDT Reference Temperature Of Nil-Ductility Transition RTP Rated Thermal Power RWCU Reactor Water Cleanup RWM Rod Worth Minimizer SAG Severe Accident Guidelines SAR Safety Analysis Report SBO Station Blackout SBPCS Steam Bypass Pressure Control System SDC Shutdown Cooling SER Safety Evaluation Report SFP Spent Fuel Pool SGTS Standby Gas Treatment System SJAE Steam Jet Air Ejector SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single (Recirculation) Loop Operation SPC Suppression Pool Cooling SR Surveillance Requirement SRM Source Range Monitor SRP Standard Review Plan SRV Safety Relief Valve SRVDL Safety Relief Valve Discharge Line TBV Turbine Bypass Valve TCV Turbine Control Valve NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION xv Term Definition TFSP Turbine First Stage Pressure T/G Turbine-Generator TIP Traversing In-Core Probe TLO Two (Recirculation) Loop Operation TLTP TPO Licensed Thermal Power TLTR NEDC-32938P-A, Thermal Power Optimization Licensing Topical Report TPO Thermal Power Optimization TSAR Thermal Power Optimization Safety Analysis Report TSV Turbine Stop Valve UBPCT Upper Bound Peak Clad Temperature UFM Ultrasonic Flow Measurement UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink USE Upper Shelf Energy VWO Valves Wide Open Wd Recirculation Drive Flow NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION xvi EXECUTIVE  
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8.3 Radiation Sources in the Reactor Core.......................................................................... 8-2 8.4 Radiation Sources in Reactor Coolant........................................................................... 8-3 8.4.1    Coolant Activation Products ................................................................................ 8-3 8.4.2    Activated Corrosion Products .............................................................................. 8-3 8.4.3    Fission Products ................................................................................................... 8-4 8.5 Radiation Levels............................................................................................................ 8-4 8.6 Normal Operation Off-Site Doses ................................................................................. 8-5 9.0 Reactor Safety Performance Evaluations ...................................................................... 9-1 9.1 Anticipated Operational Occurrences............................................................................ 9-1 9.2 Design Basis Accidents ................................................................................................. 9-1 9.3 Special Events ............................................................................................................... 9-2 9.3.1    Anticipated Transient Without Scram ................................................................. 9-2 9.3.2    Station Blackout................................................................................................... 9-8 10.0 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 ............................................................................................ 10-1 10.2 Moderate Energy Line Break ...................................................................................... 10-2 10.3 Environmental Qualification ....................................................................................... 10-3 10.3.1 Electrical Equipment.......................................................................................... 10-3 10.3.2 Mechanical Equipment With Non-Metallic Components.................................. 10-4 10.3.3 Mechanical Component Design Qualification................................................... 10-4 10.4 Testing ......................................................................................................................... 10-4 10.5 Operator Training And Human Factors....................................................................... 10-5 10.6 Plant Life ..................................................................................................................... 10-5 10.7 NRC and Industry Communications ........................................................................... 10-6 10.8 Plant Procedures and Programs ................................................................................... 10-6 10.9 Emergency Operating Procedures ............................................................................... 10-6 10.10 Individual Plant Examination ...................................................................................... 10-6 11.0 References..................................................................................................................... 11-1 vi
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION LIST OF TABLES Table No.                                            Title 1-1    Computer Codes Used for TPO Analyses 1-2    Thermal-Hydraulic Parameters at TPO Uprate Conditions 1-3    Summary of Effect of TPO Uprate on Licensing Criteria 2-1    OPRM Setpoint Versus OLMCPR Demonstration 2-2    BSP Region Intercepts for Nominal Feedwater Temperature Demonstration 2-3    BSP Region Intercepts for Minimum Feedwater Temperature Demonstration 3-1    Limerick Unit 1 Upper Shelf Energy 40-Year License (32 EFPY) 3-2    Limerick Unit 2 Upper Shelf Energy 40-Year License (32 EFPY) 3-3    Limerick Unit 1 Adjusted Reference Temperatures 40-Year License (32 EFPY) 3-4    Limerick Unit 2 Adjusted Reference Temperatures 40-Year License (32 EFPY) 3-5    Limerick 32 EFPY Effects of Irradiation on RPV Axial Weld Properties 3-6    Limerick 32 EFPY Effects of Irradiation on RPV Circumferential Weld Properties 3-7    CUF and P+Q Stress Range of Limiting Components 3-8    Governing Stress Results for RPV Internal Components 3-9    Piping Lines Recommended for Special Focus under FAC Review 4-1    Limerick ECCS-LOCA Analysis Results for GE14 Fuel 5-1    Analytical Limits that Change due to TPO 6-1    TPO Plant Electrical Characteristics 6-2    Main Generator Ratings Comparison 6-3a    Limerick 1 Generator Step-up Transformer Ratings Comparison 6-3b    Limerick 2 Generator Step-up Transformer Ratings Comparison 6-4    Unit Auxiliary Transformer Ratings Comparison 6-5a    Station Auxiliary Transformer Ratings Comparison 6-5b    Regulating Transformer Ratings Comparison 6-6    Fuel Pool Cooling and Cleanup Parameters 6-7    Effluent Discharge Comparison 9-1    Key Inputs for ATWS Analysis vii
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table No.                                        Title 9-2    Results for ATWS Analysis 9-3    Inputs for Three SLCS Pump Operation ATWS Analysis 9-4    Inputs for Two SLCS Pump Operation ATWS Analysis 9-5    Limerick SLCS Pressure Results for ATWS Analysis 9-6    MSIVC Sequence of Events 9-7    PRFO Sequence of Events 9-8    LOOP Sequence of Events viii
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION LIST OF FIGURES Figure No.                                          Title 1-1    Power/Flow Map for the TPO (101.65% of CLTP) 1-2    Reactor Heat Balance - TPO Power (101.65% of CLTP), 100% Core Flow 2-1    Illustration of OPRM Trip-Enabled Region 2-2    BSP Regions for Nominal Feedwater Temperature Demonstration 2-3    BSP Regions for Minimum Feedwater Temperature Demonstration 9-1    TPO RTP MELLLA BOC MSIVC (Short Term) 9-2    TPO RTP MELLLA BOC MSIVC (Long Term - A) 9-3    TPO RTP MELLLA BOC MSIVC (Long Term - B) 9-4    TPO RTP MELLLA BOC MSIVC (Long Term - C) 9-5    TPO RTP MELLLA BOC PRFO (Short Term) 9-6    TPO RTP MELLLA BOC PRFO (Long Term - A) 9-7    TPO RTP MELLLA BOC PRFO (Long Term - B) 9-8    TPO RTP MELLLA BOC PRFO (Long Term - C) 9-9    TPO RTP MELLLA BOC LOOP (Short Term) 9-10    TPO RTP MELLLA BOC LOOP (Long Term - A) 9-11    TPO RTP MELLLA BOC LOOP (Long Term - B) 9-12    TPO RTP MELLLA BOC LOOP (Long Term - C) 9-13    TPO RTP MELLLA EOC MSIVC (Short Term) 9-14    TPO RTP MELLLA EOC MSIVC (Long Term - A) 9-15    TPO RTP MELLLA EOC MSIVC (Long Term - B) 9-16    TPO RTP MELLLA EOC MSIVC (Long Term - C) 9-17    TPO RTP MELLLA EOC PRFO (Short Term) 9-18    TPO RTP MELLLA EOC PRFO (Long Term - A) 9-19    TPO RTP MELLLA EOC PRFO (Long Term - B) 9-20    TPO RTP MELLLA EOC PRFO (Long Term - C) 9-21    TPO RTP MELLLA EOC LOOP (Short Term) 9-22    TPO RTP MELLLA EOC LOOP (Long Term - A) ix
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure No.                                  Title 9-23    TPO RTP MELLLA EOC LOOP (Long Term - B) 9-24    TPO RTP MELLLA EOC LOOP (Long Term - C) x
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION ACRONYMS AND ABBREVIATIONS Term                                    Definition ABA        Amplitude Based Algorithm AC        Alternating Current ADS        Automatic Depressurization System AL        Analytical Limit ALARA      As Low As Reasonably Achievable AOO        Anticipated Operational Occurrence AP        Annulus Pressurization APRM      Average Power Range Monitor ART        Adjusted Reference Temperature ARTS      Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program ASME      American Society of Mechanical Engineers ATWS      Anticipated Transient Without Scram AV        Allowable Value B&PV      Boiler and Pressure Vessel BHP        Brake Horsepower BOP        Balance of Plant BSP        Backup Stability Protection BWR        Boiling Water Reactor BWRVIP    Boiling Water Reactor Vessel and Internals Project CD        Condensate Demineralizer CFR        Code of Federal Regulations CLTP      Current Licensed Thermal Power CRD        Control Rod Drive CRGT      Control Rod Guide Tube CS        Core Spray CSC        Containment Spray Cooling CSS        Core Support Structure CUF        Cumulative Usage Factor DBA        Design Basis Accident DC        Direct Current ECCS      Emergency Core Cooling System EDG        Emergency Diesel Generator EFPY      Effective Full Power Years EHC        Electro-Hydraulic Control ELTR1      NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate xi
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term                                    Definition ELTR2      NEDC-32523P-A, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate EOC        End of Cycle EOOS      Equipment Out-of-Service EOP        Emergency Operating Procedure EPG        Emergency Procedure Guidelines EPU        Extended Power Uprate EQ        Environmental Qualification Exelon    Exelon Generation Company, LLC FAC        Flow Accelerated Corrosion FFWTR      Final Feedwater Temperature Reduction FIV        Flow-Induced Vibration FPCC      Fuel Pool Cooling And Cleanup FW        Feedwater FWHOOS    Feedwater Heater(s) Out-of-Service GDC        General Design Criterion GE        General Electric Company GEH        GE Hitachi Nuclear Energy GL        Generic Letter GRA        Growth Rate Algorithm HELB      High Energy Line Break HEPA      High Efficiency Particulate Air HFCL      High Flow Control Line HPCI      High Pressure Coolant Injection HVAC      Heating, Ventilation And Air Conditioning IASCC      Irradiation Assisted Stress Corrosion Cracking ICF        Increased Core Flow ICH&GT    In-Core Housing and Guide Tube IPE        Individual Plant Examination IRM        Intermediate Range Monitor JR        Jet Reaction ksi        Kips Per Square Inch kV        Kilovolt kW        Kilowatt LBPCT      Licensing Basis Peak Clad Temperature LCO        Limiting Conditions For Operation LHGR      Linear Heat Generation Rate Limerick  Limerick Generating Station Units 1 and 2 LOCA      Loss-of-Coolant-Accident xii
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term                                    Definition LOOP      Loss of Offsite Power LPCI      Low Pressure Coolant Injection LPRM      Local Power Range Monitor LPSP      Low Power Setpoint MAPLHGR  Maximum Average Planar Linear Heat Generation Rate MCC      Motor Control Circuit/Center MCPR      Minimum Critical Power Ratio MELB      Moderate Energy Line Break MELLLA    Maximum Extended Load Line Limit Analysis MeV       Million Electron Volts MFWT     Minimum Feedwater Temperature Mlb       Millions of Pounds MOV       Motor Operated Valve MS       Main Steam MSF       Modified Shape Function MSIV     Main Steam Isolation Valve MSIVC     Main Steam Isolation Valve Closure MSL       Main Steam Line MSLB     Main Steam Line Break MVA       Million Volt Amps MWe       Megawatt-Electric MWt       Megawatt-Thermal NCL       Natural Circulation Line NFWT     Nominal Feedwater Temperature NPDES     National Pollutant Discharge Elimination System NPSH     Net Positive Suction Head NRC       Nuclear Regulatory Commission NSSS     Nuclear Steam Supply System NTSP     Nominal Trip Setpoint NUREG     Nuclear Regulations (NRC Document)
OLMCPR   Operating Limit Minimum Critical Power Ratio OOS       Out-of-Service P/F       Power/Flow P-T       Pressure-Temperature PBDA     Period Based Detection Algorithm PCS       Pressure Control System PCT       Peak Clad Temperature PRA       Probabilistic Risk Assessment xiii
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term                                   Definition PRFO     Pressure Regulator Failure Open - Maximum Steam Demand psi       Pounds Per Square Inch psia     Pounds Per Square Inch - Absolute psid     Pounds Per Square Inch - Differential psig     Pounds Per Square Inch - Gauge RBM       Rod Block Monitor RCIC     Reactor Core Isolation Cooling RCPB     Reactor Coolant Pressure Boundary RG       Regulatory Guide RHR       Residual Heat Removal RIPD     Reactor Internal Pressure Difference RIS       Regulatory Issue Summary RPT       Recirculation Pump Trip RPV       Reactor Pressure Vessel RTNDT     Reference Temperature Of Nil-Ductility Transition RTP       Rated Thermal Power RWCU     Reactor Water Cleanup RWM       Rod Worth Minimizer SAG       Severe Accident Guidelines SAR       Safety Analysis Report SBO       Station Blackout SBPCS     Steam Bypass Pressure Control System SDC       Shutdown Cooling SER       Safety Evaluation Report SFP       Spent Fuel Pool SGTS     Standby Gas Treatment System SJAE     Steam Jet Air Ejector SLCS     Standby Liquid Control System SLMCPR   Safety Limit Minimum Critical Power Ratio SLO       Single (Recirculation) Loop Operation SPC       Suppression Pool Cooling SR       Surveillance Requirement SRM       Source Range Monitor SRP       Standard Review Plan SRV       Safety Relief Valve SRVDL     Safety Relief Valve Discharge Line TBV       Turbine Bypass Valve TCV       Turbine Control Valve xiv
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term                                   Definition TFSP     Turbine First Stage Pressure T/G       Turbine-Generator TIP       Traversing In-Core Probe TLO       Two (Recirculation) Loop Operation TLTP     TPO Licensed Thermal Power TLTR     NEDC-32938P-A, Thermal Power Optimization Licensing Topical Report TPO       Thermal Power Optimization TSAR     Thermal Power Optimization Safety Analysis Report TSV       Turbine Stop Valve UBPCT     Upper Bound Peak Clad Temperature UFM       Ultrasonic Flow Measurement UFSAR     Updated Final Safety Analysis Report UHS       Ultimate Heat Sink USE       Upper Shelf Energy VWO       Valves Wide Open Wd       Recirculation Drive Flow xv
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION EXECUTIVE  


==SUMMARY==
==SUMMARY==
This report summarizes the results of all significant safety evaluations performed that justify increasing the licensed thermal power at Limerick Generating Station Units 1 and 2 (Limerick) to 3515 MWt. The requested license power leve l is 1.65% above the cu rrent licensed thermal power (CLTP) level of 3458 MWt. This report follows the Nuclear Regulatory Commission (NRC) approved format and content for Boiling Water Reactor (BWR) Thermal Power Optimization (TPO) licensing reports documented in NEDC-32938P-A, "Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization," called "TLTR."  Per the outline of the TPO Safety Analysis Report (TSAR) in the TLTR Appendix A, every safety issue that should be addressed in a plant-specific TPO licensing report is addressed in this report. For issues that have been evaluated generically, this report references the appropriate evaluation and establishes that the evaluation is applicable to the plant.
Only previously NRC approved or industry-accepted methods were used for the analysis of accidents, transients, and special events. Therefore, because the safety analysis methods have been previously addressed, they are not addressed in this report. Al so, event and analysis descriptions that are provided in other licensing documents or the Updated Final Safety Analysis Report (UFSAR) are not repeated. This report summarizes the results of the safety evaluations needed to justify a license amendment to allow for TPO operation. The TLTR addresses power increases of up to 1.5% of CLTP, which will produce up to an approximately 2% increase in steam flow to th e turbine-generator. Th e amount of power uprate
( 1.5%) contained in the TLTR was based on the expected reducti on in power level uncertainty with the instrumentation technology available in 1999. The present instrumentation technology has evolved to where a power level uncertainty is reduced to as low as 0.3%, thereby supporting the evaluation of a power level increase of up to 1.7%. A higher steam flow is achieved by increasing the reactor power along the current rod and core flow control lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are recalibrated. Plant procedures are revised, and tests similar to some of the original startup tests are performed.


Evaluations of the reactor, e ngineered safety features, power conversion, emergency power, support systems, environmental issues, design basis accidents, and previous licensing evaluations were performed. This report demonstrates that Limerick can safely operate at a power level of 3515 MWt.  
This report summarizes the results of all significant safety evaluations performed that justify increasing the licensed thermal power at Limerick Generating Station Units 1 and 2 (Limerick) to 3515 MWt. The requested license power level is 1.65% above the current licensed thermal power (CLTP) level of 3458 MWt.
This report follows the Nuclear Regulatory Commission (NRC) approved format and content for Boiling Water Reactor (BWR) Thermal Power Optimization (TPO) licensing reports documented in NEDC-32938P-A, Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization, called TLTR. Per the outline of the TPO Safety Analysis Report (TSAR) in the TLTR Appendix A, every safety issue that should be addressed in a plant-specific TPO licensing report is addressed in this report. For issues that have been evaluated generically, this report references the appropriate evaluation and establishes that the evaluation is applicable to the plant.
Only previously NRC approved or industry-accepted methods were used for the analysis of accidents, transients, and special events. Therefore, because the safety analysis methods have been previously addressed, they are not addressed in this report. Also, event and analysis descriptions that are provided in other licensing documents or the Updated Final Safety Analysis Report (UFSAR) are not repeated. This report summarizes the results of the safety evaluations needed to justify a license amendment to allow for TPO operation.
The TLTR addresses power increases of up to 1.5% of CLTP, which will produce up to an approximately 2% increase in steam flow to the turbine-generator. The amount of power uprate
( 1.5%) contained in the TLTR was based on the expected reduction in power level uncertainty with the instrumentation technology available in 1999. The present instrumentation technology has evolved to where a power level uncertainty is reduced to as low as 0.3%, thereby supporting the evaluation of a power level increase of up to 1.7%. A higher steam flow is achieved by increasing the reactor power along the current rod and core flow control lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are recalibrated. Plant procedures are revised, and tests similar to some of the original startup tests are performed.
Evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, environmental issues, design basis accidents, and previous licensing evaluations were performed. This report demonstrates that Limerick can safely operate at a power level of 3515 MWt.
The following evaluations were conducted in accordance with the criteria of TLTR Appendix B:
xvi


The following evaluations were co nducted in accordance with the criteria of TLTR Appendix B:
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION All safety aspects of the plant that are affected by a 1.65% increase in the thermal power level were evaluated, including the Nuclear Steam Supply System (NSSS) and Balance-of-Plant (BOP) systems.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION xvii All safety aspects of the plant that are affected by a 1.65% increase in the thermal power level were evaluated, including the Nuclear Steam Supply System (NSSS) and Balance-of-Plant (BOP) systems.
Evaluations and reviews were based on licensing criteria, codes, and standards applicable to the plant at the time of the TSAR submittal. There is no change in the previously established licensing basis for the plant, except for the increased power level.
Evaluations and reviews were based on licensing cr iteria, codes, and standards applicable to the plant at the time of the TSAR submittal. There is no change in the previously established licensing basis for the plant, except for the increased power level. Evaluations and/or analyses were performed using NRC-approved or industry-accepted analysis methods for the USAR accidents, transients , and special events affected by TPO. Evaluations and reviews of the NSSS systems and components, containment structures, and BOP systems and components show continued compliance to the codes and standa rds applicable to the current plant licensing basis (i.e., no change to comply with more recent codes and standards is proposed due to TPO). NSSS components and systems were reviewed to confirm that they continue to comply with the functional and regulatory requirements specified in the UFSAR and/or applicable reload license. Any modification to safety-related or non-safety-related equipment will be implemented in accordance with 10 CFR 50.59.
Evaluations and/or analyses were performed using NRC-approved or industry-accepted analysis methods for the USAR accidents, transients, and special events affected by TPO.
All plant systems and components affected by an increased thermal power level were reviewed to ensure that there is no significant increase in challenges to the safety systems. A review was performed to assure that the increased thermal power level continues to comply with the existing plant environmental regulations. An assessment, as defined in 10 CFR 50.92(C), was performed to establish that no significant hazards consideration exists as a result of operation at the increased power level.
Evaluations and reviews of the NSSS systems and components, containment structures, and BOP systems and components show continued compliance to the codes and standards applicable to the current plant licensing basis (i.e., no change to comply with more recent codes and standards is proposed due to TPO).
A review of the UFSAR and approved design changes ensures adequate evaluation of the licensing basis for the effect of TPO through the date of that evaluation. The plant licensing requirements have been review ed, and it is concluded that this TPO 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 TPO uprate does not involve a significant hazards consideration.
NSSS components and systems were reviewed to confirm that they continue to comply with the functional and regulatory requirements specified in the UFSAR and/or applicable reload license.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-1
Any modification to safety-related or non-safety-related equipment will be implemented in accordance with 10 CFR 50.59.
All plant systems and components affected by an increased thermal power level were reviewed to ensure that there is no significant increase in challenges to the safety systems.
A review was performed to assure that the increased thermal power level continues to comply with the existing plant environmental regulations.
An assessment, as defined in 10 CFR 50.92(C), was performed to establish that no significant hazards consideration exists as a result of operation at the increased power level.
A review of the UFSAR and approved design changes ensures adequate evaluation of the licensing basis for the effect of TPO through the date of that evaluation.
The plant licensing requirements have been reviewed, and it is concluded that this TPO 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 TPO uprate does not involve a significant hazards consideration.
xvii


==1.0 INTRODUCTION==
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION
1.1 OVERVIEW This document addresses a Thermal Power Optim ization (TPO) power uprate of 1.65% of the current licensed thermal power (CLTP), consistent with the magnitude of the thermal power uncertainty reduction for the Limerick Generating Station Units 1 and 2 (Limerick) plant. This will result in an increase in licensed thermal power from 3458 MWt to 3515 MWt and an increase in electrical power from 1219.7 MWe to 1240 MWe. This report follows the Nuclear Regulatory Commission (NRC)-approved format and content for Boiling Water Reactor (BWR) Thermal Power Optimization (TPO) licensing reports documented in NEDC-32938P-A, "Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization" (TLTR) (Reference 1). Power uprates in General Electric Company (GE) BWRs of up to 120% of original licensed thermal power (OLTP) are based on the generic guidelines and approach defined in the Safety Evaluation Reports provided in NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," (ELTR1) (Reference 2) and NEDC-32523P-A, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2) (Reference 3). Since their NRC approval, numerous extended power uprate (EPU) submittals have been based on these reports. The outline for the TPO Safety Analysis Report (TSAR) in TLTR Appendix A follows the same pattern as that used for the EPUs. All of the issues that


should be addressed in a plant-specific TPO lice nsing report are included in this TSAR. For issues that have been evaluate d generically, this report references the appropriate evaluation and establishes that it is applicable to Limerick. BWR plants, as currently licensed, have safety systems and component capability for operation at least 1.5% above the CLTP leve
==1.0 INTRODUCTION==
: l. The amount of power uprate ( 1.5%) contained in the TLTR was based on the expected reduction in power level uncertainty with the instrumentation technology available in 1999. The present instrumentation tec hnology has evolved to where a power level uncertainty is reduced to as lo w as 0.3%, thereby supporting the evaluation of a power level increase of up to 1.7%. Several Pr essurized Water Reactor and BWR plants have already been authorized to increase their thermal power above the OLTP based on a reduction in the uncertainty in the determination of the pow er through improved feedwater (FW) flow rate measurements. When a previous uprate (other than a TPO) has been accomplished, the  102% safety analysis basis is reestab lished above the uprated power leve
: l. Therefore, all GE Hitachi Nuclear Energy (GEH) BWR plant designs have the capability to implement a TPO uprate, whether or not the plant has previously been uprated.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-2 1.2 PURPOSE AND A PPROACH 1.2.1 TPO Analysis Basis Limerick was originally licensed at 3293 MWt. In amendments 106 and 51 for Units 1 and 2, respectively, the NRC approved a five percent power uprate to 3458 MWt, which is the CLTP. The current safety analysis basis assumes, where required, that the re actor had been operating continuously at a power level at least 1.02 times the licensed power level. The analyses performed at 102% of CLTP remain applicable at the TPO rated thermal power (RTP), because the 2% factor from Regulatory Guide (RG) 1.49, "P ower Levels of Nuclear Power Plants," is effectively reduced by the improvement in the FW flow measurements. Some analyses may be performed at TPO RTP, because the uncertainty factor is accounted for in the methods, or the additional 2% margin is not required (e.g., Anticipated Transient Without Scram (ATWS)).
Detailed descriptions of the basis for the TPO analyses are provided in the subsequent sections of this report.
The TPO uprate is based on the evaluation of the improved FW flow rate measurement provided in Section 1.4. Figure 1-1 illustrates the TPO power/flow (P/F) operating map for the analysis at 101.65% of CLTP for Limerick. The changes to the P/F operating map are consistent with the generic descriptions given in TLTR Section 5.2. The approach to achieve a higher thermal power level is to increase core flow along the established Maximum Extended Load Line Limit Analysis (MELLLA) rod lines. This strategy allows Limerick to maintain most of the existing available core flow operational flexibility while assuring that low power related issues (e.g., stability and ATWS instability) do not change because of the TPO uprate. No increase in the previously licensed maximum core flow limit is associated with the TPO uprate. When end of full power reactivity condition (all-rods-out) is reached, end-of-cycle coastdown may be used to exte nd the power generation period. Previously licensed performance improvement features are presented in Section 1.3.2. With respect to absolute thermal power and flow, there is no change in the extent of the Single (Recirculation) Loop Operation (SLO) domain as a result of the TPO uprate. Therefore, the SLO domain is not provided. For Limerick, the maximum analyzed reactor core thermal power for SLO operation remains at 2852.9 MWt. This value bounds the Technical Specification Limit of 2635 MWt.
The TPO uprate is accomplished with no increase in the nominal vessel dome pressure. This minimizes the effect of uprating on reactor thermal duty, evaluations of environmental conditions, and minimizes changes to instrument setpoints related to system pressure, etc. Satisfactory reactor pressure control capability is maintained by evaluating the steam flow margin available at the turbine inlet. This operational aspect of the TPO uprate will be demonstrated by performing controller testing as described in Sec tion 10.4. The TPO uprate does not affect the pressure control function of the turbine by pass valves (TBVs).
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-3 1.2.2 Margins The TPO analysis basis ensures that the power-dependent instrument error margin identified in RG 1.49 is maintained. NRC-approved or industry-accepted computer codes and calculation techniques are used in the safety analyses for the TPO uprate. A list of the NSSS computer codes used in the evaluations is provided in Table 1-1. Computer codes used in previous analyses (i.e., analyses at 102% of CLTP) are not listed. Similarly, factors and margins specified by the application of design code rules are maintained, as are other margin-assuring acceptance criteria used to judge the acceptability of the plant.
1.2.3 Scope of Evaluations The scope of evaluations is discussed in TLTR Appendix B. Tables B-1 through B-3 illustrate those analyses that are bounded by current analyses , those that are not signi ficantly affected, and those that require updating. The disposition of the evaluations as defined by Tables B-1 through B-3 is applicable to Limerick. This TSAR includes all of the evaluations for the plant-specific application. Many of the eval uations are supported by generic reference, some supported by rational considerations of the process differences, and some plant-specific analyses are provided. The scope of the evaluations are summarized in the fo llowing sections:
2.0  Reactor Core and Fuel Performance Overall heat balance and power-flow operatin g map information is provided. Key core performance parameters are confirmed for each fuel cycle, and will continue to be evaluated and documented for each fuel cycle.
3.0 Reactor Coolant and Connected Systems


Evaluations of the NSSS com ponents and systems are performed at the TPO conditions. These evaluations confirm the acceptabil ity of the TPO changes in process variables in the NSSS.
1.1  OVERVIEW This document addresses a Thermal Power Optimization (TPO) power uprate of 1.65% of the current licensed thermal power (CLTP), consistent with the magnitude of the thermal power uncertainty reduction for the Limerick Generating Station Units 1 and 2 (Limerick) plant. This will result in an increase in licensed thermal power from 3458 MWt to 3515 MWt and an increase in electrical power from 1219.7 MWe to 1240 MWe.
4.0  Engineered Safety Features
This report follows the Nuclear Regulatory Commission (NRC)-approved format and content for Boiling Water Reactor (BWR) Thermal Power Optimization (TPO) licensing reports documented in NEDC-32938P-A, Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization (TLTR) (Reference 1). Power uprates in General Electric Company (GE) BWRs of up to 120% of original licensed thermal power (OLTP) are based on the generic guidelines and approach defined in the Safety Evaluation Reports provided in NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate, (ELTR1) (Reference 2) and NEDC-32523P-A, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate, (ELTR2)
(Reference 3). Since their NRC approval, numerous extended power uprate (EPU) submittals have been based on these reports. The outline for the TPO Safety Analysis Report (TSAR) in TLTR Appendix A follows the same pattern as that used for the EPUs. All of the issues that should be addressed in a plant-specific TPO licensing report are included in this TSAR. For issues that have been evaluated generically, this report references the appropriate evaluation and establishes that it is applicable to Limerick.
BWR plants, as currently licensed, have safety systems and component capability for operation at least 1.5% above the CLTP level. The amount of power uprate ( 1.5%) contained in the TLTR was based on the expected reduction in power level uncertainty with the instrumentation technology available in 1999. The present instrumentation technology has evolved to where a power level uncertainty is reduced to as low as 0.3%, thereby supporting the evaluation of a power level increase of up to 1.7%. Several Pressurized Water Reactor and BWR plants have already been authorized to increase their thermal power above the OLTP based on a reduction in the uncertainty in the determination of the power through improved feedwater (FW) flow rate measurements. When a previous uprate (other than a TPO) has been accomplished, the 102%
safety analysis basis is reestablished above the uprated power level. Therefore, all GE Hitachi Nuclear Energy (GEH) BWR plant designs have the capability to implement a TPO uprate, whether or not the plant has previously been uprated.
1-1


The effects of TPO changes on the containm ent, Emergency Core Cooling Systems (ECCS), Standby Gas Treatment, and other Engineered Safety Features are evaluated for key events. The evaluations include the containment responses during limiting abnormal events, Loss-of-Coolant-Accident (LOCA), and safety relief valve (SRV) containment dynamic loads.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.2    PURPOSE AND APPROACH 1.2.1    TPO Analysis Basis Limerick was originally licensed at 3293 MWt. In amendments 106 and 51 for Units 1 and 2, respectively, the NRC approved a five percent power uprate to 3458 MWt, which is the CLTP.
5.0  Instrumentation and Control The instrumentation and control signal ranges and analytical limits for set points are evaluated to establish the effects of TPO changes in process parameters. If required, analyses are performed to determine the need for setpoint changes for various functions. In general, setpoints are NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-4 changed only to maintain adequate operating margins between plant operating parameters and trip values.
The current safety analysis basis assumes, where required, that the reactor had been operating continuously at a power level at least 1.02 times the licensed power level. The analyses performed at 102% of CLTP remain applicable at the TPO rated thermal power (RTP), because the 2% factor from Regulatory Guide (RG) 1.49, Power Levels of Nuclear Power Plants, is effectively reduced by the improvement in the FW flow measurements. Some analyses may be performed at TPO RTP, because the uncertainty factor is accounted for in the methods, or the additional 2% margin is not required (e.g., Anticipated Transient Without Scram (ATWS)).
6.0  Electrical Power and Auxiliary Systems Evaluations are performed to establish the operational capability of the plant electrical power and distribution systems and auxiliary sy stems to ensure that they are capable of supporting safe plant operation at the TPO RTP level.
Detailed descriptions of the basis for the TPO analyses are provided in the subsequent sections of this report.
7.0  Power Conversion Systems Evaluations are performed to establish the operational capability of various (non-safety) balance-of-plant (BOP) systems and components to ensure th at they are capable of delivering the increased TPO power output.  
The TPO uprate is based on the evaluation of the improved FW flow rate measurement provided in Section 1.4. Figure 1-1 illustrates the TPO power/flow (P/F) operating map for the analysis at 101.65% of CLTP for Limerick. The changes to the P/F operating map are consistent with the generic descriptions given in TLTR Section 5.2. The approach to achieve a higher thermal power level is to increase core flow along the established Maximum Extended Load Line Limit Analysis (MELLLA) rod lines. This strategy allows Limerick to maintain most of the existing available core flow operational flexibility while assuring that low power related issues (e.g.,
stability and ATWS instability) do not change because of the TPO uprate.
No increase in the previously licensed maximum core flow limit is associated with the TPO uprate. When end of full power reactivity condition (all-rods-out) is reached, end-of-cycle coastdown may be used to extend the power generation period. Previously licensed performance improvement features are presented in Section 1.3.2.
With respect to absolute thermal power and flow, there is no change in the extent of the Single (Recirculation) Loop Operation (SLO) domain as a result of the TPO uprate. Therefore, the SLO domain is not provided. For Limerick, the maximum analyzed reactor core thermal power for SLO operation remains at 2852.9 MWt. This value bounds the Technical Specification Limit of 2635 MWt.
The TPO uprate is accomplished with no increase in the nominal vessel dome pressure. This minimizes the effect of uprating on reactor thermal duty, evaluations of environmental conditions, and minimizes changes to instrument setpoints related to system pressure, etc.
Satisfactory reactor pressure control capability is maintained by evaluating the steam flow margin available at the turbine inlet. This operational aspect of the TPO uprate will be demonstrated by performing controller testing as described in Section 10.4. The TPO uprate does not affect the pressure control function of the turbine bypass valves (TBVs).
1-2


8.0 Radwaste and Radiation Sources
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.2.2    Margins The TPO analysis basis ensures that the power-dependent instrument error margin identified in RG 1.49 is maintained. NRC-approved or industry-accepted computer codes and calculation techniques are used in the safety analyses for the TPO uprate. A list of the NSSS computer codes used in the evaluations is provided in Table 1-1. Computer codes used in previous analyses (i.e., analyses at 102% of CLTP) are not listed. Similarly, factors and margins specified by the application of design code rules are maintained, as are other margin-assuring acceptance criteria used to judge the acceptability of the plant.
1.2.3    Scope of Evaluations The scope of evaluations is discussed in TLTR Appendix B. Tables B-1 through B-3 illustrate those analyses that are bounded by current analyses, those that are not significantly affected, and those that require updating. The disposition of the evaluations as defined by Tables B-1 through B-3 is applicable to Limerick. This TSAR includes all of the evaluations for the plant-specific application. Many of the evaluations are supported by generic reference, some supported by rational considerations of the process differences, and some plant-specific analyses are provided.
The scope of the evaluations are summarized in the following sections:
2.0      Reactor Core and Fuel Performance Overall heat balance and power-flow operating map information is provided. Key core performance parameters are confirmed for each fuel cycle, and will continue to be evaluated and documented for each fuel cycle.
3.0     Reactor Coolant and Connected Systems Evaluations of the NSSS components and systems are performed at the TPO conditions. These evaluations confirm the acceptability of the TPO changes in process variables in the NSSS.
4.0      Engineered Safety Features The effects of TPO changes on the containment, Emergency Core Cooling Systems (ECCS),
Standby Gas Treatment, and other Engineered Safety Features are evaluated for key events. The evaluations include the containment responses during limiting abnormal events, Loss-of-Coolant-Accident (LOCA), and safety relief valve (SRV) containment dynamic loads.
5.0      Instrumentation and Control The instrumentation and control signal ranges and analytical limits for setpoints are evaluated to establish the effects of TPO changes in process parameters. If required, analyses are performed to determine the need for setpoint changes for various functions. In general, setpoints are 1-3


The liquid and gaseous waste management systems are evaluated at TPO conditions to show that applicable release limits continue to be met during operation at the TPO RTP level. The radiological consequences are evaluated to show that applicable regulations are met for TPO including the effect on source terms, on-site doses, and off-site doses during normal operation.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION changed only to maintain adequate operating margins between plant operating parameters and trip values.
9.0 Reactor Safety Performance Evaluations  
6.0    Electrical Power and Auxiliary Systems Evaluations are 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 TPO RTP level.
7.0    Power Conversion Systems Evaluations are 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 TPO power output.
8.0    Radwaste and Radiation Sources The liquid and gaseous waste management systems are evaluated at TPO conditions to show that applicable release limits continue to be met during operation at the TPO RTP level. The radiological consequences are evaluated to show that applicable regulations are met for TPO including the effect on source terms, on-site doses, and off-site doses during normal operation.
9.0     Reactor Safety Performance Evaluations
((
                                              )) The standard reload analyses consider the plant conditions for the cycle of interest.
10.0    Other Evaluations High energy line break (HELB) and environmental qualification evaluations are performed at bounding conditions for the TPO range to show the continued operability of plant equipment under TPO conditions. The Individual Plant Examination (IPE) Probabilistic Risk Assessment (PRA) will not be updated, because the change in plant risk from the subject power uprate is insignificant. This conclusion is supported by NRC Regulatory Issue Summary (RIS) 2002-03 (Reference 4). In response to feedback received during the public workshop held on August 23, 2001, the Staff wrote, The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant effect on plant risk. Therefore, no risk information is requested to support such applications.
1-4


[[                                                                                                                                                                               
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.2.4  Exceptions to the TLTR One safety-related modification to the Standby Liquid Control System (SLCS) is proposed in this amendment request. Although the TLTR Section B.2, Licensing Criteria, states that no safety-related modifications are needed beyond potential setpoint changes, the need for this modification is discussed in separate documentation and in Section 9.3.1.5 of this report.
1.2.5  Concurrent Changes Unrelated to TPO No concurrent changes unrelated to TPO are included in this evaluation.
1.3  TPO PLANT OPERATING CONDITIONS 1.3.1  Reactor Heat Balance The reactor heat balance diagram at the TPO conditions is presented in Figure 1-2.
The small changes in thermal-hydraulic parameters for the TPO are illustrated in Figure 1-2 (3515 MWt, 100% Core Flow). These parameters are generated for TPO by performing coordinated reactor and turbine-generator heat balances that relate the reactor thermal-hydraulic parameters to the increased plant FW and steam flow conditions. Input from Limerick operation is considered (e.g., steam line pressure drop) to match expected TPO uprate conditions.
1.3.2  Reactor Performance Improvement Features The following performance improvement and equipment out-of-service (EOOS) features currently licensed at Limerick are acceptable at the TPO RTP level. Their inclusion in this analysis is appropriate, as they have been previously adopted by Limerick and incorporated in the limiting case of the initial power uprate analysis.
Performance Improvement Feature SLO Increased Core Flow (ICF) (110.0% of rated)
Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program (ARTS) / MELLLA (82.9% of Rated Core Flow at TPO RTP)
Final Feedwater Temperature Reduction (FFWTR), -105ºF Feedwater Heater (s) OOS, -60ºF SRV OOS, two valves 24 Month Cycle Recirculation Pump Trip (RPT) OOS TBV OOS Main Steam Isolation Valves (MSIV) OOS, 75% Power Turbine Control Valve (TCV) Stuck Closed Turbine Stop Valve (TSV) Stuck Closed 1-5


                                                                          ]]  The standard reload analyses consider the plant conditions for the cycle of interest.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.4   BASIS FOR TPO UPRATE The safety analyses in this report are based on a total thermal power measurement uncertainty of 0.3%. This will bound the actual power level requested. The detailed basis value is provided in separate documentation, which addresses the improved FW flow measurement accuracy using the Caldon Leading Edge Flow Meter Check-Plus system.
10.0  Other Evaluations High energy line break (HELB) and environm ental qualification evaluations are performed at bounding conditions for the TPO range to show the continued operability of plant equipment under TPO conditions. The Individual Plant Examination (IPE) Probabilistic Risk Assessment (PRA) will not be updated, because the change in plant risk from the subject power uprate is insignificant. This conclusion is supporte d by NRC Regulatory Issue Summary (RIS) 2002-03 (Reference 4). In response to feedback rece ived during the public workshop held on August 23, 2001, the Staff wrote, "The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant effect on plant risk. Therefore, no risk information is requested to support such applications."
1.5  
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-5 1.2.4 Exceptions to the TLTR One safety-related m odification to the Standby Liquid Control System (S LCS) is proposed in this amendment request. Although the TLTR Sec tion B.2, "Licensing Crite ria," states that no safety-related modifications are needed beyond pot ential setpoint changes, the need for this modification is discussed in separate documentation and in Section 9.3.1.5 of this report.
1.2.5 Concurrent Changes Unrelated to TPO No concurrent changes unrelated to TPO are included in this evaluation.
1.3 TPO PLANT OPERATING CONDITIONS 1.3.1 Reactor Heat Balance The reactor heat balance diagram at the TPO conditions is presented in Figure 1-2. The small changes in thermal-hydraulic paramete rs for the TPO are illu strated in Figure 1-2 (3515 MWt, 100% Core Flow). These parameters are generated for TPO by performing coordinated reactor and turbine-ge nerator heat balances that relate the reactor thermal-hydraulic parameters to the increased plant FW and steam flow conditions. Input from Limerick operation is considered (e.g., steam line pressure drop) to match expected TPO uprate conditions.
1.3.2 Reactor Performance Im provement Features The following performance improvement and equipment out-of-service (EOOS) features currently licensed at Limerick are acceptable at the TPO RTP level. Their inclusion in this analysis is appropriate, as they have been previously adopted by Limerick and incorporated in the limiting case of the initial power uprate analysis. Performance Improvement Feature SLO Increased Core Flow (ICF) (110.0% of rated) Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program (ARTS) / MELLLA (82.9% of Rated Core Flow at TPO RTP) Final Feedwater Temperature Reduction (FFWTR), -105ºF Feedwater Heater (s) OOS, -60ºF SRV OOS, two valves 24 Month Cycle Recirculation Pump Trip (RPT) OOS TBV OOS Main Steam Isolation Valves (MSIV) OOS, 75% Power Turbine Control Valve (TCV) Stuck Closed Turbine Stop Valve (TSV) Stuck Closed NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-6 1.4 B ASIS FOR TPO UPRATE The safety analyses in this report are based on a total thermal power measurement uncertainty of 0.3%. This will bound the actual power level requested. The detailed basis value is provided in separate documentation, which addresses the improved FW flow measurement accuracy using the Caldon Leading Edge Flow Meter Check-Plus system. 1.5  


==SUMMARY==
==SUMMARY==
AND CONCLUSIONS This evaluation has investigated a TPO uprate to 101.65% of CLTP. The strategy for achieving higher power is to increase core flow along the established MELLLA rod lines. The plant licensing challenges have been reviewed (Table 1-3) to demonstrate how the TPO 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 TPO uprate descri bed herein involves no signifi cant hazards consideration.
AND CONCLUSIONS This evaluation has investigated a TPO uprate to 101.65% of CLTP. The strategy for achieving higher power is to increase core flow along the established MELLLA rod lines. The plant licensing challenges have been reviewed (Table 1-3) to demonstrate how the TPO 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 TPO uprate described herein involves no significant hazards consideration.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-7 Table 1-1 Computer Codes For TPO Analyses Task Computer Code Version or Revision NRC Approved Comments Reactor Heat Balance ISCOR 09 Y (1) NEDE-24011-P Rev 0 Safety Evaluation Report (SER) Thermal-hydraulic Stability ISCOR PANACEA ODYSY OPRM TRACG 09 11 05 01 04 Y (1) Y Y Y(3)
1-6
N(4) NEDE-24011P Rev. 0 SER NEDE-30130-P-A (2) NEDC-32992P-A NEDO-32465-A NEDO-32465-A Reactor Internal Pressure


Differences ISCOR 09 Y (1) NEDE-24011P Rev. 0 SER Anticipated Transient Without Scram  ODYN STEMP PANACEA ISCOR 10 04 11 09 Y (5) Y (2)
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-1 Computer Codes For TPO Analyses Computer    Version or        NRC Task                                                                                  Comments Code      Revision        Approved Reactor Heat Balance                    ISCOR          09            Y (1)      NEDE-24011-P Rev 0 Safety Evaluation Report (SER)
Y (1) NEDE-24154P-A Supp. 1, Vol. 4  
Thermal-hydraulic Stability              ISCOR         09             Y (1)       NEDE-24011P Rev. 0 SER PANACEA           11               Y         NEDE-30130-P-A (2)
 
ODYSY          05              Y        NEDC-32992P-A OPRM          01            Y(3)        NEDO-32465-A TRACG          04            N(4)       NEDO-32465-A Reactor Internal Pressure                ISCOR          09            Y (1)       NEDE-24011P Rev. 0 SER Differences Anticipated Transient Without            ODYN          10              Y        NEDE-24154P-A Supp. 1, Vol. 4 Scram                                  STEMP          04              (5)        NEDE-30130-P-A PANACEA          11            Y (2)
NEDE-30130-P-A NEDE-24011-P Revision 0 SER Notes For Table 1-1: (1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P 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 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. (2) The physics code PANACEA provides inputs to the transient code ODYN. The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-24011-P-A). The use of TGBLA Version 06 and PANA CEA Version 11 in this a pplication was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (N RC) to G.A. Watford (GE)  
ISCOR          09            Y (1)      NEDE-24011-P Revision 0 SER Notes For Table 1-1:
(1)   The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P 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 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.
(2)   The physics code PANACEA provides inputs to the transient code ODYN. The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-24011-P-A). The use of TGBLA Version 06 and PANACEA Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE)


==Subject:==
==Subject:==
  "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO. MA6481), November 10, 1999. (3) The OPRM code is not Level 2. However, the methodology, as implemented in the OPRM code, has been approved by the NRC. (4) TRACG02 has been approved in NEDO-32465-A by the US NRC for the stability DIVOM analysis. The CLTP stability analysis is based on TRACG04, which has been shown to provide essentially the same or more conservative results in DIVOM applications as the previous version, TRACG02. (5) 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.
  "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO. MA6481), November 10, 1999.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-8 Table 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions Parameter CLTP TPO RTP (101.65% of CLTP) Thermal Power (MWt)  (Percent of Current Licensed Power) 3458.0 100.0 3515.0 101.65 Steam Flow (Mlb/hr)  (Percent of Current Rated) 14.997 100.0 15.287 101.9 FW Flow (Mlb/hr)
(3)   The OPRM code is not Level 2. However, the methodology, as implemented in the OPRM code, has been approved by the NRC.
(Percent of Current Rated) 14.965 100.0 15.255 101.9 Dome Pressure (psia) 1060 1060 Dome Temperature (°F) 551.5 551.5 FW Temperature (°F) 425.1 427.1 Full Power Core Flow Range (Mlb/hr)
(4)   TRACG02 has been approved in NEDO-32465-A by the US NRC for the stability DIVOM analysis. The CLTP stability analysis is based on TRACG04, which has been shown to provide essentially the same or more conservative results in DIVOM applications as the previous version, TRACG02.
(Percent of Current Rated) 81.0 to 110.0 81.0 to 110.0 82.9 to 110.0 82.9 to 110.0
(5)   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.
1-7


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-9  Table 1-3 Summary of Effect of TPO Uprate on Licensing Criteria Key Licensing Criteria Effect of 1.7% Thermal Power Increase Explanation of Effect LOCA challenges to fuel (10 CFR 50, Appendix K) No increase in peak clad temperature (PCT), no change of maximum Linear Heat Generatio n Rate (LHGR) required. Previous analysis accounted for  102% of licensed power, bounding TPO operation. No vessel pressure increase. Change of Operating Limit Minimum Critical Power Ratio (MCPR) < 0.01 increase. Minor increase (< 0.01) due to slightly higher power density and increased MCPR safety limit (slightly flatter radial power distribution). Challenges to reactor pressure vessel (RPV) overpressure No increase in peak pressure. No increase because previous analysis accounted for  102% overpower, bounding TPO operation. Primary containment pressure during a LOCA No increase in peak containment pressure. Previous analysis accounted for  102% overpower, bounding TPO operation. No vessel pressure increase. No increase in energy to the pool. Pool temperature during a
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions TPO RTP Parameter                        CLTP (101.65% of CLTP)
Thermal Power (MWt)                                 3458.0          3515.0 (Percent of Current Licensed Power)         100.0          101.65 Steam Flow (Mlb/hr)                                 14.997          15.287 (Percent of Current Rated)                  100.0            101.9 FW Flow (Mlb/hr)                                     14.965          15.255 (Percent of Current Rated)                  100.0           101.9 Dome Pressure (psia)                                 1060            1060 Dome Temperature (&deg;F)                                551.5            551.5 FW Temperature (&deg;F)                                   425.1            427.1 Full Power Core Flow Range (Mlb/hr)               81.0 to 110.0    82.9 to 110.0 (Percent of Current Rated)              81.0 to 110.0    82.9 to 110.0 1-8


LOCA No increase in peak pool temperature. Previous analysis accounted for  102% overpower, bounding TPO operation. No vessel pressure increase. No increase in energy to the pool. Offsite Radiation Release, design basis accidents No increase (remains within
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-3 Summary of Effect of TPO Uprate on Licensing Criteria Effect of 1.7%
Key Licensing Criteria                                                              Explanation of Effect Thermal Power Increase LOCA challenges to fuel          No increase in peak clad temperature  Previous analysis accounted for  102% of (10 CFR 50, Appendix K)          (PCT), no change of maximum Linear    licensed power, bounding TPO operation. No Heat Generation Rate (LHGR)            vessel pressure increase.
required.
Change of Operating Limit        < 0.01 increase.                      Minor increase (< 0.01) due to slightly higher Minimum Critical Power Ratio                                            power density and increased MCPR safety limit (MCPR)                                                                  (slightly flatter radial power distribution).
Challenges to reactor pressure  No increase in peak pressure.          No increase because previous analysis accounted vessel (RPV) overpressure                                              for  102% overpower, bounding TPO operation.
Primary containment pressure    No increase in peak containment        Previous analysis accounted for  102%
during a LOCA                   pressure.                              overpower, bounding TPO operation. No vessel pressure increase. No increase in energy to the pool.
Pool temperature during a        No increase in peak pool temperature. Previous analysis accounted for  102%
LOCA                                                                    overpower, bounding TPO operation. No vessel pressure increase. No increase in energy to the pool.
Offsite Radiation Release,       No increase (remains within            Previous analysis bounds TPO operation. No design basis accidents           10 CFR 100).                          vessel pressure increase.
Onsite Radiation Dose, normal    Approximately 1.7% increase, must      Slightly higher inventory of radionuclides in operation                        remain within 10 CFR 20.              steam/FW flow paths.
Heat discharge to environment    < 1&deg;F temperature increase.            Small % power increase.
Equipment Qualification          Remains within current pressure,      No change in Harsh Environment terms (TPO radiation, and temperature envelopes. operating conditions bounded by previous analyses); minimal change in normal operating conditions.
Fracture Toughness,              < 2&deg;F increase in Reference            Small increase in neutron fluence.
10 CFR 50, Appendix G            Temperature of the Nil-Ductility Transition (RTNDT).
Stability                        No direct effect of TPO uprate because No increase in maximum rod line boundary.
applicable stability regions and lines Characteristics of each reload core continue to be are extended beyond the absolute      evaluated as required for each stability option.
values associated with the current boundaries to preserve MWt-core flow boundaries as applicable for each stability option.
ATWS peak vessel pressure        Slight increase (15 psig).            Slightly increased power relative to SRV capacity.
Vessel and NSSS equipment        No change.                            Comply with existing ASME Code stress limits of design pressure                                                        all categories.
1-9


10 CFR 100). Previous analysis bounds TPO operation. No vessel pressure increase. Onsite Radiation Dose, normal operation Approximately 1.7% increase, must remain within 10 CFR 20. Slightly higher inventory of radionuclides in steam/FW flow paths. Heat discharge to environment < 1&deg;F temperature increase. Small % power increase. Equipment Qualification Remains within current pressure, radiation, and temperature envelopes. No change in Harsh E nvironment terms (TPO operating conditions bounded by previous analyses); minimal change in normal operating conditions. Fracture Toughness, 10 CFR 50, Appendix G < 2&deg;F increase in Reference Temperature of the Nil-Ductility Transition (RTNDT). Small increase in neutron fluence. Stability No direct effect of TPO uprate because applicable stability regions and lines are extended beyond the absolute values associated with the current boundaries to preserve MWt-core flow boundaries as applicable for each stability option. No increase in maximum rod line boundary. Characteristics of each reload core continue to be evaluated as required for each stability option. ATWS peak vessel pressure Slight increase (15 psig). Slightly increased power relative to SRV capacity. Vessel and NSSS equipment design pressure No change. Comply with existing ASME Code stress limits of all categories.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-1 Power/Flow Map for TPO (101.65% of CLTP)
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-10 Figure 1-1 Power/Flow Map for TPO (101.65% of CLTP) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 0 10 20 30 405060708090100 110120Core Flow (%)Thermal Power (%TPO RTP) 0 400 800 1200 1600 2000 2400 2800 3200 3600 4000 4400 0 10 20 30 405060708090100 110120Core Flow (Mlbm/hr)Thermal Power (MWt) 1 00.0% TPO RTP       =  3515 MWt 1 00.0% CLTP          =  3458 MWt 1 00.0% Core Flow    = 100.0 Mlbm/hr A: Natural Circulation B: Two Pump Minimum Speed C: 67.6% Power / 44.4% Flow D: 100.0% Power / 82.9% Flow D': 98.4% Power / 80.8% Flow E: 100.0% Power / 100.0% Flow E': 98.4% Power / 100.0% Flow F: 100.0% Power / 110.0% Flow F': 98.4% Power / 110.0% Flow G: 23.6% Power / 110.0% Flow H: 23.6% Power / 100.0% Flow I: 23.6% Power / 38.0% Flow MELLLA BoundaryP=(22.191+0.89714W T-0.0011905W T 2)(1.132)A D E F D'E' F' 3515 MWt3458 MWt G I Cavitation InterlockIncreased Core Flow B C H NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1-11 Figure 1-2 Reactor Heat Balance - TPO Power (101.65% of CLTP), 100% Core Flow 1060 P15.287E+06# *1190.0H *0.43M *Carryunder =0.25%1003P *3515MWtWd = 100%15.388E+06#15.255E+06#531.5H405.5H405.3H536.0&deg;F427.2&deg;F427.1&deg;F100.0E+06h = 0.9  H
Core Flow (Mlbm/hr) 0       10       20       30           40         50         60           70             80       90   100         110       120 130 4400 120       A: Natural Circulation B: Two Pump Minimum Speed C: 67.6% Power / 44.4% Flow                                                                                                       4000 110      D: 100.0% Power / 82.9% Flow D': 98.4% Power / 80.8% Flow E: 100.0% Power / 100.0% Flow                                                                 D        E          F    3515 MWt  3600 100      E': 98.4% Power / 100.0% Flow                 MELLLA Boundary F: 100.0% Power / 110.0% Flow                                       2                                                  3458 MWt P=(22.191+0.89714WT -0.0011905WT )(1.132)                D'        E'          F' F': 98.4% Power / 110.0% Flow 90 G: 23.6% Power / 110.0% Flow                                                                                                       3200 H: 23.6% Power / 100.0% Flow Thermal Power (MWt)
#1.330E+05#418.4H530.7439.0&deg;F H3.200E+04#1.330E+05#68.0H530.6H97.2&deg;F535.3&deg;F*Conditions at upstream side of TSVCore Thermal Power3515.0Pump Heating9.0Cleanup Losses-4.4Other System Losses-1.1Turbine Cycle Use3518.5MWtCleanupDemineralizerSystemMain Steam FlowMain Feed FlowControl Rod Drive Feed FlowTotal Core Flow# = Flow, lbm/hrH = Enthalpy, Btu/lbmF = Temperature, &deg;FM = Moisture, %P = Pressure, psiaLegend NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-1 2.0  REACTOR CORE AND FUEL PERFORMANCE 2.1 F UEL DESIGN AND OPERATION At the TPO RTP conditions, all fuel and core design limits are met by the deployment of fuel enrichment and burnable poison, control rod pattern management, and core flow adjustments.
I: 23.6% Power / 38.0% Flow Thermal Power (%TPO RTP) 80                                                                                                                                          2800 70                                                  C 2400 Increased Core Flow 60 2000 B
New fuel designs are not needed for the TPO to ensure safety. However, revised loading patterns, slightly larger batch sizes, and potentially new fuel designs may be used to provide additional operating flexibility and maintain fuel cycle length. NRC approved limits for burnup on the fuel are not exceeded. Therefore, the re actor core and fuel design is adequate for TPO operation. The initial TPO cycle at Limerick Unit 1 will be loaded with fresh and previously irradiated GE14 fuel assemblies. The initial TPO cycle at Limerick Unit 2 will be loaded with fresh GNF2 fuel assemblies and previously irradiated GE14 fuel assemblies. 2.2 THERMAL LIMITS ASSESSMENT Operating thermal limits ensure that regulatory and/or safety limits are not exceeded for a range of postulated events (e.g., transients, LOCA). This section addresses the effects of TPO on thermal limits. Cycle-specific core configurations, which are evaluated for each reload, confirm TPO RTP capability and establish or confirm cycle-specific limits. The historical 25% of RTP value for the Technical Specification Safety Limit, some thermal limits monitoring Limiting Conditions for Operation (LCOs) thresholds, and some Surveillance Requirements (SRs) thresholds is based on [[                                                                                               
50 A                                                                                            1600 40 1200 30 Cavitation Interlock 800 20                                        I                                                                          H           G 100.0% TPO RTP      = 3515 MWt 10                                                                                          100.0% CLTP         = 3458 MWt                  400 100.0% Core Flow   = 100.0 Mlbm/hr 0                                                                                                                                          0 0       10      20        30            40          50        60          70            80        90  100          110        120 Core Flow (%)
1-10


                                                            ]]  The historical 25% RTP value is a conservative basis, as described in the plant Te chnical Specifications, [[                                                                                       
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-2 Reactor Heat Balance - TPO Power (101.65% of CLTP), 100% Core Flow Legend
# = Flow, lbm/hr                    1060 H = Enthalpy, Btu/lbm                P F = Temperature, &deg;F M = Moisture, %                                          Main Steam Flow        15.287E+06 #
* P = Pressure, psia 1190.0 H
* 0.43 M
* Carryunder = 0.25%                1003 P
* 3515                          Main Feed Flow MWt Wd = 100 %                                      15.388E+06 #            15.255E+06 #
531.5 H                                          405.5 H                  405.3 H 536.0 &deg;F                Total                    427.2 &deg;F                427.1 &deg;F Core Flow 100.0E+06 h = 0.9 H                          #
1.330E+05 #
418.4 H 530.7                                              439.0 &deg;F H
Cleanup Demineralizer System 3.200E+04 #          Control Rod Drive                      1.330E+05 #
68.0 H              Feed Flow                              530.6 H 97.2 &deg;F                                                    535.3 &deg;F
                                                    *Conditions at upstream side of TSV Core Thermal Power                    3515.0 Pump Heating                              9.0 Cleanup Losses                          -4.4 Other System Losses                      -1.1 Turbine Cycle Use                    3518.5 MWt 1-11


                                                                                                                                                                                                                                                        ]]  Therefore, the Safety Limit percent RTP basis, some thermal limits monitoring LCOs, and SR percent RTP thresholds remain at 25% RTP for the TPO uprate.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2.0 REACTOR CORE AND FUEL PERFORMANCE 2.1    FUEL DESIGN AND OPERATION At the TPO RTP conditions, all fuel and core design limits are met by the deployment of fuel enrichment and burnable poison, control rod pattern management, and core flow adjustments.
New fuel designs are not needed for the TPO to ensure safety. However, revised loading patterns, slightly larger batch sizes, and potentially new fuel designs may be used to provide additional operating flexibility and maintain fuel cycle length. NRC approved limits for burnup on the fuel are not exceeded. Therefore, the reactor core and fuel design is adequate for TPO operation.
The initial TPO cycle at Limerick Unit 1 will be loaded with fresh and previously irradiated GE14 fuel assemblies. The initial TPO cycle at Limerick Unit 2 will be loaded with fresh GNF2 fuel assemblies and previously irradiated GE14 fuel assemblies.
2.2    THERMAL LIMITS ASSESSMENT Operating thermal limits ensure that regulatory and/or safety limits are not exceeded for a range of postulated events (e.g., transients, LOCA). This section addresses the effects of TPO on thermal limits. Cycle-specific core configurations, which are evaluated for each reload, confirm TPO RTP capability and establish or confirm cycle-specific limits.
The historical 25% of RTP value for the Technical Specification Safety Limit, some thermal limits monitoring Limiting Conditions for Operation (LCOs) thresholds, and some Surveillance Requirements (SRs) thresholds is based on ((
                                      )) The historical 25% RTP value is a conservative basis, as described in the plant Technical Specifications, ((
                                                                                  )) Therefore, the Safety Limit percent RTP basis, some thermal limits monitoring LCOs, and SR percent RTP thresholds remain at 25% RTP for the TPO uprate.
2-1


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-2 2.2.1 Safety Limit MCPR The Safety Limit Minimum Critical Power Ratio (SLMCPR) is dependent upon the nominal average power level and the uncertainty in its measurement. Consistent with approved practice, a revised SLMCPR is calculated for the first TPO fuel cycle and confirmed for each subsequent cycle. The historical uncerta inty allowance and calculational methods are discussed in TLTR Section 5.7.2.1. 2.2.2 MCPR Operating Limit TLTR Appe ndix E shows that the changes in the Operating Limit Minimum Critical Power Ratio (OLMCPR) for a TPO uprate [[                                                                                                                       
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2.2.1   Safety Limit MCPR The Safety Limit Minimum Critical Power Ratio (SLMCPR) is dependent upon the nominal average power level and the uncertainty in its measurement. Consistent with approved practice, a revised SLMCPR is calculated for the first TPO fuel cycle and confirmed for each subsequent cycle. The historical uncertainty allowance and calculational methods are discussed in TLTR Section 5.7.2.1.
2.2.2   MCPR Operating Limit TLTR Appendix E shows that the changes in the Operating Limit Minimum Critical Power Ratio (OLMCPR) for a TPO uprate ((
                                                                    )) Because the cycle-specific SLMCPR is also defined, the actual required OLMCPR can be established. This ensures an adequate fuel thermal margin for TPO uprate operation.
2.2.3  MAPLHGR and Maximum LHGR Operating Limits The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and maximum Linear Heat Generation Rate (LHGR) limits are maintained as described in TLTR Section 5.7.2.2. No significant change results due to TPO operation. The LHGR limits are fuel dependent and are not affected by the TPO. The ECCS performance is addressed in Section 4.3.
2.3  REACTIVITY CHARACTERISTICS All minimum shutdown margin requirements apply to cold shutdown conditions and are maintained without change. Checks of cold shutdown margin based on SLCS boron injection capability and shutdown using control rods with the most reactive control rod stuck out are made for each reload. The TPO uprate has no significant effect on these conditions; the shutdown margin is confirmed in the reload core design.
Operation at the TPO RTP could result in a minor decrease in the hot excess reactivity during the cycle. This loss of reactivity does not affect safety and does not affect the ability to manage the power distribution through the cycle to achieve the target power level. However, the lower hot excess reactivity can result in achieving an earlier all-rods-out condition. Through fuel cycle redesign, sufficient excess reactivity can be obtained to match the desired cycle length.
2.4  THERMAL HYDRAULIC STABILITY 2.4.1  Stability Option III Limerick Units 1 and 2 have implemented the long-term stability solution Option III (References 5 and 6). The Option III solution combines closely spaced Local Power Range Monitor (LPRM) detectors into cells to effectively detect either core-wide or regional (local) 2-2


                                                                                                                                ]]  Because the cycle-specific SLMCPR is also defined, the actual required OLMCPR can be established. This ensures an adequate fuel thermal margin for TPO uprate operation.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION modes of reactor instability. These cells are termed Oscillation Power Range Monitor (OPRM) cells and are configured to provide local area coverage with multiple channels. Plants implementing Option III have hardware to combine the LPRM signals and to evaluate the cell signals with instability detection algorithms. The Period Based Detection Algorithm (PBDA) is the only algorithm credited in the Option III licensing basis (Reference 6). Two defense-in-depth algorithms, referred to as Amplitude Based Algorithm (ABA) and the Growth Rate Algorithm (GRA), offer a high degree of assurance that fuel failure will not occur as a consequence of stability related oscillations. Because the OPRM hardware does not change, the hot channel oscillation magnitude (HCOM) portion of the Option III calculation (Reference 7) is not affected by TPO and does not need to be recalculated.
2.2.3 MAPLHGR and Maximum LHGR Operating Limits The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and maximum Linear Heat Generation Rate (LHGR) limits are maintained as described in TLTR Section 5.7.2.2. No significant change results due to TPO operation. The LHGR limits are fuel dependent and are not affected by the TPO. The ECCS performance is addressed in Section 4.3. 2.3 REACTIVITY CHARACTERISTICS All minimum shutdown margin requirements apply to cold shutdown conditions and are maintained without change. Checks of cold shutdown margin based on SLCS boron injection capability and shutdown using control rods with the most reactive control rod stuck out are made for each reload. The TPO uprate has no significant effect on these conditions; the shutdown margin is confirmed in the reload core design. Operation at the TPO RTP could result in a minor decrease in the hot excess reactivity during the cycle. This loss of reactivity does not affect safety and does not affect the ability to manage the power distribution through the cycle to achieve the target power level. However, the lower hot excess reactivity can result in achieving an earlier all-rods-out condition. Through fuel cycle redesign, sufficient excess reactivity can be obtained to match the desired cycle length. 2.4 THERMAL HYDRAULIC STABILITY  2.4.1 Stability Option III Limerick Units 1 and 2 have implemented the long-term stability solution Option III (References 5 and 6). The Option III solution combines closely spaced Local Power Range Monitor (LPRM) detectors into "cells" to effectiv ely detect either core-w ide or regional (local)
The Option III Trip-Enabled Region has been generically defined as the region ( 60% rated core flow and  30% rated power) where the OPRM system is fully armed. For TPO, the Option III Trip-Enabled Region is rescaled to maintain the same absolute P/F region boundaries. The Backup Stability Protection (BSP) evaluation described in Section 2.4.2 shows that the generic Option III Trip-Enabled Region is adequate. The Trip-Enabled Region is shown in Figure 2-1.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-3 modes of reactor instability. These cells are termed Oscillation Power Range Monitor (OPRM) cells and are configured to provide local area coverage with multiple channels. Plants implementing Option III have hardware to combine the LPRM signals and to evaluate the cell signals with instability detection algorithms. The Period Based Detection Algorithm (PBDA) is the only algorithm credited in the Option III licensing basis (Reference 6). Two defense-in-depth algorithms, referred to as Amplitude Based Algorithm (ABA) and the Growth Rate Algorithm (GRA), offer a high degree of assurance that fuel failure will not occur as a consequence of stability related oscillations. Because the OPRM hardware does not change, the hot channel oscillation magnitude (HCOM) portion of the Option III calculation (Reference 7) is not affected by TPO and does not need to be recalculated. The Option III Trip-Enabled Region has been generically defined as the region ( 60% rated core flow and  30% rated power) where the OPRM system is fully armed. For TPO, the Option III Trip-Enabled Region is rescaled to maintain the same absolute P/F region boundaries. The Backup Stability Protection (BSP) evaluation described in Secti on 2.4.2 shows that the generic Option III Trip-Enabled Region is adequate. Th e Trip-Enabled Region is shown in Figure 2-1.
Because the rated core flow is not changed, the 60% recirculation drive flow boundary is not rescaled (It should be noted that 60% recirculation drive flow bounds 60% core flow). The 30%
Because the rated core flow is not changed, th e 60% recirculation drive flow boundary is not rescaled (It should be noted that 60% recirculat ion drive flow bounds 60% core flow). The 30%
of CLTP boundary changes by the following equation:
of CLTP boundary changes by the following equation:
TPO Region Boundary = 30% CLTP * (100% / TPO (% CLTP))  
TPO Region Boundary = 30% CLTP * (100% / TPO (% CLTP))
 
Thus, for a 101.65% of CLTP TPO:
Thus, for a 101.65% of CLTP TPO:
TPO Region boundary = 30% CLTP
TPO Region boundary = 30% CLTP * (100% / 101.65%) = 29.5% TPO The minimum power at which the OPRM should be confirmed operable is 24.5% TPO. A 5%
* (100% / 101.65%) = 29.5% TPO The minimum power at which the OPRM should be confirmed operable is 24.5% TPO. A 5%
absolute power separation between the OPRM Trip-Enabled Region power boundary and the power, at which the OPRM system should be confirmed operable, is deemed adequate for the Option III application.
absolute power separation between the OPRM Trip-Enabled Region power boundary and the power, at which the OPRM system should be confirmed operable, is deemed adequate for the Option III application.
Stability Option III provides SLMCPR protection by generating a reactor scram if a reactor instability, which exceeds the specifi ed trip setpoint, is detected. The demonstration setpoint is determined per the current NRC approved methodology. The Option III stability reload licensing basis calculates the limiting OLMCPR required to protect the SLMCPR for both steady-state and transient stab ility events as specified in the Option III methodology (Reference 6). These OLMCPRs are calculated for a range of OPRM setpoints for TPO operation. Selection of an appropriate instru ment setpoint is then based upon the OLMCPR required to provide adequate SLMCPR protection. This determination relies on the DIVOM curve (Delta CPR Over Initial MCPR Versus Oscillation Magnitude) to determine an OPRM setpoint that protects the SLMCPR during an anticipated instability event. A DIVOM analysis is performed and used in Option III OPRM setpoint demonstration.
Stability Option III provides SLMCPR protection by generating a reactor scram if a reactor instability, which exceeds the specified trip setpoint, is detected. The demonstration setpoint is determined per the current NRC approved methodology. The Option III stability reload licensing basis calculates the limiting OLMCPR required to protect the SLMCPR for both steady-state and transient stability events as specified in the Option III methodology (Reference 6). These OLMCPRs are calculated for a range of OPRM setpoints for TPO operation. Selection of an appropriate instrument setpoint is then based upon the OLMCPR required to provide adequate SLMCPR protection. This determination relies on the DIVOM curve (Delta CPR Over Initial MCPR Versus Oscillation Magnitude) to determine an OPRM setpoint that protects the SLMCPR during an anticipated instability event. A DIVOM analysis is performed and used in Option III OPRM setpoint demonstration.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-4 As shown in Table 2-1, with an estimated OLMCPR of 1.34 and an estimated SLMCPR of 1.07, an OPRM trip setpoint of 1.14 with a successive confirmation count of 16 (Reference 6) is the highest setpoint that may be used without stabi lity setting the OLMCPR.
2-3
The actual setpoint will be established in accordance with Limerick Units 1 and 2 Technical Specifications at each reload. These demonstration results are base d on a power level of 101.7% CLTP, which is bounding for a power level of 101.65%. Therefore, TPO operation is justified for pl ant operation with stability Option III. 2.4.2 Stability Backup Stability Protection Limerick Units 1 and 2 have implemented th e Backup Stability Prot ection (BSP) regions (Reference 8) as the stability backup solution if the OPRM system is declared inoperable. The BSP regions consist of two regions, I-Scram and II-Controlled Entry. The Base BSP Scram Region and Base BSP Controlled Entry Region are defined by state points on the High Flow Control Line (HFCL) and on the Natural Circulation Line (NCL) in accordance with Reference 8. The bounding plan t-specific BSP region state points must enclose the corresponding Base BSP region state points on the HFCL and on the NCL. If a calculated BSP region state point is located insi de the corresponding base BSP region state point, then it must be replaced by the corresponding base BSP region state point. If a ca lculated BSP region state point is located outside the correspondin g base BSP region state point, this point is acceptable for use. That is, the selected points will result in the largest, or most conservative, region sizes. The proposed BSP Scram and Controlled Entry regi on boundaries are constructed by connecting the corresponding bounding state points on the HFCL and the NCL using a shape function. The Modified Shape Function (MSF) is demonstrated. The demonstration BSP regions for both Nominal Feedwater Temperature (NFWT) and Minimum Feedwater Temperature (MFWT) operati ons are shown in Tabl e 2-2/Figure 2-2 and Table 2-3/Figure 2-3, respectively. The OPRM Trip-Enabled Region is confirmed for NFWT operations based on the demonstration BSP regions for NFWT. The demonstration BSP regions for MFWT confirm the OPRM Trip-Enabled Region for operations on or below the 97% rod line. These demonstration results are base d on a power level of 101.7% CLTP, which is bounding for a power level of 101.65%. The BSP regions are confirmed or e xpanded on a cycle-specific basis. Therefore, TPO operation is justified for pl ant operation with stability BSP regions. 2.5 REACTIVITY CONTROL The generic discussion in TLTR Sections 5.6.3 and Appendix J.2.3.3 applies to the Limerick plant. The Control Rod Drive (CRD) and CRD hydraulic systems and supporting equipment are not affected by the TPO uprate and no further evaluation of CRD performance is necessary.
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-5 Table 2-1 OPRM Setpoint Versus OLMCPR Demonstration Limerick 1&2 TPO OPRM Set Point OLMCPR (2RPT) OLMCPR (SS) 1.05 1.144 1.160 1.06 1.163 1.179 1.07 1.183 1.199 1.08 1.204 1.220 1.09 1.225 1.242 1.10 1.247 1.264 1.11 1.268 1.285 1.12 1.290 1.307 1.13 1.312 1.330 1.14 1.335 1.354 1.15 1.359 1.378 1.16 1.384 1.403 1.17 1.408 1.428 1.18 1.434 1.454 1.19 1.461 1.481 1.20 1.489 1.509 Acceptance Criteria Rated Power OLMCPR Off-rated OLMCPR At 45% Flow NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-6 Table 2-2 BSP Region Intercepts for Nominal Feedwater Temperature Demonstration Region Boundary Intercept % TPO Power % Core Flow Scram Region (Region I) Boundary Intercept on HCFL A1 71.1 48.5 Scram Region (Region I) Boundary Intercept on NCL B1 Base 52.2 39.3 Controlled Entry Region (Region II) Boundary Intercept on HCFL A2 72.7 50.2 Controlled Entry Region (Region II) Boundary Intercept on NCL B2 Base 36.0 38.6 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-7 Table 2-3 BSP Region Intercepts for Minimum Feedwater Temperature Demonstration Region Boundary Intercept % TPO Power % Core Flow Scram Region (Region I) Boundary Intercept on HCFL A1 87.3 67.2 Scram Region (Region I) Boundary Intercept on NCL B1 45.5 39.1 Controlled Entry Region (Region II) Boundary Intercept on HCFL A2 88.4 68.6 Controlled Entry Region (Region II) Boundary Intercept on NCL B2 Base 36.0 38.6 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-8  Figure 2-1 Illustration of OPRM Trip-Enabled Region
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION As shown in Table 2-1, with an estimated OLMCPR of 1.34 and an estimated SLMCPR of 1.07, an OPRM trip setpoint of 1.14 with a successive confirmation count of 16 (Reference 6) is the highest setpoint that may be used without stability setting the OLMCPR. The actual setpoint will be established in accordance with Limerick Units 1 and 2 Technical Specifications at each reload. These demonstration results are based on a power level of 101.7% CLTP, which is bounding for a power level of 101.65%.
Therefore, TPO operation is justified for plant operation with stability Option III.
2.4.2 Stability Backup Stability Protection Limerick Units 1 and 2 have implemented the Backup Stability Protection (BSP) regions (Reference 8) as the stability backup solution if the OPRM system is declared inoperable.
The BSP regions consist of two regions, I-Scram and II-Controlled Entry. The Base BSP Scram Region and Base BSP Controlled Entry Region are defined by state points on the High Flow Control Line (HFCL) and on the Natural Circulation Line (NCL) in accordance with Reference 8. The bounding plant-specific BSP region state points must enclose the corresponding Base BSP region state points on the HFCL and on the NCL. If a calculated BSP region state point is located inside the corresponding base BSP region state point, then it must be replaced by the corresponding base BSP region state point. If a calculated BSP region state point is located outside the corresponding base BSP region state point, this point is acceptable for use.
That is, the selected points will result in the largest, or most conservative, region sizes. The proposed BSP Scram and Controlled Entry region boundaries are constructed by connecting the corresponding bounding state points on the HFCL and the NCL using a shape function. The Modified Shape Function (MSF) is demonstrated.
The demonstration BSP regions for both Nominal Feedwater Temperature (NFWT) and Minimum Feedwater Temperature (MFWT) operations are shown in Table 2-2/Figure 2-2 and Table 2-3/Figure 2-3, respectively. The OPRM Trip-Enabled Region is confirmed for NFWT operations based on the demonstration BSP regions for NFWT. The demonstration BSP regions for MFWT confirm the OPRM Trip-Enabled Region for operations on or below the 97% rod line. These demonstration results are based on a power level of 101.7% CLTP, which is bounding for a power level of 101.65%.
The BSP regions are confirmed or expanded on a cycle-specific basis.
Therefore, TPO operation is justified for plant operation with stability BSP regions.
2.5   REACTIVITY CONTROL The generic discussion in TLTR Sections 5.6.3 and Appendix J.2.3.3 applies to the Limerick plant. The Control Rod Drive (CRD) and CRD hydraulic systems and supporting equipment are not affected by the TPO uprate and no further evaluation of CRD performance is necessary.
2-4


Limerick TPO OPRM Trip-Enabled Region 0 10 20 30 40 50 60 70 80 90 100 110 120 1300102030405060708090100110120Core Flow (%)Thermal Power (% TLTP) 0400800 1200 1600 200024002800 3200 3600 4000 44000102030405060708090100110120Core Flow (Mlbm/hr)Thermal Power (MWt)OPRM Trip-Enabled Region NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-9  Figure 2-2 BSP Regions for Nominal Feedwater Temperature Demonstration
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 2-1 OPRM Setpoint Versus OLMCPR Demonstration Limerick 1&2 TPO OPRM Set Point OLMCPR (2RPT)                    OLMCPR (SS) 1.05                          1.144                            1.160 1.06                          1.163                            1.179 1.07                          1.183                            1.199 1.08                          1.204                            1.220 1.09                          1.225                            1.242 1.10                          1.247                            1.264 1.11                          1.268                            1.285 1.12                          1.290                            1.307 1.13                          1.312                            1.330 1.14                          1.335                            1.354 1.15                          1.359                            1.378 1.16                          1.384                            1.403 1.17                          1.408                            1.428 1.18                          1.434                            1.454 1.19                          1.461                            1.481 1.20                          1.489                            1.509 Rated Power                  Off-rated OLMCPR Acceptance Criteria OLMCPR                        At 45% Flow 2-5


Limerick TPO NFWT Proposed BSP Regions 0 10 20 30 40 50 60 70 80 90 100 110 120 1300102030405060708090100110120Core Flow (%)Thermal Power (% TLTP) 0 400 800 1200 1600 2000 2400 2800 3200 3600 4000 44000102030405060708090100110120Core Flow (Mlbm/hr)Thermal Power (MWt)Scram RegionControlled Entry RegionBSP Endpoints A 1B1 BaseB2 Base A2 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2-10 Figure 2-3 BSP Regions for Minimum Feedwater Temperature Demonstration  
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 2-2 BSP Region Intercepts for Nominal Feedwater Temperature Demonstration Region Boundary Intercept                    % TPO Power                          % Core Flow Scram Region (Region I) Boundary Intercept on HCFL A1                                    71.1                              48.5 Scram Region (Region I) Boundary Intercept on NCL B1 Base                                  52.2                              39.3 Controlled Entry Region (Region II) Boundary Intercept on HCFL A2                                    72.7                              50.2 Controlled Entry Region (Region II) Boundary Intercept on NCL B2 Base                                  36.0                              38.6 2-6


Limerick TPO MFWT Proposed BSP Regions 0 10 20 30 40 50 60 70 80 90 100 110 120 13001 02 03 04 050607 080901 001 101 2 0Core Flow (%)Thermal Power (% TLTP) 0 400 800 1200 1600 2000 2400 2800 3200 3600 4000 44000102030405060708090100110120Core Flow (Mlbm/hr)Thermal Power (MWt)Scram RegionControlled Entry RegionBSP EndpointsB2 Base A2 A1 B1 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-1 3.0  REACTOR COOLANT AND CONNECTED SYSTEMS 3.1 NUCLEAR S YSTEM P RESSURE RELIEF / OVERPRESSURE PROTECTION The pressure relief system prevents over-pressurization of the nuclear system during abnormal operational transients. The SRVs along with other functions provide this protection. Evaluations and analyses for the CLTP have been performed at 102% of CLTP to demonstrate that the reactor vessel conformed to ASME Boil er and Pressure Vessel (B&PV) Code and plant Technical Specification requirements. There is no increase in nominal operating pressure for the Limerick TPO uprate. There are no changes in the SRV setpoints or valve OOS options. There is no change in the methodology or the limiting overpressure event. Therefore, the generic evaluation contained in the TLTR is applicable. The analysis for each fuel reload, which is current practice, confirms the capability of the system to meet the ASME design criteria. 3.2 REACTOR V ESSEL  The RPV structure and support components form a pressure boundary to contain 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.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 2-3 BSP Region Intercepts for Minimum Feedwater Temperature Demonstration Region Boundary Intercept                    % TPO Power                          % Core Flow Scram Region (Region I) Boundary Intercept on HCFL A1                                    87.3                               67.2 Scram Region (Region I) Boundary Intercept on NCL B1                                    45.5                               39.1 Controlled Entry Region (Region II) Boundary Intercept on HCFL A2                                    88.4                               68.6 Controlled Entry Region (Region II) Boundary Intercept on NCL B2 Base                                  36.0                               38.6 2-7
3.2.1 Fracture Toughness The TLTR, Secti on 5.5.1.5, describes the RPV fracture toughness evaluation process. RPV embrittlement is caused by neutron exposure of the wall adjacent to the co re including the regions above and below the core that experience fluence  1.0E+17 n/cm
: 2. This region is defined as the "beltline" region. Opera tion at TPO conditions results in a high er neutron flux, which increases the integrated fluence over the period of plant life. Limerick Units 1 and 2 are evaluated for a fluence that bounds the required value for operation at TPO conditions. The neutron fluence for TPO is calculated using two-dimensional neutron transport theory. The neutron transport methodology is consistent with RG 1.190. A bounding peak fluence, 1.9E+18 n/cm 2, is used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G (Reference 9). The results of these evaluations indicate that: (a) The upper shelf energy (USE) will remain > 50 ft-lb for the design life of the vessel or maintain the margin requirements of 10 CFR 50, Appendix G as defined in RG 1.99 (Reference 10). The minimum USE for the Unit 1 beltline materials is 24 ft-lb for 32 effective full power years (EFPY) and for th e Unit 2 beltline materials is 25 ft-lbs for 32 EFPY. Many of the Limerick RPV materials do not have sufficient unirradiated USE data, and Charpy data from low temperature tests were used to develop an initial USE. Therefore, Equivalent Margin Analyses were perfor med for the limiting beltline plate, weld, and nozzle forging NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-2 materials to assure qualification. These values are provided in Tables 3-1 and 3-2 for Limerick Units 1 an d 2, respectively. (b) The beltline material RTNDT remains below the 200&deg;F screening criteria as defined in Reference 10. These values are provided in Tables 3-3 and 3-4 for Limerick Units 1 and 2, respectively. (c) The CLTP Pressure-Temperatu re (P-T) curves (References 11 and 12) remain bounding for TPO, limited to the currently approved fluence. The current Adjusted Reference Temperature (ART) values for the beltline pl ates and welds remain bounding for TPO. The currently licensed P-T curves include the Low Pressure Coolant Injection (LPCI) nozzle. The water level instrumentation no zzle that occurs within the beltline region is bounded by the CLTP curves. (d) The surveillance program consists of three capsules in each vessel. No capsules have been removed from either vessel. These three capsules have been in each reactor vessel since plant startup. Limerick is a participant in the Integrated Surveillance Program (Reference 13), currently administrated by EPRI, and is not designated as a representative plant; therefore, no capsules are slated for removal at this time. TPO has no effect on the existing surveillance schedule. (e) The 32 EFPY beltline axial an d circumferential weld materi al RTNDT remains bounded by the requirements of Boiling Water Reactor Vessel and Internals Project (BWRVIP)-05 as defined in References 14 and 15. This comparison is provided in Tables 3-5 and 3-6 for axial and circumferential welds, respectively. The maximum normal operating dome pressure for TPO is unchanged from that for CLTP power operation. Therefore, the hydrostatic and leakage test pressures and associated temperatures are acceptable for the TPO. Because the vessel is still in compliance with the regulatory requirements as demonstrated above, operation with TPO does not have an adverse effect (not exceeding regulatory requirements) on the reactor vessel fracture toughness.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-3 3.2.2 Reactor Vessel Structural Evaluation 
[[                                                                                                                                                                                         


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-4                                                               
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 2-1    Illustration of OPRM Trip-Enabled Region Limerick TPO OPRM Trip-Enabled Region Core Flow (Mlbm/hr) 0  10  20    30      40        50        60        70  80  90  100  110  120 130 4400 120 4000 110 3600 100 90                                                                                        3200 Thermal Power (% TLTP)                                                                                                    Thermal Power (MWt) 80                                                                                        2800 70                                                                                        2400 60 OPRM Trip-                                            2000 50                                Enabled Region 1600 40 1200 30 800 20 10                                                                                        400 0                                                                                        0 0  10  20    30      40        50        60        70  80  90  100  110  120 Core Flow (%)
2-8


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-5                                                       
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 2-2          BSP Regions for Nominal Feedwater Temperature Demonstration Limerick TPO NFWT Proposed BSP Regions Core Flow (Mlbm/hr) 0  10      20          30              40        50          60            70  80  90  100  110  120 130 4400 120            Scram Region 4000 110            Controlled Entry Region BSP Endpoints                                                                                      3600 100 90                                                                                                              3200 Thermal Power (% TLTP)                                                                                                                            Thermal Power (MWt) 80                                                                                                              2800 A2 A1 70                                                                                                              2400 60 2000 B1 Base 50 1600 40 B2 Base                                                                      1200 30 800 20 10                                                                                                              400 0                                                                                                                0 0  10      20          30              40        50          60            70  80  90  100  110  120 Core Flow (%)
2-9


                                                                                        ]] [[                                                                                                                                                                                           
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 2-3        BSP Regions for Minimum Feedwater Temperature Demonstration Limerick TPO MFWT Proposed BSP Regions Core Flow (Mlbm/hr) 0  10    20              30            40        50          60              70    80  90  100  110  120 130 4400 120 Scram Region 4000 110              Controlled Entry Region BSP Endpoints 3600 100 A2                            3200 90                                                                                A1 80                                                                                                                    2800 Thermal Power (% TLTP)                                                                                                                                Thermal Power (MWt) 70 2400 60 2000 50 B1                                                                        1600 40 B2 Base                                                                        1200 30 800 20 10                                                                                                                    400 0                                                                                                                    0 0  10    20              30            40        50          60              70    80  90  100  110  120 Core Flow (%)
2-10


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-6                                   
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.0 REACTOR COOLANT AND CONNECTED SYSTEMS 3.1    NUCLEAR SYSTEM PRESSURE RELIEF / OVERPRESSURE PROTECTION The pressure relief system prevents over-pressurization of the nuclear system during abnormal operational transients. The SRVs along with other functions provide this protection.
Evaluations and analyses for the CLTP have been performed at 102% of CLTP to demonstrate that the reactor vessel conformed to ASME Boiler and Pressure Vessel (B&PV) Code and plant Technical Specification requirements. There is no increase in nominal operating pressure for the Limerick TPO uprate. There are no changes in the SRV setpoints or valve OOS options. There is no change in the methodology or the limiting overpressure event. Therefore, the generic evaluation contained in the TLTR is applicable.
The analysis for each fuel reload, which is current practice, confirms the capability of the system to meet the ASME design criteria.
3.2    REACTOR VESSEL The RPV structure and support components form a pressure boundary to contain 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.
3.2.1  Fracture Toughness The TLTR, Section 5.5.1.5, describes the RPV fracture toughness evaluation process. RPV embrittlement is caused by neutron exposure of the wall adjacent to the core including the regions above and below the core that experience fluence  1.0E+17 n/cm2. This region is defined as the beltline region. Operation at TPO conditions results in a higher neutron flux, which increases the integrated fluence over the period of plant life. Limerick Units 1 and 2 are evaluated for a fluence that bounds the required value for operation at TPO conditions.
The neutron fluence for TPO is calculated using two-dimensional neutron transport theory. The neutron transport methodology is consistent with RG 1.190. A bounding peak fluence, 1.9E+18 n/cm2, is used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G (Reference 9). The results of these evaluations indicate that:
(a)  The upper shelf energy (USE) will remain > 50 ft-lb for the design life of the vessel or maintain the margin requirements of 10 CFR 50, Appendix G as defined in RG 1.99 (Reference 10). The minimum USE for the Unit 1 beltline materials is 24 ft-lb for 32 effective full power years (EFPY) and for the Unit 2 beltline materials is 25 ft-lbs for 32 EFPY. Many of the Limerick RPV materials do not have sufficient unirradiated USE data, and Charpy data from low temperature tests were used to develop an initial USE. Therefore, Equivalent Margin Analyses were performed for the limiting beltline plate, weld, and nozzle forging 3-1


]] High and low pressure seal leak detection nozzles were not considered to be pressure boundary components at the time that the OLTP evaluation was performed, and have not been evaluated for TPO. The effect of TPO 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, 1968 Edition with addenda to and including Summer 1969 (except that Figure N-462(e)(2) of the Summer 1970 Addenda were applied) was used as the governing code and is considered the Code of Construction. However, if a component's design has been modified, the governing code for that component was the code used in the stress analysis of the modified component. The following components were modified since the original construction of Limerick Units 1 and 2: FW Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION materials to assure qualification. These values are provided in Tables 3-1 and 3-2 for Limerick Units 1 and 2, respectively.
LPCI Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressu re Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1975 and 1968 Edition with Addenda to and including Winter 1969.
(b)  The beltline material RTNDT remains below the 200&deg;F screening criteria as defined in Reference 10. These values are provided in Tables 3-3 and 3-4 for Limerick Units 1 and 2, respectively.
CRD Hydraulic System Return Nozzle (Unit 2): This component was modified and the governing Code for the modifica tion is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976.
(c)  The CLTP Pressure-Temperature (P-T) curves (References 11 and 12) remain bounding for TPO, limited to the currently approved fluence. The current Adjusted Reference Temperature (ART) values for the beltline plates and welds remain bounding for TPO. The currently licensed P-T curves include the Low Pressure Coolant Injection (LPCI) nozzle. The water level instrumentation nozzle that occurs within the beltline region is bounded by the CLTP curves.
CS Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1975 and 1968 Edition with Addenda to and including Winter 1969.
(d)  The surveillance program consists of three capsules in each vessel. No capsules have been removed from either vessel. These three capsules have been in each reactor vessel since plant startup. Limerick is a participant in the Integrated Surveillance Program (Reference 13),
Recirculation Inlet Nozz le: This component was modified and the governing Code for the modification is the ASME Boile r and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and in cluding Summer 1 976 and 1968 Edition with Addenda to and including Summer 1969.
currently administrated by EPRI, and is not designated as a representative plant; therefore, no capsules are slated for removal at this time. TPO has no effect on the existing surveillance schedule.
Universal Dry Tube, Power Range Detector, and In-Core Detector Assembly: These components were modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Ve ssel Code, Section III, 1968 Edition with Addenda to and including Winter 1969, 1971 Edition with Addenda to and including Summer 1973, and NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-7 1977 Edition with Addenda to and including Summer 1977 (varies depending upon specific component).
(e)  The 32 EFPY beltline axial and circumferential weld material RTNDT remains bounded by the requirements of Boiling Water Reactor Vessel and Internals Project (BWRVIP)-05 as defined in References 14 and 15. This comparison is provided in Tables 3-5 and 3-6 for axial and circumferential welds, respectively.
Typically, new stresses are determined by scali ng the "original" stresses based on the TPO conditions (pressure, temperatur e, and flow). The bounding analyses were performed for the design, normal & upset, and emergency & faulted c onditions. If there is an increase in annulus pressurization, jet reaction, pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency and faulted conditions.
The maximum normal operating dome pressure for TPO is unchanged from that for CLTP power operation. Therefore, the hydrostatic and leakage test pressures and associated temperatures are acceptable for the TPO. Because the vessel is still in compliance with the regulatory requirements as demonstrated above, operation with TPO does not have an adverse effect (not exceeding regulatory requirements) on the reactor vessel fracture toughness.
3-2
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.2.2 Reactor Vessel Structural Evaluation
((
3-3
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-4
 
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              ))
((
3-5
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION
                                                                    ))
High and low pressure seal leak detection nozzles were not considered to be pressure boundary components at the time that the OLTP evaluation was performed, and have not been evaluated for TPO.
The effect of TPO 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, 1968 Edition with addenda to and including Summer 1969 (except that Figure N-462(e)(2) of the Summer 1970 Addenda were applied) was used as the governing code and is considered the Code of Construction. However, if a components design has been modified, the governing code for that component was the code used in the stress analysis of the modified component. The following components were modified since the original construction of Limerick Units 1 and 2:
* FW Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976.
* LPCI Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1975 and 1968 Edition with Addenda to and including Winter 1969.
* CRD Hydraulic System Return Nozzle (Unit 2): This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976.
* CS Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1975 and 1968 Edition with Addenda to and including Winter 1969.
* Recirculation Inlet Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976 and 1968 Edition with Addenda to and including Summer 1969.
* Universal Dry Tube, Power Range Detector, and In-Core Detector Assembly: These components were modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section III, 1968 Edition with Addenda to and including Winter 1969, 1971 Edition with Addenda to and including Summer 1973, and 3-6
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1977 Edition with Addenda to and including Summer 1977 (varies depending upon specific component).
Typically, new stresses are determined by scaling the original stresses based on the TPO conditions (pressure, temperature, and flow). The bounding analyses were performed for the design, normal & upset, and emergency & faulted conditions. If there is an increase in annulus pressurization, jet reaction, pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency and faulted conditions.
3.2.2.1 Design Conditions Because there are no changes in the design conditions due to TPO, the design stresses are unchanged and the Code requirements are met.
3.2.2.1 Design Conditions Because there are no changes in the design conditions due to TPO, the design stresses are unchanged and the Code requirements are met.
3.2.2.2 Normal & Upset Conditions The reactor coolant temperatur e and flows at TPO conditions are unchanged from those at current rated conditions, because the 105% OLTP power uprate evaluations were performed at conditions [[                                ]] that bound the change in operating conditions from CLTP to TPO. The evaluation type is mainly reconciliation of the stresses and usage factors to reflect TPO conditions. A primary plus secondary stress analysis was performed showing TPO stresses still meet the requirements of the ASME Code, Section III, and Subsection NB for all components. Lastly, the fatigue usage was evaluated for the limiting location of components [[                                                                                ]]  The Limerick Units 1 and 2 fatigue analysis results for the limiting components are provided in Table 3-7. The Limerick Units 1 and 2 analysis results for TPO show that all components meet their ASME Code requirements and no further analysis is required.
3.2.2.2 Normal & Upset Conditions The reactor coolant temperature and flows at TPO conditions are unchanged from those at current rated conditions, because the 105% OLTP power uprate evaluations were performed at conditions ((                    )) that bound the change in operating conditions from CLTP to TPO. The evaluation type is mainly reconciliation of the stresses and usage factors to reflect TPO conditions. A primary plus secondary stress analysis was performed showing TPO stresses still meet the requirements of the ASME Code, Section III, and Subsection NB for all components. Lastly, the fatigue usage was evaluated for the limiting location of components ((
3.2.2.3 Emergency & Faulted Conditions The stresses due to Emergency & Faulted cond itions are based on load s such as peak dome pressure, which are unchanged for TPO. Therefore, Code requirements are met for all RPV components under Emergency & Faulted conditions.
                                        )) The Limerick Units 1 and 2 fatigue analysis results for the limiting components are provided in Table 3-7. The Limerick Units 1 and 2 analysis results for TPO show that all components meet their ASME Code requirements and no further analysis is required.
3.3 REACTOR INTERNALS The reactor internals include core support structure (CSS) a nd non-core support structure (non-CSS) components.  
3.2.2.3 Emergency & Faulted Conditions The stresses due to Emergency & Faulted conditions are based on loads such as peak dome pressure, which are unchanged for TPO. Therefore, Code requirements are met for all RPV components under Emergency & Faulted conditions.
3.3   REACTOR INTERNALS The reactor internals include core support structure (CSS) and non-core support structure (non-CSS) components.
3.3.1  Reactor Internal Pressure Difference The reactor internal pressure differences (RIPDs) are affected more by the maximum licensed core flow rate than by the power level. The maximum licensed core flow rate is not changed for the TPO uprate. The effect due to the changes in loads for both Normal and Upset conditions is reported in Section 3.3.2. The Normal and Upset evaluations of RIPDs for the TPO uprate are 3-7
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION bounded by the current analyses that conservatively assumed an initial power level of 110% of OLTP for Normal conditions and 112% of OLTP for Upset conditions. The Emergency and Faulted evaluations of RIPDs for the TPO uprate are bounded by the current analyses that conservatively assumed an initial power level of 112% of OLTP.
Fuel Bundle Lift Margins and Control Rod Guide Tube (CRGT) Lift Forces are calculated at the Faulted condition to demonstrate that fuel bundles would not lift under the worst conditions. The current analysis conservatively assumed of 112% of OLTP and 110% core flow, which bounds the TPO. The Fuel Lift Margins for the normal and upset conditions at the TPO RTP decrease slightly from CLTP. The CRGT Lift Forces for the normal condition at the TPO RTP increase slightly from CLTP. The Fuel Lift Margins and CRGT Lift Forces at Normal and Upset conditions are bounded by Emergency and Faulted conditions. The effect due to the changes in Fuel Lift Margins and CRGT Lift Forces is reported in Section 3.3.2.
Acoustic and flow-induced loads on jet pump, core shroud and shroud support due to recirculation line break are bounded by the current analyses that conservatively assumed an initial power level of 102% of CLTP.
3.3.2    Reactor Internals Structural Evaluation The RPV internals consist of the CSS components and non-CSS components. The RPV Internals are not ASME Code components, however, the requirements of the ASME Code are used as guidelines in their design/analysis. The evaluations/stress reconciliation in support of the TPO was performed consistent with the design basis analysis of the components. The reactor internal components evaluated are:
CSS Components
* Shroud Support
* Shroud
* Core Plate
* Top Guide
* Control Rod Drive Housing
* Control Rod Guide Tube
* Orificed Fuel Support Non-CSS Components
* FW Sparger 3-8
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION
* Jet Pump
* CS Line and Sparger
* Access Hole Cover
* Shroud Head and Steam Separator Assembly
* In-Core Housing and Guide Tube
* Vessel Head Cooling Spray Nozzle
* Core Differential Pressure and Liquid Control Line
* LPCI Coupling
* Steam Dryer The original configurations of the RPV internals are considered in the TPO evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation (e.g., jet pumps).
The loads considered in the evaluation of the RPV internals include RIPDs, dead weight, seismic, SRV, LOCA, Annulus Pressurization/Jet Reaction (AP/JR), acoustic and flow induced loads due to Recirculation Line Break (RLB), fuel lift, hydraulic flow and thermal loads.
RPV design pressure remains unchanged. RIPD loads are bounded by the existing design basis values (110% of OLTP and 110% ICF). Seismic, SRV, LOCA and AP/JR loads remain unchanged. Acoustic and flow induced loads due to RLB, hydraulic flow and thermal loads remain bounded. The effect of weight change on load due to jet pump repair is insignificant.
Fuel lift loads increased in Service Level D. The stresses of core plate, top guide and control rod guide tube were reconciled for the increase of the fuel lift loads to show that adequate stress margins exist, and the stresses remain within the allowable limits. The limiting stresses of other RPV internal components remain bounded by the existing design basis values (110% OLTP and 110% ICF) (See Table 3-8). The input loads to existing flaw analysis are not impacted by TPO.
Hence, the existing flaw evaluations remain valid to TPO. Therefore the RPV internal components are demonstrated to be structurally qualified for operation in the TPO conditions.
Qualitative analysis was performed to assess the potential for the Steam Dryer flow induced vibration (FIV) using 1/5th scale model testing and an analytical approach. The scale model testing has indicated some potential propagation of acoustic resonance in two of the four main steam lines (MSLs) at CLTP conditions. The Steam Dryer has operated satisfactorily for approximately 14 years at CLTP with no adverse flow effects. The testing has also shown a downward trend (i.e., reduction) of the normalized root mean square (RMS) pressure in the 3-9
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION MSLs as flow conditions are changed from CLTP to TPO conditions. Additionally, for the other two MSLs, the normalized RMS pressure was determined to be constant and below any potential acoustic resonance propagation thresholds. Other loads applicable to the Steam Dryer evaluation are deadweight, seismic, RIPD, SRV, LOCA, AP, JR and fuel lift loads. Dead weight and seismic loads remain unchanged for the TPO conditions. RIPDs contributing to the Steam Dryer load remain bounded by previous analyses. SRV, LOCA, AP, JR, and fuel lift loads remain bounded in the TPO conditions by the CLTP values. All applicable loads to the Steam Dryer are bounded by the existing design basis for the TPO conditions. Hence, the Steam Dryer remains structurally qualified for plant operation at the TPO conditions.
3.3.3    Steam Separator and Dryer Performance For Limerick, the TPO performance of the steam dryer/separator was evaluated based on a plant-specific evaluation using recent fuel cycle core design for Unit 1 Fuel Cycles 12 and 13, which also bounds the current Unit 2 Fuel Cycle 11. The results of the evaluation demonstrated that the steam dryer/separator will maintain moisture content  0.1 weight % at TPO operating conditions except under some ICF conditions or above L7. TPO results in an increase in the amount of saturated steam generated in the reactor core. For constant core flow, this results in an increase in the separator inlet quality, an increase in the steam 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 dryer/separator performance. The net effect of changes due to TPO up to 100 % RCF does not result in exceeding a moisture content of 0.1 wt % leaving the steam dryer. TPO does not significantly affect the performance of the steam dryer/separator.
3.4    FLOW-INDUCED VIBRATION The process for the reactor vessel internals vibration assessment is described in Section 5.5.1.3 of the TLTR (Reference 1). An evaluation determined the effects of FIV on the reactor internals at 110% rated core flow and TPO RTP of 101.7%. The vibration levels for the TPO uprate conditions were estimated from vibration data recorded during startup testing of the NRC designated prototype plant (Browns Ferry Unit 1) and during other tests. These expected vibration levels were compared with established vibration acceptance limits. The following components were evaluated for the TPO uprate:
Component(s)                Process Parameter(s)                  TPO Evaluation Shroud                    Steam Flow at TPO RTP is ~2%      Slight increase in FIV. Extrapolation of Shroud Head and Separator  greater than CLTP.                measured data shows stresses are within limits.
Jet Pumps                  Jet Pump flow at TPO is unchanged No change from CLTP.
Jet Pump Riser Braces      Resonance due to vane passing    No resonance due to vane passing frequency                        frequency Jet Pump Sensing Lines    Resonance due to vane passing    No resonance due to vane passing frequency.                        frequency.
3-10


3.3.1 Reactor Internal Pressure Difference The reactor internal pressure differences (RIPDs) are affected more by the maximum licensed core flow rate than by the power level. The maximum licensed core flow rate is not changed for the TPO uprate. The effect due to the changes in loads for both Normal and Upset conditions is reported in Section 3.3.2. The Normal and Upset evaluations of RIPDs for the TPO uprate are NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-8 bounded by the current analyses that conservatively assumed an initial pow er level of 110% of OLTP for Normal conditions and 112% of OLTP for Upset conditions. The Emergency and Faulted evaluations of RIPDs for the TPO uprate are bounded by the current analyses that conservatively assumed an initia l power level of 112% of OLTP.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Component(s)                Process Parameter(s)                  TPO Evaluation FW Sparger                  FW Flow at TPO RTP is ~ 2%       Slight increase in FIV. Extrapolation of greater than CLTP.               measured data shows stresses are well within limits.
Fuel Bundle Lift Margins and Control Rod Guide T ube (CRGT) Lift Forces are calculated at the Faulted condition to demonstrate that fuel bundles would not lift under the worst conditions. The current analysis conservatively assumed of 112
Control Rod Guide Tube and Core Flow at TPO Licensed Thermal No change In-Core Guide Tubes        Power (TLTP) is unchanged from CLTP.
% of OLTP and 110% co re flow, which bounds the TPO. The Fuel Lift Margins for the normal and upset conditions at the TPO RTP decrease slightly from CLTP. The CRGT Lift Forces for the normal condition at the TPO RTP increase slightly from CLTP. The Fuel Lift Margins and CRGT Lift Forces at Normal and Upset conditions are bounded by Emergency and Faulted conditions. The e ffect due to the changes in Fuel Lift Margins and CRGT Lift Forces is reported in Section 3.3.2.
The calculations for the TPO uprate conditions indicate that vibrations of all safety-related reactor internal components are within the GEH acceptance criteria. For some components, FIV is a function of core flow. Because the maximum licensed core flow is unchanged for TPO, FIV for those components is not affected. The analysis is conservative for the following reasons:
Acoustic and flow-induced loads on jet pum p, core shroud and shroud support due to recirculation line break are bounded by the current analyses that conservatively assumed an initial power level of 102% of CLTP.
* The GEH criteria of 10,000 psi peak stress intensity is much more conservative than the ASME allowable peak stress intensity of 13,600 psi for service cycles in excess of 1011.
3.3.2 Reactor Internals Structural Evaluation The RPV internals consist of the CSS co mponents and non-CSS co mponents. The RPV Internals are not ASME Code components, however, the requirements of the ASME Code are used as guidelines in their design/analysis. The evaluations/stress reconcil iation in support of the TPO was performed consistent with the design basis analysis of the components. The reactor internal components evaluated are:
* Conservatively, the peak responses of the applicable modes are absolute summed.
CSS Components Shroud Support Shroud Core Plate  Top Guide Control Rod Drive Housing Control Rod Guide Tube Orificed Fuel Support Non-CSS Components FW Sparger NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-9 Jet Pump CS Line and Sparger Access Hole Cover Shroud Head and Steam Separator Assembly In-Core Housing and Guide Tube Vessel Head Cooling Spray Nozzle Core Differential Pressure and Liquid Control Line  LPCI Coupling Steam Dryer The original configurations of the RPV internals are considered in the TPO evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation (e.g., jet pumps). The loads considered in the evaluation of the RPV internals include RIPDs, dead weight, seismic, SRV, LOCA, Annulus Pressurization/Je t Reaction (AP/JR), acous tic and flow induced loads due to Recirculation Line Break (RLB), fuel lift, hydraulic flow and thermal loads. RPV design pressure remains unchanged. RIPD loads are bounded by the existing design basis values (110% of OLTP and 110% ICF). Seismic, SRV, LOCA and AP/JR loads remain unchanged. Acoustic and flow induced loads due to RLB, hydraulic flow and thermal loads remain bounded. The effect of weight change on load due to jet pump repair is insignificant. Fuel lift loads increased in Service Level D. The stresses of core plate, top guide and control rod guide tube were reconciled for the increase of the fuel lift loads to show that adequate stress margins exist, and the stresses remain within the allowable limits. The limiting stresses of other RPV internal components remain bounded by the existing design basis values (110% OLTP and 110% ICF) (See Table 3-8). The in put loads to existing flaw analysis are not impacted by TPO. Hence, the existing flaw evaluations remain valid to TPO. Therefore the RPV internal components are demonstrated to be structurally qualified for operation in the TPO conditions. Qualitative analysis was performed to assess the potential for the Steam Dryer flow induced vibration (FIV) using 1/5th scale model testing and an analytical approach. The scale model testing has indicated some potential propagation of acoustic resonance in two of the four main steam lines (MSLs) at CLTP conditions. The Steam Dryer has operated satisfactorily for approximately 14 years at CLTP with no adverse flow effects. The testing has also shown a downward trend (i.e., reduction) of the normalized root mean square (RMS) pressure in the NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-10 MSLs as flow conditions are changed from CLTP to TPO conditions. Additionally, for the other two MSLs, the normalized RMS pressure was determined to be constant and below any potential acoustic resonance propagation thresholds. Other loads applicable to the Steam Dryer evaluation are deadweight, seismic, RIPD, SRV, LOCA, AP, JR and fuel lift loads.
* Although the maximum vibration stress amplitude of each mode is used in the absolute sum process, the maximum vibration modal amplitude actually differs with time.
Dead weight and seismic loads remain unchanged for the TPO cond itions. RIPDs contributing to the Steam Dryer load remain bounded by previous analyses. SRV, LOCA, AP, JR, and fuel lift loads remain bounded in the TPO conditions by the CLTP values. All applicable loads to the Steam Dryer are
Therefore, it is concluded that the flow-induced vibrations for all evaluated components remain within acceptable limits.
The safety-related Main Steam (MS) and Feedwater (FW) piping have minor increased flow rates and flow velocities resulting from the TPO uprate. The MS and FW piping experience increased vibration levels, approximately proportional to the increase in the square of the flow velocities and also in proportion to any increase in fluid density. The decrease in FW fluid density for TPO uprate conditions, as a result of the ~ 2&#xba; F increase in FW temperature, is insignificant. The MS and FW piping vibration is expected to increase only by about 4%. A MS and FW piping FIV test program, during initial plant startup, showed that vibration levels were within acceptance criteria and operating experience shows that there are no existing vibration problems in MS and FW lines at CLTP operating conditions. Therefore, the MS and FW lines vibration will remain within acceptable limits during TPO. Analytical evaluation has shown that the safety-related thermowells in the MS, FW, and Recirculation piping systems are structurally adequate for the TPO operating condition.
3.5    PIPING EVALUATION 3.5.1 Reactor Coolant Pressure Boundary Piping The methods used for the piping and pipe support evaluations are described in TLTR Appendix K. These approaches are identical to those used in the evaluation of previous BWR power uprates of up to 20% power. The effect of the TPO uprate with no nominal vessel dome 3-11


bounded by the existing design basis for the TPO conditions. Hence, the Steam Dryer remains structurally qualified for plant operation at the TPO conditions.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION pressure increase is negligible for the Reactor Coolant Pressure Boundary (RCPB) portion of all piping except for portions of the FW lines, MS lines, and piping connected to the FW and MS lines. The following table summarizes the evaluation of the piping Inside Containment.
3.3.3 Steam Separator and Dryer Performance For Limerick, the TPO performance of the steam dryer/separator was evaluated based on a plant-specific evaluation using recent fuel cycle core design for Unit 1 Fuel Cycles 12 and 13, which also bounds the current Unit 2 Fuel Cycle 11. The re sults of the evaluation demonstrated that the steam dryer/separator will maintain moisture content  0.1 weight % at TPO operating conditions except under some ICF conditions or a bove L7. TPO results in an increase in the amount of saturated steam generated in the reactor co re. For constant core flow, this results in an increase in the separator inlet quality, an increase in the steam 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 dryer/separator performance. The net effect of changes due to TPO up to 100 % RCF does not result in exceeding a moisture content of 0.1 wt % leaving the steam dryer. TPO does not significantly affect the performance of the steam dryer/separator. 3.4 F LOW-INDUCED VIBRATION The process for the reactor vessel internals vibr ation assessment is described in Section 5.5.1.3 of the TLTR (Reference 1). An evaluation determined the effects of FIV on the reactor internals at 110% rated core flow and TPO RTP of 101.7%. The vibra tion levels for the TPO uprate conditions were estimated from vibration data recorded duri ng startup testing of the NRC designated prototype plant (Browns Ferry Unit 1) and during other tests. These expected vibration levels were compared with established vibration acceptance limits. The following components were evaluated for the TPO uprate:
Component(s) / Concern                           Process Parameter(s)                 TPO Evaluation Recirculation System                     Nominal dome pressure at TPO RTP is identical to   No change in pipe stress CLTP                                               No effect on pipe Recirculation flow at TPO RTP is identical to CLTP supports Pipe Stresses                            No change in core pressure Pipe Supports                            Small decrease in Recirculation fluid temperature MS and Attached Piping                   Nominal dome pressure at TPO RTP is identical to  Current Licensing Basis (Inside Containment) (e.g., SRV         CLTP                                              envelops TPO Discharge Line (SRVDL) piping up to     Steam flow at TPO RTP is ~2% greater than CLTP    conditions; therefore, first anchor, Reactor Core Isolation     No change in MSL pressure                          piping system is Cooling (RCIC) / High Pressure Coolant                                                     acceptable for TPO.
Component(s) Process Parameter(s) TPO Evaluation Shroud Shroud Head and Separator Steam Flow at TPO RTP is ~2%
Injection (HPCI) piping (Steam Side), MS drain lines, RPV head vent line piping                                                     Negligible change in located Inside Containment)                                                                 pipe stress; negligible effect on pipe supports Minor increase in the Pipe Stresses                                                                               potential for FAC (FAC Pipe Supports                                                                               concerns are covered by existing piping monitoring program)
greater than CLTP. Slight increase in FIV. Extrapolation of measured data shows stresses are within limits. Jet Pumps Jet Pump flow at TPO is unchanged from CLTP.
Flow-accelerated erosion/corrosion (FAC)
No change Jet Pump Riser Braces Resonance due to vane passing frequency No resonance due to vane passing frequency Jet Pump Sensing Lines Res onance due to vane passing frequency. No resonance due to vane passing frequency.
FW and Attached Piping                  Nominal dome pressure at TPO RTP is identical to   Current Licensing Basis (Inside Containment)                    CLTP                                               envelops TPO FW flow at TPO RTP is ~2% greater than CLTP       conditions; therefore, Minor change in FW line pressure                  piping system is Pipe Stresses                            Fluid temperature increases 2&deg;F                    acceptable for TPO.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-11 Component(s) Process Parameter(s) TPO Evaluation FW Sparger FW Flow at TPO RTP is ~ 2% greater than CLTP. Slight increase in FIV. Extrapolation of measured data shows stresses are well within limits. Control Rod Guide Tube and In-Core Guide Tubes Core Flow at TPO Licensed Thermal Power (TLTP) is unchanged from CLTP. No change The calculations for the TPO uprate conditions i ndicate that vibrations of all safety-related reactor internal components are within the GEH acceptance criteria. For some components, FIV is a function of core flow. Because the maximum licensed core flow is unchanged for TPO, FIV for those components is not affected. The anal ysis is conservative for the following reasons: The GEH criteria of 10,000 psi peak stress intensity is much more conservative than the ASME allowable peak stress intensity of 13,600 psi for service cycles in excess of 10 11. Conservatively, the peak responses of the applicable modes are absolute summed. Although the maximum vibration stress amplitude of each mode is used in the absolute sum process, the maximum vibration modal amplitude actually differs with time. Therefore, it is concluded that the flow-induced vibrations for all evaluated components remain within acceptable limits. The safety-related Main Steam (MS) and Feedwater (FW) piping have minor increased flow rates and flow velocities resulting from the TPO uprate. The MS and FW piping experience increased vibration levels, approx imately proportional to the increase in the square of the flow velocities and also in proportion to any increase in fluid density. The decrease in FW fluid density for TPO uprate conditions, as a result of the ~ 2&#xba; F increase in FW temperature, is insignificant. The MS and FW piping vibration is expected to increase only by about 4%. A MS and FW piping FIV test program, during initial plan t startup, showed that vibration levels were within acceptance criteria and operating experien ce shows that there are no existing vibration problems in MS and FW lines at CLTP operating conditions. Therefore, the MS and FW lines vibration will remain within acceptable limits during TPO. Analytical evaluation has shown that the safety-related thermowells in the MS, FW, a nd Recirculation piping systems are structurally adequate for the TPO operating condition. 3.5 PIPING EVALUATION 3.5.1 Reactor Coolant Pressure Boundary Piping The methods used for the piping and pipe support evaluations ar e described in TLTR Appendix K. These approaches are identical to those used in the evaluation of previous BWR power uprates of up to 20% power. The effect of the TPO uprate with no nominal vessel dome NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-12 pressure increase is negligible for the Reactor Coolant Pressure Boundary (RCPB) portion of all piping except for portions of the FW lines, MS lines, and piping connected to the FW and MS lines. The following table summarizes the evaluation of the piping Inside Containment. Component(s) / Concern Process Parameter(s) TPO Evaluation Recirculation System Pipe Stresses Pipe Supports Nominal dome pressure at TPO RTP is identical to CLTP Recirculation flow at TPO RTP is identical to CLTP No change in core pressure Small decrease in Recirculation fluid temperature No change in pipe stressNo effect on pipe supports MS and Attached Piping (Inside Containment) (e.g., SRV Discharge Line (SRVDL) piping up to first anchor, Reactor Core Isolation Cooling (RCIC) / High Pressure Coolant Injection (HPCI) piping (Steam Side), MS drain lines, RPV head vent line piping located Inside Containment)
Pipe Supports                                                                              Negligible change in pipe stress Negligible effect on pipe supports FAC Minor increase in the potential for FAC (FAC concerns are covered by existing piping monitoring program)
Pipe Stresses Pipe Supports Flow-accelerated erosion/corrosion (FAC) Nominal dome pressure at TPO RTP is identical to CLTP Steam flow at TPO RTP is ~2% greater than CLTP No change in MSL pressure Current Licensing Basis
RPV bottom head drain line, RCIC        Nominal dome pressure at TPO RTP is identical to  Negligible change in piping, HPCI piping, LPCI piping, CS    CLTP                                              pipe stress piping, Standby Liquid Control System    Small Increase in core pressure drop of < 1 psi    Negligible effect on (SLCS) piping, and Reactor Water        Small decrease in Recirculation fluid temperature  pipe supports Cleanup (RWCU) piping Minor increase in the potential for FAC Pipe Stresses                                                                              (FAC concerns are Pipe Supports                                                                              covered by existing piping monitoring FAC                                                                                        program)
For the MS and FW lines, supports, and connected lines, the methodologies as described in TLTR Section 5.5.2 and Appendix K were used to determine the percent increases in applicable ASME Code stresses, displacements, CUF, and pipe interface component loads (including 3-12


envelops TPO conditions; therefore, piping system is acceptable for TPO. Negligible change in pipe stress; negligible effect on pipe supports Minor increase in the potential for FAC (FAC concerns are covered by existing piping monitoring program)
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION supports) as a function of percentage increase in pressure (where applicable), temperature, and flow due to TPO conditions. As necessary, the percentage increases were applied to the highest calculated stresses, displacements, and the CUF at applicable piping system node points to conservatively determine the maximum TPO calculated stresses, displacements and usage factors. This approach is conservative because the TPO does not affect weight and all building filtered loads (i.e., seismic loads are not affected by the TPO). The factors were also applied to nozzle load, support loads, penetration loads, valves, pumps, heat exchangers and anchors so that these components could be evaluated for acceptability, where required. No new computer codes were used or new assumptions introduced for this evaluation.
FW and Attached Piping (Inside Containment)
MS and Attached Piping System Evaluation The MS piping system (Inside Containment) was evaluated for compliance with the ASME code stress criteria, and for the effects of thermal displacements on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, flanges and valves were also evaluated.
Pipe Stresses Pipe Supports FAC Nominal dome pressure at TPO RTP is identical to CLTP FW flow at TPO RTP is ~2% greater than CLTP Minor change in FW line pressure Fluid temperature increases 2 F Current Licensing Basis envelops TPO conditions; therefore, piping system is acceptable for TPO. Negligible change in pipe stress Negligible effect on pipe supports Minor increase in the potential for FAC  (FAC concerns are covered by existing piping monitoring program) RPV bottom head drain line, RCIC piping, HPCI piping, LPCI piping, CS piping, Standby Liquid Control System (SLCS) piping, and Reactor Water Cleanup (RWCU) piping Pipe Stresses Pipe Supports FAC Nominal dome pressure at TPO RTP is identical to CLTP Small Increase in core pressure drop of < 1 psi Small decrease in Recirculation fluid temperature Negligible change in pipe stress Negligible effect on pipe supports Minor increase in the potential for FAC  (FAC concerns are covered by existing piping monitoring program) For the MS and FW lines, supports, and connected lines, the methodologies as described in TLTR Section 5.5.2 and Appendix K were used to determine the percent increases in applicable ASME Code stresses, displacements, CUF, a nd pipe interface component loads (including NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-13 supports) as a function of percentage increase in pressure (where applicable), temperature, and flow due to TPO conditions. As necessary, the percentage increases were applied to the highest calculated stresses, displacements, and the CUF at applicable piping system node points to conservatively determine the maximum TPO calculated stresses, displacements and usage factors. This approach is conservative because the TPO does not affect weight and all building filtered loads (i.e., seismic loads are not affected by the TPO). The factors were also applied to nozzle load, support loads, penetration loads, valves, pumps, heat ex changers and anchors so that these components could be evaluated for acceptability, where required. No new computer codes were used or new assumptions in troduced for this evaluation. MS and Attached Piping System Evaluation The MS piping system (Inside Containment) was evaluated for compliance with the ASME code stress criteria, and for the effects of thermal displacements on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, flanges and valves were also evaluated.
Pipe Stresses The evaluation shows that the increase in flow associated with the TPO uprate does not result in load limits being exceeded for the MS piping system or for the RPV nozzles. The current licensing basis design analyses have sufficient design margin between calculated stresses and ASME Code allowable limits to justify operation at the TPO uprate conditions. The temperature of the MS piping (Inside Containment) is unchanged for the TPO.
Pipe Stresses The evaluation shows that the incr ease in flow associated with the TPO uprate doe s not result in load limits being exceeded for the MS piping system or for the RPV nozzles. The current licensing basis design analyses have sufficient design margin between calculated stresses and ASME Code allowable limits to justify operation at the TPO uprate conditions. The temperature of the MS piping (Inside Containm ent) is unchanged for the TPO.
The design adequacy evaluation results show that the requirements of ASME, Section III, Subsection NB/ND (as applicable) requirements are satisfied for the evaluated piping systems.
The design adequacy evaluation results show that the requirements of ASME, Section III, Subsection NB/ND (as applicable) requirements are satisfied for the evaluated piping systems.
Therefore, the TPO does not have an adverse effect on the MS piping design.
Therefore, the TPO does not have an adverse effect on the MS piping design.
Pipe Supports  
Pipe Supports The current licensing basis MS piping was reviewed for the effects of transient loading on the piping snubbers, hangers, struts, and pipe whip restraints. A review of the increases in MS flow associated with the TPO uprate indicates that piping load changes do not result in any load limit being exceeded.
Erosion / Corrosion The carbon steel MS piping can be affected by FAC. FAC is affected by changes in fluid velocity, temperature and moisture content. Limerick has an established FAC monitoring program for monitoring pipe wall thinning in single and two-phase high-energy carbon steel piping. The variation in velocity, temperature, and moisture content resulting from the TPO uprate are minor changes to parameters affecting FAC. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the MS and attached piping indicates minimal 3-13


The current licensing basis MS piping was reviewed for the effects of transient loading on the piping snubbers, hangers, struts, and pipe whip restraints. A review of the increases in MS flow associated with the TPO uprate indicates that pipi ng load changes do not result in any load limit being exceeded.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION effect. Table 3-9 shows piping line segments that are recommended for additional review under the station FAC program.
Erosion / Corrosion  The carbon steel MS piping can be affected by FAC. FAC is affected by changes in fluid velocity, temperature and moisture content. Limerick has an established FAC monitoring program for monitoring pipe wall thinning in single and two-phase high-energy carbon steel piping. The variation in velocity, temperature, and moisture content resulting from the TPO uprate are minor changes to parameters affecting FAC. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the MS and attached piping indicates minimal NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-14 effect. Table 3-9 shows piping line segments that are recomme nded for additional review under the station FAC program. The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO uprate has no adverse effect on high-energy piping systems potentially su sceptible to pipe wall thinning due to FAC.
The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO uprate has no adverse effect on high-energy piping systems potentially susceptible to pipe wall thinning due to FAC.
FW Piping System Evaluation The current licensing basis FW piping system (Inside Containment) reports were reviewed for compliance with the ASME Section III Code stress criteria, and for the effects of thermal expansion displacements on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, and valves were also evaluated.
FW Piping System Evaluation The current licensing basis FW piping system (Inside Containment) reports were reviewed for compliance with the ASME Section III Code stress criteria, and for the effects of thermal expansion displacements on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, and valves were also evaluated.
Pipe Stresses A review of the change in temperature, pressure, and flow associated with the TPO uprate indicates that piping load changes do not result in load limits being exceeded for the FW piping system or for RPV nozzles. The current licensi ng basis design analyses have adequate design margin between calculated stresses and ASME Code allowable limits to justify operation at the TPO uprate conditions. The design adequacy evaluation shows that the requirements of ASME, Section III, Subsection NB/NC/ND-3600 requirements remain satisfied. Therefore, the TPO does not have an adverse effect on the FW piping design.
Pipe Stresses A review of the change in temperature, pressure, and flow associated with the TPO uprate indicates that piping load changes do not result in load limits being exceeded for the FW piping system or for RPV nozzles. The current licensing basis design analyses have adequate design margin between calculated stresses and ASME Code allowable limits to justify operation at the TPO uprate conditions.
Pipe Supports The TPO does not affect the FW piping snubbers, ha ngers, and struts. A re view of the increase in FW temperature and flow associated with the TPO indicates that piping load changes do not result in any load limit being exceeded at the TPO uprate conditions.
The design adequacy evaluation shows that the requirements of ASME, Section III, Subsection NB/NC/ND-3600 requirements remain satisfied. Therefore, the TPO does not have an adverse effect on the FW piping design.
Erosion / Corrosion The carbon steel FW piping can be affected by FA C. FAC in the FW piping is affected by changes in fluid velocity and temperature. Limerick has an established program for monitoring pipe wall thinning in single and two-phase high-energy carbon st eel piping. The variation in velocity and temperature resulting from the TPO uprate are minor changes to parameters affecting FAC. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of component s susceptible to FAC. For TP O, the evaluation of predicted wall thinning of the FW Piping System indicates minimal effect. Table 3-9 shows piping line segments that are recommended for additional review under the station FAC program.
Pipe Supports The TPO does not affect the FW piping snubbers, hangers, and struts. A review of the increase in FW temperature and flow associated with the TPO indicates that piping load changes do not result in any load limit being exceeded at the TPO uprate conditions.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-15 The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO uprate has no adverse effect on high energy piping systems potentially su sceptible to pipe wall thinning due to FAC.
Erosion / Corrosion The carbon steel FW piping can be affected by FAC. FAC in the FW piping is affected by changes in fluid velocity and temperature. Limerick has an established program for monitoring pipe wall thinning in single and two-phase high-energy carbon steel piping. The variation in velocity and temperature resulting from the TPO uprate are minor changes to parameters affecting FAC. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the FW Piping System indicates minimal effect. Table 3-9 shows piping line segments that are recommended for additional review under the station FAC program.
3.5.2 Balance-of-Plant Piping Evaluation This section addresses the adequacy of the BOP piping design (outside of the RCPB) for operation at the TPO 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).
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Pipe Supports For the condensate, FW, extraction steam, heater drain, and main steam systems, operating system pressures and temperatures under TPO will remain within design ratings. Because there is no change in the MS operating temperature, i.e., from the reactor to the MS stop valves, there is no change in the thermal expansion stress for TPO. For systems with increased operating temperatures (i.e., MS downstream of the stop valves, condensate, feedwater, extraction steam, heater drains), changes to thermal expansion stresses are small and acceptable.
 
Pipe support loads will experience a small increase in the thermal load (< 1%). However, when considering the combination with other loads that are not affected by the TPO uprate (e.g., deadweight) the combined support load increase is insignificant. This pi ping has been analyzed to conditions which envelope operations under TPO. For the MS system piping outside containment, the turbine stop valve (TSV) closure transient was reviewed against conditions that bound operations under TPO.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO uprate has no adverse effect on high energy piping systems potentially susceptible to pipe wall thinning due to FAC.
Available stress and support load margins are adequate to accommodate the increase in loading associated with this fluid transient.
3.5.2   Balance-of-Plant Piping Evaluation This section addresses the adequacy of the BOP piping design (outside of the RCPB) for operation at the TPO 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).
Pipe Supports For the condensate, FW, extraction steam, heater drain, and main steam systems, operating system pressures and temperatures under TPO will remain within design ratings.
Because there is no change in the MS operating temperature, i.e., from the reactor to the MS stop valves, there is no change in the thermal expansion stress for TPO. For systems with increased operating temperatures (i.e., MS downstream of the stop valves, condensate, feedwater, extraction steam, heater drains), changes to thermal expansion stresses are small and acceptable.
Pipe support loads will experience a small increase in the thermal load (< 1%). However, when considering the combination with other loads that are not affected by the TPO uprate (e.g.,
deadweight) the combined support load increase is insignificant. This piping has been analyzed to conditions which envelope operations under TPO.
For the MS system piping outside containment, the turbine stop valve (TSV) closure transient was reviewed against conditions that bound operations under TPO. Available stress and support load margins are adequate to accommodate the increase in loading associated with this fluid transient.
For the FW system piping outside containment, changes to fluid transient loading such as for feed pump trip are small. The station design for fluid transients was reviewed and no changes are required for TPO.
For the FW system piping outside containment, changes to fluid transient loading such as for feed pump trip are small. The station design for fluid transients was reviewed and no changes are required for TPO.
Erosion / Corrosion The integrity of high-energy piping system s is assured by proper design in accordance with the applicable codes and standards. Piping thickness of carbon steel components can be affected by FAC. Limerick has an established program for monitoring pipe wall thinni ng in single phase and two-phase high-energy carbon steel piping. FAC rates may be influenced by changes in fluid velocity, temperature, and moisture content. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the BOP piping indicates minimal effect. Table 3-9 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-16 shows piping line segments that are recommen ded for additional review under the station FAC program. Operation at the TPO RTP results in some changes to parameters affecting FAC in those systems associated with the turbine cycle (e.g., condensate, FW, MS). The evaluation of and inspection for FAC in BOP systems is addressed by compliance with Generic Letter (GL) 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning."  The plant FAC program currently monitors the affected systems. Continued monitoring of the systems provides confidence in the integrity of susceptible high-energy piping systems. Appropriate changes to piping inspection frequency will be implemented to ensure adequate margin exists for those systems with changing process conditions. This action takes into consideration adjustments to predicted ma terial loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO has no adverse effect on high-energy piping systems potentially suscepti ble to pipe wall thinning due to FAC. 3.6 REACTOR RECIRCULATION S YSTEM  The Reactor Recirculation System (RRS) evaluation process is described in TLTR Section 5.6.2.
Erosion / Corrosion The integrity of high-energy piping systems is assured by proper design in accordance with the applicable codes and standards. Piping thickness of carbon steel components can be affected by FAC. Limerick has an established program for monitoring pipe wall thinning in single phase and two-phase high-energy carbon steel piping. FAC rates may be influenced by changes in fluid velocity, temperature, and moisture content. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the BOP piping indicates minimal effect. Table 3-9 3-15
The TPO uprate has a minor effect on the RRS and its components. The TPO uprate does not require an increase in the maximum core flow. No significant reduction of the maximum flow capability occurs due to the TPO uprate because of the small increase in core pressure drop  
 
(< 1 psi). The effect on pump net positive suction head (NPSH) at TPO conditions is negligible. An evaluation has confirmed that no significant increase in RRS vibration occurs from the TPO operating conditions.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION shows piping line segments that are recommended for additional review under the station FAC program.
The cavitation protection interlock for the recirculation pumps and jet pumps is expressed in terms of FW flow. This interlock is based on s ub-cooling 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 TPO. 3.7 M AIN STEAM LINE FLOW RESTRICTORS The generic evaluation provided in TLTR Appendix J.2.3.7 is applicable to Limerick. The requirements for the MSL flow restrictors rema in unchanged for TPO uprate conditions. No change in steam line break flow rate occurs because the operating pre ssure is unchanged. All safety and operational aspects of the MSL flow restrictors are within previous evaluations. 3.8 M AIN STEAM ISOLATION V ALVES  The generic evaluation provided in TLTR Appendix J.2.3.7 is applicable to Limerick. The requirements for the main steam isolation valves (MSIVs) remain unchanged for TPO uprate conditions. All safety and operational aspects of the MSIVs are within previous evaluations.
Operation at the TPO RTP results in some changes to parameters affecting FAC in those systems associated with the turbine cycle (e.g., condensate, FW, MS). The evaluation of and inspection for FAC in BOP systems is addressed by compliance with Generic Letter (GL) 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning. The plant FAC program currently monitors the affected systems. Continued monitoring of the systems provides confidence in the integrity of susceptible high-energy piping systems. Appropriate changes to piping inspection frequency will be implemented to ensure adequate margin exists for those systems with changing process conditions. This action takes into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO has no adverse effect on high-energy piping systems potentially susceptible to pipe wall thinning due to FAC.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-17 3.9 REACTOR C ORE ISOLATION COOLING The RCIC system provides inventory makeup to the reactor vessel when the vessel is isolated from the normal high-pressure makeup systems. The generic evaluation provided in TLTR Section 5.6.7 is applicable to Limerick. Th e TPO uprate does not affect the RCIC system operation, initiation, or capability requirements. 3.10 RESIDUAL H EAT REMOVAL S YSTEM  The Residual Heat Removal (RHR) system is de signed to restore and maintain the coolant inventory in the reactor vessel and to remove sensible and decay heat from the primary system and containment following reactor shutdown for both normal and post accident conditions. The RHR system is designed to function in several operating modes. The generic evaluation provided in TLTR Sections 5.6.4 and Appendices J.2.3.1 and J.2.3.13 are applicable to Limerick. The following table summarizes the effect of the TPO on the design basis of the RHR system. Operating Mode Key Function TPO Evaluation LPCI Mode Core Cool ing See Section 4.2.4 Suppression Pool Cooling (SPC) and Containment Spray Cooling (CSC)
3.6   REACTOR RECIRCULATION SYSTEM The Reactor Recirculation System (RRS) evaluation process is described in TLTR Section 5.6.2.
Modes Normal SPC function is to maintain pool temperature below the limit. For Abnormal events or accidents, the SPC mode maintains the long-term pool temperature below the design limit. The CSC mode sprays water into the containment to reduce post-accident containment pressure and temperature. Containment Analyses have been performed at 102% of CLTP. Shutdown Cooling (SDC) Mode Removes sensible and decay heat from the reactor primary system during a normal reactor shutdown. The slightly higher decay heat has negligible effect on the SDC mode, which has no safety function.
The TPO uprate has a minor effect on the RRS and its components. The TPO uprate does not require an increase in the maximum core flow. No significant reduction of the maximum flow capability occurs due to the TPO uprate because of the small increase in core pressure drop
Steam Condensing Mode Decay Heat removal Limerick does not have a Steam Condensing Mode of RHR. Fuel Pool Cooling Assist Supplemental fuel pool cooling in the event that the fuel pool heat load exceeds the heat removal capability of the Fuel Pool Cooling system. See Section 6.3.1 The ability of the RHR system to perform required safety functions is demonstrated with analyses based on 102% of CLTP.
(< 1 psi). The effect on pump net positive suction head (NPSH) at TPO conditions is negligible.
Therefore, all safety aspects of the RHR system are within previous evaluations. The requirements for the RHR system remain unchanged for TPO uprate conditions. 3.11 REACTOR W ATER CLEANUP S YSTEM  The generic evaluation of the RWCU system provided in TLTR Sections 5.6.6 and J.2.3.4 is applicable to Limerick. The performance requirements of the RWCU system are negligibly NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-18 affected by TPO uprate. There is no significan t effect on operating temp erature and pressure conditions in the high-pressure portion of the system. RWCU flow is not changed for TPO conditions. Steady-state power level changes for much larger power uprates have shown no effect on reactor water chemistry and the performance of the RWCU system. Power transients that result in crud bursts causing high intermediate loading on the system capacity are the primary source of challenge to the system, so safety and operational aspects of water chemistry performance are not affected by the TPO.
An evaluation has confirmed that no significant increase in RRS vibration occurs from the TPO operating conditions.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-19 Table 3-1 Limerick Unit 1 Upper Shelf Energy 40-Year License (32 EFPY)
The cavitation protection interlock for the recirculation pumps and jet pumps is expressed in terms of FW flow. This interlock is based on sub-cooling 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 TPO.
LocationHeatTest TemperatureShearInitial Longitudinal USEInitia l Transverse USE  
3.7   MAIN STEAM LINE FLOW RESTRICTORS The generic evaluation provided in TLTR Appendix J.2.3.7 is applicable to Limerick. The requirements for the MSL flow restrictors remain unchanged for TPO uprate conditions. No change in steam line break flow rate occurs because the operating pressure is unchanged. All safety and operational aspects of the MSL flow restrictors are within previous evaluations.
[1]%Cu32 EFPY 1/4T Fluence% Decrease USE
3.8   MAIN STEAM ISOLATION VALVES The generic evaluation provided in TLTR Appendix J.2.3.7 is applicable to Limerick. The requirements for the main steam isolation valves (MSIVs) remain unchanged for TPO uprate conditions. All safety and operational aspects of the MSIVs are within previous evaluations.
[2]32 EFPY Transverse USE [3]32 EFPY Equivalent Margin Results [4,6](&deg;F)%(ft-lb)(ft-lb)(n/cm 2)(ft-lb)Plates:LowerC7688-1 40 50 85 55.30.121.3E+1813.5 4813.5% < 21%C7698-2 40 50 100 650.111.3E+1812.5 57C7688-2 40 70 104 67.60.121.3E+1813.5 58Lower-IntermediateC7689-1 40 60 93 60.50.111.3E+1812.5 53C7677-1 40 40 71 46.20.111.3E+1812.5 4012.5% < 21%C7698-1 40 50 96 62.40.111.3E+1812.5 55Welds:Vertical: BE411A3531/H004A27A 10 60 680.021.3E+18 8.5 62BA,BB,BD,BF06L165/F017A27A 10 70 620.031.3E+18 10 56BA,BD,BE,BF662A746/H013A27A-20 65 950.031.3E+18 10 86BA,BB,BC3P4000/3932-989 10 80 970.021.3E+18 8.5 89 BFS3986/RUN 934 10 40 510.0541.3E+1812.5 4522% < 34%BA,BB,BE1P4218/3929-989 10 83 1020.0531.3E+18 12 90TEST PLATE421A6811/F022A27A 10 75 910.091.3E+1814.5 78Girth:AB07L857/B101A27A 10 50 390.031.3E+18 10 3522% < 34%AB402C4371/C115A27A 10 70 920.021.3E+18 8.5 84AB411A3531/H004A27A 10 60 680.021.3E+18 8.5 62AB09M057/C109A27A 10 50 440.031.3E+18 10 4022% < 34%AB412P3611/J417B27AF130 100 1360.031.3E+18 10 122AB03M014/C118A27A 10 40 470.011.3E+18 7 4422% < 34%ABL83355/S411B27AD 130 100 1500.031.3E+18 10 135AB640892/J424B27AE 130 100 1180.091.3E+1814.5 101AB401P6741/S419B27AG130 100 1360.031.3E+18 10 122AB 5P6756 0 100 1210.0831.3E+18 14 104Nozzles:Water Level InstrumentationForgingSB166 [5]40 40 71 46.20.111.9E+17 8 428% < 21%LPCI NozzleWeld07L669/K004A27A 10 40 540.031.9E+17 6.5 50Weld401Z9711/A022A27A 10 80 1040.021.9E+17 6 98Forging: 45&deg;Q2Q25W-20 30 58 37.70.181.9E+17 11 3411% < 21%Forging: 135&deg;Q2Q35W-20 70 77 50.10.181.9E+17 11 4511% < 21%Forging: 225&deg;Q2Q25W-20 40 60 390.181.9E+17 11 3511% < 21%Forging: 315&deg;Q2Q35W-20 40 41 26.70.181.9E+17 11 2411% < 21%BEST ESTIMATE CHEMISTRIES:per BWRVIP-135 R1 [7]:BA,BB,BC 3P4000/3932-989 10 80 970.021.3E+18 9.5 88 BFS3986/RUN 934 10 40 510.0581.3E+18 12 4522% < 34%BA,BB,BE1P4218/3929-989 10 83 1020.0581.3E+1812.5 89AB 5P6756 0 100 1210.081.3E+1813.5 105Integrated Surveillance Program (ISP):PlateC2761-2 127.20.101.3E+18 15 108Weld 5P6756104.40.06/0.08 [8]1.3E+1813.5 90Notes:supersede the plant-specific chemistries.[8] The first %Cu value is from Appendix B of BWRVIP-135 Revision 1; the second value is from Appendix D and represents the best estimate chemistry. The bounding value (Appendix D) is conservatively used in this evaluation.[7] The best estimate chemistries from BWRVIP-135 Revision 1 are provided in a separate section in order to discern between the original plant-specific properties. It is intended that the best estimate chemistries [6]  These values have been determined using NEDO-32205-A. Note that the weld materials required adjustment because the measured results exceeded Regulatory Guide 1.99, Revision 2 predicted results.[1] Transverse USE for plate and forging materials is obtained using 65% of the longitudinal USE.[3] 32 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE /100)].[4] The initial USE for the materials evaluated in this column is very low due to a lack of sufficient test data. Although conservatively evaluated, this column demonstrates that these materials are bounded by an Equivalent Margin Analysis.[5] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.[2]  USE Decrease is obtained from Regulatory Guide 1.99, Revision 2, Figure 2.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-20 Table 3-2 Limerick Unit 2 Upper Shelf Energy 40-Year License (32 EFPY)LocationHeatTest TemperatureShearInitial Longitudinal USEInitial Transverse USE[1]%Cu32 EFPY 1/4T Fluence% Decrease USE[2]32 EFPY Transverse USE[3]32 EFPY Equivalent Margin Results [4](&deg;F)%(ft-lb)(ft-lb)(n/cm 2)(ft-lb)PLATES:LowerB3312-140507850.70.131.3E+18144414% < 21%B3416-140506139.70.141.3E+1814.53414.5% < 21%
 
C9621-240308957.90.151.3E+18154915% < 21%Lower-IntermediateC9569-240408756.60.111.3E+1812.550 C9526-140408957.90.111.3E+1812.551 C9526-240509763.10.111.3E+1812.555 WELDS:Vertical:BA,BB,BD,BE,BF432A2671/H019A27A-2040 540.041.3E+18114822% < 34%BA,BC03R728/L910A27A 10 70720.031.3E+1810 65BA,BB,BC,BD,BE,BF3P4000/3933(Single Wire) 10 80950.021.3E+188.5 87BA,BB,BC,BD,BE,BF3P4000/3933 (Tandem Wire) 10 98910.021.3E+188.5 83 BB401Z9711/A022A27A10801040.021.3E+188.595BC662A746/H013A27A-2065950.031.3E+1810 86BC402A0462/B023A27A 10 62860.021.3E+188.5 79BD,BE09L853/A111A27A 10 60790.031.3E+1810 71BC,BD,BE,BF07L669/K004A27A-2040540.031.3E+1810 4922% < 34%Girth:AB07L857/B101A27A1050 390.031.3E+18103522% < 34%ABL83355/S411B27AD1301001500.031.3E+1810135AB402C4371/C115A27A 10 70920.021.3E+188.5 84AB03M014/C118A27A 10 40470.011.3E+187 4422% < 34%AB411A3531/H004A27A1060 680.021.3E+188.562AB09M057/C109A27A 10 50440.031.3E+1810 4022% < 34%AB640892/J424B27AE1301001180.091.3E+1814.5101AB401P6741/S419B27AG1301001360.031.3E+1810122AB412P3611/J417B27AF1301001360.031.3E+1810122NOZZLES:Water Level InstrumentationForgingSB166 [5]4040 8756.60.111.9E+1710.551LPCIWeld KAC3L46C/J020A27A1060 400.021.9E+1763814% < 34%Weld KA422B7201/L030A27A-2030380.041.9E+177.5 3514% < 34%Weld KA4P4784/3930 (Single Wire)4090 970.061.9E+178.589Weld KA4P4784/3930 (Tandem Wire)4075 950.061.9E+178.587Forging 892L-1Q2Q33W-2040 4629.90.151.9E+17102710% < 21%Forging 892L-2Q2Q33W-204043280.151.9E+1710 2510% < 21%Forging 892L-3Q2Q33W-2050 5334.50.151.9E+17103110% < 21%Forging 892L-4Q2Q33W-2050 4831.20.151.9E+17102810% < 21%BEST ESTIMATE CHEMISTRIESper BWRVIP-135 R1 [7]:BA,BB,BC,BD,BE,BF3P4000/3933(Single Wire)1080950.021.3E+189.586BA,BB,BC,BD,BE,BF3P4000/3933 (Tandem Wire) 10 98910.021.3E+189.5 82Weld [6]CTY538/A027A27A1070 94 610.031.3E+181055Weld5P6756104.40.081.3E+1813.590INTEGRATED SURVEILLANCE PROGRAM (ISP):
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.9   REACTOR CORE ISOLATION COOLING The RCIC system provides inventory makeup to the reactor vessel when the vessel is isolated from the normal high-pressure makeup systems. The generic evaluation provided in TLTR Section 5.6.7 is applicable to Limerick. The TPO uprate does not affect the RCIC system operation, initiation, or capability requirements.
PlateB0673-1158.10.151.3E+1815134Weld5P6756104.40.06/0.08[8]1.3E+1813.590Notes:[5] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.[8] The first %Cu value is from Appendix B of BWRVIP-135 Revision 1; the second value is from Appendix D and represents the best estimate chemistry. The bounding value (Appendix D) is conservatively used in this evaluation.[7] The best estimate chemistries obtained from BWRVIP-135 Revision 1 have been provided in a separate section in order to discern between the original plant-specific properties. It is intended that the best estimate chemistries supersede the plant-specific chemistries.[6]  CMTR records do not indicate that this is a surveillance weld. However, the CMTRs demonstrate that this heat is a weld in the vessel; therefore, it is evaluated using the best estimate chemistry from BWRVIP-135 Revision 1.[1]  Transverse USE for plate and forging materials is obtained using 65% of the longitudinal USE.[2]  USE Decrease is obtained from Regulatory Guide 1.99, Revision 2, Figure 2.[3] 32 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE /100)].[4]  The initial USE for the materials evaluated in this column is very low due to a lack of sufficient test data. Although conservatively evaluated, this column demonstrates that these materials are bounded by an Equivalent Margin Analysis. The weld materials require adjustment because the measured decrease exceeds the predicted decrease. This has been performed in accordance with Regulatory Guide 1.99, Revision 2.
3.10 RESIDUAL HEAT REMOVAL SYSTEM The Residual Heat Removal (RHR) system is designed to restore and maintain the coolant inventory in the reactor vessel and to remove sensible and decay heat from the primary system and containment following reactor shutdown for both normal and post accident conditions. The RHR system is designed to function in several operating modes. The generic evaluation provided in TLTR Sections 5.6.4 and Appendices J.2.3.1 and J.2.3.13 are applicable to Limerick.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-21 Table 3-3 Limerick Unit 1 Adjusted Reference Temperatures 40-Year License (32 EFPY)
The following table summarizes the effect of the TPO on the design basis of the RHR system.
Plates and WeldsThickness =6.19inches32 EFPY Peak I.D. fluence =1.9E+18n/cm 232 EFPY Peak 1/4 T fluence =1.3E+18n/cm 232 EFPY Peak 1/4 T fluence =1.3E+18n/cm 2Nozzle Forgings and WeldsThickness = 6.19inches32 EFPY Peak I.D. fluence =2.8E+17n/cm 232 EFPY Peak 1/4 T fluence =1.9E+17n/cm 232 EFPY Peak 1/4 T fluence =1.9E+17n/cm 2 Initial1/4 T 32EFPY I32EFPY32EFPYComponentHeat or Heat/Lot%Cu%Ni CF RTNDTFluenceRTNDTMarginShiftART
Operating Mode                             Key Function                         TPO Evaluation LPCI Mode                           Core Cooling                                  See Section 4.2.4 Suppression Pool Cooling (SPC) and   Normal SPC function is to maintain pool      Containment Analyses have Containment Spray Cooling (CSC)     temperature below the limit.                 been performed at 102% of Modes                                                                              CLTP.
&deg;Fn/cm 2&deg;F&deg;F&deg;F&deg;FPLATES:Lower14-1C7688-10.120.5181101.3E+183801734728214-2C7698-20.110.4873101.3E+183501734697914-3C7688-20.120.5181101.3E+1838017347282Lower-Intermediate17-1C7689-10.110.4873101.3E+183501734697917-2C7677-10.110.5073201.3E+183501734698917-3C7698-10.110.4873101.3E+1835017346979WELDS:Vertical (shop) [1]BE411A3531/H004A27A0.020.9627-501.3E+1813061326-24BA,BB,BD,BF06L165/F017A27A0.030.9941-501.3E+18190101939-11BA,BD,BE,BF662A746/H013A27A0.030.8841-201.3E+1819010193919BA,BB,BC3P4000/3932-989 [2]0.020.92827-501.3E+1813061326-24BFS3986/RUN 934 [2]0.0540.96974-421.3E+1835018357028BA,BB,BE1P4218/3929-989 [2]0.0530.8972-501.3E+1834017346818TEST PLATE421A6811/F022A27A0.090.81122-501.3E+18580285611464Girth (field)
For Abnormal events or accidents, the SPC mode maintains the long-term pool temperature below the design limit.
AB07L857/B101A27A0.030.9741-61.3E+1819010193933 AB402C4371/C115A27A0.020.9227-501.3E+1813061326-24 AB411A3531/H004A27A0.020.9627-501.3E+1813061326-24 AB09M057/C109A27A0.030.8941-361.3E+18190101939 3 AB412P3611/J417B27AF0.030.9341-801.3E+18190101939-41 AB03M014/C118A27A0.010.9420-341.3E+18905919-15 ABL83355/S411B27AD0.031.0841-701.3E+18190101939-31 AB640892/J424B27AE0.091.00122-601.3E+18580285611454 AB401P6741/S419B27AG0.030.9241-601.3E+18190101939-21 AB5P6756 [2]0.0830.943112-601.3E+18530265310646NOZZLES:Water Level InstrumentationForgingSB166 [4]0.110.5073201.9E+171206122545LPCI NozzleWeld07L669/K004A27A0.031.0241-501.9E+17703714-36Weld401Z9711/A022A27A0.020.8327-501.9E+1750259-41Forging: 45&deg; & 225&deg;Q2Q25W0.180.78140-61.9E+1724012244741Forging:  135&deg; & 315&deg;Q2Q35W0.180.85142-81.9E+1724012244840BEST ESTIMATE CHEMISTRIESper BWRVIP-135 R1 [3]:BA,BB,BC3P40000.020.93527-501.3E+1813061326-24BFS39860.0580.94979.2-421.3E+1837019377533BA,BB,BE1P42180.0580.86579.2-501.3E+1837019377525 AB5P6756 [8]0.080.936108-601.3E+18510265110242 AB5P6756 [8]0.080.936154[7]-601.3E+18730142810141INTEGRATED SURVEILLANCE PROGRAM:PlateC2761-2 [5]0.100.5465201.3E+1831015316282Weld5P6756 [6]0.06/0.080.93/0.936154[7]-601.3E+18730142810141Notes:[1] Welds BA, BB, BC occur in the lower shell (Shell Ring #1) and welds BD, BE, BF occur in the lower-intermediate shell (Shell Ring #2).[3] It is intended that the best estimate chemistries supersede the plant-specific chemistries.[4] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.[5] The ISP plate material is NOT the same heat as the target vessel plate material. The results provided are for information only and do not affect the Limerick Unit 1 PT curves.[6] The ISP weld material is NOT the same heat as the target vessel weld material. However, because this heat does occur in the beltline region, the material is evaluated as   defined in Section 3 of BWRVIP-135 Revision 1, and considered applicable to the beltline region.[8]  This heat is presented with the CF prior to adjustment and after adjustment in order to provide both sets of data.[2]  This material is evaluated below using the best estimate chemistry for this heat of material as provided in BWRVIP-135, Revision 1.[7] The Adjusted CF is calculated as: (108/82)
The CSC mode sprays water into the containment to reduce post-accident containment pressure and temperature.
Shutdown Cooling (SDC) Mode         Removes sensible and decay heat from the     The slightly higher decay heat reactor primary system during a normal       has negligible effect on the SDC reactor shutdown.                            mode, which has no safety function.
Steam Condensing Mode               Decay Heat removal                           Limerick does not have a Steam Condensing Mode of RHR.
Fuel Pool Cooling Assist             Supplemental fuel pool cooling in the event   See Section 6.3.1 that the fuel pool heat load exceeds the heat removal capability of the Fuel Pool Cooling system.
The ability of the RHR system to perform required safety functions is demonstrated with analyses based on 102% of CLTP. Therefore, all safety aspects of the RHR system are within previous evaluations. The requirements for the RHR system remain unchanged for TPO uprate conditions.
3.11 REACTOR WATER CLEANUP SYSTEM The generic evaluation of the RWCU system provided in TLTR Sections 5.6.6 and J.2.3.4 is applicable to Limerick. The performance requirements of the RWCU system are negligibly 3-17
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION affected by TPO uprate. There is no significant effect on operating temperature and pressure conditions in the high-pressure portion of the system. RWCU flow is not changed for TPO conditions. Steady-state power level changes for much larger power uprates have shown no effect on reactor water chemistry and the performance of the RWCU system. Power transients that result in crud bursts causing high intermediate loading on the system capacity are the primary source of challenge to the system, so safety and operational aspects of water chemistry performance are not affected by the TPO.
3-18
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-1 Limerick Unit 1 Upper Shelf Energy 40-Year License (32 EFPY)
Initial            Initial Test                              Longitudinal     Transverse USE                   32 EFPY 1/4T                    32 EFPY Transverse    32 EFPY Equivalent
[1]
Location                                        Heat                Temperature            Shear              USE                              %Cu          Fluence     % Decrease USE [2]       USE [3]       Margin Results [4,6]
2
(&deg;F)               %               (ft-lb)           (ft-lb)                       (n/cm )                             (ft-lb)
Plates:
Lower                                                                    C7688-1                     40                 50                 85               55.3          0.12          1.3E+18            13.5                 48              13.5% < 21%
C7698-2                     40                 50               100                 65            0.11          1.3E+18            12.5                 57 C7688-2                     40                 70               104               67.6          0.12          1.3E+18            13.5                 58 Lower-Intermediate                                                        C7689-1                     40                 60                 93               60.5          0.11          1.3E+18            12.5                 53 C7677-1                     40                 40                 71               46.2          0.11          1.3E+18            12.5                 40              12.5% < 21%
C7698-1                     40                 50                 96               62.4          0.11          1.3E+18            12.5                 55 Welds:
Vertical:
BE                                  411A3531/H004A27A                   10                 60                                   68            0.02          1.3E+18             8.5                 62 BA,BB,BD,BF                                06L165/F017A27A                   10                 70                                   62            0.03          1.3E+18             10                 56 BA,BD,BE,BF                              662A746/H013A27A                 -20                 65                                   95            0.03          1.3E+18             10                 86 BA,BB,BC                                  3P4000/3932-989                   10                 80                                   97            0.02          1.3E+18             8.5                 89 BF                                      S3986/RUN 934                   10                 40                                   51          0.054          1.3E+18            12.5                 45              22% < 34%
BA,BB,BE                                  1P4218/3929-989                   10                 83                                   102          0.053          1.3E+18             12                 90 TEST PLATE                              421A6811/F022A27A                   10                 75                                   91            0.09          1.3E+18            14.5                 78 Girth:
AB                                    07L857/B101A27A                   10                 50                                   39            0.03          1.3E+18             10                 35              22% < 34%
AB                                  402C4371/C115A27A                   10                 70                                   92            0.02          1.3E+18             8.5                 84 AB                                  411A3531/H004A27A                   10                 60                                   68            0.02          1.3E+18             8.5                 62 AB                                    09M057/C109A27A                   10                 50                                   44            0.03          1.3E+18             10                 40              22% < 34%
AB                                  412P3611/J417B27AF                130                100                                   136            0.03          1.3E+18             10                 122 AB                                    03M014/C118A27A                   10                 40                                   47            0.01          1.3E+18             7                 44              22% < 34%
AB                                  L83355/S411B27AD                 130                 100                                   150            0.03          1.3E+18             10                 135 AB                                    640892/J424B27AE                 130                 100                                   118            0.09          1.3E+18            14.5               101 AB                                  401P6741/S419B27AG                130                100                                   136            0.03          1.3E+18             10                 122 AB                                          5P6756                       0               100                                   121          0.083          1.3E+18             14                 104 Nozzles:
Water Level Instrumentation Forging                                      SB166 [5]                     40                 40                 71               46.2          0.11          1.9E+17             8                  42              8% < 21%
LPCI Nozzle Weld                                  07L669/K004A27A                   10                 40                                   54            0.03          1.9E+17             6.5                 50 Weld                                  401Z9711/A022A27A                   10                 80                                   104            0.02          1.9E+17             6                 98 Forging: 45&deg;                                   Q2Q25W                     -20                 30                 58               37.7          0.18          1.9E+17             11                  34              11% < 21%
Forging: 135&deg;                                   Q2Q35W                     -20                 70                 77               50.1          0.18          1.9E+17             11                  45              11% < 21%
Forging: 225&deg;                                   Q2Q25W                     -20                 40                 60                 39            0.18          1.9E+17             11                  35              11% < 21%
Forging: 315&deg;                                   Q2Q35W                     -20                 40                 41               26.7          0.18          1.9E+17             11                  24              11% < 21%
BEST ESTIMATE CHEMISTRIES:
per BWRVIP-135 R1 [7]:
BA,BB,BC                                 3P4000/3932-989                   10                 80                                   97            0.02          1.3E+18             9.5                 88 BF                                      S3986/RUN 934                   10                 40                                   51          0.058          1.3E+18             12                 45              22% < 34%
BA,BB,BE                                  1P4218/3929-989                   10                 83                                   102          0.058          1.3E+18            12.5                 89 AB                                          5P6756                       0               100                                   121            0.08          1.3E+18            13.5               105 Integrated Surveillance Program (ISP):
Plate                                      C2761-2                                                                             127.2          0.10          1.3E+18             15                 108 Weld                                        5P6756                                                                              104.4      0.06/0.08 [8]     1.3E+18            13.5                 90 Notes:
[1] Transverse USE for plate and forging materials is obtained using 65% of the longitudinal USE.
[2] USE Decrease is obtained from Regulatory Guide 1.99, Revision 2, Figure 2.
[3] 32 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE /100)].
[4] The initial USE for the materials evaluated in this column is very low due to a lack of sufficient test data. Although conservatively evaluated, this column demonstrates that these materials are bounded by an Equivalent Margin Analysis.
[5] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.
[6] These values have been determined using NEDO-32205-A. Note that the weld materials required adjustment because the measured results exceeded Regulatory Guide 1.99, Revision 2 predicted results.
[7] The best estimate chemistries from BWRVIP-135 Revision 1 are provided in a separate section in order to discern between the original plant-specific properties. It is intended that the best estimate chemistries supersede the plant-specific chemistries.
[8] The first %Cu value is from Appendix B of BWRVIP-135 Revision 1; the second value is from Appendix D and represents the best estimate chemistry. The bounding value (Appendix D) is conservatively used in this evaluation.
3-19
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-2 Limerick Unit 2 Upper Shelf Energy 40-Year License (32 EFPY) 32 EFPY Test                        Initial            Initial                  32 EFPY 1/4T   % Decrease     Transverse    32 EFPY Equivalent Margin
[1]                                      [2]             [3]                   [4]
Location                                            Heat                      Temperature      Shear    Longitudinal USE Transverse USE          %Cu        Fluence          USE            USE                Results
(&deg;F)         %           (ft-lb)             (ft-lb)                     (n/cm2)                         (ft-lb)
PLATES:
Lower                                                                      B3312-1                            40          50            78                50.7          0.13        1.3E+18          14              44              14% < 21%
B3416-1                            40          50            61                39.7          0.14        1.3E+18          14.5            34              14.5% < 21%
C9621-2                            40          30            89                57.9          0.15        1.3E+18          15              49              15% < 21%
Lower-Intermediate                                                        C9569-2                            40          40            87                56.6          0.11        1.3E+18          12.5            50 C9526-1                            40          40            89                57.9          0.11        1.3E+18          12.5            51 C9526-2                            40          50            97                63.1          0.11        1.3E+18          12.5            55 WELDS:
Vertical:
BA,BB,BD,BE,BF                                432A2671/H019A27A                       -20          40                                54            0.04        1.3E+18          11              48              22% < 34%
BA,BC                                      03R728/L910A27A                       10           70                                72            0.03        1.3E+18          10              65 BA,BB,BC,BD,BE,BF                            3P4000/3933(Single Wire)                   10           80                                95            0.02        1.3E+18          8.5             87 BA,BB,BC,BD,BE,BF                          3P4000/3933 (Tandem Wire)                     10           98                                91            0.02        1.3E+18          8.5             83 BB                                      401Z9711/A022A27A                        10          80                                104            0.02        1.3E+18          8.5            95 BC                                      662A746/H013A27A                       -20          65                                95            0.03        1.3E+18          10              86 BC                                      402A0462/B023A27A                       10           62                                86            0.02        1.3E+18          8.5             79 BD,BE                                      09L853/A111A27A                       10           60                                79            0.03        1.3E+18          10              71 BC,BD,BE,BF                                    07L669/K004A27A                       -20          40                                54            0.03        1.3E+18          10              49              22% < 34%
Girth:
AB                                        07L857/B101A27A                        10          50                                39            0.03        1.3E+18          10              35              22% < 34%
AB                                      L83355/S411B27AD                      130          100                                150            0.03        1.3E+18          10            135 AB                                    402C4371/C115A27A                         10           70                                92            0.02        1.3E+18          8.5             84 AB                                      03M014/C118A27A                         10           40                                47            0.01        1.3E+18            7              44              22% < 34%
AB                                    411A3531/H004A27A                        10          60                                68            0.02        1.3E+18          8.5            62 AB                                      09M057/C109A27A                         10           50                                44            0.03        1.3E+18          10              40              22% < 34%
AB                                      640892/J424B27AE                      130          100                                118            0.09        1.3E+18          14.5            101 AB                                    401P6741/S419B27AG                      130          100                                136            0.03        1.3E+18          10            122 AB                                    412P3611/J417B27AF                      130          100                                136            0.03        1.3E+18          10            122 NOZZLES:
Water Level Instrumentation Forging                                          SB166 [5]                         40          40            87                56.6          0.11        1.9E+17          10.5            51 LPCI Weld KA                                      C3L46C/J020A27A                        10          60                                40            0.02        1.9E+17            6              38              14% < 34%
Weld KA                                    422B7201/L030A27A                       -20          30                                38            0.04        1.9E+17          7.5             35              14% < 34%
Weld KA                                4P4784/3930 (Single Wire)                   40          90                                97            0.06        1.9E+17          8.5            89 Weld KA                              4P4784/3930 (Tandem Wire)                     40          75                                95            0.06        1.9E+17          8.5            87 Forging 892L-1                                      Q2Q33W                            -20          40            46                29.9          0.15        1.9E+17          10              27              10% < 21%
Forging 892L-2                                      Q2Q33W                            -20          40            43                  28            0.15        1.9E+17          10              25              10% < 21%
Forging 892L-3                                      Q2Q33W                            -20          50            53                34.5          0.15        1.9E+17          10              31              10% < 21%
Forging 892L-4                                      Q2Q33W                            -20           50            48                31.2          0.15        1.9E+17          10              28              10% < 21%
BEST ESTIMATE CHEMISTRIES per BWRVIP-135 R1 [7]:
BA,BB,BC,BD,BE,BF                            3P4000/3933(Single Wire)                   10          80                                95            0.02        1.3E+18          9.5            86 BA,BB,BC,BD,BE,BF                          3P4000/3933 (Tandem Wire)                     10           98                                91            0.02        1.3E+18          9.5             82 Weld [6]                                     CTY538/A027A27A                        10          70            94                 61            0.03        1.3E+18          10              55 Weld                                            5P6756                                                                          104.4          0.08        1.3E+18          13.5            90 INTEGRATED SURVEILLANCE PROGRAM (ISP):
Plate                                          B0673-1                                                                          158.1          0.15        1.3E+18          15            134 Weld                                            5P6756                                                                          104.4      0.06/0.08[8]   1.3E+18          13.5            90 Notes:
[1] Transverse USE for plate and forging materials is obtained using 65% of the longitudinal USE.
[2] USE Decrease is obtained from Regulatory Guide 1.99, Revision 2, Figure 2.
[3] 32 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE /100)].
[4] The initial USE for the materials evaluated in this column is very low due to a lack of sufficient test data. Although conservatively evaluated, this column demonstrates that these materials are bounded by an Equivalent Margin Analysis. The weld materials require adjustment because the measured decrease exceeds the predicted decrease. This has been performed in accordance with Regulatory Guide 1.99, Revision 2.
[5] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.
[6] CMTR records do not indicate that this is a surveillance weld. However, the CMTRs demonstrate that this heat is a weld in the vessel; therefore, it is evaluated using the best estimate chemistry from BWRVIP-135 Revision 1.
[7] The best estimate chemistries obtained from BWRVIP-135 Revision 1 have been provided in a separate section in order to discern between the original plant-specific properties. It is intended that the best estimate chemistries supersede the plant-specific chemistries.
[8] The first %Cu value is from Appendix B of BWRVIP-135 Revision 1; the second value is from Appendix D and represents the best estimate chemistry. The bounding value (Appendix D) is conservatively used in this evaluation.
3-20
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-3 Limerick Unit 1 Adjusted Reference Temperatures 40-Year License (32 EFPY)
Plates and Welds 2
Thickness =              6.19              inches                                                                        32 EFPY Peak I.D. fluence =  1.9E+18      n/cm 2
32 EFPY Peak 1/4 T fluence =   1.3E+18      n/cm 2
32        EFPY Peak 1/4 T fluence =   1.3E+18      n/cm Nozzle Forgings and Welds 2
Thickness =               6.19              inches                                                                        32 EFPY Peak I.D. fluence =   2.8E+17      n/cm 2
32 EFPY Peak 1/4 T fluence =   1.9E+17      n/cm 2
32        EFPY Peak 1/4 T fluence =   1.9E+17      n/cm Initial      1/4 T       32 EFPY            I                        32 EFPY        32 EFPY Component                        Heat or Heat/Lot             %Cu         %Ni         CF           RTNDT        Fluence            RTNDT                        Margin        Shift          ART 2
                                                                                                                        &deg;F        n/cm           &deg;F                                 &deg;F           &deg;F           &deg;F PLATES:
Lower 14-1                              C7688-1                  0.12      0.51        81              10      1.3E+18            38            0          17      34            72            82 14-2                              C7698-2                  0.11      0.48        73              10      1.3E+18            35            0          17      34            69            79 14-3                              C7688-2                  0.12      0.51        81              10      1.3E+18            38            0          17      34            72            82 Lower-Intermediate 17-1                              C7689-1                  0.11      0.48        73              10      1.3E+18            35            0          17      34            69            79 17-2                              C7677-1                  0.11      0.50        73              20      1.3E+18            35            0          17      34            69            89 17-3                              C7698-1                  0.11      0.48        73              10      1.3E+18            35            0          17      34            69            79 WELDS:
Vertical (shop) [1]
BE                      411A3531/H004A27A                0.02      0.96        27              -50      1.3E+18            13            0          6      13            26          -24 BA,BB,BD,BF                      06L165/F017A27A                0.03      0.99        41              -50      1.3E+18            19            0          10      19            39          -11 BA,BD,BE,BF                    662A746/H013A27A                0.03      0.88        41              -20      1.3E+18            19            0          10      19            39            19 BA,BB,BC                    3P4000/3932-989 [2]             0.02      0.928        27              -50      1.3E+18            13            0          6      13            26          -24 BF                        S3986/RUN 934 [2]             0.054      0.969        74              -42        1.3E+18            35            0          18      35            70            28 BA,BB,BE                    1P4218/3929-989 [2]             0.053      0.89        72              -50        1.3E+18            34            0          17      34            68            18 TEST PLATE                    421A6811/F022A27A                0.09      0.81        122              -50      1.3E+18            58            0          28      56          114            64 Girth (field)
AB                          07L857/B101A27A                0.03      0.97        41              -6      1.3E+18            19            0          10      19            39            33 AB                        402C4371/C115A27A                0.02      0.92        27              -50      1.3E+18            13            0          6      13            26          -24 AB                      411A3531/H004A27A                0.02      0.96        27              -50      1.3E+18            13            0          6      13            26          -24 AB                        09M057/C109A27A                0.03      0.89        41              -36      1.3E+18            19            0          10      19            39            3 AB                      412P3611/J417B27AF                0.03      0.93        41              -80      1.3E+18            19            0          10      19            39          -41 AB                        03M014/C118A27A                0.01      0.94        20              -34      1.3E+18              9            0          5      9            19          -15 AB                        L83355/S411B27AD                0.03      1.08        41              -70      1.3E+18            19            0          10      19            39          -31 AB                        640892/J424B27AE                0.09      1.00        122              -60      1.3E+18            58            0          28      56          114            54 AB                      401P6741/S419B27AG                0.03      0.92        41              -60      1.3E+18            19            0          10      19            39          -21 AB                              5P6756 [2]               0.083      0.943        112              -60      1.3E+18            53            0          26      53          106            46 NOZZLES:
Water Level Instrumentation Forging                          SB166 [4]                 0.11      0.50        73              20      1.9E+17            12            0          6      12            25            45 LPCI Nozzle Weld                        07L669/K004A27A                0.03      1.02        41              -50      1.9E+17              7            0          3      7            14          -36 Weld                      401Z9711/A022A27A                0.02      0.83        27              -50      1.9E+17              5            0          2      5            9            -41 Forging: 45&deg; & 225&deg;                     Q2Q25W                    0.18      0.78        140              -6      1.9E+17            24            0          12      24            47            41 Forging: 135&deg; & 315&deg;                     Q2Q35W                    0.18      0.85        142              -8      1.9E+17            24            0          12      24            48            40 BEST ESTIMATE CHEMISTRIES per BWRVIP-135 R1 [3]:
BA,BB,BC                            3P4000                    0.02      0.935        27              -50      1.3E+18            13            0          6      13            26          -24 BF                                S3986                  0.058      0.949        79.2             -42      1.3E+18            37            0          19      37            75            33 BA,BB,BE                            1P4218                  0.058      0.865        79.2             -50      1.3E+18            37            0          19      37            75            25 AB                            5P6756 [8]                 0.08      0.936        108              -60      1.3E+18            51            0          26      51          102            42 AB                            5P6756 [8]                 0.08      0.936        154      [7]     -60      1.3E+18            73            0          14      28          101            41 INTEGRATED SURVEILLANCE PROGRAM:
Plate                          C2761-2 [5]                 0.10      0.54        65              20      1.3E+18            31            0          15      31            62            82 Weld                            5P6756 [6]             0.06/0.08 0.93/0.936      154      [7]     -60      1.3E+18            73            0          14      28          101            41 Notes:
[1] Welds BA, BB, BC occur in the lower shell (Shell Ring #1) and welds BD, BE, BF occur in the lower-intermediate shell (Shell Ring #2).
[2] This material is evaluated below using the best estimate chemistry for this heat of material as provided in BWRVIP-135, Revision 1.
[3] It is intended that the best estimate chemistries supersede the plant-specific chemistries.
[4] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.
[5] The ISP plate material is NOT the same heat as the target vessel plate material. The results provided are for information only and do not affect the Limerick Unit 1 PT curves.
[6] The ISP weld material is NOT the same heat as the target vessel weld material. However, because this heat does occur in the beltline region, the material is evaluated as defined in Section 3 of BWRVIP-135 Revision 1, and considered applicable to the beltline region.
[7] The Adjusted CF is calculated as: (108/82)
* 116.9 = 154&deg;F. Note that the chemistry values provided represent the BWRVIP-135 Revision 1 Appendix B values (first value) and the BWRVIP-135 Revision 1 Appendix D best estimate chemistry (second value). These are provided for clarity. The Appendix B chemistry results in a CF = 82&deg;F; the Appendix D chemstry results in a CF = 108&deg;F, and the fitted CF from Section 2 = 116.9&deg;F. In accordance with Regulatory Guide 1.99, Revision 2, the  has been reduced by 0.5.
* 116.9 = 154&deg;F. Note that the chemistry values provided represent the BWRVIP-135 Revision 1 Appendix B values (first value) and the BWRVIP-135 Revision 1 Appendix D best estimate chemistry (second value). These are provided for clarity. The Appendix B chemistry results in a CF = 82&deg;F; the Appendix D chemstry results in a CF = 108&deg;F, and the fitted CF from Section 2 = 116.9&deg;F. In accordance with Regulatory Guide 1.99, Revision 2, the  has been reduced by 0.5.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-22 Table 3-4 Limerick Unit 2 Adjusted Reference Temperatures 40-Year License (32 EFPY) Plates and WeldsThickness = 6.19inches32 EFPY Peak I.D. fluence =1.9E+18n/cm 232 EFPY Peak 1/4 T fluence =1.3E+18n/cm 232 EFPY Peak 1/4 T fluence =1.3E+18n/cm 2Nozzle Forgings and WeldsThickness = 6.19inches32 EFPY Peak I.D. fluence =2.8E+17n/cm 232 EFPY Peak 1/4 T fluence =1.9E+17n/cm 232 EFPY Peak 1/4 T fluence =1.9E+17n/cm 2 Initial1/4 T 32EFPY I32EFPY32EFPYComponentHeat or Heat/Lot%Cu%NiCF RTNDTFluenceRTNDT  MarginShiftART&deg;Fn/cm 2&deg;F&deg;F&deg;F&deg;FPLATES:Lower14-1B3312-10.130.5890101.3E+184301734 778714-2B3416-10.140.65101401.3E+184801734 82 12214-3C9621-20.150.60110221.3E+185201734 86 108Lower-Intermediate17-1C9569-20.110.5173101.3E+183501734 697917-2C9526-10.110.5674101.3E+183501734 697917-3C9526-20.110.5674101.3E+183501734 6979WELDS:Vertical [1]BA,BB,BD,BE,BF432A2671/H019A27A0.041.0854-121.3E+182601326 5139BA,BC03R728/L910A27A0.030.9241-501.3E+18190101939-11BA,BB,BC,BD,BE,BF3P4000/3933 [2,7]0.020.92827-501.3E+18130613 26-24BB401Z9711/A022A27A0.020.8327-501.3E+18130613 26-24BC662A746/H013A27A [6]0.030.8841-201.3E+181901019 3919BC402A0462/B023A27A0.020.9027-501.3E+18130613 26-24BD,BE09L853/A111A27A0.030.8641-501.3E+181901019 39-11BC,BD,BE,BF07L669/K004A27A0.031.0241-501.3E+181901019 39-11 GirthAB07L857/B101A27A0.030.9741-61.3E+181901019 3933ABL83355/S411B27AD0.031.0841-701.3E+181901019 39-31AB402C4371/C115A27A0.020.9227-501.3E+18130613 26-24AB03M014/C118A27A0.010.9420-341.3E+18905 9 19-15AB411A3531/H004A27A0.020.9627-501.3E+18130613 26-24AB09M057/C109A27A0.030.8941-361.3E+181901019 39 3AB640892/J424B27AE0.091.00122-601.3E+185802856 11454AB401P6741/S419B27AG0.030.9241-601.3E+181901019 39-21AB412P3611/J417B27AF0.030.9341-801.3E+181901019 39-41NOZZLES:Water Level InstrumentationForgingSB166 [3]0.110.5173101.9E+17120612 2535LPCIWeld KA432A2671/H019A27A0.041.0854-121.9E+17905 9 18 6Weld KA07L669/K004A27A0.031.0241-501.9E+17703 7 14-36Weld KAC3L46C/J020A27A0.020.8727-201.9E+17502 5 9-11Weld KA422B7201/L030A27A0.040.9054-401.9E+17905 9 18-22Weld KA09L853/A111A27A0.030.8641-501.9E+17703 7 14-36Weld KA4P4784/3930 (single wire)0.060.8782-501.9E+17140714 28-22Weld KA4P4784/3930 (tandem wire)0.060.8782-201.9E+1714071428 8Forging 892L-1Q2Q33W0.150.83115-201.9E+171901019 3919Forging 892L-2Q2Q33W0.150.81115-61.9E+1719010193933Forging 892L-3Q2Q33W0.150.82115-41.9E+1719010193935Forging 892L-4Q2Q33W0.150.82115-201.9E+171901019 3919BEST ESTIMATE CHEMISTRIESper BWRVIP-135 R1:BA,BB,BC,BD,BE,BF3P4000/3933 0.020.93527-501.3E+18130613 26-24Weld [8]CTY538/A027A27A0.030.8341-501.3E+181901019 39-11Weld5P6756 [5,10]0.080.936108-61.3E+185102651 10296Weld5P6756 [5,10]0.080.936154[9]-61.3E+187301428 10195INTEGRATED SURVEILLANCE PROGRAM (ISP):PlateB0673-1 [4]0.150.65111401.3E+185301734 87 127Weld5P6756 [5]0.06/0.080.93/0.936154[9]-61.3E+187301428 10195Notes:[1] Welds BA, BB, BC occur in the lower shell (Shell Ring #1) and welds BD, BE, BF occur in the lower-intermediate shell (Shell Ring #2).[2] This material is evaluated below for the best estimate chemistry for this heat of material as provided in BWRVIP-135, Revision 1.[3] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.[4] The ISP plate material is NOT the same heat as the target vessel plate material. The results provided are for information only and do not affect the Limerick Unit 2 PT curves.[5] The ISP weld material is NOT the same heat as the target vessel weld material. The results provided are for information only and do not affect the Limerick Unit 2 PT curves.[6] This heat number was provided in the DIR as 661A746; review of the CMTRs has determined this heat to be 662A746.[8] CMTR records do not indicate that this is a surveillance weld. However, the CMTRs demonstrate that this heat is a weld in the vessel; therefore, it is evaluated using the best estimate chemistry from BWRVIP-135 Revision 1.[9]  The Adjusted CF is calculated as:  (108/82)
[8] This heat is presented with the CF prior to adjustment and after adjustment in order to provide both sets of data.
* 116.9 = 154&deg;F. Note that the chemistry values provide represent the BWRVIP-135 Revision 1 Appendix B values (first value) and the BWRVIP-135 Revision 1 Appendix D best estimate chemistry (second value). These are provided for clarity. The Appendix B chemistry results in a CF = 82&deg;F; the Appendix D chemstry results in a CF = 108&deg;F, and the fitted CF from Section 2 = 116.9&deg;F. In accordance with Regulatory Guide 1.99, Revision 2, the has been reduced by 0.5.[10]  This heat is presented with the CF prior to adjustment and after adjustment in order to provide both sets of data.[7]  3P4000 data is available for both tandem and single wire; there is no change in %Cu. The limiting %Ni is used (single = 0.89; tandem = 0.95). The plant-specific %Ni used is 0.928; this value agrees with the NRC database RVID2.
3-21
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-23 Table 3-5 Limerick 32 EFPY Effects of Irradiation on RPV Axial Weld Properties (CB&I RPV)(CB&I Vessel)(CB&I Vessel)
 
Cu% 0.100.0580.04 Ni% 1.080.951.08 CF 135 79 54 Fluence at clad/weld interface (10 19 n/cm 2)0.690.190.19 RT NDT(U) (&deg;F)-30-42-12  RT NDT  w/o margin (&deg;F)(See Note 3) 121 44 30 Mean RT NDT  (&deg;F) 91 2 18 P (F/E) NRC (See Note 1) 1.42E-01(Note 2)(Note 2)
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-4 Limerick Unit 2 Adjusted Reference Temperatures 40-Year License (32 EFPY)
Notes: [3]    RT NDT = CF
Plates and Welds 2
* f  (0.28 - 0.10 log f ) Parameter [1] P (F/E) stands for "Probabilit y of a failure event." [2] Although a conditional failure probability has not been calculated, the fact that the Limerick values at the end of license are less than the 32 EFPY value provided by the NRC leads to the conclusion that the Limerick RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL 98-05.[4] This data is obtained from GL 98-05.
Thickness =                     6.19                inches                                                                                    32 EFPY Peak I.D. fluence =     1.9E+18      n/cm 2
NRC Staff Assessment for 32 EFPY      (Axial Welds )[4]Limerick Unit 1 32 EFPYLimerick Unit 2              32 EFPY NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-24 Table 3-6 Limerick 32 EFPY Effects of Irradiation on RPV Circumferential Weld Properties (CB&I RPV)(CB&I Vessel)(CB&I Vessel)
32 EFPY Peak 1/4 T fluence =       1.3E+18      n/cm 2
Cu% 0.10.090.09 Ni% 0.991.001.00 CF 134.9 122 122 Fluence at clad/weld interface (10 19 n/cm 2)0.510.190.19 RT NDT(U) (&deg;F) 60-60 RT NDT w/o margin (&deg;F)(See Note 3) 109.5 68 68 Mean RT NDT (&deg;F) 44.5 8 8 P (F/E) NRC (See Note 1) 2.00E-07(Note 2)(Note 2) Notes: [3]   RT NDT  = CF
32            EFPY Peak 1/4 T fluence =     1.3E+18      n/cm Nozzle Forgings and Welds 2
* f  (0.28 - 0.10 log f ) [2] Although a conditional failure probability has not been calculated, the fact that the Limerick values at the end of license are less than the 32 EFPY value provided by the NRC leads to the conclusion that the Limerick RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL 98-05.[1] P (F/E) stands for "Probability of a failure event." Parameter NRC Staff Assessment for 32 EFPY          (Circ Welds )[4]Limerick Unit 1              32 EFPY[4]  This data is obtained from GL 98-05. The CF = 134.9&deg;F was corrected in BWRVIP-05 SE dated 3/7/00 (previously shown to be 109.5&deg;F). Limerick Unit 2 32 EFPY NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-25 Table 3-7 CUF and P+Q Stress Range of Limiting Components P + Q Stress (Kips Per Square Inch (ksi)) CUF[4,5] Component[1, 6] Current (3458 MWt)
Thickness =                     6.19                inches                                                                                    32 EFPY Peak I.D. fluence =     2.8E+17      n/cm 2
TPO (3517 MWt)[3] Allowable (ASME Code Limit) Current (3458 MWt)
32 EFPY Peak 1/4 T fluence =       1.9E+17      n/cm 2
TPO (3517 MWt)[3] Allowable (ASME Code Limit) Feedwater Nozzle[7] 79.6/ 19.4[2] 79.6/ 19.4[2,6]  53.1 0. 9957 0. 9957 1.0 Closure Flange 95.0/ 47.5[2] 95.0/ 47.5[2]  80.1 0.78 0.78 1.0 Closure Bolts 62.7/ 114.7 62.7/ 114.7 79.4 (2S m)/ 118.5 (3S m) 0.95 0.95 1.0 Stabilizer Bracket 79.7 79.7 80.1 0.94 0.94 1.0 Support Skirt Units 1&2 115.0/ 69.7[2] 115.0/
32            EFPY Peak 1/4 T fluence =     1.9E+17      n/cm Initial        1/4 T     32 EFPY            I                              32 EFPY          32 EFPY Component                                Heat or Heat/Lot                 %Cu           %Ni          CF              RTNDT          Fluence          RTNDT                              Margin        Shift              ART 2
69.7[2]  80.1 0.83[4] 0.83 1.0 Steam Outlet Nozzle 37.7 37.7 40.1 0.85 0.85 1.0 LPCI Nozzle  
                                                                                                                                          &deg;F          n/cm         &deg;F                                       &deg;F             &deg;F               &deg;F PLATES:
-/68.5[2] -/68.5[2] 69.9 0.79 0.79 1.0 Core P & Liquid Control Nozzle 100.3/ 44.2[2] 100.3/
Lower 14-1                                    B3312-1                      0.13          0.58          90                10          1.3E+18          43              0            17        34            77                87 14-2                                    B3416-1                      0.14          0.65        101                40          1.3E+18          48              0            17        34            82               122 14-3                                    C9621-2                      0.15          0.60        110                22          1.3E+18          52              0            17        34            86               108 Lower-Intermediate 17-1                                    C9569-2                      0.11          0.51          73                10          1.3E+18          35              0            17        34            69                79 17-2                                    C9526-1                      0.11          0.56          74                10          1.3E+18          35              0            17        34            69                79 17-3                                    C9526-2                      0.11          0.56          74                10          1.3E+18          35              0            17        34            69                79 WELDS:
44.2[2]  69.9 0.71 0.71 1.0 Core Spray Nozzle (Low Alloy Steel) 118.2/ 55.8 118.2/ 55.8 69.9 0.510 0.510 1.0 Notes: 1. There are no changes in operating conditions from CLTP to TPO. Therefore, the CLTP evaluation remains applicable for TPO. The components presented in this table are consistent with the CLTP Safety Analysis Report (NEDC-32265P, "Limerick Generating Station Units 1 and 2 Power Rerate Engineering Report," May 1994) to demonstrate that the results remain unchanged from CLTP to TPO. 2. Thermal Bending included/Thermal bending removed. P + Q stresses are acceptable per CLTP elastic-plastic analysis where applicable, which is valid for TPO conditions. 3. [[                                                                                                                                                                                                                                             
Vertical [1]
BA,BB,BD,BE,BF                          432A2671/H019A27A                    0.04          1.08          54                -12          1.3E+18          26              0            13        26            51                39 BA,BC                                03R728/L910A27A                  0.03          0.92          41                -50          1.3E+18          19              0            10        19            39              -11 BA,BB,BC,BD,BE,BF                          3P4000/3933 [2,7]                 0.02        0.928          27                -50          1.3E+18          13              0            6        13            26               -24 BB                              401Z9711/A022A27A                  0.02          0.83          27                -50          1.3E+18          13              0            6        13            26               -24 BC                            662A746/H013A27A [6]                 0.03          0.88          41                -20          1.3E+18          19              0            10        19            39                19 BC                              402A0462/B023A27A                  0.02          0.90          27                -50          1.3E+18          13              0            6        13            26               -24 BD,BE                                09L853/A111A27A                  0.03          0.86          41                -50          1.3E+18          19              0            10        19            39               -11 BC,BD,BE,BF                              07L669/K004A27A                  0.03          1.02          41                -50          1.3E+18          19              0            10        19            39               -11 Girth AB                                07L857/B101A27A                  0.03          0.97          41                -6          1.3E+18          19              0            10        19            39                33 AB                                L83355/S411B27AD                  0.03          1.08          41                -70          1.3E+18          19              0            10        19            39               -31 AB                              402C4371/C115A27A                  0.02          0.92          27                -50          1.3E+18          13              0            6        13            26               -24 AB                                03M014/C118A27A                    0.01          0.94          20                -34          1.3E+18          9              0            5          9             19               -15 AB                              411A3531/H004A27A                    0.02          0.96          27                -50          1.3E+18          13              0            6        13            26               -24 AB                                09M057/C109A27A                    0.03          0.89          41                -36          1.3E+18          19              0            10        19            39                3 AB                                640892/J424B27AE                  0.09          1.00        122                -60          1.3E+18          58              0            28        56            114                54 AB                              401P6741/S419B27AG                  0.03          0.92          41                -60          1.3E+18          19              0            10        19            39               -21 AB                              412P3611/J417B27AF                  0.03          0.93          41                -80          1.3E+18          19              0            10        19            39               -41 NOZZLES:
Water Level Instrumentation Forging                                    SB166 [3]                     0.11          0.51          73                10          1.9E+17          12              0            6          12            25                35 LPCI Weld KA                            432A2671/H019A27A                    0.04          1.08          54                -12          1.9E+17          9             0            5          9            18                 6 Weld KA                                07L669/K004A27A                  0.03          1.02          41                -50          1.9E+17          7               0            3          7            14               -36 Weld KA                                C3L46C/J020A27A                  0.02          0.87          27                -20          1.9E+17          5               0            2          5            9               -11 Weld KA                              422B7201/L030A27A                  0.04          0.90          54                -40          1.9E+17          9               0            5          9            18              -22 Weld KA                                09L853/A111A27A                  0.03          0.86          41                -50          1.9E+17          7              0            3          7             14               -36 Weld KA                          4P4784/3930 (single wire)             0.06          0.87          82                -50          1.9E+17          14              0            7          14            28               -22 Weld KA                          4P4784/3930 (tandem wire)               0.06          0.87          82                -20          1.9E+17          14              0            7          14            28                8 Forging 892L-1                                Q2Q33W                        0.15          0.83        115                -20          1.9E+17          19              0            10        19            39                19 Forging 892L-2                                Q2Q33W                        0.15          0.81        115                -6          1.9E+17          19              0            10        19            39                33 Forging 892L-3                                Q2Q33W                        0.15          0.82        115                -4          1.9E+17          19              0            10        19            39                35 Forging 892L-4                                Q2Q33W                        0.15          0.82        115                -20          1.9E+17          19              0            10        19            39                19 BEST ESTIMATE CHEMISTRIES per BWRVIP-135 R1:
BA,BB,BC,BD,BE,BF                            3P4000/3933                     0.02        0.935          27                -50          1.3E+18          13              0            6          13            26               -24 Weld [8]                               CTY538/A027A27A                  0.03          0.83          41                -50          1.3E+18          19              0            10        19            39               -11 Weld                                  5P6756 [5,10]                   0.08        0.936        108                -6          1.3E+18          51              0            26        51            102                96 Weld                                  5P6756 [5,10]                   0.08        0.936        154      [9]       -6          1.3E+18          73              0            14        28            101                95 INTEGRATED SURVEILLANCE PROGRAM (ISP):
Plate                                  B0673-1 [4]                   0.15          0.65        111                40          1.3E+18          53              0            17        34            87               127 Weld                                    5P6756 [5]                 0.06/0.08    0.93/0.936      154      [9]       -6          1.3E+18          73              0            14        28            101                95 Notes:
[1] Welds BA, BB, BC occur in the lower shell (Shell Ring #1) and welds BD, BE, BF occur in the lower-intermediate shell (Shell Ring #2).
[2] This material is evaluated below for the best estimate chemistry for this heat of material as provided in BWRVIP-135, Revision 1.
[3] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.
[4] The ISP plate material is NOT the same heat as the target vessel plate material. The results provided are for information only and do not affect the Limerick Unit 2 PT curves.
[5] The ISP weld material is NOT the same heat as the target vessel weld material. The results provided are for information only and do not affect the Limerick Unit 2 PT curves.
[6] This heat number was provided in the DIR as 661A746; review of the CMTRs has determined this heat to be 662A746.
[7] 3P4000 data is available for both tandem and single wire; there is no change in %Cu. The limiting %Ni is used (single = 0.89; tandem = 0.95). The plant-specific %Ni used is 0.928; this value agrees with the NRC database RVID2.
[8] CMTR records do not indicate that this is a surveillance weld. However, the CMTRs demonstrate that this heat is a weld in the vessel; therefore, it is evaluated using the best estimate chemistry from BWRVIP-135 Revision 1.
[9] The Adjusted CF is calculated as: (108/82)
* 116.9 = 154&deg;F. Note that the chemistry values provide represent the BWRVIP-135 Revision 1 Appendix B values (first value) and the BWRVIP-135 Revision 1 Appendix D best estimate chemistry (second value). These are provided for clarity. The Appendix B chemistry results in a CF = 82&deg;F; the Appendix D chemstry results in a CF = 108&deg;F, and the fitted CF from Section 2 = 116.9&deg;F. In accordance with Regulatory Guide 1.99, Revision 2, the  has been reduced by 0.5.
[10] This heat is presented with the CF prior to adjustment and after adjustment in order to provide both sets of data.
3-22
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-5 Limerick 32 EFPY Effects of Irradiation on RPV Axial Weld Properties NRC Staff Assessment for    Limerick Unit 1     Limerick Unit 2 Parameter 32 EFPY        32 EFPY            32 EFPY (Axial Welds )[4]
(CB&I RPV)     (CB&I Vessel)       (CB&I Vessel)
Cu%                                                          0.10          0.058                0.04 Ni%                                                          1.08            0.95                1.08 CF                                                            135              79                  54 19 Fluence at clad/weld interface (10    n/cm 2)               0.69            0.19                0.19 RT NDT(U) (&deg;F)                                               -30            -42                -12 RT NDT w/o margin (&deg;F)(See Note 3)                          121              44                  30 Mean RT NDT (&deg;F)                                              91              2                  18 P (F/E) NRC (See Note 1)                                  1.42E-01        (Note 2)            (Note 2)
Notes:
[1] P (F/E) stands for "Probability of a failure event."
[2] Although a conditional failure probability has not been calculated, the fact that the Limerick values at the end of license are less than the 32 EFPY value provided by the NRC leads to the conclusion that the Limerick RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL 98-05.
[3] RT NDT = CF
* f (0.28 - 0.10 log f )
[4] This data is obtained from GL 98-05.
3-23
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-6 Limerick 32 EFPY Effects of Irradiation on RPV Circumferential Weld Properties NRC Staff Assessment for    Limerick Unit 1    Limerick Unit 2 Parameter 32 EFPY            32 EFPY            32 EFPY (Circ Welds )[4]
(CB&I RPV)        (CB&I Vessel)     (CB&I Vessel)
Cu%                                                           0.1              0.09                0.09 Ni%                                                         0.99              1.00                1.00 CF                                                         134.9               122                 122 19 Fluence at clad/weld interface (10     n/cm 2)               0.51              0.19                0.19 RT NDT(U) (&deg;F)                                               -65                -60                 -60 RT NDT w/o margin (&deg;F)(See Note 3)                         109.5               68                 68 Mean RT NDT (&deg;F)                                             44.5                 8                   8 P (F/E) NRC (See Note 1)                                 2.00E-07           (Note 2)           (Note 2)
Notes:
[1] P (F/E) stands for "Probability of a failure event."
[2] Although a conditional failure probability has not been calculated, the fact that the     Limerick values at the end of license are less than the 32 EFPY value provided by the NRC leads to the conclusion that the Limerick RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL 98-05.
(0.28 - 0.10 log f )
[3] RT NDT = CF
* f
[4] This data is obtained from GL 98-05. The CF = 134.9&deg;F was corrected in BWRVIP-05 SE dated 3/7/00 (previously shown to be 109.5&deg;F).
3-24
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-7 CUF and P+Q Stress Range of Limiting Components P + Q Stress (Kips Per Square Inch (ksi))                             CUF[4,5]
Component[1, 6]       Current             TPO            Allowable      Current              TPO            Allowable (3458 MWt)       (3517 MWt)[3]     (ASME Code     (3458 MWt)       (3517 MWt)[3]         (ASME Limit)                                          Code Limit) 79.6/             79.6/
Feedwater Nozzle[7]                                                            0. 9957           0. 9957           1.0
[2]
19.4              19.4[2,6]
53.1 Closure Flange                95.0/             95.0/                            0.78             0.78             1.0 47.5[2]            47.5[2]
80.1 62.7/             62.7/       79.4 (2Sm)/
Closure Bolts                                                                    0.95             0.95             1.0 114.7              114.7        118.5 (3Sm)
Stabilizer Bracket             79.7             79.7             80.1           0.94             0.94             1.0 Support Skirt Units         115.0/             115.0/                         0.83[4]             0.83             1.0 1&2                              [2]                [2]
69.7              69.7 80.1 Steam Outlet Nozzle            37.7              37.7            40.1          0.85               0.85             1.0 LPCI Nozzle                 -/68.5[2]         -/68.5[2]           69.9           0.79             0.79             1.0 Core P &
100.3/             100.3/
Liquid Control                                                                    0.71             0.71             1.0
[2]
44.2              44.2[2]            69.9 Nozzle Core Spray Nozzle            118.2/             118.2/
69.9           0.510             0.510             1.0 (Low Alloy Steel)              55.8              55.8 Notes:
: 1. There are no changes in operating conditions from CLTP to TPO. Therefore, the CLTP evaluation remains applicable for TPO. The components presented in this table are consistent with the CLTP Safety Analysis Report (NEDC-32265P, Limerick Generating Station Units 1 and 2 Power Rerate Engineering Report, May 1994) to demonstrate that the results remain unchanged from CLTP to TPO.
: 2. Thermal Bending included/Thermal bending removed. P + Q stresses are acceptable per CLTP elastic-plastic analysis where applicable, which is valid for TPO conditions.
: 3.   ((
                                                                            ))
: 4. Limiting CUF is presented.
: 5. Fatigue usage factors are for a 40-year license.
: 6. CLTP and TPO were ((                                                            )) Therefore, there is no change in values from CLTP to TPO.
: 7. Considering normal operating conditions (i.e., does not consider FFWTR or FWHOOS).
3-25


      ]] 4. Limiting CUF is presented. 5. Fatigue usage factors are for a 40-year license. 6. CLTP and TPO were [[                                                                                                              ]] Therefore, there is no change in values from CLTP to TPO. 7. Considering normal operating conditions (i.e., does not consider FFWTR or FWHOOS).
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-8 Governing Stress Results for RPV Internal Components Current Service    Stress/ Load Item         Component                                       Unit        Design          TPO            Allowable Level(3)   Category Basis(1) 1       Shroud Support           Bounded by the existing design basis. The component is qualified for TPO.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-26 Table 3-8 Governing Stress Results for RPV Internal Components Item Component Service Level (3) Stress/ Load Category Unit Current Design Basis (1) TPO Allowable 1 Shroud Support Bounded by the existing design basis. The component is qualified for TPO.
2             Shroud               Bounded by the existing design basis. The component is qualified for TPO.
2 Shroud   Bounded by the existing design basis. The component is qualified for TPO.
Bend.
B Beam Buckling Bend. Mom., In-lbs 1.21E6 1.21E6 1.51E6 3 Core Plate D Beam Buckling Bend. Mom., In-lbs 1.65E6 1.67E6 3.01E6 B Pm+Pb ksi 17.57 17.63 25.35 4 Top Guide D Pm+Pb ksi 35.26 36.20 50.70 5 CRDH Bounded by the existing design basis. The component is qualified for TPO. B Pm+Pb ksi 14.44 14.82 24.00 C Pm+Pb ksi 20.39 20.92   36.00 6 CRGT  D Pm+Pb ksi 32.39 33.23   38.40 7 Orificed Fuel Support Bounded by the existing design basis. The component is qualified for TPO. 8 FW Sparger Bounded by the existing design basis. The component is qualified for TPO. 9 Jet Pump (2) Bounded by the existing design basis. The component is qualified for TPO.
Beam B                          Mom.,         1.21E6         1.21E6             1.51E6 Buckling In-lbs 3           Core Plate Bend.
10 Core Spray Line & Sparger Bounded by the existing design basis. The component is qualified for TPO. 11 Access Hole Cover Bounded by the existing design basis. The component is qualified for TPO.
Beam D                          Mom.,         1.65E6         1.67E6             3.01E6 Buckling In-lbs B           Pm+Pb           ksi         17.57         17.63             25.35 4           Top Guide D           Pm+Pb           ksi         35.26         36.20             50.70 5             CRDH                 Bounded by the existing design basis. The component is qualified for TPO.
12 Shroud Head And Separator  Bounded by the existing design basis. The component is qualified for TPO.
B           Pm+Pb           ksi         14.44         14.82             24.00 6            CRGT            C           Pm+Pb           ksi         20.39         20.92             36.00 D           Pm+Pb           ksi         32.39         33.23             38.40 7     Orificed Fuel Support       Bounded by the existing design basis. The component is qualified for TPO.
13 In-Core Housing and Guide Tube (ICH&GT) Bounded by the existing design basis. The component is qualified for TPO.
8         FW Sparger             Bounded by the existing design basis. The component is qualified for TPO.
14 Vessel Head Cooling Spray Nozzle Bounded by the existing design basis. The component is qualified for TPO.
(2) 9           Jet Pump               Bounded by the existing design basis. The component is qualified for TPO.
15 Core DP & Liquid Control Line Bounded by the existing design basis. The component is qualified for TPO. 16 LPCI Coupling Bounded by the existing design basis. The component is qualified for TPO.
Core Spray Line &
Notes: (1). 110% OLTP and 110% ICF.  
10                                  Bounded by the existing design basis. The component is qualified for TPO.
(2). The mechanical evaluation of jet pumps is evaluated in a separate report by GEH for TPO.  
Sparger 11     Access Hole Cover           Bounded by the existing design basis. The component is qualified for TPO.
Shroud Head And 12                                  Bounded by the existing design basis. The component is qualified for TPO.
Separator In-Core Housing and 13                                  Bounded by the existing design basis. The component is qualified for TPO.
Guide Tube (ICH&GT)
Vessel Head Cooling 14                                  Bounded by the existing design basis. The component is qualified for TPO.
Spray Nozzle Core DP & Liquid 15                                  Bounded by the existing design basis. The component is qualified for TPO.
Control Line 16       LPCI Coupling             Bounded by the existing design basis. The component is qualified for TPO.
Notes:
(1). 110% OLTP and 110% ICF.
(2). The mechanical evaluation of jet pumps is evaluated in a separate report by GEH for TPO.
(3). Service level A is bounded by service level B.
(3). Service level A is bounded by service level B.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3-27 Table 3-9 Piping Lines Recommended for Special Focus under FAC Review Line Name Comment 30'' GBD-109, -209 10'' GBD-111, -211 30'' GBD-111, -211 10'' GBD-116, -216 20'' GBD-116, -216 30'' GBD-116, -216 24'' GBD-179, -279 30'' GBD-179, -279 24'' GBD-180, -280 30'' GBD-180, -280 20'' GBD-117, -217 20'' GBD-118, -218 18'' GBD-118, -218 20'' DBD-101, -201 18'' GAD-103, -203 10'' HAD-101, -201 16'' HAD-103, -203 24'' HAD-103, -203 20'' GAD-107, -207 These lines are predicted to exceed recommended flow velocity guidelines under TPO conditions.
3-26
Flow velocities that exceed these guidelines are acceptable, provided the FAC program is updated to ensure adequate inspection frequencies. The station FAC program will be updated to include the effects of TPO conditions as necessary.
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4-1 4.0 ENGINEERED SAFE TY FEATURES 4.1 CONTAINMENT S YSTEM PERFORMANCE TLTR Appendix G presents the methods, approach, and scope for the TPO uprate containment evaluation for LOCA. The current containment evaluations were performed at 102% of CLTP. Although the nominal operating conditions change slightly because of the TPO uprate, the required initial conditions for containment analysis inputs remain the same as previously documented. The following table summarizes the effect of the TPO uprate on various aspects of the containment system performance.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-9 Piping Lines Recommended for Special Focus under FAC Review Line Name                                     Comment 30'' GBD-109, -209 These lines are predicted to exceed recommended 10'' GBD-111, -211 flow velocity guidelines under TPO conditions.
Topic Key Parameters TPO Effect Short Term Pressure and Temperature Response Gas Temperature Break Flow and Energy Pressure Break Flow and Energy Long-Term Suppression Pool Temperature Response Bulk Pool Decay Heat Local Temperature with SRV Discharge Decay Heat Containment Dynamic Loads Loss-of-Coolant Accident Loads Break Flow and Energy Safety-Relief Valve Loads Decay Heat Sub compartment Pressurization Break Flow and Energy Current Analysis Based on 102% of CLTP Containment Isolation Section 4.1.1 provides confirmation that MOVs are capable of performing design basis functions at TPO conditions. The ability of containment isolation valves and operators to perform their required functions is not affected because the evaluations have been performed at 102% of CLTP. 4.1.1 Generic Letter 89-10 Program The motor-operated valve (MOV) requirements in the UFSAR were reviewed, and no changes to the functional requirements of the GL 89-10, "Saf ety-Related Motor-Operated Valve Testing and Surveillance," MOVs are identified as a result of operating at the TPO RTP level. Because previous analyses were either based on 102% of C LTP or are consistent with the plant conditions expected to result from TPO, there are no increases in the pressure or temperature at which NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4-2 MOVs are required to operate. Therefore, the GL 89-10 MOVs remain capable of performing their design basis functions.
30'' GBD-111, -211                         Flow velocities that exceed these guidelines are 10'' GBD-116, -216                         acceptable, provided the FAC program is updated to 20'' GBD-116, -216                         ensure adequate inspection frequencies. The station 30'' GBD-116, -216                         FAC program will be updated to include the effects 24'' GBD-179, -279                         of TPO conditions as necessary.
4.1.2 Generic Letter 95-07 Program The evaluation performed in support of GL 95-07, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," has been reviewed and no changes are identified as a result of operating at the TPO RTP level. The criteria for susceptibility to pressure locking or thermal binding were reviewed and it was determined that the slight changes in operating or environmental conditions expected to result fr om the TPO uprate would have no impact on the  
30'' GBD-179, -279 24'' GBD-180, -280 30'' GBD-180, -280 20'' GBD-117, -217 20'' GBD-118, -218 18'' GBD-118, -218 20'' DBD-101, -201 18'' GAD-103, -203 10'' HAD-101, -201 16'' HAD-103, -203 24'' HAD-103, -203 20'' GAD-107, -207 3-27
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4.0 ENGINEERED SAFETY FEATURES 4.1   CONTAINMENT SYSTEM PERFORMANCE TLTR Appendix G presents the methods, approach, and scope for the TPO uprate containment evaluation for LOCA. The current containment evaluations were performed at 102% of CLTP.
Although the nominal operating conditions change slightly because of the TPO uprate, the required initial conditions for containment analysis inputs remain the same as previously documented.
The following table summarizes the effect of the TPO uprate on various aspects of the containment system performance.
Topic                     Key Parameters                   TPO Effect Short Term Pressure and Temperature Response Gas Temperature         Break Flow and Energy Pressure               Break Flow and Energy Long-Term Suppression Pool Temperature Response Bulk Pool               Decay Heat Current Analysis Local Temperature with Decay Heat              Based on 102% of CLTP SRV Discharge Containment Dynamic Loads Loss-of-Coolant         Break Flow and Energy Accident Loads Safety-Relief Valve     Decay Heat Loads Sub compartment         Break Flow and Energy Pressurization Containment Isolation                                     The ability of containment isolation valves Section 4.1.1 provides                                     and operators to perform their required confirmation that MOVs are                                functions is not affected because the capable of performing design                              evaluations have been performed at 102%
basis functions at TPO                                    of CLTP.
conditions.
4.1.1   Generic Letter 89-10 Program The motor-operated valve (MOV) requirements in the UFSAR were reviewed, and no changes to the functional requirements of the GL 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance, MOVs are identified as a result of operating at the TPO RTP level. Because previous analyses were either based on 102% of CLTP or are consistent with the plant conditions expected to result from TPO, there are no increases in the pressure or temperature at which 4-1
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION MOVs are required to operate. Therefore, the GL 89-10 MOVs remain capable of performing their design basis functions.
4.1.2   Generic Letter 95-07 Program The evaluation performed in support of GL 95-07, Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves, has been reviewed and no changes are identified as a result of operating at the TPO RTP level. The criteria for susceptibility to pressure locking or thermal binding were reviewed and it was determined that the slight changes in operating or environmental conditions expected to result from the TPO uprate would have no impact on the functioning of power-operated gate valves within the scope of GL 95-07. Therefore, the valves remain capable of performing their design basis functions.
4.1.3  Generic Letter 96-06 The Limerick response to GL 96-06, Assurance of Equipment Operability and Containment Integrity during Design-Basis Accident Conditions, was reviewed for the TPO uprate. The containment design temperatures and pressures in the current GL 96-06 evaluation are not exceeded under post-accident conditions for the TPO uprate. Therefore, the Limerick response to GL 96-06 remains valid under TPO uprate conditions.
4.1.4  Containment Coatings The qualified coatings in primary containment are qualified such that they do not fail when exposed to the existing maximum post-LOCA primary containment operating conditions of 340&deg;F, 44.0 psig, and 100% relative humidity. These operating conditions bound those, which are expected after implementation of TPO since the current operating conditions are based on 102% of CLTP.
4.2    EMERGENCY CORE COOLING SYSTEMS 4.2.1  High Pressure Coolant Injection The High Pressure Coolant Injection (HPCI) system is a turbine driven system designed to pump water into the reactor vessel over a wide range of operating pressures. For the TPO uprate, there is no change to the nominal reactor operating pressure or the SRV setpoints. The primary purpose of the HPCI system is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the RPV. The generic evaluation of the HPCI system provided in TLTR Section 5.6.7 is applicable to Limerick. The ability of the HPCI system to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the HPCI system are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.
4-2


functioning of power-operated gate valves within the scope of GL 95-07. Therefore, the valves remain capable of performing their design basis functions.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4.2.2  High Pressure Core Spray The High Pressure Core Spray system is not applicable to Limerick.
4.1.3 Generic Letter 96-06  The Limerick response to GL 96-06, "Assurance of Equipment Operability and Containment Integrity during Design-Basis A ccident Conditions," was reviewed for the TPO uprate. The containment design temperatures and pressures in the current GL 96-06 evaluation are not exceeded under post-accident conditions for the TPO uprate. Therefore, the Limerick response to GL 96-06 remains valid under TPO uprate conditions.
4.2.3   Core Spray The Core Spray (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 makeup for a large break LOCA and for any small break LOCA after the RPV has depressurized. It also provides spray cooling for long-term core cooling in the event of a LOCA. The generic evaluation of the CS system provided in TLTR Section 5.6.10 is applicable to Limerick. The ability of the CS system to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the CS system are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.
4.1.4 Containment Coatings  The qualified coatings in primary containment are qualified such that they do not fail when exposed to the existing maximum post-LOCA primary containment operating conditions of 340&deg;F, 44.0 psig, and 100% relative humidity.
4.2.4   Low Pressure Coolant Injection 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 provide reactor vessel coolant makeup during a large break LOCA or small break LOCA after the RPV has depressurized. The generic evaluation of the LPCI mode provided in TLTR Section 5.6.4 is applicable to Limerick. The ability of the RHR system to perform required safety functions required by the LPCI mode is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the RHR system LPCI mode are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.
These operating conditions bound those, which are expected after implementation of TPO sin ce the current operating c onditions are based on 102% of CLTP. 4.2 EMERGENCY C ORE COOLING SYSTEMS 4.2.1 High Pressure Coolant Injection  The High Pressure Coolant Injection (HPCI) system is a turbine driven system designed to pump water into the reactor vessel over a wide range of operating pressures. For the TPO uprate, there is no change to the nominal reactor operating pressure or the SRV setpoints. The primary purpose of the HPCI system is to maintain reac tor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the RPV. The generic evaluation of the HPCI system provided in TLTR Section 5.6.7 is applicable to Limerick. The ability of the  
4.2.5  Automatic Depressurization System The Automatic Depressurization System (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 CS and LPCI to inject coolant into the RPV. The ADS initiation logic and valve control is not affected by the TPO uprate. The generic evaluation of the ADS provided in TLTR Section 5.6.8 is applicable to Limerick. The ability of the ADS to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the ADS are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.
4.2.6  ECCS Net Positive Suction Head The most limiting case for NPSH typically occurs at the peak long-term suppression pool temperature. The generic evaluation of the containment provided in TLTR Appendix G is applicable to Limerick. The CLTP containment analyses were based on 102% of CLTP and there is no change in the available NPSH for systems using suppression pool water. Therefore, the TPO uprate does not affect compliance with the ECCS pump NPSH requirements.
4-3


HPCI system to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the HPCI system are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4.3   EMERGENCY CORE COOLING SYSTEM PERFORMANCE The ECCS is designed to provide protection against a postulated LOCA caused by ruptures in the primary system piping. The current 10 CFR 50.46, or LOCA, analyses for the Limerick plant have been performed at 102% of CLTP, consistent with Appendix K. Table 4-1 shows the results of the Limerick ECCS-LOCA analysis. The ECCS-LOCA results for Limerick are in conformance with the error reporting requirements of 10 CFR 50.46. Therefore, the CLTP LOCA analysis for GE14 fuel bounds the TPO uprate for Limerick.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4-3 4.2.2 High Pressure Core Spray The High Pressure Core Spray system is not applicable to Limerick. 4.2.3 Core Spray  The Core Spray (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 makeup for a large break LOCA and for any small break LOCA after the RPV has depressurized. It also provides spray cooling for long-term core cooling in the event of a LOCA. The generi c evaluation of the CS system provided in TLTR Section 5.6.10 is applicable to Limerick. The ability of the CS system to perform required safety functions is demonstrated with prev ious analyses based on 102% of CLTP. Therefore, all safety aspects of the CS system are with in previous evaluations and the requirements are unchanged fo r the TPO uprate conditions.
Reference 16 provides justification for the elimination of the 1600&deg;F Upper Bound PCT (UBPCT) limit and generic justification that the Licensing Basis PCT (LBPCT) will be conservative with respect to the UBPCT. The NRC SER for Reference 16 accepted this position, noting that because plant-specific UBPCT calculations have been performed for all plants, other means may be used to demonstrate compliance with the original SER requirements. These other means are acceptable provided there are no significant changes to a plants configuration that would invalidate the existing UBPCT calculations. Reference 17 provided justification for the elimination of the UBPCT limit for Limerick Units 1 and 2.
4.2.4 Low Pressure Coolant Injection  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 provide reactor vessel coolant makeup during a large break LOCA or small break LOCA after the RPV has depressurized. The generic evaluation of the LPCI mode provided in TLTR Section 5.6.4 is applicable to Limerick. The ability of the RHR system to perform required safety functions required by the LPCI mode is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the RHR system LPCI mode are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.
For the TPO uprate there are no changes to the plant configuration that would invalidate the Reference 17 evaluation for conformance with Reference 16.
4.2.5 Automatic Depressurization System  The Automatic Depressurization System (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 CS and LPCI to inject coolant into the RPV. The ADS initiation logic and valve control is not affected by the TPO uprate. The generic evaluation of the ADS provided in TLTR Section 5.6.8 is applicable to Limerick. The ability of the ADS to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP.
The CLTP LOCA analysis for GE14 fuel is concluded to bound the TPO uprate for Limerick.
Therefore, all safety aspects of the ADS are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.
4.4  MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM The Main Control Room atmosphere is not affected by the TPO uprate. Control Room habitability following a postulated accident at TPO conditions is unchanged because the Main Control Room Atmosphere Control System has previously been evaluated for radiation release accident conditions at 102% of CLTP. Therefore, the system remains capable of performing its safety function at the TPO conditions.
4.2.6 ECCS Net Positive Suction Head  The most limiting case for NPSH typically occurs at the peak long-term suppression pool temperature. The generic evaluation of the containment provided in TLTR Appendix G is applicable to Limerick. The CLTP containment analyses were based on 102% of CLTP and there is no change in the available NPSH for systems using suppression pool water. Therefore, the TPO uprate does not affect compliance with the ECCS pump NPSH requirements.
4.5  STANDBY GAS TREATMENT SYSTEM The Standby Gas Treatment System (SGTS) minimizes the offsite and control room dose rates during venting and purging of the containment atmosphere under abnormal conditions. The current capacity of the SGTS was selected to maintain the secondary containment at a slightly negative pressure during such conditions. This capability is not changed by the TPO uprate conditions. The SGTS can accommodate design basis accident (DBA) conditions at 102% of CLTP. Therefore, the system remains capable of performing its safety function for the TPO uprate condition.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4-4 4.3 EMERGENCY C ORE COOLING SYSTEM PERFORMANCE The ECCS is designed to provide protection ag ainst a postulated LOCA caused by ruptures in the primary system piping. The current 10 CFR 50.46, or LOCA, analyses for the Limerick plant have been performed at 102% of CLTP, consis tent with Appendix K. Table 4-1 shows the results of the Limerick ECCS-LOCA analysis. The ECCS-LOCA results for Limerick are in conformance with the error reporting requireme nts of 10 CFR 50.46. Therefore, the CLTP LOCA analysis for GE14 fuel bounds the TPO uprate for Limerick.
4.6  MAIN STEAM ISOLATION VALVE LEAKAGE ALTERNATE DRAIN PATHWAY The MSIV Leakage Alternate Drain Pathway prevents a direct release of fission products that could leak through the closed MSIVs after a LOCA. The pathway provides control by providing 4-4
Reference 16 provides justification for the elimination of the 1600&deg;F Upper Bound PCT (UBPCT) limit and generic justification that the Licensing Basis PCT (LBPCT) will be conservative with respect to the UBPCT. Th e NRC SER for Reference 16 accepted this position, noting that because plant-specific UBPCT calculations have been performed for all plants, other means may be used to demonstrate compliance with the original SER requirements. These other means are acceptable provided there are no significant changes to a plant's configuration that would invalidate the existing UBPCT calculations. Reference 17 provided justification for the elimination of the UBPCT limit for Limerick Units 1 and 2. For the TPO uprate there are no changes to the plant configuration that would invalidate the Reference 17 evaluation for conformance with Reference 16.


The CLTP LOCA analysis for GE14 fuel is concluded to bound the TPO uprate for Limerick. 4.4 M AIN CONTROL ROOM ATMOSPHERE CONTROL S YSTEM  The Main Control Room atmosphere is not affected by the TPO uprate. Control Room habitability following a postulated accident at TPO conditions is unchanged because the Main Control Room Atmosphere Control System has pr eviously been evaluated for radiation release accident conditions at 102% of CLTP. Therefore, the system remains capable of performing its safety function at the TPO conditions. 4.5 STANDBY G AS TREATMENT S YSTEM  The Standby Gas Treatment System (SGTS) minimizes the offsite and control room dose rates during venting and purging of the containment atmosphere under abnormal conditions. The current capacity of the SGTS was selected to ma intain the secondary containment at a slightly negative pressure during such conditions. This capability is not changed by the TPO uprate conditions. The SGTS can accommodate design ba sis accident (DBA) conditions at 102% of CLTP. Therefore, the system remains capable of performing its safety function for the TPO uprate condition. 4.6 M AIN STEAM ISOLATION V ALVE LEAKAGE ALTERNATE DRAIN PATHWAY The MSIV Leakage Alternate Drain Pathway preven ts a direct release of fission products that could leak through the closed MSIVs after a LOCA. The pathway provides control by providing NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4-5 a hold-up volume for the MSIV leakage before release to the atmosphere. This is accomplished by directing the leakage through existing Main Steam Drain Lines to the High Pressure Shell of the Main Condenser. This system was previously evaluated at 102%
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION a hold-up volume for the MSIV leakage before release to the atmosphere. This is accomplished by directing the leakage through existing Main Steam Drain Lines to the High Pressure Shell of the Main Condenser.
and is therefore acceptable for TPO operations. 4.7 P OST-LOCA COMBUSTIBLE G AS CONTROL S YSTEM  The Combustible Gas Control System (CGCS) maintains the post-LOCA concentration of oxygen or hydrogen in the containment atmosphere below the flammability limit. The generic evaluation of the CGCS provided in TLTR Section J.2.3.10 is applicable to Limerick Units 1 and 2. The metal available for reaction is uncha nged by the TPO uprate and the hydrogen production due to radiolytic decomposition is unchanged because the system was previously evaluated for accident conditions from 102% of CLTP. Therefore, the current evaluation is valid for the TPO uprate.
This system was previously evaluated at 102% and is therefore acceptable for TPO operations.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4-6 Table 4-1 Limerick ECCS-LOCA Analys is Results for GE14 Fuel Parameter MELLLA Analysis Limit Nominal PCT 1007 F N/A Appendix K PCT 1666F < 2200 F* LBPCT 1675F < 2200 F* Maximum Local Oxidation
4.7  POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM The Combustible Gas Control System (CGCS) maintains the post-LOCA concentration of oxygen or hydrogen in the containment atmosphere below the flammability limit. The generic evaluation of the CGCS provided in TLTR Section J.2.3.10 is applicable to Limerick Units 1 and
<1.0%  17%* Core-Wide Metal-Water Reaction
: 2. The metal available for reaction is unchanged by the TPO uprate and the hydrogen production due to radiolytic decomposition is unchanged because the system was previously evaluated for accident conditions from 102% of CLTP. Therefore, the current evaluation is valid for the TPO uprate.
<0.1%  1.0%*
4-5
* 10 CFR 50.46 ECCS-LOCA Analysis Acceptance Criteria NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-1 5.0  INSTRUMENTATION AND CONTROL 5.1 NSSS MONITORING AND CONTROL The instruments and controls that directly interact with or control the reactor are usually considered within the NSSS. The NSSS process variables and instrument se tpoints that could be affected by the TPO uprate were evaluated.
 
5.1.1 Neutron Monitoring System 5.1.1.1 Average Power Range Monito rs, Intermediate Range Monitors, and Source Range Monitors The Average Power Range Monitors (APRMs) ar e re-calibrated to i ndicate 100% at the TPO RTP level of 3515 MWt. The APRM high flux scram and the upper limit of the rod block setpoints, expressed in units of percent of licensed power, are not changed. The flow-biased APRM trips, expressed in units of absolute thermal power (i.e., MWt), remain the same.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 4-1 Limerick ECCS-LOCA Analysis Results for GE14 Fuel Parameter            MELLLA                    Analysis Limit Nominal PCT                1007&deg;F                        N/A Appendix K PCT              1666&deg;F                    < 2200&deg;F*
However, in order to accommodate limits in the Stability Region, new flow-biased APRM Analytical Limits (ALs) were established that conservatively boun d the entire operating envelope. This approach for the Limerick TPO uprate follows th e guidelines of TLTR Section 5.6.1 and Appendix F, which is consis tent with the practi ce approved for GE BWR uprates in ELTR1 (Reference 2). For the TPO uprate, no adjustment is needed to ensure the Intermediate Range Monitors (IRMs) have adequate overlap with the SRMs and APRMs. However, normal plant surveillance procedures may be used to adjust the IRMs, the overlap with the SRMs and the APRMs. The IRM channels have sufficient margin to the upscale scram trip on the highest range when the APRM channels are reading near their downscale alarm trip because the change in APRM scaling is so small for the TPO uprate.
LBPCT                      1675&deg;F                    < 2200&deg;F*
5.1.1.2 Local Power Range Monitors an d Traversing In Core Probes At the TPO RTP level, the flux at some LPRMs increases. However, the small change in the power level is not a significant factor to the ne utronic service life of the LPRM detectors and radiation level of the traversing in core probes (TIPs). It does not change the number of cycles in the lifetime of any of the detector
Maximum Local
: s. The LPRM accuracy at the increased flux is within specified limits, and the LPRMs are designed as replaceable components. The TIPs are stored in shielded rooms. The radiation protection program for normal plant operation can accommodate a small increase in radiation levels.
                            <1.0%                        17%*
5.1.1.3 Rod Block Monitor  The Rod Block Monitor (RBM) instrumentation is referenced to an APRM channel. Because the APRM has been rescaled, there is only a small effect on the RBM performance due to the LPRM NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-2 performance at the higher average local flux. The RBM instrumentation is not significantly affected by the TPO uprate conditio ns, and no change is needed.
Oxidation Core-Wide Metal-
5.1.2 Rod Worth Minimizer  The Rod Worth Minimizer (RWM) does not perform a safety-related func tion. The function of the RWM is to support the operator by enforcing rod patterns until reactor power has reached appropriate levels. The power-dependent setpoints for the RWM are discussed in Section 5.3.8.
                            <0.1%                      1.0%*
5.2 BOP M ONITORING AND CONTROL  Operation of the plant at the TPO RTP level has minimal effect on the BOP system instrumentation and control devices. The improved FW flow measurement, which is the basis for the reduction in power uncerta inty, is addressed in Section 1.4. All instrumentation with control functions has sufficient range/adjustment capability for use at the TPO uprate conditions. No safety-related BOP system setpoint changes are required as a result of the TPO uprate. The plant-specific instrumentation and control de sign and operating conditi ons are bounded by those used in the evaluations contained in the TLTR.
Water Reaction
5.2.1 Pressure Control System  The Pressure Control System (PCS) provides a fast and stable response to steam flow changes so that reactor pressure is controlled within allowa ble values. The PCS consists of the pressure regulation system, turbine control valve system and steam bypass valve system. The main turbine speed/load control function is performed by the main tu rbine-generator Electro-hydraulic Control (EHC) system. The steam bypass valve pressure control function is performed by the Turbine Bypass Control System (TBCS). Satisfactory reactor pressure control by the pressure regulator and the turbine control valves (TCVs) requires an adequate flow margin between the TPO RTP oper ating condition and the steam flow capability of the TCVs at their maximum stroke (i.e., valves wide open (VWO)). Limerick will modify or replace the first stage nozzle plate on the main turbine in order to maintain adequate flow margin at TPO conditions. The existing electronic controls, as designed for the current 100% of CLTP cond itions, were evaluated. The results of the analysis indicated that the performance of the TCVs is not imp acted by increasing power level to TPO conditions; therefore, no modifications to th e electronic components are required. No modification is required to the steam bypass valves. No modifica tions are required to controls or alarm annunciators provided in the main control room. The appr opriate control room indicators will be adju sted, as necessary, to reflect 100% TPO power. The required adjustments are limited to "tuning" of the control settings that may be required to operate optimally at the TPO uprate power level.
* 10 CFR 50.46 ECCS-LOCA Analysis Acceptance Criteria 4-6
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-3 5.2.2 EHC Turbine Control System The turbine EHC system was reviewed for the increase in core thermal power and associated ~2% increase in rated steam flow. The control system is expected to perform normally for TPO RTP operation. Normal operator cont rols are used in conjunction with the associated operating procedures. Confirmation testing will be performed during power ascension (Section 10.4).
 
5.2.3 Feedwater Control System  An evaluation of the ability of the FW level control system, FW control valves, and/or FW turbine controls to maintain adequate water leve l control at the TPO uprat e conditions has been performed. The ~2% increase in FW flow associated with TPO uprate is within the current control margin of these systems. No changes in the operating reactor water level or reactor water level trip set points are required for the TPO uprate. Per the guidelines of TLTR Appendix L, the performance of the FW level control systems will be recorded at 95% and 100% of CLTP and confirmed at the TPO power during power ascension. These checks will demonstrate acceptable operational capability and will utilize the methods and criteria described in the original startup testing of these systems.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.0 INSTRUMENTATION AND CONTROL 5.1   NSSS MONITORING AND CONTROL The instruments and controls that directly interact with or control the reactor are usually considered within the NSSS. The NSSS process variables and instrument setpoints that could be affected by the TPO uprate were evaluated.
5.2.4 Leak Detection System  The setpoints associated with leak detection have been evaluated with respect to the ~2% higher steam flow and ~2&deg;F increase in FW temperature for the TPO uprate. Each of the systems, where leak detection potentially could be affected, is addressed below.
5.1.1   Neutron Monitoring System 5.1.1.1 Average Power Range Monitors, Intermediate Range Monitors, and Source Range Monitors The Average Power Range Monitors (APRMs) are re-calibrated to indicate 100% at the TPO RTP level of 3515 MWt. The APRM high flux scram and the upper limit of the rod block setpoints, expressed in units of percent of licensed power, are not changed. The flow-biased APRM trips, expressed in units of absolute thermal power (i.e., MWt), remain the same.
Main Steam Tunnel Temperature Based Leak Detection The ~2&deg;F increase in FW temperature for the TPO uprate decreases the leak detection trip avoidance margin. As descri bed in TLTR Section F.4.2.8, the high steam tunnel temperature setpoint remains unchanged.
However, in order to accommodate limits in the Stability Region, new flow-biased APRM Analytical Limits (ALs) were established that conservatively bound the entire operating envelope. This approach for the Limerick TPO uprate follows the guidelines of TLTR Section 5.6.1 and Appendix F, which is consistent with the practice approved for GE BWR uprates in ELTR1 (Reference 2).
For the TPO uprate, no adjustment is needed to ensure the Intermediate Range Monitors (IRMs) have adequate overlap with the SRMs and APRMs. However, normal plant surveillance procedures may be used to adjust the IRMs, the overlap with the SRMs and the APRMs. The IRM channels have sufficient margin to the upscale scram trip on the highest range when the APRM channels are reading near their downscale alarm trip because the change in APRM scaling is so small for the TPO uprate.
5.1.1.2 Local Power Range Monitors and Traversing In Core Probes At the TPO RTP level, the flux at some LPRMs increases. However, the small change in the power level is not a significant factor to the neutronic service life of the LPRM detectors and radiation level of the traversing in core probes (TIPs). It does not change the number of cycles in the lifetime of any of the detectors. The LPRM accuracy at the increased flux is within specified limits, and the LPRMs are designed as replaceable components. The TIPs are stored in shielded rooms. The radiation protection program for normal plant operation can accommodate a small increase in radiation levels.
5.1.1.3 Rod Block Monitor The Rod Block Monitor (RBM) instrumentation is referenced to an APRM channel. Because the APRM has been rescaled, there is only a small effect on the RBM performance due to the LPRM 5-1
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION performance at the higher average local flux. The RBM instrumentation is not significantly affected by the TPO uprate conditions, and no change is needed.
5.1.2  Rod Worth Minimizer The Rod Worth Minimizer (RWM) 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 power-dependent setpoints for the RWM are discussed in Section 5.3.8.
5.2    BOP MONITORING AND CONTROL Operation of the plant at the TPO RTP level has minimal effect on the BOP system instrumentation and control devices. The improved FW flow measurement, which is the basis for the reduction in power uncertainty, is addressed in Section 1.4. All instrumentation with control functions has sufficient range/adjustment capability for use at the TPO uprate conditions.
No safety-related BOP system setpoint changes are required as a result of the TPO uprate. The plant-specific instrumentation and control design and operating conditions are bounded by those used in the evaluations contained in the TLTR.
5.2.1  Pressure Control System The Pressure Control System (PCS) provides a fast and stable response to steam flow changes so that reactor pressure is controlled within allowable values. The PCS consists of the pressure regulation system, turbine control valve system and steam bypass valve system. The main turbine speed/load control function is performed by the main turbine-generator Electro-hydraulic Control (EHC) system. The steam bypass valve pressure control function is performed by the Turbine Bypass Control System (TBCS).
Satisfactory reactor pressure control by the pressure regulator and the turbine control valves (TCVs) requires an adequate flow margin between the TPO RTP operating condition and the steam flow capability of the TCVs at their maximum stroke (i.e., valves wide open (VWO)).
Limerick will modify or replace the first stage nozzle plate on the main turbine in order to maintain adequate flow margin at TPO conditions. The existing electronic controls, as designed for the current 100% of CLTP conditions, were evaluated. The results of the analysis indicated that the performance of the TCVs is not impacted by increasing power level to TPO conditions; therefore, no modifications to the electronic components are required.
No modification is required to the steam bypass valves. No modifications are required to controls or alarm annunciators provided in the main control room. The appropriate control room indicators will be adjusted, as necessary, to reflect 100% TPO power. The required adjustments are limited to tuning of the control settings that may be required to operate optimally at the TPO uprate power level.
5-2
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.2.2    EHC Turbine Control System The turbine EHC system was reviewed for the increase in core thermal power and associated
~2% increase in rated steam flow. The control system is expected to perform normally for TPO RTP operation. Normal operator controls are used in conjunction with the associated operating procedures. Confirmation testing will be performed during power ascension (Section 10.4).
5.2.3    Feedwater Control System An evaluation of the ability of the FW level control system, FW control valves, and/or FW turbine controls to maintain adequate water level control at the TPO uprate conditions has been performed. The ~2% increase in FW flow associated with TPO uprate is within the current control margin of these systems. No changes in the operating reactor water level or reactor water level trip set points are required for the TPO uprate. Per the guidelines of TLTR Appendix L, the performance of the FW level control systems will be recorded at 95% and 100% of CLTP and confirmed at the TPO power during power ascension. These checks will demonstrate acceptable operational capability and will utilize the methods and criteria described in the original startup testing of these systems.
5.2.4    Leak Detection System The setpoints associated with leak detection have been evaluated with respect to the ~2% higher steam flow and ~2&deg;F increase in FW temperature for the TPO uprate. Each of the systems, where leak detection potentially could be affected, is addressed below.
Main Steam Tunnel Temperature Based Leak Detection The ~2&deg;F increase in FW temperature for the TPO uprate decreases the leak detection trip avoidance margin. As described in TLTR Section F.4.2.8, the high steam tunnel temperature setpoint remains unchanged.
RWCU System Temperature Based Leak Detection There is no significant effect on RWCU system temperature or pressure due to the TPO uprate.
RWCU System Temperature Based Leak Detection There is no significant effect on RWCU system temperature or pressure due to the TPO uprate.
Therefore, there is no effect on the RWCU temperature based leak detection. RCIC System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the RCIC system temperat ure or pressure, and t hus, the RCIC temperature based leak detection system is not affected.
Therefore, there is no effect on the RWCU temperature based leak detection.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-4 HPCI System Temperature Based Leak Detection  The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the HPCI system temperat ure or pressure, and thus, the HPCI temperature based leak detection system is not affected. RHR System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the RHR system temperature or pressure, and thus, the RHR temperature based leak detection system is not affected.
RCIC System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the RCIC system temperature or pressure, and thus, the RCIC temperature based leak detection system is not affected.
Non-Temperature Based Leak Detection The non-temperature based leak detection systems are not affected by the TPO uprate. 5.3 TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS  The determination of instrument setpoints is based on plant operating experience, conservative licensing analyses or limiting design/operating values. Standard GE setpoint methodologies (References 18 and 19) are used to generate the allowable values (AV) and nominal trip setpoints (NTSP) related to any AL change, as applicable. Each actual trip setting is established to preclude inadvertent initiation of the protective action, while assuring adequate allowances for instrument accuracy, calibration, drift and applicable normal and accident design basis events. Table 5-1 lists the ALs that change based on results from the TPO evaluations and safety analyses. In general, if th e AL does not change in the units shown in the Technical Specifications, then no change in its associated plant AV and NTSP is required, as shown in the Technical Specifications. Changes in the setpoint margins due to changes in instrument accuracy and calibration errors caused by the change in environmental conditions around the instrument due to the TPO uprate are negligible. Maintaining constant nominal dome pressure for the TPO uprate minimizes the potential effect on these instruments by maintaining the same fluid properties at the instruments. The setpoint evaluations are based on the guidelines in TLTR Sections 5.8 and F.4 and on Section 5.3 of Reference 18.
5-3
5.3.1 High-Pressure Scram  The high-pressure scram terminates a pressure increase transient not terminated by direct or high flux scram. Because there is no increase in nominal reactor operating pressure with the TPO uprate, the scram AL on reacto r high pressure is unchanged.
5.3.2 Hydraulic Pressure Scram The AL for the turbine hydraulic pressure (low oil pressure trip) that initiates the Turbine-Generator (T/G) trip scram at high power remains the same as for the CLTP. No modifications NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-5 are being made to the turbine hydraulic control systems for TP O; actuation of these safety functions remain unchanged for TPO.
5.3.3 High-Pressure Recirculation Pump Trip  The ATWS-RPT trips the pumps during plant transients with increases in reactor vessel dome pressure. The ATWS-RPT provides negative reac tivity by reducing core flow during the initial part of an ATWS. The evaluation in Section 9.3.1 demonstrates that th e current high pressure ATWS-RPT AL is acceptable for the TPO uprate.
5.3.4 Safety Relief Valve Because there is no increase in reactor operating dome pressure , the SRV ALs are not changed.
5.3.5 Main Steam Line High Flow Isolation The Technical Specification AV of this function is expressed in terms of psid. The corresponding percent thermal power is not changed. The existing AV setpoint in terms of psid has sufficient trip avoidance margin to support the small increase in the TPO rated steam flow. Therefore, the AV at the TPO remains unchanged from CLTP. Because of the large spurious trip margin, sufficient margin to the trip setpoint exists to allow for normal plant testing of the MSIVs and turbine stop and control valves. This is consistent with TLTR Section F.4.2.5.
5.3.6 Fixed APRM Scram The fixed APRM ALs, for both two (recirculation) loop operation (TLO) and SLO, expressed in percent of RTP do not change for the TPO upr ate. The generic ev aluation and guidelines presented in TLTR Section F.4.2.2 are applicable to Limerick. The limiting transient that relies on the fixed APRM trip is the vessel overpressure transient (MSIV closure) with indirect scram. 


This event has been analyzed assuming 102% of CLTP and is reanalyzed on a cycle specific basis. 5.3.7 APRM Flow-Biased Scram The flow-referenced APRM ALs, for both TLO a nd SLO, are unchanged in units of absolute core thermal power versus recirculation drive flow. Because the setpoints are expressed in percent of RTP, they decrease in proportion to the power uprate or CLTP RTP / TPO RTP. This is the same approach taken for generic BWR up rates described in ELTR1 (Reference 2). There is no significant effect on the instrument errors or uncertainties from the TPO uprate. Therefore, the AV and NTSP are established by directly incorporating the change in the AL.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION HPCI System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the HPCI system temperature or pressure, and thus, the HPCI temperature based leak detection system is not affected.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-6 5.3.8 Rod Worth Minimizer Low Power Setpoint The Rod Worth Minimizer (RWM) Low Power Setp oint (LPSP) is used to enforce the rod patterns established for the control rod drop accident at low power levels. The generic guidelines in TLTR Section F.4.2.9 are applicable to Limerick. The RWM LPSP AL is kept the same in terms of percent power, and is therefore higher in terms of abso lute power. This new higher absolute power is conservative for the RWM LPSP.
RHR System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the RHR system temperature or pressure, and thus, the RHR temperature based leak detection system is not affected.
5.3.9 Rod Block Monitor The severity of the Rod Withdrawal Error (RWE) during powe r operation event is dependent upon the RBM rod block setpoint. The power-dependent ALs are maintained at the same percent power. The cycle specific reload analysis is used to determine any changes in the rod block setpoint.
Non-Temperature Based Leak Detection The non-temperature based leak detection systems are not affected by the TPO uprate.
5.3.10 Flow-Biased Rod Block Monitor (%RTP) Limerick does not have a flow-biased RBM system.
5.3  TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS The determination of instrument setpoints is based on plant operating experience, conservative licensing analyses or limiting design/operating values. Standard GE setpoint methodologies (References 18 and 19) are used to generate the allowable values (AV) and nominal trip setpoints (NTSP) related to any AL change, as applicable. Each actual trip setting is established to preclude inadvertent initiation of the protective action, while assuring adequate allowances for instrument accuracy, calibration, drift and applicable normal and accident design basis events.
Table 5-1 lists the ALs that change based on results from the TPO evaluations and safety analyses. In general, if the AL does not change in the units shown in the Technical Specifications, then no change in its associated plant AV and NTSP is required, as shown in the Technical Specifications. Changes in the setpoint margins due to changes in instrument accuracy and calibration errors caused by the change in environmental conditions around the instrument due to the TPO uprate are negligible. Maintaining constant nominal dome pressure for the TPO uprate minimizes the potential effect on these instruments by maintaining the same fluid properties at the instruments. The setpoint evaluations are based on the guidelines in TLTR Sections 5.8 and F.4 and on Section 5.3 of Reference 18.
5.3.1  High-Pressure Scram The high-pressure scram terminates a pressure increase transient not terminated by direct or high flux scram. Because there is no increase in nominal reactor operating pressure with the TPO uprate, the scram AL on reactor high pressure is unchanged.
5.3.2  Hydraulic Pressure Scram The AL for the turbine hydraulic pressure (low oil pressure trip) that initiates the Turbine-Generator (T/G) trip scram at high power remains the same as for the CLTP. No modifications 5-4
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION are being made to the turbine hydraulic control systems for TPO; actuation of these safety functions remain unchanged for TPO.
5.3.3  High-Pressure Recirculation Pump Trip The ATWS-RPT trips the pumps during plant transients with increases in reactor vessel dome pressure. The ATWS-RPT provides negative reactivity by reducing core flow during the initial part of an ATWS. The evaluation in Section 9.3.1 demonstrates that the current high pressure ATWS-RPT AL is acceptable for the TPO uprate.
5.3.4  Safety Relief Valve Because there is no increase in reactor operating dome pressure, the SRV ALs are not changed.
5.3.5  Main Steam Line High Flow Isolation The Technical Specification AV of this function is expressed in terms of psid. The corresponding percent thermal power is not changed. The existing AV setpoint in terms of psid has sufficient trip avoidance margin to support the small increase in the TPO rated steam flow.
Therefore, the AV at the TPO remains unchanged from CLTP.
Because of the large spurious trip margin, sufficient margin to the trip setpoint exists to allow for normal plant testing of the MSIVs and turbine stop and control valves. This is consistent with TLTR Section F.4.2.5.
5.3.6  Fixed APRM Scram The fixed APRM ALs, for both two (recirculation) loop operation (TLO) and SLO, expressed in percent of RTP do not change for the TPO uprate. The generic evaluation and guidelines presented in TLTR Section F.4.2.2 are applicable to Limerick. The limiting transient that relies on the fixed APRM trip is the vessel overpressure transient (MSIV closure) with indirect scram.
This event has been analyzed assuming 102% of CLTP and is reanalyzed on a cycle specific basis.
5.3.7   APRM Flow-Biased Scram The flow-referenced APRM ALs, for both TLO and SLO, are unchanged in units of absolute core thermal power versus recirculation drive flow. Because the setpoints are expressed in percent of RTP, they decrease in proportion to the power uprate or CLTP RTP / TPO RTP. This is the same approach taken for generic BWR uprates described in ELTR1 (Reference 2). There is no significant effect on the instrument errors or uncertainties from the TPO uprate. Therefore, the AV and NTSP are established by directly incorporating the change in the AL.
5-5
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.3.8   Rod Worth Minimizer Low Power Setpoint The Rod Worth Minimizer (RWM) Low Power Setpoint (LPSP) is used to enforce the rod patterns established for the control rod drop accident at low power levels. The generic guidelines in TLTR Section F.4.2.9 are applicable to Limerick. The RWM LPSP AL is kept the same in terms of percent power, and is therefore higher in terms of absolute power. This new higher absolute power is conservative for the RWM LPSP.
5.3.9   Rod Block Monitor The severity of the Rod Withdrawal Error (RWE) during power operation event is dependent upon the RBM rod block setpoint. The power-dependent ALs are maintained at the same percent power. The cycle specific reload analysis is used to determine any changes in the rod block setpoint.
5.3.10 Flow-Biased Rod Block Monitor (%RTP)
Limerick does not have a flow-biased RBM system.
5.3.11 Main Steam Line High Radiation Isolation The MSL high radiation signal isolation function has been eliminated at Limerick.
5.3.11 Main Steam Line High Radiation Isolation The MSL high radiation signal isolation function has been eliminated at Limerick.
5.3.12 Low Steam Line Pressure MSIV Closure (RUN Mode)
5.3.12 Low Steam Line Pressure MSIV Closure (RUN Mode)
The purpose of this function is to initiate MSIV closure on low steam line pressure when the reactor is in the RUN mode. This AL is not changed for the TPO as discussed in TLTR Section F.4.2.7.
The purpose of this function is to initiate MSIV closure on low steam line pressure when the reactor is in the RUN mode. This AL is not changed for the TPO as discussed in TLTR Section F.4.2.7.
5.3.13 Reactor Water Level Instruments As described in TLTR Section F.4.2.10, the TPO uprate does not result in a significant increase in the possibility of a reactor scram, equipment trip, or ECCS actuation. Use of the current ALs maintains acceptable safety system performan ce. The low reactor water level Technical Specification setpoints for scram, high-pressure injection, and ADS/ECCS are not changed for the TPO uprate. The high water level ALs for trip of the main turbine, FW pumps, and reactor scram are not changed for the TPO uprate.
5.3.13 Reactor Water Level Instruments As described in TLTR Section F.4.2.10, the TPO uprate does not result in a significant increase in the possibility of a reactor scram, equipment trip, or ECCS actuation. Use of the current ALs maintains acceptable safety system performance. The low reactor water level Technical Specification setpoints for scram, high-pressure injection, and ADS/ECCS are not changed for the TPO uprate. The high water level ALs for trip of the main turbine, FW pumps, and reactor scram are not changed for the TPO uprate.
Water level change during operational transients (e.g., trip of a recirculation pump, FW controller failure, loss of one FW pump) is slightly affected by the TPO uprate. The plant response following the trip of one FW pump does not change significantly, because the maximum operating rod line is not being increase
Water level change during operational transients (e.g., trip of a recirculation pump, FW controller failure, loss of one FW pump) is slightly affected by the TPO uprate. The plant response following the trip of one FW pump does not change significantly, because the maximum operating rod line is not being increased. Therefore, the final power level following a single FW pump trip at TPO uprate conditions would not change relative to the remaining FW flow as exists at CLTP.
: d. Therefore, the final power level following a single FW pump trip at TPO uprat e conditions would not change relative to the remaining FW flow as exists at CLTP.
5-6
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-7 5.3.14 Main Steam Line Tunnel Hig h Temperature Isolations As noted in Section 5.2.4 above, the high steam tunnel temperature AL remains unchanged for the TPO uprate.
5.3.15 Low Condenser Vacuum  In order to produce more electrica l power, the amount of heat discharged to the main condenser increases slightly. This added heat load may s lightly increase condenser backpressure, but the increase would be insignificant
(< 0.15 in. HgA). The slight change in condenser vacuum after implementation of TPO will not adversely affect any trip setpoints associated with low condenser vacuum (turbine trip / MSIV closure).
5.3.16 TSV Closure Scram, TCV Fast Closure Scram Bypasses The turbine first-stage pressure bypass allows the Turbine Stop Valve (TSV) closure scram and Turbine Control Valve (TCV) fast closure scram to be bypassed, when reactor power is sufficiently low, such that the scram functions are not needed to mitigate a T/G trip. This power level is the AL for determining the actual trip setpoint, which comes from the turbine first-stage pressure (TFSP). The TFSP setpoint is chosen to allow operational margin so that scrams can be avoided, by transferring steam to the turbine bypass system dur ing T/G trips at low power.
Based on the guidelines in TLTR Section F.4.2.3, the TSV closure scram, TCV fast closure scram, and End of Cycle (EOC)-RPT bypass AL in percent of RTP is reduced by the ratio of the power increase. The new AL does not change with respect to absolute thermal power.  [[             


                    ]]  The maneuvering range for plant startup is maximized. The High Pressure turbine first stage nozzle plates for Limerick, Units 1 and 2, will be modified for TPO operation as described in Section 7.1. Th e first stage pressure se tpoint for the bypass of the TSV closure scram, TCV fast closure scram, and EOC-RPT will be adjusted as necessary to implement the new AL.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.3.14 Main Steam Line Tunnel High Temperature Isolations As noted in Section 5.2.4 above, the high steam tunnel temperature AL remains unchanged for the TPO uprate.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5-8 Table 5-1 Analytical Limits that Change due to TPO Parameter Current TPO Justification APRM High Neutron Flux Scram 121 No change APRM Flow-biased Scram Fixed (%RTP) 119 No change TLO Flow-biased (%RTP)
5.3.15 Low Condenser Vacuum In order to produce more electrical power, the amount of heat discharged to the main condenser increases slightly. This added heat load may slightly increase condenser backpressure, but the increase would be insignificant (< 0.15 in. HgA). The slight change in condenser vacuum after implementation of TPO will not adversely affect any trip setpoints associated with low condenser vacuum (turbine trip / MSIV closure).
(1)(2) 0.66W + 66.2 0.65W + 65.1 (3)  SLO Flow-biased (%RTP)
5.3.16 TSV Closure Scram, TCV Fast Closure Scram Bypasses The turbine first-stage pressure bypass allows the Turbine Stop Valve (TSV) closure scram and Turbine Control Valve (TCV) fast closure scram to be bypassed, when reactor power is sufficiently low, such that the scram functions are not needed to mitigate a T/G trip. This power level is the AL for determining the actual trip setpoint, which comes from the turbine first-stage pressure (TFSP). The TFSP setpoint is chosen to allow operational margin so that scrams can be avoided, by transferring steam to the turbine bypass system during T/G trips at low power.
(1)(2) 0.66W + 61.2 0.65(W - W) + 65.1 0.65W + 60.1 (3) APRM Flow-biased Rod Block AVs (1)    Fixed (%RTP) 108.4 No change TLO Flow-biased (%RTP)
Based on the guidelines in TLTR Section F.4.2.3, the TSV closure scram, TCV fast closure scram, and End of Cycle (EOC)-RPT bypass AL in percent of RTP is reduced by the ratio of the power increase. The new AL does not change with respect to absolute thermal power. ((
(1)(2) 0.66W + 55.7 0.65W + 54.8 (3)  SLO Flow-biased (%RTP)
          )) The maneuvering range for plant startup is maximized.
(1)(2) 0.66W + 50.7 0.65(W - W) + 54.8 0.65W + 49.8 (3) TSV & TCV Scram & EOC-RPT Bypasses
The High Pressure turbine first stage nozzle plates for Limerick, Units 1 and 2, will be modified for TPO operation as described in Section 7.1. The first stage pressure setpoint for the bypass of the TSV closure scram, TCV fast closure scram, and EOC-RPT will be adjusted as necessary to implement the new AL.
(%RTP) 30 29.5 (4)
5-7
MSL High Flow Isolation  % rated steam flow psid  140 126.5  140 128.9  (4) Rod Worth Minimizer  LPSP (%RTP)  LPSP (Steam flow) 10 1.199 Mlbm/hr 10 1.222 Mlbm/hr (4) Notes: (1) No credit is taken in any safety analysis for the flow-biased setpoints. (2) W is % recirculation drive flow where 100% drive flow is that required to achieve 100% core flow at 100% power, and W is the difference between the TLO and SLO drive flow at the same core flow. The current value of W is 7.6% and is not changed. (3) These changes to the ALs are based upon the methodology approved by the NRC in Reference 1. (4) All limits scaled for an uprate of 1.65% thermal.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-1 6.0  ELECTRICAL POWER AND AUXILIARY SYSTEMS 6.1 AC POWER  Plant electrical characterist ics are given in Table 6-1. A detailed comparison of existing ratings with uprated ratings and the effect of the power uprate on the main generator, generator step-up transformers, unit auxiliary transformer, station auxiliary transformer, and regulating transformer are shown in Tables 6-2, 6-3a, 6-3b, 6-4, 6-5a, and 6-5b.
6.1.1 Off-Site Power  The generator, main transformer and isolated phase bus nameplate ratings are listed below: Generator:  The generator is a direct-driven 3-phase 60-HZ, 22,000-Volt, 1800-rpm, hydrogen inner-cooled, synchronous genera tor rated for: 1,264,970 kVA at a 0.9 power factor (PF), with a .58 short circuit ratio at a nominal hydrogen pressure of 75 psig. Unit 1 Main Transformer:  The 1266 Million Volt Amps (MVA) MPT consists of three single-phase, 20.9-230 kV 422 MVA, oil immersed, forced air-cooled, 65&deg;C rise, 60 Hz, outdoor Westinghouse transformers. Unit 2 Main Transformer:  The 1575 MVA MPT consists of three single-phase, 22-531.6 kV 525 MVA, oil immersed, forced ai r-cooled, 65&deg;C rise, 60 Hz, outdoor ABB transformers. Isolated Phase Bus Duct:  The isolated phase bus duct continuous current rating is based on a 65&deg;C rise above a 40&deg;C ambient with forced air cooling. The Main bus is rated at 35,000A, the Auxiliary Branch bus is rated at 2,000A and the transformer Delta bus is rated at 20,000A. The moment ary fault current rating for the Main and Delta buses is 350,000A and the Auxiliary Br anch is 550,000A. The voltage rating of the system is 25,000V. The forced cooling is handled by two air handling units with a design heat transfer ca pacity of 1,810,000 Btu/hr.
The review of the existing off-site electrical equipment concluded the following: The Main Generators will be operating within the existing generating capability curve for TPO uprate. For summer and winter operations, the gross generator MWe output is on the existing generator capability curve with greater than 0.9PF. The isolated phase bus duct is adequate for both rated voltage and low voltage current output.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-2 The main transformers and the associated switchyard components (rated for maximum transformer output) are adequate for the TPO uprate-related transformer output. A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criterion (GDC) 17 (10 CFR 50, Appendix A).
GDC-17 addresses on-site and off-site electrical supply and distribution systems for safety-related components. There is no significant effect on grid stability or reliability. There are no modifications associated with the TPO uprat e, which would increase electrical loads beyond those levels previously incl uded or revise the logic of the distribution systems.
6.1.2 On-Site Power The on-site power distribution system consists of transformers, numerous buses, and switchgear. Alternating current (AC) power to the distribution system is provided from the transmission system or from onsite diesel generators. The on-site power distribution system loads were reviewed under both normal and emergency operating scenarios. In both cases, loads are computed based primarily on equipment nameplate data or brake horsepower (BHP). These loads are used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP level is achieved in both normal and emergency conditions by operating equipment at or below the nameplate rating running KW or BHP. Therefore, there are negligible changes to the load, volta ge drop or short circuit current values. The only identifiable change in electrical load demand is associated with the condensate pumps. These pumps experience increased flow and a sm all change in horsepower duty (~1%) due to the TPO uprate conditions. Accordingly, there are negligible changes in the on-site distribution system design basis loads or voltages due to the TPO conditions. The system environmental design bases are unchanged. Operation at the TPO RTP level is achieved by utilizing existing equipment operating at or below the nameplate rating; therefore, under normal conditions, the


electrical supply and distribu tion components (e.g., switchgear, motor contro l centers (MCCs), cables) are adequate. Station loads under emergency operation and distribution conditions (emergency diesel generators) are based on operational requirements. The ECCS pump loading is based on station UFSAR design basis requirements. Emergency ope ration at the TPO RTP levels is achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the stated pumps. Therefore, under emergency conditions the electrical supply and distribution components are adequate.  
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 5-1 Analytical Limits that Change due to TPO Parameter                                Current                TPO                  Justification APRM High Neutron Flux Scram                                  121                No change APRM Flow-biased Scram Fixed (%RTP)                                        119              No change (1)(2)
TLO Flow-biased (%RTP)                          0.66W + 66.2          0.65W + 65.1                  (3)
(1)(2)
SLO Flow-biased (%RTP)                          0.66W + 61.2      0.65(W - W) + 65.1              (3) 0.65W + 60.1 APRM Flow-biased Rod Block AVs (1)
Fixed (%RTP)                                        108.4              No change (1)(2)
TLO Flow-biased (%RTP)                          0.66W + 55.7          0.65W + 54.8                  (3)
(1)(2)
SLO Flow-biased (%RTP)                          0.66W + 50.7      0.65(W - W) + 54.8              (3) 0.65W + 49.8 TSV & TCV Scram & EOC-RPT Bypasses                              30                  29.5                    (4)
(%RTP)
MSL High Flow Isolation
          % rated steam flow                                  140                  140 (4) psid                                                126.5                128.9 Rod Worth Minimizer LPSP (%RTP)                                          10                    10                      (4)
LPSP (Steam flow)                             1.199 Mlbm/hr        1.222 Mlbm/hr Notes:
(1) No credit is taken in any safety analysis for the flow-biased setpoints.
(2) W is % recirculation drive flow where 100% drive flow is that required to achieve 100% core flow at 100% power, and W is the difference between the TLO and SLO drive flow at the same core flow. The current value of W is 7.6% and is not changed.
(3) These changes to the ALs are based upon the methodology approved by the NRC in Reference 1.
(4) All limits scaled for an uprate of 1.65% thermal.
5-8


No increase in flow or pressure is required of any AC-powered ECCS equipment for the TPO uprate. Therefore, the amount of power required to perform safety-related functions (pump and valve loads) does not increase, and the current emergency power system remains adequate. The systems have sufficient capacity to support all required loads for safe shutdown, to maintain a NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-3 safe shutdown condition, and to operate the engineered safety feature equipment following postulated accidents. Since the duty cycle and duration for design basis EDG loads is ba sed on analytical power levels of at least 102% of the current licensed thermal power
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.0 ELECTRICAL POWER AND AUXILIARY SYSTEMS 6.1  AC POWER Plant electrical characteristics are given in Table 6-1.
, these will remain unchanged by TPO. Hence, the required reserve volume of emergency fuel oil is not changed. Therefore, useable emergency fuel oil reserves will be adequate to support TPO.
A detailed comparison of existing ratings with uprated ratings and the effect of the power uprate on the main generator, generator step-up transformers, unit auxiliary transformer, station auxiliary transformer, and regulating transformer are shown in Tables 6-2, 6-3a, 6-3b, 6-4, 6-5a, and 6-5b.
6.2 DC POWER The direct current (DC) loading requirements documented in the UFSAR and station load calculations were reviewed, and no reactor power-dependent loads were identified. The DC power distribution system provides control and motive power for various systems and components. These loads are used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP level does not increase any loads or revise control logic. Therefore, there are no changes to the load, voltage drop or short circ uit current values. 6.3 F UEL P OOL  The following subsections address fuel pool coo ling, crud and corrosion products in the fuel pool, radiation levels and structural adequacy of the fuel racks. The changes due to TPO are within the design limits of the system and its components. The fuel pool cooling system meets the UFSAR requirements at the TPO conditions.
6.1.1   Off-Site Power The generator, main transformer and isolated phase bus nameplate ratings are listed below:
6.3.1 Fuel Pool Cooling The Spent Fuel Pool (SFP) heat load increases slightly at TPO but remains within the capability of the Fuel Pool Cooling and Cleanup System (FPCC) as supplemented by the Residual Heat Removal System (RHR). The TPO uprate does not affect the heat removal capability of the FPCC or RHR systems and the TPO heat load is within the design basis heat load for the FPCC and RHR systems as shown in Table 6-6. The FPCC and RHR heat exchangers are sufficient to remove the decay heat during a normal batch and full-core refueling. The SFP makeup requirement increases slightly at TPO, however makeup capacity remains sufficient.
* Generator: The generator is a direct-driven 3-phase 60-HZ, 22,000-Volt, 1800-rpm, hydrogen inner-cooled, synchronous generator rated for: 1,264,970 kVA at a 0.9 power factor (PF), with a .58 short circuit ratio at a nominal hydrogen pressure of 75 psig.
6.3.2 Crud Activity and Corrosion Products The crud activity and corrosion products associated with spent fuel can increase very slightly due to the TPO. The increase is insignificant and SFP water quality is maintained by the FPCC.
* Unit 1 Main Transformer: The 1266 Million Volt Amps (MVA) MPT consists of three single-phase, 20.9-230 kV 422 MVA, oil immersed, forced air-cooled, 65&deg;C rise, 60 Hz, outdoor Westinghouse transformers.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-4 6.3.3 Radiation Levels The normal radiation levels around the SFP may increa se slightly during fuel handling operation. This increase is acceptable and does not significantly increase the operational doses to personnel or equipment.
* Unit 2 Main Transformer: The 1575 MVA MPT consists of three single-phase, 22-531.6 kV 525 MVA, oil immersed, forced air-cooled, 65&deg;C rise, 60 Hz, outdoor ABB transformers.
6.3.4 Fuel Racks There is no effect on the design of the fuel racks because the maximum allowable spent fuel temperature is not being increased. 6.4 W ATER SYSTEMS The safety-related and non-safety-related cooling water loads potentially affected by TPO are addressed in the following sections. The environmental effects of TPO are controlled such that none of the present limits (e.g., maximum allowed cooling water discharge temperature) are increased.
* Isolated Phase Bus Duct: The isolated phase bus duct continuous current rating is based on a 65&deg;C rise above a 40&deg;C ambient with forced air cooling. The Main bus is rated at 35,000A, the Auxiliary Branch bus is rated at 2,000A and the transformer Delta bus is rated at 20,000A. The momentary fault current rating for the Main and Delta buses is 350,000A and the Auxiliary Branch is 550,000A. The voltage rating of the system is 25,000V. The forced cooling is handled by two air handling units with a design heat transfer capacity of 1,810,000 Btu/hr.
6.4.1 Service Water Systems 6.4.1.1 Safety-Related Loads Emergency Service Wa ter The safety-related Emergency Service Water (ESW) system provides cooling water to essential equipment during and following a design basis accident, such as a Loss of Offsite Power (LOOP) or LOCA. The performance of the ESW system during these events does not change for TPO because the original LOCA analysis and containment response analysis were based on 102% of CLTP, the bounding power level for the TPO analys is. The increases in the heat loads to equipment cooled by ESW are within the existing capacity of the ESW system. Residual Heat Removal Service Water The required design performance of the Residual Heat Removal Service Water System (RHRSW) does not change for TPO because the original LOCA analysis and containment response analysis were based on at least 102% of CLTP, the bounding analytical power level for TPO. The increases in the heat loads to equipment cooled by RHRSW are within the existing capacity of the RHRSW system.
The review of the existing off-site electrical equipment concluded the following:
6.4.1.2 Non-safety-Related Loads  The major operational heat load increases to the Service Water (SW) system from TPO reflect an operational increase in main generator losses rejected to the genera tor hydrogen coolers, generator stator coolers, Alterrex air cooler, and Iso-phase bus coolers. The resulting design heat loads for the SW system are <0.2% above CLTP. The increases in heat loads to equipment NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-5 cooled by the SW system are insignificant and the design of the system is adequate to accommodate for TPO.
* The Main Generators will be operating within the existing generating capability curve for TPO uprate. For summer and winter operations, the gross generator MWe output is on the existing generator capability curve with greater than 0.9PF.
6.4.2 Main Condenser/Circulating Water/N ormal Heat Sink Performance  The main condenser, circulating water, and normal heat sink systems are designed to remove the heat rejected to the condenser and thereby ma intain adequately low condenser pressure as recommended by the turbine vendor.
* The isolated phase bus duct is adequate for both rated voltage and low voltage current output.
TPO operation increases the he at rejected to the condenser and may reduce the difference between the operating pressure and the required minimum condenser vacuum. The performance of the main condenser was evaluated for operation at the TPO RTP. The evaluation confirms that the condenser, circulating water system and heat sink are adequate for TPO operation. Note that based on weather conditions, unit derates may be necessary to maintain adequa te condenser backpressure.
6-1
6.4.2.1 Discharge Limits  The Pennsylvania Department of Environmental Quality National Pollutant Discharge Elimination System (NPDES) Permit provides the effluent limitations and monitoring requirements for discharge wastewater at the site. The discharge limits on residual oxides and Spectrus CT1300 are daily maximums of 0.2 mg/l and 0.4 mg/l (respectivel y). The discharge water temperature shall not exceed a maximum of 110&deg;F, or raise the river temperature 5&deg;F above its ambient temperature. When river temperatures exceed 87&deg;F, discharge from the plant cannot raise the river temperature at all. Frequent monitoring of these parameters ensures that permit limits are not exceeded. The TPO uprate has minimal effect on the parameters, and no changes to NPDES permit requirements are needed. The state thermal discharge limits, the current discharges, and bounding analysis discharges for the TPO uprate are shown in Table 6-7. This comparison demonstrates that the plant remains within the state discharge limits, during operation at TPO conditions.
6.4.3 Reactor Enclosure Cooling Water System  The heat loads on the Reactor Enclosure Cooling Water (RECW) system do not increase significantly due to TPO. The main power-dependent heat loads on the RECW system that are increased by the TPO, are those related to the operation of the Reactor Water Cleanup non-regenerative heat exchangers, Reactor Water Cleanup recirculation pumps, and the reactor recirculation pumps. The design of the RECW heat exchangers is adequate to accommodate a heat load increase of <1.2% for normal operations and ~2.3% for emergency operations. Changes to the RECW system heat loads are minimal and will result in a negligible temperature increase of ~0.4&deg;F for the RECW system during normal operation. The RECW system experiences a slight heat load increase associated with the Fuel Pool Coolers heat exchangers during emergency operations. However, the system has adequate design margin to remove the additional heat. Therefore, the RECW system is acceptable for the TPO uprate.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-6 6.4.4 Turbine Enclosure Cooling Water System  The power-dependent heat loads on the Turbine Enclosure Cooling Water (TECW) system that are increased by the TPO, are those related to the operation of the condensate pump motor bearing oil coolers and turbine enclosure sample station. The remaining TECW heat loads are not strongly dependent upon reactor power and do not significantly increase. The TECW system has sufficient capacity to assure that adequate heat removal capability is available for TPO operation.
6.4.5 Ultimate Heat Sink  The ultimate heat sink (UHS) for Limerick is the spray pond. The ESW and RHRSW systems provide water from the UHS for equipment coo ling throughout the plant.
As a result of operation at the TPO RTP level, the post-LOCA heat load increases slightly, primarily due to higher reactor decay heat. However, the ability of the UHS to perform re quired safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the UHS are within previous evaluations a nd the requirements are unchanged for TPO uprate conditions. The current Techni cal Specifications for UHS limits are adequate due to conservatism in the current design. 6.5 STANDBY L IQUID CONTROL S YSTEM  The SLCS is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that all or some of the control rods cannot be inserted. This manually or automatically operated system pumps a highly enriched sodium pentaborate solution into the vessel to achieve a sub critic al condition. The generic ev aluation presented in TLTR Section 5.6.5 (SLCS) and Appendix L.3 (ATWS Evaluation) was not applicable to the Limerick TPO uprate. The SLCS performance was analyzed to compare the maximum discharge pressure of the SLCS pump to the SLCS relief valve nominal set pressure.
The SLCS analysis concluded that there is insufficient margin between the relief valve nominal setpoint and the maximum expected pump discharge pressure based on a nominal SLCS set pressure of 1400 psig for three SLCS pump operation. However, two SLCS pump operation results maintain adequate margin between the discharge pressure of the SLCS pump and the nominal set pressure of the SLCS relief valve op ening setpoint accounting fo r setpoint tolerance.
The TPO uprate does not affect shutdown or injection capability of the SLCS system in two SLCS pump operation. Because the shutdown margin is reload dependent, the shutdown margin and the required reactor boron concentration are confirmed for each reload core. The SLCS ATWS performance is evaluated in TSAR Section 9.3.1. The evaluation shows that the TPO has no adverse effect on the ability of the SLCS to mitigate an ATWS in two SLCS pump operation.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-7 6.6 P OWER-DEPENDENT HEATING , VENTILATION AND A IR CONDITIONING The Heating, Ventilation and Air Conditioning (HVAC) systems that are potentially affected by the TPO uprate consist mainly of heating, cooling supply, exhaust, and recirculation units in the turbine enclosure, reactor enclosure (including steam tunnel), and primary containment. TPO results in a minor increase in the heat load caused by the slightly higher FW process temperature (~2&deg;F). The increased heat load is within the margin of the steam tunnel area coolers. In the drywell, the increase in heat load due to the FW process temperature is within the system capacity. In the turbine enclosure, the temperature increases are expected to be very low due to the increase in the FW process temperatures. In the reactor enclosure, the increase in heat load caused by the slightly higher FW process temperature is within the margin of the area coolers. Other areas are unaffected by the TPO because the process temperatures and electrical heat loads remain constant.
Therefore, the power-dependent HVAC system s are adequate to support the TPO uprate. 6.7 FIRE PROTECTION Operation of the plant at the TP O RTP level does not affect the fire suppression or detection systems. There is no change in the physical plant configuration and the potential for minor changes to combustible loading as a result of the TPO uprate are addressed by controlled design change procedures (e.g., the new FW Ultrasonic Flow Measurement (UFM) equipment).
The Limerick fire safe shutdown analysis was performed at 3622 MWt (i.e., 110% of OLTP) and thus bounds operation at the TPO power level of 3515 MWt. The fire safe shutdown analysis includes consideration of equipment needed to achieve and maintain hot shutdown, fire barriers, operator manual actions, personnel resources, a nd repair activities credited to achieve and maintain cold shutdown. Thus, the fire safe s hutdown analysis is acceptable for TPO operation.
6.7.1 10 CFR 50 Appendix R Fire Event  The Limerick fire safe shutdown analysis was performed at 3622 MWt (i.e., 110% of OLTP) and thus bounds operation at the TPO power level of 3515 MWt. Thus, the fire safe shutdown analysis is acceptable for TPO operation. 6.8 S YSTEMS N OT AFFECTED B Y TPO UPRATE  Based on experience and previous NRC reviews, all systems that are significantly affected by TPO are addressed in this report. Other systems not addressed by this report are not significantly affected by TPO. The systems unaffected by TPO at Limerick are confirmed to be consistent with the generic description provided in the TLTR.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-8 Table 6-1 TPO Plant Electrical Characteristics Parameter Value Generator Output (MWe) 1138.47 Rated Voltage (kV) 22 Power Factor 0.90 Generator Output (MVA) 1264.97 Current Output (Amps) 33,197 Isolated Phase Bus Duct Rating: (Amps) Main Section  Branch Section 35,000 2,000 Main Transformers Rating (MVA)
Unit 1 Unit 2  1,266 1,575  Table 6-2 Main Generator Ratings Comparison Design Max. Nominal Power Level MVA @ 75 psig H2 MWe @ 75 psig H2 MVAR @ 75 psig H2 Existing 1264.97 1138.47 551.39 Uprated (1) 1264.97 1138.47 551.39 (1) Operation at the uprate d condition is not expected to have any effect on the operation of the Main Generator. Operation in this range is still within the op erating boundaries specified in station design analysis and operating procedures.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-9 Table 6-3a Limerick 1 Generator Step-up Transformer Ratings Comparison Power Level Design MVA @ 65&#xba;C MVA Loading Existing 1266 1211 Uprated (1) 1266 1211 (1) Operation at the uprated condition is not expect ed to have any effect on the operation of the Generator Step-up Transformer. The generator MWe will increase and MVAR will decrease, thus MVA will remain the same. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.
Table 6-3b Limerick 2 Generator Step-up Transformer Ratings Comparison Power Level Design MVA @ 65&#xba;C MVA Loading Existing 1266 1213 Uprated (1)(2) 1575 1213 (1) Operation at the uprated condition is not expect ed to have any effect on the operation of the Generator Step-up Transformer. The generator MWe will increase and MVAR will decrease, thus MVA will remain the same. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures. (2) The Unit 2 Generator Step-up Transformer is planned for replacement in 2011. The new transformer will have an increased design MVA rating.
Table 6-4 Unit Auxiliary Transformer Ratings Comparison Power Level Rated MVA @ 65&#xba;C Existing MVA Loading TPO MVA Loading Unit 1 (1)  52.5 49.025 49.139 Unit 2 (1)  52.5 47.3 47.414 (1) Operation at the uprated condition is not exp ected to have any effect on the operation of the Unit Auxiliary Transformer. Operation in this range is still within the operating boundaries specified in stat ion design analysis and operating procedures.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-10 Table 6-5a Station Auxiliary Transformer Ratings Comparison Power Level Rated MVA @ 65&#xba;C Existing MVA Loading TPO MVA Loading Unit 1 (1)  61.6 45.444 45.558 (1) Operation at the uprated condition is not exp ected to have any effect on the operation of the Station Auxiliary Transformer. Operation in this range is still within the operating boundaries specified in stat ion design analysis and operating procedures.
Table 6-5b Regulating Transformer Ratings Comparison Power Level Rated MVA @ 65&#xba;C Existing MVA Loading TPO MVA Loading Unit 2 (1) 58 43.109 43.223 (1) Operation at the uprated condition is not exp ected to have any effect on the operation of the Regulating Transformer. Operation in this range is still within the operating boundaries specified in stat ion design analysis and operating procedures.
Table 6-6: Fuel Pool Cooling and Cleanup Parameters CLTP TPO Parameter Normal Batch Full Core Normal Batch Full Core Bundles offloaded 280 764 280 764 Maximum fuel moves per hour 10 10 10 10 Time from shutdown to beginning of fuel transfer  (hours) 40 40 40 40 Heat load at maximum temperature (MBTU/hr) 24.40 46.76 24.77 48.01 Maximum bulk temperature  (&#xba;F) 117 138 117 139 Maximum bulk temperature limit  (&#xba;F) 140 1 140 140 1 140 Time to boil  (hrs) 9.25 3.57 9.10 3.46 Note 1:  143 F when utilizing FPCC only


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6-11 Table 6-7 Effluent Discharge Comparison Parameter State Limit Current TPO Discharge Temperature ( F) 110  89.1 1 Insignificant change Residual Oxides mg/L 0.2 0.1 2 Insignificant change Spectrus CT1300 mg/L 0.4 < 0.05 2 Insignificant change 1. Maximum discharge temperature for 2009. Taken from NPDES Sampling at Outfall 001 on July 29, 2009. 2. Maximum value from weekly grab samples during April 2009.  
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION
* The main transformers and the associated switchyard components (rated for maximum transformer output) are adequate for the TPO uprate-related transformer output.
A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criterion (GDC) 17 (10 CFR 50, Appendix A).
GDC-17 addresses on-site and off-site electrical supply and distribution systems for safety-related components. There is no significant effect on grid stability or reliability. There are no modifications associated with the TPO uprate, which would increase electrical loads beyond those levels previously included or revise the logic of the distribution systems.
6.1.2  On-Site Power The on-site power distribution system consists of transformers, numerous buses, and switchgear.
Alternating current (AC) power to the distribution system is provided from the transmission system or from onsite diesel generators. The on-site power distribution system loads were reviewed under both normal and emergency operating scenarios. In both cases, loads are computed based primarily on equipment nameplate data or brake horsepower (BHP). These loads are used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP level is achieved in both normal and emergency conditions by operating equipment at or below the nameplate rating running KW or BHP. Therefore, there are negligible changes to the load, voltage drop or short circuit current values.
The only identifiable change in electrical load demand is associated with the condensate pumps.
These pumps experience increased flow and a small change in horsepower duty (~1%) due to the TPO uprate conditions. Accordingly, there are negligible changes in the on-site distribution system design basis loads or voltages due to the TPO conditions. The system environmental design bases are unchanged. Operation at the TPO RTP level is achieved by utilizing existing equipment operating at or below the nameplate rating; therefore, under normal conditions, the electrical supply and distribution components (e.g., switchgear, motor control centers (MCCs),
cables) are adequate.
Station loads under emergency operation and distribution conditions (emergency diesel generators) are based on operational requirements. The ECCS pump loading is based on station UFSAR design basis requirements. Emergency operation at the TPO RTP levels is achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the stated pumps. Therefore, under emergency conditions the electrical supply and distribution components are adequate.
No increase in flow or pressure is required of any AC-powered ECCS equipment for the TPO uprate. Therefore, the amount of power required to perform safety-related functions (pump and valve loads) does not increase, and the current emergency power system remains adequate. The systems have sufficient capacity to support all required loads for safe shutdown, to maintain a 6-2


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 7-1 7.0 POWER CONVERSION SYSTEMS 7.1 TURBINE-GENERATOR General Electric Energy Services and Siemens performed the evaluation of the steam turbine, valves, turbine auxiliary systems, cross around relief valves and piping for the TPO condition. A summary of the results of the evaluation are presented as follows:
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION safe shutdown condition, and to operate the engineered safety feature equipment following postulated accidents.
For the turbine high pre ssure (HP) section, the existing nozzle plates ar e not able to pass the required additional steam flow at the TPO operation point and still maintain sufficient flow margin of approximately 3% for reactor pressure control. New first stage nozzle plates designed with increased flow area are required; these modified nozzle plates will allow the HP turbine to maintain flow margin of approximately 3% and thus maintain adequate pressure control. All other components in the HP sec tion are within allowable design limits and no other changes are recommended or required.
Since the duty cycle and duration for design basis EDG loads is based on analytical power levels of at least 102% of the current licensed thermal power, these will remain unchanged by TPO.
Hence, the required reserve volume of emergency fuel oil is not changed. Therefore, useable emergency fuel oil reserves will be adequate to support TPO.
6.2    DC POWER The direct current (DC) loading requirements documented in the UFSAR and station load calculations were reviewed, and no reactor power-dependent loads were identified. The DC power distribution system provides control and motive power for various systems and components. These loads are used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP level does not increase any loads or revise control logic. Therefore, there are no changes to the load, voltage drop or short circuit current values.
6.3    FUEL POOL The following subsections address fuel pool cooling, crud and corrosion products in the fuel pool, radiation levels and structural adequacy of the fuel racks. The changes due to TPO are within the design limits of the system and its components. The fuel pool cooling system meets the UFSAR requirements at the TPO conditions.
6.3.1  Fuel Pool Cooling The Spent Fuel Pool (SFP) heat load increases slightly at TPO but remains within the capability of the Fuel Pool Cooling and Cleanup System (FPCC) as supplemented by the Residual Heat Removal System (RHR). The TPO uprate does not affect the heat removal capability of the FPCC or RHR systems and the TPO heat load is within the design basis heat load for the FPCC and RHR systems as shown in Table 6-6.
The FPCC and RHR heat exchangers are sufficient to remove the decay heat during a normal batch and full-core refueling. The SFP makeup requirement increases slightly at TPO, however makeup capacity remains sufficient.
6.3.2  Crud Activity and Corrosion Products The crud activity and corrosion products associated with spent fuel can increase very slightly due to the TPO. The increase is insignificant and SFP water quality is maintained by the FPCC.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.3.3  Radiation Levels The normal radiation levels around the SFP may increase slightly during fuel handling operation.
This increase is acceptable and does not significantly increase the operational doses to personnel or equipment.
6.3.4  Fuel Racks There is no effect on the design of the fuel racks because the maximum allowable spent fuel temperature is not being increased.
6.4  WATER SYSTEMS The safety-related and non-safety-related cooling water loads potentially affected by TPO are addressed in the following sections. The environmental effects of TPO are controlled such that none of the present limits (e.g., maximum allowed cooling water discharge temperature) are increased.
6.4.1  Service Water Systems 6.4.1.1 Safety-Related Loads Emergency Service Water The safety-related Emergency Service Water (ESW) system provides cooling water to essential equipment during and following a design basis accident, such as a Loss of Offsite Power (LOOP) or LOCA. The performance of the ESW system during these events does not change for TPO because the original LOCA analysis and containment response analysis were based on 102% of CLTP, the bounding power level for the TPO analysis. The increases in the heat loads to equipment cooled by ESW are within the existing capacity of the ESW system.
Residual Heat Removal Service Water The required design performance of the Residual Heat Removal Service Water System (RHRSW) does not change for TPO because the original LOCA analysis and containment response analysis were based on at least 102% of CLTP, the bounding analytical power level for TPO. The increases in the heat loads to equipment cooled by RHRSW are within the existing capacity of the RHRSW system.
6.4.1.2 Non-safety-Related Loads The major operational heat load increases to the Service Water (SW) system from TPO reflect an operational increase in main generator losses rejected to the generator hydrogen coolers, generator stator coolers, Alterrex air cooler, and Iso-phase bus coolers. The resulting design heat loads for the SW system are <0.2% above CLTP. The increases in heat loads to equipment 6-4
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION cooled by the SW system are insignificant and the design of the system is adequate to accommodate for TPO.
6.4.2  Main Condenser/Circulating Water/Normal Heat Sink Performance The main condenser, circulating water, and normal heat sink systems are designed to remove the heat rejected to the condenser and thereby maintain adequately low condenser pressure as recommended by the turbine vendor. TPO operation increases the heat rejected to the condenser and may reduce the difference between the operating pressure and the required minimum condenser vacuum. The performance of the main condenser was evaluated for operation at the TPO RTP. The evaluation confirms that the condenser, circulating water system and heat sink are adequate for TPO operation. Note that based on weather conditions, unit derates may be necessary to maintain adequate condenser backpressure.
6.4.2.1 Discharge Limits The Pennsylvania Department of Environmental Quality National Pollutant Discharge Elimination System (NPDES) Permit provides the effluent limitations and monitoring requirements for discharge wastewater at the site. The discharge limits on residual oxides and Spectrus CT1300 are daily maximums of 0.2 mg/l and 0.4 mg/l (respectively). The discharge water temperature shall not exceed a maximum of 110&deg;F, or raise the river temperature 5&deg;F above its ambient temperature. When river temperatures exceed 87&deg;F, discharge from the plant cannot raise the river temperature at all. Frequent monitoring of these parameters ensures that permit limits are not exceeded. The TPO uprate has minimal effect on the parameters, and no changes to NPDES permit requirements are needed.
The state thermal discharge limits, the current discharges, and bounding analysis discharges for the TPO uprate are shown in Table 6-7. This comparison demonstrates that the plant remains within the state discharge limits, during operation at TPO conditions.
6.4.3  Reactor Enclosure Cooling Water System The heat loads on the Reactor Enclosure Cooling Water (RECW) system do not increase significantly due to TPO. The main power-dependent heat loads on the RECW system that are increased by the TPO, are those related to the operation of the Reactor Water Cleanup non-regenerative heat exchangers, Reactor Water Cleanup recirculation pumps, and the reactor recirculation pumps. The design of the RECW heat exchangers is adequate to accommodate a heat load increase of <1.2% for normal operations and ~2.3% for emergency operations.
Changes to the RECW system heat loads are minimal and will result in a negligible temperature increase of ~0.4&deg;F for the RECW system during normal operation. The RECW system experiences a slight heat load increase associated with the Fuel Pool Coolers heat exchangers during emergency operations. However, the system has adequate design margin to remove the additional heat. Therefore, the RECW system is acceptable for the TPO uprate.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.4.4  Turbine Enclosure Cooling Water System The power-dependent heat loads on the Turbine Enclosure Cooling Water (TECW) system that are increased by the TPO, are those related to the operation of the condensate pump motor bearing oil coolers and turbine enclosure sample station. The remaining TECW heat loads are not strongly dependent upon reactor power and do not significantly increase. The TECW system has sufficient capacity to assure that adequate heat removal capability is available for TPO operation.
6.4.5  Ultimate Heat Sink The ultimate heat sink (UHS) for Limerick is the spray pond. The ESW and RHRSW systems provide water from the UHS for equipment cooling throughout the plant. As a result of operation at the TPO RTP level, the post-LOCA heat load increases slightly, primarily due to higher reactor decay heat. However, the ability of the UHS to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the UHS are within previous evaluations and the requirements are unchanged for TPO uprate conditions. The current Technical Specifications for UHS limits are adequate due to conservatism in the current design.
6.5    STANDBY LIQUID CONTROL SYSTEM The SLCS is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that all or some of the control rods cannot be inserted. This manually or automatically operated system pumps a highly enriched sodium pentaborate solution into the vessel to achieve a sub critical condition. The generic evaluation presented in TLTR Section 5.6.5 (SLCS) and Appendix L.3 (ATWS Evaluation) was not applicable to the Limerick TPO uprate. The SLCS performance was analyzed to compare the maximum discharge pressure of the SLCS pump to the SLCS relief valve nominal set pressure.
The SLCS analysis concluded that there is insufficient margin between the relief valve nominal setpoint and the maximum expected pump discharge pressure based on a nominal SLCS set pressure of 1400 psig for three SLCS pump operation. However, two SLCS pump operation results maintain adequate margin between the discharge pressure of the SLCS pump and the nominal set pressure of the SLCS relief valve opening setpoint accounting for setpoint tolerance.
The TPO uprate does not affect shutdown or injection capability of the SLCS system in two SLCS pump operation. Because the shutdown margin is reload dependent, the shutdown margin and the required reactor boron concentration are confirmed for each reload core.
The SLCS ATWS performance is evaluated in TSAR Section 9.3.1. The evaluation shows that the TPO has no adverse effect on the ability of the SLCS to mitigate an ATWS in two SLCS pump operation.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.6  POWER-DEPENDENT HEATING, VENTILATION AND AIR CONDITIONING The Heating, Ventilation and Air Conditioning (HVAC) systems that are potentially affected by the TPO uprate consist mainly of heating, cooling supply, exhaust, and recirculation units in the turbine enclosure, reactor enclosure (including steam tunnel), and primary containment.
TPO results in a minor increase in the heat load caused by the slightly higher FW process temperature (~2&deg;F). The increased heat load is within the margin of the steam tunnel area coolers. In the drywell, the increase in heat load due to the FW process temperature is within the system capacity. In the turbine enclosure, the temperature increases are expected to be very low due to the increase in the FW process temperatures. In the reactor enclosure, the increase in heat load caused by the slightly higher FW process temperature is within the margin of the area coolers. Other areas are unaffected by the TPO because the process temperatures and electrical heat loads remain constant.
Therefore, the power-dependent HVAC systems are adequate to support the TPO uprate.
6.7  FIRE PROTECTION Operation of the plant at the TPO RTP level does not affect the fire suppression or detection systems. There is no change in the physical plant configuration and the potential for minor changes to combustible loading as a result of the TPO uprate are addressed by controlled design change procedures (e.g., the new FW Ultrasonic Flow Measurement (UFM) equipment).
The Limerick fire safe shutdown analysis was performed at 3622 MWt (i.e., 110% of OLTP) and thus bounds operation at the TPO power level of 3515 MWt. The fire safe shutdown analysis includes consideration of equipment needed to achieve and maintain hot shutdown, fire barriers, operator manual actions, personnel resources, and repair activities credited to achieve and maintain cold shutdown. Thus, the fire safe shutdown analysis is acceptable for TPO operation.
6.7.1  10 CFR 50 Appendix R Fire Event The Limerick fire safe shutdown analysis was performed at 3622 MWt (i.e., 110% of OLTP) and thus bounds operation at the TPO power level of 3515 MWt. Thus, the fire safe shutdown analysis is acceptable for TPO operation.
6.8  SYSTEMS NOT AFFECTED BY TPO UPRATE Based on experience and previous NRC reviews, all systems that are significantly affected by TPO are addressed in this report. Other systems not addressed by this report are not significantly affected by TPO. The systems unaffected by TPO at Limerick are confirmed to be consistent with the generic description provided in the TLTR.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-1 TPO Plant Electrical Characteristics Parameter                          Value Generator Output (MWe)                                1138.47 Rated Voltage (kV)                                        22 Power Factor                                            0.90 Generator Output (MVA)                                1264.97 Current Output (Amps)                                  33,197 Isolated Phase Bus Duct Rating: (Amps)
Main Section                                      35,000 Branch Section                                    2,000 Main Transformers Rating (MVA)
Unit 1                                            1,266 Unit 2                                            1,575 Table 6-2 Main Generator Ratings Comparison Power Level            Design            Max. Nominal MVA @ 75 psig MWe @ 75 psig  MVAR @ 75 H2          H2          psig H2 Existing                      1264.97      1138.47        551.39 (1)
Uprated                      1264.97      1138.47        551.39 (1)
Operation at the uprated condition is not expected to have any effect on the operation of the Main Generator. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-3a Limerick 1 Generator Step-up Transformer Ratings Comparison Power Level            Design MVA @ 65&#xba;C                MVA Loading Existing                            1266                          1211 Uprated(1)                          1266                          1211 (1)
Operation at the uprated condition is not expected to have any effect on the operation of the Generator Step-up Transformer. The generator MWe will increase and MVAR will decrease, thus MVA will remain the same. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.
Table 6-3b Limerick 2 Generator Step-up Transformer Ratings Comparison Power Level            Design MVA @ 65&#xba;C                MVA Loading Existing                            1266                          1213 (1)(2)
Uprated                              1575                          1213 (1)
Operation at the uprated condition is not expected to have any effect on the operation of the Generator Step-up Transformer. The generator MWe will increase and MVAR will decrease, thus MVA will remain the same. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.
(2)
The Unit 2 Generator Step-up Transformer is planned for replacement in 2011. The new transformer will have an increased design MVA rating.
Table 6-4 Unit Auxiliary Transformer Ratings Comparison Power Level          Rated MVA @ 65&#xba;C        Existing MVA Loading    TPO MVA Loading Unit 1(1)                          52.5                    49.025                49.139 (1)
Unit 2                              52.5                      47.3                47.414 (1)
Operation at the uprated condition is not expected to have any effect on the operation of the Unit Auxiliary Transformer. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-5a Station Auxiliary Transformer Ratings Comparison Power Level              Rated MVA @ 65&#xba;C                  Existing MVA Loading      TPO MVA Loading (1)
Unit 1                                  61.6                            45.444                  45.558 (1)
Operation at the uprated condition is not expected to have any effect on the operation of the Station Auxiliary Transformer. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.
Table 6-5b Regulating Transformer Ratings Comparison Power Level              Rated MVA @ 65&#xba;C                  Existing MVA Loading      TPO MVA Loading Unit 2(1)                                58                              43.109                  43.223 (1)
Operation at the uprated condition is not expected to have any effect on the operation of the Regulating Transformer. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.
Table 6-6: Fuel Pool Cooling and Cleanup Parameters CLTP                  TPO Parameter Normal                  Normal Full Core              Full Core Batch                  Batch Bundles offloaded                                                    280            764      280          764 Maximum fuel moves per hour                                          10            10      10            10 Time from shutdown to beginning of fuel transfer (hours)              40            40      40            40 Heat load at maximum temperature (MBTU/hr)                          24.40          46.76    24.77        48.01 Maximum bulk temperature (&#xba;F)                                        117            138      117          139 1                      1 Maximum bulk temperature limit (&#xba;F)                                  140            140      140          140 Time to boil (hrs)                                                  9.25          3.57    9.10          3.46 Note 1: 143&deg;F when utilizing FPCC only 6-10
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-7 Effluent Discharge Comparison Parameter                        State        Current              TPO Limit Discharge Temperature (&deg;F)                        110          89.11          Insignificant change Residual Oxides mg/L                              0.2            0.12          Insignificant change Spectrus CT1300 mg/L                              0.4          < 0.052          Insignificant change
: 1. Maximum discharge temperature for 2009. Taken from NPDES Sampling at Outfall 001 on July 29, 2009.
: 2. Maximum value from weekly grab samples during April 2009.
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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 7.0 POWER CONVERSION SYSTEMS 7.1   TURBINE-GENERATOR General Electric Energy Services and Siemens performed the evaluation of the steam turbine, valves, turbine auxiliary systems, cross around relief valves and piping for the TPO condition. A summary of the results of the evaluation are presented as follows:
For the turbine high pressure (HP) section, the existing nozzle plates are not able to pass the required additional steam flow at the TPO operation point and still maintain sufficient flow margin of approximately 3% for reactor pressure control. New first stage nozzle plates designed with increased flow area are required; these modified nozzle plates will allow the HP turbine to maintain flow margin of approximately 3% and thus maintain adequate pressure control. All other components in the HP section are within allowable design limits and no other changes are recommended or required.
The turbine low pressure (LP) section rotor and all LP components are within allowable design margins and no changes are recommended or required.
The turbine low pressure (LP) section rotor and all LP components are within allowable design margins and no changes are recommended or required.
Main stop valves, control valves, and combined intermediate valves are all within allowable design margins to operate at the TPO flow condition.
Main stop valves, control valves, and combined intermediate valves are all within allowable design margins to operate at the TPO flow condition.
The turbine auxiliary systems were evaluated; no modifications are ne eded to support operation at the TPO uprate condition.
The turbine auxiliary systems were evaluated; no modifications are needed to support operation at the TPO uprate condition.
The existing missile analysis was evaluated for TPO conditions. The assumptions in the existing missile analysis bound operation at TPO conditions. Thus, the missile generation probability remains unchanged and is therefore acceptable. The overspeed evaluation was reviewed for TPO conditions. The assumptions in the existing overspeed evaluation bound operation at TPO conditions. Thus, the overspeed evaluation remains acceptable for TPO operation. No change in the overspeed trip settings is required. 7.2 CONDENSER A ND STEAM J ET A IR EJECTORS The main condenser capability was evaluated for performance at the TPO uprate conditions in  
The existing missile analysis was evaluated for TPO conditions. The assumptions in the existing missile analysis bound operation at TPO conditions. Thus, the missile generation probability remains unchanged and is therefore acceptable.
The overspeed evaluation was reviewed for TPO conditions. The assumptions in the existing overspeed evaluation bound operation at TPO conditions. Thus, the overspeed evaluation remains acceptable for TPO operation. No change in the overspeed trip settings is required.
7.2   CONDENSER AND STEAM JET AIR EJECTORS The main condenser capability was evaluated for performance at the TPO uprate conditions in Section 6.4.2. Air leakage into the condenser does not increase as a result of the TPO uprate.
The small increase in hydrogen and oxygen flows from the reactor does not affect the Steam Jet Air Ejector (SJAE) capacity because the design was based on operation at greater than required flows at uprate conditions. Therefore, the condenser air removal system is not affected by the TPO uprate and the SJAEs are adequate for operation at the TPO uprate conditions.
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Section 6.4.2. Air leakage into the condenser does not increase as a result of the TPO uprate. The small increase in hydrogen and oxygen flows from the reactor does not affect the Steam Jet Air Ejector (SJAE) capacity beca use the design was based on opera tion at greater than required flows at uprate conditions. Therefore, the condenser air removal system is not affected by the TPO uprate and the SJAEs are adequate for operation at the TPO uprate conditions.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 7.3   TURBINE STEAM BYPASS The Steam Bypass Pressure Control System (SBPCS) was originally designed for a steam flow capacity of approximately 25% of the 100% rated flow at CLTP. The steam bypass capacity at the TPO RTP is approximately 24.3% of the 100% TPO RTP steam flow rate. The steam bypass system is non-safety-related. While the bypass capacity as a percent of rated steam flow is reduced, the actual steam bypass capacity is unchanged. The transient analyses that credit the turbine bypass system use a bypass capacity that is less than the actual capacity. Therefore, the turbine bypass capacity remains adequate for TPO operation because the actual capacity (unchanged) continues to bound the value used in the analyses.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 7-2 7.3 TURBINE STEAM BYPASS The Steam Bypass Pressure Control System (SBPCS) was originally designed for a steam flow capacity of approximately 25% of the 100% rated flow at CLTP. The steam bypass capacity at the TPO RTP is approximately 24.3% of the 100% TPO RTP steam flow rate. The steam bypass system is non-safety-related. While the bypass capacity as a percent of rated steam flow is reduced, the actual steam bypass capacity is unchanged. The transien t analyses that credit the turbine bypass system use a bypass ca pacity that is less th an the actual capac ity. Therefore, the turbine bypass capacity remains adequate for TPO operation because the actual capacity (unchanged) continues to bound the value used in the analyses. 7.4 FEEDWATER A ND CONDENSATE SYSTEMS The FW and condensate systems are designed to provide FW at the temperature, pressure, quality, and flow rate required by the reactor. These systems are not safety-related; however, their performance may have an effect on plant availability and the capability to operate reliably at the TPO uprate condition. A review of the Limerick FW heaters, heater drain system, condensate demineralizers, and the pumps (FW and condensate) demonstrated that the components are capable of performing in the proper design range to provide the slightly higher TPO uprate FW flow rate at the desired temperature and pressure. A review of the Limeri ck heater drain system demonstrated that the components are capable of supporting the slightly higher TPO uprate extraction flow rates. The No. 3 heater drain valves are currently undersized, but will be repl aced prior to TPO. Performance evaluations were based on an assessment of the capability of the condensate and FW systems and equipment to remain within the design limitations of the following parameters: Pump NPSH Ability to avoid suction pressure trip Flow capacity Bearing cooling capability Rated driver horsepower Vibration The FW system run-out and loss of FW heating events are expected to see very small changes from the TPO uprate as shown by the experience with substantially larger power uprates.
7.4   FEEDWATER AND CONDENSATE SYSTEMS The FW and condensate systems are designed to provide FW at the temperature, pressure, quality, and flow rate required by the reactor. These systems are not safety-related; however, their performance may have an effect on plant availability and the capability to operate reliably at the TPO uprate condition.
7.4.1 Normal Operation System operating flows for the TPO uprate increase approximately 2.3%. Operation at the TPO  
A review of the Limerick FW heaters, heater drain system, condensate demineralizers, and the pumps (FW and condensate) demonstrated that the components are capable of performing in the proper design range to provide the slightly higher TPO uprate FW flow rate at the desired temperature and pressure. A review of the Limerick heater drain system demonstrated that the components are capable of supporting the slightly higher TPO uprate extraction flow rates. The No. 3 heater drain valves are currently undersized, but will be replaced prior to TPO.
Performance evaluations were based on an assessment of the capability of the condensate and FW systems and equipment to remain within the design limitations of the following parameters:
* Pump NPSH
* Ability to avoid suction pressure trip
* Flow capacity
* Bearing cooling capability
* Rated driver horsepower
* Vibration The FW system run-out and loss of FW heating events are expected to see very small changes from the TPO uprate as shown by the experience with substantially larger power uprates.
7.4.1   Normal Operation System operating flows for the TPO uprate increase approximately 2.3%. Operation at the TPO RTP level does not significantly affect operating conditions of these systems. Discharge pressure of the condensate pumps decreases due to the pump head characteristics at increased flows. Discharge pressure of the FW pumps will increase to compensate for the increase in FW friction losses due to higher flow. To accomplish this function, opening the flow control valves to the feed pump turbine increases the feed pump speed. During steady-state conditions, the 7-2


RTP level does not significantly affect operating conditions of these systems. Discharge pressure of the condensate pumps decreases due to the pump head characteristics at increased flows. Discharge pressure of the FW pumps will increase to compensate for the increase in FW friction losses due to higher flow. To accomplish this function, opening the flow control valves to the feed pump turbine increases the feed pump speed. Du ring steady-state conditions, the NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 7-3 condensate and FW systems have available NPSH for all of the pumps to operate without cavitation at the TPO uprate conditions. Adequate margin during steady-state conditions exists between the calculated minimum pump suction pressure and the minimum pump suction pressure trip set points. The existing FW design pressure and temperature requirements bound operating conditions with adequate margin. The FW heaters are ASME Sec tion VIII pressure vessels. The heaters were analyzed and verified to be acceptable for the slightly higher FW heater temperatures and pressures for the TPO uprate.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION condensate and FW systems have available NPSH for all of the pumps to operate without cavitation at the TPO uprate conditions. Adequate margin during steady-state conditions exists between the calculated minimum pump suction pressure and the minimum pump suction pressure trip set points.
7.4.2 Transient Operation To account for FW demand transients, the condensate and FW systems were evaluated to ensure that sufficient margin above the TPO uprated flow is available. For system operation with all system pumps available, the predicted operating parameters were acceptable and within the component capabilities. Following a single FW pump trip with low reactor water level, the reactor recirculation system would runback recirculation flow, such that the steam production rate is within th e flow capacity of the remaining FW pumps. The runback setting prevents a reactor low water level scram, and is sufficient to maintain adequate margin to the potential P/F instability regions. Operation at the TPO condition does not degrade this capability.
The existing FW design pressure and temperature requirements bound operating conditions with adequate margin. The FW heaters are ASME Section VIII pressure vessels. The heaters were analyzed and verified to be acceptable for the slightly higher FW heater temperatures and pressures for the TPO uprate.
7.4.3 Condensate Filters and Condensate Deep Bed Demineralizers The effect of the TPO uprate on the condensate filter demineralizers (CFDs) and the condensate deep bed demineralizers (CDBDs) was reviewed. The CFD and CDBD systems can accommodate (without bypass) TPO uprate operations with one vessel removed from service (when backwash/resin change out is required).
7.4.2   Transient Operation To account for FW demand transients, the condensate and FW systems were evaluated to ensure that sufficient margin above the TPO uprated flow is available. For system operation with all system pumps available, the predicted operating parameters were acceptable and within the component capabilities.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8-1 8.0  RADWASTE AND RADIATION SOURCES 8.1 LIQUID AND S OLID W ASTE MANAGEMENT The liquid radwaste system collects, monitors , processes, stores, and returns processed radioactive waste to the plant for reuse, discharge, or shipment. Major sources of liquid and wet solid waste are from the condensate filters and deep bed condensate demineralizers. The TPO uprate results in a ~ 2% increased flow rate through the condensate system, potentially resu lting in a reduction in the average time between backwashes of the condensate filters and replacement of the deep bed condensate demineralizer resin. This potential reduction in the service time of the condensate filters and deep bed condensate demineralizers does not affect plant safety. The liquid collection subsystem and the solid collection subsystem both r eceive periodic inputs from a variety of sources. Neither subsystem experiences a significant increase in volume due to operation at the TPO uprate condition.
Following a single FW pump trip with low reactor water level, the reactor recirculation system would runback recirculation flow, such that the steam production rate is within the flow capacity of the remaining FW pumps. The runback setting prevents a reactor low water level scram, and is sufficient to maintain adequate margin to the potential P/F instability regions. Operation at the TPO condition does not degrade this capability.
The activated corrosion products in the waste stream are expected to in crease proportionally to the TPO uprate. However, the total volume of processed waste is not expected to increase appreciably because the only significant increase in processed waste is due to the more frequent backwashes of the condensate filters and deep bed condensate demine ralizers. A review of plant operating effluent reports and the slight increase expected from the TPO uprate, leads to the conclusion that the requirements of 10 CFR 20 and 10 CFR 50, Appendix I will continue to be met. Therefore, the TPO uprate does not adversely affect the processing of liquid radwaste and there are no significant environmental effects. 8.2 G ASEOUS W ASTE M ANAGEMENT The gaseous waste systems collect , control, process, and dispos e of gaseous radioactive waste generated during normal operation and abnormal ope rational occurrences. The gaseous waste management systems include the offgas system and various building ventilation systems. The systems are designed to meet the requirem ents of 10 CFR 20 and 10 CFR 50, Appendix I. Non-condensable radioactive gas from the main condenser normally contains activation gases and fission product radioactive noble gas parents. These are the major sources of radioactive gas, and are greater than all other sources co mbined. These non-condensable gases, along with non-radioactive air in leakage, are continuously removed from the main condensers by the SJAEs that discharge into the offgas system. Building ventilation systems control airborne radioactive gases by using devices such as High Efficiency Particulate Air (HEPA) and charcoal filters, and radiation monitors that activate isolation dampers or trip supply and exhaust fans, or by maintaining negative or positive air pressure to limit migration of gases. The changes to the gaseous radwaste releases are NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8-2 proportional to the change in core power and the to tal releases are a small fraction of the design basis releases. The release limit is an administratively controlled variable and is not a function of core power. The gaseous effluents are well within limits at CLTP operation and remain well within limits following implementation of the TPO uprate. Th ere are no significant environmental effects due to the TPO uprate.
7.4.3   Condensate Filters and Condensate Deep Bed Demineralizers The effect of the TPO uprate on the condensate filter demineralizers (CFDs) and the condensate deep bed demineralizers (CDBDs) was reviewed. The CFD and CDBD systems can accommodate (without bypass) TPO uprate operations with one vessel removed from service (when backwash/resin change out is required).
The off gas system was evaluated for the TPO uprat
7-3
: e. Radiolysis of water in the core region, which forms H 2 and O 2, increases linearly with core power, thus increasing the heat load on the recombiner and related components. The Offgas system design basis H 2 is 138.4 scfm (with a corresponding stoichiometric O 2 of 69.7 scfm). The expected H 2 flow rate for the TPO uprate is 96.4 scfm (48.2 scfm of O 2). The increase in H 2 and O 2 due to the TPO uprate remains well with the capacity of the system. Therefore, the TPO uprate does not affect the offgas system design or operation. 8.3 RADIATION SOURCES IN THE REACTOR CORE  TLTR Appendix H describes the methodology and assumptions for the evaluation of radiological effects for the TPO uprate.


During power operation, the radiation sources in the core are directly related to the fission rate. These sources include radiation from the fission process, accumulated fission products and neutron reactions as a secondary result of fission.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8.0 RADWASTE AND RADIATION SOURCES 8.1    LIQUID AND SOLID WASTE MANAGEMENT The liquid radwaste system collects, monitors, processes, stores, and returns processed radioactive waste to the plant for reuse, discharge, or shipment.
Historically, these sources have been defined in terms of energy released per unit of reactor power. Therefore, for TPO, the percent increase in the operating source terms is no greater than the percent increase in power. The source term increases due to the TPO uprate are bounded by the safety margins of the design basis sources. The post-operation radiation sources in the core are primarily the result of accumulated fission products. Two separate forms of post-operation source data are normally applied. The first is the core gamma-ray source, which is used in shielding calculations for the core and for individual fuel bundles. This source term is defined in terms of MeV/sec per watt of reactor thermal power (or equivalent) at various times after shutdown. Therefore, the total gamma energy source increases in proportion to reactor power.
Major sources of liquid and wet solid waste are from the condensate filters and deep bed condensate demineralizers. The TPO uprate results in a ~ 2% increased flow rate through the condensate system, potentially resulting in a reduction in the average time between backwashes of the condensate filters and replacement of the deep bed condensate demineralizer resin. This potential reduction in the service time of the condensate filters and deep bed condensate demineralizers does not affect plant safety.
The second set of post-operation source data consists primarily of nuclide activity inventories for fission products in the fuel. These are needed for post-accident and spent fuel pool evaluations, which are performed in compliance with regulatory guidance that applies different release and transport assumptions to different fission products. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which de velops "equilibrium" activities in the fuel (typically three years). Most radiological ly significant fission products reach equilibrium within a 60-day period. The calculated inventories are approximately proportional to core thermal power. Conseque ntly, for TPO, the inventories of those radionuclides, which reached or approached equilibrium, are expected to increase in proportion NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8-3 to the thermal power increase. The inventories of the very long-lived radionuclides, which did not approach equilibrium, ar e both power and exposure depende nt. They are expected to increase proportionally with power if the fuel irradiation time remains within the current basis.
The liquid collection subsystem and the solid collection subsystem both receive periodic inputs from a variety of sources. Neither subsystem experiences a significant increase in volume due to operation at the TPO uprate condition.
Thus, the long-lived radionuclides are expected to increase pr oportionally to power. The radionuclide inventories are provided in terms of curies per megawatt of reactor thermal power at various times after shutdown.
The activated corrosion products in the waste stream are expected to increase proportionally to the TPO uprate. However, the total volume of processed waste is not expected to increase appreciably because the only significant increase in processed waste is due to the more frequent backwashes of the condensate filters and deep bed condensate demineralizers. A review of plant operating effluent reports and the slight increase expected from the TPO uprate, leads to the conclusion that the requirements of 10 CFR 20 and 10 CFR 50, Appendix I will continue to be met. Therefore, the TPO uprate does not adversely affect the processing of liquid radwaste and there are no significant environmental effects.
[[                                                                                                                                                                               
8.2    GASEOUS WASTE MANAGEMENT The gaseous waste systems collect, control, process, and dispose of gaseous radioactive waste generated during normal operation and abnormal operational occurrences. The gaseous waste management systems include the offgas system and various building ventilation systems. The systems are designed to meet the requirements of 10 CFR 20 and 10 CFR 50, Appendix I.
Non-condensable radioactive gas from the main condenser normally contains activation gases and fission product radioactive noble gas parents. These are the major sources of radioactive gas, and are greater than all other sources combined. These non-condensable gases, along with non-radioactive air in leakage, are continuously removed from the main condensers by the SJAEs that discharge into the offgas system.
Building ventilation systems control airborne radioactive gases by using devices such as High Efficiency Particulate Air (HEPA) and charcoal filters, and radiation monitors that activate isolation dampers or trip supply and exhaust fans, or by maintaining negative or positive air pressure to limit migration of gases. The changes to the gaseous radwaste releases are 8-1


                                                                  ]] 8.4 RADIATION SOURCES IN REACTOR COOLANT 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 is the dominant source in the turbine building and in the lower regions of the drywell. Because these sources are produced by interactions in the core region, their ra tes of production are proportional to power. However, the concentration in the steam remains nearly constant, because the increase in activation production is balanced by the increase in steam flow. As a result, the activation produc ts, observed in th e reactor water and steam, increase in approximate proportion to the increase in thermal power.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION proportional to the change in core power and the total releases are a small fraction of the design basis releases.
8.4.2 Activated Corrosion Products The reactor coolant contains activated corrosion products from metallic materials entering the water and being activated in th e reactor region. Under the TPO uprate conditions, the FW flow increases with power, the activation rate in the reactor region increases with power, and the filter efficiency of the condensate demineralizers may decrease as a result of the FW flow increase. The net result may be an increase in the activ ated corrosion product production. However, the NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8-4 TPO uprate corrosion product concentrations are not expected to exceed the design basis concentrations. TPO activated corrosion product activity levels in the reactor water remain less than one-half of the design basi s activated corrosion product activity. Therefor e, no change is required in the design basis activated corrosi on product concentrations for the TPO uprate.
The release limit is an administratively controlled variable and is not a function of core power.
8.4.3 Fission Products 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. The noble gases released during plant operation result from the escape of minute fractions of the fission products from the fuel rods. Noble gas release rates increase approximately w ith power level. This activity is the noble gas offgas that is included in the Limerick design. The offgas rate s for TPO uprate operations are well below the original design ba sis. Therefore, the design ba sis release rates are bounding for the TPO uprate. The fission product activity in the reactor water, like the activity in the steam, is the result of minute releases from the fuel rods. As is the case for the noble gases, there is no expectation that releases from the fuel increase due to the TPO uprate. Activity levels in the reactor water are expected to be approximately equal to current measured data, which are fractions of the design basis values. Therefore, the design basis values are unchanged. 8.5 RADIATION LEVELS Normal operation radiation levels increase slightly for the TPO uprate. Limerick was designed with substantial conservatism for higher-than-expect ed 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 design, source terms, and analytical techniques.
The gaseous effluents are well within limits at CLTP operation and remain well within limits following implementation of the TPO uprate. There are no significant environmental effects due to the TPO uprate.
Post-operation radiation levels in most areas of the plant increase by no more than the percentage increase in power level. In a few areas near the SFP cooling system piping and the reactor water piping, where accumulation of corrosion product crud is expected, as well as near some liquid radwaste equipment, the increase could be slightly higher. The radiation levels in areas with significant N-16 radiation ar e expected to increase by slightly more than the percentage increase in power level. Regardless, individual worker exposures will be mainta ined within acceptable limits by the site As Low As Reasonably Achievable (ALARA) program, which controls access to radiation areas.
The off gas system was evaluated for the TPO uprate. Radiolysis of water in the core region, which forms H2 and O2, increases linearly with core power, thus increasing the heat load on the recombiner and related components. The Offgas system design basis H2 is 138.4 scfm (with a corresponding stoichiometric O2 of 69.7 scfm). The expected H2 flow rate for the TPO uprate is 96.4 scfm (48.2 scfm of O2). The increase in H2 and O2 due to the TPO uprate remains well with the capacity of the system. Therefore, the TPO uprate does not affect the offgas system design or operation.
8.3    RADIATION SOURCES IN THE REACTOR CORE TLTR Appendix H describes the methodology and assumptions for the evaluation of radiological effects for the TPO uprate.
During power operation, the radiation sources in the core are directly related to the fission rate.
These sources include radiation from the fission process, accumulated fission products and neutron reactions as a secondary result of fission. Historically, these sources have been defined in terms of energy released per unit of reactor power. Therefore, for TPO, the percent increase in the operating source terms is no greater than the percent increase in power. The source term increases due to the TPO uprate are bounded by the safety margins of the design basis sources.
The post-operation radiation sources in the core are primarily the result of accumulated fission products. Two separate forms of post-operation source data are normally applied. The first is the core gamma-ray source, which is used in shielding calculations for the core and for individual fuel bundles. This source term is defined in terms of MeV/sec per watt of reactor thermal power (or equivalent) at various times after shutdown. Therefore, the total gamma energy source increases in proportion to reactor power.
The second set of post-operation source data consists primarily of nuclide activity inventories for fission products in the fuel. These are needed for post-accident and spent fuel pool evaluations, which are performed in compliance with regulatory guidance that applies different release and transport assumptions to different fission products. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops equilibrium activities in the fuel (typically three years). Most radiologically significant fission products reach equilibrium within a 60-day period. The calculated inventories are approximately proportional to core thermal power. Consequently, for TPO, the inventories of those radionuclides, which reached or approached equilibrium, are expected to increase in proportion 8-2
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION to the thermal power increase. The inventories of the very long-lived radionuclides, which did not approach equilibrium, are both power and exposure dependent. They are expected to increase proportionally with power if the fuel irradiation time remains within the current basis.
Thus, the long-lived radionuclides are expected to increase proportionally to power. The radionuclide inventories are provided in terms of curies per megawatt of reactor thermal power at various times after shutdown.
((
                                    ))
8.4   RADIATION SOURCES IN REACTOR COOLANT 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 is the dominant source in the turbine building and in the lower regions of the drywell. Because these sources are produced by interactions in the core region, their rates of production are proportional to power. However, the concentration in the steam remains nearly constant, because the increase in activation production is balanced by the increase in steam flow. As a result, the activation products, observed in the reactor water and steam, increase in approximate proportion to the increase in thermal power.
8.4.2   Activated Corrosion Products The reactor coolant contains activated corrosion products from metallic materials entering the water and being activated in the reactor region. Under the TPO uprate conditions, the FW flow increases with power, the activation rate in the reactor region increases with power, and the filter efficiency of the condensate demineralizers may decrease as a result of the FW flow increase.
The net result may be an increase in the activated corrosion product production. However, the 8-3
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION TPO uprate corrosion product concentrations are not expected to exceed the design basis concentrations. TPO activated corrosion product activity levels in the reactor water remain less than one-half of the design basis activated corrosion product activity. Therefore, no change is required in the design basis activated corrosion product concentrations for the TPO uprate.
8.4.3   Fission Products 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. The noble gases released during plant operation result from the escape of minute fractions of the fission products from the fuel rods.
Noble gas release rates increase approximately with power level. This activity is the noble gas offgas that is included in the Limerick design. The offgas rates for TPO uprate operations are well below the original design basis. Therefore, the design basis release rates are bounding for the TPO uprate.
The fission product activity in the reactor water, like the activity in the steam, is the result of minute releases from the fuel rods. As is the case for the noble gases, there is no expectation that releases from the fuel increase due to the TPO uprate. Activity levels in the reactor water are expected to be approximately equal to current measured data, which are fractions of the design basis values. Therefore, the design basis values are unchanged.
8.5   RADIATION LEVELS Normal operation radiation levels increase slightly for the TPO uprate. Limerick was designed with substantial conservatism 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 design, source terms, and analytical techniques.
Post-operation radiation levels in most areas of the plant increase by no more than the percentage increase in power level. In a few areas near the SFP cooling system piping and the reactor water piping, where accumulation of corrosion product crud is expected, as well as near some liquid radwaste equipment, the increase could be slightly higher. The radiation levels in areas with significant N-16 radiation are expected to increase by slightly more than the percentage increase in power level.
Regardless, individual worker exposures will be maintained within acceptable limits by the site As Low As Reasonably Achievable (ALARA) program, which controls access to radiation areas.
Procedural controls compensate for increased radiation levels.
Procedural controls compensate for increased radiation levels.
The change in core activity inventory resulting from the TPO uprate (S ection 8.3) increases post-accident radiation levels by no more than approximately the percentage increase in power level. The slight increase in the post-accident radiation levels has no significant effect on the plant or the habitability of the on-site Emergency Respons e facilities. A review of areas requiring post-NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8-5 accident occupancy concluded that access needed for accident mitigation is not significantly affected by the TPO uprate.
The change in core activity inventory resulting from the TPO uprate (Section 8.3) increases post-accident radiation levels by no more than approximately the percentage increase in power level.
Section 9.2 addresses the Main Control Room doses for the worst-case accident. 8.6 N ORMAL OPERATION O FF-S ITE D OSES  The Technical Specification limits implement the guidelines of 10 CFR 50, Appendix I. A review of the normal radiologi cal effluent doses shows that at CLTP, the annual doses are a small fraction of the doses allowed by Technical Specification limits. The TPO uprate does not involve significant increases in the offsite dose from noble gases, airborne particulates, iodine, tritium or liquid effluents. In addition, radiation from shine is not a significant exposure pathway. Present offsite radiation levels are a negligible portion of background radiation. Therefore, the normal offsite doses are not significantly affected by operation at the TPO RTP level and remain below the limits of 10 CFR 20 and 10 CFR 50, Appendix I.  
The slight increase in the post-accident radiation levels has no significant effect on the plant or the habitability of the on-site Emergency Response facilities. A review of areas requiring post-8-4
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION accident occupancy concluded that access needed for accident mitigation is not significantly affected by the TPO uprate.
Section 9.2 addresses the Main Control Room doses for the worst-case accident.
8.6   NORMAL OPERATION OFF-SITE DOSES The Technical Specification limits implement the guidelines of 10 CFR 50, Appendix I. A review of the normal radiological effluent doses shows that at CLTP, the annual doses are a small fraction of the doses allowed by Technical Specification limits. The TPO uprate does not involve significant increases in the offsite dose from noble gases, airborne particulates, iodine, tritium or liquid effluents. In addition, radiation from shine is not a significant exposure pathway. Present offsite radiation levels are a negligible portion of background radiation.
Therefore, the normal offsite doses are not significantly affected by operation at the TPO RTP level and remain below the limits of 10 CFR 20 and 10 CFR 50, Appendix I.
8-5
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9.0 REACTOR SAFETY PERFORMANCE EVALUATIONS 9.1    ANTICIPATED OPERATIONAL OCCURRENCES TLTR Appendix E provides a generic evaluation of the AOOs for TPO uprate plants. ((
                              )) Also included are the analytical methods to be used and operating conditions to be assumed. The AOO events are organized into two major groups: Fuel Thermal Margin Events, and Transient Overpressure Events.
TLTR Table E-2 illustrates the effect of a 1.5% power uprate on the OLMCPR. ((
            )) The OLMCPR changes for the 1.7% uprate may be slightly larger than shown in Table E- 2, but the changes are expected to be within the normal cycle-to-cycle variation. The overpressure events and loss of FW transient are currently performed with the assumption of 2%
overpower. Therefore, they are applicable and bounding for the TPO uprate.
The reload transient analysis includes the worst overpressure event, which is usually the closure of all MSIVs with high neutron flux scram.
The evaluations and conclusions of TLTR Appendix E are applicable to the Limerick TPO uprate. Therefore, it is sufficient for the plant to perform the standard reload analyses at the first fuel cycle that implement the TPO uprate.
9.2    DESIGN BASIS ACCIDENTS The radiological consequences of a DBA are basically proportional to the quantity of radioactivity released to the environment. This quantity is a function of the fission products released from the core as well as the transport mechanisms from the core to the release point.
The radiological releases at the TPO uprate power are generally expected to increase in proportion to the core inventory increase, which is in proportion to the power increase.
Postulated DBA events have been evaluated and analyzed to show that NRC regulations are met for 2% above the CLTP. DBA events have either been previously analyzed at 102% of CLTP or are not dependent on core thermal power. The Main Steam Line Break Accident outside containment was evaluated using a 4 &#xb5;Ci/g dose equivalent I-131 limit on reactor coolant activity. The limit on reactor coolant activity is unchanged for the TPO uprate condition. The evaluation/analysis was based on the methodology, assumptions, and analytical techniques 9-1


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-1 9.0  REACTOR SAFETY PERF ORMANCE EVALUATIONS 9.1 ANTICIPATED OPERATIONAL OCCURRENCES TLTR Appendix E provides a generic evaluati on of the AOOs for TP O uprate plants. [[               
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION described in the Regulatory Guides, the Standard Review Plan (SRP) (where applicable), and in previous SEs.
9.3  SPECIAL EVENTS 9.3.1   Anticipated Transient Without Scram TLTR Section 5.3.5 and TLTR Appendix L, present a generic evaluation of the sensitivity of an ATWS to a change in power typical of the TPO uprate. The evaluation is based on previous analyses for power uprate projects. For a TPO uprate, if a plant has sufficient margin for the projected changes in peak parameters given in TLTR Section L.3.5, ((
                                                                                            ))
The previous ATWS analysis, performed at 100% of CLTP, did not demonstrate the required margins for generic evaluation to the peak vessel bottom head pressure limit and to the pool temperature limit. ((
                            ))
NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate, Class III, July 2003 (also referred to as the CLTR) was approved by the NRC as an acceptable method for evaluating the effects of Constant Pressure Power Uprates (CPPUs). Section 9.3.1 of the CLTR addresses the effect of CPPU on ATWS. The CLTR methodology was used to analyze and evaluate the Limerick ATWS event.
((                      )) ATWS analysis is required for TPO RTP to ensure that the following ATWS acceptance criteria are met:
* Maintain reactor vessel integrity (i.e., peak vessel bottom pressure less than the ASME Service Level C limit of 1500 psig).
* Maintain containment integrity (i.e., maximum containment pressure and temperature less than the design pressure (55 psig) and temperature (190&deg;F) of the containment structure).
* Maintain coolable core geometry.
9-2
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION The TPO RTP ATWS analysis is performed using the NRC approved code ODYN (Table 1-1).
The key inputs to the ATWS analysis are provided in Table 9-1. The results of the analysis are provided in Table 9-2.
The results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the Limerick response to an ATWS event at TPO is acceptable. The potential for thermal-hydraulic instability in conjunction with ATWS events is evaluated in Section 9.3.1.4.
Limerick also meets the ATWS mitigation requirements defined in 10 CFR 50.62:
* Installation of an ARI system;
* Boron injection equivalent to 86 gpm; and
* Installation of automatic Recirculation Pump Trip (RPT) logic (i.e., ATWS-RPT).
There are no changes to the assumed operator actions for the TPO RTP ATWS analysis.
When required by changes in plant configuration (as identified by the design change process),
changes to Emergency Operating Procedures (EOPs), including changes to EOP calculations and plant data, are developed and implemented in accordance with plant administrative procedure for EOP program maintenance.
Limerick performs EOP calculations consistent with the BWR Owners Group Emergency Procedure Guidelines (EPGs) / Severe Accident Guidelines (SAGs) Appendix C. Critical software is verified and validated by Design Engineering to generate EOP results. The EOP calculation input and output data is reviewed and verified by Design Engineering. Changes to the EOP calculation outputs are forwarded to Operations for use in revising the EOP Procedures/Flow Charts and the SAGs and supporting documents. Finally, the EOP flow charts are verified and validated by Operations, including trial use in the simulator.
The ATWS mitigation strategy is based on the BWROG EPGs, which are incorporated in the existing Limerick EOPs. TPO implementation does not significantly change the transient sequence of events. Therefore, there is no change in operator strategy on ATWS level reduction or early boron injection. TPO may affect some of the calculated curves, but does not affect stability mitigation actions. The changes due to TPO do not require modification of operator instructions.
Limerick meets all CLTR dispositions and the results in this evaluation are described below.
The topics addressed in this evaluation are:
9-3


                                                    ]]  Also included are the analytical methods to be used and operating conditions to be assumed. The AOO events are organized into two major groups: Fuel Thermal Margin Events, and Transient Overpressure Events.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Topic                      CLTR Disposition        Limerick Result Meets CLTR ATWS (Overpressure) - Event Selection              ((
TLTR Table E-2 illustrates the effect of a 1.5% power uprate on the OLMCPR. [[                         
Disposition Meets CLTR ATWS (Overpressure) - Limiting Events Disposition ATWS (Suppression Pool Temperature) -                                    Meets CLTR Event Selection                                                          Disposition ATWS (Suppression Pool Temperature) -                                    Meets CLTR Limiting Events                                                          Disposition Meets CLTR ATWS (Peak Cladding Temperature)                          ))
Disposition 9.3.1.1 ATWS (Overpressure)
As stated in Section 9.3.1 of the CLTR, the higher operating steam flow may result in higher peak vessel pressures. The higher power and decay heat will result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could affect the capability of the SLCS to mitigate the consequences of an ATWS event.
The overpressure evaluation includes consideration of the most limiting RPV overpressure case.
TLTR Appendix L considers four ATWS events: ((
)) The ATWS (Overpressure) - Event Selection meets all CLTR dispositions.
As shown in Section 3.7 of ELTR2, ((
                                                )) The MSIVC, PRFO and LOOP cases were performed for Limerick. The analysis results are given in Table 9-2. The MSIVC, PRFO and LOOP sequence of events are given in Tables 9-6 through 9-8. The short-term and long-term transient response to the MSIVC, PRFO and LOOP ATWS events are presented in Figures 9-1 through 9-24. Therefore, ATWS (Overpressure) - Limiting Events meet all CLTR dispositions.
9.3.1.2 ATWS (Suppression Pool Temperature)
As stated in Section 9.3.1 of the CLTR, the higher operating steam flow will result in higher peak vessel pressures. The higher power and decay heat may result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could impact the capability of the SLCS to mitigate the consequences of an ATWS event.
The suppression pool temperature evaluation includes consideration of the most limiting RHR pool cooling capability case. TLTR Appendix L considered four ATWS events: ((
9-4


                    ]]  The OLMCPR changes for the 1.7% uprate may be slightly larger than shown in Table E- 2, but the changes are expected to be within the normal cycle-to-cycle variation. The overpressure events and loss of FW transient are currently performed with the assumption of 2%
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION
overpower. Therefore, they are appl icable and bounding for the TPO uprate. The reload transient analysis includes the worst overpressure event, which is usually the closure of all MSIVs with high neutron flux scram. The evaluations and conclusions of TLTR Appe ndix E are applicable to the Limerick TPO uprate. Therefore, it is sufficient for the plant to perform the standard reload analyses at the first fuel cycle that implement the TPO uprate. 9.2 DESIGN BASIS ACCIDENTS  The radiological consequences of a DBA ar e basically proportional to the quantity of radioactivity released to the environment. This quantity is a function of the fission products released from the core as well as the transport mechanisms from the core to the release point. The radiological releases at the TPO uprate power are generally expected to increase in proportion to the core inventory increase, whic h is in proportion to the power increase. Postulated DBA events have been evaluated and analyzed to show that NRC regulations are met for 2% above the CLTP. DBA events have either been previously analyzed at 102% of CLTP or are not dependent on core thermal power. The Main Steam Line Break Accident outside containment was evaluated using a 4 &#xb5;Ci/g dose equivalent I-131 limit on reactor coolant activity. The limit on reactor coolant activity is unchanged for the TPO uprate condition. The evaluation/analysis was based on the methodol ogy, assumptions, and anal ytical techniques NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-2 described in the Regulatory Guides , the Standard Review Plan (SRP) (where applicable), and in previous SEs. 9.3 S PECIAL EVENTS 9.3.1 Anticipated Transient Without Scram TLTR Section 5.3.5 and TLTR Appendix L, present a generic evaluation of the sensitivity of an ATWS to a change in power typical of the TP O uprate. The evaluation is based on previous analyses for power uprate projects. For a TPO uprate, if a plant has sufficient margin for the projected changes in peak parameters given in TLTR Section L.3.5, [[                                                 
                                                                                  )) The ATWS (Suppression Pool Temperature) - Event Selection meets all CLTR dispositions.
The MSIVC, PRFO and LOOP cases were performed for Limerick. The key inputs to the ATWS analysis are provided in Table 9-1. The ATWS analysis results are given in Table 9-2.
The MSIVC, PRFO and LOOP sequence of events are given in Tables 9-6 through 9-8. The ATWS (Suppression Pool Temperature) - Limiting Events meet all CLTR dispositions.
9.3.1.3 ATWS (Peak Cladding Temperature)
The TLTR in Appendix L.3 states that power uprate has a negligible effect on the PCT or local cladding oxidation. ((
                                                                                            ))
For ATWS events, the acceptance criteria for PCT and local cladding oxidation for emergency core cooling systems, defined in 10 CFR 50.46, are adopted to ensure an ATWS event does not impede core cooling.
For TPO, PCT and local cladding oxidation are not required to be explicitly analyzed per Appendix L.3 of TLTR. Therefore, ATWS (PCT) is in compliance with the acceptance criteria of 10 CFR 50.46; subsequently, coolable core geometry is assured by meeting the 2200&#xba;F PCT and the 17% local cladding oxidation acceptance criteria stated in 10 CFR 50.46.
9.3.1.4 ATWS with Core Instability The CLTR in Section 9.3.3 states that 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 TPO operation because the MELLLA upper boundary is not increased. The core design necessary to achieve TPO 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 References 20 and 21 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum flow point. ((
9-5


                                                                                                                                                                                ]]  The previous ATWS analysis, performed at 100
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION
% of CLTP, did not demonstrate the required margins for generic evaluation to the peak vessel bottom head pressure limit and to the pool temperature limit. [[                                                                                                                                           
                    ))
TPO allows plants to increase their operating thermal power but does not allow an increase in control rod line. ((
                                                  ))
((
                                                                                ))
Initial operating conditions of FWHOOS and FFWTR do not significantly affect the ATWS instability response reported in References 20 and 21. The limiting ATWS evaluation assumes that all FW heating is lost during the event and the injected FW temperature approaches the lowest achievable main condenser hot well temperature. ((
            ))
((
                                      )) Therefore, the TPO effect on ATWS with core instability at Limerick meets all CLTR dispositions.
9.3.1.5 SLCS System Performance and Hardware Based on the results of the ((                    )) ATWS analysis, the maximum reactor upper plenum pressure following the limiting ATWS event reaches 1224 psig (1239 psia) during the time the SLCS is analyzed to be in operation. Consequently, there is a corresponding increase in the maximum two pump discharge pressure to 1336 psig and 1330 psig for Limerick Units 1 and 2 respectively and a decrease in the operating pressure margin for the pump discharge relief valves. The relief valve margin is not satisfied with three SLCS pump operation; however, the relief valve margin is satisfied with two SLCS pump operation. This is acceptable because they meet the injection requirements of 10 CFR 50.62 and the injection requirements of Information Notice 2001-13 with two pumps. The key SLCS input parameters are summarized in Table 9-3 and Table 9-4 for three and two SLCS pump operation, respectively. The key SLCS 9-6


                                                      ]] NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate,"
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION performance results are presented in Table 9-5. Consideration was also given to system flow, head losses for full injection, and cyclic pressure pulsations due to the positive displacement pump operation in determining the pressure margin to the opening set point for the pump discharge relief valves. The relief valve setpoint margin for two SLCS pump operation is 19.7 psi and 25.8 psi for Limerick Units 1 and 2, respectively. This margin is based on a SLCS pump relief valve setpoint of 1400 psig. The pump discharge relief valves are periodically tested to maintain this tolerance. Therefore, the current SLCS process parameters associated with the minimum boron injection rate are not changed.
Class III, July 2003 (also referred to as the "CLTR") was approved by the NRC as an acceptable me thod for evaluating the effects of Constant Pressure Power Uprates (CPPUs). Section 9.3.1 of the CLTR addresses the effect of CPPU on ATWS. The CLTR methodology was used to analyze and evaluate the Limerick ATWS event.  
The SLCS ATWS performance is evaluated for a representative core design for TPO. The evaluation shows that TPO has no adverse effect on the ability of the SLCS to mitigate an ATWS in two SLCS pump operation. There are no timer setting changes for TPO for Limerick, and the ATWS analysis confirms acceptable results. Therefore, the system performance and hardware meets all CLTR dispositions.
[[                                      ]] ATWS analysis is required for TPO RTP to ensure that the following ATWS acceptance criteria are met:  Maintain reactor vessel integrity (i.e., peak vessel bottom pressure less than the ASME Service Level C limit of 1500 psig). Maintain containment integrity (i.e., maximum containment pressure and temperature less than the design pressure (55 psig) and temperature (190F) of the containment structure). Maintain coolable core geometry.
9.3.1.6 Suppression Pool Temperature following ATWS Event As stated in Section 6.5 of the CLTR, changes in the fuel design for TPO may require modifications to the SLCS as a result of the increase in the suppression pool temperature for the limiting ATWS event.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-3 The TPO RTP ATWS analysis is performed using the NRC approved code ODYN (Table 1-1). The key inputs to the ATWS analysis are provided in Table 9-1. The results of the analysis are provided in Table 9-2. The results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the Limerick response to an ATWS event at TPO is acceptable. The potential for thermal-hydraulic instability in conjunction with ATWS events is evaluated in Section 9.3.1.4. Limerick also meets the ATWS mitigation requirements defined in 10 CFR 50.62: Installation of an ARI system; Boron injection equivalent to 86 gpm; and Installation of automatic Recirculation Pump Trip (RPT) logic (i.e., ATWS-RPT). There are no changes to the assumed operator actions for the TPO RTP ATWS analysis.
The boron injection rate requirement for maintaining the peak suppression pool water temperature limits, following the limiting ATWS event with SLCS injection, is not increased for TPO. Therefore, the Suppression Pool temperature following an ATWS event meets all CLTR dispositions.
When required by changes in plan t configuration (as identified by the design change process), changes to Emergency Operating Procedures (EOP s), including changes to EOP calculations and plant data, are developed and implemented in accordance with plant administrative procedure for EOP program maintenance.
9.3.1.7 Equipment Out-of-Service and Flexibility Options MELLLA, ICF and SLO: The TPO ATWS analyses were performed along the MELLLA boundary. The TPO ATWS analysis at MELLLA conditions bounds operation at ICF and in SLO. Therefore, TPO continues to support these performance improvement features.
Limerick performs EOP calculations consistent with the BWR Owners Group Emergency Procedure Guidelines (EPGs) / Severe Accident Guidelines (S AGs) Appendix C. Critical software is verified and valid ated by Design Engineering to generate EOP results. The EOP calculation input and output data is reviewed and verified by Design Engineering. Changes to the EOP calculation outputs are forwarded to Operations for use in revising the EOP Procedures/Flow Charts and the SAGs and supporting documents. Finally, the EOP flow charts are verified and validated by Operations, including trial use in the simulator. The ATWS mitigation strategy is based on the BWROG EPGs, which are incorporated in the existing Limerick EOPs. TPO implementation does not significantly change the transient sequence of events. Therefore, there is no change in operator strategy on ATWS level reduction or early boron injection. TPO may affect some of the calculated curves, but does not affect stability mitigation actions. The changes due to TPO do not require modification of operator instructions. Limerick meets all CLTR dispositions and the re sults in this evaluation are described below.
SRV OOS: The TPO ATWS analysis was performed with two SRVs OOS. Therefore, TPO continues to support this EOOS option.
The topics addressed in this evaluation are:
FWHOOS and FFWTR: FWHOOS and FFWTR are operational flexibility options that allow continued operation with reduced FW temperature. Initial power is unchanged for both the FWHOOS and FFWTR conditions. The additional reactivity associated with the reduced FW temperature is typically offset with control rods, as needed. This makes the core less reactive due to the lower void fraction. Thus, use of normal FW temperature is conservative for ATWS analyses.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-4 Topic CLTR Disposition Limerick Result ATWS (Overpressure) - Event Selection [[
MSIV OOS: The TPO ATWS analysis bounds the MSIV OOS condition for TPO. Limerick operation with MSIV OOS is limited to  75% rated power. With this restriction, the severity of the limiting ATWS events is reduced. The lower initial steaming rate reduces the peak vessel 9-7
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION pressure, peak power, PCT, and integrated SRV flow. The reduction in integrated SRV flow thereby reduces the peak suppression pool temperature and containment pressure.
ARTS, TBV OOS, RPT OOS, 24 Month Cycle, TCV Stuck Closed and TSV Stuck Closed: The TPO ATWS analysis is not impacted by these performance improvement features.
9.3.2  Station Blackout The Limerick Station Blackout (SBO) evaluation has previously been performed assuming 102% of CLTP. Therefore, the postulated SBO scenarios for TPO operation are bounded by the current evaluations.
9-8
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-1 Key Inputs for ATWS Analysis Input Variable                                    CLTP                    TPO RTP Reactor power (MWt)                                                      3458                    3517*
Reactor dome pressure (psia)                                            1060                      1060 Each SRV capacity at 1090 psig (Mlbm/hr)                                 0.870                    0.870 High pressure ATWS-RPT (psig)                                            1156                      1156 Number of SRVs OOS                                                        2                        2 Number of Auto Start SLCS Pumps                                            2                        2
* Performed at 101.7% of CLTP Table 9-2 Results for ATWS Analysis Acceptance Criteria                                CLTP 1, 2                TPO RTP1 Peak vessel bottom pressure (psig)                                    1458                      1473 Peak suppression pool temperature (&deg;F)                                181                        182 Peak containment pressure (psig)                                      10.3                      10.6 Peak cladding temperature (&deg;F)                                Generic Assessment          Generic Assessment Local cladding oxidation (%)                                  Generic Assessment          Generic Assessment Notes:
: 1. Cladding temperature and oxidation calculations are not required per Appendix L.3 of TLTR.
: 2. To maximize the effect of TPO, a baseline is established at the CLTP level, assuming the current licensed equipment performance assumptions and plant parameters.
Table 9-3 Inputs for Limerick Three SLCS Pump Operation ATWS Analysis Parameter                                      Limerick Unit 1              Limerick Unit 2 Reactor Water Elevation Head (psi)                                                2.5                            2.5 CS Nozzle Differential Pressure (psi)                                            20.8                          20.8 SLCS Pipe Elevation (psi)                                                        10.0                          10.0 Flow Friction Loss (psi)                                                          126.0                        112.1 Total Three Pump Operation System Losses (psi)                                    159.3                        145.4 9-9
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-4 Inputs for Limerick Two SLCS Pump Operation ATWS Analysis Parameter                          Limerick Unit 1 Limerick Unit 2 Reactor Water Elevation Head (psi)                                  2.5            2.5 CS Nozzle Differential Pressure (psi)                              20.3            20.3 SLCS Pipe Elevation (psi)                                          10.0            10.0 Flow Friction Loss (psi)                                            79.5            73.4 Total Two Pump Operation System Losses (psi)                        112.3          106.2 Table 9-5 Limerick SLCS Pressure Results for ATWS Analysis Parameter                          Limerick Unit 1 Limerick Unit 2 SLCS RV Setpoint (psig)                                            1400            1400 RV Tolerance (%)                                                    1.0            1.0 Pulsating Margin (psi)                                              30.0            30.0 SLCS Piping Flow Loss (psi) - 3 Pump Operation                      159.3          145.4 SLCS Piping Flow Loss (psi) - 2 Pump Operation                      112.3          106.2 Pressure at Point of SLCS Injection (psig)                          1224            1224 Pump Discharge Pressure (psig) - 3 Pump Operation                  1383.3          1369.4 Pump Discharge Pressure (psig) - 2 Pump Operation                  1336.3          1330.2 9-10
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-6 MSIVC Sequence of Events Item                      Event              TPO RTP BOC TPO RTP EOC Event Time Event Time (sec)      (sec) 1  MSIV Isolation Initiated                      0.0        0.0 2  MSIVs Fully Closed                            4.0        4.0 3  Peak Neutron Flux                              4.1        4.0 4  High Pressure ATWS Setpoint                    4.1        4.1 5  Recirculation Pumps Trip                      4.7        4.6 6  Start Opening of the First Relief Valve        4.9        4.8 7  Peak Heat Flux                                5.2        5.1 8  Peak Vessel Pressure                          9.5        9.1 9  Feedwater Reduction Initiated                30.0      30.0 10  BIIT Reached                                  53.0        54.0 11  SLCS Pumps Start                            124.1      124.1 12  Hot Shutdown Achieved 422        444 (Neutron Flux Remains < 0.1%)
13  RHR Cooling Initiated                          660        660 14  Peak Suppression Pool Temperature            8527      8322 9-11
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-7 PRFO Sequence of Events Item                      Event              TPO RTP BOC TPO RTP EOC Event Time Event Time (sec)     (sec) 1  TCV and Bypass Valves Start Open              0.1        0.1 MSIV Closure Initiated by Low Steamline 2                                                20.2      19.3 Pressure 3  MSIVs Fully Closed                            24.2        23.3 4  Peak Neutron Flux                            25.2        24.8 5  High Pressure ATWS Setpoint                  27.8        27.0 6  Recirculation Pumps Trip                      28.4      27.5 7  Start Opening of the First Relief Valve      28.6        27.7 8  Peak Heat Flux                                28.8        27.5 9  Peak Vessel Pressure                          35.2        33.9 10  Feedwater Reduction Initiated                53.4      53.4 11  BIIT Reached                                  69.0        70.0 12  SLCS Pumps Start                            147.8      147.0 Hot Shutdown Achieved 13                                                451        477 (Neutron Flux Remains < 0.1%)
14  RHR Cooling Initiated                          660        660 15  Peak Suppression Pool Temperature            7871      7131 9-12
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-8 LOOP Sequence of Events Item                      Event              TPO RTP BOC TPO RTP EOC Event Time Event Time (sec)      (sec) 1  Loss of Auxiliary Power                        0.0        0.0 2  Recirculation Pumps Trip                      0.0        0.0 3  Feedwater Reduction Initiated                  0.0        0.0 4  TCV Closure                                    0.0        0.0 5  Peak Neutron Flux                              0.7        0.7 6  High Pressure ATWS Setpoint                    1.0        0.9 7  Start Opening of the First Relief Valve        2.0        1.9 8  MSIV Closure Initiated                        2.0        2.0 9  Peak Heat Flux                                2.3        2.2 10  MSIVs Fully Closed                            6.0        6.0 11  Peak Vessel Pressure                          8.3        7.8 12  BIIT Reached                                  54.0        55.0 13  SLCS Pumps Start                              121.0      120.9 Hot Shutdown Achieved 14                                                419        454 (Neutron Flux Remains < 0.1%)
15  RHR Cooling Initiated                          660        660 16  Peak Suppression Pool Temperature            ~29000    ~29000 9-13
 
NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-1:
Figure 9-1 : TPO TPO RTP RTP MELLLA MELLLA BOC      MSIVC (Short DOC MSIVC  (Short Term)
Term)
((
                                                          ))
9-14
 
NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-2:
Figure 9-2: TPO TPO RTP RTP MELLLA MELLLA HOC BOC MSIVC MSIVC (Long (Long Term Term -- A)
A)
((
                                                              ))
9-15
 
NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-3:
Figure      TPO RTP 9-3: TPO RTP MELLLA MELLLA BOC BOC MSIVC MSIVC (Long (Long Term Term -- B)
B)
((
                                                              ))
9-16
 
NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY INFORMATION NON-PROPRIETARY   INFORMATION Figure Figure 9-4 9-4:: TPO TPO RTP RTP MELLLA MELLLA BOC BOC MSIVC MSIVC (Long (Long Term Term -- C)
C)
((
                                                                ))
9-17
 
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure Figure 9-5 9-5:: TPO TPO RTP RTP MELLLA MELLLA BOCDOC PRFO PRFO (Short (Short Term)
Term)
((
                                                            ))
9-18
 
NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-6:
Figure 9-6: TPO TPO RTP RTP MELLLA MELLLA BOC      PRFO (Long DOC PRFO  (Long Term Term -- A)
A)
((
                                                              ))
9-19
 
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-7:
Figure 9-7 : TPO TPO RTP RTP MELLLA MELLLA BOC BOC PRFO  (Long Term PRFO (Long Term -- B)
B)
((
                                                              ))
9-20
 
NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-8:
Figure 9-S : TPO TPO RTP RTP MELLLA MELLLA BOC BOC PRFO PRFO (Long (Long Term Term -- C)
C)
((
                                                              ))
9-21
 
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-9: TPO Figure 9-9: TPO RTP RTP MELLLA MELLLA BOCBOC LOOP  (Short Term)
LOOP (Short  Term)
((
                                                          ))
9-22
 
NEDO-33484 REVISION NEDO-33484  REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY    INFORMATION Figure 9-10:
Figure 9-10: TPO TPO RTP RTP MELLLA MELLLA DOC BOC LOOP LOOP (Long (Long Term Term -- A)
A)
((
                                                                ))
9-23


Meets CLTR Disposition ATWS (Overpressure) - Limiting Events Meets CLTR Disposition ATWS (Suppression Pool Temperature) - Event Selection
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure Figure 9-11 9-11:: TPO TPO RTP RTP MELLLA MELLLA BOCBOC LOOP LOOP (Long (Long Term Term -- B)
B)
((
                                                                ))
9-24


Meets CLTR Disposition ATWS (Suppression Pool Temperature) -
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure Figure 9-12: TPO RTP 9-12: TPO RTP MELLLA MELLLA BOCBOC LOOP  (Long Term LOOP (Long  Term -- C)
Limiting Events Meets CLTR Disposition ATWS (Peak Cladding Temperature)  
C)
                    ]] Meets CLTR Disposition 9.3.1.1 ATWS (Overpressure)
((
As stated in Section 9.3.1 of the CLTR, the higher operating steam flow may result in higher peak vessel pressures. The higher power and decay heat will result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could affect the capability of the SLCS to mitigate the consequences of an ATWS event.
                                                              ))
The overpressure evaluation includes consideration of the most limiting RPV overpressure case. TLTR Appendix L considers four ATWS events:  [[                                                                                 
9-25


                                                                                                                                                                                      ]]  The ATWS (Overpressure) - Event Selection meets all CLTR dispositions.
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure Figure 9-13 9-13:: TPO TPO RTP RTP MELLLA MELLLA EOC EOC MSIVC  (Short Term)
MSIVC (Short Term)
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                                                              ))
9-26


As shown in Section 3.7 of ELTR2, [[                                                                                                           
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-14:
                                                                          ]]  The MSIVC, PRFO and LOOP cases were performed for Limerick. The analysis results are given in Table 9-2. The MSIVC, PRFO and LOOP sequence of events are given in Tables 9-6 through 9-8. The short-term and long-term transient response to the MSIVC, PRFO and LOOP ATWS events are pres ented in Figures 9-1 through 9-24. Therefore, ATWS (Overpressure) - Limiting Events meet all CLTR dispositions.
Figure      TPO RTP 9-14: TPO RTP MELLLA MELLLA EOC EOC MSIVC  (Long Term MSIVC (Long  Term -- A)
9.3.1.2 ATWS (Suppression Pool Temperature) As stated in Section 9.3.1 of the CLTR, the higher operating steam flow will result in higher peak vessel pressures. The higher power and decay heat may result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could impact the capability of the SLCS to mitigate the
A)
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                                                                ))
9-27


consequences of an ATWS event. The suppression pool temperature evaluation includes consideration of the most limiting RHR pool cooling capability case. TLTR Appendi x L considered four ATWS events:  [[                       
NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-15: TPO Figure 9-15: TPO RTP RTP MELLLA MELLLA EOC EOC MSIVC  (Long Term MSIVC (Long Term -- B)
B)
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                                                                ))
9-28


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-5                                                                                                                                                                                     
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY  INFORMATION NON-PROPRIETARY INFORMATION Figure Figure 9-16 9-16:: TPO TPO RTP RTP MELLLA MELLLA EOC Eoe MSIVC  (Long Term MSIVe (Long  Term -- C)
C)
((
                                                                ))
9-29


                                                                                                                                                            ]]  The ATWS (Suppression Pool Temperature) - Event Selection meets all CLTR dispositions. The MSIVC, PRFO and LOOP cases were performed for Limerick. The key inputs to the ATWS analysis are provided in Table 9-1. The ATWS analysis results are given in Table 9-2.
NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-17:
The MSIVC, PRFO and LOOP sequence of events are given in Tables 9-6 through 9-8. The ATWS (Suppression Pool Temperature) - Limiting Events meet all CLTR dispositions.
Figure 9-1 7: TPO TPO RTP RTP MELLLA MELLLA EOCEOC PRFO  (Short Term)
9.3.1.3 ATWS (Peak Cladding Temperature)
PRFO (Short  Term)
The TLTR in Appendix L.3 states that power uprate has a negligible effect on the PCT or local cladding oxidation.  [[                                                                                                                                         
((
                                                            ))
9-30


                                                                                                                                                                              ]] For ATWS events, the acceptance criteria for PCT and local cladding oxidation for emergency core cooling systems, defined in 10 CFR 50.46, are adopted to ensure an ATWS event does not impede core cooling.
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-Figure    1 8: TPO 9-18:  TPO RTP RTP MELLLA MELLLA EOC EOC PRFO PRFO (Long (Long Term Term -- A)
For TPO, PCT and local cladding oxidation are not required to be explicitly analyzed per Appendix L.3 of TLTR. Therefore, ATWS (PCT) is in compliance with the acceptance criteria of 10 CFR 50.46; subsequently, coolable core geometry is assured by meeting the 2200&#xba;F PCT and the 17% local cladding oxidation accep tance criteria stated in 10 CFR 50.46.
A)
9.3.1.4 ATWS with Core Instability The CLTR in Section 9.3.3 states that 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 TPO operation because the MELLLA upper boundary is not increased. The core
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                                                                ))
9-31


design necessary to achieve TPO operations may affect the susceptibility to coupled thermal-hydraulic/neutronic core oscillations at th e natural circulation condition, but would not significantly affect the event progression.
NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure Figure 9-19:
Several factors affect the response of an ATWS instability event, including operating power and flow conditions and core design. The limiting ATWS core instability ev aluation presented in References 20 and 21 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum fl ow point. [[                                                                                                                 
9-19: TPO TPO RTP RTP MELLLA MELLLA EOC EOC PRFO (Long Term PRFO (Long  Term -- B)
B)
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                                                              ))
9-32


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-6                                                                                                                                                                                     
NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-20:
                                      ]] TPO allows plants to increase their operating thermal power but does not allow an increase in control rod line. [[                                                                                                                                               
Figure 9-20: TPO TPO RTP RTP MELLLA MELLLA EOC EOC PRFO (Long Term PRFO (Long Term -- C)
C)
((
                                                              ))
9-33


                                                                                              ]] [[                                         
NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-21:
Figure 9-21 : TPO TPO RTP RTP MELLLA MELLLA EOCEOC LOOP LOOP (Short (Short Term)
Term)
((
                                                            ))
9-34


                                                                                                                                                          ]] Initial operating conditions of FWHOOS and FFWTR do not significantly affect the ATWS instability response reported in References 20 and 21. The limiting ATWS evaluation assumes that all FW heating is lost during the event and the injected FW temperature approaches the lowest achievable main condenser hot well temperature. [[                                                                     
NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-22:
Figure 9-22 : TPO TPO RTP RTP MELLLA MELLLA Eoe EOC LOOP LOOP (Long (Long Term Term -- A)
A)
((
                                                                ))
9-35


                      ]] [[                                                                                                                                                                               
NEDO-33484 REVISION NEDO-33484  REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY  INFORMATION Figure 9-23:
Figure 9-23: TPO TPO RTP RTP MELLLA MELLLA EOC EOC LOOP LOOP (Long (Long Term Term -- B)
B)
((
                                                                ))
9-36


                                                                      ]]  Therefore, the TPO effect on ATWS with core instability at Limerick meets all CLTR dispositions.
NEDO-33484 REVISION NEDO-33484  REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY   INFORMATION Figure 9-24:
9.3.1.5 SLCS System Performance and Hardware Based on the results of the [[    ]] ATWS analysis, the maximum reactor upper plenum pressure following the limiting ATWS event reaches 1224 psig (1239 psia) during the time the SLCS is analyzed to be in operation. Consequently, there is a corresponding increase in the maximum two pump discharge pressure to 1336 psig and 1330 psig for Limerick Units 1 and 2 respectively and a decrease in the operating pressure margin for the pump discharge relief valves. The relief valve margin is not satisfied with three SLCS pump operation; however, the relief valve margin is satisfied with two SLCS pump operation. This is acceptable because they meet the injection requirements of 10 CFR 50.62 and the injection requirements of Information Notice 2001-13 with two pumps. The key SLCS input parameters are summ arized in Table 9-3 and Table 9-4 for three and two SLCS pump operation, respectively. The key SLCS NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-7 performance results are presented in Table 9-5. Consideration was also given to system flow, head losses for full injection, and cyclic pressure pulsations due to the positive displacement pump operation in determining the pressure margin to the opening set point for the pump discharge relief valves. The relief valve setpoint margin for two SLCS pump operation is 19.7 psi and 25.8 psi for Limerick Units 1 and 2, respectively. This margin is based on a SLCS pump relief valve setpoint of 1400 psig. The pump discharge relief valves are periodically tested to maintain this tolerance. Therefore, the current SLCS process parameters associated with the minimum boron injection rate are not changed. The SLCS ATWS performance is evaluated for a representative core design for TPO. The evaluation shows that TPO has no adverse effect on the ability of the SLCS to mitigate an ATWS in two SLCS pump operation. There are no timer setting changes for TPO for Limerick, and the ATWS analysis confirms acceptable results. Therefore, the system performance and hardware meets all CLTR dispositions.
Figure 9-24: TPO TPO RTP RTP MELLLA MELLLA EOC EOC LOOP LOOP (Long (Long Term Term -- C)
9.3.1.6 Suppression Pool Temperature following ATWS Event As stated in Section 6.5 of the CLTR, chan ges in the fuel design for TPO may require modifications to the SLCS as a result of the increase in the suppression pool temperature for the limiting ATWS event.
C)
The boron injection rate requirement for maintaining the peak suppression pool water temperature limits, following the limiting ATWS event with SLCS injection, is not increased for TPO. Therefore, the Suppression Pool temperature following an ATWS event meets all CLTR dispositions.
((
9.3.1.7 Equipment Out-of-Service and Flexibility Options MELLLA, ICF and SLO:  The TPO ATWS analyses were performed along the MELLLA boundary. The TPO ATWS analysis at MELLLA conditions bounds operation at ICF and in SLO. Therefore, TPO continues to support these performance improvement features. SRV OOS:  The TPO ATWS analysis was performed with two SRVs OOS. Therefore, TPO continues to support this EOOS option. FWHOOS and FFWTR:  FWHOOS and FFWTR are operational flexibility options that allow continued operation with reduced FW temperatur
                                                                ))
: e. Initial power is unchanged for both the FWHOOS and FFWTR conditions. The additional reactivity associated with the reduced FW temperature is typically offset with control rods, as needed. This makes the core less reactive due to the lower void fraction. Thus, use of normal FW temperature is conservative for ATWS analyses.
9-37
MSIV OOS:  The TPO ATWS analysis bounds the MSIV OOS condition for TPO. Limerick operation with MSIV OOS is limited to  75% rated power. With this restriction, the severity of the limiting ATWS events is reduced. The lower initial steaming rate reduces the peak vessel NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-8 pressure, peak power, PCT, and integrated SRV flow. The reduction in integrated SRV flow thereby reduces the peak suppression pool temperature and containment pressure. ARTS, TBV OOS, RPT OOS, 24 Month Cycle, TCV Stuck Closed and TSV Stuck Closed:  The TPO ATWS analysis is not impacted by these performance improvement features.
9.3.2 Station Blackout The Limerick Station Blackout (SBO) evaluation has previously been performed assuming  102% of CLTP. Therefore, the postulated SBO scenarios for TPO operation are bounded by the current evaluations.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-9 Table 9-1 Key Inputs for ATWS Analysis Input Variable CLTP TPO RTP Reactor power (MWt) 3458 3517* Reactor dome pressure (psia) 1060 1060 Each SRV capacity at 1090 psig (Mlbm/hr) 0.870 0.870 High pressure ATWS-RPT (psig) 1156 1156 Number of SRVs OOS 2 2 Number of Auto Start SLCS Pumps 2 2
* Performed at 101.7% of CLTP Table 9-2 Results for ATWS Analysis Acceptance Criteria CLTP 1, 2 TPO RTP 1 Peak vessel bottom pressure (psig) 1458 1473 Peak suppression pool temperature ( F) 181 182 Peak containment pressure (psig) 10.3 10.6 Peak cladding temperature ( F) Generic Assessment Generic Assessment Local cladding oxidation (%)
Generic Assessment Gene ric Assessment Notes: 1. Cladding temperature and oxi dation calculations are not required per Appendix L.3 of TLTR. 2. To maximize the effect of TPO, a baseline is established at the CLTP level, assuming the current licensed equipment performance assumpti ons and plant parameters.
Table 9-3 Inputs for Limerick Three SL CS Pump Operation ATWS Analysis Parameter Limerick Unit 1 Limerick Unit 2 Reactor Water Elevation Head (psi) 2.5 2.5 CS Nozzle Differential Pressure  (psi) 20.8 20.8 SLCS Pipe Elevation (psi) 10.0 10.0 Flow Friction Loss (psi) 126.0 112.1 Total Three Pump Operation System Losses (psi) 159.3 145.4 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-10 Table 9-4 Inputs for Limerick Two SLCS Pump Operation ATWS Analysis Parameter Limerick Unit 1 Limerick Unit 2 Reactor Water Elevation Head (psi) 2.5 2.5 CS Nozzle Differential Pressure  (psi) 20.3 20.3 SLCS Pipe Elevation (psi) 10.0 10.0 Flow Friction Loss (psi) 79.5 73.4 Total Two Pump Operation System Losses (psi) 112.3 106.2  Table 9-5 Limerick SLCS Pressu re Results for ATWS Analysis Parameter Limerick Unit 1 Limerick Unit 2 SLCS RV Setpoint (psig) 1400 1400 RV Tolerance (%)
1.0 1.0 Pulsating Margin (psi) 30.0 30.0 SLCS Piping Flow Loss (psi) - 3 Pump Operation 159.3 145.4 SLCS Piping Flow Loss (psi) - 2 Pump Operation 112.3 106.2 Pressure at Point of SLCS Injection (psig) 1224 1224 Pump Discharge Pressure (psig) - 3 Pump Operation 1383.3 1369.4 Pump Discharge Pressure (psig) - 2 Pump Operation 1336.3 1330.2 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-11 Table 9-6 MSIVC Sequence of Events Item Event TPO RTP BOC Event Time  (sec) TPO RTP EOC Event Time  (sec) 1 MSIV Isolation Initiated 0.0 0.0 2 MSIVs Fully Closed 4.0 4.0 3 Peak Neutron Flux 4.1 4.0 4 High Pressure ATWS Setpoint 4.1 4.1 5 Recirculation Pumps Trip 4.7 4.6 6 Start Opening of the First Relief Valve 4.9 4.8 7 Peak Heat Flux 5.2 5.1 8 Peak Vessel Pressure 9.5 9.1 9 Feedwater Reduction Initiated 30.0 30.0 10 BIIT Reached 53.0 54.0 11 SLCS Pumps Start 124.1 124.1 12 Hot Shutdown Achieved (Neutron Flux Remains < 0.1%)
422 444 13 RHR Cooling Initiated 660 660 14 Peak Suppression Pool Temperature 8527 8322 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-12 Table 9-7 PRFO Sequence of Events Item Event TPO RTP BOC Event Time  (sec) TPO RTP EOC Event Time  (sec) 1 TCV and Bypass Valves Start Open 0.1 0.1 2 MSIV Closure Initiated by Low Steamline Pressure 20.2 19.3 3 MSIVs Fully Closed 24.2 23.3 4 Peak Neutron Flux 25.2 24.8 5 High Pressure ATWS Setpoint 27.8 27.0 6 Recirculation Pumps Trip 28.4 27.5 7 Start Opening of the First Relief Valve 28.6 27.7 8 Peak Heat Flux 28.8 27.5 9 Peak Vessel Pressure 35.2 33.9 10 Feedwater Reduction Initiated 53.4 53.4 11 BIIT Reached 69.0 70.0 12 SLCS Pumps Start 147.8 147.0 13 Hot Shutdown Achieved (Neutron Flux Remains < 0.1%)
451 477 14 RHR Cooling Initiated 660 660 15 Peak Suppression Pool Temperature 7871 7131 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9-13 Table 9-8 LOOP Sequence of Events Item Event TPO RTP BOC Event Time  (sec) TPO RTP EOC Event Time  (sec) 1 Loss of Auxiliary Power 0.0 0.0 2 Recirculation Pumps Trip 0.0 0.0 3 Feedwater Reduction Initiated 0.0 0.0 4 TCV Closure 0.0 0.0 5 Peak Neutron Flux 0.7 0.7 6 High Pressure ATWS Setpoint 1.0 0.9 7 Start Opening of the First Relief Valve 2.0 1.9 8 MSIV Closure Initiated 2.0 2.0 9 Peak Heat Flux 2.3 2.2 10 MSIVs Fully Closed 6.0 6.0 11 Peak Vessel Pressure 8.3 7.8 12 BIIT Reached 54.0 55.0 13 SLCS Pumps Start 121.0 120.9 14 Hot Shutdown Achieved (Neutron Flux Remains < 0.1%)
419 454 15 RHR Cooling Initiated 660 660 16 Peak Suppression Pool Temperature ~29000 ~29000 NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-1: TPO RTP MELLLA BOC MSIVC (Short Term) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-1: TPO RTP MELLLA DOC MSIVC (Short Term)9-14]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-2: TPO RTP MELLLA BOC MSIVC (Long Term - A) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-2: TPO RTP MELLLA HOC MSIVC (Long Term-A)9-15]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-3: TPO RTP MELLLA BOC MSIVC (Long Term - B) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-3: TPO RTP MELLLA BOC MSIVC (Long Term-B)9-16]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-4: TPO RTP MELLLA BOC MSIVC (Long Term - C) [[NEDO-33484 REVISION 0 NON-PROPRIETARY IN FORMA TION Figure 9-4: TPO RTP MELLLA BOC MSIVC (Long Term-C)9-17]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-5: TPO RTP MELLLA BOC PRFO (Short Term) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-5: TPO RTP MELLLA DOC PRFO (Short Term)9-18]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-6: TPO RTP MELLLA BOC PRFO (Long Term - A) [[NEDO-33484 REVISION 0 NON-PROPRIETARY IN FORMA TION Figure 9-6: TPO RTP MELLLA DOC PRFO (Long Term-A)9-19]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-7: TPO RTP MELLLA BOC PRFO (Long Term - B) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-7: TPO RTP MELLLA BOC PRFO (Long Term-B)9-20]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-S: TPO RTP MELLLA BOC PRFO (Long Term - C) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-8: TPO RTP MELLLA BOC PRFO (Long Term-C)9-21]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-9: TPO RTP MELLLA BOC LOOP (Short Term) [[NEDO-33484 REVISION 0 NON-PROPRIETARY IN FORMA TION Figure 9-9: TPO RTP MELLLA BOC LOOP (Short Term)9-22]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-10: TPO RTP MELLLA BOC LOOP (Long Term - A) [[NEDO-33484 REVISION 0 NON-PROPRIETARY IN FORMA TION Figure 9-10: TPO RTP MELLLA DOC LOOP (Long Term-A)9-23]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-11: TPO RTP MELLLA BOC LOOP (Long Term - B) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-11: TPO RTP MELLLA BOC LOOP (Long Term-B)9-24]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-12: TPO RTP MELLLA BOC LOOP (Long Term - C) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-12: TPO RTP MELLLA BOC LOOP (Long Term-C)9-25]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-13: TPO RTP MELLLA EOC MSIVC (Short Term) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-13: TPO RTP MELLLA EOC MSIVC (Short Term)9-26]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-1 4: TPO RTP MELLLA EOC MSIVC (Long Term - A) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-14: TPO RTP MELLLA EOC MSIVC (Long Term-A)9-27]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-15: TPO RTP MELLLA EOC MSIVC (Long Term - B) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-15: TPO RTP MELLLA EOC MSIVC (Long Term-B)9-28]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-16: TPO RTP MELLLA EOC MSIVC (Long Term - C) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-16: TPO RTP MELLLA Eoe MSIVe (Long Term-C)9-29]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-1 7: TPO RTP MELLLA EOC PRFO (Short Term) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-17: TPO RTP MELLLA EOC PRFO (Short Term)9-30]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-1 8: TPO RTP MELLLA EOC PRFO (Long Term - A) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-18: TPO RTP MELLLA EOC PRFO (Long Term-A)9-31]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-1 9: TPO RTP MELLLA EOC PRFO (Long Term - B) [[NEDO-33484 REVISION 0 NON-PROPRIETARY IN FORMA TION Figure 9-19: TPO RTP MELLLA EOC PRFO (Long Term-B)9-32]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-2 0: TPO RTP MELLLA EOC PRFO (Long Term - C) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-20: TPO RTP MELLLA EOC PRFO (Long Term-C)9-33]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-21: TPO RTP MELLLA EOC LOOP (Short Term) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-21: TPO RTP MELLLA EOC LOOP (Short Term)9-34]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-22: TPO RTP MELLLA EOC LOOP (Long Term - A) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-22: TPO RTP MELLLA Eoe LOOP (Long Term-A)9-35]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-2 3: TPO RTP MELLLA EOC LOOP (Long Term - B) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-23: TPO RTP MELLLA EOC LOOP (Long Term-B)9-36]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-24: TPO RTP MELLLA EOC LOOP (Long Term - C) [[NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 9-24: TPO RTP MELLLA EOC LOOP (Long Term-C)9-37]]
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10-1 10.0 OTHER EVALUATIONS 10.1 H IGH ENERGY LINE BREAK  Because the TPO uprate system operating temperatures and pressures change only slightly, there is no significant change in HELB mass and energy releases. The FW lines, near the pump discharge, increase < 2&deg;F and ~ 4 psi. The recirculation line temperature decreases approximately 0.1&deg;F with a 0.1 BTU/lbm enthalpy d ecrease. These changes are insignificant in relation to the effect on line break calculations.
Vessel dome pressure and other portions of the RCPB remain at current operating pressure or lower. Therefore, the consequences of any postulated HELB would not significantly change. The postulated break locations remain the same because the piping configuration does not change due to the TPO uprate.
The HELB evaluation was performed for all systems evaluated in the UFSAR. At the TPO RTP level, HELBs outside the drywell would result in an insignificant change in the sub-compartment pressure and temperature profiles. The affected building and cubicles that support safety-related functions are designed to withstand the resulti ng pressure and thermal loading following an HELB at the TPO RTP. A brief di scussion of each break follows.
10.1.1 Steam Line Breaks  The critical parameter affecting the high-energy steam line break analysis is the reactor vessel dome pressure. Because there is no pressure increase for the TPO, the main steam line (MSL) pressure decreases and there is a slight decrease in the main steam line break (MSLB) blowdown rate. The MSLB is used to establish the peak pressure and the temperature environment in the MS tunnel. Design margins within the HELB analysis for a MSLB with a concurrent FW line break provide adequate margin to the limits in the steam tunnel.
10.1.2 Liquid Line Breaks 10.1.2.1 Feedwater Line Breaks  The TPO uprate increases the FW temperature < 2&deg; F and pressure ~ 4 psi, which results in an insignificant increase in the FW mass and energy release. As a result of the small increase in FW temperature and pressure, the blowdown rate changes marginally and the energy increases slightly. The original analysis was generally performed with conservative modeling assumptions. These conservatisms more than offset the effects of the temperature change. 


NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10.0 OTHER EVALUATIONS 10.1 HIGH ENERGY LINE BREAK Because the TPO uprate system operating temperatures and pressures change only slightly, there is no significant change in HELB mass and energy releases. The FW lines, near the pump discharge, increase < 2&deg;F and ~ 4 psi. The recirculation line temperature decreases approximately 0.1&deg;F with a 0.1 BTU/lbm enthalpy decrease. These changes are insignificant in relation to the effect on line break calculations. Vessel dome pressure and other portions of the RCPB remain at current operating pressure or lower. Therefore, the consequences of any postulated HELB would not significantly change. The postulated break locations remain the same because the piping configuration does not change due to the TPO uprate.
The HELB evaluation was performed for all systems evaluated in the UFSAR. At the TPO RTP level, HELBs outside the drywell would result in an insignificant change in the sub-compartment pressure and temperature profiles. The affected building and cubicles that support safety-related functions are designed to withstand the resulting pressure and thermal loading following an HELB at the TPO RTP. A brief discussion of each break follows.
10.1.1 Steam Line Breaks The critical parameter affecting the high-energy steam line break analysis is the reactor vessel dome pressure. Because there is no pressure increase for the TPO, the main steam line (MSL) pressure decreases and there is a slight decrease in the main steam line break (MSLB) blowdown rate. The MSLB is used to establish the peak pressure and the temperature environment in the MS tunnel. Design margins within the HELB analysis for a MSLB with a concurrent FW line break provide adequate margin to the limits in the steam tunnel.
10.1.2 Liquid Line Breaks 10.1.2.1 Feedwater Line Breaks The TPO uprate increases the FW temperature < 2&deg;F and pressure ~ 4 psi, which results in an insignificant increase in the FW mass and energy release. As a result of the small increase in FW temperature and pressure, the blowdown rate changes marginally and the energy increases slightly. The original analysis was generally performed with conservative modeling assumptions. These conservatisms more than offset the effects of the temperature change.
Therefore, the original HELB analysis is bounding.
Therefore, the original HELB analysis is bounding.
10.1.2.2 ECCS Line Breaks Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase. Therefore, the previous HELB analysis is bounding for the TPO uprate condition.
10.1.2.2 ECCS Line Breaks Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase. Therefore, the previous HELB analysis is bounding for the TPO uprate condition.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10-2 Because these lines are normally isolated, the TPO uprate does not affect their line break analyses, for breaks outside drywell.
10-1
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Because these lines are normally isolated, the TPO uprate does not affect their line break analyses, for breaks outside drywell.
10.1.2.3 RCIC System Line Breaks Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase. Therefore, the previous HELB analysis is bounding for the TPO uprate conditions.
10.1.2.3 RCIC System Line Breaks Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase. Therefore, the previous HELB analysis is bounding for the TPO uprate conditions.
10.1.2.4 RWCU System Line Breaks As a result of the small decrease in recirculation temperature with a negligible increase in pressure, the blowdown rate increases slightly and the energy decreases slightly. The original analysis was generally performed with conservative modeling assumptions. These conservatisms more than offset the effects of the temperature change, so the original HELB analysis is bounding.
10.1.2.4 RWCU System Line Breaks As a result of the small decrease in recirculation temperature with a negligible increase in pressure, the blowdown rate increases slightly and the energy decreases slightly. The original analysis was generally performed with conservative modeling assumptions.                   These conservatisms more than offset the effects of the temperature change, so the original HELB analysis is bounding.
10.1.2.5 CRD System Line Breaks The CRD pipe rupture analysis is not affected by the TPO uprate.
10.1.2.5 CRD System Line Breaks The CRD pipe rupture analysis is not affected by the TPO uprate.
10.1.2.6 Building Heating Line Breaks Building heating lines are not connected to the reactor-turbine primary loop. Therefore, building heating lines are not affected.
10.1.2.6 Building Heating Line Breaks Building heating lines are not connected to the reactor-turbine primary loop. Therefore, building heating lines are not affected.
10.1.2.7 Pipe Whip and Jet Impingement Because there is no change in the nominal vessel dome pressure, pipe whip and jet impingement loads do not significantly change. Existing calculati ons supporting the dispos itions of potential targets of pipe whip and jet impingement from postulated HELBs have been reviewed and determined to be adequate for the safe shut down effects in the TPO RTP conditions. Existing pipe whip restraints, jet impingement shields, a nd their supporting structures are also adequate for the TPO uprate conditions.
10.1.2.7 Pipe Whip and Jet Impingement Because there is no change in the nominal vessel dome pressure, pipe whip and jet impingement loads do not significantly change. Existing calculations supporting the dispositions of potential targets of pipe whip and jet impingement from postulated HELBs have been reviewed and determined to be adequate for the safe shutdown effects in the TPO RTP conditions. Existing pipe whip restraints, jet impingement shields, and their supporting structures are also adequate for the TPO uprate conditions.
10.1.2.8 Internal Flooding from HELB None of the plant flooding zones contains a potential HELB location affected by the reactor operating conditions changed for the TPO uprate. The high-energy line systems' operational modes evaluated for HELB are not affected by the TPO uprate, nor are the plant internal  
10.1.2.8 Internal Flooding from HELB None of the plant flooding zones contains a potential HELB location affected by the reactor operating conditions changed for the TPO uprate. The high-energy line systems operational modes evaluated for HELB are not affected by the TPO uprate, nor are the plant internal flooding analysis or safe shutdown analysis.
10.2 MODERATE ENERGY LINE BREAK None of the plant flooding zones contains a potential Moderate Energy Line Break (MELB) location affected by the reactor operating conditions changed for the TPO uprate. The following 10-2


flooding analysis or sa fe shutdown analysis. 10.2 MODERATE E NERGY LINE BREAK  None of the plant flooding z ones contains a potential Mode rate Energy Line Break (MELB) location affected by the reactor operating conditions changed for the TPO uprate. The following NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10-3 systems contain potential MELB locations in pl ant flooding zones: Condensate, Condensate and Refueling Water Storage, SW, Emergency Service Water, RHR, RHR Service Water, Reactor Enclosure Cooling Water, Condensate Filter Demineralizers, Make-up Demineralizer, Fuel and Diesel Oil Storage and Transfer, Auxiliary Steam, Fire Protection, RWCU System, Clean-up Filter Demineralizer, Control Rod Drive Hydraulic, SLCS, RCIC, RCIC Pump Turbine, Core Spray and Safeguard Piping, Fuel Pool Cooling and Cleanup, Fuel Pool Filter/Demineralizer, HPCI, HPCI Pump Turbine, Liquid Radwaste Collection, Liquid Radwaste-Chemical and Laundry Processing, Drywell Chilled Water, and Control Structure Chilled Water. No new moderate energy lines are identified. Protection requirements for safe-shutdown equipment for a postulated MELB are not depende nt on power level.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION systems contain potential MELB locations in plant flooding zones: Condensate, Condensate and Refueling Water Storage, SW, Emergency Service Water, RHR, RHR Service Water, Reactor Enclosure Cooling Water, Condensate Filter Demineralizers, Make-up Demineralizer, Fuel and Diesel Oil Storage and Transfer, Auxiliary Steam, Fire Protection, RWCU System, Clean-up Filter Demineralizer, Control Rod Drive Hydraulic, SLCS, RCIC, RCIC Pump Turbine, Core Spray and Safeguard Piping, Fuel Pool Cooling and Cleanup, Fuel Pool Filter/Demineralizer, HPCI, HPCI Pump Turbine, Liquid Radwaste Collection, Liquid Radwaste-Chemical and Laundry Processing, Drywell Chilled Water, and Control Structure Chilled Water.
All sources of and protection measures against flooding are independent of power level. Internal flooding will not alter the ability of the plant to reach safe shutdown under TPO. Therefore, the plant internal flooding analysis is not affected. 10.3 ENVIRONMENTAL QUALIFICATION Safety-related components must be qualified for the environment in which they operate. The TPO increase in power level increases the radiation levels experienced by equipment during normal operation and accident conditions. Because the TPO uprate does not increase the nominal vessel dome pressure, there is a very small effect on pressure and temperature conditions experienced by equipment during norm al operation and accide nt conditions. The resulting environmental conditions are bounded by the existing environmental parameters specified for use in the environmental qualification program.
No new moderate energy lines are identified. Protection requirements for safe-shutdown equipment for a postulated MELB are not dependent on power level. All sources of and protection measures against flooding are independent of power level. Internal flooding will not alter the ability of the plant to reach safe shutdown under TPO. Therefore, the plant internal flooding analysis is not affected.
10.3 ENVIRONMENTAL QUALIFICATION Safety-related components must be qualified for the environment in which they operate. The TPO increase in power level increases the radiation levels experienced by equipment during normal operation and accident conditions. Because the TPO uprate does not increase the nominal vessel dome pressure, there is a very small effect on pressure and temperature conditions experienced by equipment during normal operation and accident conditions. The resulting environmental conditions are bounded by the existing environmental parameters specified for use in the environmental qualification program.
10.3.1 Electrical Equipment The environmental conditions for safety-related electrical equipment were reviewed to ensure that the existing qualification for the normal and accident conditions expected in the area where the devices are located remain adequate. Conservatisms in the equipment qualifications were originally applied to the environmental parameters, and no change is needed for the TPO uprate.
10.3.1 Electrical Equipment The environmental conditions for safety-related electrical equipment were reviewed to ensure that the existing qualification for the normal and accident conditions expected in the area where the devices are located remain adequate. Conservatisms in the equipment qualifications were originally applied to the environmental parameters, and no change is needed for the TPO uprate.
10.3.1.1 Inside Containment Environmental qualification (EQ) for safety-related electrical equipment located inside the containment is based on DBA-LOCA conditions and their resultant temperature, pressure, humidity and radiation consequences, and includes the environments expected to exist during normal plant operation. The current accident conditions for temperature and pressure are based on analyses initiated from  102% of CLTP. Normal temperatures may increase slightly near the FW and reactor recirculation lines and will be evaluated through the EQ temperature monitoring program, which tracks such information for equipment aging considerations. The current radiation levels under normal plant conditions also increase s lightly. The current plant environmental envelope for radiation is not exceeded by the changes resulting from the TPO uprate.
10.3.1.1 Inside Containment Environmental qualification (EQ) for safety-related electrical equipment located inside the containment is based on DBA-LOCA conditions and their resultant temperature, pressure, humidity and radiation consequences, and includes the environments expected to exist during normal plant operation. The current accident conditions for temperature and pressure are based on analyses initiated from  102% of CLTP. Normal temperatures may increase slightly near the FW and reactor recirculation lines and will be evaluated through the EQ temperature monitoring program, which tracks such information for equipment aging considerations. The current radiation levels under normal plant conditions also increase slightly. The current plant environmental envelope for radiation is not exceeded by the changes resulting from the TPO uprate.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10-4 10.3.1.2 Outside Containment Accident temperature, pressure, and humidity environments used for qualification of equipment outside containment result from an MSLB in th e pipe tunnel, or other HELBs, whichever is limiting for each area. The HELB pressure and temperature profiles bound the TPO uprate conditions. There is adequate margin in the qualification envelopes to accommodate the small changes due to TPO conditions. Maximum accident radiation levels used for qualification of equipment outside containment are from a DBA-LOCA.
10-3
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10.3.1.2 Outside Containment Accident temperature, pressure, and humidity environments used for qualification of equipment outside containment result from an MSLB in the pipe tunnel, or other HELBs, whichever is limiting for each area. The HELB pressure and temperature profiles bound the TPO uprate conditions. There is adequate margin in the qualification envelopes to accommodate the small changes due to TPO conditions. Maximum accident radiation levels used for qualification of equipment outside containment are from a DBA-LOCA.
10.3.2 Mechanical Equipment With Non-Metallic Components Operation at the TPO RPT level increases the normal process temperature very slightly in the FW piping. The slight increase in normal and accident radiation was evaluated in Section 10.3.
10.3.2 Mechanical Equipment With Non-Metallic Components Operation at the TPO RPT level increases the normal process temperature very slightly in the FW piping. The slight increase in normal and accident radiation was evaluated in Section 10.3.
10.3.3 Mechanical Component Design Qualification The increase in power level increases the ra diation levels experienced by equipment during normal operation. However, where the previous accident analyses have been based on 102% of CLTP, the accident pressures, temperatures and radiation levels do not change. The mechanical design of equipment/components (valves, heat exchangers, pumps, snubbers, etc.) in certain systems is affected by operation at the TPO RTP level because of the slightly increased temperature and sometimes flow rate. The revised operating conditio ns do not significantly affect the cumulative usage fatigue factors of mechanical components.
10.3.3 Mechanical Component Design Qualification The increase in power level increases the radiation levels experienced by equipment during normal operation. However, where the previous accident analyses have been based on 102% of CLTP, the accident pressures, temperatures and radiation levels do not change. The mechanical design of equipment/components (valves, heat exchangers, pumps, snubbers, etc.) in certain systems is affected by operation at the TPO RTP level because of the slightly increased temperature and sometimes flow rate. The revised operating conditions do not significantly affect the cumulative usage fatigue factors of mechanical components.
The effects of increased fluid induced loads on safety-related components are described in Section 3.4. As stated in Section 4.1, the containment loads for the TPO uprate are bounded by previous analyses at 102% of CLTP. Increased nozzle loads and component support loads due to the revised operating conditions we re evaluated in the piping assessments in Section 3.4. These increased loads are insignificant, and become negligible when combined with the dynamic loads. Therefore, the mechanical components and component supports are adequately designed for the TPO uprate conditions.
The effects of increased fluid induced loads on safety-related components are described in Section 3.4. As stated in Section 4.1, the containment loads for the TPO uprate are bounded by previous analyses at 102% of CLTP. Increased nozzle loads and component support loads due to the revised operating conditions were evaluated in the piping assessments in Section 3.4. These increased loads are insignificant, and become negligible when combined with the dynamic loads.
10.4 TESTING The TPO uprate power ascension is based on the guidelines in TLTR Section L.2. Pre-operational tests are not needed because there are no significant changes to any plant systems or components that require such testing.
Therefore, the mechanical components and component supports are adequately designed for the TPO uprate conditions.
10.4 TESTING The TPO uprate power ascension is based on the guidelines in TLTR Section L.2. Pre-operational tests are not needed because there are no significant changes to any plant systems or components that require such testing.
In preparation for operation at TPO uprate conditions, routine measurements of reactor and system pressures, flows, and selected major rotating equipment vibration are taken near 95% and 100% of CLTP, and at 100% of TPO RTP. The measurements will be taken along the same rod pattern line used for the increase to TPO RTP. Core power from the APRMs is re-scaled to the TPO RTP before exceeding the CLTP and any necessary adjustments will be made to the APRM alarm and trip settings.
In preparation for operation at TPO uprate conditions, routine measurements of reactor and system pressures, flows, and selected major rotating equipment vibration are taken near 95% and 100% of CLTP, and at 100% of TPO RTP. The measurements will be taken along the same rod pattern line used for the increase to TPO RTP. Core power from the APRMs is re-scaled to the TPO RTP before exceeding the CLTP and any necessary adjustments will be made to the APRM alarm and trip settings.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10-5 The turbine pressure controller setpoint will be readjusted at 95% of CLTP and held constant. The setpoint is reduced so the reactor dome pressure is the same at TPO RTP as for the CLTP. Adjustment of the pressure setpoint before taki ng the baseline power ascension data establishes a consistent basis for measuring the performance of the reactor and the turbine control valves. Demonstration of acceptable fuel thermal margin will be performed prior to and during power ascension to the TPO RTP at each steady-state heat balance point defined above. Fuel thermal margin will be projected to the TPO RTP point after the measurements taken at 95% and 100% of CLTP to show the estimated margin. The thermal margin will be confirmed by the measurements taken at full TPO RTP conditions. The demonstration of core and fuel conditions will be performed with the methods currently used at Limerick. Performance of the pressure and FW/level control systems will be recorded at each steady-state point defined above. The checks will utilize the methods and criter ia described in the original startup testing of these systems to demonstrate acceptable operational capability. Water level changes of +/-3 inches and pressure setpoint step changes of +/-3 psi will be used. If necessary, adjustments will be made to the controllers and actuator elements. The increase in power for the TPO uprate is sufficiently small that large transient tests are not necessary. High power testing performed during initial startup demonstrated the adequacy of the safety and protection systems for such large transients. Operational occu rrences have shown the unit response is clearly bounded by the sa fety analyses for these events.  [[                                       
10-4


                                                                                                ]] 10.5 OPERATOR TRAINING A ND H UMAN F ACTORS  No additional training (apart from normal training for plant changes) is required to operate the plant in the TPO uprate condition.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION The turbine pressure controller setpoint will be readjusted at 95% of CLTP and held constant.
For TPO uprate conditions, opera tor response to transient, accident, and special events is not affected. Operator actions for maintaining safe shutdown, core cooling, containment cooling, etc., do not ch ange for the TPO uprate.
The setpoint is reduced so the reactor dome pressure is the same at TPO RTP as for the CLTP.
Minor changes to the P/F map, flow-referenced setpoint, and the like, will be communicated through normal operator training. Simulator changes and validation for the TPO uprate will be performed in accordance with established Limerick plant certification testing procedures. 10.6 PLANT LIFE Two degradation mechanisms may be influenced by the TPO uprate: (1) Irradiation Assisted Stress Corrosion Cracking (IASCC) and (2) FAC. The increase in irradiation of the core internal components influences IASCC. The increases in steam and FW flow rate influence FAC.
Adjustment of the pressure setpoint before taking the baseline power ascension data establishes a consistent basis for measuring the performance of the reactor and the turbine control valves.
However, the sensitivity to the TPO uprate is small and various programs are currently implemented to monitor the aging of plant components, including EQ, FAC, and In-service Inspection. EQ is addressed in Section 10.3, and FAC is addressed in Section 3.5. These programs address the degradation mechanisms and do not change for the TPO uprate. The core NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10-6 internals see a slight increase in fluence, but th e inspection strategy used at Limerick, based on the BWRVIP, is sufficient to address the incr ease. The Maintenance Rule also provides oversight for the other mechanical and electrical components, important to plant safety, to guard against age-related degradation. The longevity of most equipment is not affected by the TPO uprate because there is no significant change in the operating conditions. No additional maintenance, inspection, testing, or surveillance procedures are required.
Demonstration of acceptable fuel thermal margin will be performed prior to and during power ascension to the TPO RTP at each steady-state heat balance point defined above. Fuel thermal margin will be projected to the TPO RTP point after the measurements taken at 95% and 100%
10.7 NRC AND INDUSTRY COMMUNICATIONS NRC and Industry communications are generically addressed in the TLTR, Section 10.8. Per the TLTR, it is not necessary to review prior dispositions of NRC and industry communications and no additional information is required in this area. 10.8 PLANT PROCEDURES AND PROGRAMS  Plant procedures and programs are in place to: 1. Monitor and maintain instrument calibration during normal plant operation to assure that instrument uncertainty is not greater than the uncertainty used to justify the TPO uprate; 2. Control the software and ha rdware configuration of the associated instrumentation; 3. Perform corrective actions, where required, to maintain instrument uncertainty within limits; 4. Report deficiencies of the associated instruments to the manufacturer; and 5. Receive and resolve the manuf acturer's deficiency reports. 10.9 EMERGENCY OPERATING PROCEDURES The Emergency Operating Procedures (EOPs) action thresholds are plant unique and will be addressed using standard procedure updating processe
of CLTP to show the estimated margin. The thermal margin will be confirmed by the measurements taken at full TPO RTP conditions. The demonstration of core and fuel conditions will be performed with the methods currently used at Limerick.
: s. It is expected that the TPO uprate will have a negligible or no effect on the operator ac tion thresholds and to the EOPs in general. 10.10 INDIVIDUAL PLANT EXAMINATION Limerick maintains and regularly updates a station probabilistic risk assessment (PRA) model. 
Performance of the pressure and FW/level control systems will be recorded at each steady-state point defined above. The checks will utilize the methods and criteria described in the original startup testing of these systems to demonstrate acceptable operational capability. Water level changes of +/-3 inches and pressure setpoint step changes of +/-3 psi will be used. If necessary, adjustments will be made to the controllers and actuator elements.
The increase in power for the TPO uprate is sufficiently small that large transient tests are not necessary. High power testing performed during initial startup demonstrated the adequacy of the safety and protection systems for such large transients. Operational occurrences have shown the unit response is clearly bounded by the safety analyses for these events. ((
                                                    ))
10.5 OPERATOR TRAINING AND HUMAN FACTORS No additional training (apart from normal training for plant changes) is required to operate the plant in the TPO uprate condition. For TPO uprate conditions, operator response to transient, accident, and special events is not affected. Operator actions for maintaining safe shutdown, core cooling, containment cooling, etc., do not change for the TPO uprate. Minor changes to the P/F map, flow-referenced setpoint, and the like, will be communicated through normal operator training. Simulator changes and validation for the TPO uprate will be performed in accordance with established Limerick plant certification testing procedures.
10.6 PLANT LIFE Two degradation mechanisms may be influenced by the TPO uprate: (1) Irradiation Assisted Stress Corrosion Cracking (IASCC) and (2) FAC. The increase in irradiation of the core internal components influences IASCC. The increases in steam and FW flow rate influence FAC.
However, the sensitivity to the TPO uprate is small and various programs are currently implemented to monitor the aging of plant components, including EQ, FAC, and In-service Inspection. EQ is addressed in Section 10.3, and FAC is addressed in Section 3.5. These programs address the degradation mechanisms and do not change for the TPO uprate. The core 10-5


Use of the model is integrated with station operations and decision-making. The Limerick IPE (PRA) model and analysis will not be specifically updated for TPO, because the change in plant risk from the subject power uprate is insi gnificant. This conclusion is supported by NRC RIS 2002-03 (Refer ence 4). In response to f eedback received during the public workshop held on August 23, 2001, the NRC wrote, "The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant effect on plant risk. Therefore, no risk information is requested to support such applications."
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION internals see a slight increase in fluence, but the inspection strategy used at Limerick, based on the BWRVIP, is sufficient to address the increase. The Maintenance Rule also provides oversight for the other mechanical and electrical components, important to plant safety, to guard against age-related degradation.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 11-1
The longevity of most equipment is not affected by the TPO uprate because there is no significant change in the operating conditions. No additional maintenance, inspection, testing, or surveillance procedures are required.
10.7 NRC AND INDUSTRY COMMUNICATIONS NRC and Industry communications are generically addressed in the TLTR, Section 10.8. Per the TLTR, it is not necessary to review prior dispositions of NRC and industry communications and no additional information is required in this area.
10.8 PLANT PROCEDURES AND PROGRAMS Plant procedures and programs are in place to:
: 1. Monitor and maintain instrument calibration during normal plant operation to assure that instrument uncertainty is not greater than the uncertainty used to justify the TPO uprate;
: 2. Control the software and hardware configuration of the associated instrumentation;
: 3. Perform corrective actions, where required, to maintain instrument uncertainty within limits;
: 4. Report deficiencies of the associated instruments to the manufacturer; and
: 5. Receive and resolve the manufacturers deficiency reports.
10.9 EMERGENCY OPERATING PROCEDURES The Emergency Operating Procedures (EOPs) action thresholds are plant unique and will be addressed using standard procedure updating processes. It is expected that the TPO uprate will have a negligible or no effect on the operator action thresholds and to the EOPs in general.
10.10 INDIVIDUAL PLANT EXAMINATION Limerick maintains and regularly updates a station probabilistic risk assessment (PRA) model.
Use of the model is integrated with station operations and decision-making.
The Limerick IPE (PRA) model and analysis will not be specifically updated for TPO, because the change in plant risk from the subject power uprate is insignificant. This conclusion is supported by NRC RIS 2002-03 (Reference 4). In response to feedback received during the public workshop held on August 23, 2001, the NRC wrote, The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant effect on plant risk. Therefore, no risk information is requested to support such applications.
10-6
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION


==11.0 REFERENCES==
==11.0 REFERENCES==
: 1. GE Nuclear Energy, "Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization," Li censing Topical Report, NEDC-32938P-A, Revision 2, May 2003; and NEDO-32938-A, Revision 2, May 2003. 2. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," (ELTR1), Licensing Topical Reports NEDC-32424P-A, February 1999; and NEDO-32424, April 1995. 3. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2), Licensing Topical Reports NEDC-32523P-A, February 2000; and NEDO-32523, April 1991; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999; and NEDO-32523, Supplement 1, January 1999. 4. NRC Regulatory Issue Summary 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002. 5. GE Nuclear Energy, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960-A and NEDO-31960-A Supplement 1, November 1995. 6. GE Nuclear Energy, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," NEDO-32465-A, August 1996. 7. GE Nuclear Energy, "Reactor Long-Term Stability Solution Option III: Licensing Basis Hot Channel Oscillation Magnitude for Limerick 1 and 2," GE-NE-0000-0035-6037-R1, February 2006. 8. OG 02-0119-260, "Backup Stability Protection (B SP) for Inoperable Option III Solution,"
: 1. GE Nuclear Energy, Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization, Licensing Topical Report, NEDC-32938P-A, Revision 2, May 2003; and NEDO-32938-A, Revision 2, May 2003.
July 17, 2002. 9. GE Nuclear Energy, "10 CFR 50 Appendix G E quivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vesse ls," NEDO-32205-A, Revision 1, February 1994. 10. Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988. 11. GE Nuclear Energy, "Pressure-Temperature Curves for PECO Energy Company Limerick Unit 1," GE-NE-B11-00836-00-01, Revision 0, April 2000 and GE-NE-B11-00836-00-01a NP, Revision 0, April 2000.
: 2. GE Nuclear Energy, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate, (ELTR1), Licensing Topical Reports NEDC-32424P-A, February 1999; and NEDO-32424, April 1995.
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 11-2 12. GE Nuclear Energy, "Pressure-Temperature Curves for PECO Energy Company Limerick Unit 2," GE-NE-B11-00836-00-02, Revision 0, July 2000 and GE-NE-B11-00836-00-02a NP, Revision 0, July 2000 13. BWRVIP-135, Revision 1, "BWR Vessel and In ternals Project Integr ated Surveillance Program (ISP) Data Source Book and Plant Evaluations," EPRI, Palo Alto, CA, June 2007 (TR-1013400). 14. C.I. Grimes (NRC) to Carl Terry (Niagara Mohawk Power Company), "Acceptance For Referencing Of EPRI Proprietary Report TR-113596, 'BWR Vessel And Internals Project, BWR Reactor Pressure Vessel Inspection And Flaw Evaluation Guidelines (BWRVIP-74)," and Appendix A, "Demonstration Of Compliance With the Technical Information Requirements Of The License Renewal Rule (10 CFR 54.21)," October 18, 2001. 15. NRC Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," November 1998. 16. GE Nuclear Energy, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1 Additional Information for Upper Bound PCT Calculation," NEDE-23785P-A, Supplement 1, Revision 1, March 2002. 17. GE Nuclear Energy, "Relaxation of UB PCT Limit for Limerick Units 1 and 2," GE-NE-0000-0010-5377-R0, December 2002. 18. GE Nuclear Energy, "Constant Pressure Power Uprate," Licensing Topical Report, NEDC-33004P-A, Revision 4, June 2003. 19. GE Nuclear Energy, "General Electric Instrument Setpoint Methodology," NEDC-31336P-A, September 1996 and NEDO-31336-A, September 1996. 20. GE Nuclear Energy, "Mitigation of BWR Core Thermal-Hydraulic Instabilities in ATWS," NEDO-32164, December 1992. 21. GE Nuclear Energy, "ATWS Rule Issues Relative to BWR Core Thermal-Hydraulics Stability," NEDO-32047-A, June 1995.}}
: 3. GE Nuclear Energy, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate, (ELTR2), Licensing Topical Reports NEDC-32523P-A, February 2000; and NEDO-32523, April 1991; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999; and NEDO-32523, Supplement 1, January 1999.
: 4. NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, January 31, 2002.
: 5. GE Nuclear Energy, BWR Owners Group Long-Term Stability Solutions Licensing Methodology, NEDO-31960-A and NEDO-31960-A Supplement 1, November 1995.
: 6. GE Nuclear Energy, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, NEDO-32465-A, August 1996.
: 7. GE Nuclear Energy, Reactor Long-Term Stability Solution Option III: Licensing Basis Hot Channel Oscillation Magnitude for Limerick 1 and 2, GE-NE-0000-0035-6037-R1, February 2006.
: 8. OG 02-0119-260, Backup Stability Protection (BSP) for Inoperable Option III Solution, July 17, 2002.
: 9. GE Nuclear Energy, 10 CFR 50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels, NEDO-32205-A, Revision 1, February 1994.
: 10. Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988.
: 11. GE Nuclear Energy, Pressure-Temperature Curves for PECO Energy Company Limerick Unit 1, GE-NE-B11-00836-00-01, Revision 0, April 2000 and GE-NE-B11-00836-00-01a NP, Revision 0, April 2000.
11-1
 
NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION
: 12. GE Nuclear Energy, Pressure-Temperature Curves for PECO Energy Company Limerick Unit 2, GE-NE-B11-00836-00-02, Revision 0, July 2000 and GE-NE-B11-00836-00-02a NP, Revision 0, July 2000
: 13. BWRVIP-135, Revision 1, BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations, EPRI, Palo Alto, CA, June 2007 (TR-1013400).
: 14. C.I. Grimes (NRC) to Carl Terry (Niagara Mohawk Power Company), Acceptance For Referencing Of EPRI Proprietary Report TR-113596, BWR Vessel And Internals Project, BWR Reactor Pressure Vessel Inspection And Flaw Evaluation Guidelines (BWRVIP-74),
and Appendix A, Demonstration Of Compliance With the Technical Information Requirements Of The License Renewal Rule (10 CFR 54.21), October 18, 2001.
: 15. NRC Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds, November 1998.
: 16. GE Nuclear Energy, GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1 Additional Information for Upper Bound PCT Calculation, NEDE-23785P-A, Supplement 1, Revision 1, March 2002.
: 17. GE Nuclear Energy, Relaxation of UB PCT Limit for Limerick Units 1 and 2, GE-NE-0000-0010-5377-R0, December 2002.
: 18. GE Nuclear Energy, Constant Pressure Power Uprate, Licensing Topical Report, NEDC-33004P-A, Revision 4, June 2003.
: 19. GE Nuclear Energy, General Electric Instrument Setpoint Methodology, NEDC-31336P-A, September 1996 and NEDO-31336-A, September 1996.
: 20. GE Nuclear Energy, Mitigation of BWR Core Thermal-Hydraulic Instabilities in ATWS, NEDO-32164, December 1992.
: 21. GE Nuclear Energy, ATWS Rule Issues Relative to BWR Core Thermal-Hydraulics Stability, NEDO-32047-A, June 1995.
11-2}}

Latest revision as of 18:57, 21 March 2020

NEDO-33484, Rev. 0, Safety Analysis Report for Limerick Generating Station Units 1 & 2 Thermal Power Optimization
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ATTACHMENT 8 GEH Nuclear Energy Safety Analysis Report for Limerick Generating Station, Units 1 and 2 Thermal Power Optimization, NEDO-33484 (Non-Proprietary Version)

HITACHI GE Hitachi Nuclear Energy NEDO-33484 Revision 0 Class I DRF 0000-0095-5957 March 2010 Non-Proprietary Information SAFETY ANALYSIS REPORT FOR LIMERICK GENERATING STATION UNITS 1 AND 2 THERMAL POWER OPTIMIZATION Copyright 2010 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33484P, Revision 0, from which the proprietary information has been removed. Portions of the document that have been removed are identified by white space within double square brackets, as shown here (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The design, engineering, and other information contained in this document is furnished for the purposes of supporting the Exelon Generation Company, LLC (Exelon) license amendment request for a thermal power uprate at Limerick Generating Station Units 1 and 2 to 3515 MWt in proceedings before the U.S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilities, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

No use of or right to copy any of this information contained in this document, other than by the NRC and its contractors in support of GEHs application, is authorized except by contract with GEH, as noted above. The information provided in this document is part of and dependent upon a larger set of knowledge, technology, and intellectual property rights pertaining to the design of standardized, nuclear powered, electric generating facilities. Without access and a GEH grant of rights to that larger set of knowledge, technology, and intellectual property rights, this document is not practically or rightfully usable by others, except by the NRC or through contractual agreements with Exelon, as set forth in the previous paragraph.

Copyright 2010 GE-Hitachi Nuclear Energy Americas LLC, All Rights Reserved ii

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION TABLE OF CONTENTS Page Acronyms and Abbreviations ..................................................................................................... xi Executive Summary ................................................................................................................... xvi 1.0 Introduction ...................................................................................................................... 1-1 1.1 Overview ....................................................................................................................... 1-1 1.2 Purpose and Approach................................................................................................... 1-2 1.2.1 TPO Analysis Basis ............................................................................................. 1-2 1.2.2 Margins ................................................................................................................ 1-3 1.2.3 Scope of Evaluations............................................................................................ 1-3 1.2.4 Exceptions to the TLTR....................................................................................... 1-5 1.2.5 Concurrent Changes Unrelated to TPO ............................................................... 1-5 1.3 TPO Plant Operating Conditions................................................................................... 1-5 1.3.1 Reactor Heat Balance........................................................................................... 1-5 1.3.2 Reactor Performance Improvement Features....................................................... 1-5 1.4 Basis for TPO Uprate .................................................................................................... 1-6 1.5 Summary and Conclusions ............................................................................................ 1-6 2.0 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.............................................................................................. 2-2 2.4 Thermal Hydraulic Stability .......................................................................................... 2-2 2.4.1 Stability Option III............................................................................................... 2-2 2.4.2 Stability Backup Stability Protection................................................................... 2-4 2.5 Reactivity Control ......................................................................................................... 2-4 3.0 Reactor Coolant and Connected Systems ...................................................................... 3-1 3.1 Nuclear System Pressure Relief / Overpressure Protection .......................................... 3-1 3.2 Reactor Vessel ............................................................................................................... 3-1 3.2.1 Fracture Toughness.............................................................................................. 3-1 3.2.2 Reactor Vessel Structural Evaluation .................................................................. 3-3 3.3 Reactor Internals............................................................................................................ 3-7 3.3.1 Reactor Internal Pressure Difference ................................................................... 3-7 3.3.2 Reactor Internals Structural Evaluation ............................................................... 3-8 3.3.3 Steam Separator and Dryer Performance........................................................... 3-10 3.4 Flow-Induced Vibration .............................................................................................. 3-10 3.5 Piping Evaluation ........................................................................................................ 3-11 iii

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.5.1 Reactor Coolant Pressure Boundary Piping....................................................... 3-11 3.5.2 Balance-of-Plant Piping Evaluation................................................................... 3-15 3.6 Reactor Recirculation System ..................................................................................... 3-16 3.7 Main Steam Line Flow Restrictors.............................................................................. 3-16 3.8 Main Steam Isolation Valves....................................................................................... 3-16 3.9 Reactor Core Isolation Cooling ................................................................................... 3-17 3.10 Residual Heat Removal System .................................................................................. 3-17 3.11 Reactor Water Cleanup System................................................................................... 3-17 4.0 Engineered Safety Features............................................................................................. 4-1 4.1 Containment System Performance ................................................................................ 4-1 4.1.1 Generic Letter 89-10 Program ............................................................................. 4-1 4.1.2 Generic Letter 95-07 Program ............................................................................. 4-2 4.1.3 Generic Letter 96-06 ............................................................................................ 4-2 4.1.4 Containment Coatings.......................................................................................... 4-2 4.2 Emergency Core Cooling Systems ................................................................................ 4-2 4.2.1 High Pressure Coolant Injection .......................................................................... 4-2 4.2.2 High Pressure Core Spray.................................................................................... 4-3 4.2.3 Core Spray ........................................................................................................... 4-3 4.2.4 Low Pressure Coolant Injection........................................................................... 4-3 4.2.5 Automatic Depressurization System.................................................................... 4-3 4.2.6 ECCS Net Positive Suction Head ........................................................................ 4-3 4.3 Emergency Core Cooling System Performance ............................................................ 4-4 4.4 Main Control Room Atmosphere Control System ........................................................ 4-4 4.5 Standby Gas Treatment System..................................................................................... 4-4 4.6 Main Steam Isolation Valve Leakage Alternate Drain Pathway................................... 4-4 4.7 Post-LOCA Combustible Gas Control System ............................................................. 4-5 5.0 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.......................................................................................... 5-2 5.2 BOP Monitoring and Control ........................................................................................ 5-2 5.2.1 Pressure Control System ...................................................................................... 5-2 5.2.2 EHC Turbine Control System.............................................................................. 5-3 5.2.3 Feedwater Control System................................................................................... 5-3 5.2.4 Leak Detection System ........................................................................................ 5-3 5.3 Technical Specification Instrument Setpoints ............................................................... 5-4 5.3.1 High-Pressure Scram ........................................................................................... 5-4 5.3.2 Hydraulic Pressure Scram.................................................................................... 5-4 5.3.3 High-Pressure Recirculation Pump Trip.............................................................. 5-5 5.3.4 Safety Relief Valve .............................................................................................. 5-5 5.3.5 Main Steam Line High Flow Isolation................................................................. 5-5 5.3.6 Fixed APRM Scram............................................................................................. 5-5 iv

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.3.7 APRM Flow-Biased Scram.................................................................................. 5-5 5.3.8 Rod Worth Minimizer Low Power Setpoint........................................................ 5-6 5.3.9 Rod Block Monitor .............................................................................................. 5-6 5.3.10 Flow-Biased Rod Block Monitor (%RTP) .......................................................... 5-6 5.3.11 Main Steam Line High Radiation Isolation ......................................................... 5-6 5.3.12 Low Steam Line Pressure MSIV Closure (RUN Mode) ..................................... 5-6 5.3.13 Reactor Water Level Instruments ........................................................................ 5-6 5.3.14 Main Steam Line Tunnel High Temperature Isolations ...................................... 5-7 5.3.15 Low Condenser Vacuum...................................................................................... 5-7 5.3.16 TSV Closure Scram, TCV Fast Closure Scram Bypasses ................................... 5-7 6.0 Electrical Power and Auxiliary Systems ........................................................................ 6-1 6.1 AC Power ...................................................................................................................... 6-1 6.1.1 Off-Site Power ..................................................................................................... 6-1 6.1.2 On-Site Power...................................................................................................... 6-2 6.2 DC Power ...................................................................................................................... 6-3 6.3 Fuel Pool........................................................................................................................ 6-3 6.3.1 Fuel Pool Cooling ................................................................................................ 6-3 6.3.2 Crud Activity and Corrosion Products................................................................. 6-3 6.3.3 Radiation Levels .................................................................................................. 6-4 6.3.4 Fuel Racks............................................................................................................ 6-4 6.4 Water Systems ............................................................................................................... 6-4 6.4.1 Service Water Systems ........................................................................................ 6-4 6.4.2 Main Condenser/Circulating Water/Normal Heat Sink Performance ................. 6-5 6.4.3 Reactor Enclosure Cooling Water System........................................................... 6-5 6.4.4 Turbine Enclosure Cooling Water System .......................................................... 6-6 6.4.5 Ultimate Heat Sink............................................................................................... 6-6 6.5 Standby Liquid Control System .................................................................................... 6-6 6.6 Power-dependent Heating, Ventilation and Air Conditioning ...................................... 6-7 6.7 Fire Protection ............................................................................................................... 6-7 6.7.1 10 CFR 50 Appendix R Fire Event...................................................................... 6-7 6.8 Systems Not Affected By TPO Uprate.......................................................................... 6-7 7.0 Power Conversion Systems.............................................................................................. 7-1 7.1 Turbine-Generator ......................................................................................................... 7-1 7.2 Condenser And Steam Jet Air Ejectors ......................................................................... 7-1 7.3 Turbine Steam Bypass................................................................................................... 7-2 7.4 Feedwater And Condensate Systems............................................................................. 7-2 7.4.1 Normal Operation ................................................................................................ 7-2 7.4.2 Transient Operation ............................................................................................. 7-3 7.4.3 Condensate Filters and Condensate Deep Bed Demineralizers........................... 7-3 8.0 Radwaste and Radiation Sources.................................................................................... 8-1 8.1 Liquid and Solid Waste Management ........................................................................... 8-1 8.2 Gaseous Waste Management......................................................................................... 8-1 v

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8.3 Radiation Sources in the Reactor Core.......................................................................... 8-2 8.4 Radiation Sources in Reactor Coolant........................................................................... 8-3 8.4.1 Coolant Activation Products ................................................................................ 8-3 8.4.2 Activated Corrosion Products .............................................................................. 8-3 8.4.3 Fission Products ................................................................................................... 8-4 8.5 Radiation Levels............................................................................................................ 8-4 8.6 Normal Operation Off-Site Doses ................................................................................. 8-5 9.0 Reactor Safety Performance Evaluations ...................................................................... 9-1 9.1 Anticipated Operational Occurrences............................................................................ 9-1 9.2 Design Basis Accidents ................................................................................................. 9-1 9.3 Special Events ............................................................................................................... 9-2 9.3.1 Anticipated Transient Without Scram ................................................................. 9-2 9.3.2 Station Blackout................................................................................................... 9-8 10.0 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 ............................................................................................ 10-1 10.2 Moderate Energy Line Break ...................................................................................... 10-2 10.3 Environmental Qualification ....................................................................................... 10-3 10.3.1 Electrical Equipment.......................................................................................... 10-3 10.3.2 Mechanical Equipment With Non-Metallic Components.................................. 10-4 10.3.3 Mechanical Component Design Qualification................................................... 10-4 10.4 Testing ......................................................................................................................... 10-4 10.5 Operator Training And Human Factors....................................................................... 10-5 10.6 Plant Life ..................................................................................................................... 10-5 10.7 NRC and Industry Communications ........................................................................... 10-6 10.8 Plant Procedures and Programs ................................................................................... 10-6 10.9 Emergency Operating Procedures ............................................................................... 10-6 10.10 Individual Plant Examination ...................................................................................... 10-6 11.0 References..................................................................................................................... 11-1 vi

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION LIST OF TABLES Table No. Title 1-1 Computer Codes Used for TPO Analyses 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions 1-3 Summary of Effect of TPO Uprate on Licensing Criteria 2-1 OPRM Setpoint Versus OLMCPR Demonstration 2-2 BSP Region Intercepts for Nominal Feedwater Temperature Demonstration 2-3 BSP Region Intercepts for Minimum Feedwater Temperature Demonstration 3-1 Limerick Unit 1 Upper Shelf Energy 40-Year License (32 EFPY) 3-2 Limerick Unit 2 Upper Shelf Energy 40-Year License (32 EFPY) 3-3 Limerick Unit 1 Adjusted Reference Temperatures 40-Year License (32 EFPY) 3-4 Limerick Unit 2 Adjusted Reference Temperatures 40-Year License (32 EFPY) 3-5 Limerick 32 EFPY Effects of Irradiation on RPV Axial Weld Properties 3-6 Limerick 32 EFPY Effects of Irradiation on RPV Circumferential Weld Properties 3-7 CUF and P+Q Stress Range of Limiting Components 3-8 Governing Stress Results for RPV Internal Components 3-9 Piping Lines Recommended for Special Focus under FAC Review 4-1 Limerick ECCS-LOCA Analysis Results for GE14 Fuel 5-1 Analytical Limits that Change due to TPO 6-1 TPO Plant Electrical Characteristics 6-2 Main Generator Ratings Comparison 6-3a Limerick 1 Generator Step-up Transformer Ratings Comparison 6-3b Limerick 2 Generator Step-up Transformer Ratings Comparison 6-4 Unit Auxiliary Transformer Ratings Comparison 6-5a Station Auxiliary Transformer Ratings Comparison 6-5b Regulating Transformer Ratings Comparison 6-6 Fuel Pool Cooling and Cleanup Parameters 6-7 Effluent Discharge Comparison 9-1 Key Inputs for ATWS Analysis vii

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table No. Title 9-2 Results for ATWS Analysis 9-3 Inputs for Three SLCS Pump Operation ATWS Analysis 9-4 Inputs for Two SLCS Pump Operation ATWS Analysis 9-5 Limerick SLCS Pressure Results for ATWS Analysis 9-6 MSIVC Sequence of Events 9-7 PRFO Sequence of Events 9-8 LOOP Sequence of Events viii

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION LIST OF FIGURES Figure No. Title 1-1 Power/Flow Map for the TPO (101.65% of CLTP) 1-2 Reactor Heat Balance - TPO Power (101.65% of CLTP), 100% Core Flow 2-1 Illustration of OPRM Trip-Enabled Region 2-2 BSP Regions for Nominal Feedwater Temperature Demonstration 2-3 BSP Regions for Minimum Feedwater Temperature Demonstration 9-1 TPO RTP MELLLA BOC MSIVC (Short Term) 9-2 TPO RTP MELLLA BOC MSIVC (Long Term - A) 9-3 TPO RTP MELLLA BOC MSIVC (Long Term - B) 9-4 TPO RTP MELLLA BOC MSIVC (Long Term - C) 9-5 TPO RTP MELLLA BOC PRFO (Short Term) 9-6 TPO RTP MELLLA BOC PRFO (Long Term - A) 9-7 TPO RTP MELLLA BOC PRFO (Long Term - B) 9-8 TPO RTP MELLLA BOC PRFO (Long Term - C) 9-9 TPO RTP MELLLA BOC LOOP (Short Term) 9-10 TPO RTP MELLLA BOC LOOP (Long Term - A) 9-11 TPO RTP MELLLA BOC LOOP (Long Term - B) 9-12 TPO RTP MELLLA BOC LOOP (Long Term - C) 9-13 TPO RTP MELLLA EOC MSIVC (Short Term) 9-14 TPO RTP MELLLA EOC MSIVC (Long Term - A) 9-15 TPO RTP MELLLA EOC MSIVC (Long Term - B) 9-16 TPO RTP MELLLA EOC MSIVC (Long Term - C) 9-17 TPO RTP MELLLA EOC PRFO (Short Term) 9-18 TPO RTP MELLLA EOC PRFO (Long Term - A) 9-19 TPO RTP MELLLA EOC PRFO (Long Term - B) 9-20 TPO RTP MELLLA EOC PRFO (Long Term - C) 9-21 TPO RTP MELLLA EOC LOOP (Short Term) 9-22 TPO RTP MELLLA EOC LOOP (Long Term - A) ix

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure No. Title 9-23 TPO RTP MELLLA EOC LOOP (Long Term - B) 9-24 TPO RTP MELLLA EOC LOOP (Long Term - C) x

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION ACRONYMS AND ABBREVIATIONS Term Definition ABA Amplitude Based Algorithm AC Alternating Current ADS Automatic Depressurization System AL Analytical Limit ALARA As Low As Reasonably Achievable AOO Anticipated Operational Occurrence AP Annulus Pressurization APRM Average Power Range Monitor ART Adjusted Reference Temperature ARTS Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program ASME American Society of Mechanical Engineers ATWS Anticipated Transient Without Scram AV Allowable Value B&PV Boiler and Pressure Vessel BHP Brake Horsepower BOP Balance of Plant BSP Backup Stability Protection BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Project CD Condensate Demineralizer CFR Code of Federal Regulations CLTP Current Licensed Thermal Power CRD Control Rod Drive CRGT Control Rod Guide Tube CS Core Spray CSC Containment Spray Cooling CSS Core Support Structure CUF Cumulative Usage Factor DBA Design Basis Accident DC Direct Current ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EFPY Effective Full Power Years EHC Electro-Hydraulic Control ELTR1 NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate xi

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term Definition ELTR2 NEDC-32523P-A, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate EOC End of Cycle EOOS Equipment Out-of-Service EOP Emergency Operating Procedure EPG Emergency Procedure Guidelines EPU Extended Power Uprate EQ Environmental Qualification Exelon Exelon Generation Company, LLC FAC Flow Accelerated Corrosion FFWTR Final Feedwater Temperature Reduction FIV Flow-Induced Vibration FPCC Fuel Pool Cooling And Cleanup FW Feedwater FWHOOS Feedwater Heater(s) Out-of-Service GDC General Design Criterion GE General Electric Company GEH GE Hitachi Nuclear Energy GL Generic Letter GRA Growth Rate Algorithm HELB High Energy Line Break HEPA High Efficiency Particulate Air HFCL High Flow Control Line HPCI High Pressure Coolant Injection HVAC Heating, Ventilation And Air Conditioning IASCC Irradiation Assisted Stress Corrosion Cracking ICF Increased Core Flow ICH&GT In-Core Housing and Guide Tube IPE Individual Plant Examination IRM Intermediate Range Monitor JR Jet Reaction ksi Kips Per Square Inch kV Kilovolt kW Kilowatt LBPCT Licensing Basis Peak Clad Temperature LCO Limiting Conditions For Operation LHGR Linear Heat Generation Rate Limerick Limerick Generating Station Units 1 and 2 LOCA Loss-of-Coolant-Accident xii

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term Definition LOOP Loss of Offsite Power LPCI Low Pressure Coolant Injection LPRM Local Power Range Monitor LPSP Low Power Setpoint MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCC Motor Control Circuit/Center MCPR Minimum Critical Power Ratio MELB Moderate Energy Line Break MELLLA Maximum Extended Load Line Limit Analysis MeV Million Electron Volts MFWT Minimum Feedwater Temperature Mlb Millions of Pounds MOV Motor Operated Valve MS Main Steam MSF Modified Shape Function MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSL Main Steam Line MSLB Main Steam Line Break MVA Million Volt Amps MWe Megawatt-Electric MWt Megawatt-Thermal NCL Natural Circulation Line NFWT Nominal Feedwater Temperature NPDES National Pollutant Discharge Elimination System NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NTSP Nominal Trip Setpoint NUREG Nuclear Regulations (NRC Document)

OLMCPR Operating Limit Minimum Critical Power Ratio OOS Out-of-Service P/F Power/Flow P-T Pressure-Temperature PBDA Period Based Detection Algorithm PCS Pressure Control System PCT Peak Clad Temperature PRA Probabilistic Risk Assessment xiii

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term Definition PRFO Pressure Regulator Failure Open - Maximum Steam Demand psi Pounds Per Square Inch psia Pounds Per Square Inch - Absolute psid Pounds Per Square Inch - Differential psig Pounds Per Square Inch - Gauge RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RG Regulatory Guide RHR Residual Heat Removal RIPD Reactor Internal Pressure Difference RIS Regulatory Issue Summary RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RTNDT Reference Temperature Of Nil-Ductility Transition RTP Rated Thermal Power RWCU Reactor Water Cleanup RWM Rod Worth Minimizer SAG Severe Accident Guidelines SAR Safety Analysis Report SBO Station Blackout SBPCS Steam Bypass Pressure Control System SDC Shutdown Cooling SER Safety Evaluation Report SFP Spent Fuel Pool SGTS Standby Gas Treatment System SJAE Steam Jet Air Ejector SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single (Recirculation) Loop Operation SPC Suppression Pool Cooling SR Surveillance Requirement SRM Source Range Monitor SRP Standard Review Plan SRV Safety Relief Valve SRVDL Safety Relief Valve Discharge Line TBV Turbine Bypass Valve TCV Turbine Control Valve xiv

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Term Definition TFSP Turbine First Stage Pressure T/G Turbine-Generator TIP Traversing In-Core Probe TLO Two (Recirculation) Loop Operation TLTP TPO Licensed Thermal Power TLTR NEDC-32938P-A, Thermal Power Optimization Licensing Topical Report TPO Thermal Power Optimization TSAR Thermal Power Optimization Safety Analysis Report TSV Turbine Stop Valve UBPCT Upper Bound Peak Clad Temperature UFM Ultrasonic Flow Measurement UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink USE Upper Shelf Energy VWO Valves Wide Open Wd Recirculation Drive Flow xv

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION EXECUTIVE

SUMMARY

This report summarizes the results of all significant safety evaluations performed that justify increasing the licensed thermal power at Limerick Generating Station Units 1 and 2 (Limerick) to 3515 MWt. The requested license power level is 1.65% above the current licensed thermal power (CLTP) level of 3458 MWt.

This report follows the Nuclear Regulatory Commission (NRC) approved format and content for Boiling Water Reactor (BWR) Thermal Power Optimization (TPO) licensing reports documented in NEDC-32938P-A, Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization, called TLTR. Per the outline of the TPO Safety Analysis Report (TSAR) in the TLTR Appendix A, every safety issue that should be addressed in a plant-specific TPO licensing report is addressed in this report. For issues that have been evaluated generically, this report references the appropriate evaluation and establishes that the evaluation is applicable to the plant.

Only previously NRC approved or industry-accepted methods were used for the analysis of accidents, transients, and special events. Therefore, because the safety analysis methods have been previously addressed, they are not addressed in this report. Also, event and analysis descriptions that are provided in other licensing documents or the Updated Final Safety Analysis Report (UFSAR) are not repeated. This report summarizes the results of the safety evaluations needed to justify a license amendment to allow for TPO operation.

The TLTR addresses power increases of up to 1.5% of CLTP, which will produce up to an approximately 2% increase in steam flow to the turbine-generator. The amount of power uprate

( 1.5%) contained in the TLTR was based on the expected reduction in power level uncertainty with the instrumentation technology available in 1999. The present instrumentation technology has evolved to where a power level uncertainty is reduced to as low as 0.3%, thereby supporting the evaluation of a power level increase of up to 1.7%. A higher steam flow is achieved by increasing the reactor power along the current rod and core flow control lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are recalibrated. Plant procedures are revised, and tests similar to some of the original startup tests are performed.

Evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, environmental issues, design basis accidents, and previous licensing evaluations were performed. This report demonstrates that Limerick can safely operate at a power level of 3515 MWt.

The following evaluations were conducted in accordance with the criteria of TLTR Appendix B:

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION All safety aspects of the plant that are affected by a 1.65% increase in the thermal power level were evaluated, including the Nuclear Steam Supply System (NSSS) and Balance-of-Plant (BOP) systems.

Evaluations and reviews were based on licensing criteria, codes, and standards applicable to the plant at the time of the TSAR submittal. There is no change in the previously established licensing basis for the plant, except for the increased power level.

Evaluations and/or analyses were performed using NRC-approved or industry-accepted analysis methods for the USAR accidents, transients, and special events affected by TPO.

Evaluations and reviews of the NSSS systems and components, containment structures, and BOP systems and components show continued compliance to the codes and standards applicable to the current plant licensing basis (i.e., no change to comply with more recent codes and standards is proposed due to TPO).

NSSS components and systems were reviewed to confirm that they continue to comply with the functional and regulatory requirements specified in the UFSAR and/or applicable reload license.

Any modification to safety-related or non-safety-related equipment will be implemented in accordance with 10 CFR 50.59.

All plant systems and components affected by an increased thermal power level were reviewed to ensure that there is no significant increase in challenges to the safety systems.

A review was performed to assure that the increased thermal power level continues to comply with the existing plant environmental regulations.

An assessment, as defined in 10 CFR 50.92(C), was performed to establish that no significant hazards consideration exists as a result of operation at the increased power level.

A review of the UFSAR and approved design changes ensures adequate evaluation of the licensing basis for the effect of TPO through the date of that evaluation.

The plant licensing requirements have been reviewed, and it is concluded that this TPO 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 TPO uprate does not involve a significant hazards consideration.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION

1.0 INTRODUCTION

1.1 OVERVIEW This document addresses a Thermal Power Optimization (TPO) power uprate of 1.65% of the current licensed thermal power (CLTP), consistent with the magnitude of the thermal power uncertainty reduction for the Limerick Generating Station Units 1 and 2 (Limerick) plant. This will result in an increase in licensed thermal power from 3458 MWt to 3515 MWt and an increase in electrical power from 1219.7 MWe to 1240 MWe.

This report follows the Nuclear Regulatory Commission (NRC)-approved format and content for Boiling Water Reactor (BWR) Thermal Power Optimization (TPO) licensing reports documented in NEDC-32938P-A, Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization (TLTR) (Reference 1). Power uprates in General Electric Company (GE) BWRs of up to 120% of original licensed thermal power (OLTP) are based on the generic guidelines and approach defined in the Safety Evaluation Reports provided in NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate, (ELTR1) (Reference 2) and NEDC-32523P-A, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate, (ELTR2)

(Reference 3). Since their NRC approval, numerous extended power uprate (EPU) submittals have been based on these reports. The outline for the TPO Safety Analysis Report (TSAR) in TLTR Appendix A follows the same pattern as that used for the EPUs. All of the issues that should be addressed in a plant-specific TPO licensing report are included in this TSAR. For issues that have been evaluated generically, this report references the appropriate evaluation and establishes that it is applicable to Limerick.

BWR plants, as currently licensed, have safety systems and component capability for operation at least 1.5% above the CLTP level. The amount of power uprate ( 1.5%) contained in the TLTR was based on the expected reduction in power level uncertainty with the instrumentation technology available in 1999. The present instrumentation technology has evolved to where a power level uncertainty is reduced to as low as 0.3%, thereby supporting the evaluation of a power level increase of up to 1.7%. Several Pressurized Water Reactor and BWR plants have already been authorized to increase their thermal power above the OLTP based on a reduction in the uncertainty in the determination of the power through improved feedwater (FW) flow rate measurements. When a previous uprate (other than a TPO) has been accomplished, the 102%

safety analysis basis is reestablished above the uprated power level. Therefore, all GE Hitachi Nuclear Energy (GEH) BWR plant designs have the capability to implement a TPO uprate, whether or not the plant has previously been uprated.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.2 PURPOSE AND APPROACH 1.2.1 TPO Analysis Basis Limerick was originally licensed at 3293 MWt. In amendments 106 and 51 for Units 1 and 2, respectively, the NRC approved a five percent power uprate to 3458 MWt, which is the CLTP.

The current safety analysis basis assumes, where required, that the reactor had been operating continuously at a power level at least 1.02 times the licensed power level. The analyses performed at 102% of CLTP remain applicable at the TPO rated thermal power (RTP), because the 2% factor from Regulatory Guide (RG) 1.49, Power Levels of Nuclear Power Plants, is effectively reduced by the improvement in the FW flow measurements. Some analyses may be performed at TPO RTP, because the uncertainty factor is accounted for in the methods, or the additional 2% margin is not required (e.g., Anticipated Transient Without Scram (ATWS)).

Detailed descriptions of the basis for the TPO analyses are provided in the subsequent sections of this report.

The TPO uprate is based on the evaluation of the improved FW flow rate measurement provided in Section 1.4. Figure 1-1 illustrates the TPO power/flow (P/F) operating map for the analysis at 101.65% of CLTP for Limerick. The changes to the P/F operating map are consistent with the generic descriptions given in TLTR Section 5.2. The approach to achieve a higher thermal power level is to increase core flow along the established Maximum Extended Load Line Limit Analysis (MELLLA) rod lines. This strategy allows Limerick to maintain most of the existing available core flow operational flexibility while assuring that low power related issues (e.g.,

stability and ATWS instability) do not change because of the TPO uprate.

No increase in the previously licensed maximum core flow limit is associated with the TPO uprate. When end of full power reactivity condition (all-rods-out) is reached, end-of-cycle coastdown may be used to extend the power generation period. Previously licensed performance improvement features are presented in Section 1.3.2.

With respect to absolute thermal power and flow, there is no change in the extent of the Single (Recirculation) Loop Operation (SLO) domain as a result of the TPO uprate. Therefore, the SLO domain is not provided. For Limerick, the maximum analyzed reactor core thermal power for SLO operation remains at 2852.9 MWt. This value bounds the Technical Specification Limit of 2635 MWt.

The TPO uprate is accomplished with no increase in the nominal vessel dome pressure. This minimizes the effect of uprating on reactor thermal duty, evaluations of environmental conditions, and minimizes changes to instrument setpoints related to system pressure, etc.

Satisfactory reactor pressure control capability is maintained by evaluating the steam flow margin available at the turbine inlet. This operational aspect of the TPO uprate will be demonstrated by performing controller testing as described in Section 10.4. The TPO uprate does not affect the pressure control function of the turbine bypass valves (TBVs).

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.2.2 Margins The TPO analysis basis ensures that the power-dependent instrument error margin identified in RG 1.49 is maintained. NRC-approved or industry-accepted computer codes and calculation techniques are used in the safety analyses for the TPO uprate. A list of the NSSS computer codes used in the evaluations is provided in Table 1-1. Computer codes used in previous analyses (i.e., analyses at 102% of CLTP) are not listed. Similarly, factors and margins specified by the application of design code rules are maintained, as are other margin-assuring acceptance criteria used to judge the acceptability of the plant.

1.2.3 Scope of Evaluations The scope of evaluations is discussed in TLTR Appendix B. Tables B-1 through B-3 illustrate those analyses that are bounded by current analyses, those that are not significantly affected, and those that require updating. The disposition of the evaluations as defined by Tables B-1 through B-3 is applicable to Limerick. This TSAR includes all of the evaluations for the plant-specific application. Many of the evaluations are supported by generic reference, some supported by rational considerations of the process differences, and some plant-specific analyses are provided.

The scope of the evaluations are summarized in the following sections:

2.0 Reactor Core and Fuel Performance Overall heat balance and power-flow operating map information is provided. Key core performance parameters are confirmed for each fuel cycle, and will continue to be evaluated and documented for each fuel cycle.

3.0 Reactor Coolant and Connected Systems Evaluations of the NSSS components and systems are performed at the TPO conditions. These evaluations confirm the acceptability of the TPO changes in process variables in the NSSS.

4.0 Engineered Safety Features The effects of TPO changes on the containment, Emergency Core Cooling Systems (ECCS),

Standby Gas Treatment, and other Engineered Safety Features are evaluated for key events. The evaluations include the containment responses during limiting abnormal events, Loss-of-Coolant-Accident (LOCA), and safety relief valve (SRV) containment dynamic loads.

5.0 Instrumentation and Control The instrumentation and control signal ranges and analytical limits for setpoints are evaluated to establish the effects of TPO changes in process parameters. If required, analyses are performed to determine the need for setpoint changes for various functions. In general, setpoints are 1-3

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION changed only to maintain adequate operating margins between plant operating parameters and trip values.

6.0 Electrical Power and Auxiliary Systems Evaluations are 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 TPO RTP level.

7.0 Power Conversion Systems Evaluations are 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 TPO power output.

8.0 Radwaste and Radiation Sources The liquid and gaseous waste management systems are evaluated at TPO conditions to show that applicable release limits continue to be met during operation at the TPO RTP level. The radiological consequences are evaluated to show that applicable regulations are met for TPO including the effect on source terms, on-site doses, and off-site doses during normal operation.

9.0 Reactor Safety Performance Evaluations

((

)) The standard reload analyses consider the plant conditions for the cycle of interest.

10.0 Other Evaluations High energy line break (HELB) and environmental qualification evaluations are performed at bounding conditions for the TPO range to show the continued operability of plant equipment under TPO conditions. The Individual Plant Examination (IPE) Probabilistic Risk Assessment (PRA) will not be updated, because the change in plant risk from the subject power uprate is insignificant. This conclusion is supported by NRC Regulatory Issue Summary (RIS) 2002-03 (Reference 4). In response to feedback received during the public workshop held on August 23, 2001, the Staff wrote, The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant effect on plant risk. Therefore, no risk information is requested to support such applications.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.2.4 Exceptions to the TLTR One safety-related modification to the Standby Liquid Control System (SLCS) is proposed in this amendment request. Although the TLTR Section B.2, Licensing Criteria, states that no safety-related modifications are needed beyond potential setpoint changes, the need for this modification is discussed in separate documentation and in Section 9.3.1.5 of this report.

1.2.5 Concurrent Changes Unrelated to TPO No concurrent changes unrelated to TPO are included in this evaluation.

1.3 TPO PLANT OPERATING CONDITIONS 1.3.1 Reactor Heat Balance The reactor heat balance diagram at the TPO conditions is presented in Figure 1-2.

The small changes in thermal-hydraulic parameters for the TPO are illustrated in Figure 1-2 (3515 MWt, 100% Core Flow). These parameters are generated for TPO by performing coordinated reactor and turbine-generator heat balances that relate the reactor thermal-hydraulic parameters to the increased plant FW and steam flow conditions. Input from Limerick operation is considered (e.g., steam line pressure drop) to match expected TPO uprate conditions.

1.3.2 Reactor Performance Improvement Features The following performance improvement and equipment out-of-service (EOOS) features currently licensed at Limerick are acceptable at the TPO RTP level. Their inclusion in this analysis is appropriate, as they have been previously adopted by Limerick and incorporated in the limiting case of the initial power uprate analysis.

Performance Improvement Feature SLO Increased Core Flow (ICF) (110.0% of rated)

Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program (ARTS) / MELLLA (82.9% of Rated Core Flow at TPO RTP)

Final Feedwater Temperature Reduction (FFWTR), -105ºF Feedwater Heater (s) OOS, -60ºF SRV OOS, two valves 24 Month Cycle Recirculation Pump Trip (RPT) OOS TBV OOS Main Steam Isolation Valves (MSIV) OOS, 75% Power Turbine Control Valve (TCV) Stuck Closed Turbine Stop Valve (TSV) Stuck Closed 1-5

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1.4 BASIS FOR TPO UPRATE The safety analyses in this report are based on a total thermal power measurement uncertainty of 0.3%. This will bound the actual power level requested. The detailed basis value is provided in separate documentation, which addresses the improved FW flow measurement accuracy using the Caldon Leading Edge Flow Meter Check-Plus system.

1.5

SUMMARY

AND CONCLUSIONS This evaluation has investigated a TPO uprate to 101.65% of CLTP. The strategy for achieving higher power is to increase core flow along the established MELLLA rod lines. The plant licensing challenges have been reviewed (Table 1-3) to demonstrate how the TPO 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 TPO uprate described herein involves no significant hazards consideration.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-1 Computer Codes For TPO Analyses Computer Version or NRC Task Comments Code Revision Approved Reactor Heat Balance ISCOR 09 Y (1) NEDE-24011-P Rev 0 Safety Evaluation Report (SER)

Thermal-hydraulic Stability ISCOR 09 Y (1) NEDE-24011P Rev. 0 SER PANACEA 11 Y NEDE-30130-P-A (2)

ODYSY 05 Y NEDC-32992P-A OPRM 01 Y(3) NEDO-32465-A TRACG 04 N(4) NEDO-32465-A Reactor Internal Pressure ISCOR 09 Y (1) NEDE-24011P Rev. 0 SER Differences Anticipated Transient Without ODYN 10 Y NEDE-24154P-A Supp. 1, Vol. 4 Scram STEMP 04 (5) NEDE-30130-P-A PANACEA 11 Y (2)

ISCOR 09 Y (1) NEDE-24011-P Revision 0 SER Notes For Table 1-1:

(1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P 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 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.

(2) The physics code PANACEA provides inputs to the transient code ODYN. The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-24011-P-A). The use of TGBLA Version 06 and PANACEA Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE)

Subject:

"Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO. MA6481), November 10, 1999.

(3) The OPRM code is not Level 2. However, the methodology, as implemented in the OPRM code, has been approved by the NRC.

(4) TRACG02 has been approved in NEDO-32465-A by the US NRC for the stability DIVOM analysis. The CLTP stability analysis is based on TRACG04, which has been shown to provide essentially the same or more conservative results in DIVOM applications as the previous version, TRACG02.

(5) 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.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions TPO RTP Parameter CLTP (101.65% of CLTP)

Thermal Power (MWt) 3458.0 3515.0 (Percent of Current Licensed Power) 100.0 101.65 Steam Flow (Mlb/hr) 14.997 15.287 (Percent of Current Rated) 100.0 101.9 FW Flow (Mlb/hr) 14.965 15.255 (Percent of Current Rated) 100.0 101.9 Dome Pressure (psia) 1060 1060 Dome Temperature (°F) 551.5 551.5 FW Temperature (°F) 425.1 427.1 Full Power Core Flow Range (Mlb/hr) 81.0 to 110.0 82.9 to 110.0 (Percent of Current Rated) 81.0 to 110.0 82.9 to 110.0 1-8

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-3 Summary of Effect of TPO Uprate on Licensing Criteria Effect of 1.7%

Key Licensing Criteria Explanation of Effect Thermal Power Increase LOCA challenges to fuel No increase in peak clad temperature Previous analysis accounted for 102% of (10 CFR 50, Appendix K) (PCT), no change of maximum Linear licensed power, bounding TPO operation. No Heat Generation Rate (LHGR) vessel pressure increase.

required.

Change of Operating Limit < 0.01 increase. Minor increase (< 0.01) due to slightly higher Minimum Critical Power Ratio power density and increased MCPR safety limit (MCPR) (slightly flatter radial power distribution).

Challenges to reactor pressure No increase in peak pressure. No increase because previous analysis accounted vessel (RPV) overpressure for 102% overpower, bounding TPO operation.

Primary containment pressure No increase in peak containment Previous analysis accounted for 102%

during a LOCA pressure. overpower, bounding TPO operation. No vessel pressure increase. No increase in energy to the pool.

Pool temperature during a No increase in peak pool temperature. Previous analysis accounted for 102%

LOCA overpower, bounding TPO operation. No vessel pressure increase. No increase in energy to the pool.

Offsite Radiation Release, No increase (remains within Previous analysis bounds TPO operation. No design basis accidents 10 CFR 100). vessel pressure increase.

Onsite Radiation Dose, normal Approximately 1.7% increase, must Slightly higher inventory of radionuclides in operation remain within 10 CFR 20. steam/FW flow paths.

Heat discharge to environment < 1°F temperature increase. Small % power increase.

Equipment Qualification Remains within current pressure, No change in Harsh Environment terms (TPO radiation, and temperature envelopes. operating conditions bounded by previous analyses); minimal change in normal operating conditions.

Fracture Toughness, < 2°F increase in Reference Small increase in neutron fluence.

10 CFR 50, Appendix G Temperature of the Nil-Ductility Transition (RTNDT).

Stability No direct effect of TPO uprate because No increase in maximum rod line boundary.

applicable stability regions and lines Characteristics of each reload core continue to be are extended beyond the absolute evaluated as required for each stability option.

values associated with the current boundaries to preserve MWt-core flow boundaries as applicable for each stability option.

ATWS peak vessel pressure Slight increase (15 psig). Slightly increased power relative to SRV capacity.

Vessel and NSSS equipment No change. Comply with existing ASME Code stress limits of design pressure all categories.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-1 Power/Flow Map for TPO (101.65% of CLTP)

Core Flow (Mlbm/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 4400 120 A: Natural Circulation B: Two Pump Minimum Speed C: 67.6% Power / 44.4% Flow 4000 110 D: 100.0% Power / 82.9% Flow D': 98.4% Power / 80.8% Flow E: 100.0% Power / 100.0% Flow D E F 3515 MWt 3600 100 E': 98.4% Power / 100.0% Flow MELLLA Boundary F: 100.0% Power / 110.0% Flow 2 3458 MWt P=(22.191+0.89714WT -0.0011905WT )(1.132) D' E' F' F': 98.4% Power / 110.0% Flow 90 G: 23.6% Power / 110.0% Flow 3200 H: 23.6% Power / 100.0% Flow Thermal Power (MWt)

I: 23.6% Power / 38.0% Flow Thermal Power (%TPO RTP) 80 2800 70 C 2400 Increased Core Flow 60 2000 B

50 A 1600 40 1200 30 Cavitation Interlock 800 20 I H G 100.0% TPO RTP = 3515 MWt 10 100.0% CLTP = 3458 MWt 400 100.0% Core Flow = 100.0 Mlbm/hr 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-2 Reactor Heat Balance - TPO Power (101.65% of CLTP), 100% Core Flow Legend

  1. = Flow, lbm/hr 1060 H = Enthalpy, Btu/lbm P F = Temperature, °F M = Moisture, % Main Steam Flow 15.287E+06 #
  • P = Pressure, psia 1190.0 H
  • 0.43 M
  • Carryunder = 0.25% 1003 P
  • 3515 Main Feed Flow MWt Wd = 100 % 15.388E+06 # 15.255E+06 #

531.5 H 405.5 H 405.3 H 536.0 °F Total 427.2 °F 427.1 °F Core Flow 100.0E+06 h = 0.9 H #

1.330E+05 #

418.4 H 530.7 439.0 °F H

Cleanup Demineralizer System 3.200E+04 # Control Rod Drive 1.330E+05 #

68.0 H Feed Flow 530.6 H 97.2 °F 535.3 °F

  • Conditions at upstream side of TSV Core Thermal Power 3515.0 Pump Heating 9.0 Cleanup Losses -4.4 Other System Losses -1.1 Turbine Cycle Use 3518.5 MWt 1-11

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2.0 REACTOR CORE AND FUEL PERFORMANCE 2.1 FUEL DESIGN AND OPERATION At the TPO RTP conditions, all fuel and core design limits are met by the deployment of fuel enrichment and burnable poison, control rod pattern management, and core flow adjustments.

New fuel designs are not needed for the TPO to ensure safety. However, revised loading patterns, slightly larger batch sizes, and potentially new fuel designs may be used to provide additional operating flexibility and maintain fuel cycle length. NRC approved limits for burnup on the fuel are not exceeded. Therefore, the reactor core and fuel design is adequate for TPO operation.

The initial TPO cycle at Limerick Unit 1 will be loaded with fresh and previously irradiated GE14 fuel assemblies. The initial TPO cycle at Limerick Unit 2 will be loaded with fresh GNF2 fuel assemblies and previously irradiated GE14 fuel assemblies.

2.2 THERMAL LIMITS ASSESSMENT Operating thermal limits ensure that regulatory and/or safety limits are not exceeded for a range of postulated events (e.g., transients, LOCA). This section addresses the effects of TPO on thermal limits. Cycle-specific core configurations, which are evaluated for each reload, confirm TPO RTP capability and establish or confirm cycle-specific limits.

The historical 25% of RTP value for the Technical Specification Safety Limit, some thermal limits monitoring Limiting Conditions for Operation (LCOs) thresholds, and some Surveillance Requirements (SRs) thresholds is based on ((

)) The historical 25% RTP value is a conservative basis, as described in the plant Technical Specifications, ((

)) Therefore, the Safety Limit percent RTP basis, some thermal limits monitoring LCOs, and SR percent RTP thresholds remain at 25% RTP for the TPO uprate.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 2.2.1 Safety Limit MCPR The Safety Limit Minimum Critical Power Ratio (SLMCPR) is dependent upon the nominal average power level and the uncertainty in its measurement. Consistent with approved practice, a revised SLMCPR is calculated for the first TPO fuel cycle and confirmed for each subsequent cycle. The historical uncertainty allowance and calculational methods are discussed in TLTR Section 5.7.2.1.

2.2.2 MCPR Operating Limit TLTR Appendix E shows that the changes in the Operating Limit Minimum Critical Power Ratio (OLMCPR) for a TPO uprate ((

)) Because the cycle-specific SLMCPR is also defined, the actual required OLMCPR can be established. This ensures an adequate fuel thermal margin for TPO uprate operation.

2.2.3 MAPLHGR and Maximum LHGR Operating Limits The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and maximum Linear Heat Generation Rate (LHGR) limits are maintained as described in TLTR Section 5.7.2.2. No significant change results due to TPO operation. The LHGR limits are fuel dependent and are not affected by the TPO. The ECCS performance is addressed in Section 4.3.

2.3 REACTIVITY CHARACTERISTICS All minimum shutdown margin requirements apply to cold shutdown conditions and are maintained without change. Checks of cold shutdown margin based on SLCS boron injection capability and shutdown using control rods with the most reactive control rod stuck out are made for each reload. The TPO uprate has no significant effect on these conditions; the shutdown margin is confirmed in the reload core design.

Operation at the TPO RTP could result in a minor decrease in the hot excess reactivity during the cycle. This loss of reactivity does not affect safety and does not affect the ability to manage the power distribution through the cycle to achieve the target power level. However, the lower hot excess reactivity can result in achieving an earlier all-rods-out condition. Through fuel cycle redesign, sufficient excess reactivity can be obtained to match the desired cycle length.

2.4 THERMAL HYDRAULIC STABILITY 2.4.1 Stability Option III Limerick Units 1 and 2 have implemented the long-term stability solution Option III (References 5 and 6). The Option III solution combines closely spaced Local Power Range Monitor (LPRM) detectors into cells to effectively detect either core-wide or regional (local) 2-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION modes of reactor instability. These cells are termed Oscillation Power Range Monitor (OPRM) cells and are configured to provide local area coverage with multiple channels. Plants implementing Option III have hardware to combine the LPRM signals and to evaluate the cell signals with instability detection algorithms. The Period Based Detection Algorithm (PBDA) is the only algorithm credited in the Option III licensing basis (Reference 6). Two defense-in-depth algorithms, referred to as Amplitude Based Algorithm (ABA) and the Growth Rate Algorithm (GRA), offer a high degree of assurance that fuel failure will not occur as a consequence of stability related oscillations. Because the OPRM hardware does not change, the hot channel oscillation magnitude (HCOM) portion of the Option III calculation (Reference 7) is not affected by TPO and does not need to be recalculated.

The Option III Trip-Enabled Region has been generically defined as the region ( 60% rated core flow and 30% rated power) where the OPRM system is fully armed. For TPO, the Option III Trip-Enabled Region is rescaled to maintain the same absolute P/F region boundaries. The Backup Stability Protection (BSP) evaluation described in Section 2.4.2 shows that the generic Option III Trip-Enabled Region is adequate. The Trip-Enabled Region is shown in Figure 2-1.

Because the rated core flow is not changed, the 60% recirculation drive flow boundary is not rescaled (It should be noted that 60% recirculation drive flow bounds 60% core flow). The 30%

of CLTP boundary changes by the following equation:

TPO Region Boundary = 30% CLTP * (100% / TPO (% CLTP))

Thus, for a 101.65% of CLTP TPO:

TPO Region boundary = 30% CLTP * (100% / 101.65%) = 29.5% TPO The minimum power at which the OPRM should be confirmed operable is 24.5% TPO. A 5%

absolute power separation between the OPRM Trip-Enabled Region power boundary and the power, at which the OPRM system should be confirmed operable, is deemed adequate for the Option III application.

Stability Option III provides SLMCPR protection by generating a reactor scram if a reactor instability, which exceeds the specified trip setpoint, is detected. The demonstration setpoint is determined per the current NRC approved methodology. The Option III stability reload licensing basis calculates the limiting OLMCPR required to protect the SLMCPR for both steady-state and transient stability events as specified in the Option III methodology (Reference 6). These OLMCPRs are calculated for a range of OPRM setpoints for TPO operation. Selection of an appropriate instrument setpoint is then based upon the OLMCPR required to provide adequate SLMCPR protection. This determination relies on the DIVOM curve (Delta CPR Over Initial MCPR Versus Oscillation Magnitude) to determine an OPRM setpoint that protects the SLMCPR during an anticipated instability event. A DIVOM analysis is performed and used in Option III OPRM setpoint demonstration.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION As shown in Table 2-1, with an estimated OLMCPR of 1.34 and an estimated SLMCPR of 1.07, an OPRM trip setpoint of 1.14 with a successive confirmation count of 16 (Reference 6) is the highest setpoint that may be used without stability setting the OLMCPR. The actual setpoint will be established in accordance with Limerick Units 1 and 2 Technical Specifications at each reload. These demonstration results are based on a power level of 101.7% CLTP, which is bounding for a power level of 101.65%.

Therefore, TPO operation is justified for plant operation with stability Option III.

2.4.2 Stability Backup Stability Protection Limerick Units 1 and 2 have implemented the Backup Stability Protection (BSP) regions (Reference 8) as the stability backup solution if the OPRM system is declared inoperable.

The BSP regions consist of two regions, I-Scram and II-Controlled Entry. The Base BSP Scram Region and Base BSP Controlled Entry Region are defined by state points on the High Flow Control Line (HFCL) and on the Natural Circulation Line (NCL) in accordance with Reference 8. The bounding plant-specific BSP region state points must enclose the corresponding Base BSP region state points on the HFCL and on the NCL. If a calculated BSP region state point is located inside the corresponding base BSP region state point, then it must be replaced by the corresponding base BSP region state point. If a calculated BSP region state point is located outside the corresponding base BSP region state point, this point is acceptable for use.

That is, the selected points will result in the largest, or most conservative, region sizes. The proposed BSP Scram and Controlled Entry region boundaries are constructed by connecting the corresponding bounding state points on the HFCL and the NCL using a shape function. The Modified Shape Function (MSF) is demonstrated.

The demonstration BSP regions for both Nominal Feedwater Temperature (NFWT) and Minimum Feedwater Temperature (MFWT) operations are shown in Table 2-2/Figure 2-2 and Table 2-3/Figure 2-3, respectively. The OPRM Trip-Enabled Region is confirmed for NFWT operations based on the demonstration BSP regions for NFWT. The demonstration BSP regions for MFWT confirm the OPRM Trip-Enabled Region for operations on or below the 97% rod line. These demonstration results are based on a power level of 101.7% CLTP, which is bounding for a power level of 101.65%.

The BSP regions are confirmed or expanded on a cycle-specific basis.

Therefore, TPO operation is justified for plant operation with stability BSP regions.

2.5 REACTIVITY CONTROL The generic discussion in TLTR Sections 5.6.3 and Appendix J.2.3.3 applies to the Limerick plant. The Control Rod Drive (CRD) and CRD hydraulic systems and supporting equipment are not affected by the TPO uprate and no further evaluation of CRD performance is necessary.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 2-1 OPRM Setpoint Versus OLMCPR Demonstration Limerick 1&2 TPO OPRM Set Point OLMCPR (2RPT) OLMCPR (SS) 1.05 1.144 1.160 1.06 1.163 1.179 1.07 1.183 1.199 1.08 1.204 1.220 1.09 1.225 1.242 1.10 1.247 1.264 1.11 1.268 1.285 1.12 1.290 1.307 1.13 1.312 1.330 1.14 1.335 1.354 1.15 1.359 1.378 1.16 1.384 1.403 1.17 1.408 1.428 1.18 1.434 1.454 1.19 1.461 1.481 1.20 1.489 1.509 Rated Power Off-rated OLMCPR Acceptance Criteria OLMCPR At 45% Flow 2-5

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 2-2 BSP Region Intercepts for Nominal Feedwater Temperature Demonstration Region Boundary Intercept  % TPO Power  % Core Flow Scram Region (Region I) Boundary Intercept on HCFL A1 71.1 48.5 Scram Region (Region I) Boundary Intercept on NCL B1 Base 52.2 39.3 Controlled Entry Region (Region II) Boundary Intercept on HCFL A2 72.7 50.2 Controlled Entry Region (Region II) Boundary Intercept on NCL B2 Base 36.0 38.6 2-6

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 2-3 BSP Region Intercepts for Minimum Feedwater Temperature Demonstration Region Boundary Intercept  % TPO Power  % Core Flow Scram Region (Region I) Boundary Intercept on HCFL A1 87.3 67.2 Scram Region (Region I) Boundary Intercept on NCL B1 45.5 39.1 Controlled Entry Region (Region II) Boundary Intercept on HCFL A2 88.4 68.6 Controlled Entry Region (Region II) Boundary Intercept on NCL B2 Base 36.0 38.6 2-7

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 2-1 Illustration of OPRM Trip-Enabled Region Limerick TPO OPRM Trip-Enabled Region Core Flow (Mlbm/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 4400 120 4000 110 3600 100 90 3200 Thermal Power (% TLTP) Thermal Power (MWt) 80 2800 70 2400 60 OPRM Trip- 2000 50 Enabled Region 1600 40 1200 30 800 20 10 400 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 2-2 BSP Regions for Nominal Feedwater Temperature Demonstration Limerick TPO NFWT Proposed BSP Regions Core Flow (Mlbm/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 4400 120 Scram Region 4000 110 Controlled Entry Region BSP Endpoints 3600 100 90 3200 Thermal Power (% TLTP) Thermal Power (MWt) 80 2800 A2 A1 70 2400 60 2000 B1 Base 50 1600 40 B2 Base 1200 30 800 20 10 400 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Figure 2-3 BSP Regions for Minimum Feedwater Temperature Demonstration Limerick TPO MFWT Proposed BSP Regions Core Flow (Mlbm/hr) 0 10 20 30 40 50 60 70 80 90 100 110 120 130 4400 120 Scram Region 4000 110 Controlled Entry Region BSP Endpoints 3600 100 A2 3200 90 A1 80 2800 Thermal Power (% TLTP) Thermal Power (MWt) 70 2400 60 2000 50 B1 1600 40 B2 Base 1200 30 800 20 10 400 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.0 REACTOR COOLANT AND CONNECTED SYSTEMS 3.1 NUCLEAR SYSTEM PRESSURE RELIEF / OVERPRESSURE PROTECTION The pressure relief system prevents over-pressurization of the nuclear system during abnormal operational transients. The SRVs along with other functions provide this protection.

Evaluations and analyses for the CLTP have been performed at 102% of CLTP to demonstrate that the reactor vessel conformed to ASME Boiler and Pressure Vessel (B&PV) Code and plant Technical Specification requirements. There is no increase in nominal operating pressure for the Limerick TPO uprate. There are no changes in the SRV setpoints or valve OOS options. There is no change in the methodology or the limiting overpressure event. Therefore, the generic evaluation contained in the TLTR is applicable.

The analysis for each fuel reload, which is current practice, confirms the capability of the system to meet the ASME design criteria.

3.2 REACTOR VESSEL The RPV structure and support components form a pressure boundary to contain 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.

3.2.1 Fracture Toughness The TLTR, Section 5.5.1.5, describes the RPV fracture toughness evaluation process. RPV embrittlement is caused by neutron exposure of the wall adjacent to the core including the regions above and below the core that experience fluence 1.0E+17 n/cm2. This region is defined as the beltline region. Operation at TPO conditions results in a higher neutron flux, which increases the integrated fluence over the period of plant life. Limerick Units 1 and 2 are evaluated for a fluence that bounds the required value for operation at TPO conditions.

The neutron fluence for TPO is calculated using two-dimensional neutron transport theory. The neutron transport methodology is consistent with RG 1.190. A bounding peak fluence, 1.9E+18 n/cm2, is used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G (Reference 9). The results of these evaluations indicate that:

(a) The upper shelf energy (USE) will remain > 50 ft-lb for the design life of the vessel or maintain the margin requirements of 10 CFR 50, Appendix G as defined in RG 1.99 (Reference 10). The minimum USE for the Unit 1 beltline materials is 24 ft-lb for 32 effective full power years (EFPY) and for the Unit 2 beltline materials is 25 ft-lbs for 32 EFPY. Many of the Limerick RPV materials do not have sufficient unirradiated USE data, and Charpy data from low temperature tests were used to develop an initial USE. Therefore, Equivalent Margin Analyses were performed for the limiting beltline plate, weld, and nozzle forging 3-1

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION materials to assure qualification. These values are provided in Tables 3-1 and 3-2 for Limerick Units 1 and 2, respectively.

(b) The beltline material RTNDT remains below the 200°F screening criteria as defined in Reference 10. These values are provided in Tables 3-3 and 3-4 for Limerick Units 1 and 2, respectively.

(c) The CLTP Pressure-Temperature (P-T) curves (References 11 and 12) remain bounding for TPO, limited to the currently approved fluence. The current Adjusted Reference Temperature (ART) values for the beltline plates and welds remain bounding for TPO. The currently licensed P-T curves include the Low Pressure Coolant Injection (LPCI) nozzle. The water level instrumentation nozzle that occurs within the beltline region is bounded by the CLTP curves.

(d) The surveillance program consists of three capsules in each vessel. No capsules have been removed from either vessel. These three capsules have been in each reactor vessel since plant startup. Limerick is a participant in the Integrated Surveillance Program (Reference 13),

currently administrated by EPRI, and is not designated as a representative plant; therefore, no capsules are slated for removal at this time. TPO has no effect on the existing surveillance schedule.

(e) The 32 EFPY beltline axial and circumferential weld material RTNDT remains bounded by the requirements of Boiling Water Reactor Vessel and Internals Project (BWRVIP)-05 as defined in References 14 and 15. This comparison is provided in Tables 3-5 and 3-6 for axial and circumferential welds, respectively.

The maximum normal operating dome pressure for TPO is unchanged from that for CLTP power operation. Therefore, the hydrostatic and leakage test pressures and associated temperatures are acceptable for the TPO. Because the vessel is still in compliance with the regulatory requirements as demonstrated above, operation with TPO does not have an adverse effect (not exceeding regulatory requirements) on the reactor vessel fracture toughness.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.2.2 Reactor Vessel Structural Evaluation

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High and low pressure seal leak detection nozzles were not considered to be pressure boundary components at the time that the OLTP evaluation was performed, and have not been evaluated for TPO.

The effect of TPO 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, 1968 Edition with addenda to and including Summer 1969 (except that Figure N-462(e)(2) of the Summer 1970 Addenda were applied) was used as the governing code and is considered the Code of Construction. However, if a components design has been modified, the governing code for that component was the code used in the stress analysis of the modified component. The following components were modified since the original construction of Limerick Units 1 and 2:

  • FW Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1976.
  • LPCI Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1975 and 1968 Edition with Addenda to and including Winter 1969.
  • CRD Hydraulic System Return Nozzle (Unit 2): This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1976.
  • CS Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1975 and 1968 Edition with Addenda to and including Winter 1969.
  • Recirculation Inlet Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1976 and 1968 Edition with Addenda to and including Summer 1969.
  • Universal Dry Tube, Power Range Detector, and In-Core Detector Assembly: These components were modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code,Section III, 1968 Edition with Addenda to and including Winter 1969, 1971 Edition with Addenda to and including Summer 1973, and 3-6

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 1977 Edition with Addenda to and including Summer 1977 (varies depending upon specific component).

Typically, new stresses are determined by scaling the original stresses based on the TPO conditions (pressure, temperature, and flow). The bounding analyses were performed for the design, normal & upset, and emergency & faulted conditions. If there is an increase in annulus pressurization, jet reaction, pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency and faulted conditions.

3.2.2.1 Design Conditions Because there are no changes in the design conditions due to TPO, the design stresses are unchanged and the Code requirements are met.

3.2.2.2 Normal & Upset Conditions The reactor coolant temperature and flows at TPO conditions are unchanged from those at current rated conditions, because the 105% OLTP power uprate evaluations were performed at conditions (( )) that bound the change in operating conditions from CLTP to TPO. The evaluation type is mainly reconciliation of the stresses and usage factors to reflect TPO conditions. A primary plus secondary stress analysis was performed showing TPO stresses still meet the requirements of the ASME Code,Section III, and Subsection NB for all components. Lastly, the fatigue usage was evaluated for the limiting location of components ((

)) The Limerick Units 1 and 2 fatigue analysis results for the limiting components are provided in Table 3-7. The Limerick Units 1 and 2 analysis results for TPO show that all components meet their ASME Code requirements and no further analysis is required.

3.2.2.3 Emergency & Faulted Conditions The stresses due to Emergency & Faulted conditions are based on loads such as peak dome pressure, which are unchanged for TPO. Therefore, Code requirements are met for all RPV components under Emergency & Faulted conditions.

3.3 REACTOR INTERNALS The reactor internals include core support structure (CSS) and non-core support structure (non-CSS) components.

3.3.1 Reactor Internal Pressure Difference The reactor internal pressure differences (RIPDs) are affected more by the maximum licensed core flow rate than by the power level. The maximum licensed core flow rate is not changed for the TPO uprate. The effect due to the changes in loads for both Normal and Upset conditions is reported in Section 3.3.2. The Normal and Upset evaluations of RIPDs for the TPO uprate are 3-7

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION bounded by the current analyses that conservatively assumed an initial power level of 110% of OLTP for Normal conditions and 112% of OLTP for Upset conditions. The Emergency and Faulted evaluations of RIPDs for the TPO uprate are bounded by the current analyses that conservatively assumed an initial power level of 112% of OLTP.

Fuel Bundle Lift Margins and Control Rod Guide Tube (CRGT) Lift Forces are calculated at the Faulted condition to demonstrate that fuel bundles would not lift under the worst conditions. The current analysis conservatively assumed of 112% of OLTP and 110% core flow, which bounds the TPO. The Fuel Lift Margins for the normal and upset conditions at the TPO RTP decrease slightly from CLTP. The CRGT Lift Forces for the normal condition at the TPO RTP increase slightly from CLTP. The Fuel Lift Margins and CRGT Lift Forces at Normal and Upset conditions are bounded by Emergency and Faulted conditions. The effect due to the changes in Fuel Lift Margins and CRGT Lift Forces is reported in Section 3.3.2.

Acoustic and flow-induced loads on jet pump, core shroud and shroud support due to recirculation line break are bounded by the current analyses that conservatively assumed an initial power level of 102% of CLTP.

3.3.2 Reactor Internals Structural Evaluation The RPV internals consist of the CSS components and non-CSS components. The RPV Internals are not ASME Code components, however, the requirements of the ASME Code are used as guidelines in their design/analysis. The evaluations/stress reconciliation in support of the TPO was performed consistent with the design basis analysis of the components. The reactor internal components evaluated are:

CSS Components

  • Shroud Support
  • Shroud
  • Core Plate
  • Top Guide
  • Orificed Fuel Support Non-CSS Components
  • FW Sparger 3-8

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION

  • Jet Pump
  • CS Line and Sparger
  • Access Hole Cover
  • Shroud Head and Steam Separator Assembly
  • In-Core Housing and Guide Tube
  • Vessel Head Cooling Spray Nozzle
  • Core Differential Pressure and Liquid Control Line
  • Steam Dryer The original configurations of the RPV internals are considered in the TPO evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation (e.g., jet pumps).

The loads considered in the evaluation of the RPV internals include RIPDs, dead weight, seismic, SRV, LOCA, Annulus Pressurization/Jet Reaction (AP/JR), acoustic and flow induced loads due to Recirculation Line Break (RLB), fuel lift, hydraulic flow and thermal loads.

RPV design pressure remains unchanged. RIPD loads are bounded by the existing design basis values (110% of OLTP and 110% ICF). Seismic, SRV, LOCA and AP/JR loads remain unchanged. Acoustic and flow induced loads due to RLB, hydraulic flow and thermal loads remain bounded. The effect of weight change on load due to jet pump repair is insignificant.

Fuel lift loads increased in Service Level D. The stresses of core plate, top guide and control rod guide tube were reconciled for the increase of the fuel lift loads to show that adequate stress margins exist, and the stresses remain within the allowable limits. The limiting stresses of other RPV internal components remain bounded by the existing design basis values (110% OLTP and 110% ICF) (See Table 3-8). The input loads to existing flaw analysis are not impacted by TPO.

Hence, the existing flaw evaluations remain valid to TPO. Therefore the RPV internal components are demonstrated to be structurally qualified for operation in the TPO conditions.

Qualitative analysis was performed to assess the potential for the Steam Dryer flow induced vibration (FIV) using 1/5th scale model testing and an analytical approach. The scale model testing has indicated some potential propagation of acoustic resonance in two of the four main steam lines (MSLs) at CLTP conditions. The Steam Dryer has operated satisfactorily for approximately 14 years at CLTP with no adverse flow effects. The testing has also shown a downward trend (i.e., reduction) of the normalized root mean square (RMS) pressure in the 3-9

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION MSLs as flow conditions are changed from CLTP to TPO conditions. Additionally, for the other two MSLs, the normalized RMS pressure was determined to be constant and below any potential acoustic resonance propagation thresholds. Other loads applicable to the Steam Dryer evaluation are deadweight, seismic, RIPD, SRV, LOCA, AP, JR and fuel lift loads. Dead weight and seismic loads remain unchanged for the TPO conditions. RIPDs contributing to the Steam Dryer load remain bounded by previous analyses. SRV, LOCA, AP, JR, and fuel lift loads remain bounded in the TPO conditions by the CLTP values. All applicable loads to the Steam Dryer are bounded by the existing design basis for the TPO conditions. Hence, the Steam Dryer remains structurally qualified for plant operation at the TPO conditions.

3.3.3 Steam Separator and Dryer Performance For Limerick, the TPO performance of the steam dryer/separator was evaluated based on a plant-specific evaluation using recent fuel cycle core design for Unit 1 Fuel Cycles 12 and 13, which also bounds the current Unit 2 Fuel Cycle 11. The results of the evaluation demonstrated that the steam dryer/separator will maintain moisture content 0.1 weight % at TPO operating conditions except under some ICF conditions or above L7. TPO results in an increase in the amount of saturated steam generated in the reactor core. For constant core flow, this results in an increase in the separator inlet quality, an increase in the steam 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 dryer/separator performance. The net effect of changes due to TPO up to 100 % RCF does not result in exceeding a moisture content of 0.1 wt % leaving the steam dryer. TPO does not significantly affect the performance of the steam dryer/separator.

3.4 FLOW-INDUCED VIBRATION The process for the reactor vessel internals vibration assessment is described in Section 5.5.1.3 of the TLTR (Reference 1). An evaluation determined the effects of FIV on the reactor internals at 110% rated core flow and TPO RTP of 101.7%. The vibration levels for the TPO uprate conditions were estimated from vibration data recorded during startup testing of the NRC designated prototype plant (Browns Ferry Unit 1) and during other tests. These expected vibration levels were compared with established vibration acceptance limits. The following components were evaluated for the TPO uprate:

Component(s) Process Parameter(s) TPO Evaluation Shroud Steam Flow at TPO RTP is ~2% Slight increase in FIV. Extrapolation of Shroud Head and Separator greater than CLTP. measured data shows stresses are within limits.

Jet Pumps Jet Pump flow at TPO is unchanged No change from CLTP.

Jet Pump Riser Braces Resonance due to vane passing No resonance due to vane passing frequency frequency Jet Pump Sensing Lines Resonance due to vane passing No resonance due to vane passing frequency. frequency.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Component(s) Process Parameter(s) TPO Evaluation FW Sparger FW Flow at TPO RTP is ~ 2% Slight increase in FIV. Extrapolation of greater than CLTP. measured data shows stresses are well within limits.

Control Rod Guide Tube and Core Flow at TPO Licensed Thermal No change In-Core Guide Tubes Power (TLTP) is unchanged from CLTP.

The calculations for the TPO uprate conditions indicate that vibrations of all safety-related reactor internal components are within the GEH acceptance criteria. For some components, FIV is a function of core flow. Because the maximum licensed core flow is unchanged for TPO, FIV for those components is not affected. The analysis is conservative for the following reasons:

  • The GEH criteria of 10,000 psi peak stress intensity is much more conservative than the ASME allowable peak stress intensity of 13,600 psi for service cycles in excess of 1011.
  • Conservatively, the peak responses of the applicable modes are absolute summed.
  • Although the maximum vibration stress amplitude of each mode is used in the absolute sum process, the maximum vibration modal amplitude actually differs with time.

Therefore, it is concluded that the flow-induced vibrations for all evaluated components remain within acceptable limits.

The safety-related Main Steam (MS) and Feedwater (FW) piping have minor increased flow rates and flow velocities resulting from the TPO uprate. The MS and FW piping experience increased vibration levels, approximately proportional to the increase in the square of the flow velocities and also in proportion to any increase in fluid density. The decrease in FW fluid density for TPO uprate conditions, as a result of the ~ 2º F increase in FW temperature, is insignificant. The MS and FW piping vibration is expected to increase only by about 4%. A MS and FW piping FIV test program, during initial plant startup, showed that vibration levels were within acceptance criteria and operating experience shows that there are no existing vibration problems in MS and FW lines at CLTP operating conditions. Therefore, the MS and FW lines vibration will remain within acceptable limits during TPO. Analytical evaluation has shown that the safety-related thermowells in the MS, FW, and Recirculation piping systems are structurally adequate for the TPO operating condition.

3.5 PIPING EVALUATION 3.5.1 Reactor Coolant Pressure Boundary Piping The methods used for the piping and pipe support evaluations are described in TLTR Appendix K. These approaches are identical to those used in the evaluation of previous BWR power uprates of up to 20% power. The effect of the TPO uprate with no nominal vessel dome 3-11

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION pressure increase is negligible for the Reactor Coolant Pressure Boundary (RCPB) portion of all piping except for portions of the FW lines, MS lines, and piping connected to the FW and MS lines. The following table summarizes the evaluation of the piping Inside Containment.

Component(s) / Concern Process Parameter(s) TPO Evaluation Recirculation System Nominal dome pressure at TPO RTP is identical to No change in pipe stress CLTP No effect on pipe Recirculation flow at TPO RTP is identical to CLTP supports Pipe Stresses No change in core pressure Pipe Supports Small decrease in Recirculation fluid temperature MS and Attached Piping Nominal dome pressure at TPO RTP is identical to Current Licensing Basis (Inside Containment) (e.g., SRV CLTP envelops TPO Discharge Line (SRVDL) piping up to Steam flow at TPO RTP is ~2% greater than CLTP conditions; therefore, first anchor, Reactor Core Isolation No change in MSL pressure piping system is Cooling (RCIC) / High Pressure Coolant acceptable for TPO.

Injection (HPCI) piping (Steam Side), MS drain lines, RPV head vent line piping Negligible change in located Inside Containment) pipe stress; negligible effect on pipe supports Minor increase in the Pipe Stresses potential for FAC (FAC Pipe Supports concerns are covered by existing piping monitoring program)

Flow-accelerated erosion/corrosion (FAC)

FW and Attached Piping Nominal dome pressure at TPO RTP is identical to Current Licensing Basis (Inside Containment) CLTP envelops TPO FW flow at TPO RTP is ~2% greater than CLTP conditions; therefore, Minor change in FW line pressure piping system is Pipe Stresses Fluid temperature increases 2°F acceptable for TPO.

Pipe Supports Negligible change in pipe stress Negligible effect on pipe supports FAC Minor increase in the potential for FAC (FAC concerns are covered by existing piping monitoring program)

RPV bottom head drain line, RCIC Nominal dome pressure at TPO RTP is identical to Negligible change in piping, HPCI piping, LPCI piping, CS CLTP pipe stress piping, Standby Liquid Control System Small Increase in core pressure drop of < 1 psi Negligible effect on (SLCS) piping, and Reactor Water Small decrease in Recirculation fluid temperature pipe supports Cleanup (RWCU) piping Minor increase in the potential for FAC Pipe Stresses (FAC concerns are Pipe Supports covered by existing piping monitoring FAC program)

For the MS and FW lines, supports, and connected lines, the methodologies as described in TLTR Section 5.5.2 and Appendix K were used to determine the percent increases in applicable ASME Code stresses, displacements, CUF, and pipe interface component loads (including 3-12

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION supports) as a function of percentage increase in pressure (where applicable), temperature, and flow due to TPO conditions. As necessary, the percentage increases were applied to the highest calculated stresses, displacements, and the CUF at applicable piping system node points to conservatively determine the maximum TPO calculated stresses, displacements and usage factors. This approach is conservative because the TPO does not affect weight and all building filtered loads (i.e., seismic loads are not affected by the TPO). The factors were also applied to nozzle load, support loads, penetration loads, valves, pumps, heat exchangers and anchors so that these components could be evaluated for acceptability, where required. No new computer codes were used or new assumptions introduced for this evaluation.

MS and Attached Piping System Evaluation The MS piping system (Inside Containment) was evaluated for compliance with the ASME code stress criteria, and for the effects of thermal displacements on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, flanges and valves were also evaluated.

Pipe Stresses The evaluation shows that the increase in flow associated with the TPO uprate does not result in load limits being exceeded for the MS piping system or for the RPV nozzles. The current licensing basis design analyses have sufficient design margin between calculated stresses and ASME Code allowable limits to justify operation at the TPO uprate conditions. The temperature of the MS piping (Inside Containment) is unchanged for the TPO.

The design adequacy evaluation results show that the requirements of ASME,Section III, Subsection NB/ND (as applicable) requirements are satisfied for the evaluated piping systems.

Therefore, the TPO does not have an adverse effect on the MS piping design.

Pipe Supports The current licensing basis MS piping was reviewed for the effects of transient loading on the piping snubbers, hangers, struts, and pipe whip restraints. A review of the increases in MS flow associated with the TPO uprate indicates that piping load changes do not result in any load limit being exceeded.

Erosion / Corrosion The carbon steel MS piping can be affected by FAC. FAC is affected by changes in fluid velocity, temperature and moisture content. Limerick has an established FAC monitoring program for monitoring pipe wall thinning in single and two-phase high-energy carbon steel piping. The variation in velocity, temperature, and moisture content resulting from the TPO uprate are minor changes to parameters affecting FAC. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the MS and attached piping indicates minimal 3-13

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION effect. Table 3-9 shows piping line segments that are recommended for additional review under the station FAC program.

The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO uprate has no adverse effect on high-energy piping systems potentially susceptible to pipe wall thinning due to FAC.

FW Piping System Evaluation The current licensing basis FW piping system (Inside Containment) reports were reviewed for compliance with the ASME Section III Code stress criteria, and for the effects of thermal expansion displacements on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, and valves were also evaluated.

Pipe Stresses A review of the change in temperature, pressure, and flow associated with the TPO uprate indicates that piping load changes do not result in load limits being exceeded for the FW piping system or for RPV nozzles. The current licensing basis design analyses have adequate design margin between calculated stresses and ASME Code allowable limits to justify operation at the TPO uprate conditions.

The design adequacy evaluation shows that the requirements of ASME,Section III, Subsection NB/NC/ND-3600 requirements remain satisfied. Therefore, the TPO does not have an adverse effect on the FW piping design.

Pipe Supports The TPO does not affect the FW piping snubbers, hangers, and struts. A review of the increase in FW temperature and flow associated with the TPO indicates that piping load changes do not result in any load limit being exceeded at the TPO uprate conditions.

Erosion / Corrosion The carbon steel FW piping can be affected by FAC. FAC in the FW piping is affected by changes in fluid velocity and temperature. Limerick has an established program for monitoring pipe wall thinning in single and two-phase high-energy carbon steel piping. The variation in velocity and temperature resulting from the TPO uprate are minor changes to parameters affecting FAC. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the FW Piping System indicates minimal effect. Table 3-9 shows piping line segments that are recommended for additional review under the station FAC program.

3-14

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO uprate has no adverse effect on high energy piping systems potentially susceptible to pipe wall thinning due to FAC.

3.5.2 Balance-of-Plant Piping Evaluation This section addresses the adequacy of the BOP piping design (outside of the RCPB) for operation at the TPO 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).

Pipe Supports For the condensate, FW, extraction steam, heater drain, and main steam systems, operating system pressures and temperatures under TPO will remain within design ratings.

Because there is no change in the MS operating temperature, i.e., from the reactor to the MS stop valves, there is no change in the thermal expansion stress for TPO. For systems with increased operating temperatures (i.e., MS downstream of the stop valves, condensate, feedwater, extraction steam, heater drains), changes to thermal expansion stresses are small and acceptable.

Pipe support loads will experience a small increase in the thermal load (< 1%). However, when considering the combination with other loads that are not affected by the TPO uprate (e.g.,

deadweight) the combined support load increase is insignificant. This piping has been analyzed to conditions which envelope operations under TPO.

For the MS system piping outside containment, the turbine stop valve (TSV) closure transient was reviewed against conditions that bound operations under TPO. Available stress and support load margins are adequate to accommodate the increase in loading associated with this fluid transient.

For the FW system piping outside containment, changes to fluid transient loading such as for feed pump trip are small. The station design for fluid transients was reviewed and no changes are required for TPO.

Erosion / Corrosion The integrity of high-energy piping systems is assured by proper design in accordance with the applicable codes and standards. Piping thickness of carbon steel components can be affected by FAC. Limerick has an established program for monitoring pipe wall thinning in single phase and two-phase high-energy carbon steel piping. FAC rates may be influenced by changes in fluid velocity, temperature, and moisture content. The FAC monitoring program includes the use of a predictive method to calculate wall thinning of components susceptible to FAC. For TPO, the evaluation of predicted wall thinning of the BOP piping indicates minimal effect. Table 3-9 3-15

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION shows piping line segments that are recommended for additional review under the station FAC program.

Operation at the TPO RTP results in some changes to parameters affecting FAC in those systems associated with the turbine cycle (e.g., condensate, FW, MS). The evaluation of and inspection for FAC in BOP systems is addressed by compliance with Generic Letter (GL) 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning. The plant FAC program currently monitors the affected systems. Continued monitoring of the systems provides confidence in the integrity of susceptible high-energy piping systems. Appropriate changes to piping inspection frequency will be implemented to ensure adequate margin exists for those systems with changing process conditions. This action takes into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO has no adverse effect on high-energy piping systems potentially susceptible to pipe wall thinning due to FAC.

3.6 REACTOR RECIRCULATION SYSTEM The Reactor Recirculation System (RRS) evaluation process is described in TLTR Section 5.6.2.

The TPO uprate has a minor effect on the RRS and its components. The TPO uprate does not require an increase in the maximum core flow. No significant reduction of the maximum flow capability occurs due to the TPO uprate because of the small increase in core pressure drop

(< 1 psi). The effect on pump net positive suction head (NPSH) at TPO conditions is negligible.

An evaluation has confirmed that no significant increase in RRS vibration occurs from the TPO operating conditions.

The cavitation protection interlock for the recirculation pumps and jet pumps is expressed in terms of FW flow. This interlock is based on sub-cooling 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 TPO.

3.7 MAIN STEAM LINE FLOW RESTRICTORS The generic evaluation provided in TLTR Appendix J.2.3.7 is applicable to Limerick. The requirements for the MSL flow restrictors remain unchanged for TPO uprate conditions. No change in steam line break flow rate occurs because the operating pressure is unchanged. All safety and operational aspects of the MSL flow restrictors are within previous evaluations.

3.8 MAIN STEAM ISOLATION VALVES The generic evaluation provided in TLTR Appendix J.2.3.7 is applicable to Limerick. The requirements for the main steam isolation valves (MSIVs) remain unchanged for TPO uprate conditions. All safety and operational aspects of the MSIVs are within previous evaluations.

3-16

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 3.9 REACTOR CORE ISOLATION COOLING The RCIC system provides inventory makeup to the reactor vessel when the vessel is isolated from the normal high-pressure makeup systems. The generic evaluation provided in TLTR Section 5.6.7 is applicable to Limerick. The TPO uprate does not affect the RCIC system operation, initiation, or capability requirements.

3.10 RESIDUAL HEAT REMOVAL SYSTEM The Residual Heat Removal (RHR) system is designed to restore and maintain the coolant inventory in the reactor vessel and to remove sensible and decay heat from the primary system and containment following reactor shutdown for both normal and post accident conditions. The RHR system is designed to function in several operating modes. The generic evaluation provided in TLTR Sections 5.6.4 and Appendices J.2.3.1 and J.2.3.13 are applicable to Limerick.

The following table summarizes the effect of the TPO on the design basis of the RHR system.

Operating Mode Key Function TPO Evaluation LPCI Mode Core Cooling See Section 4.2.4 Suppression Pool Cooling (SPC) and Normal SPC function is to maintain pool Containment Analyses have Containment Spray Cooling (CSC) temperature below the limit. been performed at 102% of Modes CLTP.

For Abnormal events or accidents, the SPC mode maintains the long-term pool temperature below the design limit.

The CSC mode sprays water into the containment to reduce post-accident containment pressure and temperature.

Shutdown Cooling (SDC) Mode Removes sensible and decay heat from the The slightly higher decay heat reactor primary system during a normal has negligible effect on the SDC reactor shutdown. mode, which has no safety function.

Steam Condensing Mode Decay Heat removal Limerick does not have a Steam Condensing Mode of RHR.

Fuel Pool Cooling Assist Supplemental fuel pool cooling in the event See Section 6.3.1 that the fuel pool heat load exceeds the heat removal capability of the Fuel Pool Cooling system.

The ability of the RHR system to perform required safety functions is demonstrated with analyses based on 102% of CLTP. Therefore, all safety aspects of the RHR system are within previous evaluations. The requirements for the RHR system remain unchanged for TPO uprate conditions.

3.11 REACTOR WATER CLEANUP SYSTEM The generic evaluation of the RWCU system provided in TLTR Sections 5.6.6 and J.2.3.4 is applicable to Limerick. The performance requirements of the RWCU system are negligibly 3-17

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION affected by TPO uprate. There is no significant effect on operating temperature and pressure conditions in the high-pressure portion of the system. RWCU flow is not changed for TPO conditions. Steady-state power level changes for much larger power uprates have shown no effect on reactor water chemistry and the performance of the RWCU system. Power transients that result in crud bursts causing high intermediate loading on the system capacity are the primary source of challenge to the system, so safety and operational aspects of water chemistry performance are not affected by the TPO.

3-18

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-1 Limerick Unit 1 Upper Shelf Energy 40-Year License (32 EFPY)

Initial Initial Test Longitudinal Transverse USE 32 EFPY 1/4T 32 EFPY Transverse 32 EFPY Equivalent

[1]

Location Heat Temperature Shear USE %Cu Fluence  % Decrease USE [2] USE [3] Margin Results [4,6]

2

(°F)  % (ft-lb) (ft-lb) (n/cm ) (ft-lb)

Plates:

Lower C7688-1 40 50 85 55.3 0.12 1.3E+18 13.5 48 13.5% < 21%

C7698-2 40 50 100 65 0.11 1.3E+18 12.5 57 C7688-2 40 70 104 67.6 0.12 1.3E+18 13.5 58 Lower-Intermediate C7689-1 40 60 93 60.5 0.11 1.3E+18 12.5 53 C7677-1 40 40 71 46.2 0.11 1.3E+18 12.5 40 12.5% < 21%

C7698-1 40 50 96 62.4 0.11 1.3E+18 12.5 55 Welds:

Vertical:

BE 411A3531/H004A27A 10 60 68 0.02 1.3E+18 8.5 62 BA,BB,BD,BF 06L165/F017A27A 10 70 62 0.03 1.3E+18 10 56 BA,BD,BE,BF 662A746/H013A27A -20 65 95 0.03 1.3E+18 10 86 BA,BB,BC 3P4000/3932-989 10 80 97 0.02 1.3E+18 8.5 89 BF S3986/RUN 934 10 40 51 0.054 1.3E+18 12.5 45 22% < 34%

BA,BB,BE 1P4218/3929-989 10 83 102 0.053 1.3E+18 12 90 TEST PLATE 421A6811/F022A27A 10 75 91 0.09 1.3E+18 14.5 78 Girth:

AB 07L857/B101A27A 10 50 39 0.03 1.3E+18 10 35 22% < 34%

AB 402C4371/C115A27A 10 70 92 0.02 1.3E+18 8.5 84 AB 411A3531/H004A27A 10 60 68 0.02 1.3E+18 8.5 62 AB 09M057/C109A27A 10 50 44 0.03 1.3E+18 10 40 22% < 34%

AB 412P3611/J417B27AF 130 100 136 0.03 1.3E+18 10 122 AB 03M014/C118A27A 10 40 47 0.01 1.3E+18 7 44 22% < 34%

AB L83355/S411B27AD 130 100 150 0.03 1.3E+18 10 135 AB 640892/J424B27AE 130 100 118 0.09 1.3E+18 14.5 101 AB 401P6741/S419B27AG 130 100 136 0.03 1.3E+18 10 122 AB 5P6756 0 100 121 0.083 1.3E+18 14 104 Nozzles:

Water Level Instrumentation Forging SB166 [5] 40 40 71 46.2 0.11 1.9E+17 8 42 8% < 21%

LPCI Nozzle Weld 07L669/K004A27A 10 40 54 0.03 1.9E+17 6.5 50 Weld 401Z9711/A022A27A 10 80 104 0.02 1.9E+17 6 98 Forging: 45° Q2Q25W -20 30 58 37.7 0.18 1.9E+17 11 34 11% < 21%

Forging: 135° Q2Q35W -20 70 77 50.1 0.18 1.9E+17 11 45 11% < 21%

Forging: 225° Q2Q25W -20 40 60 39 0.18 1.9E+17 11 35 11% < 21%

Forging: 315° Q2Q35W -20 40 41 26.7 0.18 1.9E+17 11 24 11% < 21%

BEST ESTIMATE CHEMISTRIES:

per BWRVIP-135 R1 [7]:

BA,BB,BC 3P4000/3932-989 10 80 97 0.02 1.3E+18 9.5 88 BF S3986/RUN 934 10 40 51 0.058 1.3E+18 12 45 22% < 34%

BA,BB,BE 1P4218/3929-989 10 83 102 0.058 1.3E+18 12.5 89 AB 5P6756 0 100 121 0.08 1.3E+18 13.5 105 Integrated Surveillance Program (ISP):

Plate C2761-2 127.2 0.10 1.3E+18 15 108 Weld 5P6756 104.4 0.06/0.08 [8] 1.3E+18 13.5 90 Notes:

[1] Transverse USE for plate and forging materials is obtained using 65% of the longitudinal USE.

[2] USE Decrease is obtained from Regulatory Guide 1.99, Revision 2, Figure 2.

[3] 32 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE /100)].

[4] The initial USE for the materials evaluated in this column is very low due to a lack of sufficient test data. Although conservatively evaluated, this column demonstrates that these materials are bounded by an Equivalent Margin Analysis.

[5] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.

[6] These values have been determined using NEDO-32205-A. Note that the weld materials required adjustment because the measured results exceeded Regulatory Guide 1.99, Revision 2 predicted results.

[7] The best estimate chemistries from BWRVIP-135 Revision 1 are provided in a separate section in order to discern between the original plant-specific properties. It is intended that the best estimate chemistries supersede the plant-specific chemistries.

[8] The first %Cu value is from Appendix B of BWRVIP-135 Revision 1; the second value is from Appendix D and represents the best estimate chemistry. The bounding value (Appendix D) is conservatively used in this evaluation.

3-19

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-2 Limerick Unit 2 Upper Shelf Energy 40-Year License (32 EFPY) 32 EFPY Test Initial Initial 32 EFPY 1/4T  % Decrease Transverse 32 EFPY Equivalent Margin

[1] [2] [3] [4]

Location Heat Temperature Shear Longitudinal USE Transverse USE %Cu Fluence USE USE Results

(°F)  % (ft-lb) (ft-lb) (n/cm2) (ft-lb)

PLATES:

Lower B3312-1 40 50 78 50.7 0.13 1.3E+18 14 44 14% < 21%

B3416-1 40 50 61 39.7 0.14 1.3E+18 14.5 34 14.5% < 21%

C9621-2 40 30 89 57.9 0.15 1.3E+18 15 49 15% < 21%

Lower-Intermediate C9569-2 40 40 87 56.6 0.11 1.3E+18 12.5 50 C9526-1 40 40 89 57.9 0.11 1.3E+18 12.5 51 C9526-2 40 50 97 63.1 0.11 1.3E+18 12.5 55 WELDS:

Vertical:

BA,BB,BD,BE,BF 432A2671/H019A27A -20 40 54 0.04 1.3E+18 11 48 22% < 34%

BA,BC 03R728/L910A27A 10 70 72 0.03 1.3E+18 10 65 BA,BB,BC,BD,BE,BF 3P4000/3933(Single Wire) 10 80 95 0.02 1.3E+18 8.5 87 BA,BB,BC,BD,BE,BF 3P4000/3933 (Tandem Wire) 10 98 91 0.02 1.3E+18 8.5 83 BB 401Z9711/A022A27A 10 80 104 0.02 1.3E+18 8.5 95 BC 662A746/H013A27A -20 65 95 0.03 1.3E+18 10 86 BC 402A0462/B023A27A 10 62 86 0.02 1.3E+18 8.5 79 BD,BE 09L853/A111A27A 10 60 79 0.03 1.3E+18 10 71 BC,BD,BE,BF 07L669/K004A27A -20 40 54 0.03 1.3E+18 10 49 22% < 34%

Girth:

AB 07L857/B101A27A 10 50 39 0.03 1.3E+18 10 35 22% < 34%

AB L83355/S411B27AD 130 100 150 0.03 1.3E+18 10 135 AB 402C4371/C115A27A 10 70 92 0.02 1.3E+18 8.5 84 AB 03M014/C118A27A 10 40 47 0.01 1.3E+18 7 44 22% < 34%

AB 411A3531/H004A27A 10 60 68 0.02 1.3E+18 8.5 62 AB 09M057/C109A27A 10 50 44 0.03 1.3E+18 10 40 22% < 34%

AB 640892/J424B27AE 130 100 118 0.09 1.3E+18 14.5 101 AB 401P6741/S419B27AG 130 100 136 0.03 1.3E+18 10 122 AB 412P3611/J417B27AF 130 100 136 0.03 1.3E+18 10 122 NOZZLES:

Water Level Instrumentation Forging SB166 [5] 40 40 87 56.6 0.11 1.9E+17 10.5 51 LPCI Weld KA C3L46C/J020A27A 10 60 40 0.02 1.9E+17 6 38 14% < 34%

Weld KA 422B7201/L030A27A -20 30 38 0.04 1.9E+17 7.5 35 14% < 34%

Weld KA 4P4784/3930 (Single Wire) 40 90 97 0.06 1.9E+17 8.5 89 Weld KA 4P4784/3930 (Tandem Wire) 40 75 95 0.06 1.9E+17 8.5 87 Forging 892L-1 Q2Q33W -20 40 46 29.9 0.15 1.9E+17 10 27 10% < 21%

Forging 892L-2 Q2Q33W -20 40 43 28 0.15 1.9E+17 10 25 10% < 21%

Forging 892L-3 Q2Q33W -20 50 53 34.5 0.15 1.9E+17 10 31 10% < 21%

Forging 892L-4 Q2Q33W -20 50 48 31.2 0.15 1.9E+17 10 28 10% < 21%

BEST ESTIMATE CHEMISTRIES per BWRVIP-135 R1 [7]:

BA,BB,BC,BD,BE,BF 3P4000/3933(Single Wire) 10 80 95 0.02 1.3E+18 9.5 86 BA,BB,BC,BD,BE,BF 3P4000/3933 (Tandem Wire) 10 98 91 0.02 1.3E+18 9.5 82 Weld [6] CTY538/A027A27A 10 70 94 61 0.03 1.3E+18 10 55 Weld 5P6756 104.4 0.08 1.3E+18 13.5 90 INTEGRATED SURVEILLANCE PROGRAM (ISP):

Plate B0673-1 158.1 0.15 1.3E+18 15 134 Weld 5P6756 104.4 0.06/0.08[8] 1.3E+18 13.5 90 Notes:

[1] Transverse USE for plate and forging materials is obtained using 65% of the longitudinal USE.

[2] USE Decrease is obtained from Regulatory Guide 1.99, Revision 2, Figure 2.

[3] 32 EFPY Transverse USE = Initial Transverse USE * [1 - (% Decrease USE /100)].

[4] The initial USE for the materials evaluated in this column is very low due to a lack of sufficient test data. Although conservatively evaluated, this column demonstrates that these materials are bounded by an Equivalent Margin Analysis. The weld materials require adjustment because the measured decrease exceeds the predicted decrease. This has been performed in accordance with Regulatory Guide 1.99, Revision 2.

[5] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.

[6] CMTR records do not indicate that this is a surveillance weld. However, the CMTRs demonstrate that this heat is a weld in the vessel; therefore, it is evaluated using the best estimate chemistry from BWRVIP-135 Revision 1.

[7] The best estimate chemistries obtained from BWRVIP-135 Revision 1 have been provided in a separate section in order to discern between the original plant-specific properties. It is intended that the best estimate chemistries supersede the plant-specific chemistries.

[8] The first %Cu value is from Appendix B of BWRVIP-135 Revision 1; the second value is from Appendix D and represents the best estimate chemistry. The bounding value (Appendix D) is conservatively used in this evaluation.

3-20

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-3 Limerick Unit 1 Adjusted Reference Temperatures 40-Year License (32 EFPY)

Plates and Welds 2

Thickness = 6.19 inches 32 EFPY Peak I.D. fluence = 1.9E+18 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.3E+18 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.3E+18 n/cm Nozzle Forgings and Welds 2

Thickness = 6.19 inches 32 EFPY Peak I.D. fluence = 2.8E+17 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.9E+17 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.9E+17 n/cm Initial 1/4 T 32 EFPY I 32 EFPY 32 EFPY Component Heat or Heat/Lot %Cu %Ni CF RTNDT Fluence RTNDT Margin Shift ART 2

°F n/cm °F °F °F °F PLATES:

Lower 14-1 C7688-1 0.12 0.51 81 10 1.3E+18 38 0 17 34 72 82 14-2 C7698-2 0.11 0.48 73 10 1.3E+18 35 0 17 34 69 79 14-3 C7688-2 0.12 0.51 81 10 1.3E+18 38 0 17 34 72 82 Lower-Intermediate 17-1 C7689-1 0.11 0.48 73 10 1.3E+18 35 0 17 34 69 79 17-2 C7677-1 0.11 0.50 73 20 1.3E+18 35 0 17 34 69 89 17-3 C7698-1 0.11 0.48 73 10 1.3E+18 35 0 17 34 69 79 WELDS:

Vertical (shop) [1]

BE 411A3531/H004A27A 0.02 0.96 27 -50 1.3E+18 13 0 6 13 26 -24 BA,BB,BD,BF 06L165/F017A27A 0.03 0.99 41 -50 1.3E+18 19 0 10 19 39 -11 BA,BD,BE,BF 662A746/H013A27A 0.03 0.88 41 -20 1.3E+18 19 0 10 19 39 19 BA,BB,BC 3P4000/3932-989 [2] 0.02 0.928 27 -50 1.3E+18 13 0 6 13 26 -24 BF S3986/RUN 934 [2] 0.054 0.969 74 -42 1.3E+18 35 0 18 35 70 28 BA,BB,BE 1P4218/3929-989 [2] 0.053 0.89 72 -50 1.3E+18 34 0 17 34 68 18 TEST PLATE 421A6811/F022A27A 0.09 0.81 122 -50 1.3E+18 58 0 28 56 114 64 Girth (field)

AB 07L857/B101A27A 0.03 0.97 41 -6 1.3E+18 19 0 10 19 39 33 AB 402C4371/C115A27A 0.02 0.92 27 -50 1.3E+18 13 0 6 13 26 -24 AB 411A3531/H004A27A 0.02 0.96 27 -50 1.3E+18 13 0 6 13 26 -24 AB 09M057/C109A27A 0.03 0.89 41 -36 1.3E+18 19 0 10 19 39 3 AB 412P3611/J417B27AF 0.03 0.93 41 -80 1.3E+18 19 0 10 19 39 -41 AB 03M014/C118A27A 0.01 0.94 20 -34 1.3E+18 9 0 5 9 19 -15 AB L83355/S411B27AD 0.03 1.08 41 -70 1.3E+18 19 0 10 19 39 -31 AB 640892/J424B27AE 0.09 1.00 122 -60 1.3E+18 58 0 28 56 114 54 AB 401P6741/S419B27AG 0.03 0.92 41 -60 1.3E+18 19 0 10 19 39 -21 AB 5P6756 [2] 0.083 0.943 112 -60 1.3E+18 53 0 26 53 106 46 NOZZLES:

Water Level Instrumentation Forging SB166 [4] 0.11 0.50 73 20 1.9E+17 12 0 6 12 25 45 LPCI Nozzle Weld 07L669/K004A27A 0.03 1.02 41 -50 1.9E+17 7 0 3 7 14 -36 Weld 401Z9711/A022A27A 0.02 0.83 27 -50 1.9E+17 5 0 2 5 9 -41 Forging: 45° & 225° Q2Q25W 0.18 0.78 140 -6 1.9E+17 24 0 12 24 47 41 Forging: 135° & 315° Q2Q35W 0.18 0.85 142 -8 1.9E+17 24 0 12 24 48 40 BEST ESTIMATE CHEMISTRIES per BWRVIP-135 R1 [3]:

BA,BB,BC 3P4000 0.02 0.935 27 -50 1.3E+18 13 0 6 13 26 -24 BF S3986 0.058 0.949 79.2 -42 1.3E+18 37 0 19 37 75 33 BA,BB,BE 1P4218 0.058 0.865 79.2 -50 1.3E+18 37 0 19 37 75 25 AB 5P6756 [8] 0.08 0.936 108 -60 1.3E+18 51 0 26 51 102 42 AB 5P6756 [8] 0.08 0.936 154 [7] -60 1.3E+18 73 0 14 28 101 41 INTEGRATED SURVEILLANCE PROGRAM:

Plate C2761-2 [5] 0.10 0.54 65 20 1.3E+18 31 0 15 31 62 82 Weld 5P6756 [6] 0.06/0.08 0.93/0.936 154 [7] -60 1.3E+18 73 0 14 28 101 41 Notes:

[1] Welds BA, BB, BC occur in the lower shell (Shell Ring #1) and welds BD, BE, BF occur in the lower-intermediate shell (Shell Ring #2).

[2] This material is evaluated below using the best estimate chemistry for this heat of material as provided in BWRVIP-135, Revision 1.

[3] It is intended that the best estimate chemistries supersede the plant-specific chemistries.

[4] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.

[5] The ISP plate material is NOT the same heat as the target vessel plate material. The results provided are for information only and do not affect the Limerick Unit 1 PT curves.

[6] The ISP weld material is NOT the same heat as the target vessel weld material. However, because this heat does occur in the beltline region, the material is evaluated as defined in Section 3 of BWRVIP-135 Revision 1, and considered applicable to the beltline region.

[7] The Adjusted CF is calculated as: (108/82)

  • 116.9 = 154°F. Note that the chemistry values provided represent the BWRVIP-135 Revision 1 Appendix B values (first value) and the BWRVIP-135 Revision 1 Appendix D best estimate chemistry (second value). These are provided for clarity. The Appendix B chemistry results in a CF = 82°F; the Appendix D chemstry results in a CF = 108°F, and the fitted CF from Section 2 = 116.9°F. In accordance with Regulatory Guide 1.99, Revision 2, the has been reduced by 0.5.

[8] This heat is presented with the CF prior to adjustment and after adjustment in order to provide both sets of data.

3-21

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-4 Limerick Unit 2 Adjusted Reference Temperatures 40-Year License (32 EFPY)

Plates and Welds 2

Thickness = 6.19 inches 32 EFPY Peak I.D. fluence = 1.9E+18 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.3E+18 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.3E+18 n/cm Nozzle Forgings and Welds 2

Thickness = 6.19 inches 32 EFPY Peak I.D. fluence = 2.8E+17 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.9E+17 n/cm 2

32 EFPY Peak 1/4 T fluence = 1.9E+17 n/cm Initial 1/4 T 32 EFPY I 32 EFPY 32 EFPY Component Heat or Heat/Lot %Cu %Ni CF RTNDT Fluence RTNDT Margin Shift ART 2

°F n/cm °F °F °F °F PLATES:

Lower 14-1 B3312-1 0.13 0.58 90 10 1.3E+18 43 0 17 34 77 87 14-2 B3416-1 0.14 0.65 101 40 1.3E+18 48 0 17 34 82 122 14-3 C9621-2 0.15 0.60 110 22 1.3E+18 52 0 17 34 86 108 Lower-Intermediate 17-1 C9569-2 0.11 0.51 73 10 1.3E+18 35 0 17 34 69 79 17-2 C9526-1 0.11 0.56 74 10 1.3E+18 35 0 17 34 69 79 17-3 C9526-2 0.11 0.56 74 10 1.3E+18 35 0 17 34 69 79 WELDS:

Vertical [1]

BA,BB,BD,BE,BF 432A2671/H019A27A 0.04 1.08 54 -12 1.3E+18 26 0 13 26 51 39 BA,BC 03R728/L910A27A 0.03 0.92 41 -50 1.3E+18 19 0 10 19 39 -11 BA,BB,BC,BD,BE,BF 3P4000/3933 [2,7] 0.02 0.928 27 -50 1.3E+18 13 0 6 13 26 -24 BB 401Z9711/A022A27A 0.02 0.83 27 -50 1.3E+18 13 0 6 13 26 -24 BC 662A746/H013A27A [6] 0.03 0.88 41 -20 1.3E+18 19 0 10 19 39 19 BC 402A0462/B023A27A 0.02 0.90 27 -50 1.3E+18 13 0 6 13 26 -24 BD,BE 09L853/A111A27A 0.03 0.86 41 -50 1.3E+18 19 0 10 19 39 -11 BC,BD,BE,BF 07L669/K004A27A 0.03 1.02 41 -50 1.3E+18 19 0 10 19 39 -11 Girth AB 07L857/B101A27A 0.03 0.97 41 -6 1.3E+18 19 0 10 19 39 33 AB L83355/S411B27AD 0.03 1.08 41 -70 1.3E+18 19 0 10 19 39 -31 AB 402C4371/C115A27A 0.02 0.92 27 -50 1.3E+18 13 0 6 13 26 -24 AB 03M014/C118A27A 0.01 0.94 20 -34 1.3E+18 9 0 5 9 19 -15 AB 411A3531/H004A27A 0.02 0.96 27 -50 1.3E+18 13 0 6 13 26 -24 AB 09M057/C109A27A 0.03 0.89 41 -36 1.3E+18 19 0 10 19 39 3 AB 640892/J424B27AE 0.09 1.00 122 -60 1.3E+18 58 0 28 56 114 54 AB 401P6741/S419B27AG 0.03 0.92 41 -60 1.3E+18 19 0 10 19 39 -21 AB 412P3611/J417B27AF 0.03 0.93 41 -80 1.3E+18 19 0 10 19 39 -41 NOZZLES:

Water Level Instrumentation Forging SB166 [3] 0.11 0.51 73 10 1.9E+17 12 0 6 12 25 35 LPCI Weld KA 432A2671/H019A27A 0.04 1.08 54 -12 1.9E+17 9 0 5 9 18 6 Weld KA 07L669/K004A27A 0.03 1.02 41 -50 1.9E+17 7 0 3 7 14 -36 Weld KA C3L46C/J020A27A 0.02 0.87 27 -20 1.9E+17 5 0 2 5 9 -11 Weld KA 422B7201/L030A27A 0.04 0.90 54 -40 1.9E+17 9 0 5 9 18 -22 Weld KA 09L853/A111A27A 0.03 0.86 41 -50 1.9E+17 7 0 3 7 14 -36 Weld KA 4P4784/3930 (single wire) 0.06 0.87 82 -50 1.9E+17 14 0 7 14 28 -22 Weld KA 4P4784/3930 (tandem wire) 0.06 0.87 82 -20 1.9E+17 14 0 7 14 28 8 Forging 892L-1 Q2Q33W 0.15 0.83 115 -20 1.9E+17 19 0 10 19 39 19 Forging 892L-2 Q2Q33W 0.15 0.81 115 -6 1.9E+17 19 0 10 19 39 33 Forging 892L-3 Q2Q33W 0.15 0.82 115 -4 1.9E+17 19 0 10 19 39 35 Forging 892L-4 Q2Q33W 0.15 0.82 115 -20 1.9E+17 19 0 10 19 39 19 BEST ESTIMATE CHEMISTRIES per BWRVIP-135 R1:

BA,BB,BC,BD,BE,BF 3P4000/3933 0.02 0.935 27 -50 1.3E+18 13 0 6 13 26 -24 Weld [8] CTY538/A027A27A 0.03 0.83 41 -50 1.3E+18 19 0 10 19 39 -11 Weld 5P6756 [5,10] 0.08 0.936 108 -6 1.3E+18 51 0 26 51 102 96 Weld 5P6756 [5,10] 0.08 0.936 154 [9] -6 1.3E+18 73 0 14 28 101 95 INTEGRATED SURVEILLANCE PROGRAM (ISP):

Plate B0673-1 [4] 0.15 0.65 111 40 1.3E+18 53 0 17 34 87 127 Weld 5P6756 [5] 0.06/0.08 0.93/0.936 154 [9] -6 1.3E+18 73 0 14 28 101 95 Notes:

[1] Welds BA, BB, BC occur in the lower shell (Shell Ring #1) and welds BD, BE, BF occur in the lower-intermediate shell (Shell Ring #2).

[2] This material is evaluated below for the best estimate chemistry for this heat of material as provided in BWRVIP-135, Revision 1.

[3] The forging is Alloy 600 material; therefore, the properties for the lower-intermediate shell materials are used to represent the location of this nozzle.

[4] The ISP plate material is NOT the same heat as the target vessel plate material. The results provided are for information only and do not affect the Limerick Unit 2 PT curves.

[5] The ISP weld material is NOT the same heat as the target vessel weld material. The results provided are for information only and do not affect the Limerick Unit 2 PT curves.

[6] This heat number was provided in the DIR as 661A746; review of the CMTRs has determined this heat to be 662A746.

[7] 3P4000 data is available for both tandem and single wire; there is no change in %Cu. The limiting %Ni is used (single = 0.89; tandem = 0.95). The plant-specific %Ni used is 0.928; this value agrees with the NRC database RVID2.

[8] CMTR records do not indicate that this is a surveillance weld. However, the CMTRs demonstrate that this heat is a weld in the vessel; therefore, it is evaluated using the best estimate chemistry from BWRVIP-135 Revision 1.

[9] The Adjusted CF is calculated as: (108/82)

  • 116.9 = 154°F. Note that the chemistry values provide represent the BWRVIP-135 Revision 1 Appendix B values (first value) and the BWRVIP-135 Revision 1 Appendix D best estimate chemistry (second value). These are provided for clarity. The Appendix B chemistry results in a CF = 82°F; the Appendix D chemstry results in a CF = 108°F, and the fitted CF from Section 2 = 116.9°F. In accordance with Regulatory Guide 1.99, Revision 2, the has been reduced by 0.5.

[10] This heat is presented with the CF prior to adjustment and after adjustment in order to provide both sets of data.

3-22

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-5 Limerick 32 EFPY Effects of Irradiation on RPV Axial Weld Properties NRC Staff Assessment for Limerick Unit 1 Limerick Unit 2 Parameter 32 EFPY 32 EFPY 32 EFPY (Axial Welds )[4]

(CB&I RPV) (CB&I Vessel) (CB&I Vessel)

Cu% 0.10 0.058 0.04 Ni% 1.08 0.95 1.08 CF 135 79 54 19 Fluence at clad/weld interface (10 n/cm 2) 0.69 0.19 0.19 RT NDT(U) (°F) -30 -42 -12 RT NDT w/o margin (°F)(See Note 3) 121 44 30 Mean RT NDT (°F) 91 2 18 P (F/E) NRC (See Note 1) 1.42E-01 (Note 2) (Note 2)

Notes:

[1] P (F/E) stands for "Probability of a failure event."

[2] Although a conditional failure probability has not been calculated, the fact that the Limerick values at the end of license are less than the 32 EFPY value provided by the NRC leads to the conclusion that the Limerick RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL 98-05.

[3] RT NDT = CF

  • f (0.28 - 0.10 log f )

[4] This data is obtained from GL 98-05.

3-23

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-6 Limerick 32 EFPY Effects of Irradiation on RPV Circumferential Weld Properties NRC Staff Assessment for Limerick Unit 1 Limerick Unit 2 Parameter 32 EFPY 32 EFPY 32 EFPY (Circ Welds )[4]

(CB&I RPV) (CB&I Vessel) (CB&I Vessel)

Cu% 0.1 0.09 0.09 Ni% 0.99 1.00 1.00 CF 134.9 122 122 19 Fluence at clad/weld interface (10 n/cm 2) 0.51 0.19 0.19 RT NDT(U) (°F) -65 -60 -60 RT NDT w/o margin (°F)(See Note 3) 109.5 68 68 Mean RT NDT (°F) 44.5 8 8 P (F/E) NRC (See Note 1) 2.00E-07 (Note 2) (Note 2)

Notes:

[1] P (F/E) stands for "Probability of a failure event."

[2] Although a conditional failure probability has not been calculated, the fact that the Limerick values at the end of license are less than the 32 EFPY value provided by the NRC leads to the conclusion that the Limerick RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL 98-05.

(0.28 - 0.10 log f )

[3] RT NDT = CF

  • f

[4] This data is obtained from GL 98-05. The CF = 134.9°F was corrected in BWRVIP-05 SE dated 3/7/00 (previously shown to be 109.5°F).

3-24

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-7 CUF and P+Q Stress Range of Limiting Components P + Q Stress (Kips Per Square Inch (ksi)) CUF[4,5]

Component[1, 6] Current TPO Allowable Current TPO Allowable (3458 MWt) (3517 MWt)[3] (ASME Code (3458 MWt) (3517 MWt)[3] (ASME Limit) Code Limit) 79.6/ 79.6/

Feedwater Nozzle[7] 0. 9957 0. 9957 1.0

[2]

19.4 19.4[2,6]

53.1 Closure Flange 95.0/ 95.0/ 0.78 0.78 1.0 47.5[2] 47.5[2]

80.1 62.7/ 62.7/ 79.4 (2Sm)/

Closure Bolts 0.95 0.95 1.0 114.7 114.7 118.5 (3Sm)

Stabilizer Bracket 79.7 79.7 80.1 0.94 0.94 1.0 Support Skirt Units 115.0/ 115.0/ 0.83[4] 0.83 1.0 1&2 [2] [2]

69.7 69.7 80.1 Steam Outlet Nozzle 37.7 37.7 40.1 0.85 0.85 1.0 LPCI Nozzle -/68.5[2] -/68.5[2] 69.9 0.79 0.79 1.0 Core P &

100.3/ 100.3/

Liquid Control 0.71 0.71 1.0

[2]

44.2 44.2[2] 69.9 Nozzle Core Spray Nozzle 118.2/ 118.2/

69.9 0.510 0.510 1.0 (Low Alloy Steel) 55.8 55.8 Notes:

1. There are no changes in operating conditions from CLTP to TPO. Therefore, the CLTP evaluation remains applicable for TPO. The components presented in this table are consistent with the CLTP Safety Analysis Report (NEDC-32265P, Limerick Generating Station Units 1 and 2 Power Rerate Engineering Report, May 1994) to demonstrate that the results remain unchanged from CLTP to TPO.
2. Thermal Bending included/Thermal bending removed. P + Q stresses are acceptable per CLTP elastic-plastic analysis where applicable, which is valid for TPO conditions.
3. ((

))

4. Limiting CUF is presented.
5. Fatigue usage factors are for a 40-year license.
6. CLTP and TPO were (( )) Therefore, there is no change in values from CLTP to TPO.
7. Considering normal operating conditions (i.e., does not consider FFWTR or FWHOOS).

3-25

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-8 Governing Stress Results for RPV Internal Components Current Service Stress/ Load Item Component Unit Design TPO Allowable Level(3) Category Basis(1) 1 Shroud Support Bounded by the existing design basis. The component is qualified for TPO.

2 Shroud Bounded by the existing design basis. The component is qualified for TPO.

Bend.

Beam B Mom., 1.21E6 1.21E6 1.51E6 Buckling In-lbs 3 Core Plate Bend.

Beam D Mom., 1.65E6 1.67E6 3.01E6 Buckling In-lbs B Pm+Pb ksi 17.57 17.63 25.35 4 Top Guide D Pm+Pb ksi 35.26 36.20 50.70 5 CRDH Bounded by the existing design basis. The component is qualified for TPO.

B Pm+Pb ksi 14.44 14.82 24.00 6 CRGT C Pm+Pb ksi 20.39 20.92 36.00 D Pm+Pb ksi 32.39 33.23 38.40 7 Orificed Fuel Support Bounded by the existing design basis. The component is qualified for TPO.

8 FW Sparger Bounded by the existing design basis. The component is qualified for TPO.

(2) 9 Jet Pump Bounded by the existing design basis. The component is qualified for TPO.

Core Spray Line &

10 Bounded by the existing design basis. The component is qualified for TPO.

Sparger 11 Access Hole Cover Bounded by the existing design basis. The component is qualified for TPO.

Shroud Head And 12 Bounded by the existing design basis. The component is qualified for TPO.

Separator In-Core Housing and 13 Bounded by the existing design basis. The component is qualified for TPO.

Guide Tube (ICH&GT)

Vessel Head Cooling 14 Bounded by the existing design basis. The component is qualified for TPO.

Spray Nozzle Core DP & Liquid 15 Bounded by the existing design basis. The component is qualified for TPO.

Control Line 16 LPCI Coupling Bounded by the existing design basis. The component is qualified for TPO.

Notes:

(1). 110% OLTP and 110% ICF.

(2). The mechanical evaluation of jet pumps is evaluated in a separate report by GEH for TPO.

(3). Service level A is bounded by service level B.

3-26

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-9 Piping Lines Recommended for Special Focus under FAC Review Line Name Comment 30 GBD-109, -209 These lines are predicted to exceed recommended 10 GBD-111, -211 flow velocity guidelines under TPO conditions.

30 GBD-111, -211 Flow velocities that exceed these guidelines are 10 GBD-116, -216 acceptable, provided the FAC program is updated to 20 GBD-116, -216 ensure adequate inspection frequencies. The station 30 GBD-116, -216 FAC program will be updated to include the effects 24 GBD-179, -279 of TPO conditions as necessary.

30 GBD-179, -279 24 GBD-180, -280 30 GBD-180, -280 20 GBD-117, -217 20 GBD-118, -218 18 GBD-118, -218 20 DBD-101, -201 18 GAD-103, -203 10 HAD-101, -201 16 HAD-103, -203 24 HAD-103, -203 20 GAD-107, -207 3-27

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4.0 ENGINEERED SAFETY FEATURES 4.1 CONTAINMENT SYSTEM PERFORMANCE TLTR Appendix G presents the methods, approach, and scope for the TPO uprate containment evaluation for LOCA. The current containment evaluations were performed at 102% of CLTP.

Although the nominal operating conditions change slightly because of the TPO uprate, the required initial conditions for containment analysis inputs remain the same as previously documented.

The following table summarizes the effect of the TPO uprate on various aspects of the containment system performance.

Topic Key Parameters TPO Effect Short Term Pressure and Temperature Response Gas Temperature Break Flow and Energy Pressure Break Flow and Energy Long-Term Suppression Pool Temperature Response Bulk Pool Decay Heat Current Analysis Local Temperature with Decay Heat Based on 102% of CLTP SRV Discharge Containment Dynamic Loads Loss-of-Coolant Break Flow and Energy Accident Loads Safety-Relief Valve Decay Heat Loads Sub compartment Break Flow and Energy Pressurization Containment Isolation The ability of containment isolation valves Section 4.1.1 provides and operators to perform their required confirmation that MOVs are functions is not affected because the capable of performing design evaluations have been performed at 102%

basis functions at TPO of CLTP.

conditions.

4.1.1 Generic Letter 89-10 Program The motor-operated valve (MOV) requirements in the UFSAR were reviewed, and no changes to the functional requirements of the GL 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance, MOVs are identified as a result of operating at the TPO RTP level. Because previous analyses were either based on 102% of CLTP or are consistent with the plant conditions expected to result from TPO, there are no increases in the pressure or temperature at which 4-1

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION MOVs are required to operate. Therefore, the GL 89-10 MOVs remain capable of performing their design basis functions.

4.1.2 Generic Letter 95-07 Program The evaluation performed in support of GL 95-07, Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves, has been reviewed and no changes are identified as a result of operating at the TPO RTP level. The criteria for susceptibility to pressure locking or thermal binding were reviewed and it was determined that the slight changes in operating or environmental conditions expected to result from the TPO uprate would have no impact on the functioning of power-operated gate valves within the scope of GL 95-07. Therefore, the valves remain capable of performing their design basis functions.

4.1.3 Generic Letter 96-06 The Limerick response to GL 96-06, Assurance of Equipment Operability and Containment Integrity during Design-Basis Accident Conditions, was reviewed for the TPO uprate. The containment design temperatures and pressures in the current GL 96-06 evaluation are not exceeded under post-accident conditions for the TPO uprate. Therefore, the Limerick response to GL 96-06 remains valid under TPO uprate conditions.

4.1.4 Containment Coatings The qualified coatings in primary containment are qualified such that they do not fail when exposed to the existing maximum post-LOCA primary containment operating conditions of 340°F, 44.0 psig, and 100% relative humidity. These operating conditions bound those, which are expected after implementation of TPO since the current operating conditions are based on 102% of CLTP.

4.2 EMERGENCY CORE COOLING SYSTEMS 4.2.1 High Pressure Coolant Injection The High Pressure Coolant Injection (HPCI) system is a turbine driven system designed to pump water into the reactor vessel over a wide range of operating pressures. For the TPO uprate, there is no change to the nominal reactor operating pressure or the SRV setpoints. The primary purpose of the HPCI system is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the RPV. The generic evaluation of the HPCI system provided in TLTR Section 5.6.7 is applicable to Limerick. The ability of the HPCI system to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the HPCI system are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

4-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4.2.2 High Pressure Core Spray The High Pressure Core Spray system is not applicable to Limerick.

4.2.3 Core Spray The Core Spray (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 makeup for a large break LOCA and for any small break LOCA after the RPV has depressurized. It also provides spray cooling for long-term core cooling in the event of a LOCA. The generic evaluation of the CS system provided in TLTR Section 5.6.10 is applicable to Limerick. The ability of the CS system to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the CS system are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

4.2.4 Low Pressure Coolant Injection 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 provide reactor vessel coolant makeup during a large break LOCA or small break LOCA after the RPV has depressurized. The generic evaluation of the LPCI mode provided in TLTR Section 5.6.4 is applicable to Limerick. The ability of the RHR system to perform required safety functions required by the LPCI mode is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the RHR system LPCI mode are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

4.2.5 Automatic Depressurization System The Automatic Depressurization System (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 CS and LPCI to inject coolant into the RPV. The ADS initiation logic and valve control is not affected by the TPO uprate. The generic evaluation of the ADS provided in TLTR Section 5.6.8 is applicable to Limerick. The ability of the ADS to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the ADS are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

4.2.6 ECCS Net Positive Suction Head The most limiting case for NPSH typically occurs at the peak long-term suppression pool temperature. The generic evaluation of the containment provided in TLTR Appendix G is applicable to Limerick. The CLTP containment analyses were based on 102% of CLTP and there is no change in the available NPSH for systems using suppression pool water. Therefore, the TPO uprate does not affect compliance with the ECCS pump NPSH requirements.

4-3

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE The ECCS is designed to provide protection against a postulated LOCA caused by ruptures in the primary system piping. The current 10 CFR 50.46, or LOCA, analyses for the Limerick plant have been performed at 102% of CLTP, consistent with Appendix K. Table 4-1 shows the results of the Limerick ECCS-LOCA analysis. The ECCS-LOCA results for Limerick are in conformance with the error reporting requirements of 10 CFR 50.46. Therefore, the CLTP LOCA analysis for GE14 fuel bounds the TPO uprate for Limerick.

Reference 16 provides justification for the elimination of the 1600°F Upper Bound PCT (UBPCT) limit and generic justification that the Licensing Basis PCT (LBPCT) will be conservative with respect to the UBPCT. The NRC SER for Reference 16 accepted this position, noting that because plant-specific UBPCT calculations have been performed for all plants, other means may be used to demonstrate compliance with the original SER requirements. These other means are acceptable provided there are no significant changes to a plants configuration that would invalidate the existing UBPCT calculations. Reference 17 provided justification for the elimination of the UBPCT limit for Limerick Units 1 and 2.

For the TPO uprate there are no changes to the plant configuration that would invalidate the Reference 17 evaluation for conformance with Reference 16.

The CLTP LOCA analysis for GE14 fuel is concluded to bound the TPO uprate for Limerick.

4.4 MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM The Main Control Room atmosphere is not affected by the TPO uprate. Control Room habitability following a postulated accident at TPO conditions is unchanged because the Main Control Room Atmosphere Control System has previously been evaluated for radiation release accident conditions at 102% of CLTP. Therefore, the system remains capable of performing its safety function at the TPO conditions.

4.5 STANDBY GAS TREATMENT SYSTEM The Standby Gas Treatment System (SGTS) minimizes the offsite and control room dose rates during venting and purging of the containment atmosphere under abnormal conditions. The current capacity of the SGTS was selected to maintain the secondary containment at a slightly negative pressure during such conditions. This capability is not changed by the TPO uprate conditions. The SGTS can accommodate design basis accident (DBA) conditions at 102% of CLTP. Therefore, the system remains capable of performing its safety function for the TPO uprate condition.

4.6 MAIN STEAM ISOLATION VALVE LEAKAGE ALTERNATE DRAIN PATHWAY The MSIV Leakage Alternate Drain Pathway prevents a direct release of fission products that could leak through the closed MSIVs after a LOCA. The pathway provides control by providing 4-4

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION a hold-up volume for the MSIV leakage before release to the atmosphere. This is accomplished by directing the leakage through existing Main Steam Drain Lines to the High Pressure Shell of the Main Condenser.

This system was previously evaluated at 102% and is therefore acceptable for TPO operations.

4.7 POST-LOCA COMBUSTIBLE GAS CONTROL SYSTEM The Combustible Gas Control System (CGCS) maintains the post-LOCA concentration of oxygen or hydrogen in the containment atmosphere below the flammability limit. The generic evaluation of the CGCS provided in TLTR Section J.2.3.10 is applicable to Limerick Units 1 and

2. The metal available for reaction is unchanged by the TPO uprate and the hydrogen production due to radiolytic decomposition is unchanged because the system was previously evaluated for accident conditions from 102% of CLTP. Therefore, the current evaluation is valid for the TPO uprate.

4-5

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 4-1 Limerick ECCS-LOCA Analysis Results for GE14 Fuel Parameter MELLLA Analysis Limit Nominal PCT 1007°F N/A Appendix K PCT 1666°F < 2200°F*

LBPCT 1675°F < 2200°F*

Maximum Local

<1.0% 17%*

Oxidation Core-Wide Metal-

<0.1% 1.0%*

Water Reaction

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.0 INSTRUMENTATION AND CONTROL 5.1 NSSS MONITORING AND CONTROL The instruments and controls that directly interact with or control the reactor are usually considered within the NSSS. The NSSS process variables and instrument setpoints that could be affected by the TPO uprate were evaluated.

5.1.1 Neutron Monitoring System 5.1.1.1 Average Power Range Monitors, Intermediate Range Monitors, and Source Range Monitors The Average Power Range Monitors (APRMs) are re-calibrated to indicate 100% at the TPO RTP level of 3515 MWt. The APRM high flux scram and the upper limit of the rod block setpoints, expressed in units of percent of licensed power, are not changed. The flow-biased APRM trips, expressed in units of absolute thermal power (i.e., MWt), remain the same.

However, in order to accommodate limits in the Stability Region, new flow-biased APRM Analytical Limits (ALs) were established that conservatively bound the entire operating envelope. This approach for the Limerick TPO uprate follows the guidelines of TLTR Section 5.6.1 and Appendix F, which is consistent with the practice approved for GE BWR uprates in ELTR1 (Reference 2).

For the TPO uprate, no adjustment is needed to ensure the Intermediate Range Monitors (IRMs) have adequate overlap with the SRMs and APRMs. However, normal plant surveillance procedures may be used to adjust the IRMs, the overlap with the SRMs and the APRMs. The IRM channels have sufficient margin to the upscale scram trip on the highest range when the APRM channels are reading near their downscale alarm trip because the change in APRM scaling is so small for the TPO uprate.

5.1.1.2 Local Power Range Monitors and Traversing In Core Probes At the TPO RTP level, the flux at some LPRMs increases. However, the small change in the power level is not a significant factor to the neutronic service life of the LPRM detectors and radiation level of the traversing in core probes (TIPs). It does not change the number of cycles in the lifetime of any of the detectors. The LPRM accuracy at the increased flux is within specified limits, and the LPRMs are designed as replaceable components. The TIPs are stored in shielded rooms. The radiation protection program for normal plant operation can accommodate a small increase in radiation levels.

5.1.1.3 Rod Block Monitor The Rod Block Monitor (RBM) instrumentation is referenced to an APRM channel. Because the APRM has been rescaled, there is only a small effect on the RBM performance due to the LPRM 5-1

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION performance at the higher average local flux. The RBM instrumentation is not significantly affected by the TPO uprate conditions, and no change is needed.

5.1.2 Rod Worth Minimizer The Rod Worth Minimizer (RWM) 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 power-dependent setpoints for the RWM are discussed in Section 5.3.8.

5.2 BOP MONITORING AND CONTROL Operation of the plant at the TPO RTP level has minimal effect on the BOP system instrumentation and control devices. The improved FW flow measurement, which is the basis for the reduction in power uncertainty, is addressed in Section 1.4. All instrumentation with control functions has sufficient range/adjustment capability for use at the TPO uprate conditions.

No safety-related BOP system setpoint changes are required as a result of the TPO uprate. The plant-specific instrumentation and control design and operating conditions are bounded by those used in the evaluations contained in the TLTR.

5.2.1 Pressure Control System The Pressure Control System (PCS) provides a fast and stable response to steam flow changes so that reactor pressure is controlled within allowable values. The PCS consists of the pressure regulation system, turbine control valve system and steam bypass valve system. The main turbine speed/load control function is performed by the main turbine-generator Electro-hydraulic Control (EHC) system. The steam bypass valve pressure control function is performed by the Turbine Bypass Control System (TBCS).

Satisfactory reactor pressure control by the pressure regulator and the turbine control valves (TCVs) requires an adequate flow margin between the TPO RTP operating condition and the steam flow capability of the TCVs at their maximum stroke (i.e., valves wide open (VWO)).

Limerick will modify or replace the first stage nozzle plate on the main turbine in order to maintain adequate flow margin at TPO conditions. The existing electronic controls, as designed for the current 100% of CLTP conditions, were evaluated. The results of the analysis indicated that the performance of the TCVs is not impacted by increasing power level to TPO conditions; therefore, no modifications to the electronic components are required.

No modification is required to the steam bypass valves. No modifications are required to controls or alarm annunciators provided in the main control room. The appropriate control room indicators will be adjusted, as necessary, to reflect 100% TPO power. The required adjustments are limited to tuning of the control settings that may be required to operate optimally at the TPO uprate power level.

5-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.2.2 EHC Turbine Control System The turbine EHC system was reviewed for the increase in core thermal power and associated

~2% increase in rated steam flow. The control system is expected to perform normally for TPO RTP operation. Normal operator controls are used in conjunction with the associated operating procedures. Confirmation testing will be performed during power ascension (Section 10.4).

5.2.3 Feedwater Control System An evaluation of the ability of the FW level control system, FW control valves, and/or FW turbine controls to maintain adequate water level control at the TPO uprate conditions has been performed. The ~2% increase in FW flow associated with TPO uprate is within the current control margin of these systems. No changes in the operating reactor water level or reactor water level trip set points are required for the TPO uprate. Per the guidelines of TLTR Appendix L, the performance of the FW level control systems will be recorded at 95% and 100% of CLTP and confirmed at the TPO power during power ascension. These checks will demonstrate acceptable operational capability and will utilize the methods and criteria described in the original startup testing of these systems.

5.2.4 Leak Detection System The setpoints associated with leak detection have been evaluated with respect to the ~2% higher steam flow and ~2°F increase in FW temperature for the TPO uprate. Each of the systems, where leak detection potentially could be affected, is addressed below.

Main Steam Tunnel Temperature Based Leak Detection The ~2°F increase in FW temperature for the TPO uprate decreases the leak detection trip avoidance margin. As described in TLTR Section F.4.2.8, the high steam tunnel temperature setpoint remains unchanged.

RWCU System Temperature Based Leak Detection There is no significant effect on RWCU system temperature or pressure due to the TPO uprate.

Therefore, there is no effect on the RWCU temperature based leak detection.

RCIC System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the RCIC system temperature or pressure, and thus, the RCIC temperature based leak detection system is not affected.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION HPCI System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the HPCI system temperature or pressure, and thus, the HPCI temperature based leak detection system is not affected.

RHR System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the RHR system temperature or pressure, and thus, the RHR temperature based leak detection system is not affected.

Non-Temperature Based Leak Detection The non-temperature based leak detection systems are not affected by the TPO uprate.

5.3 TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS The determination of instrument setpoints is based on plant operating experience, conservative licensing analyses or limiting design/operating values. Standard GE setpoint methodologies (References 18 and 19) are used to generate the allowable values (AV) and nominal trip setpoints (NTSP) related to any AL change, as applicable. Each actual trip setting is established to preclude inadvertent initiation of the protective action, while assuring adequate allowances for instrument accuracy, calibration, drift and applicable normal and accident design basis events.

Table 5-1 lists the ALs that change based on results from the TPO evaluations and safety analyses. In general, if the AL does not change in the units shown in the Technical Specifications, then no change in its associated plant AV and NTSP is required, as shown in the Technical Specifications. Changes in the setpoint margins due to changes in instrument accuracy and calibration errors caused by the change in environmental conditions around the instrument due to the TPO uprate are negligible. Maintaining constant nominal dome pressure for the TPO uprate minimizes the potential effect on these instruments by maintaining the same fluid properties at the instruments. The setpoint evaluations are based on the guidelines in TLTR Sections 5.8 and F.4 and on Section 5.3 of Reference 18.

5.3.1 High-Pressure Scram The high-pressure scram terminates a pressure increase transient not terminated by direct or high flux scram. Because there is no increase in nominal reactor operating pressure with the TPO uprate, the scram AL on reactor high pressure is unchanged.

5.3.2 Hydraulic Pressure Scram The AL for the turbine hydraulic pressure (low oil pressure trip) that initiates the Turbine-Generator (T/G) trip scram at high power remains the same as for the CLTP. No modifications 5-4

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION are being made to the turbine hydraulic control systems for TPO; actuation of these safety functions remain unchanged for TPO.

5.3.3 High-Pressure Recirculation Pump Trip The ATWS-RPT trips the pumps during plant transients with increases in reactor vessel dome pressure. The ATWS-RPT provides negative reactivity by reducing core flow during the initial part of an ATWS. The evaluation in Section 9.3.1 demonstrates that the current high pressure ATWS-RPT AL is acceptable for the TPO uprate.

5.3.4 Safety Relief Valve Because there is no increase in reactor operating dome pressure, the SRV ALs are not changed.

5.3.5 Main Steam Line High Flow Isolation The Technical Specification AV of this function is expressed in terms of psid. The corresponding percent thermal power is not changed. The existing AV setpoint in terms of psid has sufficient trip avoidance margin to support the small increase in the TPO rated steam flow.

Therefore, the AV at the TPO remains unchanged from CLTP.

Because of the large spurious trip margin, sufficient margin to the trip setpoint exists to allow for normal plant testing of the MSIVs and turbine stop and control valves. This is consistent with TLTR Section F.4.2.5.

5.3.6 Fixed APRM Scram The fixed APRM ALs, for both two (recirculation) loop operation (TLO) and SLO, expressed in percent of RTP do not change for the TPO uprate. The generic evaluation and guidelines presented in TLTR Section F.4.2.2 are applicable to Limerick. The limiting transient that relies on the fixed APRM trip is the vessel overpressure transient (MSIV closure) with indirect scram.

This event has been analyzed assuming 102% of CLTP and is reanalyzed on a cycle specific basis.

5.3.7 APRM Flow-Biased Scram The flow-referenced APRM ALs, for both TLO and SLO, are unchanged in units of absolute core thermal power versus recirculation drive flow. Because the setpoints are expressed in percent of RTP, they decrease in proportion to the power uprate or CLTP RTP / TPO RTP. This is the same approach taken for generic BWR uprates described in ELTR1 (Reference 2). There is no significant effect on the instrument errors or uncertainties from the TPO uprate. Therefore, the AV and NTSP are established by directly incorporating the change in the AL.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.3.8 Rod Worth Minimizer Low Power Setpoint The Rod Worth Minimizer (RWM) Low Power Setpoint (LPSP) is used to enforce the rod patterns established for the control rod drop accident at low power levels. The generic guidelines in TLTR Section F.4.2.9 are applicable to Limerick. The RWM LPSP AL is kept the same in terms of percent power, and is therefore higher in terms of absolute power. This new higher absolute power is conservative for the RWM LPSP.

5.3.9 Rod Block Monitor The severity of the Rod Withdrawal Error (RWE) during power operation event is dependent upon the RBM rod block setpoint. The power-dependent ALs are maintained at the same percent power. The cycle specific reload analysis is used to determine any changes in the rod block setpoint.

5.3.10 Flow-Biased Rod Block Monitor (%RTP)

Limerick does not have a flow-biased RBM system.

5.3.11 Main Steam Line High Radiation Isolation The MSL high radiation signal isolation function has been eliminated at Limerick.

5.3.12 Low Steam Line Pressure MSIV Closure (RUN Mode)

The purpose of this function is to initiate MSIV closure on low steam line pressure when the reactor is in the RUN mode. This AL is not changed for the TPO as discussed in TLTR Section F.4.2.7.

5.3.13 Reactor Water Level Instruments As described in TLTR Section F.4.2.10, the TPO uprate does not result in a significant increase in the possibility of a reactor scram, equipment trip, or ECCS actuation. Use of the current ALs maintains acceptable safety system performance. The low reactor water level Technical Specification setpoints for scram, high-pressure injection, and ADS/ECCS are not changed for the TPO uprate. The high water level ALs for trip of the main turbine, FW pumps, and reactor scram are not changed for the TPO uprate.

Water level change during operational transients (e.g., trip of a recirculation pump, FW controller failure, loss of one FW pump) is slightly affected by the TPO uprate. The plant response following the trip of one FW pump does not change significantly, because the maximum operating rod line is not being increased. Therefore, the final power level following a single FW pump trip at TPO uprate conditions would not change relative to the remaining FW flow as exists at CLTP.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 5.3.14 Main Steam Line Tunnel High Temperature Isolations As noted in Section 5.2.4 above, the high steam tunnel temperature AL remains unchanged for the TPO uprate.

5.3.15 Low Condenser Vacuum In order to produce more electrical power, the amount of heat discharged to the main condenser increases slightly. This added heat load may slightly increase condenser backpressure, but the increase would be insignificant (< 0.15 in. HgA). The slight change in condenser vacuum after implementation of TPO will not adversely affect any trip setpoints associated with low condenser vacuum (turbine trip / MSIV closure).

5.3.16 TSV Closure Scram, TCV Fast Closure Scram Bypasses The turbine first-stage pressure bypass allows the Turbine Stop Valve (TSV) closure scram and Turbine Control Valve (TCV) fast closure scram to be bypassed, when reactor power is sufficiently low, such that the scram functions are not needed to mitigate a T/G trip. This power level is the AL for determining the actual trip setpoint, which comes from the turbine first-stage pressure (TFSP). The TFSP setpoint is chosen to allow operational margin so that scrams can be avoided, by transferring steam to the turbine bypass system during T/G trips at low power.

Based on the guidelines in TLTR Section F.4.2.3, the TSV closure scram, TCV fast closure scram, and End of Cycle (EOC)-RPT bypass AL in percent of RTP is reduced by the ratio of the power increase. The new AL does not change with respect to absolute thermal power. ((

)) The maneuvering range for plant startup is maximized.

The High Pressure turbine first stage nozzle plates for Limerick, Units 1 and 2, will be modified for TPO operation as described in Section 7.1. The first stage pressure setpoint for the bypass of the TSV closure scram, TCV fast closure scram, and EOC-RPT will be adjusted as necessary to implement the new AL.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 5-1 Analytical Limits that Change due to TPO Parameter Current TPO Justification APRM High Neutron Flux Scram 121 No change APRM Flow-biased Scram Fixed (%RTP) 119 No change (1)(2)

TLO Flow-biased (%RTP) 0.66W + 66.2 0.65W + 65.1 (3)

(1)(2)

SLO Flow-biased (%RTP) 0.66W + 61.2 0.65(W - W) + 65.1 (3) 0.65W + 60.1 APRM Flow-biased Rod Block AVs (1)

Fixed (%RTP) 108.4 No change (1)(2)

TLO Flow-biased (%RTP) 0.66W + 55.7 0.65W + 54.8 (3)

(1)(2)

SLO Flow-biased (%RTP) 0.66W + 50.7 0.65(W - W) + 54.8 (3) 0.65W + 49.8 TSV & TCV Scram & EOC-RPT Bypasses 30 29.5 (4)

(%RTP)

MSL High Flow Isolation

% rated steam flow 140 140 (4) psid 126.5 128.9 Rod Worth Minimizer LPSP (%RTP) 10 10 (4)

LPSP (Steam flow) 1.199 Mlbm/hr 1.222 Mlbm/hr Notes:

(1) No credit is taken in any safety analysis for the flow-biased setpoints.

(2) W is % recirculation drive flow where 100% drive flow is that required to achieve 100% core flow at 100% power, and W is the difference between the TLO and SLO drive flow at the same core flow. The current value of W is 7.6% and is not changed.

(3) These changes to the ALs are based upon the methodology approved by the NRC in Reference 1.

(4) All limits scaled for an uprate of 1.65% thermal.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.0 ELECTRICAL POWER AND AUXILIARY SYSTEMS 6.1 AC POWER Plant electrical characteristics are given in Table 6-1.

A detailed comparison of existing ratings with uprated ratings and the effect of the power uprate on the main generator, generator step-up transformers, unit auxiliary transformer, station auxiliary transformer, and regulating transformer are shown in Tables 6-2, 6-3a, 6-3b, 6-4, 6-5a, and 6-5b.

6.1.1 Off-Site Power The generator, main transformer and isolated phase bus nameplate ratings are listed below:

  • Generator: The generator is a direct-driven 3-phase 60-HZ, 22,000-Volt, 1800-rpm, hydrogen inner-cooled, synchronous generator rated for: 1,264,970 kVA at a 0.9 power factor (PF), with a .58 short circuit ratio at a nominal hydrogen pressure of 75 psig.
  • Unit 1 Main Transformer: The 1266 Million Volt Amps (MVA) MPT consists of three single-phase, 20.9-230 kV 422 MVA, oil immersed, forced air-cooled, 65°C rise, 60 Hz, outdoor Westinghouse transformers.
  • Unit 2 Main Transformer: The 1575 MVA MPT consists of three single-phase,22-531.6 kV 525 MVA, oil immersed, forced air-cooled, 65°C rise, 60 Hz, outdoor ABB transformers.
  • Isolated Phase Bus Duct: The isolated phase bus duct continuous current rating is based on a 65°C rise above a 40°C ambient with forced air cooling. The Main bus is rated at 35,000A, the Auxiliary Branch bus is rated at 2,000A and the transformer Delta bus is rated at 20,000A. The momentary fault current rating for the Main and Delta buses is 350,000A and the Auxiliary Branch is 550,000A. The voltage rating of the system is 25,000V. The forced cooling is handled by two air handling units with a design heat transfer capacity of 1,810,000 Btu/hr.

The review of the existing off-site electrical equipment concluded the following:

  • The Main Generators will be operating within the existing generating capability curve for TPO uprate. For summer and winter operations, the gross generator MWe output is on the existing generator capability curve with greater than 0.9PF.
  • The isolated phase bus duct is adequate for both rated voltage and low voltage current output.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION

  • The main transformers and the associated switchyard components (rated for maximum transformer output) are adequate for the TPO uprate-related transformer output.

A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criterion (GDC) 17 (10 CFR 50, Appendix A).

GDC-17 addresses on-site and off-site electrical supply and distribution systems for safety-related components. There is no significant effect on grid stability or reliability. There are no modifications associated with the TPO uprate, which would increase electrical loads beyond those levels previously included or revise the logic of the distribution systems.

6.1.2 On-Site Power The on-site power distribution system consists of transformers, numerous buses, and switchgear.

Alternating current (AC) power to the distribution system is provided from the transmission system or from onsite diesel generators. The on-site power distribution system loads were reviewed under both normal and emergency operating scenarios. In both cases, loads are computed based primarily on equipment nameplate data or brake horsepower (BHP). These loads are used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP level is achieved in both normal and emergency conditions by operating equipment at or below the nameplate rating running KW or BHP. Therefore, there are negligible changes to the load, voltage drop or short circuit current values.

The only identifiable change in electrical load demand is associated with the condensate pumps.

These pumps experience increased flow and a small change in horsepower duty (~1%) due to the TPO uprate conditions. Accordingly, there are negligible changes in the on-site distribution system design basis loads or voltages due to the TPO conditions. The system environmental design bases are unchanged. Operation at the TPO RTP level is achieved by utilizing existing equipment operating at or below the nameplate rating; therefore, under normal conditions, the electrical supply and distribution components (e.g., switchgear, motor control centers (MCCs),

cables) are adequate.

Station loads under emergency operation and distribution conditions (emergency diesel generators) are based on operational requirements. The ECCS pump loading is based on station UFSAR design basis requirements. Emergency operation at the TPO RTP levels is achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the stated pumps. Therefore, under emergency conditions the electrical supply and distribution components are adequate.

No increase in flow or pressure is required of any AC-powered ECCS equipment for the TPO uprate. Therefore, the amount of power required to perform safety-related functions (pump and valve loads) does not increase, and the current emergency power system remains adequate. The systems have sufficient capacity to support all required loads for safe shutdown, to maintain a 6-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION safe shutdown condition, and to operate the engineered safety feature equipment following postulated accidents.

Since the duty cycle and duration for design basis EDG loads is based on analytical power levels of at least 102% of the current licensed thermal power, these will remain unchanged by TPO.

Hence, the required reserve volume of emergency fuel oil is not changed. Therefore, useable emergency fuel oil reserves will be adequate to support TPO.

6.2 DC POWER The direct current (DC) loading requirements documented in the UFSAR and station load calculations were reviewed, and no reactor power-dependent loads were identified. The DC power distribution system provides control and motive power for various systems and components. These loads are used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP level does not increase any loads or revise control logic. Therefore, there are no changes to the load, voltage drop or short circuit current values.

6.3 FUEL POOL The following subsections address fuel pool cooling, crud and corrosion products in the fuel pool, radiation levels and structural adequacy of the fuel racks. The changes due to TPO are within the design limits of the system and its components. The fuel pool cooling system meets the UFSAR requirements at the TPO conditions.

6.3.1 Fuel Pool Cooling The Spent Fuel Pool (SFP) heat load increases slightly at TPO but remains within the capability of the Fuel Pool Cooling and Cleanup System (FPCC) as supplemented by the Residual Heat Removal System (RHR). The TPO uprate does not affect the heat removal capability of the FPCC or RHR systems and the TPO heat load is within the design basis heat load for the FPCC and RHR systems as shown in Table 6-6.

The FPCC and RHR heat exchangers are sufficient to remove the decay heat during a normal batch and full-core refueling. The SFP makeup requirement increases slightly at TPO, however makeup capacity remains sufficient.

6.3.2 Crud Activity and Corrosion Products The crud activity and corrosion products associated with spent fuel can increase very slightly due to the TPO. The increase is insignificant and SFP water quality is maintained by the FPCC.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.3.3 Radiation Levels The normal radiation levels around the SFP may increase slightly during fuel handling operation.

This increase is acceptable and does not significantly increase the operational doses to personnel or equipment.

6.3.4 Fuel Racks There is no effect on the design of the fuel racks because the maximum allowable spent fuel temperature is not being increased.

6.4 WATER SYSTEMS The safety-related and non-safety-related cooling water loads potentially affected by TPO are addressed in the following sections. The environmental effects of TPO are controlled such that none of the present limits (e.g., maximum allowed cooling water discharge temperature) are increased.

6.4.1 Service Water Systems 6.4.1.1 Safety-Related Loads Emergency Service Water The safety-related Emergency Service Water (ESW) system provides cooling water to essential equipment during and following a design basis accident, such as a Loss of Offsite Power (LOOP) or LOCA. The performance of the ESW system during these events does not change for TPO because the original LOCA analysis and containment response analysis were based on 102% of CLTP, the bounding power level for the TPO analysis. The increases in the heat loads to equipment cooled by ESW are within the existing capacity of the ESW system.

Residual Heat Removal Service Water The required design performance of the Residual Heat Removal Service Water System (RHRSW) does not change for TPO because the original LOCA analysis and containment response analysis were based on at least 102% of CLTP, the bounding analytical power level for TPO. The increases in the heat loads to equipment cooled by RHRSW are within the existing capacity of the RHRSW system.

6.4.1.2 Non-safety-Related Loads The major operational heat load increases to the Service Water (SW) system from TPO reflect an operational increase in main generator losses rejected to the generator hydrogen coolers, generator stator coolers, Alterrex air cooler, and Iso-phase bus coolers. The resulting design heat loads for the SW system are <0.2% above CLTP. The increases in heat loads to equipment 6-4

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION cooled by the SW system are insignificant and the design of the system is adequate to accommodate for TPO.

6.4.2 Main Condenser/Circulating Water/Normal Heat Sink Performance The main condenser, circulating water, and normal heat sink systems are designed to remove the heat rejected to the condenser and thereby maintain adequately low condenser pressure as recommended by the turbine vendor. TPO operation increases the heat rejected to the condenser and may reduce the difference between the operating pressure and the required minimum condenser vacuum. The performance of the main condenser was evaluated for operation at the TPO RTP. The evaluation confirms that the condenser, circulating water system and heat sink are adequate for TPO operation. Note that based on weather conditions, unit derates may be necessary to maintain adequate condenser backpressure.

6.4.2.1 Discharge Limits The Pennsylvania Department of Environmental Quality National Pollutant Discharge Elimination System (NPDES) Permit provides the effluent limitations and monitoring requirements for discharge wastewater at the site. The discharge limits on residual oxides and Spectrus CT1300 are daily maximums of 0.2 mg/l and 0.4 mg/l (respectively). The discharge water temperature shall not exceed a maximum of 110°F, or raise the river temperature 5°F above its ambient temperature. When river temperatures exceed 87°F, discharge from the plant cannot raise the river temperature at all. Frequent monitoring of these parameters ensures that permit limits are not exceeded. The TPO uprate has minimal effect on the parameters, and no changes to NPDES permit requirements are needed.

The state thermal discharge limits, the current discharges, and bounding analysis discharges for the TPO uprate are shown in Table 6-7. This comparison demonstrates that the plant remains within the state discharge limits, during operation at TPO conditions.

6.4.3 Reactor Enclosure Cooling Water System The heat loads on the Reactor Enclosure Cooling Water (RECW) system do not increase significantly due to TPO. The main power-dependent heat loads on the RECW system that are increased by the TPO, are those related to the operation of the Reactor Water Cleanup non-regenerative heat exchangers, Reactor Water Cleanup recirculation pumps, and the reactor recirculation pumps. The design of the RECW heat exchangers is adequate to accommodate a heat load increase of <1.2% for normal operations and ~2.3% for emergency operations.

Changes to the RECW system heat loads are minimal and will result in a negligible temperature increase of ~0.4°F for the RECW system during normal operation. The RECW system experiences a slight heat load increase associated with the Fuel Pool Coolers heat exchangers during emergency operations. However, the system has adequate design margin to remove the additional heat. Therefore, the RECW system is acceptable for the TPO uprate.

6-5

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.4.4 Turbine Enclosure Cooling Water System The power-dependent heat loads on the Turbine Enclosure Cooling Water (TECW) system that are increased by the TPO, are those related to the operation of the condensate pump motor bearing oil coolers and turbine enclosure sample station. The remaining TECW heat loads are not strongly dependent upon reactor power and do not significantly increase. The TECW system has sufficient capacity to assure that adequate heat removal capability is available for TPO operation.

6.4.5 Ultimate Heat Sink The ultimate heat sink (UHS) for Limerick is the spray pond. The ESW and RHRSW systems provide water from the UHS for equipment cooling throughout the plant. As a result of operation at the TPO RTP level, the post-LOCA heat load increases slightly, primarily due to higher reactor decay heat. However, the ability of the UHS to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the UHS are within previous evaluations and the requirements are unchanged for TPO uprate conditions. The current Technical Specifications for UHS limits are adequate due to conservatism in the current design.

6.5 STANDBY LIQUID CONTROL SYSTEM The SLCS is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that all or some of the control rods cannot be inserted. This manually or automatically operated system pumps a highly enriched sodium pentaborate solution into the vessel to achieve a sub critical condition. The generic evaluation presented in TLTR Section 5.6.5 (SLCS) and Appendix L.3 (ATWS Evaluation) was not applicable to the Limerick TPO uprate. The SLCS performance was analyzed to compare the maximum discharge pressure of the SLCS pump to the SLCS relief valve nominal set pressure.

The SLCS analysis concluded that there is insufficient margin between the relief valve nominal setpoint and the maximum expected pump discharge pressure based on a nominal SLCS set pressure of 1400 psig for three SLCS pump operation. However, two SLCS pump operation results maintain adequate margin between the discharge pressure of the SLCS pump and the nominal set pressure of the SLCS relief valve opening setpoint accounting for setpoint tolerance.

The TPO uprate does not affect shutdown or injection capability of the SLCS system in two SLCS pump operation. Because the shutdown margin is reload dependent, the shutdown margin and the required reactor boron concentration are confirmed for each reload core.

The SLCS ATWS performance is evaluated in TSAR Section 9.3.1. The evaluation shows that the TPO has no adverse effect on the ability of the SLCS to mitigate an ATWS in two SLCS pump operation.

6-6

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 6.6 POWER-DEPENDENT HEATING, VENTILATION AND AIR CONDITIONING The Heating, Ventilation and Air Conditioning (HVAC) systems that are potentially affected by the TPO uprate consist mainly of heating, cooling supply, exhaust, and recirculation units in the turbine enclosure, reactor enclosure (including steam tunnel), and primary containment.

TPO results in a minor increase in the heat load caused by the slightly higher FW process temperature (~2°F). The increased heat load is within the margin of the steam tunnel area coolers. In the drywell, the increase in heat load due to the FW process temperature is within the system capacity. In the turbine enclosure, the temperature increases are expected to be very low due to the increase in the FW process temperatures. In the reactor enclosure, the increase in heat load caused by the slightly higher FW process temperature is within the margin of the area coolers. Other areas are unaffected by the TPO because the process temperatures and electrical heat loads remain constant.

Therefore, the power-dependent HVAC systems are adequate to support the TPO uprate.

6.7 FIRE PROTECTION Operation of the plant at the TPO RTP level does not affect the fire suppression or detection systems. There is no change in the physical plant configuration and the potential for minor changes to combustible loading as a result of the TPO uprate are addressed by controlled design change procedures (e.g., the new FW Ultrasonic Flow Measurement (UFM) equipment).

The Limerick fire safe shutdown analysis was performed at 3622 MWt (i.e., 110% of OLTP) and thus bounds operation at the TPO power level of 3515 MWt. The fire safe shutdown analysis includes consideration of equipment needed to achieve and maintain hot shutdown, fire barriers, operator manual actions, personnel resources, and repair activities credited to achieve and maintain cold shutdown. Thus, the fire safe shutdown analysis is acceptable for TPO operation.

6.7.1 10 CFR 50 Appendix R Fire Event The Limerick fire safe shutdown analysis was performed at 3622 MWt (i.e., 110% of OLTP) and thus bounds operation at the TPO power level of 3515 MWt. Thus, the fire safe shutdown analysis is acceptable for TPO operation.

6.8 SYSTEMS NOT AFFECTED BY TPO UPRATE Based on experience and previous NRC reviews, all systems that are significantly affected by TPO are addressed in this report. Other systems not addressed by this report are not significantly affected by TPO. The systems unaffected by TPO at Limerick are confirmed to be consistent with the generic description provided in the TLTR.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-1 TPO Plant Electrical Characteristics Parameter Value Generator Output (MWe) 1138.47 Rated Voltage (kV) 22 Power Factor 0.90 Generator Output (MVA) 1264.97 Current Output (Amps) 33,197 Isolated Phase Bus Duct Rating: (Amps)

Main Section 35,000 Branch Section 2,000 Main Transformers Rating (MVA)

Unit 1 1,266 Unit 2 1,575 Table 6-2 Main Generator Ratings Comparison Power Level Design Max. Nominal MVA @ 75 psig MWe @ 75 psig MVAR @ 75 H2 H2 psig H2 Existing 1264.97 1138.47 551.39 (1)

Uprated 1264.97 1138.47 551.39 (1)

Operation at the uprated condition is not expected to have any effect on the operation of the Main Generator. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-3a Limerick 1 Generator Step-up Transformer Ratings Comparison Power Level Design MVA @ 65ºC MVA Loading Existing 1266 1211 Uprated(1) 1266 1211 (1)

Operation at the uprated condition is not expected to have any effect on the operation of the Generator Step-up Transformer. The generator MWe will increase and MVAR will decrease, thus MVA will remain the same. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.

Table 6-3b Limerick 2 Generator Step-up Transformer Ratings Comparison Power Level Design MVA @ 65ºC MVA Loading Existing 1266 1213 (1)(2)

Uprated 1575 1213 (1)

Operation at the uprated condition is not expected to have any effect on the operation of the Generator Step-up Transformer. The generator MWe will increase and MVAR will decrease, thus MVA will remain the same. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.

(2)

The Unit 2 Generator Step-up Transformer is planned for replacement in 2011. The new transformer will have an increased design MVA rating.

Table 6-4 Unit Auxiliary Transformer Ratings Comparison Power Level Rated MVA @ 65ºC Existing MVA Loading TPO MVA Loading Unit 1(1) 52.5 49.025 49.139 (1)

Unit 2 52.5 47.3 47.414 (1)

Operation at the uprated condition is not expected to have any effect on the operation of the Unit Auxiliary Transformer. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.

6-9

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-5a Station Auxiliary Transformer Ratings Comparison Power Level Rated MVA @ 65ºC Existing MVA Loading TPO MVA Loading (1)

Unit 1 61.6 45.444 45.558 (1)

Operation at the uprated condition is not expected to have any effect on the operation of the Station Auxiliary Transformer. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.

Table 6-5b Regulating Transformer Ratings Comparison Power Level Rated MVA @ 65ºC Existing MVA Loading TPO MVA Loading Unit 2(1) 58 43.109 43.223 (1)

Operation at the uprated condition is not expected to have any effect on the operation of the Regulating Transformer. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.

Table 6-6: Fuel Pool Cooling and Cleanup Parameters CLTP TPO Parameter Normal Normal Full Core Full Core Batch Batch Bundles offloaded 280 764 280 764 Maximum fuel moves per hour 10 10 10 10 Time from shutdown to beginning of fuel transfer (hours) 40 40 40 40 Heat load at maximum temperature (MBTU/hr) 24.40 46.76 24.77 48.01 Maximum bulk temperature (ºF) 117 138 117 139 1 1 Maximum bulk temperature limit (ºF) 140 140 140 140 Time to boil (hrs) 9.25 3.57 9.10 3.46 Note 1: 143°F when utilizing FPCC only 6-10

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-7 Effluent Discharge Comparison Parameter State Current TPO Limit Discharge Temperature (°F) 110 89.11 Insignificant change Residual Oxides mg/L 0.2 0.12 Insignificant change Spectrus CT1300 mg/L 0.4 < 0.052 Insignificant change

1. Maximum discharge temperature for 2009. Taken from NPDES Sampling at Outfall 001 on July 29, 2009.
2. Maximum value from weekly grab samples during April 2009.

6-11

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 7.0 POWER CONVERSION SYSTEMS 7.1 TURBINE-GENERATOR General Electric Energy Services and Siemens performed the evaluation of the steam turbine, valves, turbine auxiliary systems, cross around relief valves and piping for the TPO condition. A summary of the results of the evaluation are presented as follows:

For the turbine high pressure (HP) section, the existing nozzle plates are not able to pass the required additional steam flow at the TPO operation point and still maintain sufficient flow margin of approximately 3% for reactor pressure control. New first stage nozzle plates designed with increased flow area are required; these modified nozzle plates will allow the HP turbine to maintain flow margin of approximately 3% and thus maintain adequate pressure control. All other components in the HP section are within allowable design limits and no other changes are recommended or required.

The turbine low pressure (LP) section rotor and all LP components are within allowable design margins and no changes are recommended or required.

Main stop valves, control valves, and combined intermediate valves are all within allowable design margins to operate at the TPO flow condition.

The turbine auxiliary systems were evaluated; no modifications are needed to support operation at the TPO uprate condition.

The existing missile analysis was evaluated for TPO conditions. The assumptions in the existing missile analysis bound operation at TPO conditions. Thus, the missile generation probability remains unchanged and is therefore acceptable.

The overspeed evaluation was reviewed for TPO conditions. The assumptions in the existing overspeed evaluation bound operation at TPO conditions. Thus, the overspeed evaluation remains acceptable for TPO operation. No change in the overspeed trip settings is required.

7.2 CONDENSER AND STEAM JET AIR EJECTORS The main condenser capability was evaluated for performance at the TPO uprate conditions in Section 6.4.2. Air leakage into the condenser does not increase as a result of the TPO uprate.

The small increase in hydrogen and oxygen flows from the reactor does not affect the Steam Jet Air Ejector (SJAE) capacity because the design was based on operation at greater than required flows at uprate conditions. Therefore, the condenser air removal system is not affected by the TPO uprate and the SJAEs are adequate for operation at the TPO uprate conditions.

7-1

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 7.3 TURBINE STEAM BYPASS The Steam Bypass Pressure Control System (SBPCS) was originally designed for a steam flow capacity of approximately 25% of the 100% rated flow at CLTP. The steam bypass capacity at the TPO RTP is approximately 24.3% of the 100% TPO RTP steam flow rate. The steam bypass system is non-safety-related. While the bypass capacity as a percent of rated steam flow is reduced, the actual steam bypass capacity is unchanged. The transient analyses that credit the turbine bypass system use a bypass capacity that is less than the actual capacity. Therefore, the turbine bypass capacity remains adequate for TPO operation because the actual capacity (unchanged) continues to bound the value used in the analyses.

7.4 FEEDWATER AND CONDENSATE SYSTEMS The FW and condensate systems are designed to provide FW at the temperature, pressure, quality, and flow rate required by the reactor. These systems are not safety-related; however, their performance may have an effect on plant availability and the capability to operate reliably at the TPO uprate condition.

A review of the Limerick FW heaters, heater drain system, condensate demineralizers, and the pumps (FW and condensate) demonstrated that the components are capable of performing in the proper design range to provide the slightly higher TPO uprate FW flow rate at the desired temperature and pressure. A review of the Limerick heater drain system demonstrated that the components are capable of supporting the slightly higher TPO uprate extraction flow rates. The No. 3 heater drain valves are currently undersized, but will be replaced prior to TPO.

Performance evaluations were based on an assessment of the capability of the condensate and FW systems and equipment to remain within the design limitations of the following parameters:

  • Ability to avoid suction pressure trip
  • Flow capacity
  • Bearing cooling capability
  • Rated driver horsepower
  • Vibration The FW system run-out and loss of FW heating events are expected to see very small changes from the TPO uprate as shown by the experience with substantially larger power uprates.

7.4.1 Normal Operation System operating flows for the TPO uprate increase approximately 2.3%. Operation at the TPO RTP level does not significantly affect operating conditions of these systems. Discharge pressure of the condensate pumps decreases due to the pump head characteristics at increased flows. Discharge pressure of the FW pumps will increase to compensate for the increase in FW friction losses due to higher flow. To accomplish this function, opening the flow control valves to the feed pump turbine increases the feed pump speed. During steady-state conditions, the 7-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION condensate and FW systems have available NPSH for all of the pumps to operate without cavitation at the TPO uprate conditions. Adequate margin during steady-state conditions exists between the calculated minimum pump suction pressure and the minimum pump suction pressure trip set points.

The existing FW design pressure and temperature requirements bound operating conditions with adequate margin. The FW heaters are ASME Section VIII pressure vessels. The heaters were analyzed and verified to be acceptable for the slightly higher FW heater temperatures and pressures for the TPO uprate.

7.4.2 Transient Operation To account for FW demand transients, the condensate and FW systems were evaluated to ensure that sufficient margin above the TPO uprated flow is available. For system operation with all system pumps available, the predicted operating parameters were acceptable and within the component capabilities.

Following a single FW pump trip with low reactor water level, the reactor recirculation system would runback recirculation flow, such that the steam production rate is within the flow capacity of the remaining FW pumps. The runback setting prevents a reactor low water level scram, and is sufficient to maintain adequate margin to the potential P/F instability regions. Operation at the TPO condition does not degrade this capability.

7.4.3 Condensate Filters and Condensate Deep Bed Demineralizers The effect of the TPO uprate on the condensate filter demineralizers (CFDs) and the condensate deep bed demineralizers (CDBDs) was reviewed. The CFD and CDBD systems can accommodate (without bypass) TPO uprate operations with one vessel removed from service (when backwash/resin change out is required).

7-3

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 8.0 RADWASTE AND RADIATION SOURCES 8.1 LIQUID AND SOLID WASTE MANAGEMENT The liquid radwaste system collects, monitors, processes, stores, and returns processed radioactive waste to the plant for reuse, discharge, or shipment.

Major sources of liquid and wet solid waste are from the condensate filters and deep bed condensate demineralizers. The TPO uprate results in a ~ 2% increased flow rate through the condensate system, potentially resulting in a reduction in the average time between backwashes of the condensate filters and replacement of the deep bed condensate demineralizer resin. This potential reduction in the service time of the condensate filters and deep bed condensate demineralizers does not affect plant safety.

The liquid collection subsystem and the solid collection subsystem both receive periodic inputs from a variety of sources. Neither subsystem experiences a significant increase in volume due to operation at the TPO uprate condition.

The activated corrosion products in the waste stream are expected to increase proportionally to the TPO uprate. However, the total volume of processed waste is not expected to increase appreciably because the only significant increase in processed waste is due to the more frequent backwashes of the condensate filters and deep bed condensate demineralizers. A review of plant operating effluent reports and the slight increase expected from the TPO uprate, leads to the conclusion that the requirements of 10 CFR 20 and 10 CFR 50, Appendix I will continue to be met. Therefore, the TPO uprate does not adversely affect the processing of liquid radwaste and there are no significant environmental effects.

8.2 GASEOUS WASTE MANAGEMENT The gaseous waste systems collect, control, process, and dispose of gaseous radioactive waste generated during normal operation and abnormal operational occurrences. The gaseous waste management systems include the offgas system and various building ventilation systems. The systems are designed to meet the requirements of 10 CFR 20 and 10 CFR 50, Appendix I.

Non-condensable radioactive gas from the main condenser normally contains activation gases and fission product radioactive noble gas parents. These are the major sources of radioactive gas, and are greater than all other sources combined. These non-condensable gases, along with non-radioactive air in leakage, are continuously removed from the main condensers by the SJAEs that discharge into the offgas system.

Building ventilation systems control airborne radioactive gases by using devices such as High Efficiency Particulate Air (HEPA) and charcoal filters, and radiation monitors that activate isolation dampers or trip supply and exhaust fans, or by maintaining negative or positive air pressure to limit migration of gases. The changes to the gaseous radwaste releases are 8-1

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION proportional to the change in core power and the total releases are a small fraction of the design basis releases.

The release limit is an administratively controlled variable and is not a function of core power.

The gaseous effluents are well within limits at CLTP operation and remain well within limits following implementation of the TPO uprate. There are no significant environmental effects due to the TPO uprate.

The off gas system was evaluated for the TPO uprate. Radiolysis of water in the core region, which forms H2 and O2, increases linearly with core power, thus increasing the heat load on the recombiner and related components. The Offgas system design basis H2 is 138.4 scfm (with a corresponding stoichiometric O2 of 69.7 scfm). The expected H2 flow rate for the TPO uprate is 96.4 scfm (48.2 scfm of O2). The increase in H2 and O2 due to the TPO uprate remains well with the capacity of the system. Therefore, the TPO uprate does not affect the offgas system design or operation.

8.3 RADIATION SOURCES IN THE REACTOR CORE TLTR Appendix H describes the methodology and assumptions for the evaluation of radiological effects for the TPO uprate.

During power operation, the radiation sources in the core are directly related to the fission rate.

These sources include radiation from the fission process, accumulated fission products and neutron reactions as a secondary result of fission. Historically, these sources have been defined in terms of energy released per unit of reactor power. Therefore, for TPO, the percent increase in the operating source terms is no greater than the percent increase in power. The source term increases due to the TPO uprate are bounded by the safety margins of the design basis sources.

The post-operation radiation sources in the core are primarily the result of accumulated fission products. Two separate forms of post-operation source data are normally applied. The first is the core gamma-ray source, which is used in shielding calculations for the core and for individual fuel bundles. This source term is defined in terms of MeV/sec per watt of reactor thermal power (or equivalent) at various times after shutdown. Therefore, the total gamma energy source increases in proportion to reactor power.

The second set of post-operation source data consists primarily of nuclide activity inventories for fission products in the fuel. These are needed for post-accident and spent fuel pool evaluations, which are performed in compliance with regulatory guidance that applies different release and transport assumptions to different fission products. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops equilibrium activities in the fuel (typically three years). Most radiologically significant fission products reach equilibrium within a 60-day period. The calculated inventories are approximately proportional to core thermal power. Consequently, for TPO, the inventories of those radionuclides, which reached or approached equilibrium, are expected to increase in proportion 8-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION to the thermal power increase. The inventories of the very long-lived radionuclides, which did not approach equilibrium, are both power and exposure dependent. They are expected to increase proportionally with power if the fuel irradiation time remains within the current basis.

Thus, the long-lived radionuclides are expected to increase proportionally to power. The radionuclide inventories are provided in terms of curies per megawatt of reactor thermal power at various times after shutdown.

((

))

8.4 RADIATION SOURCES IN REACTOR COOLANT 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 is the dominant source in the turbine building and in the lower regions of the drywell. Because these sources are produced by interactions in the core region, their rates of production are proportional to power. However, the concentration in the steam remains nearly constant, because the increase in activation production is balanced by the increase in steam flow. As a result, the activation products, observed in the reactor water and steam, increase in approximate proportion to the increase in thermal power.

8.4.2 Activated Corrosion Products The reactor coolant contains activated corrosion products from metallic materials entering the water and being activated in the reactor region. Under the TPO uprate conditions, the FW flow increases with power, the activation rate in the reactor region increases with power, and the filter efficiency of the condensate demineralizers may decrease as a result of the FW flow increase.

The net result may be an increase in the activated corrosion product production. However, the 8-3

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION TPO uprate corrosion product concentrations are not expected to exceed the design basis concentrations. TPO activated corrosion product activity levels in the reactor water remain less than one-half of the design basis activated corrosion product activity. Therefore, no change is required in the design basis activated corrosion product concentrations for the TPO uprate.

8.4.3 Fission Products 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. The noble gases released during plant operation result from the escape of minute fractions of the fission products from the fuel rods.

Noble gas release rates increase approximately with power level. This activity is the noble gas offgas that is included in the Limerick design. The offgas rates for TPO uprate operations are well below the original design basis. Therefore, the design basis release rates are bounding for the TPO uprate.

The fission product activity in the reactor water, like the activity in the steam, is the result of minute releases from the fuel rods. As is the case for the noble gases, there is no expectation that releases from the fuel increase due to the TPO uprate. Activity levels in the reactor water are expected to be approximately equal to current measured data, which are fractions of the design basis values. Therefore, the design basis values are unchanged.

8.5 RADIATION LEVELS Normal operation radiation levels increase slightly for the TPO uprate. Limerick was designed with substantial conservatism 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 design, source terms, and analytical techniques.

Post-operation radiation levels in most areas of the plant increase by no more than the percentage increase in power level. In a few areas near the SFP cooling system piping and the reactor water piping, where accumulation of corrosion product crud is expected, as well as near some liquid radwaste equipment, the increase could be slightly higher. The radiation levels in areas with significant N-16 radiation are expected to increase by slightly more than the percentage increase in power level.

Regardless, individual worker exposures will be maintained within acceptable limits by the site As Low As Reasonably Achievable (ALARA) program, which controls access to radiation areas.

Procedural controls compensate for increased radiation levels.

The change in core activity inventory resulting from the TPO uprate (Section 8.3) increases post-accident radiation levels by no more than approximately the percentage increase in power level.

The slight increase in the post-accident radiation levels has no significant effect on the plant or the habitability of the on-site Emergency Response facilities. A review of areas requiring post-8-4

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION accident occupancy concluded that access needed for accident mitigation is not significantly affected by the TPO uprate.

Section 9.2 addresses the Main Control Room doses for the worst-case accident.

8.6 NORMAL OPERATION OFF-SITE DOSES The Technical Specification limits implement the guidelines of 10 CFR 50, Appendix I. A review of the normal radiological effluent doses shows that at CLTP, the annual doses are a small fraction of the doses allowed by Technical Specification limits. The TPO uprate does not involve significant increases in the offsite dose from noble gases, airborne particulates, iodine, tritium or liquid effluents. In addition, radiation from shine is not a significant exposure pathway. Present offsite radiation levels are a negligible portion of background radiation.

Therefore, the normal offsite doses are not significantly affected by operation at the TPO RTP level and remain below the limits of 10 CFR 20 and 10 CFR 50, Appendix I.

8-5

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 9.0 REACTOR SAFETY PERFORMANCE EVALUATIONS 9.1 ANTICIPATED OPERATIONAL OCCURRENCES TLTR Appendix E provides a generic evaluation of the AOOs for TPO uprate plants. ((

)) Also included are the analytical methods to be used and operating conditions to be assumed. The AOO events are organized into two major groups: Fuel Thermal Margin Events, and Transient Overpressure Events.

TLTR Table E-2 illustrates the effect of a 1.5% power uprate on the OLMCPR. ((

)) The OLMCPR changes for the 1.7% uprate may be slightly larger than shown in Table E- 2, but the changes are expected to be within the normal cycle-to-cycle variation. The overpressure events and loss of FW transient are currently performed with the assumption of 2%

overpower. Therefore, they are applicable and bounding for the TPO uprate.

The reload transient analysis includes the worst overpressure event, which is usually the closure of all MSIVs with high neutron flux scram.

The evaluations and conclusions of TLTR Appendix E are applicable to the Limerick TPO uprate. Therefore, it is sufficient for the plant to perform the standard reload analyses at the first fuel cycle that implement the TPO uprate.

9.2 DESIGN BASIS ACCIDENTS The radiological consequences of a DBA are basically proportional to the quantity of radioactivity released to the environment. This quantity is a function of the fission products released from the core as well as the transport mechanisms from the core to the release point.

The radiological releases at the TPO uprate power are generally expected to increase in proportion to the core inventory increase, which is in proportion to the power increase.

Postulated DBA events have been evaluated and analyzed to show that NRC regulations are met for 2% above the CLTP. DBA events have either been previously analyzed at 102% of CLTP or are not dependent on core thermal power. The Main Steam Line Break Accident outside containment was evaluated using a 4 µCi/g dose equivalent I-131 limit on reactor coolant activity. The limit on reactor coolant activity is unchanged for the TPO uprate condition. The evaluation/analysis was based on the methodology, assumptions, and analytical techniques 9-1

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION described in the Regulatory Guides, the Standard Review Plan (SRP) (where applicable), and in previous SEs.

9.3 SPECIAL EVENTS 9.3.1 Anticipated Transient Without Scram TLTR Section 5.3.5 and TLTR Appendix L, present a generic evaluation of the sensitivity of an ATWS to a change in power typical of the TPO uprate. The evaluation is based on previous analyses for power uprate projects. For a TPO uprate, if a plant has sufficient margin for the projected changes in peak parameters given in TLTR Section L.3.5, ((

))

The previous ATWS analysis, performed at 100% of CLTP, did not demonstrate the required margins for generic evaluation to the peak vessel bottom head pressure limit and to the pool temperature limit. ((

))

NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate, Class III, July 2003 (also referred to as the CLTR) was approved by the NRC as an acceptable method for evaluating the effects of Constant Pressure Power Uprates (CPPUs). Section 9.3.1 of the CLTR addresses the effect of CPPU on ATWS. The CLTR methodology was used to analyze and evaluate the Limerick ATWS event.

(( )) ATWS analysis is required for TPO RTP to ensure that the following ATWS acceptance criteria are met:

  • Maintain reactor vessel integrity (i.e., peak vessel bottom pressure less than the ASME Service Level C limit of 1500 psig).
  • Maintain containment integrity (i.e., maximum containment pressure and temperature less than the design pressure (55 psig) and temperature (190°F) of the containment structure).
  • Maintain coolable core geometry.

9-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION The TPO RTP ATWS analysis is performed using the NRC approved code ODYN (Table 1-1).

The key inputs to the ATWS analysis are provided in Table 9-1. The results of the analysis are provided in Table 9-2.

The results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the Limerick response to an ATWS event at TPO is acceptable. The potential for thermal-hydraulic instability in conjunction with ATWS events is evaluated in Section 9.3.1.4.

Limerick also meets the ATWS mitigation requirements defined in 10 CFR 50.62:

  • Installation of an ARI system;
  • Boron injection equivalent to 86 gpm; and
  • Installation of automatic Recirculation Pump Trip (RPT) logic (i.e., ATWS-RPT).

There are no changes to the assumed operator actions for the TPO RTP ATWS analysis.

When required by changes in plant configuration (as identified by the design change process),

changes to Emergency Operating Procedures (EOPs), including changes to EOP calculations and plant data, are developed and implemented in accordance with plant administrative procedure for EOP program maintenance.

Limerick performs EOP calculations consistent with the BWR Owners Group Emergency Procedure Guidelines (EPGs) / Severe Accident Guidelines (SAGs) Appendix C. Critical software is verified and validated by Design Engineering to generate EOP results. The EOP calculation input and output data is reviewed and verified by Design Engineering. Changes to the EOP calculation outputs are forwarded to Operations for use in revising the EOP Procedures/Flow Charts and the SAGs and supporting documents. Finally, the EOP flow charts are verified and validated by Operations, including trial use in the simulator.

The ATWS mitigation strategy is based on the BWROG EPGs, which are incorporated in the existing Limerick EOPs. TPO implementation does not significantly change the transient sequence of events. Therefore, there is no change in operator strategy on ATWS level reduction or early boron injection. TPO may affect some of the calculated curves, but does not affect stability mitigation actions. The changes due to TPO do not require modification of operator instructions.

Limerick meets all CLTR dispositions and the results in this evaluation are described below.

The topics addressed in this evaluation are:

9-3

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Topic CLTR Disposition Limerick Result Meets CLTR ATWS (Overpressure) - Event Selection ((

Disposition Meets CLTR ATWS (Overpressure) - Limiting Events Disposition ATWS (Suppression Pool Temperature) - Meets CLTR Event Selection Disposition ATWS (Suppression Pool Temperature) - Meets CLTR Limiting Events Disposition Meets CLTR ATWS (Peak Cladding Temperature) ))

Disposition 9.3.1.1 ATWS (Overpressure)

As stated in Section 9.3.1 of the CLTR, the higher operating steam flow may result in higher peak vessel pressures. The higher power and decay heat will result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could affect the capability of the SLCS to mitigate the consequences of an ATWS event.

The overpressure evaluation includes consideration of the most limiting RPV overpressure case.

TLTR Appendix L considers four ATWS events: ((

)) The ATWS (Overpressure) - Event Selection meets all CLTR dispositions.

As shown in Section 3.7 of ELTR2, ((

)) The MSIVC, PRFO and LOOP cases were performed for Limerick. The analysis results are given in Table 9-2. The MSIVC, PRFO and LOOP sequence of events are given in Tables 9-6 through 9-8. The short-term and long-term transient response to the MSIVC, PRFO and LOOP ATWS events are presented in Figures 9-1 through 9-24. Therefore, ATWS (Overpressure) - Limiting Events meet all CLTR dispositions.

9.3.1.2 ATWS (Suppression Pool Temperature)

As stated in Section 9.3.1 of the CLTR, the higher operating steam flow will result in higher peak vessel pressures. The higher power and decay heat may result in higher suppression pool temperatures. The increased core power and reactor steam flow rates, in conjunction with the SRV capacity and response times, could impact the capability of the SLCS to mitigate the consequences of an ATWS event.

The suppression pool temperature evaluation includes consideration of the most limiting RHR pool cooling capability case. TLTR Appendix L considered four ATWS events: ((

9-4

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION

)) The ATWS (Suppression Pool Temperature) - Event Selection meets all CLTR dispositions.

The MSIVC, PRFO and LOOP cases were performed for Limerick. The key inputs to the ATWS analysis are provided in Table 9-1. The ATWS analysis results are given in Table 9-2.

The MSIVC, PRFO and LOOP sequence of events are given in Tables 9-6 through 9-8. The ATWS (Suppression Pool Temperature) - Limiting Events meet all CLTR dispositions.

9.3.1.3 ATWS (Peak Cladding Temperature)

The TLTR in Appendix L.3 states that power uprate has a negligible effect on the PCT or local cladding oxidation. ((

))

For ATWS events, the acceptance criteria for PCT and local cladding oxidation for emergency core cooling systems, defined in 10 CFR 50.46, are adopted to ensure an ATWS event does not impede core cooling.

For TPO, PCT and local cladding oxidation are not required to be explicitly analyzed per Appendix L.3 of TLTR. Therefore, ATWS (PCT) is in compliance with the acceptance criteria of 10 CFR 50.46; subsequently, coolable core geometry is assured by meeting the 2200ºF PCT and the 17% local cladding oxidation acceptance criteria stated in 10 CFR 50.46.

9.3.1.4 ATWS with Core Instability The CLTR in Section 9.3.3 states that 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 TPO operation because the MELLLA upper boundary is not increased. The core design necessary to achieve TPO 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 References 20 and 21 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum flow point. ((

9-5

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION

))

TPO allows plants to increase their operating thermal power but does not allow an increase in control rod line. ((

))

((

))

Initial operating conditions of FWHOOS and FFWTR do not significantly affect the ATWS instability response reported in References 20 and 21. The limiting ATWS evaluation assumes that all FW heating is lost during the event and the injected FW temperature approaches the lowest achievable main condenser hot well temperature. ((

))

((

)) Therefore, the TPO effect on ATWS with core instability at Limerick meets all CLTR dispositions.

9.3.1.5 SLCS System Performance and Hardware Based on the results of the (( )) ATWS analysis, the maximum reactor upper plenum pressure following the limiting ATWS event reaches 1224 psig (1239 psia) during the time the SLCS is analyzed to be in operation. Consequently, there is a corresponding increase in the maximum two pump discharge pressure to 1336 psig and 1330 psig for Limerick Units 1 and 2 respectively and a decrease in the operating pressure margin for the pump discharge relief valves. The relief valve margin is not satisfied with three SLCS pump operation; however, the relief valve margin is satisfied with two SLCS pump operation. This is acceptable because they meet the injection requirements of 10 CFR 50.62 and the injection requirements of Information Notice 2001-13 with two pumps. The key SLCS input parameters are summarized in Table 9-3 and Table 9-4 for three and two SLCS pump operation, respectively. The key SLCS 9-6

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION performance results are presented in Table 9-5. Consideration was also given to system flow, head losses for full injection, and cyclic pressure pulsations due to the positive displacement pump operation in determining the pressure margin to the opening set point for the pump discharge relief valves. The relief valve setpoint margin for two SLCS pump operation is 19.7 psi and 25.8 psi for Limerick Units 1 and 2, respectively. This margin is based on a SLCS pump relief valve setpoint of 1400 psig. The pump discharge relief valves are periodically tested to maintain this tolerance. Therefore, the current SLCS process parameters associated with the minimum boron injection rate are not changed.

The SLCS ATWS performance is evaluated for a representative core design for TPO. The evaluation shows that TPO has no adverse effect on the ability of the SLCS to mitigate an ATWS in two SLCS pump operation. There are no timer setting changes for TPO for Limerick, and the ATWS analysis confirms acceptable results. Therefore, the system performance and hardware meets all CLTR dispositions.

9.3.1.6 Suppression Pool Temperature following ATWS Event As stated in Section 6.5 of the CLTR, changes in the fuel design for TPO may require modifications to the SLCS as a result of the increase in the suppression pool temperature for the limiting ATWS event.

The boron injection rate requirement for maintaining the peak suppression pool water temperature limits, following the limiting ATWS event with SLCS injection, is not increased for TPO. Therefore, the Suppression Pool temperature following an ATWS event meets all CLTR dispositions.

9.3.1.7 Equipment Out-of-Service and Flexibility Options MELLLA, ICF and SLO: The TPO ATWS analyses were performed along the MELLLA boundary. The TPO ATWS analysis at MELLLA conditions bounds operation at ICF and in SLO. Therefore, TPO continues to support these performance improvement features.

SRV OOS: The TPO ATWS analysis was performed with two SRVs OOS. Therefore, TPO continues to support this EOOS option.

FWHOOS and FFWTR: FWHOOS and FFWTR are operational flexibility options that allow continued operation with reduced FW temperature. Initial power is unchanged for both the FWHOOS and FFWTR conditions. The additional reactivity associated with the reduced FW temperature is typically offset with control rods, as needed. This makes the core less reactive due to the lower void fraction. Thus, use of normal FW temperature is conservative for ATWS analyses.

MSIV OOS: The TPO ATWS analysis bounds the MSIV OOS condition for TPO. Limerick operation with MSIV OOS is limited to 75% rated power. With this restriction, the severity of the limiting ATWS events is reduced. The lower initial steaming rate reduces the peak vessel 9-7

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION pressure, peak power, PCT, and integrated SRV flow. The reduction in integrated SRV flow thereby reduces the peak suppression pool temperature and containment pressure.

ARTS, TBV OOS, RPT OOS, 24 Month Cycle, TCV Stuck Closed and TSV Stuck Closed: The TPO ATWS analysis is not impacted by these performance improvement features.

9.3.2 Station Blackout The Limerick Station Blackout (SBO) evaluation has previously been performed assuming 102% of CLTP. Therefore, the postulated SBO scenarios for TPO operation are bounded by the current evaluations.

9-8

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-1 Key Inputs for ATWS Analysis Input Variable CLTP TPO RTP Reactor power (MWt) 3458 3517*

Reactor dome pressure (psia) 1060 1060 Each SRV capacity at 1090 psig (Mlbm/hr) 0.870 0.870 High pressure ATWS-RPT (psig) 1156 1156 Number of SRVs OOS 2 2 Number of Auto Start SLCS Pumps 2 2

  • Performed at 101.7% of CLTP Table 9-2 Results for ATWS Analysis Acceptance Criteria CLTP 1, 2 TPO RTP1 Peak vessel bottom pressure (psig) 1458 1473 Peak suppression pool temperature (°F) 181 182 Peak containment pressure (psig) 10.3 10.6 Peak cladding temperature (°F) Generic Assessment Generic Assessment Local cladding oxidation (%) Generic Assessment Generic Assessment Notes:
1. Cladding temperature and oxidation calculations are not required per Appendix L.3 of TLTR.
2. To maximize the effect of TPO, a baseline is established at the CLTP level, assuming the current licensed equipment performance assumptions and plant parameters.

Table 9-3 Inputs for Limerick Three SLCS Pump Operation ATWS Analysis Parameter Limerick Unit 1 Limerick Unit 2 Reactor Water Elevation Head (psi) 2.5 2.5 CS Nozzle Differential Pressure (psi) 20.8 20.8 SLCS Pipe Elevation (psi) 10.0 10.0 Flow Friction Loss (psi) 126.0 112.1 Total Three Pump Operation System Losses (psi) 159.3 145.4 9-9

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-4 Inputs for Limerick Two SLCS Pump Operation ATWS Analysis Parameter Limerick Unit 1 Limerick Unit 2 Reactor Water Elevation Head (psi) 2.5 2.5 CS Nozzle Differential Pressure (psi) 20.3 20.3 SLCS Pipe Elevation (psi) 10.0 10.0 Flow Friction Loss (psi) 79.5 73.4 Total Two Pump Operation System Losses (psi) 112.3 106.2 Table 9-5 Limerick SLCS Pressure Results for ATWS Analysis Parameter Limerick Unit 1 Limerick Unit 2 SLCS RV Setpoint (psig) 1400 1400 RV Tolerance (%) 1.0 1.0 Pulsating Margin (psi) 30.0 30.0 SLCS Piping Flow Loss (psi) - 3 Pump Operation 159.3 145.4 SLCS Piping Flow Loss (psi) - 2 Pump Operation 112.3 106.2 Pressure at Point of SLCS Injection (psig) 1224 1224 Pump Discharge Pressure (psig) - 3 Pump Operation 1383.3 1369.4 Pump Discharge Pressure (psig) - 2 Pump Operation 1336.3 1330.2 9-10

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-6 MSIVC Sequence of Events Item Event TPO RTP BOC TPO RTP EOC Event Time Event Time (sec) (sec) 1 MSIV Isolation Initiated 0.0 0.0 2 MSIVs Fully Closed 4.0 4.0 3 Peak Neutron Flux 4.1 4.0 4 High Pressure ATWS Setpoint 4.1 4.1 5 Recirculation Pumps Trip 4.7 4.6 6 Start Opening of the First Relief Valve 4.9 4.8 7 Peak Heat Flux 5.2 5.1 8 Peak Vessel Pressure 9.5 9.1 9 Feedwater Reduction Initiated 30.0 30.0 10 BIIT Reached 53.0 54.0 11 SLCS Pumps Start 124.1 124.1 12 Hot Shutdown Achieved 422 444 (Neutron Flux Remains < 0.1%)

13 RHR Cooling Initiated 660 660 14 Peak Suppression Pool Temperature 8527 8322 9-11

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-7 PRFO Sequence of Events Item Event TPO RTP BOC TPO RTP EOC Event Time Event Time (sec) (sec) 1 TCV and Bypass Valves Start Open 0.1 0.1 MSIV Closure Initiated by Low Steamline 2 20.2 19.3 Pressure 3 MSIVs Fully Closed 24.2 23.3 4 Peak Neutron Flux 25.2 24.8 5 High Pressure ATWS Setpoint 27.8 27.0 6 Recirculation Pumps Trip 28.4 27.5 7 Start Opening of the First Relief Valve 28.6 27.7 8 Peak Heat Flux 28.8 27.5 9 Peak Vessel Pressure 35.2 33.9 10 Feedwater Reduction Initiated 53.4 53.4 11 BIIT Reached 69.0 70.0 12 SLCS Pumps Start 147.8 147.0 Hot Shutdown Achieved 13 451 477 (Neutron Flux Remains < 0.1%)

14 RHR Cooling Initiated 660 660 15 Peak Suppression Pool Temperature 7871 7131 9-12

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Table 9-8 LOOP Sequence of Events Item Event TPO RTP BOC TPO RTP EOC Event Time Event Time (sec) (sec) 1 Loss of Auxiliary Power 0.0 0.0 2 Recirculation Pumps Trip 0.0 0.0 3 Feedwater Reduction Initiated 0.0 0.0 4 TCV Closure 0.0 0.0 5 Peak Neutron Flux 0.7 0.7 6 High Pressure ATWS Setpoint 1.0 0.9 7 Start Opening of the First Relief Valve 2.0 1.9 8 MSIV Closure Initiated 2.0 2.0 9 Peak Heat Flux 2.3 2.2 10 MSIVs Fully Closed 6.0 6.0 11 Peak Vessel Pressure 8.3 7.8 12 BIIT Reached 54.0 55.0 13 SLCS Pumps Start 121.0 120.9 Hot Shutdown Achieved 14 419 454 (Neutron Flux Remains < 0.1%)

15 RHR Cooling Initiated 660 660 16 Peak Suppression Pool Temperature ~29000 ~29000 9-13

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-1:

Figure 9-1 : TPO TPO RTP RTP MELLLA MELLLA BOC MSIVC (Short DOC MSIVC (Short Term)

Term)

((

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9-14

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-2:

Figure 9-2: TPO TPO RTP RTP MELLLA MELLLA HOC BOC MSIVC MSIVC (Long (Long Term Term -- A)

A)

((

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9-15

NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-3:

Figure TPO RTP 9-3: TPO RTP MELLLA MELLLA BOC BOC MSIVC MSIVC (Long (Long Term Term -- B)

B)

((

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9-16

NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure Figure 9-4 9-4:: TPO TPO RTP RTP MELLLA MELLLA BOC BOC MSIVC MSIVC (Long (Long Term Term -- C)

C)

((

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9-17

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure Figure 9-5 9-5:: TPO TPO RTP RTP MELLLA MELLLA BOCDOC PRFO PRFO (Short (Short Term)

Term)

((

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9-18

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-6:

Figure 9-6: TPO TPO RTP RTP MELLLA MELLLA BOC PRFO (Long DOC PRFO (Long Term Term -- A)

A)

((

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9-19

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-7:

Figure 9-7 : TPO TPO RTP RTP MELLLA MELLLA BOC BOC PRFO (Long Term PRFO (Long Term -- B)

B)

((

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9-20

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-8:

Figure 9-S : TPO TPO RTP RTP MELLLA MELLLA BOC BOC PRFO PRFO (Long (Long Term Term -- C)

C)

((

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9-21

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-9: TPO Figure 9-9: TPO RTP RTP MELLLA MELLLA BOCBOC LOOP (Short Term)

LOOP (Short Term)

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9-22

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-10:

Figure 9-10: TPO TPO RTP RTP MELLLA MELLLA DOC BOC LOOP LOOP (Long (Long Term Term -- A)

A)

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9-23

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure Figure 9-11 9-11:: TPO TPO RTP RTP MELLLA MELLLA BOCBOC LOOP LOOP (Long (Long Term Term -- B)

B)

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9-24

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure Figure 9-12: TPO RTP 9-12: TPO RTP MELLLA MELLLA BOCBOC LOOP (Long Term LOOP (Long Term -- C)

C)

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9-25

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure Figure 9-13 9-13:: TPO TPO RTP RTP MELLLA MELLLA EOC EOC MSIVC (Short Term)

MSIVC (Short Term)

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9-26

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-14:

Figure TPO RTP 9-14: TPO RTP MELLLA MELLLA EOC EOC MSIVC (Long Term MSIVC (Long Term -- A)

A)

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9-27

NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-15: TPO Figure 9-15: TPO RTP RTP MELLLA MELLLA EOC EOC MSIVC (Long Term MSIVC (Long Term -- B)

B)

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9-28

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure Figure 9-16 9-16:: TPO TPO RTP RTP MELLLA MELLLA EOC Eoe MSIVC (Long Term MSIVe (Long Term -- C)

C)

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9-29

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-17:

Figure 9-1 7: TPO TPO RTP RTP MELLLA MELLLA EOCEOC PRFO (Short Term)

PRFO (Short Term)

((

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9-30

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-Figure 1 8: TPO 9-18: TPO RTP RTP MELLLA MELLLA EOC EOC PRFO PRFO (Long (Long Term Term -- A)

A)

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9-31

NEDO-33484 REVISION 00 NEDO-33484 REVISION NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure Figure 9-19:

9-19: TPO TPO RTP RTP MELLLA MELLLA EOC EOC PRFO (Long Term PRFO (Long Term -- B)

B)

((

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9-32

NEDO-33484 NEDO-33484 REVISION REVISION 00 NON-PROPRIETARY NON-PROPRIETARY INFORMATION INFORMATION Figure 9-20:

Figure 9-20: TPO TPO RTP RTP MELLLA MELLLA EOC EOC PRFO (Long Term PRFO (Long Term -- C)

C)

((

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9-33

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-21:

Figure 9-21 : TPO TPO RTP RTP MELLLA MELLLA EOCEOC LOOP LOOP (Short (Short Term)

Term)

((

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9-34

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-22:

Figure 9-22 : TPO TPO RTP RTP MELLLA MELLLA Eoe EOC LOOP LOOP (Long (Long Term Term -- A)

A)

((

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9-35

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-23:

Figure 9-23: TPO TPO RTP RTP MELLLA MELLLA EOC EOC LOOP LOOP (Long (Long Term Term -- B)

B)

((

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9-36

NEDO-33484 REVISION NEDO-33484 REVISION 00 NON-PROPRIETARY INFORMATION NON-PROPRIETARY INFORMATION Figure 9-24:

Figure 9-24: TPO TPO RTP RTP MELLLA MELLLA EOC EOC LOOP LOOP (Long (Long Term Term -- C)

C)

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9-37

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10.0 OTHER EVALUATIONS 10.1 HIGH ENERGY LINE BREAK Because the TPO uprate system operating temperatures and pressures change only slightly, there is no significant change in HELB mass and energy releases. The FW lines, near the pump discharge, increase < 2°F and ~ 4 psi. The recirculation line temperature decreases approximately 0.1°F with a 0.1 BTU/lbm enthalpy decrease. These changes are insignificant in relation to the effect on line break calculations. Vessel dome pressure and other portions of the RCPB remain at current operating pressure or lower. Therefore, the consequences of any postulated HELB would not significantly change. The postulated break locations remain the same because the piping configuration does not change due to the TPO uprate.

The HELB evaluation was performed for all systems evaluated in the UFSAR. At the TPO RTP level, HELBs outside the drywell would result in an insignificant change in the sub-compartment pressure and temperature profiles. The affected building and cubicles that support safety-related functions are designed to withstand the resulting pressure and thermal loading following an HELB at the TPO RTP. A brief discussion of each break follows.

10.1.1 Steam Line Breaks The critical parameter affecting the high-energy steam line break analysis is the reactor vessel dome pressure. Because there is no pressure increase for the TPO, the main steam line (MSL) pressure decreases and there is a slight decrease in the main steam line break (MSLB) blowdown rate. The MSLB is used to establish the peak pressure and the temperature environment in the MS tunnel. Design margins within the HELB analysis for a MSLB with a concurrent FW line break provide adequate margin to the limits in the steam tunnel.

10.1.2 Liquid Line Breaks 10.1.2.1 Feedwater Line Breaks The TPO uprate increases the FW temperature < 2°F and pressure ~ 4 psi, which results in an insignificant increase in the FW mass and energy release. As a result of the small increase in FW temperature and pressure, the blowdown rate changes marginally and the energy increases slightly. The original analysis was generally performed with conservative modeling assumptions. These conservatisms more than offset the effects of the temperature change.

Therefore, the original HELB analysis is bounding.

10.1.2.2 ECCS Line Breaks Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase. Therefore, the previous HELB analysis is bounding for the TPO uprate condition.

10-1

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION Because these lines are normally isolated, the TPO uprate does not affect their line break analyses, for breaks outside drywell.

10.1.2.3 RCIC System Line Breaks Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase. Therefore, the previous HELB analysis is bounding for the TPO uprate conditions.

10.1.2.4 RWCU System Line Breaks As a result of the small decrease in recirculation temperature with a negligible increase in pressure, the blowdown rate increases slightly and the energy decreases slightly. The original analysis was generally performed with conservative modeling assumptions. These conservatisms more than offset the effects of the temperature change, so the original HELB analysis is bounding.

10.1.2.5 CRD System Line Breaks The CRD pipe rupture analysis is not affected by the TPO uprate.

10.1.2.6 Building Heating Line Breaks Building heating lines are not connected to the reactor-turbine primary loop. Therefore, building heating lines are not affected.

10.1.2.7 Pipe Whip and Jet Impingement Because there is no change in the nominal vessel dome pressure, pipe whip and jet impingement loads do not significantly change. Existing calculations supporting the dispositions of potential targets of pipe whip and jet impingement from postulated HELBs have been reviewed and determined to be adequate for the safe shutdown effects in the TPO RTP conditions. Existing pipe whip restraints, jet impingement shields, and their supporting structures are also adequate for the TPO uprate conditions.

10.1.2.8 Internal Flooding from HELB None of the plant flooding zones contains a potential HELB location affected by the reactor operating conditions changed for the TPO uprate. The high-energy line systems operational modes evaluated for HELB are not affected by the TPO uprate, nor are the plant internal flooding analysis or safe shutdown analysis.

10.2 MODERATE ENERGY LINE BREAK None of the plant flooding zones contains a potential Moderate Energy Line Break (MELB) location affected by the reactor operating conditions changed for the TPO uprate. The following 10-2

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION systems contain potential MELB locations in plant flooding zones: Condensate, Condensate and Refueling Water Storage, SW, Emergency Service Water, RHR, RHR Service Water, Reactor Enclosure Cooling Water, Condensate Filter Demineralizers, Make-up Demineralizer, Fuel and Diesel Oil Storage and Transfer, Auxiliary Steam, Fire Protection, RWCU System, Clean-up Filter Demineralizer, Control Rod Drive Hydraulic, SLCS, RCIC, RCIC Pump Turbine, Core Spray and Safeguard Piping, Fuel Pool Cooling and Cleanup, Fuel Pool Filter/Demineralizer, HPCI, HPCI Pump Turbine, Liquid Radwaste Collection, Liquid Radwaste-Chemical and Laundry Processing, Drywell Chilled Water, and Control Structure Chilled Water.

No new moderate energy lines are identified. Protection requirements for safe-shutdown equipment for a postulated MELB are not dependent on power level. All sources of and protection measures against flooding are independent of power level. Internal flooding will not alter the ability of the plant to reach safe shutdown under TPO. Therefore, the plant internal flooding analysis is not affected.

10.3 ENVIRONMENTAL QUALIFICATION Safety-related components must be qualified for the environment in which they operate. The TPO increase in power level increases the radiation levels experienced by equipment during normal operation and accident conditions. Because the TPO uprate does not increase the nominal vessel dome pressure, there is a very small effect on pressure and temperature conditions experienced by equipment during normal operation and accident conditions. The resulting environmental conditions are bounded by the existing environmental parameters specified for use in the environmental qualification program.

10.3.1 Electrical Equipment The environmental conditions for safety-related electrical equipment were reviewed to ensure that the existing qualification for the normal and accident conditions expected in the area where the devices are located remain adequate. Conservatisms in the equipment qualifications were originally applied to the environmental parameters, and no change is needed for the TPO uprate.

10.3.1.1 Inside Containment Environmental qualification (EQ) for safety-related electrical equipment located inside the containment is based on DBA-LOCA conditions and their resultant temperature, pressure, humidity and radiation consequences, and includes the environments expected to exist during normal plant operation. The current accident conditions for temperature and pressure are based on analyses initiated from 102% of CLTP. Normal temperatures may increase slightly near the FW and reactor recirculation lines and will be evaluated through the EQ temperature monitoring program, which tracks such information for equipment aging considerations. The current radiation levels under normal plant conditions also increase slightly. The current plant environmental envelope for radiation is not exceeded by the changes resulting from the TPO uprate.

10-3

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION 10.3.1.2 Outside Containment Accident temperature, pressure, and humidity environments used for qualification of equipment outside containment result from an MSLB in the pipe tunnel, or other HELBs, whichever is limiting for each area. The HELB pressure and temperature profiles bound the TPO uprate conditions. There is adequate margin in the qualification envelopes to accommodate the small changes due to TPO conditions. Maximum accident radiation levels used for qualification of equipment outside containment are from a DBA-LOCA.

10.3.2 Mechanical Equipment With Non-Metallic Components Operation at the TPO RPT level increases the normal process temperature very slightly in the FW piping. The slight increase in normal and accident radiation was evaluated in Section 10.3.

10.3.3 Mechanical Component Design Qualification The increase in power level increases the radiation levels experienced by equipment during normal operation. However, where the previous accident analyses have been based on 102% of CLTP, the accident pressures, temperatures and radiation levels do not change. The mechanical design of equipment/components (valves, heat exchangers, pumps, snubbers, etc.) in certain systems is affected by operation at the TPO RTP level because of the slightly increased temperature and sometimes flow rate. The revised operating conditions do not significantly affect the cumulative usage fatigue factors of mechanical components.

The effects of increased fluid induced loads on safety-related components are described in Section 3.4. As stated in Section 4.1, the containment loads for the TPO uprate are bounded by previous analyses at 102% of CLTP. Increased nozzle loads and component support loads due to the revised operating conditions were evaluated in the piping assessments in Section 3.4. These increased loads are insignificant, and become negligible when combined with the dynamic loads.

Therefore, the mechanical components and component supports are adequately designed for the TPO uprate conditions.

10.4 TESTING The TPO uprate power ascension is based on the guidelines in TLTR Section L.2. Pre-operational tests are not needed because there are no significant changes to any plant systems or components that require such testing.

In preparation for operation at TPO uprate conditions, routine measurements of reactor and system pressures, flows, and selected major rotating equipment vibration are taken near 95% and 100% of CLTP, and at 100% of TPO RTP. The measurements will be taken along the same rod pattern line used for the increase to TPO RTP. Core power from the APRMs is re-scaled to the TPO RTP before exceeding the CLTP and any necessary adjustments will be made to the APRM alarm and trip settings.

10-4

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION The turbine pressure controller setpoint will be readjusted at 95% of CLTP and held constant.

The setpoint is reduced so the reactor dome pressure is the same at TPO RTP as for the CLTP.

Adjustment of the pressure setpoint before taking the baseline power ascension data establishes a consistent basis for measuring the performance of the reactor and the turbine control valves.

Demonstration of acceptable fuel thermal margin will be performed prior to and during power ascension to the TPO RTP at each steady-state heat balance point defined above. Fuel thermal margin will be projected to the TPO RTP point after the measurements taken at 95% and 100%

of CLTP to show the estimated margin. The thermal margin will be confirmed by the measurements taken at full TPO RTP conditions. The demonstration of core and fuel conditions will be performed with the methods currently used at Limerick.

Performance of the pressure and FW/level control systems will be recorded at each steady-state point defined above. The checks will utilize the methods and criteria described in the original startup testing of these systems to demonstrate acceptable operational capability. Water level changes of +/-3 inches and pressure setpoint step changes of +/-3 psi will be used. If necessary, adjustments will be made to the controllers and actuator elements.

The increase in power for the TPO uprate is sufficiently small that large transient tests are not necessary. High power testing performed during initial startup demonstrated the adequacy of the safety and protection systems for such large transients. Operational occurrences have shown the unit response is clearly bounded by the safety analyses for these events. ((

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10.5 OPERATOR TRAINING AND HUMAN FACTORS No additional training (apart from normal training for plant changes) is required to operate the plant in the TPO uprate condition. For TPO uprate conditions, operator response to transient, accident, and special events is not affected. Operator actions for maintaining safe shutdown, core cooling, containment cooling, etc., do not change for the TPO uprate. Minor changes to the P/F map, flow-referenced setpoint, and the like, will be communicated through normal operator training. Simulator changes and validation for the TPO uprate will be performed in accordance with established Limerick plant certification testing procedures.

10.6 PLANT LIFE Two degradation mechanisms may be influenced by the TPO uprate: (1) Irradiation Assisted Stress Corrosion Cracking (IASCC) and (2) FAC. The increase in irradiation of the core internal components influences IASCC. The increases in steam and FW flow rate influence FAC.

However, the sensitivity to the TPO uprate is small and various programs are currently implemented to monitor the aging of plant components, including EQ, FAC, and In-service Inspection. EQ is addressed in Section 10.3, and FAC is addressed in Section 3.5. These programs address the degradation mechanisms and do not change for the TPO uprate. The core 10-5

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION internals see a slight increase in fluence, but the inspection strategy used at Limerick, based on the BWRVIP, is sufficient to address the increase. The Maintenance Rule also provides oversight for the other mechanical and electrical components, important to plant safety, to guard against age-related degradation.

The longevity of most equipment is not affected by the TPO uprate because there is no significant change in the operating conditions. No additional maintenance, inspection, testing, or surveillance procedures are required.

10.7 NRC AND INDUSTRY COMMUNICATIONS NRC and Industry communications are generically addressed in the TLTR, Section 10.8. Per the TLTR, it is not necessary to review prior dispositions of NRC and industry communications and no additional information is required in this area.

10.8 PLANT PROCEDURES AND PROGRAMS Plant procedures and programs are in place to:

1. Monitor and maintain instrument calibration during normal plant operation to assure that instrument uncertainty is not greater than the uncertainty used to justify the TPO uprate;
2. Control the software and hardware configuration of the associated instrumentation;
3. Perform corrective actions, where required, to maintain instrument uncertainty within limits;
4. Report deficiencies of the associated instruments to the manufacturer; and
5. Receive and resolve the manufacturers deficiency reports.

10.9 EMERGENCY OPERATING PROCEDURES The Emergency Operating Procedures (EOPs) action thresholds are plant unique and will be addressed using standard procedure updating processes. It is expected that the TPO uprate will have a negligible or no effect on the operator action thresholds and to the EOPs in general.

10.10 INDIVIDUAL PLANT EXAMINATION Limerick maintains and regularly updates a station probabilistic risk assessment (PRA) model.

Use of the model is integrated with station operations and decision-making.

The Limerick IPE (PRA) model and analysis will not be specifically updated for TPO, because the change in plant risk from the subject power uprate is insignificant. This conclusion is supported by NRC RIS 2002-03 (Reference 4). In response to feedback received during the public workshop held on August 23, 2001, the NRC wrote, The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant effect on plant risk. Therefore, no risk information is requested to support such applications.

10-6

NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION

11.0 REFERENCES

1. GE Nuclear Energy, Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization, Licensing Topical Report, NEDC-32938P-A, Revision 2, May 2003; and NEDO-32938-A, Revision 2, May 2003.
2. GE Nuclear Energy, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate, (ELTR1), Licensing Topical Reports NEDC-32424P-A, February 1999; and NEDO-32424, April 1995.
3. GE Nuclear Energy, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate, (ELTR2), Licensing Topical Reports NEDC-32523P-A, February 2000; and NEDO-32523, April 1991; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999; and NEDO-32523, Supplement 1, January 1999.
4. NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, January 31, 2002.
5. GE Nuclear Energy, BWR Owners Group Long-Term Stability Solutions Licensing Methodology, NEDO-31960-A and NEDO-31960-A Supplement 1, November 1995.
6. GE Nuclear Energy, Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, NEDO-32465-A, August 1996.
7. GE Nuclear Energy, Reactor Long-Term Stability Solution Option III: Licensing Basis Hot Channel Oscillation Magnitude for Limerick 1 and 2, GE-NE-0000-0035-6037-R1, February 2006.
8. OG 02-0119-260, Backup Stability Protection (BSP) for Inoperable Option III Solution, July 17, 2002.
9. GE Nuclear Energy, 10 CFR 50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 through BWR/6 Vessels, NEDO-32205-A, Revision 1, February 1994.
10. Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988.
11. GE Nuclear Energy, Pressure-Temperature Curves for PECO Energy Company Limerick Unit 1, GE-NE-B11-00836-00-01, Revision 0, April 2000 and GE-NE-B11-00836-00-01a NP, Revision 0, April 2000.

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NEDO-33484 REVISION 0 NON-PROPRIETARY INFORMATION

12. GE Nuclear Energy, Pressure-Temperature Curves for PECO Energy Company Limerick Unit 2, GE-NE-B11-00836-00-02, Revision 0, July 2000 and GE-NE-B11-00836-00-02a NP, Revision 0, July 2000
13. BWRVIP-135, Revision 1, BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations, EPRI, Palo Alto, CA, June 2007 (TR-1013400).
14. C.I. Grimes (NRC) to Carl Terry (Niagara Mohawk Power Company), Acceptance For Referencing Of EPRI Proprietary Report TR-113596, BWR Vessel And Internals Project, BWR Reactor Pressure Vessel Inspection And Flaw Evaluation Guidelines (BWRVIP-74),

and Appendix A, Demonstration Of Compliance With the Technical Information Requirements Of The License Renewal Rule (10 CFR 54.21), October 18, 2001.

15. NRC Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds, November 1998.
16. GE Nuclear Energy, GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1 Additional Information for Upper Bound PCT Calculation, NEDE-23785P-A, Supplement 1, Revision 1, March 2002.
17. GE Nuclear Energy, Relaxation of UB PCT Limit for Limerick Units 1 and 2, GE-NE-0000-0010-5377-R0, December 2002.
18. GE Nuclear Energy, Constant Pressure Power Uprate, Licensing Topical Report, NEDC-33004P-A, Revision 4, June 2003.
19. GE Nuclear Energy, General Electric Instrument Setpoint Methodology, NEDC-31336P-A, September 1996 and NEDO-31336-A, September 1996.
20. GE Nuclear Energy, Mitigation of BWR Core Thermal-Hydraulic Instabilities in ATWS, NEDO-32164, December 1992.
21. GE Nuclear Energy, ATWS Rule Issues Relative to BWR Core Thermal-Hydraulics Stability, NEDO-32047-A, June 1995.

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