Text

NEDO-33477, Revision 0, Safety Analysis Report for Grand Gulf Nuclear Station Constant Pressure Power Uprate.
	ML102660399
Person / Time
Site:	Grand Gulf
Issue date:	08/31/2010
From:	; GE-Hitachi Nuclear Energy Americas
To:	; Office of Nuclear Reactor Regulation
References
	DRF 0000-0094-4784, GNRO-2010/00056 NEDO-33477, Revision 0
	Download: ML102660399 (537)
	v • d • e

{{#Wiki_filter:Attachment 5A GNRO-2010/00056 Safety Analysis Report for Grand Gulf Nuclear Station Constant Pressure Power Uprate (Non-Proprietary) This is a non-proprietary version of Attachment 5B from which the proprietary information has been removed. The proprietary portions that have been removed are indicated by double square brackets as shown here: [[ ]].

NEDO-33477 Revision 0 Class I DRF 0000-0094-4784 August 2010 Non-Proprietary Information SAFETY ANALYSIS REPORT FOR GRAND GULF NUCLEAR STATION CONSTANT PRESSURE POWER UPRATE

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION ii INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33477P, Revision 0, which has the proprietary information removed. Portions of the document that have been removed are indicated

by an open and closed bracket 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 purpose of supporting the Grand Gulf Nuclear Station license amendment request for an extended power uprate 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 par ticipating 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 inte nded 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.

iii TABLE OF CONTENTS Page EXECUTIVE

SUMMARY

.........................................................................................................xxv 1 INTRODUCTION....................................................................................................................1-1 1.1 Report Approach........................................................................................................1-1 1.1.1 Generic Assessments.....................................................................................1-1 1.1.2 Plant-Specific Evaluation..............................................................................1-3 1.2 Purpose and Approach...............................................................................................1-3 1.2.1 Uprate Analysis Basis....................................................................................1-4

1.2.2 Computer

Codes.............................................................................................1-4 1.2.3 Approach........................................................................................................1-4 1.3 EPU Plant Operating Conditions...............................................................................1-7 1.3.1 Reactor Heat Balance.....................................................................................1-7

1.3.2 Reactor

Performance Improvement Features.................................................1-7 1.4 Summary and Conclusions........................................................................................1-7 2 SAFETY EVALUATION........................................................................................................2-1 2.1 Materials and Chemical Engineering.........................................................................2-1 2.1.1 Reactor Vessel Material Surveillance Program.............................................2-1 2.1.2 Pressure-Temperature Limits and Upper Shelf Energy.................................2-2

2.1.3 Reactor

Internal and Core Support Materials................................................2-4

2.1.4 Reactor

Coolant Pressure Boundary Materials..............................................2-7

2.1.5 Protective

Coating Systems (Paints) - Organic Materials.............................2-9 2.1.6 Flow Accelerated Corrosion........................................................................2-11

2.1.7 Reactor

Water Cleanup System...................................................................2-15 2.2 Mechanical and Civil Engineering..........................................................................2-40 2.2.1 Pipe Rupture Locations and Associated Dynamic Effects..........................2-40 2.2.2 Pressure-Retaining Components and Component Supports........................2-44 2.2.3 Reactor Pressure Vessel Internals and Core Supports.................................2-66 2.2.4 Safety-Related Valves and Pumps...............................................................2-80

2.2.5 Seismic

and Dynamic Qualification of Mechanical and Electrical Equipment....................................................................................................2-87 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

iv 2.3 Electrical Engineering............................................................................................2-127 2.3.1 Environmental Qualification of Electrical Equipment..............................2-127 2.3.2 Off-site Power System...............................................................................2-131 2.3.3 AC On-Site Power System.........................................................................2-134 2.3.4 DC On-Site Power System.........................................................................2-136

2.3.5 Station

Blackout.........................................................................................2-138 2.4 Instrumentation and Controls.................................................................................2-153 2.4.1 Reactor Protection, Safety Features Actuation, and Control Systems.......2-153 2.4.2 MSL High Flow Group 1 Isolation Instrument Setpoint Sample Calculation (Input/Output Document).......................................................2-164 2.5 Plant Systems.........................................................................................................2-179

2.5.1 Internal

Hazards.........................................................................................2-179

2.5.2 Fission

Product Control.............................................................................2-196

2.5.3 Component

Cooling and Decay Heat Removal.........................................2-203 2.5.4 Balance-of-Plant Systems..........................................................................2-216

2.5.5 Waste

Management Systems.....................................................................2-225

2.5.6 Additional

Considerations.........................................................................2-234

2.5.7 Additional

Review Areas (Plant Systems).................................................2-236 2.6 Containment Review Considerations.....................................................................2-258 2.6.1 Primary Containment Functional Design..................................................2-258

2.6.2 Subcompartment

Analyses.........................................................................2-265 2.6.3 Mass and Energy Release..........................................................................2-272

2.6.4 Combustible

Gas Control in Containment.................................................2-274

2.6.5 Containment

Heat Removal.......................................................................2-275

2.6.6 Secondary

Containment Functional Design..............................................2-278 2.7 Habitability, Filtration, and Ventilation.................................................................2-293 2.7.1 Control Room Habitability System............................................................2-293

2.7.2 Engineered

Safety Feature Atmosphere Cleanup......................................2-294

2.7.3 Control

Room Area Ventilation System....................................................2-296

2.7.4 Spent

Fuel Pool Area Ventilation System.................................................2-298

2.7.5 Auxiliary

and Radwaste Area and Turbine Areas Ventilation Systems....2-299

2.7.6 Engineered

Safety Feature Ventilation System.........................................2-301 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

v 2.8 Reactor Systems.....................................................................................................2-304 2.8.1 Fuel System Design...................................................................................2-304 2.8.2 Nuclear Design..........................................................................................2-305

2.8.3 Thermal

and Hydraulic Design..................................................................2-316

2.8.4 Emergency

Systems...................................................................................2-321

2.8.5 Accident

and Transient Analyses..............................................................2-338 2.8.6 Fuel Storage...............................................................................................2-371 2.9 Source Terms and Radiological Consequences Analyses.....................................2-422 2.9.1 Source Terms for Radwaste Systems Analyses.........................................2-422

2.9.2 Radiological

Consequences Analyses Using Alternative Source Terms...2-426

2.9.3 Additional

Review Areas (Radiological Consequences Analyses)...........2-429 2.10 Health Physics........................................................................................................2-44 4 2.10.1 Occupational and Public Radiation Doses.................................................2-444 2.11 Human Performance..............................................................................................2-458 2.11.1 Human Factors...........................................................................................2-458 2.12 Power Ascension and Testing Plan........................................................................2-464 2.12.1 Approach to EPU Power Level and Test Plan...........................................2-464 2.13 Risk Evaluation......................................................................................................2-467 2.13.1 Risk Evaluation of EPU.............................................................................2-467 3 REFERENCES.........................................................................................................................3-1 Appendices A - Limitations from Safety Evaluation for LTR NEDC-33173P-.............----..---A-1 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

vi TABLES Table 1-1 Computer Codes Used For EPU Table 1-2 Current and EPU Plant Operating Conditions Table 2.1-1 Upper Shelf Energy - 40 Year License (35 EFPY) Table 2.1-2 Adjusted Reference Temperatures Year License (35 EFPY) Table 2.1-3 35 EFPY Effects of Irradiation on RPV Circumferential Weld Properties Table 2.1-4 Comparison of Key Parameters Influencing FAC Wear Rate, GGNS CLTP vs. EPU Table 2.1-5 Sample of Components with Highest Predicted Wear Rates, GGNS CHECWORKS TM SFA-Predicted Thickness vs. Measured Thickness Table 2.1-6 Comparison of RWCU System Operating Conditions Table 2.1-7 Comparisons of Chemistry Parameters for CLTP and EPU Cases Table 2.1-8 Selection Process Criteria for Components in the FAC Program Table 2.2-1 High Energy Line Breaks Table 2.2-2 RCPB Structural Evaluation Table 2.2-3a Percentage Increase In Class 1 Pipe Stresses, Usage Factors, Interface Loads, and Thermal Displacements for GGNS Piping Systems Due to EPU Conditions Table 2.2-3b Percent Increase In Class 2 and/or 3 Pipe Stresses, Interface Loads, and Displacements for GGNS Piping Systems Due To EPU Conditions Table 2.2-4a Summary of MS ASME Cla ss 1 Piping, Pipe Stresses, and CUFs Table 2.2-4b Check of Pipe Breaks Against All Nodes Above Maximum Code Stresses Table 2.2-4c Summary of MS ASME Cla ss 1 Piping, Pipe Stresses, and CUFs Table 2.2-4d Check of Pipe Breaks Against All Nodes Above Maximum Code Stresses Table 2.2-4e MS (Lines A & D) Penetration Load Summary (Inside Containment) Table 2.2-4f MS (Lines B & C) Penetration Load Summary (Inside Containment) Table 2.2-4g MS (Lines A & D) Nozzle Loading Summary Table 2.2-4h MS (Lines B & C) Nozzle Loading Summary Table 2.2-4i MS (Lines A & D) SRV Flange Moment Summary (Inlet & Outlet) Table 2.2-4j MS (Lines B & C) SRV Flange Moment Summary (Inlet & Outlet) Table 2.2-4k MS (Lines A & D) Pipe Support Load Summary Table 2.2-4l MS (Lines B & C) Pipe Support Load Summary Table 2.2-5a BOP MS System Class 1 Piping (Outside Containment) Table 2.2-5b BOP MS System Class 2 and Non-Safety Related Piping (Outside Containment) NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

vii Table 2.2-5c BOP FW System Class 1 Piping (Outside Containment) Table 2.2-5d BOP FW System Class 2 and Non-Safety Related Piping (Outside Containment) Table 2.2-5e BOP Piping - Extraction Steam Table 2.2-5f BOP Piping - FW Heater Drains & Vents (HDL and HDH) Table 2.2-5g BOP Piping - Condensate Table 2.2-5h BOP Piping - ECCS Piping Table 2.2-6 BOP Piping System Evaluation Table 2.2-7 CUFs and S p+q Values of Limiting Components Table 2.2-8 RIPDs for Normal Conditions Table 2.2-9 RIPDs for Upset Conditions Table 2.2-10 RIPDs for Faulted Conditions Table 2.2-11 Governing Stress Results for RPV Internal Components Table 2.2-12 Fatigue Usage for RPV Internal Components for Plant Life of 40-Years Table 2.2-13 GGNS Program Pumps and Valves Table 2.2-14 Ambient Temperature Effects to GGNS Program Valves Table 2.3-1 Group II Partially Qualified Components Table 2.3-2 Group III, Non-Qualified Components Table 2.3-3 Electrical Equipment Ratings and Margins Table 2.3-4 Electrical Distribution System Load Increases Table 2.3-5 Battery Margin at CLTP and EPU Table 2.3-6 GGNS SBO Sequence of Events Table 2.4-1 Technical Specification Function Information Table 2.4-2 Changes to Instrumentation and Controls Table 2.5-1 Appendix R Fire Event Evaluation Results Table 2.5-2 SGTS Iodine Removal Capacity Parameters Table 2.5-3 Spent Fuel Pool Response Table 2.5-4 EPU TBCW Effect Table 2.5-5 Basis for Classifica tion of No Significant Effect Table 2.6-1 GGNS Containment Performance Results Table 2.6-2 Long-Term Containment Response Key Analysis Input Values Table 2.6-3 GGNS Peak SP Temperatures for Postulated ATWS, SBO, and 10 CFR 50 Appendix R Events NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

viii Table 2.6-4 Overview of AP Load Evaluation Methodologies Table 2.8-1 Peak Nodal Exposures Table 2.8-2 Option III Setpoint Demonstration Table 2.8-3 ODYSY Decay Ratios at BSP Region Boundary Endpoints Table 2.8-4 Parameters Used For Transient Analysis Table 2.8-5 Transient Analysis MCPR Results Table 2.8-6 Key Input Parameters for SAFER/GESTR LOCA Evaluation Table 2.8-7 ECCS Conformance Results Table 2.8-8 GGNS Key Inputs for ATWS Analysis Table 2.8-9 GGNS Results for ATWS Analysis Table 2.8-10 MSIVC Sequence of Events Table 2.8-11 PRFO Sequence of Events Table 2.9-1 Total Activity Levels Table 2.9-2 Activity Concentrations of Principal Radionuclides in Fluid Streams for Normal EPU Operation Table 2.9-3 EPU Noble Gas Radionuclide Source Term and Design Basis Comparison Table 2.9-4 EPU Fission Product Reactor Water Comparisons to Design Basis Values Table 2.9-5 EPU Activated Corrosion Product Reactor Water Comparisons to Design Basis Values Table 2.9-6 EPU Coolant Activation Product Reactor Water and Steam Comparisons to Design Basis Values Table 2.9-7 LOCA Radiological Consequences Table 2.9-8 FHA Radiological Consequences Table 2.9-9 CRDA Radiological Consequences Table 2.9-10 MSLBA Pre-Incident Iodine Spike Radiological Consequences Table 2.9-11 MSLBA Equilibrium Iodine C oncentration Radiological Consequences Table 2.9-12 Pressure Controller Failure Radiological Consequences Table 2.9-13 MSIVC Radiological Consequences Table 2.9-14 Offgas System Leak or Failure Radiological Consequences Table 2.9-15 Radioactive Liquid Waste System Le ak or Failure Radiological Consequences Table 2.9-16 Liquid Radwaste Tank Failure (Release to Groundwater) Radiological Consequences Table 2.9-17 Recirculation Pump Seizure Radiological Consequences NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

ix Table 2.10-1 Current and Anticipated Measured Radiation Fields in Selected Areas Table 2.10-2 EPU Vital Area Access Mission Doses Table 2.10-3 Estimated Annual Doses to Members of the Public Due to Normal Operation Gaseous and Liquid Radwaste Effluents Table 2.10-4 Direct Shine Annual Dose to Members of the Public

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

x FIGURES Figure 1-1 Power/Flow Operating Map for EPU Figure 1-2 EPU Heat Balance - Nominal Figure 1-3 EPU Heat Balance - Overpressure Protection Analysis Figure 2.3-1 Worst Case EQ Enveloping Accident Temperature Profiles (All Plant EQ Zones and Elevations) Figure 2.5-1 Appendix R Evaluation Results EPU Containment (SHEX) - SP Temperature Figure 2.5-2 Appendix R Evaluation Results 102.46% CLTP Containment (SHEX) - SP Temperature Figure 2.5-3 Appendix R Evaluation Results EP U Fuel Heatup (SAFER) - Hot and Average Channel Water Level Figure 2.5-4 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) - Reactor Vessel Pressure Figure 2.5-5 Appendix R Evaluation Results EP U Fuel Heatup (SAFER) - Peak Cladding Temperature Figure 2.5-6 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) - Water Level Outside the Shroud Figure 2.5-7 Reactive Capability Curve Figure 2.6-1 Long-Term DBA LOCA Temperature Response at EPU Figure 2.6-2 Long-Term ASDC SP and WW Temperature Response at EPU Figure 2.6-3 Long-Term ASDC DW and WW Temperature Response at EPU Figure 2.6-4 Short-Term DBA LOCA MSLB Pressure Response at EPU Figure 2.6-5 Short-Term DBA LOCA MSLB Differential Pressure Response at EPU Figure 2.8-1 Power of Peak Bundle versus Cycle Exposure Figure 2.8-2 Coolant Flow for Peak Bundle versus Cycle Exposure Figure 2.8-3 Exit Void Fraction for Peak Power Bundle versus Cycle Exposure Figure 2.8-4 Maximum Channel Exit Void Fraction versus Cycle Exposure Figure 2.8-5 Core Average Exit Void Fraction versus Cycle Exposure Figure 2.8-6 Peak LHGR versus Cycle Exposure Figure 2.8-7 Dimensionless Bundle Power at BOC (200 MWd/ST) Figure 2.8-8 Dimensionless Bundle Power at MOC (9000 MWd/ST) Figure 2.8-9 Dimensionless Bundle Power at EOC (18615 MWd/ST) NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xi Figure 2.8-10 Bundle Operating LHGR (KW/ft) at BOC (200 MWd/ST) Figure 2.8-11 Bundle Operating LHGR (kw/ft) at MOC (9000 MWd/ST) Figure 2.8-12 Bundle Operating LHGR (kw/ft) at EOC (18615 MWd/ST) Figure 2.8-13 Bundle Operating MCPR at BOC (200 MWd/ST) Figure 2.8-14 Bundle Operating MCPR at MOC (9000 MWd/ST) Figure 2.8-15 Bundle Operating MCPR at EOC (18615 MWd/ST) Figure 2.8-16 Bundle Operating LHGR (kw/ft) at 11000 MWd/ST (peak MFLPD point) Figure 2.8-17 Bundle Operating MCPR at 2000A MWd/ST (peak MFLCPR point) Figure 2.8-18 Bundle Average Void Fraction vs. MCPR Figure 2.8-19 Illustration of OPRM Tr ip-Enabled Region and BSP Regions Figure 2.8-20 Response to MSIVF (MSIV Closure with High Flux Scram) Figure 2.8-21 Loss of Feedwater Flow Figure 2.8-22 EPU MELLLA BOC MSIVC (Short-Term) Figure 2.8-23 EPU MELLLA BOC MSIVC (Long-Term) Figure 2.8-24 EPU MELLLA BOC MSIVC (Long-Term) Figure 2.8-25 EPU MELLLA BOC MSIVC (Long-Term) Figure 2.8-26 EPU MELLLA BOC PRFO (Short-Term) Figure 2.8-27 EPU MELLLA BOC PRFO (Long-Term) Figure 2.8-28 EPU MELLLA BOC PRFO (Long-Term) Figure 2.8-29 EPU MELLLA BOC PRFO (Long-Term) Figure 2.8-30 EPU MELLLA EOC MSIVC (Short-Term) Figure 2.8-31 EPU MELLLA EOC MSIVC (Long-Term) Figure 2.8-32 EPU MELLLA EOC MSIVC (Long-Term) Figure 2.8-33 EPU MELLLA EOC MSIVC (Long-Term) Figure 2.8-34 EPU MELLLA EOC PRFO (Short-Term) Figure 2.8-35 EPU MELLLA EOC PRFO (Long-Term) Figure 2.8-36 EPU MELLLA EOC PRFO (Long-Term) Figure 2.8-37 EPU MELLLA EOC PRFO (Long-Term) NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xii ACRONYMS AND ABBREVIATIONS Term Definition AAC Alternate AC Sources ABA Amplitude Based Algorithm ABVS Auxiliary Building Ventilation System AC Alternating Current ACT Auxiliary Cooling Tower ADS Automatic Depressurization System ADSOOS Automatic Depressurization System Out-of-Service AHC Access Hole Cover AIZ Above Instrument Zero AL Analytical Limit ALARA As Low As Is Reasonably Achievable ANS American Nuclear Society ANSI American National Standards Institute AOO Anticipated Operational Occurrence (moderate frequency transient event) AOP Abnormal Operating Procedure AOR Analysis of Record AOV Air-Operated Valve AP Annulus Pressurization APEA Accuracy - Primary Element Accuracy APLHGR Average Planar Linear Heat Generation Rate APRM Average Power Range Monitor ARI Alternate Rod Insertion ARAVS Auxiliary and Radwaste Area Ventilation System ART Adjusted Reference Temperature ARTS APRM/RBM/Technical Specifications ASME American Society of Mechanical Engineers ASDC Alternate Shutdown Cooling AST Alternate Source Term ATWS Anticipated Transient Without Scram AV Allowable Value NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xiii Term Definition AVZ Above Vessel Zero bhp Brake Horsepower BIIT Boron Injection Initiation Temperature BOC Beginning of Cycle BOP Balance-of-Plant B&PV Boiler and Pressure Vessel BPWS Banked Position Withdrawal Sequence BSP Backup Stability Protection BSW Biological Shield Wall BTU British Thermal Unit BWR Boiling Water Reactor BWROG BWR Owners' Group BWRVIP BWR Vessel and Internals Project CCW Component Cooling Water CDF Core Damage Frequency CF Chemistry Factor CFFF Condensate Full Flow Filtration cfm Cubic Feet Per Minute CFR Code of Federal Regulations CFS Condensate and Feedwater System CGCS Combustible Gas Control System CLR Containment Load Report CLTP Current Licensed Thermal Power CLTR Constant Pressure Power Uprate Licensing Topical Report CO Condensation Oscillation COLR Core Operating Limits Report CPPU Constant Pressure Power Uprate CR Control Room CRAVS Control Room Area Ventilation System CRD Control Rod Drive CRDA Control Rod Drop Accident NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xiv Term Definition CRDH Control Rod Drive Housing CREFS Control Room Emergency Filtration System CRGT Control Rod Guide Tube CS Core Spray CSC Containment Spray Cooling CST Condensate Storage Tank CT Current Transformer CUF Cumulative Usage Factor CWS Circulating Water System DBA Design Basis Accident DBLOCA Design Basis Loss-of-Coolant Accident DC Direct Current DHR Decay Heat Removal DIVOM Delta CPR over Initial CPR Versus Oscillation Magnitude DLO Dual (Recirculation) Loop Operation DP Differential Pressure DPEA Drift - Primary Element Accuracy DW Drywell EAB Exclusion Area Boundary ECCS Emergency Core Cooling System(s) EFDS Equipment and Floor Drainage System EFPY Effective Full Power Years ELLLA Extended Load Line Limit Analysis ELTR1 Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate Licensing Topical Report ELTR2 Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate Licensing Topical Report EOC End of Cycle EOP Emergency Operating Procedure EP Emergency Procedure EPG Emergency Procedure Guideline EPRI Electric Power Research Institute NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xv Term Definition EPU Extended Power Uprate EQ Environmental Qualification EQDP Environmental Qualification Documentation Package ESF Engineered Safety Feature ESFAS Engineered Safety Feature Actuation System ESFVS Engineered Safety Feature Ventilation System FAC Flow Accelerated Corrosion FEA Finite Element Analysis FFWTR Final Feedwater Temperature Reduction FHA Fuel Handling Accident FIL Flow Induced Load FIV Flow Induced Vibration FPC Fuel Pool Cooling FPCCS Fuel Pool Cooling and Cleanup System FPP Fire Protection Program FV Fussell-Vesely FW Feedwater FWCF Feedwater Controller Failure Maximum Demand FWH Feedwater Heater FWHOOS Feedwater Heater Out-of-Service FWLB Feedwater Line Break FWLC Feedwater Leakage Control GDC General Design Criterion GE General Electric GEH GE-Hitachi Nuclear Energy Americas LLC GGNS Grand Gulf Nuclear Station GI-Lli Gastro-Intestinal - Lower Large Intestine GL Generic Letter gm Gram GNF Global Nuclear Fuel LLC GRA Growth Rate Based Algorithm NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xvi Term Definition GWd/MTU Gigawatt Days per Metric Ton Uranium GWMS Gaseous Waste Management System HCTL Heat Capacity Temperature Limit HCU Hydraulic Control Unit HDP Heater Drain Pump HDT Heater Drain Tank HELB High Energy Line Break HEM Homogeneous Equilibrium Model HEP Human Error Probability HEPA High Efficiency Particulate Air HFCL High Flow Control Line Hg a Inches of Mercury Absolute HP High Pressure HPCS High Pressure Core Spray HPSP High Power Setpoint HRA Human Reliability Analysis HSR Hydraulic System Return HVAC Heating, Ven tilating, and Air Conditioning HWC Hydrogen Water Chemistry HWL High Water Level HX Heat Exchanger I&C Instrumentation and Control IASCC Irradiation-Assisted Stress Corrosion Cracking ICA Interim Corrective Action ICF Increased Core Flow IEB Inspection and Enforcement Bulletin ICHGT In-Core Housing and Guide Tube IEEE Institute of Electrical and Electronics Engineers IGSCC Intergranular Stress Corrosion Cracking IPB Isolated Phase Bus IPE Individual Plant Examination NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xvii Term Definition IPEEE Individual Plant Examination of External Event IRM Intermediate Range Monitor ISI In-Service Inspection ISP Integrated Surveillance Program IST Inservice Testing JI Jet Impingement JR Jet Reaction LAR License Amendment Request LCO Limiting Condition for Operation LCS Leakage Control System LDFS Low-Density Fuel Storage LDI Liquid Droplet Impingement LDS Leak Detection System LER Licensee Event Report LERF Large Early Release Frequency LFWH Loss of Feedwater Heating LHGR Linear Heat Generation Rate LLHS Light Load Handling System LLS Low-Low Set LOCA Loss-of-Coolant Accident LOCV Loss of Condenser Vacuum LOFW Loss of Feedwater LOOP Loss of Off-Site Power LP Low Pressure LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray LPRM Local Power Range Monitor LPSP Low Power Setpoint LPZ Low Population Zone LRNBP Generator Load Rejection with Steam Bypass Failure LTR Licensing Topical Report NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xviii Term Definition LWL Low Water Level LWMS Liquid Waste Management System MAAP Modular Accident Analysis Program MAPLHGR Maximum Average Planar Linear Heat Generation Rate MBTU Millions of BTUs MC Main Condenser MCC Motor Control Center MCES Main Condenser Evacuation System MCPR Minimum Critical Power Ratio MEDP Maximum Expected Differential Pressure MELB Moderate Energy Line Break MELLLA Maximum Extended Load Line Limit Analysis MEOD Maximum Extended Operating Domain MeV Million Electron Volts MFLCPR Maximum Fraction of Limiting Critical Power Ratio MFLPD Maximum Fraction of Limiting Power Density Mlb Millions of Pounds MOC Middle of Cycle MOP Member of the Public MOV Motor-Operated Valve MPS Minimum Pump Speed MRS Main and Reheat Steam MS Main Steam MSDT Moisture Separator (Chevron) Drain Tank MSF Modified Shape Function MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSIVD Closure of Main Steam Isolation Valves with Direct Scram MSIVF Main Steam Isolation Valve Closure with Scram on High Flux MSIVOOS Main Steam Isolation Valve Out-of-Service MSL Main Steam Line NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xix Term Definition MSLB Main Steam Line Break MSLBA Main Steam Line Break Accident MSR Moisture Separator Reheater MSRS Main Steam and Reheat System MSRV Main Steam Relief Valve MSS Moisture Separator Shell MSSDT Moisture Separator Shell Drain Tank MST Main Steam Tunnel MVA Million Volt Amps Mvar Megavar MWd Megawatt Days MWd/ST Megawatt Days per Short Ton MWe Megawatts-Electric MWt Megawatt-Thermal NA Not Applicable NCL Natural Circulation Line NDE Non-Destructive Examination NEI Nuclear Energy Institute NHRX Non-Regenerative Heat Exchanger NPSH Net Positive Suction Head NPSH A Net Positive Suction Head Available NPSH R Net Positive Suction Head Required NR Not Reported NRC Nuclear Regulatory Commission NSI Next Scheduled Inspection NSSS Nuclear Steam Supply System NTSP Nominal Trip Setpoint NUMAC Nuclear Measurement Analysis and Control NUREG Nuclear Regulatory Commission Technical Report Designation OD Outside Diameter ODB Original Design Basis NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xx Term Definition OE Operating Experience OEM Original Equipment Manufacturer OFS Orificed Fuel Support OLMCPR Operating Limit Minimum Critical Power Ratio OLTP Original Licensed Thermal Power OM Operation and Maintenance ONEP Off-Normal Event Procedure OOS Out-of-Service OPRM Oscillation Power Range Monitor P Differential Pressure - psi P 25 25% of EPU Rated Thermal Power Pa Peak Containment (Internal) Pressure PBDA Period Based Detection Algorithm PCS Pressure Control System PCT Peak Cladding Temperature PDA Pipe Dynamic Analysis PDT Pressure Differential Transmitter PEA Primary Element Accuracy PF Power Factor PFS Peripheral Fuel Support PMA Process Measurement Accuracy PRA Probabilistic Risk Assessment PRD Power Range Detector PRFD Pressure Regulator Failure Downscale PRFO Pressure Regulator Failure - Open PRNM Power Range Neutron Monitoring psi Pounds per Square Inch psia Pounds per Square Inch - Absolute psid Pounds per Square Inch - Differential psig Pounds per Square Inch - Gauge PSP Pressure Suppression Pressure NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xxi Term Definition PSW Plant Service Water P-T Pressure-Temperature PUSAR Power Uprate Safety Analysis Report PWR Pipe Whip Restraint QST Quality Steam Turbine RAVS Radwaste Area Ventilation System RAW Risk Achievement Worth RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCIS Rod Control and Information System RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RDLB Recirculation Discharge Line Break RDT Reheater Drain Tank rem Roentgen Equivalent Man RFO Refueling Outage RFP Reactor Feed Pump RFPT Reactor Feed Pump Turbine RFW Reactor Feedwater RFWT Reduced Feedwater Temperature RG Regulatory Guide RH Relative Humidity RHR Residual Heat Removal RHX Regenerative Heat Exchanger RIPD Reactor Internal Pressure Difference RLA Reload Licensing Analysis RLB Recirculation Line Break RPC Rod Pattern Controller RPS Reactor Protection System RPT Recirculation Pump Trip RPTOOS Recirculation Pump Trip Out-of-Service NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xxii Term Definition RPV Reactor Pressure Vessel RRS Reactor Recirculation System RSL Remaining Service Life RSLB Recirculation Suction Line Break RT NDT Reference Temperature of the Nil-Ductility Transition RTP Rated Thermal Power RWCU Reactor Water Cleanup RWE Rod Withdrawal Error RWM Rod Worth Minimizer S m Code Allowable Stress Limit SAFDL Specified Acceptable Fuel Design Limit SAG Severe Accident Guideline SAP Severe Accident Procedure SAW Submerged Arc Weld SBO Station Blackout SCC Stress Corrosion Cracking scfh Standard Cubic Feet Per Hour scfm Standard Cubic Feet Per Minute SDC Shutdown Cooling SE Safety Evaluation SER Safety Evaluation Report SFIE Steam Flow Induced Error SF Spent Fuel SFP Spent Fuel Pool SFPAVS Spent Fuel Pool Area Ventilation System SGTS Standby Gas Treatment System SIL Service Information Letter SJAE Steam Jet Air Ejector SLC Standby Liquid Control SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xxiii Term Definition SLO Single Loop Operation SMA Seismic Margins Analysis SMAW Shielded Metal Arc Weld S-NM Susceptible - Not Modeled SORV Stuck Open Relief Valve SOV Solenoid-Operated Valve SP Suppression Pool SPC Suppression Pool Cooling SPDS Safety Parameter Display System SPMU Suppression Pool Make-up SRLR Supplemental Reload Licensing Report SRM Source Range Monitor SRP Standard Review Plan SRSS Square Root of the Sum of the Squares SRV Safety Relief Valve SRVDL Safety Relief Valve Discharge Line SRVOOS Safety Relief Valve Out-of-Service SS Seal Steam SSC Structure, System, and Component SSE Safe Shutdown Earthquake SSG Seal Steam Generator SSW Standby Service Water SWMS Solid Waste Management System SWS Station Service Water System TAF Top of Active Fuel TAVS Turbine Area Ventilation System TBCW Turbine Building Cooling Water TBS Turbine Bypass System TBVOOS Turbine Bypass Valve Out-of-Service TCV Turbine Control Valve TEDE Total Effective Dose Equivalent NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xxiv Term Definition TFSP Turbine First-Stage Pressure T-G Turbine-Generator TID Total Integrated Dose TLD Thermal Luminescence Dosimeter TRM Technical Requirements Manual TS Technical Specification TSC Technical Support Center TSV Turbine Stop Valve TSVC Turbine Stop Valve Closure TT Turbine Trip TTNBP Turbine Trip with Steam Bypass Failure UCP Upper Containment Pool UHS Ultimate Heat Sink UFSAR Updated Final Safety Analysis Report URL Upper Range Limit USE Upper Shelf Energy VB Vacuum Breaker VHCSN Vessel Head Cooling Spray Nozzle VWO Valves Wide Open WB Whole Body WW Wetwell

ZPA Zero Period Acceleration

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xxv EXECUTIVE

SUMMARY

This report summarizes the results of safety evaluations performed that justify uprating the licensed thermal power at Grand Gulf Nuclear Station (GGNS). The requested licensed power level is an increase to 4,408 MWt from the current licensed reactor thermal power of 3,898 MWt. GE-Hitachi Nuclear Energy Americas LLC (GEH) has previously developed and implemented Extended Power Uprate (EPU) at several nuclear power plants. Based on EPU experience, GEH developed an approach to uprate reactor power that maintains the current plant maximum normal operating reactor dome pressure. This approach is referred to as Constant Pressure Power Uprate (CPPU) and was approved by the Nuclear Regulatory Commission (NRC) in the Licensing Topical Report (LTR) NEDC-33004P-A , Revision 4, "Constant Pressure Power Uprate," hereafter referred to as the CLTR. The CLTR was approved for Boiling Water Reactor (BWR) plants containing General Electric (G E) fuel types through GE14 and using GEH accident analysis methods. GGNS contains onl y GE fuel types, through and including GNF2, and this evaluation uses only GEH accident analysis methods. Because GGNS uses GNF2 fuel, the CLTR is not applicable for fuel design dependent evaluations and the transients performed in s upport of the generic disposition in the CLTR are not applicable. Therefore, for fuel-dependent topics, this report follows the NRC approved generic content for BWR EPU licensing reports, documented in NEDC-32424P-A, "Generic Guidelines For General Electric Boiling Water Reactor Extended Power Uprate," commonly called "ELTR1." Per ELTR1, every safety issue th at should be addressed in a plant-specific EPU licensing report is addressed in this report. For issues that have been evaluated generically, this report references the NRC approved gene ric evaluations in NEDC-32523P-A, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," which is commonly called "ELTR2." This report reflects the systematic application of the CLTR approach to GGNS for topics that are not fuel-dependent, including the performance of plant-specific engineering assessments and confirmation of the applicability of the CLTR generic assessments required to support an EPU. By performing the power uprate in accordan ce with the CLTR, ELTR1, ELTR2, and their NRC Safety Evaluation Reports (SERs), the evaluation scope of the plant safety analyses and system performance is reduced, thus allowing for a more streamlined process. It is not the intent of this report to explicitly address all the details of the analyses and evaluations described herein. For example, onl y previously NRC-approved or industry-accepted methods were used for the analyses of accident s and transients, as referred to in the CLTR, ELTR1, or ELTR2. The safety analysis methods ha ve been previously addressed, and thus, are not explicitly addressed in this report. Also, event and analysis descriptions that are already

provided in other licensing reports or the Updated Final Safety Analysis Report (UFSAR) are not NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

xxvi repeated within this report. This report summa rizes the significant evaluations needed to support a license amendment to allow for uprated power operation. Uprating the power level of nuclear power plants can be done safely within plant-specific limits and is a cost-effective way to increase installe d electrical generating capacity. Many light water reactors have already been uprated worldwide, including many BWR plants. An increase in the electrical output of a BWR plant is accomplished primarily by generating and supplying higher steam flow to the turbine-genera tor (T-G). GGNS, as originally licensed, has an as-designed equipment and system capability to accommodate steam flow rates above the current rating. Also, the plant has sufficient design margins to allow the plant to be safely uprated significantly beyond its original licensed power level. A higher steam flow is achieved by increasing th e reactor power along specified control rod and core flow lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are recalibrated. Plant procedures are revised, and tests similar to some of the original startup tests are performed. Detailed evaluations of the reactor, engineered safety features (ESFs), power conversion, emergency power, support systems, and design basis accidents (DBAs) were performed. This report demonstrates that GGNS can safely operate at the requested EPU level. However, non-safety power generation modifications must be implemented in order to obtain the electrical power output associated with the uprate power. Until these modifications are completed, the non-safety, balance-of-plant (BOP) equipment may limit the electrical power output, which in turn may limit the operating thermal power level to less than the rated thermal power (RTP) level. All safety aspects of GGNS that are affected by the increase in thermal power have been evaluated. The evaluations and reviews were conducted in accordance with the CLTR and the criteria in ELTR1 using NRC-approved or industry-accepted analysis methods. The results of these evaluations and reviews presented in this report are as follows: Systems and components affected by EPU were reviewed to ensure there is no significant challenge to any safety system; No changes are requested that would require compliance with more recent industry standards and codes; and Limited hardware modifications are required to meet safety requirements, and any modification to power generation equipment will be implemented per 10 CFR 50.59.

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1-1 1 INTRODUCTION

1.1 Report

Approach This report summarizes the results of safety evaluations that were performed to justify uprating the licensed thermal power at GGNS. The request ed license power level is an increase to 4,408 MWt from the current licensed thermal power (CLTP) of 3,898 MWt. GE-Hitachi Nuclear Energy Americas LLC (GEH) has previously developed and implemented EPU at several nuclear power plants. Base d on EPU experience, GEH has developed an approach to uprating reactor power that maintains the current plant maximum normal operating reactor dome pressure. This approach is refe rred to as CPPU and is contained in the LTR NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," (Reference 1) hereafter referred to as the "CLTR." The NRC approve d the CLTR in the staff SER contained in Reference 1 for BWR plants containing GE fu el types through GE14 and using GEH accident analysis methods. The GGNS Cycle 19 core desi gn contains only GE fuel types, through and

including GNF2, and this evaluation uses only GEH accident analysis methods. By performing

the power uprate in accordance with the CLTR and within the constraints of the NRC SER, the evaluation scope of the plant safety analyses and system performance is reduced, thus allowing for a more streamlined process. Because GGNS uses GNF2 fuel, the CLTR is not applicable for fuel design dependent evaluations and the transients performed in s upport of the generic disposition in the CLTR are not applicable. Therefore, for fuel-dependent topics, this report follows the NRC-approved generic content for EPU licensing reports, as desc ribed in Section 3.0 and Appendices A & B of the

"Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," (Reference 2) hereafter referred to as ELTR1, and the NRC staff position letter reprinted in ELTR1. The analytical methodologies used for Emergency Core Cooling System (ECCS) - Loss-of-Coolant Accident (LOCA) evaluations and transient evaluations are documented in Appendices D and E, respectively in ELTR1. This evaluation justifies an EPU to 4,408 MWt, which corresponds to 115% of the original licensed thermal power (OLTP) for GGNS. This report is presented in a format consistent with the template SER contained in Section 3.2 of the US NRC, Office of Nuclear Reactor Regulation, Review Standard for Extended Power Uprates, RS-001, December 2003 (Reference 3). The Regulatory Evaluations from the template SER have been modified to reflect the licensing basis of GGNS.

1.1.1 Generic

Assessments Many of the component, system, and performance ev aluations contained within this report have been generically evaluated in the CLTR and th e "Generic Evaluations of General Electric NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-2 Boiling Water Reactor Extended Power Uprate," (Reference 4) hereafter referred to as ELTR2, and found to be acceptable by the NRC. The pl ant-specific applicability of these generic assessments is identified and confirmed in the applicable sections of this report. Generic assessments are those safety evaluations that can be dispositioned for a group or all BWR plants by: A bounding analysis for the limiting conditions, Demonstrating that there is a negligible effect due to EPU, or Demonstrating that the required plant cycle-specific reload analyses are sufficient and appropriate for establishing the EPU licensing basis. Bounding analyses may be based on either: (1) a demonstration that assessments provided in previous EPU LTRs that included a pressure increase (References 2 and 4) are bounding; or (2) on specific generic studies provided in the CLTR. For these bounding analyses, the current EPU experience is provided in the CLTR, ELTR 1, and ELTR2, along with the basis and results of the assessment. For those EPU assessments having a negligible effect, the current EPU experience plus a phenomenological discussion of the basis for the assessment is provided in the CLTR. For generic assessments that are GNF2 fuel design dependent, the assessments contained in ELTR1 and ELTR2 are applicable and analyzed with GEH methodology. Some of the safety evaluations affected by EP U are fuel cycle (reload) dependent. Reload dependent evaluations require that the reload fuel design, core loading pattern, and operational plan be established so that analyses can be performed to establish core operating limits. The reload analysis demonstrates that the core design for EPU meets the applicable NRC evaluation criteria and limits documented in Reference 5. Because of the lead tim e required for the NRC review of this power uprate submittal, the GGNS re load core design for the initial fuel cycle at uprated power conditions are not established at the time of this submittal. As discussed in Section 2.8.2, the EPU has a relatively small effect on core operating and safety limits. Therefore, the reload fuel design and core loading pattern dependent plant evaluations for EPU operation are performed with the reload analys is as part of the standard reload licensing process. No plant can implement a power uprate unless the appropriate reload core analysis is performed and all criteria and limits documented in Reference 5 are satisfied. Otherwise, the plant would be in an unanalyzed condition. Based on current requirements, the reload analysis results are documented in the Supplemental Relo ad Licensing Report (SRLR), and the applicable core operating limits are documented in the plant-specific Core Operating Limits Report (COLR). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-3 1.1.2 Plant-Specific Evaluation Plant-specific evaluations are assessments of the principal evaluations that are not addressed by the generic assessments described in Section 1.1.1. The relative effect of EPU on the plant-specific evaluations and the methods used for their performance are provided in this report. Where applicable, the assessment methodology is referenced. If a specific computer code is used, the name of this computer code is provided in the subsection. Table 1-1 provides a summary of the computer codes used. The plant-specific evaluations performed and reported in this document use plant-specific values to model the actual plant systems, transient response, and operating conditions. These plant-specific analyses are considered reload independent and are performed using a conservative core representative of GGNS design for operation at 115% of OLTP for a cycle length of 24 months. Although the GGNS fuel cycle length is currently 18 months, the plant-specific analyses contained in this report have been conservatively evaluated at a fuel cycle length of 24 months.

1.2 Purpose

and Approach An increase in electrical output of a BWR is accomplished primarily by generation and supply of higher steam flow to the T-G. Most BWRs, as originally licensed, have an as-designed equipment and system capability to accommodate steam flow rates at least 5% above the original rating. In addition, continuing improvements in the analytical techniques (computer codes) based on several decades of BWR safety technology, plant performance feedback, operating experience (OE), and improved fuel and core designs have resulted in significant increases in the design and operating margins between the calculated safety analyses results and the current plant licensing limits. The available margins in calculated results, combined with the as-designed excess equipment, system, and component capabilities (1) have allowed many BWRs to increase their thermal power ratings by 5% without any Nuclear Steam Supply System (NSSS) hardware modification, and (2) provide for power increases up to 20% with some non-safety hardware modifications. These power increases involve no significant increase in the hazards presented by the plants as approved by the NRC in the original license. The method for achieving higher power is to extend the power/flow map (Figure 1-1) along the Maximum Extended Load Line Limit Analysis (MELLLA) line. However, there is no increase in the maximum normal operating reactor vessel dome pressure or the maximum licensed core flow over their pre-EPU values. EPU operation does not involve increasing the maximum normal operating reactor vessel dome pressure, because the plant, after modifications to non-safety power generation equipment, has sufficient pressure control and turbine flow capabilities to control the inlet pressure conditions at the turbine. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-4 1.2.1 Uprate Analysis Basis GGNS is currently licensed at the 100% CLTP level of 3,898 MWt. The EPU RTP level included in this evaluation is 4,408 MWt, or 115% of the OLTP (113% of the CLTP). Plant-specific EPU parameters are listed in Table 1-2. In addition, as noted in Table 1-2, various performance enhancements have been conservativ ely considered in the EPU analyses. The EPU safety analyses are based on a power level of 1.02 times the EPU power level unless the Regulatory Guide (RG) 1.49 (Reference 6) two per cent power factor (PF) is already accounted for in the analysis methods consistent with the methodology described in Reference 5, or RG 1.49 does not apply (e.g., Anticipated Transient Without Scram (ATWS) and Station Blackout (SBO) events). GGNS is not requesting NRC approval of a cycle-length extension to 24 months with this power uprate. However, for the GGNS EPU core performance evaluations, an equilibrium core based on a 24-month reference fuel cycle has been applied. The EPU evaluations that require core and/or fuel characteristic data can be categorized as two types: (1) those that are performed to evaluate system or component design basis capabilities, and (2) those that are also performed on a cycle-specific basis, to establish or confirm core design limits. The GGNS EPU 24-month fuel cycle has been applied conservatively for all evaluations that require core and/or fuel characteristic data to evaluate design basis cap abilities to bound shorter fuel cycle lengths. The GGNS EPU 24-month fuel cycle applied in EPU analyses that are also performed for

cycle-specific basis are appropriate only to the 24-month cycle reference core. The future cycle-specific evaluations utilize approved methods, and conservative results are verified in the applicable SRLR, which utilizes the cycle-specific parameters for the fuel and cycle duration.

1.2.2 Computer

Codes NRC-approved or industry-accepted computer code s and calculational techniques are used to demonstrate compliance with the applicable regul atory acceptance criteria. The application of these codes to the EPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The limitations on use of these codes and methods as defined in the N RC staff position letter reprinted in ELTR1 and the NRC SER for ELTR2 were followed for this EPU an alysis. Any exceptions to the use of the code or conditions of the applicable SERs ar e noted in Table 1-1. The application of the computer codes in Table 1-1 is consistent w ith the current GGNS licensing basis except where

noted in this report.

1.2.3 Approach

The planned approach to achieving the higher power level consists of the change to the GGNS licensing and design basis to increase the licensed power level to 4,408 MWt, consistent with the approach outlined in the CLTR, except as specifically noted in this report, and with the approach NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-5 outlined in ELTR1 for fuel-dependent evaluations. Consistent with the CLTR, the following plant-specific exclusions are exercised: No increase in maximum normal operating reactor dome pressure No increase to maximum licensed core flow No increase to currently licensed MELLLA upper boundary No change to source term methodology No new fuel product line introduction. No change to fuel cycle length No additions to currently licensed operational enhancements The plant-specific evaluations are based on a re view of plant design and operating data, as applicable, to confirm excess design capab ilities; and, if necessary, identify required modifications associated with EPU. All changes to the plant-licensing basis have been identified in this report. For specified topics, generic an alyses and evaluations in the CLTR, or ELTR1 and ELTR2 as applicable, demonstrate plant operability and safety. The dispositions in the CLTR are based on a 20% increase of OLTP, which is gr eater than the requested power uprate of 15% for GGNS. For this increase in power, the conclusions of system/component acceptability stated in the CLTR and ELTR2 are bounding and have been confirmed for GGNS. The scope and depth of the evaluation results provided herein are established based on the approach in the CLTR and ELTR2 and unique features of the plant. The results of the following evaluations are presented in this report: Reactor Core and Fuel Performance: Specific analyses required for EPU have been performed for a representative fuel cycle with the reactor core operating at EPU conditions. Specific core and fuel performance parameters are evaluated and documented each operating cycle, and will continue to be evaluated and documented for the operating cycles that implement EPU. Reactor Coolant System (RCS) and Connected Systems: Evaluations of the NSSS components and systems have been performed at EPU conditions. These evaluations confirm the acceptability of the effects of the higher power and the associated change in process variables (i.e., increased steam and feedwater (FW) flows). Safety-related equipment performance is the primary focus in this report, but key aspects of reactor operational capability are also included. Engineered Safety Feature Systems: The effects of EPU power operation on the Containment, ECCS, Standby Gas Treatment System (SGTS) and other ESFs have been NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-6 evaluated for key events. The evaluations include the containment responses during limiting Anticipated Operational Occurrences (AOOs) a nd special events, ECCS-LOCA, and safety relief valve (SRV) containment dynamic loads. Control and Instrumentation: The control and instrumentation signal ranges and analytical limits (ALs) for setpoints have been evaluated to establish the effects of the changes in various process parameters such as power, neutron flux, steam flow and FW flow. As required, evaluations have been performed to determine the need for any Technical Specification (TS) allowable value (AV) changes for various functions (e.g., main steam line (MSL) high flow isolation setpoints). Electrical Power and Auxiliary Systems: Evaluations have been performed to establish the operational capability of the plant electrical power and distribution systems and auxiliary systems to ensure that they are capable of supporting safe plant operation at the EPU power level. Power Conversion Systems: Evaluations have been performed to establish the operational capability of various non-safety BOP systems and components to ensure that they are

capable of delivering the increased power output, and/or the modifications necessary to obtain full EPU power. Radwaste Systems and Radiation Sources: The liquid and gaseous waste management systems (GWMSs) have been evaluated at limiting conditions for EPU to show that applicable release limits continue to be met during operation at higher power. The radiological consequences have been evaluated for EPU to show that applicable regulations have been met for the EPU power conditions. This evaluation includes the effect of higher power level on source terms, on-site doses and off-site doses, during normal operation. Reactor Safety Performance Evaluations: The limiting UFSAR analyses for design basis events have been addressed as part of the EPU evaluation. All limiting accidents, AOOs, and special events have been analyzed or generically dispositioned consistent with the CLTR, or with ELTR1 and ELTR2, as applicable, and show continued compliance with regulatory requirements. Additional Aspects of EPU: High-energy line break (HELB) and environmental qualification (EQ) evaluations have been performed at bounding conditions for EPU to show the continued operability of plant equipment under EPU conditions. The effects of EPU on the GGNS Probabilistic Risk Assessment (PRA) have been analyzed to demonstrate that there are no new vulnerabilities to severe accidents. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-7 1.3 EPU Plant Operating Conditions

1.3.1 Reactor

Heat Balance The operating pressure, the total core flow, and the coolant thermodynamic state characterize the thermal hydraulic performance of a BWR reactor core. The EPU values of these parameters are used to establish the steady-state operating conditions and as initial and boundary conditions for the required safety analyses. The EPU values for these parameters are determined by performing heat (energy) balance calculations for the reactor system at EPU conditions. The reactor heat balance relates the thermal-hydraulic parameters to the plant steam and FW flow conditions for the selected core thermal power level and operating pressure. Operational parameters from actual plant operation are considered (e.g., steam line pressure drop) when determining the expected EPU conditions. The thermal-hydraulic parameters define the conditions for evaluating the operation of the plant at EPU conditions. The thermal-hydraulic parameters obtained for the EPU conditions also define the steady-state operating conditions for equipment evaluations. Heat balances at appropriately selected conditions define the initial and

boundary conditions for plant safety analyses. Figure 1-2 shows the EPU heat balance at 100% of EPU RTP and 100% rated core flow. Figure 1-3 shows the EPU heat balance at 102% of EPU RTP and 105% core flow with dome pressure at 1060 psia. Table 1-2 provides a summary of the reactor thermal-hydraulic parameters for the current rated and EPU conditions. At EPU conditions, the maximum nominal operating reactor vessel dome pressure is maintained at the current value, which minimizes the need for plant and licensing changes. With the increased steam flow and associated non-safety BOP modifications, the current dome pressure provides sufficient operating turbine inlet pressure to assure good pressure

control characteristics. 1.3.2 Reactor Performance Improvement Features The reactor performance improvement features and the equipment allowed to be out-of-service (OOS) are listed in Table 1-2. When limiting, the input parameters related to the performance improvement features or the equipment OOS have been considered in the safety analyses for EPU, and as applicable, will be included in the reload core analyses. The use of these performance improvement features and allowing for equipment OOS are allowed during EPU operation. The evaluations that are dependent upon cycle length are performed for EPU assuming a 24-month fuel cycle length. 1.4 Summary and Conclusions This evaluation covers an EPU to 115% of OLTP. The strategy for achieving higher power is to extend the MELLLA power/flow map region along the upper boundary extension. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-8 The GGNS licensing bases have been reviewed to demonstrate how this uprate can be accommodated without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any existing regulatory limits or design allowable limits applicable to the plant which might cause a reduction in a margin of safety. The EPU described herein i nvolves no significant hazard consideration. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-9 Table 1-1 Computer Codes Used For EPU Task Computer Code* Version or Revision NRC Approved Comments Nominal Reactor Heat Balance ISCOR 09 Y(2) NEDE-24011P Rev. 0 SER Reactor Core and Fuel Performance TGBLA PANACEA ISCOR 06 11 09 Y Y Y(2) NEDE-30130-P-A (4)

NEDE-30130-P-A (4) NEDE-24011P Rev. 0 SER Thermal Hydraulic Stability ISCOR PANACEA ODYSY TRACG 09 11 05 04 Y(2) Y Y N(15) NEDE-24011P Rev. 0 SER

NEDE-30130-P-A (4)

NEDE-33213P-A

NEDO-32465-A Reactor Pressure Vessel (RPV) Fluence TGBLA DORTG 06 01 Y N (14) (12) (13) RPV Fluid Induced Vibration ANSYS 10.1.3 N (1) Reactor Internal Pressure

Differences (RIPDs) ISCOR LAMB TRACG 09 07 02 Y(2) (3) Y NEDE-24011P Rev. 0 SER NEDE-

20566-P-A NEDE-32176P Rev. 2 NEDC-32177P Rev. 2 NRC TAC No. M90270 Transient Analysis PANACEA ISCOR ODYN SAFER 11 09 09 04 Y Y(2) Y (5) NEDE-30130-P-A (4) NEDE-24011P Rev. 0 SER

NEDO-24154-A NEDC-32424P-A, NEDC-32523P-A

(8) (9) (10) Anticipated Transient Without Scram ODYN STEMP PANACEA ISCOR TASC 09 04 11 09 03A Y (6) Y(4) Y(2) Y NEDE-24154P-A Supp. 1, Vol. 4

NEDE-30130-P-A NEDE-24011P Rev. 0 SER NEDC-32084P-A Rev. 2 (11) Containment System Response SHEX M3CPT LAMB 06 05 08 Y Y (3) (7) NEDO-10320, April 1971 (NUREG-0978) NEDE-20566-P-A, Sept. 1986 Appendix R Fire Protection GESTR SAFER SHEX 08 04 06 (5) (5) Y NEDE-23785-1-PA Rev. 1

(8) (9) (10)

(7) Reactor Recirculation System (RRS) BILBO 04V NA NEDE-23504, February 1977 (1) NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-10 Task Computer Code* Version or Revision NRC Approved Comments ECCS-LOCA LAMB GESTR SAFER ISCOR TASC 08 08 04 09 03A Y Y Y Y(2) Y NEDO-20566A NEDE-23785-1-PA Rev. 1

(8) (9) (10) NEDE-24011P Rev. 0 SER NEDC-32084P-A, Rev. 2 (11) Fission Product Inventory ORIGEN 2.1 N Isotope Generation and Depletion Code Break Flow Mass/Energy

Release Rates TRACG 04 N(16) NEDE-32176P R2, December 1999 NEDE-32177P R2, January 2000 NEDO-33083-A, October 2005 Annulus Pressurization (AP) Loads TRACG 04 N(17) NEDE-33440P, Revision 2, March 2010 AP Loads- Reactor Vessel

and Internal Structural Analysis SAP4G PDA ANSYS 07 02 11 N(1) N(1) N(1) NEDO-10909, Revision 7, December 1979 NEDE-10813A, February 1976 Station Blackout SHEX 06A Y (7)

The application of these codes to the EPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER wher e applicable for each code. The application of the codes also complies with the SERs for the EPU programs.

(1) Not a safety analysis code that requires NRC approval. The code application is reviewed and approved by GEH for "Level-2" application and is part of GEH's standard design process. Also, the application of this code has been used in previous power uprate submittals. (2) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011P 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 RIPDs, Transient, ATWS, Stability, Reactor Core and Fuel Performance and LOCA applications is consistent with the approved models and methods. (3) The LAMB code is approved for use in ECCS-LOCA applications (NEDE-20566-P-A and NEDO-20566A), but no approving SER exists for the use of LAMB in the evaluation of RIPDs or containment system response. The use of LAMB for these applications is consistent with the model description of NEDE-20566-P-A. (4) 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 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-11 PANACEA Version 11 in this a pplication 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.

TGBLA06 with Error Correction 5 was used in the GGNS Core Design analysis and it meets the requirements established by the SER for LTR NEDC-33173P (Reference 7). (5) The ECCS-LOCA codes are not explicitly approved for Transient or Appendix R usage. The staff concluded that SAFER is qualified as a code for best estimate modeling of LOCAs and loss of inventory events via the approval letter and evaluation for NEDE-23785P, Revision 1, Volume II. (Letter, C. O. Thomas (See NRC) to J. F. Quirk (GE), "Review of NEDE-23785-1 (P), "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volumes I and II," August 29, 1983.) In addition, the use of SAFER in the analysis of long-term Loss-of-Feedwater (LOFW) events is specified in the approved LTRs for power uprate: "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32424P-A, February 1999 and "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32523P-A, February 2000. The Appendix R events are similar to the LOFW and small break LOCA events. (6) The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool (SP) 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. (7) The application of the methodology in the SHEX code to the containment response is accepted by the NRC in the letter to G. L. Sozzi (GE) from A. Thadani (NRC), "Use of the SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993 (Reference 8). (8) Letter, J. F. Klapproth (GE) to NRC, Transmittal of GE Proprietary Report NEDC-32950P, "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model," dated January 2000 by letter dated January 27, 2000. (9) Letter, S. A. Richards (NRC) to J. F. Klapprot h (GE), "General Electric Nuclear Energy (GENE) Topical Reports GENE (NEDC)-32950P and GE NE (NEDC)-32084P Acceptability Review," May 24, 2000. (10) "SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-Jet Pump Plants," NEDE-30996P-A, General Electric Company, October 1987. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-12 (11) The NRC approved the TASC-03A code by letter from S. A. Richards (NRC) to J. F. Klapproth (GE Nuclear Energy),

Subject:

"Review of NEDC-32084P, TASC-03A, A Computer Code for Transient Analysis of a Single Fuel Channel," TAC NO. MB0564, March 13, 2002.

(12) CCC-543, "TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport Version 2.8.14," Radiation Shielding Information Center (RSIC), January 1994. (13) Letter, H. N. Berkow (NRC) to G. B. Stramback (GE), "Final Safety Evaluation Regarding Removal of Methodology Limitations for NEDC-32983PA, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations (TAC No. MC3788)," November 17, 2005. (14) Letter, S.A. Richards (NRC) to G. A. Watford (GE), "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. (15) TRACG02 has been approved in NEDO-32465-A by the US NRC for the stability Delta CPR over Initial CPR Versus Oscillation Magnitude (DIVOM) analys is. 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 pr evious version, TRACG02. (16) The TRACG break flow model and qualification basis is described in NEDE-32176P and NEDE-32177P. The application of TRACG04 for the calculation of break flow mass/energy release rates has been approved for ESBWR LOCA application in NEDO-33083-A. (17) The application of TRACG04 for the calculation of AP loads has been described for ESBWR AP application in NEDE-33440P. The application of TRACG04 for the GGNS EPU has been applied in a manner consistent with NEDE-33440P. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-13 Table 1-2 Current and EPU Plant Operating Conditions Parameter Current Plant Value 1 EPU Value 4 Thermal Power (MWt) 3,898 4,408 Vessel Steam Flow (Mlb/hr) 2 16.774 18.968 Full Power Core Flow Range Mlb/hr 86.7 to 118.1 104.4 to 118.1 % Rated 77.1 to 105.0 92.8 to 105.0 Nominal Dome Pressure (psia) 1040 1040 Maximum Nominal Dome Temperature ( F) 549.4 549.4 Pressure at Upstream Side of Turbine Stop Valve (TSV) (psia) 985 970 Full Power FW Flow (Mlb/hr) 16.741 18.935 Temperature ( F) 420.0 420.0 Core Inlet Enthalpy (Btu/lb) 3 527.6 525.1

Notes: 1. Based on current reactor heat balance.

2. At normal FW heating.

3. At 100% core flow conditions.

4. Performance improvement features and/or equipment OOS that are included in EPU evaluations:

a. Maximum Extended Operating Domain (MEOD)

i. MELLLA ii. Increased Core Flow (ICF)

b. Single Loop Operation (SLO)

c. 7 SRVs Out-of-Service (SRVOOS)

d. Automatic Depressurization System Out-of-Service (ADSOOS), 1 Valve

e. Turbine Bypass Valves OOS (TBVOOS)

f. End of Cycle Recirculation Pump Trip (EOC RPT) OOS (RPTOOS)

g. Feedwater Heaters Out-of-Service (FWHOOS), licensed for 50°F, evaluated for 100°F Temperature Reduction

h. MSIV Out-of-Service (MSIVOOS)

i. 24 Month Cycle

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1-14 Figure 1-1 Power/Flow Operating Map for EPU 0 10 20 30 40 50 60 70 80 90100 1101200102030405060708090100110120Core Flow (%)Thermal Power (% of Rated) 0400800120016002000240028003200360040004400480052000153045607590105120135Core Flow (Mlbm/hr)Thermal Power (MWt) Jet Pump and Recirc Pump Cavitation Interlock100% EPU Power = 4408 MWt100% CLTP = 3898 MWt 100% OLTP = 3833 MWt 100% Core Flow = 112.5 Mlbm/hrA: 46.9% Power/ 25.9% Flow B: 87.0% Power/ 75.0% Flow C: 100.0% Power/ 92.8% FlowD: 100.0% Power/100.0% FlowE: 100.0% Power/105.0% Flow F: 87.0% Power/105.0% Flow G: 36.5% Power/105.0% Flow H: 21.7% Power/ 73.5% Flow I: 21.7% Power/ 34.0% FlowMELLLA Boundary 4408 MWt Natural CirculationIncreased Core Flow Region 3898 MWt 3833 MWt A G FEDC B H ICavitation Interlock

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-15 Figure 1-2 EPU Heat Balance - Nominal

(@ 100% Power and 100% Core Flow) 1040 P18.968E+06# *1190.8H *0.49M *Carryunder =0.35%970P *4408MWtWd = 100%19.113E+06#18.935E+06#526.1H397.8H397.6H531.7°F420.1°F420.0°F112.5E+06h = 1.1  H#1.780E+05#412.8H525.1433.9°F H3.300E+04#1.780E+05#90.7H525.0H120.0°F530.8°F*Conditions at upstream side of TSVCore Thermal Power4408.0 Pump Heating11.2 Cleanup Losses-5.9 Other System Losses-1.1Turbine Cycle Use4412.2MWtCleanupDemineralizerSystemMain Steam FlowMain Feed FlowControl Rod Drive Feed FlowTotal Core Flow# = Flow, lbm/hr H = Enthalpy, Btu/lbmF = Temperature, °F M = Moisture, %

P = Pressure, psiaLegend NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

1-16 Figure 1-3 EPU Heat Balance - Overpressure Protection Analysis (@ 102% Power and 105% Core Flow) 1060 P19.428E+06# *1190.0H *0.52M *Carryunder =0.35%987P

4496MWtWd = 105%19.573E+06#19.395E+06#529.6H400.3H400.1H534.5°F422.4°F422.3°F118.1E+06h = 1.1 H

1.780E+05#416.3H528.7437.1°F H3.300E+04#1.780E+05#90.7H528.5H120.0°F533.6°F*Conditions at upstream side of TSVCore Thermal Power4496.0Pump Heating11.3Cleanup Losses-5.9Other System Losses-1.1Turbine Cycle Use4500.3MWtCleanupDemineralizerSystemMain Steam FlowMain Feed FlowControl Rod Drive Feed FlowTotal Core Flow# = Flow, lbm/hrH = Enthalpy, Btu/lbmF = Temperature, °FM = Moisture, %P = Pressure, psiaLegend NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-1 2 SAFETY EVALUATION

2.1 Materials

and Chemical Engineering

2.1.1 Reactor

Vessel Material Surveillance Program Regulatory Evaluation The reactor vessel material surveillance program provides a means for determining and monitoring the fracture toughness of the reactor vessel beltline materials to support analyses for ensuring the structural integrity of the ferritic components of the reactor vessel. The review primarily focused on the effects of the proposed EPU on the reactor vessel surveillance capsule withdrawal schedule. The regulatory acceptan ce criteria are based on: (1) General Design Criterion (GDC)-14, insofar as it requires the r eactor coolant pressure boundary (RCPB) be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; (2) GDC-31, insofar as it requires the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a nonbrittle manner and the probability of a rapidly propagating fracture is minimized; (3) 10 CFR 50, Appendix H, which provides for monitoring changes in the fracture toughness properties of materials in the reactor

vessel beltline region; and (4) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR 50, Appendix H. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The Reactor Vessel Material Surveillance Program is described in UFSAR Sections 5.3.1, "Reactor Vessel Materials," and 5.3.3.7, "In-service Surveillance."

Technical Evaluation The RPV fracture toughness evaluation proce ss is described in Section 2.1.2. RPV embrittlement is caused by neutron exposure of the wall adjacent to the core including the regions above and below the core that experi ence fluence greater than or equal to 1 x 10 17 n/cm 2. This region is defined as the "beltline" region. Operation at EPU conditions results in a higher neutron flux, which increases the integrated flue nce over the period of plant life. For GGNS, a more conservative fluence was used for CLTP; therefore, the EPU fluence is slightly less than that considered for CLTP. The GGNS surveillance program consists of th ree capsules. One capsule containing Charpy specimens was removed from the vessel during Refu eling Outage 7 (RFO 7), but it was returned NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-2 to the vessel during Refueling Outage 8 (RFO 8) with all specimens intact. The remaining two capsules have been in the reactor vessel since plant startup. GGNS is a participant in the BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP), currently administrated by Electric Power Research Ins titute (EPRI), and is not designated as a representative plant; therefore, no capsules are slated for removal at this time. EPU has no effect on the existing surveillance schedule. The normal operating dome pressure for EPU is unchanged from that for original thermal power operation. Therefore, the hydrostatic and leakage test pressures are acceptable for EPU. Operation with EPU does not have an advers e effect on the reactor vessel fracture toughness because the vessel remains in compliance with the regulatory requirements as demonstrated in Section 2.1.2.

Conclusion The effects of the proposed EPU on the reactor vessel surveillance withdrawal schedule have been reviewed. Entergy concludes that the changes in neutron fluence and their effects on the schedule have been adequately addressed. Entergy further concludes that continued participation in the BWRVIP Integrated Surveillance Program is appropriate to ensure the material surveillance program will continue to meet the requirements of 10 CFR 50, Appendix H, and 10 CFR 50.60, and will provide Entergy with information to ensure continued compliance with GDC-14 and GDC-31 in this respect following implementation of the proposed EPU. Therefore, Entergy has determined that the proposed EPU is acceptable with respect to the reactor vessel material surveillance program. 2.1.2 Pressure-Temperature Limits and Upper Shelf Energy Regulatory Evaluation Pressure-temperature (P-T) limits are established to ensure the structural integrity of the ferritic components of the RCPB during any condition of normal operation, including AOOs and hydrostatic tests. The review of P-T limits covered the P-T limits methodology and the calculations for the number of e ffective full power years (EFPY) specified for the proposed EPU, considering neutron embrittlement effects and using linear elastic fracture mechanics. The regulatory acceptance criteria for P-T limits are ba sed on: (1) GDC-14, insofar as it requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; (2) GDC-31, inso far as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a nonbrittle manner and the probability of a rapidly propagating fracture is minimized; (3) 10 CFR 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB; and (4) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR 50, Appendix G. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-3 GGNS Current Licensing Basis NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The P-T limits are described in UFSAR Section 5.3.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 3.2.1 of the CLTR addresses the effect of CPPU on P-T Limits and Upper Shelf Energy (USE). The results of this evaluation are described below. As explicitly stated in Section 3.2.1 of the CLTR, EPU may result in a higher operating neutron flux at the vessel wall, consequently increasing the integrated flux over time (neutron fluence). The neutron fluence is recalculated using the NRC-approved GEH neutron fluence methodology (Reference 9). This method is consistent with RG 1.190 (Reference 10). GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Fracture Toughness [[

     ]] Meets CLTR Disposition The revised fluence is used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G. The results of these evaluations indicate that: (a) The USE will remain greater than 50 ft-lb for the design life of the vessel and maintain the margin requirements of 10 CFR 50, Appendix G. The minimum USE for the beltline materials is 79.6 ft-lb for 35 EFPY. These values are provided in Table 2.1-1. (b) The beltline material reference temperature of the nil-ductility transition (RT NDT) remains below 200°F. (c) The CLTP P-T curves require revision for EPU to include the effects of the N12 Water Level Instrumentation Nozzle that occurs within the beltline region. The current adjusted reference temperature (ART) values for the beltline plates and welds remain bounding for EPU due to the conservative fluence previously considered. (d) The 35 EFPY shift is decreased, and consequen tly, results in a change in the ART, which is the initial RT NDT plus the shift. These values are provided in Table 2.1-2.

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2-4 (e) The 35 EFPY beltline circumferential weld material RT NDT remains bounded by the requirements of GL 98-05 (Reference 11) and BW RVIP-05 (Reference 12) as defined in the SE to BWRVIP-74 (Reference 13). This comparison is provided in Table 2.1-3. Therefore, GGNS meets all CLTR dispositions for fracture toughness. Conclusion The effects of the proposed EPU on the P-T limits for the plant have been reviewed. Entergy concludes that the changes in neutron fluence and their effects on the P-T limits necessitate the introduction of revised P-T limits to accommodate the N12 water level nozzle. Entergy further concludes it has demonstrated the validity of the proposed P-T limits for operation under the

proposed EPU conditions. Based on this, Entergy concludes the proposed P-T limits will continue to meet the requirements of 10 CF R 50, Appendix G, and 10 CFR 50.60 and will enable GGNS to continue to comply with GDC-14 and GDC-31 following implementation of the proposed EPU. Therefore, Entergy finds th e proposed EPU acceptable with respect to the proposed P-T limits.

2.1.3 Reactor

Internal and Core Support Materials Regulatory Evaluation The reactor internals and core supports include structures, systems, and components (SSCs) that perform safety functions or whose failure could affect safety functions performed by other SSCs. These safety functions include reactivity mon itoring and control, core cooling, and fission product confinement (within both the fuel cla dding and the RCS). The review covered the materials' specifications and mechanical prope rties, welds, weld controls, non-destructive examination (NDE) procedures, corrosion resist ance, and susceptibility to degradation. The regulatory acceptance criteria for reactor internal and core support materials are based on GDC-1 and 10 CFR 50.55a for material specifications, c ontrols on welding, and inspection of reactor internals and core supports. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The reactor internals are described in UFSAR Sections 3.9.5, "Reactor Pressure Vessel Internals," 4.1.2, "Reactor Internal Components," and 4.5.2, "Reactor Internal Materials." NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-5 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 10.7 of the CLTR addr esses the effect of C PPU on Reactor Internal and Core Support Materials. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Irradiated Assisted Stress Corrosion Cracking [[

     ]] Meets CLTR Disposition As explicitly stated in Section 10.7 of the CLTR, the increase in irradiation of the core internal components influences irradiation-assisted stre ss corrosion cracking (IASCC). The longevity of most equipment is not affected by EPU.  [[

     ]] is required for EPU. The reactor internal and core support materials evaluation included the materials' specifications and mechanical properties, welds, weld controls, NDE procedures, corrosion resistance, and susceptibility to degradation. This evaluation of the reactor internals and core supports includes SSCs that perform safety functions or whose failure could affect safety functions performed by other SSCs. None of these requirements, specifica tions, or controls is changed as a result of the EPU; therefore, these continue to be acceptable. GGNS has a procedurally controlled program for the augmented NDE of selected RPV internal components in order to ensure their continued structural integrity. The inspection techniques utilized are primarily for the detection and characterization of service-induced, surface-connected planar discontinuities, such as in tergranular stress corrosion cracking (IGSCC) and IASCC, in welds and in the adjacent base mate rial. GGNS belongs to the BWRVIP organization and implementation of the procedurally controlled program is consistent with the BWRVIP issued documents. The inspection strategies recommended by the BWRVIP consider the effects of fluence on applicable components and are based on component configuration and field experience.

Components selected for inspection include those th at are identified as susceptible to in-service degradation and augmented examination is conducted for verification of structural integrity. These components have been identified through the review of NRC Inspection and Enforcement Bulletins (IEBs), BWRVIP documents, and recommendations provided by GE Service NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-6 Information Letters (GE SILs). The inspection program provides performance frequency for NDE and associated acceptance criteria. Components inspected include the following:

1. Core spray (CS) piping

2. Core plate

3. Core spray spargers

4. Core shroud and core shroud support

5. Jet pumps and associated components

6. Top guide

7. Lower plenum

8. Vessel ID attachment welds

9. Instrumentation penetrations

10. Steam dryer drain channel welds

11. FW spargers

12. In-core flux monitoring guide tubes

13. Control rod guide tubes Inspected components are considered as being poten tially susceptible to IASCC if the end-of-life fluence is in excess of 5 x 10 20 n/cm 2. Three components have been identified as being potentially susceptible to IASCC, utilizing a conservative 54 EFPY fluence: (1) Top Guide, 5.26 x 10 21 n/cm 2; (2) Shroud, 2.88 x 10 21 n/cm 2; and (3) Core Plate, 6.57 x 10 20 n/cm 2. The BWRVIP inspection recommendations which provide the scope, sample size, inspection method, and frequency of examination used to manage the effects of IASCC are as follows: Top Guide (BWRVIP-26-A) (Reference 14) Shroud (BWRVIP-76) (Reference 15) Core Plate (BWRVIP-25) (Reference 16) Continued implementation of the current procedure program assures the prompt identification of any degradation of reactor vessel internal components experienced during EPU operating conditions. To mitigate the potential for IG SCC and IASCC, GGNS utilizes hydrogen water NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-7 chemistry (HWC). Reactor vessel water chemistry conditions are also maintained consistent with the EPRI and established industry guidelines. The service life of most equipment is not a ffected by EPU. The peak fluence increase experienced by the reactor internals does not represent a significant increase in the potential for IASCC. The current inspection strategy for the reactor internal components is expected to be adequate to manage any potential effects of EPU. Therefore, GGNS meets all CLTR dispositions for IASCC.

Conclusion The effects of the proposed EPU on the suscep tibility of reactor internal and core support materials to known degradation mechanisms ha ve been reviewed. Entergy concludes the appropriate degradation management programs have been identified to address the effects of changes in operating temperature and neutron fluen ce on the integrity of reactor internal and core support materials. Entergy further concludes the reactor internal and core support materials will continue to be acceptable and will continue to meet the requirements of GDC-1 and 10 CFR 50.55a with respect to material specifications, welding controls, and inspection following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to reactor internal and core support materials.

2.1.4 Reactor

Coolant Pressure Boundary Materials Regulatory Evaluation The RCPB defines the boundary of systems and components containing the high pressure fluids produced in the reactor. The review of RCPB materials covered their specifications, compatibility with the reactor coolant, su sceptibility to degradation, and degradation management programs. The regulatory acceptance criteria for RCPB materials are based on: (1) 10 CFR 50.55a and GDC-1, insofar as they require SSCs important to safety be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (3) GDC-14, insofar as it requires the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; (4) GDC-31, insofar as it requires the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a nonbrittle manner and the probability of a rapidly propagating fracture is minimized; and (5) 10 CF R 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-8 GGNS Current Licensing Basis NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The RCPB is described in UFSAR Sections 3.9.3, "ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures ," 3.9.5, "Reactor Pressure Vessel Internals," 5.2, "Integrity of Reactor Coolant Pressu re Boundary," 5.3, "Reactor Vessel," and 5.4.1, "Reactor Recirculation System." Technical Evaluation The temperature and flow increase experien ced by the RCPB does not represent significant increase in the potential for IGSCC. Other degradation mechanisms are addressed in other sections of this report. Fracture Toughness of the vessel components is addressed in Section 2.1.2. Flow Accelerated Corrosion (FAC) for the pl ant is addressed in Section 2.1.6. Material Fatigue usage for the RCPB piping is addresse d in Section 2.2.2.2.1. Flow induced vibration (FIV) for the safety-related piping thermowells and probes is addressed in Section 2.2.2.1.3. Based on the conclusions provided in other sections of this report as referenced above, the current inspection strategy for the RCPB is adequate to manage any potential effects of EPU. The GGNS in-service inspection (ISI) program for RCPB piping is in accordance with American Society of Mechanical Engineers (ASME) Section XI coupled with the augmented program for reactor coolant piping based on Generic Letter (GL) 88-01 (Reference 17), NUREG-0313 (Reference 18) and BWRVIP-75-A (Reference 19). The inspection techniques utilized are in full conformance with ASME Section XI, Appendix VIII, Supplement 10 for the detection and characterization of service-induced, surface-connect ed planar discontinuities, such as IGSCC. Continued implementation of the current program assures the prompt identification of any degradation of RCPB components experien ced during EPU operating conditions. The augmented inspection program is designed to detect potential degradation from IGSCC. For IGSCC to occur, three conditions must be present: (1) a susceptible material (for a list of materials in the RCPB, see Section 5.2.2.7 of th e UFSAR); (2) the presence of residual or applied tensile stress (such as from welding); and (3) aggressive environment. Operation at EPU conditions results in an insignificant change to temperature and flow conditions for portions of the RCPB piping and does not affect the other susceptibility factors associated with IGSCC. This is consistent with the conclusions pres ented in Section 3.6.1 of ELTR2. EPU does increase the fluence for the reactor vessel beltline region; however the EPU fluence for the reactor vessel nozzle safe-end welds and piping remains well below the 5.0 x 10 20 n/cm 2 fluence threshold for IASCC concerns for stainless steel. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-9 The GGNS augmented inspection program frequency is based on BWRVIP-75-A normal water chemistry. While GGNS has implemented HWC, the augmented program includes more frequent inspections than those required by BWRVIP-75-A at this time for HWC. GGNS does not have any Category D, E, F, or G weldments. Several IGSCC mitigation processes have been applied to GGNS to reduce the RCPB components' susceptibility to IGSCC. GGNS wa s designed, fabricated, and constructed with IGSCC addressed in most welds by one of three methods: (1) corrosion resistant materials; (2) solution heat treatment; or (3) clad with resistant materials. For the weldments where these three processes were not used, stress improveme nt processes were applied to reduce IGSCC susceptibility. Stress improvement processes and original construction processes used for IGSCC resistance are not affected by EPU. Also, GGNS has implemented HWC, which reduces the potential for IGSCC of RCPB components.

Conclusion The effects of the proposed EPU on the susceptibility of RCPB materials to known degradation mechanisms have been reviewed. Entergy concludes the appropriate degradation management programs have been identified to address the effects of changes in system operating temperature on the integrity of RCPB materials. Entergy further concludes the RCPB materials will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of GDC-1, GDC-4, GDC-14, GDC-31, 10 CFR 50, Appendix G, and 10 CFR 50.55a. Therefore, Entergy finds the proposed EPU acceptable with respect to RCPB materials.

2.1.5 Protective

Coating Systems (Paints) - Organic Materials Regulatory Evaluation Protective coating systems (paints) provide a means for protecting the surfaces of facilities and equipment from corrosion and contamination from radionuclides and also provide wear protection during plant operation and maintenan ce activities. The review covered protective coating systems used inside the containment for their suitability for and stability under design-basis LOCA (DBLOCA) conditions considering radiation and chemical effe cts. The regulatory acceptance criteria for protective coating systems are based on: (1) 10 CFR 50, Appendix B, which states quality assurance requirements fo r the design, fabrication, and construction of safety-related SSCs and (2) RG 1.54, Revision 1, (R eference 20) for guidance on application and performance monitoring of coatings in nuclear power plants. GGNS Current Licensing Basis

The use of protective coatings is described in UFSAR Section 6.1.2.1, "Protective Coatings." NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-10 Technical Evaluation The use of protective coatings is described in UFSAR Section 6.1.2.1, "Protective Coatings." The Service Level 1 protective coatings used at GGNS were qualified per American National Standards Institute (ANSI) N101.2-1972 to a radiation level of 1.4 E+09 Rads and a temperature of 340°F. As reported in UFSAR Section 6.1.2.1, a small quantity of unqualified coatings has been determined to be acceptable. Bounding conditions for a DBA under EPU conditions are shown below. EQ Zone Peak Temperature (ºF) Peak Pressure (psig) Irradiation (Rads) Drywell Qualification Level 340 70 1.4 E+09 EPU Conditions <340 1 26.7 <1.4 E+09 Containment Qualification Level 280 70 1.00 E+09 EPU Conditions 154 11.9 3.33 E+08 Note: 1. Ignoring a short initial one-second transient to 347º F that would have an in significant effect on the coating temperatures. For GGNS, monitoring and maintaining protective coatings inside the primary containment are performed to ANSI N101.2-1972, ANSI N101.4-1972, ANSI N5.12-1974, RG 1.54, June 1973 (Reference 20), and ANSI/ASME NQA-1-1983. Coating condition assessments of Service Level 1 coatings inside the primary containment (drywell) are also conducted every RFO. The GGNS Protective Coating Monitoring and Maintenance Program is an existing program that is described in the GGNS response to GL 98-04, "Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coati ng Deficiencies and Foreign Material in Containment," (Reference 21). The program was developed in accordance with ANSI N101.4-1972 referenced in RG 1.54, June 1973, along with ANSI/ASME NQA-1-1983. The GGNS program is a "comparable program" as described in NUREG-1801, Chapter XI, Program XI.S8, Protective Coating Monitoring and Maintenance Program (Reference 22). The program applies to Service Level 1 protective coatings inside the primary containment. The GGNS SP (wetwell (WW)) is not included because it is stainless steel lined reinforced concrete and does not have Service Level 1 coatings. Coating conditions monitored by this program NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-11 include blistering, cracking, peeling, loose rust, and physical/mechanical damage. When localized degradation of a coating is identified, the affected area is evaluated and scheduled for repair, replacement, or removal, as needed. The condition assessments and resulting repair, replacement, or removal activities ensure that the amount of coatings subject to detachment from the substrate during a LOCA is minimized to ensure post-accident operability of the ECCS suction strainers. Conclusion The effects of the proposed EPU on protective coating systems have been reviewed. Entergy concludes the effect of changes in conditions following a DBLOCA and their effects on the protective coatings have been appropriately addressed. Entergy further concludes the protective coatings will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of 10 CFR 50 Appendix B. Therefore, Entergy finds the proposed EPU acceptable with respect to protective coatings systems. 2.1.6 Flow Accelerated Corrosion Regulatory Evaluation FAC is a corrosion mechanism occurring in carbon steel components exposed to flowing single- or two-phase water. Components made from stainless steel are immune to FAC, while FAC is significantly reduced in components containing small amounts of chromium or molybdenum. The rates of material loss due to FAC depend on velocity of flow, fluid temperature, steam quality, oxygen content, and pH. During plant operation, control of these parameters is limited and the optimum conditions for minimizing FAC effects, in most cases, cannot be achieved. Loss of material by FAC will, therefore, occur. Entergy has reviewed the effects of the proposed EPU on FAC and the adequacy of GGNS' FAC program to predict the rate of loss so that repair or replacement of damaged components can be made before they reach critical thickness. The GGNS FAC program is based on NUREG-1344 (Reference 23), GL 89-08 (Reference 24), and the guidelines in EPRI Report NSAC-202L-R3. It consists of predicting loss of material using the CHECWORKSŽ computer code, and performing visual inspections and volumetric examinations of the affected components. The regulatory acceptance criteria are based on the structural evaluation of the minimum acceptable wall thickness for the components undergoing degradation by FAC. GGNS Current Licensing Basis The FAC Program is not described in any GGNS licensing basis document; however, it is governed in accordance with Administrative Procedure 01-S-06-63, "Piping Integrity Program Review." NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-12 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 10.7 of the CLTR addresses the effect of CPPU on FAC. GGNS meets all CLTR dispositions. The results of this evaluation are described below. Topic CLTR Disposition GGNS Result Flow Accelerated Corrosion [[

     ]] Meets CLTR Disposition The CLTR states that the increase in steam and FW flow rates as a result of EPU influence FAC. In order to monitor and control FAC, GGNS maintains an effective FAC program. The EPU implementation at GGNS will change a number of water and steam system flow rates, temperatures, and enthalpies. All these factors affect FAC susceptibility status and FAC wear rates. As a result of EPU operating conditions, some lines will experience accelerated rates of

FAC, while others will have reduced rates. It s hould be noted that no lines that were previously non-susceptible to FAC became susceptible due to EPU operating conditions. [[

   ]]  The FAC program will not significantly change for EPU. The FAC program at GGNS is based on:  NRC I&E Bulletin 87-01, "Thinning Pipe Walls in Nuclear Power Plants"  GL 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning" (Reference 24)  EPRI NSAC-202L-R3, "Recommendations for an Effective Flow Accelerated Corrosion Program," Revision 3, May 2006 With regard to EPRI NSAC-202L, the choice of the method for detecting and evaluating the effect of FAC on a component is dependent on the type of component and its history. Results of the evaluation reveal if the component will remain above minimum allowable wall thickness throughout the next operating cycle and what the predicted minimum wall thickness will be at the end of the operating cycle. Additionally, th e evaluation shows the remaining service life of the component (based on the calculated minimum allowable wall thickness) and the next scheduled inspection (NSI) outage. The NSI outage is the outage prior to the time that the component reaches minimum allowable wall thickness. Piping is evaluated against the minimum NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-13 acceptable wall thickness according to the GGNS design methodology and is found acceptable only if it meets all the design requirements of GGNS. The GGNS FAC program monitors all FAC susceptible piping - both small bore and large bore - to ensure the structural integrity and functionality are maintained. FAC susceptible piping can

be divided into two categories: lines that meet the requirements to be modeled using EPRI CHECWORKSŽ SFA, and those that do not. For those that meet the requirements, GGNS uses CHECWORKSŽ SFA, in conjunction with actual measurements, to predict FAC wear rates and remaining service life for components in single phase and two phase systems. The FAC susceptible lines that do not meet the minimum requirements for modeling and analysis by CHECWORKSŽ SFA are referred to as "Susceptible - Not Modeled" (S-NM). This group is comprised of lines with unknown operating conditions that prevent the development of accurate predictive models. Small bore piping is also included in this group. Selection of this piping for inspection is t ypically the result of industry experience, GGNS experience, or engineering judgment. One of the most import aspects of the GGNS FAC program is the proper selection of locations for FAC inspection and replacement of degraded piping. This is accomplished using the following (detailed in Table 2.1-8): CHECWORKSŽ SFA predictive wear analysis Susceptibility ranking of S-NM piping OE GGNS-specific experience Trending of historical inspection data Sound engineering judgment combining all of the above The proposed EPU may affect the following aspects of the GGNS FAC program: FAC System Susceptibility Evaluation - no new lines are added to the GGNS FAC program. Wear rates - changes in operating conditions will result in some components wearing at an accelerated rate, while others will wear at a slower rate. Selection of component inspection and replacement locations and subsequent evaluation of inspection results (trending) - there could be a short-term increase in the number of inspections performed. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-14 These are evaluated as follows: FAC System Susceptibility Evaluation GGNS performed a system susceptibility sc reening based on the revised EPU heat balance, and determined that no additional lin es were required to be added to the FAC program. Wear Rates - CHECWORKS Ž SFA Model Update for EPU The proposed EPU will result in changes to several variables that may directly influence FAC wear rates. The variables include operating temperature, steam quality, and velocity. The oxygen content is not expected to appreciably change. To account for these changes, GGNS updated the affected parameters in the CHECWORKSŽ SFA model based on the EPU heat balance diagram. Table 2.1-4 contains a compilation of lines with similar operating conditions, water chemistry and usage for analysis. A comparis on of CLTP and EPU wear rate predictions identified changes ranging from a median decrease of 25% to a median increase of 100%. Of the 58 listed lines, 22 had a median increas e in the predicted wear rate while the remaining 36 exhibited a decrease or no change in median wear rate. Based on a review

of the changes in operating conditions, GGNS found the resulting predicted wear rates to be consistent with EPU conditions. Selection of Inspection and Replacement Locations

The current approach to select locations fo r FAC inspection does not change as a result of the EPU. However, there could be an increase in the number of FAC inspections performed on both CHECWORKSŽ SFA-mode led and S-NM piping over the next several RFOs to ensure the effect of pow er uprate is understood. Inspections will be selected considering the changes in predicted wear rates, actual component thicknesses, operating time since last examination, and design margin. This data will be used to further calibrate the CHECWORKSŽ SFA model and susceptibility rankings for S-NM piping. This approach will ensure that FAC susceptible components are inspected or replaced prior to reaching code minimum wa ll thickness. Based on the EPU evaluation, no significant effect on the component replacem ent schedule is anticipated in the near term. The continued implementation of the existing GGNS FAC program, updated to include EPU system parameters, will ensure that any required changes to the component inspection and replacement schedules are made prior to EPU implementation. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-15 Benchmarking CHECWORKS Ž SFA Predicted Component Thickness Table 2.1-5 presents a comparison of CHECWORKSŽ-predicted thicknesses to measured thicknesses for a selection of components from the 29 run definitions. The selection process includes components with the highest predicted wear rates prior to EPU. The measured thicknesses were determined between RFO 14 and RFO 16, the three most recent RFOs. The table shows that, in 115 of 117 cases, the measured thickness from inspection was greater than the predicted thickness, indicating that CHECWORKSŽ SFA predictions were conservative. For the two exceptions, the measured thickness exceeded the minimum allowed thickness. The GGNS FAC program does not address liquid droplet impingement (LDI). However, when LDI is discovered, the FAC program is frequently involved in its evaluation. The FAC program also inspects for cavitation per system engineering requests. The GGNS FAC program adequately manages the effects on FAC due to EPU. Therefore, GGNS meets all CLTR dispositions for FAC.

Conclusion Entergy has reviewed the effect of the proposed EPU on the FAC analysis for the plant and concludes the changes in the plant operating conditions on the FAC analysis have been adequately addressed. Entergy further conclude s that the updated analyses will predict the loss of material by FAC and will ensure timely repair or replacement of degraded components following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to FAC.

2.1.7 Reactor

Water Cleanup System Regulatory Evaluation The reactor water cleanup (RWCU) system provides a means for maintaining reactor water quality by filtration and ion exchange and a path for removing reactor coolant when necessary. Portions of the RWCU system comprise the RCPB. The review of the RWCU system included component design parameters for flow, temperature, pressure, heat removal capability, and impurity removal capability; and the instrumentation and process controls for proper system operation and isolation. The review consisted of evaluating the adequacy of the plant's TSs in these areas under the proposed EPU conditions. The regulatory acceptance criteria for the RWCU system are based on: (1) GDC-14, in sofar as it requires the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating

fracture; (2) GDC-60, insofar as it requires the plant design include means to control the release NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-16 of radioactive effluents; and (3) GDC-61, insofar as it requires systems that contain radioactivity be designed with appropriate confinement. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The RWCU system is described in UFSAR Section 5.4.8, "Reactor Water Cleanup System." Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 3.11 of the CLTR addresses the effect of CPPU on the RWCU system. The results of this evaluation are described below. The RWCU system is a normally operating system with no safety-related functions other than containment isolation. This system is designed to remove solid and dissolved impurities from recirculated reactor coolant, thereby reducing the concentration of radioactive and corrosive species in the reactor coolant. The evaluation of the system performance of the GGNS RWCU system under EPU conditions is presented below. The effects of EPU on the RWCU containment isolation function and valves are included in the containment isolation assessment in Sections 2.2.4 and 2.6.1.3. Tables 2.1-6 and 2.1-7 contain the magnitude of changes in RWCU system operating conditions and a summary of the chemistry values. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result System Performance [[ Meets CLTR Disposition Containment Isolation

     ]] Addressed in Section 2.6.1.3 As explicitly stated in Section 3.11 of the CLTR, the RWCU system may be slightly affected by the increase in FW flow due to the power uprate.

RWCU system operation at the EPU RTP level slightly decreases the temperature within the RWCU system (from 532.7°F to 530.8°F). This system currently operates at flow rates slightly NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-17 above the design flow; the design and nominal ope rating flow rates are not being changed by the EPU. Table 2.1-6 provides the magnitude of changes in RWCU system operating conditions (e.g., a decrease in operating inlet temperature). RWCU system flow is usually selected to be approximately 1% of FW system flow based on operational history. For the GGNS EPU, the RWCU system flow is not changing; the FW flow is being increased as discussed in Section 1. The existing RWCU system flow (and that analyzed for EPU) is within the BWR operational history and has additional margin. Furthermore, the evaluation of RWCU performance for the GGNS EPU considered water chemistry, heat exchanger (HX) performance, pump performance, flow control valve capability and filter / demineralizer performance. All aspects of performance were found to be within the design of the RWCU system at the analyzed flow at EPU conditions. The RWCU system analysis concludes that:

1. There is a negligible heat load effect.

2. A small increase in filter / demineralizer backwash frequency occurs, but this is within the capacity of the radwaste system.

3. The changes in operating system conditions result from a decrease in inlet temperature and increase in FW system operating pressure.

4. The RWCU system filter / demineralizer control valves operate in a more open position to compensate for the increased FW system pressure.

5. No changes to instrumentation are required, a nd setpoint changes are not required due to the negligible system process parameter changes. Previous OE has shown that the increased FW fl ow results in increases in four key reactor coolant chemistry parameters. Table 2.1-7 provides a summary of the chemistry values and the

evaluation results for each are presented below: Iron concentration - The calculated reactor water iron concentration increases from 42.74 ppb to 48.29 ppb for the as-operated case. However, this change is considered insignificant, and does not affect RWCU system performance. Note that these values are based on current levels of FW iron concentration (approximately 4.8 ppb). GGNS is installing a CFFF to reduce the FW iron concen tration below 1 ppb, which is expected to significantly lower reactor water iron concentration. Sulfates concentration - The current average level of sulfates is 1.393 ppb. The expected reactor water sulfate level for EPU, considering the FW flow increase, is 1.573 ppb for the nominal as-operated case. This level is well below the administrative limit of 5.0 ppb for sulfates. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-18 Chlorides concentration - The current average level of chlorides is 0.108 ppb. The expected reactor water chloride level for EPU, considering the FW flow increase, is 0.122 ppb for the nominal as-operated case. This level is well below the administrative limit of 5.0 ppb for chlorides. Conductivity - The calculated reactor water conductivity increases from 0.077 S/cm to 0.080 S/cm because of the increase in FW flow. This expected level is well below the administrative limit for conductivity of 0.30 S/cm. The current reactor water conductivity limits are unchanged for EPU and the actual conductivity remains within these limits. The effects of EPU on the RWCU system func tional capability have been reviewed, and the system can perform adequately at EPU RTP with the original RWCU system flow. As can be seen from Table 2.1-6, the changes in RWCU system operating conditions are small. The GGNS RWCU system has sufficient capacity to respond to the EPU conditions and maintain the chemistry parameters within administrative limits. GGNS intends to install a condensate full-flow iron filtration system, but, conservatively, the benefit of that modification has not been

credited in the evaluation of the effects of EPU on the RWCU. Therefore, GGNS meets all CLTR dispositions for system performance. Conclusion Entergy has reviewed the effects of the proposed EPU on the RWCU system and concludes the changes in impurity levels and pressure and their effects on the RWCU system have been adequately addressed. Entergy further concludes that the RWCU system will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of GDC-14, GDC-60, and GDC-61. Therefore, Entergy finds the proposed EPU acceptable with respect to the RWCU system. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-19 Table 2.1-1 Upper Shelf Energy - 40 Year License (35 EFPY) Material Heat Number Initial Transverse USE (ft-lb) %Cu 35 EFPY 1/4T Fluence (n/cm 2) % Decrease USE 35 EFPY Transverse USE 1 (ft-lb) Plates: C2593-2 100 0.04 1.72E+18 7.5 92.5 C2594-1 94 0.04 1.72E+18 7.5 87.0 C2594-2 96 0.04 1.72E+18 7.5 88.8 Shell Plate 2 A1224-1 117 0.04 1.72E+18 7.5 108.2 Axial Welds 2: 5P6214B/0331 Single 88 0.02 1.72E+18 9.5 79.6 Shell Ring 2 5P6214B/0331 Tandem 88 0.02 1.72E+18 9.5 79.6 Circumferential Welds: AB 3 4P7216/0156 Single 90.7 0.03 2.30E+17 7.0 84.4 AB 3 4P7216/0156 Tandem 90.7 0.03 2.30E+17 7.0 84.4 Nozzles: N12 4 C2593-2 100 0.04 1.91E+17 4.5 95.5 N12 4 C2594-2 96 0.04 1.91E+17 4.5 91.7 Best Estimate Chemistries from BWRVIP-135 R1 (Reference 25): Plate A1224-1 147.3 0.035 1.72E+18 7.0 137.0 Weld 5P6214B/0331 Single 90.9 0.019 1.72E+18 9.0 82.7 Weld 5P6214B/0331 Tandem 91.5 0.019 1.72E+18 9.0 83.3 Integrated Surveillance Program (BWRVIP-135 R1): Plate A1224-1 5, 7 147.3 0.03 1.72E+18 8.6 134.6 Plate A1224-1 6, 7 117 0.03 1.72E+18 8.6 106.9 Weld 5P6214B/0331 Single 5 90.9 0.02 1.72E+18 9.5 82.3 Weld 5P6214B/0331 Tandem 5 91.5 0.02 1.72E+18 9.5 82.8 Weld 5P6214B/0331 Single 6 88 0.02 1.72E+18 9.5 79.6 Weld 5P6214B/0331 Tandem 6 88 0.02 1.72E+18 9.5 79.6 Notes: 1. USE = Initial Transverse US E * [1 - (% decrease / 100)] 2. Use of Shielded Metal Arc Weld (SMAW) Heats 422K8511, 627069, 626677, and 627260 was determined to be limited to weld pick-ups at the ID/OD surfaces or initial root pass or sealing at the backing ba rs which were ground out or subsequently removed. Certified Material Test Reports indicate that no SMAW weld material is present at either the 1/4T or 3/4T location. Therefore, these heats are not required to be evaluated as part of the beltline region. 3. Weld AB occurs within the extended beltline region, defined as experiencing a fluence >1.0e17 n/cm2.

4. The N12 Water Level Instrumentation Nozzle occurs in the beltline region. Because the forging is fabricated from stainless steel, the ART is calculated using the plat e heats where the nozzles occur. For GGNS, these nozzles occur in only two (2) of the Shell 2 plates. 5. The material is evaluated using the ISP unirradiated USE to illustrate the difference and that the material is acceptable. 6. The material is evaluated using the GGNS unirradiated USE to illustrate the difference and that the material is acceptable. 7. % Decrease in USE has been adjusted because the measured decr ease exceeds the predicted RG 1.99 (Reference 26) decrease.

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2-20 Table 2.1-2 Adjusted Reference Temperatures Year License (35 EFPY) Lower-Intermediate She ll and Axial Welds Thickness in inches = 6.4375 35 EFPY Peak I.D. fluence = 2.53E+18 n/cm 2 35 EFPY Peak 1/4 T fluence = 1.72E+18 n/cm 2 N12 Nozzles Thickness in inches = 6.4375 35 EFPY Peak I.D. fluence = 2.81E+17 n/cm 2 35 EFPY Peak 1/4 T fluence = 1.91E+17 n/cm 2 Weld AB Thickness in inches = 6.4375 35 EFPY Peak I.D. fluence = 3.38E+17 n/cm 2 35 EFPY Peak 1/4 T fluence = 2.30E+17 n/cm 2 Component Heat %Cu %Ni CF Adjusted CF Initial RT NDT °F 1/4 T Fluence n/cm 2 35 EFPY RT NDT °F I Margin °F 35 EFPY Shift °F 35 EFPY ART °F Plant-Specific Chemistries: Plates: C2593-2 0.04 0.59 26 N/A -30 1.72E+18 13.9 0 6.9 13.9 27.8 -2.2 C2594-1 0.04 0.63 26 N/A -10 1.72E+18 13.9 0 6.9 13.9 27.8 17.8 C2594-2 0.04 0.63 26 N/A 0 1.72E+18 13.9 0 6.9 13.9 27.8 27.8 Shell Ring 2 A1224-1 0.04 0.65 26 N/A 0 1.72E+18 13.9 0 6.9 13.9 27.8 27.8 Axial Welds 1: 5P6214B/0331 Single 0.02 0.82 27 N/A -50 1.72E+18 14.4 0 7.2 14.4 28.8 -21.2 Shell Ring 2 5P6214B/0331 Tandem 0.02 0.82 27 N/A -40 1.72E+18 14.4 0 7.2 14.4 28.8 -11.2 Circumferential Welds: AB 2 4P7216/0156 Single 0.03 0.79 41 N/A -40 2.30E+17 7.7 0 3.8 7.7 15.4 -24.6 AB 2 4P7216/0156 Tandem 0.03 0.81 41 N/A -60 2.30E+17 7.7 0 3.8 7.7 15.4 -44.6 Nozzles: N12 3 C2593-2 0.04 0.59 26 N/A -30 1.91E+17 4.3 0 2.2 4.3 8.7 -21.3 N12 3 C2594-2 0.04 0.63 26 N/A 0 1.91E+17 4.3 0 2.2 4.3 8.7 8.7 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-21 Component Heat %Cu %Ni CF Adjusted CF Initial RT NDT °F 1/4 T Fluence n/cm 2 35 EFPY RT NDT °F I Margin °F 35 EFPY Shift °F 35 EFPY ART °F Best Estimate Chemistries from BWRVIP-135 R1: Plate A1224-1 0.035 0.65 23 N/A 0 1.72E+18 12.3 0 6.1 12.3 24.6 24.6 Weld 5P6214B/0331 Single 0.019 0.828 26.3 N/A -50 1.72E+18 14.0 0 7.0 14.0 28.1 -21.9 Weld 5P6214B/0331 Tandem 0.019 0.828 26.3 N/A -40 1.72E+18 14.0 0 7.0 14.0 28.1 -11.9 Integrated Surveillance Program (BWRVIP-135 R1): Plate A1224-1 0.03 0.65 20 47.87 4 0 1.72E+18 25.6 0 8.5 17.0 42.6 42.6 Weld 5P6214B Single 0.02 0.92 27 39.75 4,5 -50 1.72E+18 21.2 0 10.6 21.2 42.4 -7.6 Weld 5P6214B Tandem 0.02 0.92 27 39.75 4,5 -40 1.72E+18 21.2 0 10.6 21.2 42.4 2.4 Notes: 1. Use of SMAW Heats 422K8511, 627069, 626677, and 627260 was determined to be limited to we ld pick-ups at the ID/OD surfaces or initial root pass or sealing at the backing bars which were ground out or subsequently removed. Certified Material Test Reports indicate that no SMAW weld material is present at either the 1/4T or 3/4T location. Therefore, these heats are not required to be evaluated as part of the beltline region. 2. Weld AB occurs within the extended beltline region, defined as experiencing a fluence >1.0E17 n/cm

2. 3. The N12 Water Level Instrumentation Nozzle occurs in the beltline region. Because the forging is fabricated from stainless steel, the ART is calculated using the plate heats where the nozzles occur. For GGNS, these nozzles occur in only two (2) of the Shell 2 plates. 4. The adjusted Chemistry Factor (CF) is determined using the methods defined in RG1.99 Revi sion 2 (Reference 26), Position 2. 5. Weld Heat 5P6214B is represented by materials in BWRVIP-135 Revision 1 (Reference 25) with two (2) different chemistries. Recommendations provided in BWRVIP-135 Revision 1 have been employed to determine the surveillance chemistry used fo r calculating the adjusted CF.

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2-22 Table 2.1-3 35 EFPY Effects of Irradiation on RPV Circumferential Weld Properties Parameter NRC Staff Assessment for 32 EFPY (Circ Welds) Grand Gulf 35 EFPY AB (Lower Circ Seam) Submerged Arc Weld (SAW) 4 AC (Upper Circ Seam) SAW 5 AC (Upper Circ Seam) SMAW 5 (CB&I RPV) (CB&I Vessel) (CB&I Vessel) (CB&I Vessel) Cu% 0.1 0.03 0.04 0.02 Ni% 0.99 0.81 0.95 0.91 CF 134.9 41 54 27 Fluence at clad/weld

interface (10 19 n/cm 2) 0.51 0.253 0.253 0.253 RT NDT(U) (°F) 40 60 RT NDT w/o margin (°F) (See Note 3) 109.5 25.7 33.9 16.9 Mean RT NDT (°F) 44.5 -14.3 13.9 -43.1 P (F/E) NRC (See Note 1) 2.00E-07 (Note 2) (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 th e GGNS values at the end of license are less than the 32 EFPY value provided by the NRC leads to the c onclusion that the GGNS RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in

GL 98-05 (Reference 11).

3. RT NDT = CF

f (0.28 - 0.10 log f)

4. GGNS does not have a circumferential weld in the beltline region. Therefore, Weld AB, which is approximately 5 inches below the active core, is presented conservatively using the value of peak fluence at 35 EFPY. 5. GGNS does not have a circumferential weld in the beltline region. Therefore, Weld AC, which is approximately 22 inches above the active core, is presented conservatively using the value of peak fluence at

35 EFPY. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-23 Table 2.1-4 Comparison of Key Parameters Influencing FAC Wear Rate, GGNS CLTP vs. EPU Line Service Description 1 Line No. Pipe Diam (inch) Fluid State CLTP Temp. (°F) EPU Temp. (°F) Temp. Change 2 (%) CLTP Steam Quality EPU Steam Quality Quality Change 2 (%) CLTP Velocity (fps) EPU Velocity (fps) Velocity Change 2 (%) Median Change in Wear Rate (%) 3 Steam from HP Turbine to MSRs 12-GBD-32 12 Wet Steam 431.49 437.53 1.4 89 90 1.0 131.4 166.7 27 6 Steam from HP Turbine to MSRs 16-GBD-32 16 Wet Steam 431.49 437.53 1.4 89 90 1.0 81.3 103.2 27 0 Steam from HP Turbine to MSRs turbine stub 14 Wet Steam 431.49 437.53 1.4 89 90 1.0 111.9 142.1 27 18 #3 FWH Condensate Outlet 18-FBD-18A 18 Subcooled 200.80 207.00 3.1 10.8 12.4 15 20 #4 FWH Condensate Outlet 18-FBD-19A 18 Subcooled 254.40 260.30 2.3 11.1 12.7 15 0 #4 FWH Condensate Outlet 16-FBD-19A 16 Subcooled 254.40 260.30 2.3 14.0 16.1 15 17 #4 FWH Condensate Outlet 20-FBD-21 20 Subcooled 254.55 260.45 2.3 17.8 20.5 15 0 #4 FWH Condensate Outlet 24-FBD-21 24 Subcooled 254.50 260.40 2.3 18.5 21.3 15 0 Condensate Header

to RFP Suction 30-FBD-23A 30 Subcooled 254.50 260.40 2.3 11.7 13.4 15 0 Condensate to RFP Suction 24-FBD-24A 24 Subcooled 283.90 290.50 2.3 14.4 16.4 14 0 RFP Discharge 24-CBD-1A 24 Subcooled 285.10 292.10 2.5 16.9 19.3 14 0 RFP Discharge 28-DBD-14A 28 Subcooled 285.10 292.10 2.5 11.6 13.3 14 0 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-24 Line Service Description 1 Line No. Pipe Diam (inch) Fluid State CLTP Temp. (°F) EPU Temp. (°F) Temp. Change 2 (%) CLTP Steam Quality EPU Steam Quality Quality Change 2 (%) CLTP Velocity (fps) EPU Velocity (fps) Velocity Change 2 (%) Median Change in Wear Rate (%) 3 FW from FWH 5 to FWH 6 28-DBD-16A 28 Subcooled 326.00 332.90 2.1 11.9 13.6 14 0 FW From FWH 6 28-DBD-18A 28 Subcooled 417.10 420.50 0.8 12.7 14.5 14 17 FW From FWH 6 24-DBD-18A 24 Subcooled 417.10 420.50 0.8 18.3 20.9 14 5 Common FW Header 30-DBD-19 30 Subcooled 417.10 420.50 0.8 21.5 24.5 14 0 North and South

Main Headers 24-DBD-25A 24 Subcooled 417.10 420.50 0.8 18.3 20.9 14 11 FW (After 1st Branch to RPV) 18-DBA-13A 18 Subcooled 417.30 420.62 0.8 20.8 23.7 14 13 12-inch FW Branch

Lines to RPV 12-DBA-17A 12 Subcooled 417.30 420.62 0.8 20.9 23.8 14 11 Extraction Steam to FWH 1 LP to FWH 1 32 Wet Steam 145.82 149.36 2.4 33 37 11.2 48.7 63.4 30 100 Extraction Steam to FWH 2 LP to FWH 2 32 Wet Steam 174.16 178.35 2.4 86 86 0.0 127.3 146.0 15 38 Extraction Steam to FWH 3 LP to FWH 3 36 Wet Steam 209.71 214.76 2.4 98 97 -0.3 126.5 152.0 20 27 Extraction Steam to FWH 5 24-HBD-3A 24 Wet Steam 337.11 346.33 2.7 86 86 0.3 136.6 136.2 0 11 #5 Extraction Steam

to SSG 24-HBD-3C 24 Wet Steam 337.11 346.33 2.7 86 86 0.3 2.6 2.2 -11 10 Extraction Steam to FWH 6 HP stub 14 Wet Steam 426.69 432.27 1.3 89 90 1.0 321.4 343.4 7 0 Extraction Steam to FWH 6 24-GBD-1A 24 Wet Steam 426.69 432.27 1.3 89 90 1.0 107.8 115.2 7 -8 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-25 Line Service Description 1 Line No. Pipe Diam (inch) Fluid State CLTP Temp. (°F) EPU Temp. (°F) Temp. Change 2 (%) CLTP Steam Quality EPU Steam Quality Quality Change 2 (%) CLTP Velocity (fps) EPU Velocity (fps) Velocity Change 2 (%) Median Change in Wear Rate (%) 3 Drain from FWH 5 to

HDT 20-HBD-607A 20 Saturated Liquid 336.80 346.00 2.7 1.1 1.3 15 0 Drain from FWH 5 to

HDT 24-HBD-622 24 Saturated Liquid 336.80 346.00 2.7 1.5 1.8 15 17 Drain from FWH 6 to

HDT 12-GBD-56A 12 Saturated Liquid 340.70 347.20 1.9 8.1 8.4 4 0 Drain from FWH 4 to FWH 3 8-HBD-570A 8 Saturated Liquid 234.30 219.20 -6.4 2.8 3.1 12 0 Drain from FWH 4 to FWH 3 10-HCD-79A 10 Saturated Liquid 234.30 219.20 -6.4 1.7 1.9 12 0 MSR (Shell) Drains

to MSSDT 14-HBD-912A 14 Saturated Liquid 338.20 347.40 2.7 0.2 0.2 19 20 MSR (Shell) Drains

to MSSDT 20-HBD-912A 20 Saturated Liquid 338.20 347.40 2.7 0.7 0.8 19 22 MSSDT Drain to

HDT 10-HBD-1052A 10 Saturated Liquid 338.20 347.40 2.7 2.5 3.0 19 20 MSR (Chevron) Drains to MSDT 14-HBD-914A 14 Saturated Liquid 338.20 347.40 2.7 0.8 0.9 19 25 MSR (Chevron) Drains to MSDT 24-HBD-914A 24 Saturated Liquid 338.20 347.40 2.7 1.0 1.2 19 33 MSDT Drain to HDT 14-HBD-874A 14 Saturated Liquid 338.20 347.40 2.7 3.0 3.6 19 25 MSR to 2nd Stage

RDT 24-DBD-66A 24 Wet Steam 535.99 531.86 -0.8 10.0 2.5 -75.1 3.5 1.3 25 2nd Stage RDT to 6 FWH 14-DBD-64A 14 Saturated Liquid 535.95 531.86 -0.8 0.0 0.0 3.0 1.9 25 MSR to 1st Stage

RDT 18-GBD-114A 18 Wet Steam 427.87 433.26 1.3 10.0 2.5 -74.9 9.6 3.7 -61 0 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-26 Line Service Description 1 Line No. Pipe Diam (inch) Fluid State CLTP Temp. (°F) EPU Temp. (°F) Temp. Change 2 (%) CLTP Steam Quality EPU Steam Quality Quality Change 2 (%) CLTP Velocity (fps) EPU Velocity (fps) Velocity Change 2 (%) Median Change in Wear Rate (%) 3 1st Stage RDT to

HDT 8-GBD-113A 8 Saturated Liquid 427.87 433.26 1.3 0.0 0.0 4.9 4.8 -2 0 #5 Extraction Steam

to SSG no number 6 Wet Steam 339.05 348.27 2.7 85.6 85.8 0.3 75.1 67.1 -11 20 #5 Extraction Steam

to SSG no number 10 Wet Steam 339.05 348.27 2.7 85.6 85.8 0.3 25.9 23.2 -11 0 SSG to SSG Drain

Tank no number 8.625 Saturated Liquid 335.19 344.35 2.7 0.0 0.0 13.0 1.3 1.3 1 0 SSG Drain Tank to

F502 4-HBD-443 4 Saturated Liquid 335.19 344.35 2.7 0.0 0.0 13.0 0.9 0.9 1 0 Reactor Water to

DBA-11 2-DBA-40 2 Subcooled 533.00 530.80 -0.4 4.5 4.5 0 0 From DBA-40&41 to

DBA-9 4-DBA-11 4 Subcooled 533.00 530.80 -0.4 2.2 2.2 0 0 From DCA-1 to

DBA-9 4-DBA-10 4 Subcooled 533.00 530.80 -0.4 5.7 5.7 0 0 From DBA-10 to

DBA-89 6-DBA-9 6 Subcooled 533.00 530.80 -0.4 5.0 5.0 0 0 From DBA-9 to DBZ-22 6-DBA-89 6 Subcooled 533.00 530.80 -0.4 5.8 5.8 0 0 Inlet to RHX, DBZ-2 4-DBZ-22A 4 Subcooled 533.00 530.80 -0.4 13.2 13.1 0 0 Inlet to RHX, DBZ-2 4-DBZ-22B 4 Subcooled 433.00 430.80 -0.5 11.9 11.8 0 0 Inlet to RHX, DBZ-2 4-DBZ-22C 4 Subcooled 333.00 330.80 -0.7 11.0 11.0 0 0 RHX to NRHX 4-DBZ-4 4 Subcooled 233.00 230.80 -0.9 10.5 10.4 0 0 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-27 Line Service Description 1 Line No. Pipe Diam (inch) Fluid State CLTP Temp. (°F) EPU Temp. (°F) Temp. Change 2 (%) CLTP Steam Quality EPU Steam Quality Quality Change 2 (%) CLTP Velocity (fps) EPU Velocity (fps) Velocity Change 2 (%) Median Change in Wear Rate (%) 3 RHX Outlet 6-DBZ-5 6 Subcooled 437.00 433.90 -0.7 5.2 5.2 0 0 From DBZ-5 to DBB-71 & 88 6-DBB-104 6 Subcooled 437.00 433.90 -0.7 5.8 5.7 0 0 From DBB-104 to DBB-68 6-DBB-88 6 Subcooled 437.00 433.90 -0.7 2.9 2.9 0 0 RWCU Discharge to FW 12-DBB-68 12 Subcooled 437.00 433.90 -0.7 0.7 0.7 0 0 Notes: 1. Similar lines have been combined (e.g., there are #3FWH A, B and C, but they are reported only as #3FWH).

2. Percent change in wear rates is reported as EPU/CLTP. It provides a relative comparison to other systems.

3. Negative number indicates reduced wear.

4. Acronyms: FWH: Feedwater Heater NHRX: Non-Regenerative Heat Exchanger HDT: Heater Drain Tank RD T: Reheater Drain Tank HP: High Pressure RFP: Reactor Feed Pump MSDT: Moisture Separator (Chevron) Drain Tank RHX Regenerative Heat Exchanger MSR: Moisture Separator Reheater SSG: Seal Steam Generator MSSDT: Moisture Separator Shell Drain Tank NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-28 Table 2.1-5 Sample of Components with Highest Predicted Wear Rates, GGNS CHECWORKS TM SFA-Predicted Thickness vs. Measured Thickness Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 1 Condensate: Heater Drain Pump (HDP) Tee - RFP N19-538A 90-Degree Elbow 24 0.688 RFO 14 0.695 0.595 1.17 2 Condensate: HDP Tee - RFP N19-531B 90-Degree Elbow 24 0.688 RFO 14 0.716 0.595 1.20 3 Condensate: HDP Tee - RFP N19-527 Tee 30 0.75 RFO 15 1.232 0.636 1.94 4 Condensate: Heater 3 - Heater 4 N19-400A Exit Nozzle 18 0.562 RFO 14 0.525 0.387 1.36 5 Condensate: Heater 4 - HDP Tee N19-525 Straight Pipe 24 0.688 RFO 15 0.716 0.628 1.14 6 Condensate: Heater 4 - HDP Tee N19-508C 90-Degree Elbow 16 0.5 RFO 14 0.415 0.363 1.14 7 Condensate: Heater 4 - HDP Tee N19-504B Straight Pipe 16 0.5 RFO 15 0.493 0.419 1.18 8 Condensate: Heater 4 - HDP Tee N19-516 45-Degree Elbow 24 0.688 RFO 14 0.66 0.553 1.19 9 Condensate: Heater 4 - HDP Tee N19-519 90-Degree Elbow 24 0.688 RFO 16 0.654 0.537 1.22 10 Condensate: Heater 4 - HDP Tee N19-502B Concentric Reducer 18 0.562 RFO 16 0.55 0.45 1.22 11 Condensate: Heater 4 - HDP Tee N19-510C Concentric Expander 16 0.5 RFO 14 0.509 0.396 1.29 12 Condensate: Heater 4 - HDP Tee N19-526 Concentric Expander 24 0.688 RFO 16 0.782 0.574 1.36 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-29 Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 13 Condensate: Heater 4 - HDP Tee N19-511 Tee 20 0.594 RFO 15 0.856 0.48 1.78 14 Extraction Steam: HP Turbine to

Heater 5 N36-006A 45-Degree Elbow 24 0.5 RFO 15 0.604 0.5 1.21 15 Extraction Steam: HP Turbine to

Heater 5 N36-011A Straight Pipe 24 0.5 RFO 15 0.628 0.5 1.26 16 Extraction Steam: HP Turbine to

Heater 5 N36-025B 45-Degree Elbow 24 0.5 RFO 15 0.39 0.298 1.31 17 Extraction Steam: HP Turbine to

Heater 5 N36-034B 45-Degree Elbow 24 0.5 RFO 14 0.405 0.298 1.36 18 Extraction Steam: HP Turbine to

Heater 5 N36-036B 90-Degree Elbow 24 0.5 RFO 14 0.436 0.277 1.57 19 Extraction Steam: HP Turbine to

Heater 5 N36-021B 90-Degree Elbow 24 0.5 RFO 15 0.457 0.254 1.80 20 Extraction Steam: HP Turbine to

Heater 6 N36-053A Tee 24 0.562 RFO 16 0.51 0.354 1.44 21 Extraction Steam: HP Turbine to

Heater 6 N36-065B Concentric Expander 14 0.375 RFO 15 0.357 0.114 3.13 22 Extraction Steam: LP Turbine to

Heater 1 N36-097B 90-Degree Elbow 32 0.375 RFO 16 0.348 0.213 1.63 23 Extraction Steam: LP Turbine to

Heater 1 N36-099B 90-Degree Elbow 32 0.375 RFO 16 0.353 0.213 1.66 24 Extraction Steam: LP Turbine to

Heater 2 N36-031C 45-Degree Elbow 32 0.375 RFO 14 0.309 0.218 1.42 25 Extraction Steam: LP Turbine to

Heater 2 N36-134B 45-Degree Elbow 32 0.375 RFO 16 0.374 0.228 1.64 26 Extraction Steam: LP Turbine to

Heater 2 N36-117B Exit Nozzle 32 0.75 RFO 16 0.683 0.404 1.69 27 Extraction Steam: LP Turbine to

Heater 2 N36-034C Exit Nozzle 32 0.75 RFO 15 0.684 0.404 1.69 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-30 Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 28 Extraction Steam: LP Turbine to

Heater 2 N36-087A Exit Nozzle 32 0.75 RFO 16 0.695 0.404 1.72 29 Extraction Steam: LP Turbine to

Heater 3 N36-110A Exit Nozzle 44 0.75 RFO 16 0.647 0.431 1.50 30 Extraction Steam: LP Turbine to

Heater 3 N36-046C Exit Nozzle 44 0.75 RFO 15 0.667 0.431 1.55 31 FW: FW Pump Recirculation N21-248B Concentric Reducer 12.75 1 RFO 15 0.971 0.805 1.21 32 FW: FW Pump Recirculation N21-231A Concentric Expander 8.625 0.322 RFO 15 0.53 0.322 1.65 33 FW: FW Pump Recirculation N21-249B Concentric Expander 8.625 0.322 RFO 15 0.542 0.322 1.68 34 FW: FW Pump Recirculation N21-218A 90-Degree Elbow 12.75 1 RFO 16 1.03 0.542 1.90 35 FW: Heater 6 to Reactor Vessel N21-127A Straight Pipe 24 1.219 RFO 16 1.025 1.062 0.97 1 36 FW: Heater 6 to Reactor Vessel N21-091 Straight Pipe 30 1.452 RFO 15 1.379 1.345 1.03 37 FW: Heater 6 to Reactor Vessel N21-148B Straight Pipe 24 1.219 RFO 14 1.145 1.062 1.08 38 FW: Heater 6 to Reactor Vessel N21-102A Straight Pipe 24 1.531 RFO 15 1.578 1.423 1.11 39 FW: Heater 6 to Reactor Vessel N21-159B Tee 18 0.938 RFO 14 1.103 0.938 1.18 40 FW: Heater 6 to Reactor Vessel N21-137B 90-Degree Elbow 24 1.531 RFO 16 1.535 1.256 1.22 41 FW: Heater 6 to Reactor Vessel N21-157B 45-Degree Elbow 18 0.938 RFO 14 0.83 0.667 1.24 42 FW: Heater 6 to Reactor Vessel N21-077A 90-Degree Elbow 24 1.531 RFO 14 1.618 1.256 1.29 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-31 Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 43 FW: Heater 6 to Reactor Vessel N21-100A 90-Degree Elbow 24 1.531 RFO 16 1.77 1.256 1.41 44 FW: Heater 6 to Reactor Vessel N21-205 45-Degree Elbow 12.75 0.688 RFO 14 0.589 0.402 1.47 45 FW: Heater 6 to Reactor Vessel N21-180 90-Degree Elbow 12.75 0.688 RFO 15 0.552 0.367 1.50 46 FW: Heater 6 to Reactor Vessel N21-207 90-Degree Elbow 12.75 0.688 RFO 16 0.614 0.367 1.67 47 FW: Heater 6 to Reactor Vessel N21-174 90-Degree Elbow 12.75 0.688 RFO 14 0.617 0.367 1.68 48 FW: Heater 6 to Reactor Vessel N21-189 90-Degree Elbow 12.75 0.688 RFO 14 0.633 0.367 1.72 49 FW: Heater 6 to Reactor Vessel N21-163 90-Degree Elbow 12.75 0.688 RFO 14 0.634 0.367 1.73 50 FW: Heater 6 to Reactor Vessel N21-169 90-Degree Elbow 12.75 0.688 RFO 15 0.664 0.367 1.81 51 FW: Heater 6 to Reactor Vessel N21-183 90-Degree Elbow 12.75 0.688 RFO 15 0.682 0.367 1.86 52 FW: Heater 6 to Reactor Vessel N21-090 Tee 30 1.452 RFO 14 2.37 1.232 1.92 53 FW: Heater 6 to Reactor Vessel N21-193 90-Degree Elbow 12.75 0.688 RFO 16 0.738 0.367 2.01 54 FW: Heater 6 to Reactor Vessel N21-130A Tee 24 1.219 RFO 14 1.72 0.825 2.08 55 FW: Heater 6 to Reactor Vessel N21-093 Tee 30 1.403 RFO 14 2.341 1.109 2.11 56 FW: Heater 6 to Reactor Vessel N21-135A Tee 18 0.938 RFO 14 1.09 0.486 2.24 57 FW: Heater 5 to Heater 6 N21-058A 45-Degree Elbow 28 1.438 RFO 14 1.616 1.189 1.36 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-32 Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 58 FW: Heater 5 to Heater 6 N21-050A 90-Degree Elbow 28 1.438 RFO 14 1.616 1.175 1.38 59 FW: Heater 5 to Heater 6 N21-070B 45-Degree Elbow 28 1.438 RFO 15 1.653 1.189 1.39 60 FW: RFP to Heater 5 N21-005A Straight Pipe 24 1.682 RFO 16 1.604 1.459 1.10 61 FW: RFP to Heater 5 N21-024B Exit Nozzle 24 1.812 RFO 14 1.507 1.297 1.16 62 FW: RFP to Heater 5 N21-026B 90-Degree Elbow 24 1.682 RFO 14 1.827 1.307 1.40 63 FW: RFP to Heater 5 N21-013A Tee 24 1.531 RFO 16 1.874 1.232 1.52 64 FW: RFP to Heater 5 N21-009A 90-Degree Elbow 24 1.531 RFO 16 1.62 0.978 1.66 65 FW: RFP to Heater 5 N21-031B 90-Degree Elbow 24 1.531 RFO 15 1.627 0.978 1.66 66 FW: RFP to Heater 5 N21-002A Tee 24 1.682 RFO 16 2.47 1.378 1.79 67 FW: RFP to Heater 5 N21-037B Tee 24 1.531 RFO 15 2.248 1.232 1.82 68 Heater Drains: Drain Tank to Pumps N23-011 Tee 36 0.375 RFO 16 0.343 0.326 1.05 69 Heater Drains: Drain Tank to Pumps N23-064B 90-Degree Elbow 26 0.375 RFO 16 0.392 0.313 1.25 70 Heater Drains: From HDPs N23-073A 90-Degree Elbow 14 0.438 RFO 15 0.482 0.438 1.10 71 Heater Drains: From HDPs N23-079A 90-Degree Elbow 14 0.438 RFO 14 0.449 0.353 1.27 72 Heater Drains: From HDPs N23-085A 45-Degree Elbow 14 0.438 RFO 14 0.494 0.357 1.38 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-33 Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 73 Heater Drains: From HDPs N23-021 Concentric Expander 8.625 0.322 RFO 15 0.536 0.201 2.67 74 Heater Drains: HDP Recirc N23-179B Straight Pipe 8.625 0.322 RFO 15 0.285 0.248 1.15 75 Heater Drains: HDP Recirc N23-185A Orifice 8.625 0.322 RFO 16 0.293 0.13 2.25 76 Heater Drains: Heater 4 to 3 Drain N23-002C 90-Degree Elbow 8.625 0.322 RFO 15 0.32 0.114 2.81 77 Heater Drains: Heater 5 To Drain

Tank N23-001 Tee 24 0.375 RFO 16 0.838 0.375 2.23 78 Heater Drains: Heater 6 To Drain

Tank N23-049B Straight Pipe 12.75 0.375 RFO 16 0.348 0.25 1.39 79 Heater Drains: Heater 6 To Drain

Tank N23-051A 90-Degree Elbow 12.75 0.375 RFO 14 0.348 0.165 2.11 80 Main and Reheat Steam (MRS): Reheat Steam N11-A001 Tee 16 0.375 RFO 15 0.343 0.263 1.30 81 MRS: Reheat Steam N11-A020 Straight Pipe 16 0.375 RFO 15 0.348 0.265 1.31 82 MRS: Reheat Steam N11-B001 Tee 16 0.375 RFO 15 0.351 0.263 1.33 83 MRS: Reheat Steam MRS20-9 90-Degree Elbow 16 0.375 RFO 15 0.361 0.219 1.65 84 MRS: Reheat Steam N11-A027 90-Degree Elbow 16 0.375 RFO 14 0.361 0.219 1.65 85 MSR: 1st RDT to HDT N35-B304 Exit Nozzle 8.625 0.5 RFO 15 0.469 0.192 2.44 86 MSR: 1st RDT to HDT N35-A338 45-Degree Elbow 8.625 0.322 RFO 14 0.305 0.124 2.46 87 MSR: 1st RDT to HDT N35-A312 90-Degree Elbow 8.625 0.322 RFO 16 0.299 0.113 2.65 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-34 Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 88 MSR: 1st RDT to HDT N35-A331 90-Degree Elbow 8.625 0.322 RFO 14 0.301 0.113 2.66 89 MSR: 1st RDT to HDT N35-B329 Inlet Nozzle 18 0.438 RFO 14 0.389 0.383 1.02 90 MSR: Main Steam (MS) to MSDT N35-B124 Eccentric Expander 14 0.375 RFO 15 0.38 0.226 1.68 91 MSR: MS to MSDT N35-B127 Tee 24 0.375 RFO 14 0.395 0.206 1.92 92 MSR: MS to MSDT N35-B118 Exit Nozzle 14 0.438 RFO 14 0.365 0.166 2.20 93 MSR: MSDT to HDT N35-B134 Tee 14 0.375 RFO 14 0.445 0.375 1.19 94 MSR: MSDT to HDT N35-B132 90-Degree Elbow 14 0.375 RFO 16 0.36 0.26 1.38 95 MSR: MSDT to HDT N35-B130.5 Exit Nozzle 14 0.375 RFO 16 0.342 0.219 1.56 96 MSR: MSDT to HDT N35-A129 Exit Nozzle 14 0.375 RFO 15 0.43 0.219 1.96 97 MSR: MSSDT to HDT N35-A061 Tee 10.75 0.365 RFO 14 0.467 0.218 2.14 98 MSR: Moisture Separator Shell (MSS)

to MSSDT N35-A001 Exit Nozzle 14 0.438 RFO 16 0.291 0.361 0.81 2 99 MSR: MSS to MSSDT N35-B056 Tee 10.75 0.365 RFO 16 0.402 0.365 1.10 100 MSR: MSS to MSSDT N35-B034 Tee 20 0.375 RFO 16 0.342 0.222 1.54 101 MSR: MSS to MSSDT N35-B016 Exit Nozzle 14 0.438 RFO 16 0.441 0.259 1.70 102 MSR: MSS to MSSDT N35-B018 90-Degree Elbow 14 0.375 RFO 14 0.248 0.098 2.53 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-35 Item No. CHECWORKS TM Wear Rate Analysis Run Definition Name Component Name Component Type Pipe OD Size (inches) Nominal Thickness (inches) Last Inspection Period Measured Thickness (inches) Predicted Thickness (inches) Ratio of Measured to Predicted Thickness 103 MSR: MSS to MSSDT N35-A002 90-Degree Elbow 14 0.375 RFO 16 0.384 0.098 3.92 104 MSR: MSS to MSSDT N35-B031 90-Degree Elbow 14 0.375 RFO 14 0.517 0.098 5.28 105 MSR: MSS to MSSDT N35-A036 Exit Nozzle 14 0.438 RFO 14 0.323 0.056 5.77 106 MSR: MSS to MSSDT N35-A009 Exit Nozzle 14 0.438 RFO 16 0.436 0.056 7.79 107 MSR: Reheaters to 1st RDT N35-B301 45-Degree Elbow 18 0.438 RFO 16 0.386 0.307 1.26 108 MSR: Reheaters to 1st RDT N35-B300 Exit Nozzle 18 0.438 RFO 16 0.491 0.251 1.96 109 RWCU: RHX A to FW G33-179 Straight Pipe 6.625 0.562 RFO 16 0.483 0.445 1.09 110 RWCU: RHX A to FW G33-154 Straight Pipe 6.625 0.432 RFO 16 0.421 0.329 1.28 111 RWCU: RHX C to RHX B G33-278 90-Degree Elbow 4.5 0.337 RFO 15 0.308 0.229 1.34 112 RWCU: Reactor to RHX B G33-134 Tee 4.5 0.337 RFO 14 0.402 0.337 1.19 113 RWCU: Reactor to RHX B G33-005 45-Degree Elbow 4.5 0.337 RFO 14 0.302 0.225 1.34 114 Seal Steam (SS): Heating Steam to

SSG N33-016 Concentric Reducer 10.75 0.365 RFO 14 0.366 0.256 1.43 115 SS: Heating Steam to SSG N33-132 90-Degree Elbow 8.625 0.322 RFO 15 0.271 0.178 1.52 116 SS: Heating Steam to SSG N33-003 45-Degree Elbow 10.75 0.365 RFO 14 0.384 0.247 1.55 117 SS: Heating Steam to SSG N33-125 Tee 6.625 0.28 RFO 14 0.251 0.142 1.77 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-36 Notes: 1. The measured thickness exceeded the minimum allowed thickness of 0.994 inches. 2. The measured thickness exceeded the minimum allowed thickness of 0.07 inches. 3. OD: Outside Diameter

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2-37 Table 2.1-6 Comparison of RWCU System Operating Conditions Parameter Units CLTP EPU RWCU System MWt 3,898 4,408 RWCU System Inlet Temperature °F 532.7 530.8 RWCU System Inlet Pressure (RPV dome pressure, neglecting head) psia 1,040 1,040 RWCU System Outlet Temperature °F 436.1 433.9 Design RWCU System Flow lbm/hr 178,000 178,000 Current As-Operated RWCU System Flow lbm/hr 190,000 190,000

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2-38 Table 2.1-7 Comparisons of Chemistry Parameters for CLTP and EPU Cases CLTP Values EPU Values Item Parameter Units Nominal RWCU System Flow RWCU System Design Flow Nominal RWCU System Flow RWCU System Design Flow 1 Average reactor water iron concentration 1 ppb 42.74 45.62 48.29 51.54 2 Average reactor water conductivity S/cm 0.077 0.079 0.080 0.082 3 Average chloride

concentration ppb 0.108 0.115 0.122 0.130 4 Average sulfate

concentration ppb 1.393 1.487 1.573 1.680 Note: Above values for EPU are based on current levels of FW iron concentration. The installation of the CFFF system modification is expected to reduce the FW iron concentration below the BWRVIP-130 (Reference 27) recommended goal of 1 ppb for conti nuous operation and signifi cantly lower reactor water concentration. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-39 Table 2.1-8 Selection Process Criteria for Components in the FAC Program Selection Process Criteria Description CHECWORKS TM Model The GGNS FAC program selects components based on the results of the model's output (i.e., wear rate and remaining life). Components are selected from both lines that have not been inspected and from lines that have inspected components. Industry Experience The GGNS FAC program selects inspection components based on OEs from the industry that are applicable to GGNS. Periodically, OEs are reviewed for GGNS applicability. If the event is applicable, suitable components are selected to address the issue. Station Experience The GGNS FAC program selects inspection components based on station experiences. Periodically, the corrective action program is reviewed to discover if any situations had occurred that would be applicable to the program (i.e., valve leak-bys, steam leaks, abnormal valve usage (e.g., open when should be closed)). The thermal performance report is also reviewed periodically to identify any applicable leaking valves whose piping may need to be inspected. Inspection components are also selected based on requests from system engineers or from design changes. Re-Inspections The GGNS FAC program directs that previous inspection results be reviewed. It states that any component found to have less than 87.5% of nominal wall requires calculating the remaining service life (RSL), which includes a safety factor multiplier to the wear rate. If the RSL would be exceeded on the operati ng cycle following inspection, immediate action is required. If its RSL is greater than one operating cycle, but less than two operating cycles, it should be re-inspected and replaced or repaired in the fo llowing outage. For a longer RSL, it requires that the component be re-inspected no later than the outage

preceding the operating cycle it is estimated to exceed its RSL. Susceptible - Not Modeled (S-NM) The GGNS FAC program selects inspection components based on the susceptibility of the non-modeled piping. A large amount of FAC susceptible piping cannot be modele d because of a lack of operating parameter data. This includes almost all of the small-bore piping. This also includes FW heater shells. Lines that are deemed highly susceptible and could have detrimental consequences if failure occurred are slated for inspection. Engineering Judgment The GGNS FAC program also selects inspection components based on engineering judgment. Engineering judgment is used when selecting inspection locations through knowledge based on

experience.

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2-40 2.2 Mechanical and Civil Engineering 2.2.1 Pipe Rupture Locations and Associated Dynamic Effects Regulatory Evaluation SSCs important to safety could be affected by the pipe whip dynamic effects of a pipe rupture. Entergy conducted a review of pipe rupture analyses to ensure SSCs important to safety are adequately protected from the effects of pipe ruptures. The review covered: (1) implementation of criteria for defining pipe break and crack locations and configurations; (2) implementation of criteria dealing with special features, such as augmented ISI programs or the use of special protective devices such as pipe whip restraints (PWRs); (3) pipe whip dynamic analyses and results, including the jet thrust and impingement forcing functions and pipe whip dynamic effects, and (4) the design adequacy of supports for SSCs provided to ensure the intended design functions of the SSCs will not be impaired to an unacceptable level as a resu lt of pipe whip or jet impingement (JI) loadings. The review focused on the effects that the proposed EPU may have on items (1) through (4) above. The regulator y acceptance criteria are based on GDC-4, which requires SSCs important to safety to be designed to accommodate the dynamic effects of a

postulated pipe rupture. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. Dynamic effects of pipe ruptures are described in UFSAR Section 3.6.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 10.1 of the CLTR addresses the effect of CPPU on High Energy Line

Breaks (HELBs). The results of this evaluation are described below. High-energy piping systems inside and outside containment are listed in UFSAR Tables 3.6A-14 and 3.6A-15. Inside containment, the high-energy piping systems affected by EPU are: MS MS Drains NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-41 RCIC Steam Line FW MS Vent Lines MS SRV Piping (Between the MSL and Each SRV) MS drains and RCIC steam line flow rates, pressures and temperatures are unchanged from CLTP to EPU operating conditions. However, because these piping systems are connected directly to MS piping, stresses in these systems were conservatively increased commensurate with the increase in MS system piping stresses. Outside containment, the high-energy piping systems affected by EPU are: MS FW MS drains RCIC RHR RWCU A review was performed of piping stresses that increased due to EPU and postulated pipe break locations. The review was conducted in accordance with the requirements of the original license basis methodology. No changes to the implementa tion of the existing criteria for defining pipe break and crack locations and configurations are being made for EPU. No new break or crack locations are required to be postulated as a result of the increased piping stresses associated with EPU. No changes to the implementation of the existing criteria dealing with special features, such as the use of special protective devices such as PWRs, are being made for EPU. For EPU, HELBs are evaluated for their effects on equipment qualification. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-42 Topic CLTR Disposition GGNS Result Steam Lines [[ Meets CLTR Disposition Liquid Lines

     ]] Meets CLTR Disposition 2.2.1.1 Steam Line Breaks The CLTR states that there is no effect on steam line breaks at rated EPU operating conditions because steam conditions at postulated break lo cations are unchanged. Therefore, EPU has no effect on the mass and energy releases from a HELB in a steam line.

A review of the heat balances produced for the GGNS EPU confirmed that there is no effect on the steam pressure or enthalpy at the postulated break locations (MS and RCIC) at EPU operating conditions. Table 2.2-1 summarizes the high energy steam line break evaluation. The effect of HELBs outside the drywell (DW) on building environmental conditions and sub-compartment pressurization is addressed in Section 2.5.1.3.1. The effect of steam line breaks inside the DW is addressed in Sections 2.6.1, 2.6.2 and 2.6.3. The evaluation of the steam line HELB events in the GGNS licensing basis meets all CLTR dispositions.

2.2.1.2 Liquid Line Breaks As stated in Section 10.1 of the CLTR, EPU ma y increase subcooling in the reactor vessel, which may lead to increased break flow rate s for liquid line breaks. Operation at EPU conditions requires an increase in the MS and FW flows, which results in an increase in FW system pressures. This increase in pressure may lead to increased break flow rates for liquid line breaks. The mass and energy releases for HELBs outside the DW in the RWCU, FW and RHR systems were re-evaluated at EPU operating conditions. The GGNS evaluation of liquid line breaks included the RWCU, FW, and RHR systems as well as the effect of increased RWCU and FW operating pressure on pipe whip and JI. The mass and energy release results of the GGNS evaluation of liquid line breaks at EPU operating conditions (see Section 2.5.1.3.1) outside the DW are provided in Table 2.2-1. The effect of liquid line breaks inside the DW is addressed in Sections 2.6.1, 2.6.2 and 2.6.3. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-43 2.2.1.2.1 RWCU Line Breaks An evaluation of the mass and energy releases for RWCU line breaks at CLTP and EPU conditions determined that the EPU mass releases outside the DW remain unchanged from the existing (CLTP) mass releases. However, the EPU evaluation estimated a slight increase in the energy release in the RWCU HX room as a result of a more accurate representation of the break scenario. No changes are being made to the automatic leak detection logic or to any leak detection system (LDS) settings as a result of EPU because the RWCU HX room maximum temperature is not increased. 2.2.1.2.2 Feedwater/RHR Systems Line Breaks The CLTP mass and energy releases for FW line breaks are affected by changes in the FW system including increased FW flow rate. In accordance with the current licensing basis, the Main Steam Tunnel (MST) in the Auxiliary Building is considered a no-break zone for the FW system. Thus, the non-mechanistic crack break is the only one analyzed in that area. Because the EPU evaluation of the crack break is based on subcooled blowdown, the mass and energy

release rates are significantly increased. Because the primary source feeding the RHR line break is FW, the mass and energy releases were re-e valuated at EPU conditions. The evaluation determined that the EPU mass and energy release rates remain unchanged because the slightly

increased enthalpy is offset by a slightly reduced mass release rate resulting in a negligible net change in the mass and energy releases. HELB evaluations are not assessed for the Turbine Building as it does not contain any essential equipment. The results of the GGNS evaluation of the FW system line break on pressure and temperature transients are provided in Table 2.2-1. 2.2.1.2.3 Pipe Whip and Jet Impingement Pipe whip and JI loads resulting from high energy pipe breaks are a function of system pressure, temperature, and size, as well as proximity to re latively constant pressure sources connected to the line, and the effect of fric tion or line area restrictions between the break and the constant pressure source. Inside containment, the only high-energy pipi ng that experiences an increase in operating pressure due to EPU is in the FW system. Outside containment, the only high-energy piping experiencing an increase in operating pressure in proximity to essential SSCs (similar to UFSAR Table 3.6A-15) due to EPU is in the FW and connected systems (RHR, RCIC, and RWCU). The potential effect of increased FW and connected system pressures at the existing HELB break locations relative to the subsequent effects of pipe whip (targets) and JI loads were evaluated. The resulting EPU pipe whip (targets) and JI loads are bounded by the current licensing basis pipe whip and JI loads. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-44 The adequacies of supports relative to pipe whip and JI loads are evaluated in Section 2.2.2. Therefore, GGNS meets all CLTR dispositions for liquid line breaks.

Conclusion Entergy has reviewed the evaluations related to determinations of rupture locations and associated dynamic effects and concludes the effects of the proposed EPU have been adequately addressed. Entergy further concludes the SSCs important to safety will continue to meet the requirements of GDC-4 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the determination of rupture locations and dynamic effects associated with the postulated rupture of piping. 2.2.2 Pressure-Retaining Components and Component Supports Regulatory Evaluation Entergy has reviewed the structural integrity of pressure-retaining components (and their supports) designed in accordance with the ASME Boiler and Pressure Vessel Code (B&PV Code), Section III, Division 1 (Reference 28), and GDCs 1, 2, 4, 14, and 15. The review focused on the effects of the proposed EPU on the design input parameters and the design-basis loads and load combinations for normal operating, upset, em ergency, and faulted conditions. The review covered: (1) analyses of FIV; and (2) analytical methodologies, assumptions, ASME Code editions, and computer programs used for thes e analyses. The review also included a comparison of the resulting stresses and cumulative fatigue usage factors against the code-allowable limits. The regulatory acceptan ce criteria are based on: (1) 10 CFR 50.55a and GDC-1, insofar as they require SSCs important to safety be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-2, insofar as it requires SSCs important to safety be designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions; (3) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (4) GDC-14, insofar as it requires the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; and (5) GDC-15, insofar as it requires the RCS be designed with margin sufficient to ensure th at the design conditions of the RCPB are not exceeded during any condition of normal operation. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-45 The structural integrity of pressure retaining components and their supports are described in UFSAR Section 3.9, "Mechanical Systems and Components." Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 3.2.2, 3.4 and 3.5 of the CLTR addresses the effect of CPPU on Reactor Vessel Structural Evaluation, FIV and Pi ping Evaluation, respectively. The results of this evaluation are described below. 2.2.2.1 Flow Induced Vibration The FIV evaluation addresses the influence of an increase in flow during EPU on RCPB piping and RCPB piping components. Key applicable structures include the RRS piping and suspension, the MS system piping and suspension, the FW system piping and suspension and the branch lines attached to the MS system piping or FW system piping. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Structural Evaluation of Recirculation Piping [[ Meets CLTR Disposition Structural Evaluation of Main Steam and Feedwater Piping Meets CLTR Disposition Safety-Related Thermowells and Probes

     ]] Meets CLTR Disposition 2.2.2.1.1 Structural Evaluation of Recirculation Piping The CLTR states that there is no significant in crease in the recirculation flow rate at EPU conditions. The recirculation system drive flow increased from 17.0 Mlb/hr per loop at CLTP to 17.2 Mlb/hr per loop at EPU, resulting in an increase of 1.2% during EPU operation. Consequently, the FIV levels of the RRS components are expected to remain essentially the same. Because RRS flow rates for EPU are essentially the same as pr eviously experienced and tested, no further NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-46 evaluation or testing of the FIV levels of the RRS piping, branch piping (e.g., attached Residual Heat Removal (RHR) piping), or its suspension system is required. The FIV effect on RRS piping inside containment at GGNS meets all CLTR dispositions because the nominal reactor dome pressure remains the same and the RRS maximum drive flow does not increase more than 5%. 2.2.2.1.2 Structural Evaluation of Main Steam and Feedwater Piping The CLTR states that MS and FW flow rates increase due to the power uprate. As a result of the increased flow rates and flow velocities, the MS and FW piping experience increased vibration levels, approximately proportional to the square of the flow velocities. Thus, for GGNS, vibration levels may increase by up to 40%. The ASME Code (NB-3622.3) and nuclear regulatory guidelines require some vibrati on test data be taken and evaluated for these high energy piping systems during initial operation at EPU conditions. Vibration data for the MS and FW piping inside containment will be acquired using remote sensors, such as displacement probes, velocity sensors, and accelerometers. A piping vi bration startup test program, which meets the ASME code and regulatory requirements, will be performed. Therefore, the assessment of the structural evaluation of MS and FW Piping meets all CLTR dispositions. FIV testing of the MS and FW piping system will be performed during EPU power ascension. Additional information related to the MS piping is provided in Section 2.5.4.1.1. 2.2.2.1.3 Safety-Related Thermowells and Probes As explicitly stated in Section 3.4 of the CLTR, MS and FW flow rates increase due to the power uprate. [[ ]] of safety-related thermowells and probes in the MS and FW piping systems at EPU conditions. The MS system flow increased from 4.194 Mlb/ hr per line at CLTP to 4.857 Mlb/hr per line resulting in an increase of 15.8% during EPU operation. The FW system flow increased from 8.37 Mlb/hr per line at CLTP to 9.697 Mlb/hr pe r line resulting in an increase of 15.8% during EPU operation. The safety-related thermowells and probes in the MS and FW piping systems were evaluated and found to be adequate for the increased MS and FW flow as a result of EPU. The methodology used to evaluate components for FIV for EPU is described in Section 3.4.1 of the CLTR. This evaluation utilizes SAP4G07 to develop the dynamic finite element models of the MSS / FW system thermowells and sample probes. Three-dimensional beam elements are used to model the thermowell and sample probe sockolet/pipe weld. At each nodal point of the beam elements, six degrees of freedom are assumed: three translations and three rotations. At the sockolet/pipe weld to the outer pipe wall, all six degrees of freedom are fixed. The masses of the thermowells and sample probes and NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-47 sockolet/pipe weld are lumped at the nodal points, which include both the structural mass and fluid mass displaced by the thermowells and sample probes. These added masses are used to account for the effects of fluid on the thermowells and sample probes vibration responses. The non-dimensional quantity defined as V r, termed reduced velocity (Reference 28, Figure N-1323-1), is used to assess whether or not high FIV response is likely. To assess whether or not synchronization of vortex shedding frequency and tube natural frequency occurs, a damping parameter (Reference 28, Figure N-1323-1) is calculated. To calculate the structural response, a non-dimensional parameter, termed reduced damping (Reference 28, N-1324.1 Equation 76), was calculated. For off resonance (non lock-in) condition, the structural response is ordinarily small and was calculated using the standard method (Reference 28, N-1324.2, first paragraph). For thermowell and sample probe resonant structural res ponse, Reference 28, Table N-1324.2(a)-1 was used. The total vibratory stress was calculated by using the square root of the sum of the squares (SRSS) of the oscillating lift and drag forces. These are unlikely to occur at the same time and hence it is conservative to use the SRSS method. The results of the analyses are presented below: Item Component Analyzed Unit EPU Value Allowable 1 MSS Thermowell (TE-N040) psi 836 7,690 psi for Carbon Steel 2 FW System Thermowell (TE-N041) psi 2823 7,690 psi for Carbon Steel 3 RRS Thermowell (TE-N023) psi 2564 10,880 psi for Stainless Steel Therefore, GGNS meets all CLTR dispositions for safety-related thermowells and probes. 2.2.2.2 Piping Evaluation 2.2.2.2.1 Reactor Coolant Pressure Boundary Piping (Non-FIV) Evaluation The RCPB systems evaluation consists of a number of safety-related piping subsystems that move fluid through the reactor and other safety systems. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-48 Topic CLTR Disposition GGNS Result Structural Evaluation for Unaffected Safety-Related Piping [[ Meets CLTR Disposition Structural Evaluation for Affected Safety-Related

Piping

     ]] Meets CLTR Disposition 2.2.2.2.1.1 Structural Evaluation for Unaffected Safety-Related Piping As stated in Section 3.5.1 of the CLTR, the flow, pressure, temperature, and mechanical loading for most of the RCPB piping systems do not increase for EPU. Consequently, there is no change

in stress and fatigue evaluations. [[

     ]] [[

     ]] [[                                                                                                                                                                                     
                                                                        ]] and therefore, GGNS meets all CLTR dispositions for the structural evaluation for unaffected safety-relate d piping. Table 2.2-2 provi des the justification for [[

     ]] Pipe Whip and Jet Impingement Pipe whip and JI loads resulting from high energy pipe breaks are a function of system pressure, temperature, and size, as well as proximity to relatively constant pressure sources connected to the line, and the effect of friction or line area restric tions between the break and the constant pressure NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-49 source. The resulting EPU pipe whip and JI lo ads are bounded by the current licensing basis pipe whip and JI loads. Additionally, a review of pipe stress calculations determined that there are no FW temperature increases associated with EPU conditions that will result in pipe stress levels above the thresholds

required for postulating HELBs, except at locations already evaluated for breaks. As a result, EPU conditions do not result in new HELB locations, nor affect existing HELB evaluations of PWRs and jet targets. 2.2.2.2.1.2 Structural Evaluation for Affected Safety-Related Piping As stated in Section 3.5.1 of the CLTR, the FW a nd MSL piping and associated branch piping up to the first anchor or support will experience an increase in the flow, pressure, and/or temperature resulting in an increase in stress and fatigue. For all systems, the maximum stress levels and fatigue analysis results were reviewed based on specific increases in temperature, pressure, and flow rate (see Tables 2.2-3a and 2.2-3b). Simp le, conservative scaling factors are used to evaluate the effects of EPU pressure, temperature, and flow changes. Pipe stress increases with pressure increases, temperature increases, and fl ow increases. EPU operation also increases the pipe support loads due to the above effects as well as increased fluid transient Turbine Stop Valve Closure (TSVC) loads that result from the increased steam flow rates (see Tables 2.2-3a and 2.2-3b). The factors in Tables 2.2-3a and 2.2-3b were derived using NRC approved GEH methodology detailed in the CLTR. This analysis determined that the maximum increase in piping support load due to TSVC was 22.99%. Therefore, this factor was conservatively applied to the CLTP TSVC support load to determine EPU loads. This analysis method shows that interface loads on snubbers, struts, guides, and flange connections due to EPU load increases are within the design limits (capacities) of these components. For RCPB MS piping outside containment, detailed steam hammer forcing functions were developed and a piping stress analysis was performed to determine less conservative loads at EPU. Direct time history input and analyses for the TSVC transient were used in Table 2.2-5a. This evaluation determined that the piping stress due to the TSVC transient was bounded by the existing analyses. This analysis resulted in all MS piping outside containment meeting all code criteria. There are no pipe supports on the RCPB MS piping outside containment. The piping systems affected by EPU have been evaluated and found to meet the appropriate code criteria for the EPU conditions, based on the design margins between actual stresses and code limits in the original design. All piping stresses are below the code allowables of the plant code of record. The pipe supports of the systems affected by EPU loading increases were reviewed to determine if there is sufficient margin to code acceptance criteria to accommodate the increased loadings. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-50 This review shows that there is adequate design margin between the original design and code limits of all supports inside and outside containment. Main Steam and Associated Piping Sy stem Evaluation (Inside Containment) For GGNS, an increase in flow, pressure, temperature and mechanical loads was evaluated on a plant-specific basis consistent with the me thods specified in Appendix K of ELTR1. [[

                                              ]] are required to demonstrate that the calculated stresses and fatigue usage factors are less than the code allowable limits in accordance with the requirements of the applicable code of record in the existing design basis stress report.

The MS and associated branch piping inside containment was evaluated for compliance with the original code of record, ASME B&PV Code, Section III, Subsection NB, 1980 Edition stress

criteria, including the effects of EPU on piping stresses, piping supports, piping interfaces with the RPV nozzles, penetrations, flanges, and valves. Because the MS piping pressures and temperatures are not affected by EPU, there is no effect on the analyses for these parameters. Seismic inertia loads, seismic building displacement loads, and SRV discharge loads are not affected by EPU; thus, there is no effect on the analyses for these load cases. Other external loading conditions, such as AP, chugging, and condensation oscillation (CO) also are not changed by EPU (see Section 2.6.1.2.1). The increase in MS flow results in increased forces from the TSVC transient. The TSVC loads bound the main steam isolation valve closure (MSIVC) loads because the MSIVC time is significantly longer than the stop valve closure time. The TSVC transient loading is considered one of the most significant loads for the qualification of piping and supports to EPU conditions because the TSVC load was already a significant load at transient CLTP conditions. The increase in the MS flow rate for EPU will in crease this load. The EPU evaluations used conservative scaling factors (see Table 2.2-3a) to identify the components, pipe stress, and support loads that might exceed their AV.

Pipe Stresses

A review of the increase in flow associated w ith EPU indicates that piping load changes do not result in load limits being exceeded for the MS system and attached branch piping or for RPV nozzles and containment penetrations. The original design analyses have sufficient margin between calculated stresses and ASME B&PV Code, S ection III, Subsection NB, 1980 Edition allowable limits (see Tables 2.2-4a through 2.2-4l) to justify operation at EPU conditions. The pressure and temperature of the MS piping are unchanged for the EPU. Similarly, the branch pipelines (Safety Relief Valve Discharge Line (SRVDL), Reactor Core Isolation Cooling (RCIC), RPV Vent, and MS drains including Main Steam Isolation Valve (MSIV) Drain) connected to the MS headers were evaluated to determine the effect of the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-51 increased MS flow on the lines. This evaluation concluded that there is no adverse effect on the existing MS branch line qualifications due to the increased flows resulting from EPU. As with the MS piping, the pressures and temperatures fo r these branch pipelines do not change as a result of EPU. A review was performed of pos tulated pipe break locations. The review was conducted in accordance with the requirements of the current license basis methodology. As a result of this review, no new postulated break locations were identified. Based on existing margins available for the MS piping, it was conclude d that EPU does not result in reactions in excess of the current design capacity. For the portion of the MS Class 1 piping that is located outside containment, a review of the increase in flow associated with EPU indicates that piping load changes do not result in load limits being exceeded for the MS system and attached branch piping or for containment penetrations. Based on the revised analyses performed for the MS system outside containment, the original design has sufficient margin between calculated stresses and the allowable limits in the code of record, ASME B&PV Code, Section III, Subsection NB, 1974 Edition with Addenda through Summer 1975, with exceptions and use of some sub-subsections from the 1977 Edition with Addenda through Summer 1979 and the 1980 Edition with Addenda through Summer 1981 (see Table 2.2-5a) to justify operation at EPU conditions. The pressure of the MS piping is unchanged for the EPU. There is a slight change in temperature for MS piping outside containment. Pipe Supports The MS piping inside containment was evaluated for the effects of flow increase on the piping snubbers, hangers, struts, and PWRs. A review of the increase in MS flow associated with EPU indicates that piping load changes do not exceed component or structure allowables. The TSVC transient load is one of the individual loads experiencing the most significant increase for the qualification of piping and supports to EPU conditi ons. The increased load is due to the increase in the MS flow rate for EPU. The EPU eval uations used conservative scaling factors (see Table 2.2-3a) to identify the components, pipe stress, and support loads that might exceed their AVs. Applying these conservative EPU scaling factors to cover CLTP and OLTP loads resulted in no Class 1 piping supports inside containment that exceeded the allowable loads.

Main Steam Isolation Valves The MSIVs are part of the RCPB and perform the safety function of steam line isolation during certain abnormal events and accidents. The MSIVs must be able to close within a specified time range at all design and operating conditions. They are designed to satisfy leakage limits set forth in the plant TSs. These design requirements are not adversely affected by in creased EPU flow; thus the original design remains adequate for EPU conditions. The MSIVs have been evaluated, as discussed in Section 4.7 of ELTR2, Supplement 1 (Reference 4). The evaluation covers both the effects of the changes to the structural capability NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-52 of the MSIV to meet pressure boundary requireme nts, and the potential effects of EPU-related changes to the safety functions of the MSIVs. The generic evaluation from ELTR2 is based on: (1) a 20% thermal power increase, (2) an increased operating dome pressure to 1095 psia, (3) a reactor temperature increase to 556°F, and (4) steam and FW flow increases of about 24%. Table 1-2 provides the Maximum Nominal Dome Pressure and Temperature as well as the changes in steam and FW flows. From these parameters, it can be determined that the evaluation from ELTR2 is applicable to GGNS. The Hydraulic Control System controls the MSIVC time. Adjusting the pressure compensated flow control valve controls the stroke time. Once the flow control valve is adjusted for a particular extend speed, the pressure compensator feature of the valve will maintain this set speed even if external forces, such as valve disc thrust, vary. The hydraulic damper senses the combined driving force of the pneumatic cylinder, the external closing springs, the steam drag force, the dead weight of the moving components, and the friction force. The steam drag force applied on the main disc increases due to an increase in steam flow rate. This force change is transmitted from the main disc to the valve stem, and then to the connecting hydraulic damper rod. It is then transmitted to the hydraulic damper and the hydraulic control circuit. As the driving force increases due to the higher steam flow rate, a spring inside the hydr aulic control valve reduces the opening of an internal variable orifice in order to compensate for the higher closing force. The net driving force stays unchanged due to this compensating mechanism. The self-compensating feature of

the hydraulic control valve will ensure the MSIV closing time will stay essentially unchanged as a result of changes in the steam flow rate. Therefore, the MSIV performance is bounded by conclusions of the evaluation in Section 4.7 of ELTR2, and the GGNS MS IVs are acceptable for EPU operation. Feedwater System Evaluation The temperature and pressure changes are in significant for EPU. The flow change of approximately 13.1% does not affect the reactor feedwater (RFW) piping system because water hammer loads were not a design load in the orig inal stress analyses. Therefore, the current licensing basis for the RFW system inside containment complies with Section III of the ASME B&PV Code for the effect of thermal expansion displacement on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, flanges, and valves also remain valid per the current licensing basis. This discussion also applies to the FW system and associated branch piping outside containment. The piping design was evaluated for compliance w ith the code of record stress criteria (ASME B&PV Code, Section III, Subsection NB, 1974 Edition with Addenda through Summer 1975, with exceptions and use of some sub-subsections from the 1977 Edition with Addenda through Summer 1979 and the 1980 Edition with Addenda through Summer 1981) and for the effects of EPU on piping supports including the associated building structure. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-53 Because the FW system piping pressures and temperatures increase slightly due to EPU, the effect of these parameters on the existing analyses was evaluated. Seismic inertia loads, and seismic building displacement loads are not aff ected by EPU; thus, ther e is no effect on the analyses for these load cases. Other external loading conditions, such as chugging and CO are not changed by EPU. AP has no effect on piping outside containment. It is bounded by the CLTP design basis AP loads and also by other mechanical loading (e.g., SRV discharge loads). For the FW piping outside containment, there is no FW system fluid transient analysis in the existing design basis analysis so the increase in FW system flow has no effect on the original analysis. Pipe Stresses A review of the changes in pressure, temperatur e and flow associated with EPU indicates that piping load changes do not result in load limits being exceeded for the RFW piping system, for RPV nozzles, and at postulated pipe break locations. Therefore, the current licensing basis RFW stress reports are adequate and do not have an adverse effect on the RFW piping design to justify operation at EPU conditions. This discussion also pertains to that portion of the Class 1 FW piping outside containment. A review of the increase in flow, operating pressure, and temperature associated with EPU indicates that piping load changes do not result in load limits being exceeded for the FW piping system and attached branch piping. A review was also performed of postulated pipe break locations in accordance with the original license basis methodology. As a result of this review, no new postulated break locations were identified. Based on existing margins available for the FW system piping (see Table 2.2-5c), it was concluded that EPU does not have an adverse effect on the FW piping design. Pipe Supports A review of the changes in pressure, temperature a nd flow associated with EPU indicates that piping load changes do not result in load limits being exceeded for the RFW piping system; therefore, the pipe supports for the RFW piping system complies with the changes in pressure, temperature and flow at EPU conditions. A review of the changes in temperature associated with EPU indicates that the effects of thermal expansion displacements do not change for the RFW piping system; therefore, the current licensing basis remains valid. Seismic inertia loads and seismic building displacemen t loads are not affected by EPU; thus, there is no effect on the analyses for these load cases. Other external loading conditions such as AP, chugging, and CO also are not changed by EPU. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-54 The FW system piping outside containment was ev aluated for the effects of EPU flow, operating pressure, and temperature increase on the piping snubbe rs, hangers, and struts. The review indicates that piping load changes do not result in any load limit being exceeded (see Table 2.2-5c). Therefore, the existing analyses bound the EPU conditions. Other Piping Evaluation As previously noted, the nominal operating pressure and temperature of the reactor are not changed by EPU. Aside from MS and FW, no other system connected to the RCPB experiences an increased flow rate at EPU conditions. Only minor changes to fluid conditions are experienced by these systems due to higher steam flow from the reactor and the subsequent change in fluid conditions within the reactor. Other external loading conditions, such as AP, chugging, and CO also are not changed by EP U (see Section 2.6.1.2.1). Additionally, piping dynamic loads due to SRV discharge at EPU conditions are bounded by those used in the existing analyses. These effects have been eval uated for the RCPB portion of the RPV head vent line, SRV discharge piping and RWCU piping, as required. These systems were previously evaluated for compliance with the ASME Code stress criteria as required. Because none of these piping systems experience any significant change in operating conditions, they are all acceptable as currently designed. Therefore, GGNS meets all CLTR dispositions for RCPB piping. 2.2.2.2.2 Balance-of-Plant Piping Systems The BOP piping systems evaluation consists of a number of piping subsystems that move fluid through systems outside the RCPB piping. GGNS meets all CLTR dispositions. The topics considered in this section are: Topic CLTR Disposition GGNS Result Structural Evaluation for Unaffected Safety-Related Piping [[ Meets CLTR Disposition Structural Evaluation for Affected Safety-Related

Piping Meets CLTR Disposition Structural Evaluation for Unaffected Non-Safety

Related Piping

Meets CLTR Disposition

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2-55 Topic CLTR Disposition GGNS Result Structural Evaluation for Affected Non-Safety Related Piping

     ]] Meets CLTR Disposition 2.2.2.2.2.1 Structural Evaluation for Unaffected BOP Piping As stated in Section 3.5.2 of the CLTR, the flow, pressure, temperature, and mechanical loading for some BOP piping systems do not increase for EPU. Consequently, there is no change in stress evaluations, and these BOP piping systems meet all CLTR dispositions. The following BOP piping system designs at GGNS were confirmed to be unaffected by EPU conditions because either the flow, temperature, pressure, or other mechanical loads do not change in the system for EPU or the change is insignificant and has no effect on the piping system design:  Auxiliary Steam Piping  Circulating Water Piping   Component Cooling Water (CCW) Piping   Condensate and Refueling Water Storage and Transfer Piping   Condensate Cleanup Piping   Condenser Air Removal Piping   CRD Piping, Excluding CRD Insert, Withdraw and Sensing Lines Inside Containment   MSR Relief Valve Discharge Piping   DW Chilled Water Piping   Fuel Pool Cooling (FPC) and Cleanup Piping   HWC Piping   Liquid Radwaste Piping   MS Drain and MSIVs Drain Piping (Outside Containment)   Off Gas Piping NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-56 Plant Chilled Water Piping Plant Service Water (PSW) Piping Post Accident Sampling Piping Process Sampling Piping RWCU Piping Beyond the First Anchor Standby Liquid Control (SLC) Piping (Outside Containment) Standby Service Water (SSW) Piping Turbine Building Cooling Water (TBCW) Piping SRVDL Piping Beyond the First Anchor to the Quenchers 2.2.2.2.2.2 Structural Evaluation for Affected BOP Piping As stated in Section 3.5.2 of the CLTR, the FW and MS piping, including the associated branch piping, will experience an increase in the flow, pressure, and/or temperature resulting in an increase in stress. The GGNS piping systems determined to be affected by EPU operation include: High Pressure Core Spray (HPCS) Piping Beyond the First Anchor Low Pressure Core Spray (LPCS) Piping Beyond the First Anchor RCIC Piping Beyond the First Anchor RHR Piping Beyond the First Anchor Containment Spray Piping MS and Reheat Piping (Outside Containment) Extraction Steam Piping FW Piping (Outside Containment) Condensate Piping Moisture Separator Reheater Vents and Drains Piping NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-57 FW Heater Vents and Drains Piping Sealing Steam Piping For those systems with analyses, the maximum stre ss level analysis results were reviewed based on specific increases in temperature, pressure a nd flow rate (see Tables 2.2-5a through 2.2-5h and Table 2.2-6). These piping systems have been eval uated for the appropriate code criteria for the EPU conditions based on the design margins between actual stresses and code limits in the original design. All piping stresses and support loads have been found to be below the code allowable limits of the present code of record. The code of record for non-safety rela ted piping is ANSI B31.1 Power Piping Code, 1973 Edition. The code of record for Class 2 and 3 piping is ASME B&PV Code - Section III, Division I, 1974 Edition with addenda through Summer 1975, with some exceptions and use of the 1977 Edition with addenda through Winter 1979 and the 1980 Edition with addenda through Winter 1981. For those systems that do not require a detailed analysis, pipe routing and flexibility was determined to remain acceptable. A review was performed of postulated high ener gy pipe break locations in accordance with the requirements of the original license basis me thodology. As a result of this review, no new postulated break locations were identified. Details regarding analyses pertaining to the dynamic effects of high-energy piping failures outside containment are provided in Section 2.2.1 and the environmental effects of piping failures outside containment are discussed in Section 2.5.1.3.

Main Steam and Associated Piping System Evaluation The MS piping system outside containment was evaluated for compliance with all codes, standards, and criteria in the CLTP GGNS design basis, incl uding the effects of EPU on piping stresses, piping supports, and the associated building structure, equipment nozzles, pipe break postulation, flanges, and valves.

Because the effect of EPU on MS piping pressures and temperatures outside containment is minimal, there was no effect on the analyses for these parameters. The increase in MS flow results in increased transient forces from the TSV and MSIV closure. The TSVC tr ansient load is one of the individual loads experiencing the most significan t increase for the qualification of MS piping and supports at EPU conditions. For MS piping outside containment, the factors in Table 2.2-5b were derived using NRC approved GEH methodology detailed in the CLTR. This analysis determined that the maximum increase in piping and support load due to TS VC was 38.8%. This factor was conservatively applied to the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-58 CLTP stress and support loads to determine EPU loads. Application of these factors resulted in MS piping and supports exceeding code allowable limits. Therefore, detailed steam hammer forcing functions were developed and piping stress analysis was performed to determine more realistic loads at EPU for the MS system outside containment. This analysis resulted in all MS piping and pipe supports outside containment meeting all code criteria. Pipe Stresses

The pressure of the MS piping is unchanged fo r the EPU. There is a slight change in temperature for MS piping outside containment. The results of the EPU stress analysis demonstrate that the original piping design has sufficient margin between calculated stresses and code allowable limits in the Class 2 and non-safety related portions of the MS piping (see

Table 2.2-5b) to justify operation at EPU conditions. Pipe Supports

The pipe supports and turbine nozzles for the MS piping system outside containment were evaluated based on the loads developed in the MS piping stress analysis for the increased loading associated with the TSVC transients at EPU conditions. The evaluations demonstrate that all supports and turbine nozzles have adequate design margin to accommodate the increased loads and movements resulting from EPU. Based on existing margins available for the outside containment MS piping supports, it was conclude d that EPU does not result in reactions on

existing structures in excess of the current design capacity. Therefore, the MS and associated piping meet all CLTR dispositions. Feedwater System Evaluation Operation at EPU conditions increases stresses on piping and piping system components due to slightly higher operating temperatures and pressures and due to an approximately 15.9% increase in flow rates internal to the pipes. Higher FW operating pressures result from the higher head loss associated with a higher FW flow rate. The FW piping systems outside containment have been evaluated for the appropriate code crite ria for the EPU conditions based on the design margins between actual stresses and applicable code limits. All piping is below the code

allowable of the present code of record. No new postulated pipe break locations were identified. The FW stress analysis results for the non-Class 1 piping that interfaces with the Class 1 piping outside containment are provided in Table 2.2-5d. Pipe Stresses Because the FW system piping pressures and temperatures increase slightly due to EPU, the effects of these parameters on the existing analyses were evaluated. Seismic inertia loads, and seismic building displacement loads are not aff ected by EPU, thus, there is no effect on the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-59 analyses for these load cases. Other external loading conditions, such as chugging and CO also are not changed by EPU. AP has no effect on piping outside containment. It is bounded by the CLTP design basis AP loads and also by other mechanical loading (e.g., SRV discharge loads). For the BOP FW piping outside containment, there is no FW system fluid transient analysis in the existing design basis analysis, so the increase in FW system flow has no effect on the original analysis. The FW temperature and pressure changes are insignificant relative to piping design for EPU. The flow change does not affect the FW piping design system because water hammer transient was not a design load in the original stress anal yses. Therefore, the current licensing basis for the FW system complies with the code of record for the effect of thermal expansion displacement on the piping snubbers, hangers, and st ruts. Piping interfaces with penetrations, flanges, and valves also remain valid per current licensing basis. A review was also performed of postulated high energy pipe break locations in accordance with the requirements of the original license basis methodology. As a result of this review, no new postulated break locations were identified. Based on existing margins available for the FW piping, it was concluded th at EPU does not have an adverse effect on the FW piping design.

Pipe Supports

Operation at EPU conditions increases the pipe support loadings due to increases in the temperature of the affected piping systems (s ee Table 2.2-5d). The loading increases of the pipe supports were reviewed to determine if there is sufficient support capacity margin to accommodate the increased loadings. This review shows that support loads under EPU conditions are in compliance with appropriate code criteria. Additionally, thermal load changes were reviewed for their effect on existing equipment nozzle loading. This review shows that loading on affected nozzles remains within vendor specified values. Therefore, the FW system meets all CLTR dispositions.

Other Piping Evaluation A number of piping subsystems located inside containment move fluid through systems outside the RCPB piping. The flow, pressure, temperature, and mechanical loading for some of these piping systems do not increase for EPU. Conseque ntly, there is no change in stress and fatigue evaluations, and these piping systems meet all CLTR dispositions. Large bore and small bore piping and supports not addressed in Section 3.5.1 of the CLTR were evaluated for acceptability at EPU conditions. The evaluation of the piping and supports was performed in a manner similar to the evaluation of RCPB piping systems and supports using NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-60 applicable code equations. The original codes of record (as referenced in the appropriate calculations), code AVs, and analytical techniques were used, and no new assumptions were introduced. An evaluation of those GGNS piping systems where the loads and temperatures used in the CLTP analyses do not remain bounding of EPU containment hydrodynamic loads and short/long term temperatures was performed. The DBA LOCA dynamic loads, including the pool swell loads, vent thrust loads, CO loads and chugging loads were originally defined and evaluated for GGNS. The structures attached to the containment (WW) such as piping systems, vent penetrations, and valves are based on the DBA LOCA hydrodynamic loads. For EPU conditions, the DBA LOCA containment (WW) re sponse loads were evaluated and found to be unchanged by EPU (see Section 2.6.1.2.1) and thus , there are no resulting effects on the containment (WW) attached structures. EPU short/long term SP temperatures are evaluated in Section 2.6 and reported in Table 2.6-1. It was confirmed that the existing piping and pipe support analyses contain adequate margin to accommodate the effects of these small changes in SP temperatures for the following piping systems: RHR LPCI Lines Containment Spray Lines HPCS/Injection Lines (Beyond the Closed Valve) LPCS RCIC (Water Segment) As previously noted, the nominal operating pressure and temperature of the reactor are not changed by EPU. Aside from MS and FW, no other system connected to the RCPB experiences an increased flow rate at EPU conditions. Only minor changes to fluid conditions are experienced by these systems due to higher steam flow from the reactor and slightly higher FW operating pressure and the subsequent change in fluid conditions throughout the BOP systems. Other external loading conditions, such as c hugging, and CO also are not changed by EPU (see Section 2.6.1.2.1). Additionally, piping dynamic load s due to SRV discharge at EPU conditions are bounded by those used in the existing anal yses. AP has no eff ect on piping outside containment. It is bounded by the CLTP design basis AP loads and also by other mechanical

loading (e.g., SRV discharge loads). The piping and pipe supports of the other systems affected by EPU loading increases were reviewed to determine if there is sufficient capacity margin to accommodate the increased loadings. This review shows that, in most cases, piping stress and support loads under EPU NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-61 conditions are in compliance with appropriate code of record criteria. Additionally, thermal load changes were reviewed for their effect on existing equipment nozzle loading. The evaluations (see Tables 2.2-5a through 2.2-5h and 2.2-6) demonstrate that all piping, supports and equipment nozzles have adequate design margin to accommodate the increased loads and movements resulting from EPU. 2.2.2.3 Reactor Pressure Vessel Structural Evaluation 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 DW. The RPV also provides structural support for the reactor core and internals. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Reactor Vessel Structural Evaluation (Components Not Significantly Affected) [[

     ]] Meets CLTR Disposition As stated in Section 3.2.2 of the CLTR, for most RPV components, the flow, temperature, RIPDs and other mechanical loads do not increase. Consequently, there is no change in the stress or fatigue for these components. As discussed in Section 2.2.3.1.2, the steam dryer is being replaced. Therefore, the structural evaluation of the steam dryer support bracket, hol d down bracket, and guide rod bracket are not discussed in this section. These components were evaluated as part of the replacement steam dryer evaluation. Results from the structural evaluation of the steam dryer support bracket, steam dryer hold down bracket, and the guide rod bracket are provided in Attachment 11 to the EPU License Amendment Request (LAR).

[[

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2-62

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2-63

     ]] Certain reactor vessel components are dispositione d without detailed structural analysis. For components with no increase in flow, temperature, RIPD, or other mechanical loads, no further evaluation is required. The following components are confirmed to be cons istent with the dispositions provided in the CLTR (Reference 1), ELTR1 (Reference 2), and ELTR2 (Reference 4) for GGNS:  Main Closure Flange  RPV Steam Water Interface  Liquid Control - P Nozzle  CRD Penetration  CRD Housing  In-Core Penetration  In-Core Housing  FW Sparger Brackets  Main Shell  IRM / SRM / Dry Tube  Jet Pump Instrumentation Penetration Seal  PRD Dry Tube  Vent and Head Spray  Flange Closure Studs  Support Skirt  Shroud Support  CS Nozzle  Vibration Instrumentation Nozzle  RHR-LPCI Nozzle  CRD HSR Nozzle NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-64 Top Head Nozzles Refueling Bellows Stabilizer Bracket CS Bracket Top Head Lifting Lug Jet Pump Instrumentation Nozzle Drain Nozzle Water Level Instrumentation Nozzle Therefore these components are considered acceptable for EPU based on the EPU evaluation methodology. Jet Pump Riser Pads, Surveillance Brackets, a nd 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 EPU. No ASME Section XI flawed components were identified at GGNS. [[

     ]] The effect of EPU was evaluated to ensure that the reactor vessel components continue to comply with the existing structural requireme nts of the ASME B&PV Code. For the OLTP components under consideration, ASME B&PV 1971 Code with Addenda up to and including Winter 1972 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 physically modified since the original construction of GGNS:   FW Nozzle: This component was modi fied and the governing code for the evaluation/modification is the ASME B&

PV Code, Section III, 1974 Edition with Addenda up to and including Summer 1976. Recirculation Inlet Nozzle: This component was modified and the governing code for the modification is the ASME B&PV Code, S ection III, 1974 Edition with Addenda up to and including Summer 1976. The modification is to also satisfy the requirements of

ASME B&PV Code, Section III, 1971 Edition with Addenda up to and including Winter 1972. CRD HSR Nozzle: This component was m odified and the governing code for the evaluation/modification is the ASME B& PV Code, Section III, 1974 Edition with Addenda up to and including Winter 1975. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-65 In-Core Penetration: This component was modified and the governing code for the evaluation/modification is the ASME B& PV Code, Section III, 1971 Edition with Addenda up to and including Winter 1973. In-Core Housing: This component was modified and the governing code for the modification is the ASME B&PV Code, S ection III, 1971 Edition thru 1974 Edition, with Addenda up to and including Winter 1976. IRM / SRM / Dry Tube: This component was modified and the governing code for the modification is the ASME B&PV Code, S ection III, 1971 Edition with Addenda up to and including Summer 1973. The modification is to also satisfy the requirements of

ASME B&PV Code, Section III, 1977 Edition with Addenda up to and including Summer 1977. Jet Pump Instrumentation Penetration Seal: This component was modified and the governing code for the evaluation/modificati on is the ASME B&PV Code, Section III, 1974 Edition. PRD Dry Tube: ASME B&PV Code, Secti on III, 1974 Edition with Addenda up to and including Winter 1974 and ASME B&PV Code, Section III, 1977 Edition with Addenda up to and including Summer 1977. Vent and Head Spray: ASME B&PV Code, Section III, 1974 Edition with Addenda up to and including Winter 1975 and ASME B&PV Code, Section III, 1974 Edition with Addenda up to and including Summer 1976. New stresses are determined by scaling the "o riginal" stresses based on the EPU conditions (pressure, temperature, and flow). The analyses were performed for the design, the normal and upset, and the emergency and faulted conditions. If there is an increase in AP, jet reaction (JR), pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency, and faulted conditions. Design Conditions Because there are no changes in the design c onditions due to EPU, the design stresses are unchanged and the code requirements are met.

Normal and Upset Conditions The reactor coolant temperature and flows, with the exception of the core flow, FW flow, recirculation flow, and MS flow, at EPU conditions are only slightly changed from those at current rated conditions. Evaluations were performed at conditions that bound the change in operating conditions. The evaluation type is mainly reconciliation of the stresses and usage factors to reflect EPU conditions. A primary plus secondary stress analysis was performed NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-66 showing EPU stresses still meet the requirements of the ASME code, Section III, Subsection NB. The fatigue usage was evaluated for the limiting location of components with a [[

                                    ]]. The GGNS fatigue analysis results for the limiting components are provided in Table 2.2-7. The GGNS analysis results for EPU show that components requiring evaluation do not exceed ASME code AVs and no further analysis is required before these components meet their ASME code requirements.

Emergency and Faulted Conditions The stresses due to emergency and faulted conditions are based on loads such as peak dome pressure, which are unchanged for EPU. Therefore, code requirements are met for all RPV components under emergency and faulted conditions. Therefore, the reactor vessel meets all CLTR dispositions.

Conclusion Entergy has reviewed the evaluations related to the structural integrity of pressure-retaining components and their supports. For the reasons set forth above, Entergy concludes that the effects of the proposed EPU on these component s and their supports have been adequately addressed. Based on the above, Entergy further concludes that pressure-retaining components and their supports will continue to meet the requirements of 10 CFR 50.55a, GDC-1, GDC-2, GDC-4, GDC-14, and GDC-15 following implementa tion of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with re spect to the structural integrity of the pressure-retaining components and their supports.

2.2.3 Reactor

Pressure Vessel Internals and Core Supports Regulatory Evaluation RPV internals consist of structural and mechanical elements inside the reactor vessel, including core support structures. Entergy reviewed the effects of the proposed EPU on the design input parameters and the design-basis loads and load combinations for the reactor internals for normal operation, upset, emergency, and faulted conditions. These include pressure differences and thermal effects for normal operation, transient pr essure loads associated with DBLOCAs, and design transient occurrences. The review covere d: (1) analyses of FIV for safety-related and non-safety related reactor internal components; and (2) analytical methodologies, assumptions, ASME Code editions, and computer programs us ed for these analyses. The review also compared the resulting stresses and CUFs against the corresponding Code-allowable limits. The

regulatory acceptance criteria are based on: (1) 10 CFR 50.55a and GDC-1, insofar as they require SSCs important to safety be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-2, insofar as it requires SSCs important to safety be designed to withstand NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-67 the effects of earthquakes combined with the effects of normal or accident conditions; (3) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; and (4) GDC-10, insofar as it requires the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of AOOs. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The reactor internals are described in UFSA R Sections 3.9.3, "ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures," 3.9.5, "Reactor Pressure Vessel Internals," 4.1.2, "Reactor Internal Components," and 4.5.2, "Reactor Internal Materials."

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Sections 3.3 and 3.4 of the CLTR address the effect of CPPU on Reactor Vessel and Reactor Internals, respectively. The results of this evaluation are described below. 2.2.3.1 FIV Influence on Reactor Internal Components The FIV evaluation of the RPV internals addresse s the influence of an increase in flow during EPU. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Structural Evaluation of Core Flow Dependent RPV

Internals [[ Meets CLTR Disposition Structural Evaluation of Other RPV Internals

     ]] Meets CLTR Disposition 2.2.3.1.1 Structural Evaluation of Core Flow Dependent RPV Internals As stated in Section 3.4.2 of the CLTR, EPU cause s an increase in reactor coolant quality and an increase in FW, steam and recirculation pump drive flow.

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2-68 [[ ]] The core flow dependent RPV internals (in-core guide tube and control rod guide tube (CRGT) components) meet all dispositions provided in the CLTR [[

     ]] 2.2.3.1.2 Structural Evaluation of Other RPV Internals As stated in Section 3.4.2 of the CLTR, EPU cause s an increase in reactor coolant quality and an increase in FW, steam and recirculation pump drive flow. The required RPV internals vibration assessment of the other RPV internals is described in the CLTR. EPU operation increases the steam production in the core, resulting in an increase in the core pressure drop.  [[                                                                                                                                               
                                                                                ]]  The increase in power may increase the level of reactor internals vibration. Analyses were performed to evaluate the effects of FIV on the

reactor internals at EPU conditions. This ev aluation used a bounding reactor power of 102% of 4,408 MWt and 105% of rated core flow. [[

                        ]]  For components requiring an evaluation but not instrumented in the prototype plant, [[                                                                                                                                                                         
                                            ]]  The expected vibration levels for EPU were estimated by extrapolating the vibration data recorded in the prototype plant or similar plants and based on GEH BWR OE.

These expected vibration levels were then compared with the established vibration acceptance limits. The following components were evaluated: a) Shroud b) Shroud Head and Separator Assembly c) Jet Pumps d) LPCI Coupling e) FW Sparger f) Jet Pump Sensing Lines g) In-Core Guide Tubes and Control Rod Guide Tubes h) Fuel Channels i) Guide Rods NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-69 j) RPV Top Head Spare Instrument Nozzle k) RPV Top Head Vent Nozzle l) RPV Head Spray Pipe and Head Spray Nozzle m) Core Spray Piping and Sparger n) Shroud Head Stud Assembly Modification o) Steam Dryer The results of the vibration evaluation show th at continuous operation at a reactor power of 4,408 MWt and 105% of rated core flow does not result in any detrimental effects on the safety-related reactor internal components. FIV of critical reactor internal components at EPU is

predicted based on the available startup test data at [[

     ]]  Vibration amplitudes are also adjusted by a [[

          ]]  The extrapolated vibration amplitude response under EPU conditions is compared with the acceptance criterion in the percent criteria for each mode. The percentages of the criteria for all modes are cumulative as total percent criteria.  [[

                  ]]  The summary of the evaluation methods and results for the following components are: Shroud For the shroud, the measured vibrations were extrapolated to the EPU conditions. Maximum stresses during CLTP are less than 4,869 psi and will remain well within acceptance criteria during EPU. The calculated maximum stress at EPU is less than 60% of the acceptance criteria or less than 6,000 psi at EPU conditions.

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2-70 Shroud Head and Separator Assembly For the shroud head, [[

     ]]  Jet Pumps Results from strain gage measurements [[

     ]]  Low Pressure Coolant Injection Coupling The vibration amplitude response at EPU conditions is calculated; the stress is equivalent to 2,000 psi or 20% of the AV; therefore, the vibration amplitude is acceptable at EPU conditions for all flow conditions.

FW Sparger The FW sparger in the GGNS reactor vessel is of the improved triple thermal sleeve design. [[

                                                                                                                                                     ]]  Therefore, the GGNS FW sparger is acceptable under FIV for EPU conditions.

Jet Pump Sensing Lines Resonance of the recirculation pump vane passing frequency (VPF) with the natural frequency of the jet pump sensing line (JPSL) is the cause of the JPSL stress. [[

     ]] In-Core Guide Tubes and Control Rod Guide Tubes The FIVs of these components are not affected by EPU as they are a function of the core flow, and the core flow does not change during EPU.

Hence, there will be no increase in FIV stresses due to EPU. Maximum stresses during OLTP are well within the acceptance criteria and will remain approximately the same at EPU conditions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-71 Fuel Channels GGNS uses GE14 and GNF2 fuel bundle assemblies. [[

                                                                              ]]  Therefore, the GGNS fuel assemblies are acceptable under FIV for EPU conditions.

Guide Rods The guide rod is subjected to cross flow vibra tion and the ASME Code Section III procedure for lock-in phenomena is used. [[

     ]]  RPV Top Head Spare Instrument Nozzle
[[

     ]]  Thus, the stress due to FIV at EPU conditions is deemed to be negligible.

RPV Top Head Vent Nozzle

[[

                     ]]  Therefore, the top head vent nozzle will be structurally adequate from a vibration viewpoint at EPU conditions.

RPV Head Spray Pipe and Head Spray Nozzle

[[

                                                                                                                                    ]]  Thus, the stress due to FIV at EPU conditions is deemed to be negligible.

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2-72 Core Spray Piping and Sparger

[[

                                                                                                              ]]  Therefore, the FIV stress due to vortex shedding at EPU conditions is minimal.

Shroud Head Stud Assembly Modification

[[

                                                                                                                                                                   ]]  Therefore, the FIV stress due to vortex shedding at EPU conditions is minimal.

Summary During EPU, the components in the core region and components such as the CS line are primarily affected by the core flow. Components in the annulus region such as the jet pump are primarily affected by the recirculation pump driv e flow and core flow. For EPU conditions at GGNS, there is no change in the maximum licensed core flow in comparison to the CLTP condition, resulting in negligible changes in FIV on the components in the annular and core

regions. The calculations for EPU conditions indicate that vibrations of all safety-related reactor internal components are within the GEH acceptance criteri

a. The analysis is conservative for the

following reasons: The GEH criteria of [[ ]] stress intensity is less than the ASME Section III Code criteria of 13,600 psi; The modes are [[

     ]]; and  The maximum vibration amplitude in each mode is used in the absolute sum process, whereas in reality the peak vibration amplitudes are unlikely to occur at the same time.

Based on the above, the FIV effect on reactor internal components meets all CLTR dispositions. Steam Dryer During EPU, the stress levels in components in the upper zone of the reactor such as the moisture separators and steam dryer are affected by the increased steam flow. As a result, the structural integrity of the steam dryer may be affected by operating at EPU conditions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-73 The steam dryer is a non-safety related component. Recent uprate experience indicates that FIV at EPU conditions may lead to high cycle fatigue failure of some dryer components. A quantitative evaluation of the existing GGNS steam dryer has been performed. Based on that evaluation, a replacement dryer of a proven, robust design is being installed at GGNS prior to EPU. Analyses were performed to evaluate the acceptability of the replacement steam dryer under normal, upset, emergency and faulted conditions. Re sults of these analyses are within acceptable stress limits. A specific analysis was performed to quantify the FIV stress levels for the replacement steam dryer at GGNS. This analysis identified EPU stresses on the replacement steam dryer and was performed in accordance with RG 1.20, Revision 3 (Reference 30). Attachment 11 of the EPU LAR describes in detail the methods and results of the analysis. The results of this analysis demonstrated that the stress intensity is less than half the ASME code criteria of 13,600 psi (i.e., a greater than 2.0 Safety Factor to the ASME code criteria). Based on the above, it is concluded that the replacement steam dryer stress levels are within acceptable limits at EPU under all loading conditions. 2.2.3.2 Reactor Internals The RPV internals consist of the Core Support Structure components and Non-Core Support Structure components. GGNS meets all CLTR di spositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Reactor Internal Pressure Differences [[ Meets CLTR Disposition Reactor Internals Structural Evaluation

Meets CLTR Disposition Steam Dryer Separator Performance

     ]] Meets CLTR Disposition 2.2.3.2.1 Reactor Internal Pressure Differences As stated in Section 3.3.1 of the CLTR, EPU re sults in higher pressure differences across internals due to higher core exit steam flow.

The increase in core average power alone would result in higher core loads and reactor internals pressure differences (RIPDs) due to the higher core exit steam quality. The original acoustic and flow induced loads (FILs), following a postulated recirculation line break (RLB), we re also updated in accordance with current methodology. The acoustic loads are determined by a multi-dimensional method of NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-74 characteristics calculation performed specifically for the reactor vessel internals. This approach is widely used by the industry for predicting acoustic response. The FILs as a result of the RLB

are calculated using TRACG Model. The RIPDs are calculated for Normal (steady-stat e operation), Upset, and Faulted conditions for all major reactor internal components. For minor components (jet pump sensing lines, dryer/separator guide rods, and in-core guide tube braces), the pressure drops during Normal, Upset, and Faulted conditions are minimal and represent insignificant portions of the RIPDs because of the small surface area. They are not affected by EPU and are not evaluated for EPU. Tables 2.2-8 through 2.2-10 compare the RIPDs across the major reactor internal components during current and EPU operation in the Normal, Upset, and Faulted conditions, respectively. The core plate DP slightly increases from CLTP to EPU. EPU is evaluated on the basis of operation at the same dome pressure but higher core power and steam flow. Therefore, pressures up-stream from the dome, such as above th e core plate region, should increase slightly (consistent with higher steam flow resistance of approximately 1 psi in the steam dryers and separators), additionally, pressures downstream from the dome, such as the turbine inlet region, should decrease slightly (consistent with higher steam flow resistance in the steam lines).

The GGNS EPU core will include GNF2 fuel. The RIPDs are calculated for both GE14 and GNF2 fuel; the results for the GE14 fuel are demonstrated to be bounding. 2.2.3.2.2 Reactor Internals Structural Evaluation (Non-FIV) As stated in Section 3.3.2 of the CLTR, the typical loads considered in EPU structural evaluation of the internals include: deadweight, RIPDs, seismic loads, SRV, LOCA, AP/JR loads, thermal loads, flow loads, acoustic and flow induced loads due to RLB, and fuel lift loads, consistent with the design basis. The RPV internals consist of the core support structure components and non-core support structure components. The core support structure components are ASME Code components. The requirements of the ASME Code are used in their design/analysis. The non-core support structure components are not ASME Code components; however, the requirements of the ASME Code are used as guidelines in their design/an alysis. The evaluations/stress reconciliation in support of the EPU was performed consistent with the design basis analysis of the components. The reactor internal components evaluated are: Core Support Components Shroud Support Shroud NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-75 Core Plate Top Guide/Grid CRDH Control Rod Guide Tube Orificed Fuel Support (OFS) Peripheral Fuel Support (PFS) Non-Core Support Components Fuel Channel Steam Dryer FW Sparger Jet Pump Assembly CS Line and Sparger Access Hole Cover (AHC) Shroud Head and Steam Separator Assembly In-Core Housing and Guide Tube (ICHGT) Vessel Head Cooling Spray Nozzle (VHCSN) LPCI Coupling The original configurations of the RPV internal s are considered in the EPU evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation. The loads considered in the evaluation of the RPV internals include RIPDs, deadweight, seismic loads, hydrodynamic loads such as SRV, LOCA, AP /JR loads, acoustic loads and FILs due to RLB, fuel lift loads, flow loads, and thermal loads. EPU loads are compared to those used in the exis ting design basis analysis. If the EPU loads are bounded by the design basis loads for the RPV intern als, the existing design basis qualification is valid for EPU. In such cases, no further evaluations are required or performed. For RPV NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-76 internals exhibiting increases in loads, the method of analysis is to linearly scale the critical/governing stresses based on the increase in loads as applicable, and compare the resulting stresses against the allowable stress limits, consistent with the design basis. Tables 2.2-11 and 2.2-12 present the governing stresse s and fatigue values for the various reactor internal components affected by EPU. All stre sses and fatigue usage factors are within the design basis ASME Code allowable limits, and the RPV internal components are demonstrated to be structurally qualified for operation at EPU conditions.

The following reactor vessel internals are evaluate d for the effects of changes in loads due to EPU. a) Shroud Support: Quantitative analysis of the shroud support was performed for the changes in loads associated with EPU cond itions. The loads applicable to the shroud support are: deadweight, seismic, RIPD, SRV, LOCA, AP/JR, acoustic load and FIL due to RLB, and fuel lift loads. Seismic loads remain unchanged. RIPDs remain bounded by original design basis (ODB) values. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Acoustic load and FIL due to RLB increased. Fuel lift loads remain unchanged. Shroud support stresses were reconciled for EPU loads to show that the stresses remain within the allowable limits. Therefore, the shroud support, in its original configuration, is qualified for EPU conditions. b) Shroud: Quantitative analysis of the shroud was performed for the changes in loads associated with EPU conditions. The loads applicable to the shroud are: deadweight, seismic, RIPD, SRV, LOCA, AP/JR, acoustic load and FIL due to RLB, and fuel lift loads. Seismic loads remain unchanged. RIPDs increased. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Acoustic load and FIL due to RLB increased. Fuel lift loads remain unchanged. Shroud stresses were reconciled for EPU loads to show that the stresses remain within the allowable limits. Therefore, the shroud, in its original configuration, is qualified for EPU conditions. c) Core Plate: Quantitative analysis of the core plate was performed for the changes in loads associated with EPU conditions. The loads applicable to the core plate are: deadweight, seismic, RIPD, SRV, LOCA, AP/JR, and fuel lift loads. Seismic loads remain unchanged. RIPDs increased. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Fuel lift loads remain unchanged. Core plate stresses were reconciled for EPU loads to show that the stresses remain within the allowable limits. Therefore, the core plate, in its original configuration, is qualified for EPU conditions. d) Top Guide/Grid: Quantitative analysis of the top guide/grid was performed for the changes in loads associated with EPU cond itions. The loads applicable to the top guide/grid are: deadweight, seismic, RIPD, SRV, LOCA, AP/JR, and fuel lift loads. Seismic loads remain unchanged. RIPDs increased. SRV, LOCA, and AP/JR loads NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-77 remain bounded by CLTP. Fuel lift loads remain unchanged. Top guide/grid stresses were reconciled for EPU loads to show that the stresses remain within the allowable limits. Therefore, the top guide/grid, in its original configuration, is qualified for EPU conditions. e) Control Rod Drive Housing: Quantitative analysis of the CRDH was performed for the changes in loads associated with EPU conditi ons. The loads applicable to the CRDH are: deadweight, seismic, SRV, LOCA, AP/JR, fuel lift, and flow loads. Seismic loads remain unchanged. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Fuel lift loads and flow loads remain unchanged. Hence, the CLTP stresses remain valid for EPU conditions. Therefore, the CRDH, in its or iginal configuration, is qualified for EPU conditions. f) Control Rod Guide Tube: Quantitative analysis of the CRGT was performed for the changes in loads associated with EPU conditions. The loads applicable to the CRGT are: deadweight, seismic, RIPD, SRV, LOCA, AP/JR, fuel lift loads and flow loads. Seismic loads remain unchanged. RIPDs increased. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Fuel lift loads and flow loads remain unchanged. The CRGT stresses were reconciled for EPU loads to show that the stresses remain within the allowable limits. Therefore, the CRGT, in its orig inal configuration, is qualified for EPU conditions. g) Orificed Fuel Support: Qualitative analysis of the OFS was performed for the changes in loads associated with EPU conditions. The loads applicable to the OFS are: deadweight, seismic, RIPD, SRV, LOCA, AP/JR, and fuel lift loads. All applicable loads for EPU are bounded by the design basis values. The CLTP stresses remain valid for

EPU conditions. Therefore, the OFS, in its original configuration, is qualified for EPU conditions. h) Peripheral Fuel Support: Qualitative analysis of the PFS was performed for the changes in loads associated with EPU conditi ons. The loads applicable to the PFS are: deadweight, seismic, RIPD, SRV, LOCA, AP/JR, and fuel lift loads. All applicable loads for EPU are bounded by the design basis values. The CLTP stresses remain valid for

EPU conditions. Therefore, the PFS, in its original configuration, is qualified for EPU conditions. i) Fuel Channel: The RIPDs for the fuel channel are within the respective design limits for the channel. Additionally, the channel/control blade interference is not affected by EPU. Therefore, the fuel channel, in its original configuration, is qualified for EPU conditions. j) Steam Dryer: The steam dryer is evaluated in Attachment 11 to the EPU LAR.

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2-78 k) Feedwater Sparger: Quantitative analysis of the FW sparger was performed for the changes in loads associated with EPU conditions. The loads applicable to the FW sparger are: deadweight, seismic, thermal loads, SRV, LOCA, AP/JR, and flow loads. Seismic loads and thermal loads remain unchanged. SRV, LOCA and AP/JR loads remain bounded by CLTP. Flow loads increased. FW sparger stresses were reconciled for EPU loads to show that the stresses remain within the allowable limits. Fatigue usage for EPU conditions is bounded by design basis value. Therefore, the FW sparger, in its original configuration, is qualified for EPU conditions. l) Jet Pump Assembly: Qualitative analysis of the jet pump assembly was performed for the changes in loads associated with EPU c onditions. The loads applicable to the jet pump assembly are: deadweight, seismic, RIPDs, thermal loads, SRV, LOCA, AP/JR, acoustic load and FIL due to RLB, and flow loads. Seismic loads remain unchanged. RIPDs remain bounded by ODB values. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Acoustic load and FIL due to RLB remain bounded by ODB values. Thermal loads and flow loads remain unchanged. Hence, the CLTP stresses remain valid for EPU conditions. Therefore, the jet pump assembly, in its original configuration, is qualified

for EPU conditions. m) Core Spray Line and Sparger Quantitative analysis of the CS line and sparger was performed for the changes in loads associated with EPU conditions. The loads applicable to the CS line and sparger are: deadweight, seismic, thermal loads, SRV, LOCA, AP/JR, and flow loads. Seismic loads and thermal loads remain unchanged. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Flow loads remain unchanged. Hence, the CLTP stresses remain valid for EPU conditions. Therefore, the CS line and sparger, in its original configuration, is qualified for EPU conditions. n) Access Hole Cover: Qualitative analysis of the AHC was performed for the changes in loads associated with EPU conditions. The lo ads applicable to the AHC are: deadweight, seismic, RIPDs, SRV, LOCA, AP/JR, and acoustic loads and FIL due to RLB. Seismic loads remain unchanged. RIPDs increased. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Acoustic loads and FIL due to RLB increased. All applicable loads are bounded by ODB values. The CLTP stresses remain va lid for EPU conditions. Therefore, the AHC, in its original configuration, is qualified for EPU conditions. o) Shroud Head and Separator Assembly: Quantitative analysis of the shroud head and separator assembly was performed for the changes in loads associated with EPU conditions. The loads applicable to the shroud head and separator assembly are: deadweight, seismic, RIPD, thermal, SRV, LOCA, and AP/JR loads. The deadweight change due to shroud head stud repair is negligible. Seismic loads and thermal loads remain unchanged. RIPDs remain bounded by the ODB values. SRV, LOCA, and AP/JR loads remain bounded by CLTP. Hence, the CLTP stresses remain valid for EPU conditions. Therefore, the shroud head and NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-79 separator assembly, in its permanently m odified configuration, is qualified for EPU conditions. p) In-Core Housing and Guide Tube

Quantitative analysis of the ICHGT was performed for the changes in loads associated with EPU c onditions. The loads applicable to the ICHGT are: deadweight, seismic, SRV, LOCA, AP/JR, and flow loads. SRV, LOCA, and AP/JR loads remain bounded by CLTP. All other applicable loads remain unaffected for EPU conditions. Hence, the CLTP stresses remain valid for EPU conditions. Therefore, the ICHGT, in its original configuration, is qualified for EPU conditions.

q) Vessel Head Cooling Spray Nozzle: Quantitative analysis of the VHCSN was performed for the changes in loads associated with EP U conditions. The loads applicable to the VHCSN are: deadweight, seismic, thermal, SRV, LOCA, and AP/JR loads. SRV, LOCA, and AP/JR loads remain bounded by CLTP. All other applicable loads remain unaffected for EPU conditions. Hence, the CLTP stresses rema in valid for EPU conditions. Therefore, the VHCSN, in its original configuration, is qualified for EPU conditions. r) LPCI Coupling: Quantitative analysis of the LPCI coupling was performed for the changes in loads associated with EPU conditions. The loads applicable to the LPCI coupling are: deadweight, seismic, RIPDs, SRV, LOCA, and AP/JR loads. Seismic loads remain unchanged. SRV, LOCA, and AP/JR loads remain bounded by CLTP. RIPDs are bounded by design basis values for upset and emergenc y conditions and increase by 4.8% for faulted condition. The LPCI coupling stresses were r econciled for EPU loads to show that the stresses remain within the allowable limits. Th erefore, the LPCI coupling, in its original configuration, is qualified for EPU conditions. 2.2.3.2.3 Steam Dryer/Separator Performance For GGNS, the EPU performance of the steam dryer/ separator was evaluated to ensure that the quality of the steam leaving the RPV continues to meet existing operational criteria at EPU

conditions. EPU 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 results of the evaluation demonstrated that the steam dryer/separator performance remains acceptable (e.g., moisture content 0.1 weight %) at EPU conditions. Conclusion Entergy has reviewed the structur al integrity of reactor internals and core supports and concludes the effects of the proposed EPU on the reactor internals and core supports have been adequately

addressed. Entergy further concludes the reactor internals and core supports will continue to meet the requirements of 10 CFR 50.55a, GDC-1, GDC-2, GDC-4, and GDC-10 following NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-80 implementation of the proposed EPU. Therefor e, Entergy finds the proposed EPU acceptable with respect to the design of the reactor internal and core supports. 2.2.4 Safety-Related Valves and Pumps Regulatory Evaluation Entergy's review included certain safety-related pumps and valves typically designated as Class 1, 2, or 3 under Section III of the ASME B&PV Code and within the scope of Section XI of the ASME B&PV Code and the ASME Operations and Maintenance (O&M) Code, as applicable. The review focused on the effects of the proposed EPU on the required functional performance of the valves and pumps. The review also covered any effects that the proposed EPU may have on the Entergy motor-operated valve (MOV) programs related to GL 89-10, GL 96-05, and GL 95-07. Entergy also evaluated the lessons learned from the MOV program and the application of those lessons learned to othe r safety-related power-operated valves. The regulatory acceptance criteria are based on: (1) GDC-1, insofar as it requires that SSCs important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-37, GDC-40, GDC-43, and GDC-46, insofar as they require the ECCS, the containment heat removal system, the containment atmospheric cleanup systems, and the cooling water system, respectively, be designed to permit appropriate periodic testing to ensure the leak-tight integrity and performance of their active components; (3) GDC-54, insofar as it requires piping systems penetrating containment be designed with the capability to periodically test the operability of the isolation valves to determine if valve leakage is within acceptable limits; and (4) 10 CFR 50.55a(f), insofar as it requires pumps and valves subject to that section must meet the inservice testing (IST) program requirements identified in that section. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. Safety-related valves and pumps are describe d in UFSAR Section Sections 3.9.3, "ASME Code Class 1, 2, and 3 Components, Component S upports, and Core Support Structures," and 3.9.6, "Inservice Testing of Pumps and Valves." NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-81 Technical Evaluation 2.2.4.1 Background Inservice Testing of Safety-Related Pumps and Valves The GGNS Pump and Valve IST Program, hereafter referred to as the IST Program, has been prepared to summarize the test program for certain pumps and valves pursuant to the requirements of the 10 CFR 50.55a(f), except wher e specific written relief has been granted by the NRC in accordance with 10 CFR 50.55a(f)(6)(i). GGNS TS Section 5.5.6, Inservice Testing Program, states that this program provides controls for IST of ASME Class 1, 2 and 3 components. The applicable Code for the IST Program is the ASME Operation and Maintenance (OM) Code-2001 Edition with addenda thr ough and including the ASME OMb Code-2003 Addenda. Containment Leakage Rate Testing Program Containment Leakage Rate Testing is addressed in UFSAR Section 6.2.6 and GGNS TS Section 5.5.12. The GGNS Containment Leakage Rate Testing Program implements testing requirements in accordance with 10 CFR 50 Appendix J, Option B, as modified by any approved exemptions, and criteria contained in RG 1.163, Performance-Based Containment Leak Test Program (dated September 1995) (Referen ce 31); NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appe ndix J (July 26, 1995) (Reference 32); and ANSI 56.8-1994, Containment System Leakage Testing Requirements (Reference 33). Tests that measure containment isolation valve leak rates (Type C tests) are performed using the CLTP calculated peak containment pressure (Pa) value of 11.5 psig. From the containment analysis at EPU conditions, the peak containment pressure is 11.9 psig for a LOCA (Section 2.6.1). Because the containment peak pressure for EPU is higher than the CLTP Pa, the leak rate testing requirements for containment isolation valves are affected by the

proposed EPU. As a result, the GGNS Containment Leakage Rate Program will be updated to incorporate the EPU Pa value. Pumps in the IST Program The scope of the IST Program is derived from the OM Code, ISTB requirements for the ASME Code Class 1, 2 and 3 pumps providing a safety-related function. Table 2.2-13 lists the systems with pumps in the IST Program.

Valves in the IST Program The scope of the IST Program is derived from the OM Code, ISTC requirements for the active and passive ASME Code Class 1, 2, and 3 valves. The IST program includes MOVs, NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-82 Air-Operated Valves (AOVs), Solenoid-Operated Valves (SOVs), check valves, and relief valves. Table 2.2-13 lists the systems with valves in the IST Program. Motor-Operated Valve Program The GGNS MOV Program implements the recommendations and requirements of GL 89-10, "Safety-Related Motor Operated Valve Testing and Surveillance" (Reference 34). The scope of the program also includes the requirements of GL 96-05, "Verification of Design-Basis Capability of Safety-Related Motor Operated Va lves" (Reference 35). Table 2.2-13 indicates the systems which contain MOV Program valves.

Air-Operated Valve Program The GGNS AOV Program activities include the following elements: (1) performance of design basis/functional reviews; (2) performance of diagnostic testing to ensure proper maintenance setup and confirm proper assembly performance; and (3) development and implementation of improved maintenance instructions and controls. The AOV Program va lve population includes: (1) valves within the IST program; (2) thermal generation significant valves; (3) trip critical/sensitive valves; and (4) GGNS Level 1 and 2 PRA risk significant valves. Table 2.2-13 indicates the systems which contain AOV Program valves. Lessons Learned GGNS IST, Containment Leak Rate, MOV, and AOV programs utilize the Entergy Corrective Action Program to evaluate and resolve non-conforming conditions identified during program performance. The process includes the provisi on for an extent of condition review which identifies the total population of items that have or may have the same problem as identified in the original condition report problem statement. The intent of the exte nt of condition review focuses on a determination of any potential effect on the operability or functionality of similar components or equipment. EN-LI-102 is the administrative procedure that implements the requirements of the corrective action process from 10 CFR 50 Appendix B. The GGNS AOV Program was developed utilizing lessons learned from the MOV Program. Elements that have been successful for the MOV Program contained in the AOV Program include a design basis review of AOV program valves, the establishment of a priority for potential testing based on the valve's risk significance, and the trending of valve test results.

2.2.4.2 Description of Analyses and Evaluations This section addresses the effect of EPU on the performance requirements of GGNS safety-related pumps and valves in the IST, MOV, and AOV programs. The discussion is organized by system or groups of systems and the respective program pumps and valves are discussed therein. Each system was analyzed to define any parameter changes such as pressure, NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-83 flow, and process and ambient temperatures resulting from implementation of EPU. The 480 VAC motor control center (MCC) minimum voltages supplied from off-site power are only marginally affected by EPU (0.51 VAC maximum voltage drop). This 0.11% voltage drop has a negligible effect on valve torque and will be incorporated into the affected MOV calculations. The 480 VAC minimum voltages supplied from on-site power and safety related DC sources are unaffected by EPU. Systems Not Significantly Affected by EPU: The Makeup Water Treatment; Floor and Equipment Drain; Service Air; Instrument Air; Suppression Pool Cleanup; and Domestic Water systems are addressed in Section 2.5.7, which concludes that EPU does not have any significant effect on these systems. The only safety function performed by valves in the following systems is containment isolation: DW Monitoring; Containment Personnel Air Locks; Containment Leak Rate Test; Containment and DW Instrument and Control; Condensate and Refueling Water Storage and Transfer System; Suspended Floor and Equipment Drain; Plant Chilled Water System; and DW Chilled Water System. Section 2.6.1.3 states that containment isolation capabilities are not adversely affected by the EPU. In addition, the Standby and HPCS Diesel Generator systems (Sections 2.3.3 and 2.5.6.1); Control Room Heating, Ventilating, and Air Conditioning (HVAC) System (Section 2.7.3); Fire Protection System (Section 2.5.1.4); MSIV Leakage Control System (LCS) (Section 2.5.2.4); and Feedwater Leakage Control (FWLC) System (Section 2.8.4.4.5) are not significantly affected by EPU. There are no system parameter changes and no modifications being made to these systems as a result of EPU. Therefore, the GGNS program valves in these systems are not affected by EPU. Emergency Core Cooling Systems / Reactor Core Isolation Cooling System The ECCS are comprised of the RHR, LPCS, and HPCS systems. Because EPU does not affect system operating pressures, flow rates, or pump head for the ECCS or RCIC system, potential effects to these systems are limited to those caused by the increase in the peak SP temperature following certain transients. Sections 2.8.4.3 (RCIC), 2.8.4.4 (RHR), and 2.8.5.6.2 (LPCS and HPCS) conclude that pump performance is acceptable for each of these systems. As a result of the higher peak SP temperatures, the temperatures of several rooms/areas were conservatively assumed to increase by up to 8 ûF. The effect of this ambient temperature increase to the ECCS valves is discussed in Section 2.2.4.3. No other ECCS program valves are affected by EPU. RCIC valves were not affected by an ambient temp erature increase; therefore, EPU has no effect on the performance characteristics or IST Plan requirements for the valves in the RCIC system. Power-Dependent HVAC The power-dependent HVAC systems with GGNS program components are the Containment Cooling, Auxiliary Building Ventilation, and the Fuel Handling Area Ventilation systems. The NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-84 EPU effect on these systems is addressed in Sections 2.7.5 and 2.7.6. No system modifications are required, and the GGNS program valves in these systems are not affected as a result of EPU. Power Conversion Systems The following systems contain non-safety related AOV program valves only (i.e., no MOV or IST program components): Main and Reheat Steam; Condensate; Feedwater; Condensate Cleanup; Heater Vents and Drains; Seal Steam and Drains; MSR Vents and Drains; Extraction Steam; Condenser Air Removal; Offgas; and Turbine Building Closed Cooling Water. Due to the increased power associated with EPU, the components in these systems are potentially affected by temperature, flow, and/or operating pressure changes. Detailed discussions are provided in Section 2.5.2.3, Turbine Gland Sealing System; Section 2.5.3.3, Reactor Auxiliary Cooling Water Systems; Section 2.5.4.1, Main Steam; Section 2.5.4.4, Condensate and Feedwater; and Section 2.5.5.1, Gaseous Waste Management Systems. Evaluations have determined that the current design bases of the AOV program valves are acceptable for the expected EPU effect on these systems. Station Service Water Systems The station service water systems (SWSs) are addressed in Section 2.5.3.2 and consist of the SSW, PSW, and PSW Radial Well systems. The only PSW Radial Well program components are relief valves at each radial well which provide overpressure protection for the safety related components in the PSW system. EPU has no effect on these relief valves. Because the heat removal requirements of the SSW and PSW systems increase at EPU conditions, several modifications are required to increase each system's heat removal capacity. These modifications and any required system flow changes do not affect the ability of the SSW pumps to perform their safety function. One SSW valve (1P41F113) experiences a potential ambient temperature increase as a result of EPU. The effect of this ambient temperature change is evaluated in Section 2.2.4.3. There is no effect on any other SSW or PSW program valve's ability to perform its safety function as a result

of EPU. Nuclear Boiler System The Nuclear Boiler System IST components include the MSIVs addressed in Section 2.2.2.1, other containment isolation valves addressed in Section 2.6.1.3, and the MS SRVs addressed in Section 2.8.4.2. Although these components experience parameter changes due to EPU, the ability of the Nuclear Boiler System program valves to perform their safety functions is not affected by EPU. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-85 Reactor Recirculation System The RRS changes due to EPU are addressed in Section 2.8.4.6. No modifications are being made to the system as a result of EPU. The GGNS program valves are not affected by EPU.

Control Rod Drive Hydraulic System The CRD hydraulic system changes due to EPU are addressed in Section 2.8.4.1. No modifications are being made to the system as a result of EPU. The ability of the GGNS program valves to perform their associated desi gn/safety function(s) is not affected by EPU.

Standby Liquid Control System The Standby Liquid Control system (SLCS) changes due to EPU are addressed in Section 2.8.4.5. The SLC maximum pump discharge pressure is increased by 54.3 psig. This discharge pressure increase does not result in any required system modifications; however, the pressure at which the pump is tested to satisfy TS requirements must be increased and pump operation at this new pressure must be verified prior to operation at EPU conditions. As a result, the GGNS TS Surveillance Requirement and the IST Program are being updated to address the increased pressure requirement. The ability of the IST Program valves to perform their safety functions is not affected by EPU. Suppression Pool Make-Up System The Suppression Pool Make-up (SPMU) system provides water from the upper containment pool (UCP) to the SP by gravity flow following a LOCA. Because containment analysis determined that the peak WW pressure will increase (Section 2.6.1), the maximum expected differential pressure (MEDP) for the SPMU dump valves w ill increase. In addition, four of the SP instrument line isolation valves experience an elevated ambient temperature due to EPU. The effect of the MEDP and ambient temperature increases to the SPMU system motor operated valves is discussed in Section 2.2.4.3. Combustible Gas Control System

Section 2.6.4 provides the current licensing basis for the Combustible Gas Control system (CGCS). No modifications are being made to this system as a result of EPU. The system's IST program valves provide for DW vacuum relief, DW isolation, and containment isolation (Section 2.6.1). These valves are not affected by EPU. RWCU System / RWCU Filter Demineralizer System The RWCU and RWCU Filter Demineralizer system changes due to EPU are addressed in Section 2.1.7. The small RWCU parameter changes do not affect the ability of the system MOVs and AOVs to meet their design requirements. In addition, the ability of the IST Program NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-86 check valves and containment penetration relief valves to perform their safety functions is not affected by EPU. Fuel Pool Cooling System and Cleanup System The Fuel Pool Cooling and Cleanup system (FP CCS) is addressed in Section 2.5.3.1. Additional relief valves required by the modification to increase the FPCC heat removal capability will be added to the IST program scope. Other than this minor program change, the FPCC IST program pumps and GGNS program valves are not affected by EPU.

Component Cooling Water System EPU effect on the CCW system is addressed in Section 2.5.3.3. The predicted heat load increase and the CCW HX cleaning modification do not affect the ability of the GGNS program valves to perform their safety function.

Standby Gas Treatment System

The EPU effect on the Standby Gas Treatment is addressed in Section 2.5.2.1. No modifications are being made as a result of EPU. The GGNS program valves are not affected by EPU.

Generic Letter 95-07

GL 95-07 addresses "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," August 17, 1995 (Reference 36). A review of key safety-related gate valves determined that there is no change in susceptibility to pressure locking or thermal binding as a result of EPU. Key parameters that cause susceptibility to pressure locking are not affected by EPU. The susceptibility to thermal binding is not increased because SP temperature remains below 200°F or the valve/system is not required to operate during the evaluated transient. 2.2.4.3 Individual Component Evaluations Maximum Expected Differential Pressure Effect The four SPMU dump valves are required to open to dump water from the UCP to the SP following a design basis accident. They are presently assumed to open against containment pressure (11.5 psig) plus the water head between the upper pool and the valve (19 psid), which is approximately 31 psid. Containment analysis (Section 2.6.1) determined that the peak WW pressure increases to 14.8 psig for a LOCA at EPU conditions. The resulting MEDP would increase from 31 psid to 34 psid or a 9.7% increas

e. The following table provides the associated torque margins for these valves:

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2-87 Parameter Value Available Torque Margin Maximum Available Stem Torque 4,375 ft-lbs NA Required stem torque at CLTP (based on peak containment pressure of 11.5 psig) 3,765 ft-lbs 16.2% EPU Required Stem Torque Assuming 9.7% Increase in MEDP 4,130 ft-lbs 5.9% Because the 9.7% MEDP increase was directly applied to the required stem torque change, the resulting 5.9% margin is conservative. The required stem torque is also affected by unseating

and packing loads which would not change with th e differential pressure (DP) increase. Based on the stem torque margin evaluation, no physical changes are required to these valves. Ambient Temperature Effect The areas affected by an ambient temperature increase are primarily the lower Auxiliary Building rooms which contain the ECCS equipmen t that circulates the SP water during a LOCA with loss of off-site power (LOOP). The EPU temperature effect becomes less significant at higher elevations. Only one valve (1E12F008) experiences an ambient temperature increase due to a HELB. Table 2.2-14 lists the valves that are affected by the ambient temperature increase and identifies the associated torque effect. As shown, all torque effects are well within the CLTP safety margin. For this reason, no physical modifi cations will be required due to the EPU effect on ambient temperatures. Conclusion Entergy has reviewed the assessments related to the functional performance of safety-related valves and pumps and concludes the effects of the proposed EPU on safety-related pumps and valves have been adequately addressed. Enter gy further concludes the effects of the proposed EPU on its MOV programs related to GL 89-10, GL 96-05, and GL 95-07, and the lessons learned from those programs to other safety-related, power-operated valves have been adequately evaluated. Based on this, Entergy concludes the safety-related valves and pumps will continue to meet the requirements of GDC-1, GDC-37, GDC-40, GDC-43, GDC-46, GDC-54, and 10 CFR 50.55a(f) following implementation of th e proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to safety-related valves and pumps.

2.2.5 Seismic

and Dynamic Qualification of Mechanical and Electrical Equipment Regulatory Evaluation Mechanical and electrical equipment covered by this section includes equipment associated with systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal. Equipment associated with systems essential NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-88 for preventing significant releases of radioactive materials to the environment are also covered by this section. The review focused on the effects of the proposed EPU on qualifying equipment to withstand seismic events and the dynamic effect s associated pipe whip and JI forces. The primary input motions due to the safe shutdown earthquake (SSE) are not affected by an EPU. The regulatory acceptance criteria are based on: (1) GDC-1, insofar as it requires SSCs important to safety be designed, fabricate d, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-30, insofar as it requires components that are part of the RCPB be designed, fabricat ed, erected, and tested to the highest quality standards practical; (3) GDC-2, insofar as it requires SSCs important to safety be designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions; (4) 10 CFR Part 100, Appendix A, which sets forth the principal seismic and geologic considerations for evaluating the suitability of plant design bases established in considering the seismic and geologic characteristics of the plant site; (5) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (6) GDC-14, insofar as it requires the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating

fracture; and (7) 10 CFR Part 50, Appendix B, which sets quality assurance requirements for safety-related equipment. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. Seismic and dynamic qualification of equipment is described in UFSAR Section 3.10, "Seismic and Dynamic Qualification of Mechanical and Electrical Equipment." Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Sections 10.1 and 10.3 of th e CLTR address the e ffect of CPPU on the seismic and dynamic qualification of mechanical and electrical equipment. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-89 Topic CLTR Disposition GGNS Result Electrical Equipment [[ Meets CLTR Disposition Mechanical Equipment With Non-Metallic Components

Meets CLTR Disposition Mechanical Component Design Qualification

     ]] Meets CLTR Disposition 2.2.5.1 Electrical Equipment The CLTR states that the increas e in power level increases the radiation levels experienced by equipment during normal operati on and accident conditions.  [[

     ]]   The safety-related electrical equipment was reviewed for EPU to ensure the existing qualification for the normal and accident conditions expected in the area where the devices are located remains adequate. Conservatisms in accordance with Institute of Electrical and Electronics Engineers (IEEE) 323-1974 are applied to the environmental parameters as required. EQ for safety-related electrical equipment located inside the containment is based on main steam line break (MSLB) and/or DBA LOCA conditions and their resultant temperature, pressure, humidity and radiation consequences, and includes the environments expected to exist during normal plant operation.  [[

                                ]]  Temperatures are not expected to increase during normal operation because any increase in heat loads is slight (e.g., near the FW lines) and are bounded by margins in CLTP design basis calculations.  [[                                                                                                                                   
                              ]]  Radiation levels under accident conditions also increase with EPU. The plant environmental conditions changed by EPU were reviewed to determine if the current envelopes are exceeded. If the CLTP environmental e nvelopes are exceeded, the qualification of the equipment located within the containment area was reviewed.

Accident temperature, pressure, and humidity environments used for qualification of equipment outside containment result from a MSLB in the steam tunnel, other HELBs, or loss of non-safety related HVAC, whichever is limiting for each plant area. The accident temperature, pressure and humidity conditions resulting from a LOCA do not change with the power level except for areas affected by an increase in SP temperature, and some of the HELB profiles may increase by a small amount (Section 2.2). Normal temperature, pressure, and humidity environments used for qualification of equipment outside containment do not change due to EPU. The current radiation NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-90 levels under normal plant conditions will also increase. Maximum accident radiation levels used for qualification of equipment outside containment are from a DBA LOCA. The plant environmental radiation and temperature conditions were reviewed to determine if the current envelopes are exceeded. If any CLTP environmen tal envelope is exceeded, the qualification of the equipment located within the area was reviewed. The qualification of seismically qualified devices is acceptable for operation at the EPU power level. The current/existing hydrodynamic loads remain applicable and bounding for seismic loads as well as all structural response spectra curves at EPU operating conditions. Based on a review of postulated pipe break s (Sections 2.2.1 and 2.5.1.3), qualified electrical components continue to be protected from the e ffects of JI, spray, or flooding at EPU conditions. Therefore, the seismic and dynamic qualification of electrical equipment meets all CLTR dispositions.

2.2.5.2 Mechanical Equipment/Components The CLTR states that the increas e in power level increases the radiation levels experienced by equipment during normal operati on and accident conditions. [[

     ]]   The mechanical design of equipment/components (e.g., pumps, HXs) in certain systems is affected by operation at EPU due to increased accident or normal operating temperatures, pressures and flow rates. Examples of the maximum conditions are: ECCS pumps operate post-LOCA with approximately 29°F higher SP temperature (bounding value used for the evaluation), FW isolation valves operate with approximately 31 psi higher pressure, and MSIVs operate with approximately 13% higher mass flow rate. Mechanical components can also be affected by ambient temperature conditions, radiologi cal increases either internal or external to components or external loads such as seismic JI or imposed nozzle loads. Mechanical design of equipment/components (e.g., HXs) in certain systems is affected by operation at the uprate power level due to slightly increased flow and, in some cases, temperatures.  [[

     ]] Evaluation of all GGNS safety-related mechanical components for EPU conditions demonstrated that process conditions under which each item operates during both normal and accident conditions are bounded by the CLTP design conditions.

The qualification of seismically qualified devices is acceptable for operation at the EPU power level. The existing seismic and hydrodynamic loads remain applicable and bounding for all response spectra curves at EPU operating conditions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-91 The effects of increased fluid induced loads on safety-related components are described in Section 2.6.1.2. Increased nozzle loads and component support loads due to the revised operating conditions were evaluated within the piping assessments in Section 2.2. [[

     ]] The effect of dynamic forces (pipe whip and JI) is minimal because there is minimal pressure increase for EPU (see Section 2.2.1). As stated above, the primary input motions due to the SSE are not affected by EPU and therefore, there are no consequences to the existing seismic analyses. No quality standards related to the design, fabrication, erection, and testing of the RCPB or SSCs important to safety are relaxed or removed as a result of EPU and no changes have been made to the plant design bases established in consideration of the seismic and geologic characteristics of the plant site.

The GGNS design and licensing bases do not require a formal mechanical EQ program. The GGNS design control program ensures that mechanical components are specified and procured for the environment and process conditions in which they are intended to function. Periodic maintenance and testing are performed in accordance with industry OE and vendor recommendations to ensure continued functionality. Causes of failures are investigated as part of the GGNS Maintenance Rule program and incorporated into equipment reliability improvement efforts. Therefore, the mechanical component design and the seismic and dynamic qualification of mechanical equipment/components meet all CLTR dispositions.

Conclusion Entergy has reviewed the evaluations of the e ffects of the proposed EPU on the qualification of mechanical and electrical equipment and concludes it: (1) adequately addressed the effects of the proposed EPU on this equipment and (2) demonstrated the equipment will continue to meet the requirements of GDCs 1, 2, 4, 14, and 30; 10 CF R Part 100, Appendix A; and 10 CFR Part 50, Appendix B, following implementation of the pr oposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the qualification of the mechanical and electrical equipment. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-92 Table 2.2-1 High Energy Line Breaks Change Due to EPU Break Type / Location Total Mass Release (Used for Flooding Evaluation) Mass Release Rate Pressure Temperature MS Line (Slot/Crack) Break in MST in Auxiliary Building - Steam Line

Break

Note that because the MST

is a "no-break" zone for the MS line, a non-mechanistic

slot/crack break has been

evaluated. Not Applicable No Change No Change No Change MS Line (Circumferential) Break in MST in Turbine Building - Steam Line

Break Not Applicable No Change No Change Not Applicable RCIC Line (Circumferential/Crack) Break in RCIC Pump Room - Steam Line Break Not Applicable No Change No Change No Change FW Line (Crack) Break in MST in Auxiliary Building - Liquid Line Break

Note that because the MST

is a "no-break" zone for the FW line, a non-mechanistic

crack break has been

evaluated. No EPU Effect (See Note 3) Increase in mass rate and a slight increase in break effluent enthalpy (See Note 1) Bounded by CLTP which is based on the MS

line slot break No Change due to presence of a blowout panel. RHR Line (Circumferential) Break in Piping Penetration Room - Liquid Line Break No Change No Change No Change No Change RWCU Line (Circumferential/Crack) Break in RWCU HX Room - Liquid Line Break N/A (See Note 4) Increase in break effluent enthalpy (See Note 2) Bounded by CLTP which is based on the

RHR line break in the Piping Penetration Room CLTP temperature envelope during the 9.7 < t < 60 second period slightly

increased at EPU conditions.

CLTP peak temperature remains unchanged at EPU conditions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-93 Change Due to EPU Break Type / Location Total Mass Release (Used for Flooding Evaluation) Mass Release Rate Pressure Temperature RWCU Line (Circumferential/Crack)

Break in Filter/Demineralizer Room - Liquid Line Break N/A (See Note 4) No Change No Change No Change RWCU Line (Circumferential/Crack) Break in Holding Pump Room - Liquid Line Break N/A (See Note 4) No Change No Change No Change RWCU Line (Circumferential/Crack) Break in Valve Nest Room - Liquid Line Break N/A (See Note 4) No Change No Change No Change RWCU Line (Circumferential/Crack) Break in RWCU Pump Room - Liquid Line Break No Change No Change No Change No Change Notes: 1. The CLTP mass release rate was estimated based on the saturated liquid blowdown. The EPU evaluation is based on subcooled blowdown because the initial FW pipe condition would remain unchanged during the entire transient until isolation by operator. Based on subcooled Moody critical flow, the EPU mass release rate increases to 1,086 lbm/sec (CLTP value is 313 lbm/sec). 2. The CLTP analysis applied a break effluent enthalpy of 470 Btu/lbm for the en tire transient. The limiting break flow enthalpy of 528.7 Btu/lbm (based on the EPU ICF operating condition), is applicable after 9.7 seconds when the only source of break flow is the RRS. 3. The break is conservatively evaluated by assuming the entire volume from the hotwell and the available condensate system make-up from the condensate storage tank (CST) are discharged into the break area. These volumes do not

change with EPU. 4. HELB flooding is limited to breaks outside containment.

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2-94 Table 2.2-2 RCPB Structural Evaluation Temperature Pressure Flow Rate System CLTP EPU CLTP EPU CLTP EPU Mechanical Loading [[

]]

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2-95 Table 2.2-3a Percentage Increase In Class 1 Pipe Stresses, Usage Factors, Interface Loads, and Thermal Displacements for GGNS Piping Systems Due to EPU Conditions ASME CODE PIPING CATEGORIES EQUATION A B NO. 1 Main Steam Feedwater Flow 4 Pressure & TemperatureTotal Flow Temperature Pressure Total 9A N/A N/A N/A N/A N/A N/A N/A 9B 4.45 0.00 4.45 N/A N/A N/A N/A 9C N/A N/A N/A N/A N/A N/A N/A 9D 4.45 0.00 4.45 N/A N/A N/A N/A 10 2.30 0.00 2.30 N/A N/A N/A N/A 12 N/A 0.00 0.00 N/A N/A N/A N/A 13 N/A 0.00 0.00 N/A N/A N/A N/A 14 1 1.15 0.00 1.15 N/A N/A N/A N/A 22.99 0.00 22.99 2 N/A N/A N/A N/A Interface Loads 34.92 0.00 34.92 3 N/A N/A N/A N/A Thermal Displacement N/A 0.00 0.00 N/A N/A N/A N/A Notes: N/A Not affected due to EPU

1. Fatigue - CUF

2. Interface load factor = 22.99 to be applied to the load combination associated with TSV load case. 3. Interface load factor = 34.92 to be applied to the load case alone and then recombined w ith the respective load combination. 4. The percentage increase of MS due to flow was developed based on the flow rate increase from OLTP at 16.492 Mlb/hr to 102% license power uprate flow at 19.43 Mlb/hr, a 17.8% increase.

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2-96 Table 2.2-3b Percent Increase In Class 2 and/or 3 Pipe Stresses, Interface Loads, and Displacements for GGNS Piping Systems Due To EPU Conditions ASME CODE PIPING CATEGORY EQUATION A B NO. 1 Main Steam Feedwater Flow 3 Pressure & TemperatureTotal Flow Pressure Temperature Total 8 N/A N/A N/A N/A N/A N/A N/A 9B 4.45 0 4.45 N/A N/A N/A N/A 9C N/A N/A N/A N/A N/A N/A N/A 9D 4.45 0 4.45 N/A N/A N/A N/A 10 N/A 0 0 N/A N/A N/A N/A 11 N/A 0 0 N/A N/A N/A N/A 22.99 0 22.99 1 N/A N/A N/A N/A Interface Loads 34.92 0 34.92 2 N/A N/A N/A N/A Thermal Displacement N/A 0 0.00 N/A N/A N/A N/A Notes: N/A Not affected due to EPU

1. Interface load factor = 22.99 to be applied to the load combination associat ed with TSV load case. 2. Interface load factor = 34.92 to be applied to the load case alone and then recombined with the respective load combination. 3. The percentage increase of MS due to flow was developed based on the flow rate increase from OLTP at 16.492 Mlb/hr to 102% license power uprate flow at 19.43 Mlb/hr, a 17.8% increase.

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2-97 Table 2.2-4a Summary of MS ASME Class 1 Piping, Pipe Stresses, and CUFs MS Lines A & D ASME NB-3600 Criteria Node No. Original Result (psi) EPU Result (psi) Code Allowable (psi) 2 Stress Ratio Remarks Eq. 9 020 10,570 10,570 28,725 0.368 OK Level B 020 19,861 20,745 34,470 0.601 OK Level C 020 19,824 19,824 43,088 0.460 OK Level D 020 19,953 20,841 57,450 0.362 OK Eq. 10 002N 60,390 61,779 57,450 1.075 OK (Note 3) Eq. 12 002N 42,617 42,617 57,450 0.742 OK Eq. 13 028 29,772 29,772 53,100 0.560 OK Eq. 14 017 CUF = 0.0728 0.0736 CUF <1.0 OK OK Notes: 1. Pipe Break Criteria: Postulated pipe break is required to be postulated at the following locations:

1. Terminal ends.

2. Eq. 10 and either Eq. 12 or Eq. 13 exceeds 2.4 S m , where S m is the Code Allowable Stress Limit.

3. CUF exceeds 0.1.

2. S m = 19,150 psi for Nodes 020, 002N; S m =18,200 for Node 017; and S m = 17,700 for Node 028. 3. Eq. 10 can exceed 3S m , if U <1.0 4. Load combinations B1 and D2 include TSV load. However, TSV factor is applied to the maximum stress for conservative analysi

s. 5. Level C is not affected by EPU.

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2-98 Table 2.2-4b Check of Pipe Breaks Agains t All Nodes Above Maximum Code Stresses MS Lines A & D Node Eq. 10 Eq. 12 Eq.13 Eq. 14 Remarks 020 32,775 x 1.023 =33,529 < 45,960 5,656 < 45,960 16,129 < 45,960 0.0442x 1.0115 =0.0447 < 0.1 Does not result in a break location 002N 60,390 x 1.0223=61,779 42,617 < 45,960 18,899 < 45,960 0.0153 x 1.0115 =0.0155 < 0.1 Does not result in a break location 028 34,409 x 1.023 =35,200 < 42,480 635 < 42,480 29,772 < 42,480 0.0064 x 1.0115 =0.00647 < 0.1 Does not result in a break location 017 28,499 x 1.023 =29,155 < 43,680 0 < 43,680 25,642 < 43,680 0.0728 x 1.0115 =0.0736 < 0.1 Does not result in a break locatio n NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-99 Table 2.2-4c Summary of MS ASME Cla ss 1 Piping, Pipe Stresses, and CUFs MS Lines B & C ASME NB-3600 Criteria Node No. Original Result (psi) EPU Result (psi) Code Allowable (psi) (Note 2) Stress Ratio Remarks Eq. 9 022 9,704 9,704 28,725 0.338 OK Level B 125 18,137 18,944 32,760 0.578 OK Level C 125 18,103 18,103 40,950 0.442 OK Level D 125 18,223 19,034 54,600 0.349 OK Eq. 10 128 56,937 58,247 54,600 1.067 OK (Note 3) Eq. 12 002N 37,486 37,486 57,450 0.652 OK Eq. 13 030 29,706 29,706 53,100 0.559 OK Eq. 14 020 CUF = 0.0732 0.0740 CUF <1.0 OK OK Notes: 1. Pipe Break Criteria: Postulated pipe break is required to be postulated at the following locations:

1. Terminal ends.

2. Eq. 10 and either Eq. 12 or Eq. 13 exceeds 2.4 S

m. 3. CUF exceeds 0.1.

2. S m = 19,150 for Node 022, 18,200 for Node 125, 17,700 for Node 030. 3. Eq. 10 can exceed 3S m , if U <1.0 4. Load combinations B1 and D2 include TSV load. However, TSV factor is applied to the maximum stress for conservative analysis. 5. Level C is not affected by EPU.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-100 Table 2.2-4d Check of Pipe Breaks Agains t All Nodes Above Maximum Code Stresses MS Lines B & C Node Eq. 10 Eq. 12 Eq. 13 Eq. 14 Remarks 022 31,821 x 1.023 = 32,552 < 45,960 5,550 < 45,960 15,931 < 45,960 0.0414x 1.0115 = 0.0419 < 0.1 OK 002N 54,539 x 1.023= 55,793 > 45,960 37,486 < 45,960 18,699 < 45,960 0.011 x 1.0115 =0.0111 < 0.1 Eq. 12/13 satisfy 125 56,675 x 1.023 =57,978 > 43,680 13,704 < 43,680 27,101 < 43,680 0.0697 x 1.0115 =0.0705 < 0.1 Eq. 12/13 satisfy 030 34,030 x 1.023 = 34,812 < 42,480 416 < 42,480 29,706 < 42,480 0.0061 x 1.0115 =0.0062 < 0.1 OK

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2-101 Table 2.2-4e MS (Lines A & D) Penetration Load Summary (Inside Containment) Level Loading Existing Enveloped LoadLoading with TSV EPU Result Maximum of EPU & Enveloped Load Percent of Increase Node B Fx (lb) 165,479 63,700 78,290 165,479 0% 028 Fy (lb) 18,292 1,200 1,476 18,292 0% ANC Fz (lb) 15,567 1,303 1,603 15,567 0% Mx (in-lb) 514,963 90,847 111,733 514,963 0% My (in-lb) 1,303,063 615,308 756,767 1,303,063 0% Mz (in-lb) 981,248 43,834 53,911 981,248 0% D Fx (lb) 166,582 21,072 25,916 166,582 0% Fy (lb) 18,356 656 807 18,356 0% Fz (lb) 20,271 7,402 9,104 20,271 0% Mx (in-lb) 530,782 85,431 105,072 530,782 0% My (in-lb) 1,925,803 1,361,104 1,674,022 1,925,803 0% Mz (in-lb) 986,514 90,473 111,273 986,514 0% Notes: 1. Levels A and C are not affected by EPU. 2. Level B and D loads are increased by 1.2299

3. Forces are in lbs and Moments are in in-lbs.

Penetration Coordinate System: X DIR: Along the penetration Y DIR: Vertical Z DIR: Perpendicular to X and Y axes

Conclusion:

EPU loads at the penetration are the same as the existing enveloped loads. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-102 Table 2.2-4f MS (Lines B & C) Penetration Load Summary (Inside Containment) Level Loading Existing Enveloped LoadLoading with TSV EPU Result Maximum of EPU & Enveloped Load Percent of Increase Node B Fx (lb) 147,778 96,052 118,134 147,778 0% 030 Fy (lb) 9,283 3,980 4,895 9,283 0% ANC Fz (lb) 15,213 12,898 15,863 15,863 4% Mx (in-lb) 242,558 109,866 135,124 242,558 0% My (in-lb) 1,289,753 1,090,682 1,341,430 1,341,430 4% Mz (in-lb) 641,193 164,801 202,689 641,193 0% D Fx (lb) 148,966 98,452 121,086 148,966 0% Fy (lb) 9,331 4,297 5,285 9,331 0% Fz (lb) 20,777 19,414 23,877 23,877 15% Mx (in-lb) 248,701 134,077 164,901 248,701 0% My (in-lb) 2,013,484 1,899,337 2,335,995 2,335,995 16% Mz (in-lb) 645,911 200,844 247,018 645,911 0% Note: Levels A and C are not affected by EPU. The greatest increase is approximately 16% for the Level D combination. The overall increase in the Level D moment is 5.6% (SR SS summation). The stress value due to the resultant moment is 3,242 psi, which is below the allowable value.

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2-103 Table 2.2-4g MS (Lines A & D) Nozzle Loading Summary Node No. Max. Original Result Loadin g with TSV EPU Result Code Allowable (Note 1) Allowable Ratio Acceptable 001 Loading Level B Force (lbs) 19,105 19,105 23,498 438,000 0.0536 OK Moment (in-lbs) 906,605 906,605 1,115,034 9,458,021 0.096 OK Loading Level D Force (lbs) 39,709 39,709 48,838 737,000 0.066 OK Moment (in-lbs) 1,706,141 1,706,141 2,098,383 14,594,000 0.117 OK Notes: 1. Allowable values used from MS line B & C computer run.

2. Level C loads are not affected.

3. Level B, Case 1 and Level D, Case 1 containing TSV load cases are used. The SRSS method was used for load combinations.

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2-104 Table 2.2-4h MS (Lines B & C) Nozzle Loading Summary Node No. Maximum Original Result Loading with TSV EPU Result Code AllowableAllowable Ratio Acceptable 002N Loading Level B Force (lbs) 34,419 34,419 42,332 438,000 0.0966 OK Moment (in-lbs) 1,212,496 1,212,496 1,491,249 9,458,021 0.1577 OK Loading Level C Force (lbs) 17,462 N/A 17,462 438,000 0.0399 OK Moment (in-lbs) 605,446 N/A 605,446 10,054,142 0.0602 OK Loading Level D Force (lbs) 34,510 34,510 42,444 737,000 0.0576 OK Moment (in-lbs) 1,216,765 1,216,765 1,496,499 14,594,000 0.1025 OK

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2-105 Table 2.2-4i MS (Lines A & D) SRV Flange Moment Summary (Inlet & Outlet) Node Level [See Note 2] Loading with TSV (in-lbs) EPU Result (in-lbs) Existing Enveloped Load (in-lbs) Maximum of EPU & Enveloped Loads (in-lbs) Allowable Ratio Remarks [See Note 1] 048 B 70,496 86,703 442,075 442,075 1,360,000 0.3251 OK Inlet D 96,007 118,079 445,143 445,143 1,360,000 0.3273 OK 064 B 76,989 94,689 373,946 373,946 1,360,000 0.2750 OK Inlet D 97,656 120,107 376,994 376,994 1,360,000 0.2772 OK 080 B 85,317 104,931 367,031 367,031 1,360,000 0.2699 OK Inlet D 104,819 128,917 370,182 370,182 1,360,000 0.2722 OK 096 B 90,588 111,414 425,779 425,779 1,360,000 0.3131 OK Inlet D 111,894 137,618 428,921 428,921 1,360,000 0.3154 OK 050 B 41,041 50,476 70,528 70,528 600,000 0.1175 OK Outlet D 43,217 53,153 70,777 70,777 600,000 0.1180 OK 066 B 48,414 59,544 65,254 65,254 600,000 0.1088 OK Outlet D 49,711 61,140 65,761 65,761 600,000 0.1096 OK 082 B 54,825 67,429 69,613 69,613 600,000 0.1160 OK Outlet D 55,939 68,799 70,166 70,166 600,000 0.1169 OK 098 B 56,988 70,090 71,013 71,013 600,000 0.1184 OK Outlet D 57,987 71,318 71,542 71,542 600,000 0.1192 OK Notes: 1. EPU results are less than existing enveloped loads. 2. Levels A and C are not affected by EPU.

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2-106 Table 2.2-4j MS (Lines B & C) SRV Flange Moment Summary (Inlet & Outlet) Node Level [See Note 2] Loading with TSV (in-lb) EPU Result (in-lb) Existing Enveloped Load (in-lb) Maximumof EPU & Enveloped Loads (in-lb) Allowable Ratio Remarks [See Note 1] Inlet 032 B 197,081 242,390 565,988 565,988 1,360,000 0.4162 OK D 218,830 269,139 569,239 569,239 1,360,000 0.4186 OK Inlet 047 B 201,476 247,795 507,249 507,249 1,360,000 0.3730 OK D 223,077 274,362 510,920 510,920 1,360,000 0.3757 OK Inlet 062 B 208,316 256,208 554,815 554,815 1,360,000 0.4080 OK D 230,155 283,068 558,191 558,191 1,360,000 0.4104 OK Inlet 077 B 222,112 273,176 588,510 588,510 1,360,000 0.4327 OK D 243,192 299,102 591,721 591,721 1,360,000 0.4351 OK Inlet 093 B 123,388 151,755 446,080 446,080 1,360,000 0.3280 OK D 142,122 174,796 449,114 449,114 1,360,000 0.3302 OK Inlet 108 B 144,945 178,268 473,939 473,939 1,360,000 0.3485 OK D 165,351 203,365 477,152 477,152 1,360,000 0.3508 OK Outlet 034 B 119,821 147,368 149,891 149,891 600,000 0.2498 OK D 123,716 152,158 151,024 152,158 600,000 0.2536 OK Outlet 049 B 122,537 150,708 156,380 156,380 600,000 0.2606 OK D 126,769 155,913 157,489 157,489 600,000 0.2625 OK Outlet 064 B 126,848 156,010 161,531 161,531 600,000 0.2692 OK D 130,717 160,769 162,428 162,428 600,000 0.2707 OK Outlet 079 B 141,745 174,332 186,667 186,667 600,000 0.3111 OK D 145,423 178,856 187,287 187,287 600,000 0.3121 OK Notes: 1. EPU results are less than existing enveloped loads. 2. Levels A and C are not affected by EPU.

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2-107 Table 2.2-4k MS (Lines A & D) Pipe Support Load Summary Support ID Node Level Existing Enveloped Load (lb) Loading with TSV (lb) EPU Result (lb) Maximum of EPU & Enveloped Load (lb) Snubber Allowable (lb) Ratio Remarks See Note BelowS102 005 B 11,833 4,814 5,921 11,833 70,000 0.1690 OK D 33,728 8,172 10,051 33,728 105,000 0.3212 OK S103 008 B 23,283 9,100 11,192 23,283 50,000 0.4657 OK D 29,266 11,444 14,075 29,266 75,000 0.3902 OK S104 010 B 29,402 5,484 6,745 29,402 50,000 0.5880 OK D 31,123 11,586 14,250 31,123 75,000 0.4150 OK S105 010 B 24,087 6,077 7,474 24,087 50,000 0.4817 OK D 25,781 11,018 13,551 25,781 75,000 0.3437 OK Note: EPU results are less than existing enveloped loads.

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2-108 Table 2.2-4l MS (Lines B & C) Pipe Support Load Summary Support ID Node Level Existing Enveloped Load (lb) Loading with TSV (lb) EPU Result (lb) Maximum of EPU & Enveloped Load (lb) Snubber Allowable (lb) Ratio Remarks [See Note below]S101 005 B 9,879 5,260 6,469 9,879 50,000 0.1976 OK D 59,707 7,794 9,586 59,707 75,000 0.7961 OK S102 006 B 25,102 18,203 22,388 25,102 70,000 0.3586 OK D 32,344 18,203 22,388 32,344 105,000 0.3080 OK S103 007 B 12,891 4,676 5,751 12,891 50,000 0.2578 OK D 29,010 6,596 8,112 29,010 75,000 0.3868 OK S104 009 B 13,238 8,157 10,032 13,238 30,000 0.4413 OK D 25,307 9,632 11,846 25,307 45,000 0.5624 OK S105 013 B 27,884 15,929 19,591 27,884 70,000 0.3983 OK D 45,295 17,470 21,486 45,295 105,000 0.4314 OK S106 019 B 47,026 11,522 14,171 47,026 70,000 0.6718 OK D 48,362 16,133 19,842 48,362 105,000 0.4606 OK S107 019 B 25,068 7,962 9,792 25,068 50,000 0.5014 OK D 25,912 10,314 12,685 25,912 75,000 0.3455 OK S108 013 B 27,081 16,120 19,826 27,081 70,000 0.3869 OK D 28,380 18,219 22,408 28,380 105,000 0.2703 OK Note: EPU result is less than existing enveloped loads.

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2-109 Table 2.2-5a BOP MS System Class 1 Piping (Outside Containment) Maximum Stress for Class 1 Piping Criteria Per ASME III NB 3600 Node Element Type CLTP EPU Allowable Ratio EQ. 9 Deadweight 28 Penetration 6,366 6,366 26,550 0.24 EQ. 9U Normal/Upset 28 Penetration 6,809 9,450 31,860 0.30 EQ. 9E Emergency N/A N/A N/A N/A N/A N/A EQ. 9F Faulted 28 Penetration 8,138 11,300 53,100 0.21 EQ. 10 Primary + Secondary 28 Penetration 34,409 34,409 53,100 0.65 EQ. 12 Secondary 28 Penetration 435 440 42,480 0.01 EQ. 13 Primary + Secondary 28 Penetration 27,907 27,907 42,480 0.66 EQ. 14 Primary + Secondary for

Usage Factor Evaluation 28 Penetration 0.0064 0.0064 0.100 0.064

The maximum stress originally occurred at Node 28, and EPU maximum stress occurs at Node 28.

Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients (Between Containment and Outboard Isolation Valve) No change

No change 38.8% Maximum pipe support loading**: EPU (due to thermal expansion loading) EPU (due to FT loads between Containment and Outboard Isolation Valve) No change

No change ** There is no pipe support in Class 1 MS piping outside containment. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-110 Table 2.2-5b BOP MS System Class 2 and Non-Safety Related Piping (Outside Containment) Maximum Stress for Class 2 and Non-Safety Related Piping Criteria Per ASME III NC 3600 Node Element Type CLTP EPU Allowable Ratio EQ. 8 Deadweight 517 Pipe Run 10,126 10,126 17,500 0.58 EQ. 9U1 Normal/Upset 1 52 Pipe Run 14,718 14,718 21,000 0.70 EQ. 9U2 Normal/Upset 2* 520/972 Valve 16,302 11,612 18,000 0.65 EQ. 9F Faulted 52 Pipe Run 15,831 15,831 42,000 0.38 EQ. 10 Thermal Expansion 932 Valve 30,175 30,933 26,250 1.18 EQ. 11 Deadweight + Thermal Expansion 932 Valve 36,550 37,308 43,750 0.85

The maximum stress originally occurred at Node 972, but the EPU maximum stress occurs at Node 52.

Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients (Between Containment and Turbine) 2.5% No change 38.8%** Maximum pipe support loading: EPU (due to thermal expansion loading) EPU (due to FT loads between Containment and Turbine) 2.5% 38.8%** ** Piping re-analyses using the time history TSVC loading for EPU on the piping outside containment was performed and the piping stress and support are evaluated against code allowable and are acceptable. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-111 Table 2.2-5c BOP FW System Class 1 Piping (Outside Containment) Maximum Stress for Class 1 Piping Criteria Per ASME III NB 3600 Node Element Type CLTP EPU Allowable Ratio EQ. 9 Deadweight N/A N/A N/A N/A N/A N/A EQ. 9U Upset 185 Branch 14,920 14,920 33,960 0.44 EQ. 9E Emergency N/A N/A N/A N/A N/A N/A EQ. 9F Faulted N/A N/A N/A N/A N/A N/A EQ. 10 Primary + Secondary

195 Valve 74,730 76,220 59,340 1.28 EQ. 12 Secondary 190 Valve 6,220 6,250 54,340 0.12 EQ. 13 Primary + Secondary 190 Valve 40,890 41,112 54,340 0.76 EQ. 14 Primary + Secondary for

Usage Factor Evaluation 195 Valve 0.058 0.064 0.100 0.64

Exceeded EQ. 10, qualified by EQ. 12 and EQ. 13 Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients 0.5% 0%

0% Maximum pipe support loading increase (due to thermal expansion loading):

- ** There is no pipe support in Class I FW piping outside containment.

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2-112 Table 2.2-5d BOP FW System Class 2 and Non-Safety Related Piping (Outside Containment) Maximum Stress for Class 2 and Non-Safety Related Piping Criteria Per ASME III NC 3600 Node Element Type CLTP EPU Allowable Ratio EQ. 8 Deadweight 405 Tee 10,865 10,865 17,500 0.62 EQ. 9U Normal/Upset 230 Pipe Run 12,924 12,924 18,000 0.72 EQ. 9E Emergency NA NA NA NA NA NA EQ. 9F Faulted 230 Pipe Run 13,617 13,617 36,000 0.38 EQ. 10 Thermal Expansion ** 505 Tee 36,844 36,522 26,250 1.39 EQ. 11 Deadweight + Thermal Expansion ** 505 Tee 45,987 45,987 43,750 1.051*

Exceeded allowable for thermal st ratification for CLTP, as justif ied in the existing design basis calculation. The thermal increase factor of 1.016 is not used because thermal stresses are governed by thermal stratification. There is no increase in temperature during thermal stratification for EPU. ** Qualified by EQ. 11 Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients 1.6%

0.0% 0% Maximum pipe support loading increase (due to thermal expansion loading): 1.6%

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2-113 Table 2.2-5e BOP Piping - Extraction Steam Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients 1.0% 0% 0% Maximum pipe support loading increase (due to thermal expansion loading): 1.0% Table 2.2-5f BOP Piping - FW Heater Drains & Vents (HDL and HDH) Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients 59% 0% 0% Maximum pipe support loading increase (due to thermal expansion loading): 59% Table 2.2-5g BOP Piping - Condensate Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients 12.7% 0% 0% Maximum pipe support loading increase (due to thermal expansion loading): 12.7% NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-114 Table 2.2-5h BOP Piping - ECCS Piping Maximum pipe stress increase: Temperature Expansion Pressure Fluid Transients 0.0%

0.0%

0.0% Maximum pipe support loading increase (due to thermal expansion loading): 100%

The temperature increase due to EPU is not a normal or upset event and a piping stress evaluation is not required per the existing design basis NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-115 Table 2.2-6 BOP Piping System Evaluation System Temperature (F) Pressure (psig) Flow Rate (Mlb/hr) Mechanical Loading CLTP EPU CLTP EPU CLTP EPU Condensate Piping (Condenser to condensate pump) 130 136.5 -10.7 -10.7 10.96 12.80 No change in mechanical loading Extraction Steam Piping (LP turbine to 1 st point heater) 145.8 151 -11.35 -10.89 0.078 0.089 No change in mechanical loading FW Piping (RFP to 5 th point heater, outside containment) 290 293.5 1,136 1,165 8.34 9.70 No change in mechanical loading Heater Vents and Drains Piping (LP heater #3A,B,C

to LP heater #2 A,B,C) 185 199.1 0 1 0.34 0.41 No change in mechanical loading MS and Reheat Piping (outside containment, from stop valves to containment penetration) 549.4 549.4 1,025 1,025 16.70 19.43 TSV loading changes with increased flow rate ECCS Piping (From suction strainer in SP) RHR LPCI RHR Containment Spray

HPCS LPCS RCIC 185 210* No change No change No change No change No change in mechanical loading Note: Maximum changes in system parameters for the systems that are affected in piping stresses and pipe supports are shown.

Evaluated at a conservatively high bounding value.

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2-116 Table 2.2-7 CUFs and S p+q Values of Limiting Components P + Q Stress (ksi) CUF 4,8 Component 1 Current (3,898 MWt) EPU (4,408 MWt) 3 Allowable (ASME Code Limit) Current (3,898 MWt) EPU (4,408 MWt) 3 Allowable FW Nozzle 9 - Carbon Steel Replacement Safe

End 59.5/ 51.6 2 45.9 54.3 0.732 0.886 1.0 FW Nozzle 9 - Stainless Steel Clad Replacement

Safe End 73.2/ 19.4 2 80.2/ 23.1 2 50.7 0.997 0.620 5 1.0 FW Nozzle 9 - Low Alloy Steel Forging 39.5 39.5 40.05 10 0.564 1.041 11,12 1.0 Steam Outlet Nozzle - Blend Radius 6 32.5 34.8 40.05 10 0.520 0.604 1.0 Steam Outlet Nozzle - Carbon Steel Safe End 48.3 51.7 54.3 See Note 13 See Note 13 1.0 Steam Outlet Nozzle -

Low Alloy Steel Safe

End 66.7 70.4 80.1 See Note 13 See Note 13 1.0 Recirculation Inlet

Nozzle - Stainless Steel Replacement Safe End 34.6 36.1 41.17 See Note 13 See Note 13 1.0 Recirculation Inlet

Nozzle - Inconel Replacement Safe End 38.1 53.8 69.90 See Note 13 See Note 13 1.0 Recirculation Inlet

Nozzle - Nozzle Forging (Low Alloy Steel) 6 39.5 39.5 7 40.05 10 0.564 0.685 1.0 Recirculation Outlet

Nozzle - Low Alloy

Steel 56.5 58.4 80.1 See Note 13 See Note 13 1.0 Recirculation Outlet

Nozzle - Stainless Steel 39.7 41.0 49.2 See Note 13 See Note 13 1.0 Recirculation Outlet Nozzle - Blend Radius (Low Alloy Steel) 6 32.5 33.5 40.05 10 0.540 0.549 1.0 Notes: 1. Only components with usage factors greater than 0.50 and experiencing an increase in flow, temperature, RIPDs or other mechanical loads are shown. 2. Thermal Bending included/Thermal bending rem oved. P + Q stresses are acceptable per CLTP elastic-plastic analysis, which is valid for EPU conditions. 3. EPU was conservatively evaluated for 102% of EPU (4,496 MWt). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-117 4. Only the limiting CUF is presented. 5. Conservatism was removed through reduced numbers of design cycles and finite element analysis (FEA) calculation of critical transients at EPU conditions. 6. Stress Values listed for this component are thermal stresses only.

7. OLTP evaluation bounds EPU conditions.

8. Fatigue Usage Factors are for a 40-year license

9. Considering normal operating conditions (i.e., does not consider Feedwater Heater Out-of-Service).

10. 1.5S m 11. While CUF exceeds ASME allowable (1.0), qualification of this component can be performed by inspection using NUREG-0619 (Reference 37) or GE Alternate Methods (Reference 38). 12. The bounding location (nozzle blend radius) is reported. All other nozzle forging locations have a CUF <1.0. 13. CUF is less than the evaluation threshold. Component is listed in table to present change in P + Q.

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2-118 Table 2.2-8 RIPDs for Normal Conditions Parameter CLTP 1 (psid) EPU 1 (psid) Shroud Support Ring and Lower Shroud 28.0 29.5 Core Plate and Guide Tube 24.8 25.8 Upper Shroud 3.4 4.0 Shroud Head 4.3 5.0 Shroud Head to Water Level (Irreversible

2) 7.0 7.7 Shroud Head to Water Level (Elevation

2) 0.8 0.7 Top Guide 0.46 0.46 Steam Dryer 0.46 0.57 Fuel Channel Wall (Maximum Power Bundle) 13.1 14.3 Notes: 1. 105% core flow with GE14. GE14 is the limiting fuel for RIPD.

2. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud (at the midpoint between the top of fuel and the shroud dome) and the exit of the separators.

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2-119 Table 2.2-9 RIPDs for Upset Conditions Parameter CLTP 1 (psid) EPU 1 (psid) Shroud Support Ring and Lower Shroud 30.4 31.9 Core Plate and Guide Tube 27.0 3 28.0 3 Upper Shroud 4.8 5.6 Shroud Head 6.1 7.0 Shroud Head to Water Level (Irreversible

2) 9.8 10.8 Shroud Head to Water Level (Elevation

2) 1.1 1.0 Top Guide 0.54 0.54 Steam Dryer 0.70 0.86 Fuel Channel Wall (Maximum Power Bundle) 15.0 3 16.8 3 Notes: 1. 105% core flow with GE14. GE14 is the limiting fuel for RIPD.

2. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud (at the midpoint between the top of fuel and the shroud dome) and the exit of the separators.

3. Core Plate and Channel Wall RIPD values are bounded by the Faulted Condition results.

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2-120 Table 2.2-10 RIPDs for Faulted Conditions Parameter CLTP 1 (psid) EPU 1 (psid) Shroud Support Ring and Lower Shroud 51.0 52.0 Core Plate and Guide Tube 27.0 28.0 Upper Shroud 32.0 32.5 Shroud Head 24.5 25.5 Shroud Head to Water Level (Irreversible

2) 26.0 27.0 Shroud Head to Water Level (Elevation

2) 1.9 1.9 Top Guide 18.0 18.2 Steam Dryer 3 13.0 13.0 Fuel Channel Wall (Maximum Power Bundle) 15.9 16.8 Notes: 1. Values are the maximum results from either the cavitation interlock power or the high power with 105% core flow points for GE14 fuel. Evaluations at these points considered both CLTP and EPU operation, as well as both normal and reduced feedwater temperatures (RFWTs). The GE14 results bound the results for GNF2 fuel. A RFWT of 50°F was used for CLTP and 100°F for EPU.

2. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud (at the midpoint between the top of fuel and the shroud dome) and the exit of the separators.

3. The limiting steam dryer dP is at the hot standby condition, which is a zero-power condition and is not affected by EPU.

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2-121 Table 2.2-11 Governing Stress Results for RPV Internal Components Item Component Service Level Stress / Load Category 2 Unit CLTP EPU Allowable U P m ksi 18.00 18.00 20.97 U P m + P b ksi 30.70 30.70 31.45 F P m ksi 35.40 40.00 46.60 1 Shroud Support F P m + P b ksi 61.70 69.10 69.90 U P m + P b ksi 18.03 18.03 21.45 U P + Q ksi 38.63 38.63 42.90 2 Shroud F P m + P b ksi 43.10 47.06 51.48 U P m + P b ksi 19.51 19.94 21.45 3a Core Plate -Ligament

Stress F P m + P b ksi 45.16 45.16 51.48 U Buckling P/Pc 0.32 0.33 0.45 3b Core Plate -Stiffener Buckling F Buckling P/Pc 0.82 0.82 0.90 U P m + P b ksi 19.76 19.76 21.45 E P m + P b ksi 31.99 31.99 32.17 3c Core Plate - Beam

Stress F P m + P b ksi 37.03 37.03 51.48 P m ksi 9.57 9.57 14.30 U P m + P b ksi 14.30 14.30 21.45 P m ksi 19.13 20.72 34.32 4 Top Guide/Grid F P m + P b ksi 28.61 30.99 51.48 P m ksi 10.35 10.35 16.60 U P m + P b ksi 19.56 19.56 21.45 P m ksi 10.65 10.65 39.84 5 CRDH F P m + P b ksi 20.57 20.57 59.76 U P m + P b ksi 17.59 17.98 24.00 6 CRGT F P m + P b ksi 32.58 35.87 38.40 P m ksi 9.7 9.7 10.4 N/U/E P m + P b ksi 14.7 14.7 16.6 P m ksi 19.8 19.8 24.9 7 OFS F P m + P b ksi 30.8 30.8 37.4 P m ksi 1.51 1.51 14.1 8 Peripheral Fuel Support N/U/E/F P m + P b ksi 1.51 1.51 14.1 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-122 Item Component Service Level Stress / Load Category 2 Unit CLTP EPU Allowable 9 Fuel Channel Qualified for EPU conditions U P m + P b ksi 11.24 12.67 21.45 10 FW Sparger F P m + P b ksi 34.46 35.89 51.48 U P + Q m ksi 18.6 18.6 50.7 E P m + P b ksi 11.4 11.4 38.0 11 Jet Pump F P m + P b ksi 59.6 59.6 60.8 U P + Q ksi 47.45 47.45 50.78 12a Core Spray Sparger F P m + P b ksi 28.05 28.05 60.93 U P + Q m ksi 30.76 30.76 41.85 12b Core Spray Line F P m + P b ksi 49.13 49.13 61.02 N/U P m + P b ksi 4.5 4.5 20.5 13 Access Hole Cover F P m + P b ksi 42.8 42.8 49.4 U P m ksi 9.4 9.4 10.73 14a Shroud Head And Separator - Shroud

Head F P m ksi 16.12 16.12 25.74 U P m ksi 19.03 19.03 23.30 U P + Q ksi 28.04 28.04 34.26 14b Shroud Head And Separator - Stud F P m ksi 24.19 24.19 55.92 U P m + P b ksi 18.06 18.06 23.3 15 ICHGT F P m + P b ksi 21.73 21.73 55.92 U 1 P + Q m ksi 47.40 47.40 53.10 16 VHCSN F P m + P b ksi 26.55 26.55 49.00 U P m + P b ksi 14.03 14.03 14.86 E P m + P b ksi 20.25 20.25 22.29 17 LPCI Coupling F P m + P b ksi 25.07 26.27 35.66 Notes: 1. Calculated stresses are based upon faulted load combination for conservatism. 2. P = Primary stress intensity; P m = Primary membrane stress intensity; P b = Primary bending stress intensity; Q = Secondary stress intensity; Q m = Secondary membrane stress intensity; N = Normal; U = Upset; E = Emergency; F = Faulted

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2-123 Table 2.2-12 Fatigue Usage for RPV Internal Components for Plant Life of 40-Years Item Component CLTP CUF EPU CUF Allowable 1 Shroud Support 0.098 0.098 1.0 2 Shroud 0.56 0.56 1.0 3 Core Plate 0.9 0.9 1.0 4 Top Guide/Grid 0.018 0.018 1.0 5 CRDH 0.107 0.107 1.0 6 CRGT 1 Remain qualified for EPU 1.0 7 OFS 1 Remain qualified for EPU 1.0 8 Peripheral Fuel Support 1 Remain qualified for EPU 1.0 9 FW Sparger 0.961 0.961 1.0 10 Jet Pump 0.946 0.946 1.0 11 Core Spray Line and Sparger 0.83 0.83 1.0 12 Access Hole Cover 0.24 0.24 1.0 13 Shroud Head And Separator 0.906 0.906 1.0 14 ICHGT 0.013 0.013 1.0 15 VHCSN 0.79 0.79 1.0 16 LPCI Coupling 1 Remain qualified for EPU 1.0 Note: 1. Exempted from fatigue analysis per Paragraph NG-3222.4(d) of the ASME Code Section III.

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2-124 Table 2.2-13 GGNS Program Pumps and Valves System Designator IST Pumps IST Valves GL 89-10 Valves MOV Program AOV Program System Affected by EPU Nuclear Boiler System (B21) X X X X X Reactor Recirculation System (B33) X X X X Control Rod Drive Hydraulic System (C11) X X X X X Standby Liquid Control System (C41) X X X X X Drywell Monitoring System (D23) X X Residual Heat Removal System (E12) X X X X X X Low Pressure Core Spray System (E21) X X X X X High Pressure Core Spray System (E22) X X X X X Suppression Pool Make-up System (E30) X X X X MSIV Leakage Control System (E32) X X X X Feedwater Leakage Control System (E38) X X X X Reactor Core Isolation Cooling System (E51) X X X X X X Combustible Gas Control Systems (E61) X X X X X Reactor Water Cleanup System (G33) X X X X X RWCU Filter/Demineralizer System (G36) X X X Fuel Pool Cooling and Cleanup System (G41) X X X X X X FPC and Cleanup Filter/Demineralizer System (G46) X X X Containment Personnel Air Locks (M23) X Containment Cooling System (M41) X X Containment Leak Rate Test System (M61) X Containment and Drywell Instrument and Control System (M71) X X X X Main and Reheat Steam System (N11) X* X Condensate System (N19) X* X Feedwater System (N21) X* X Condensate Cleanup System (N22) X* X Heater Vents and Drains (N23) X* X Seal Steam and Drains (N33) X* X NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-125 System Designator IST Pumps IST Valves GL 89-10 Valves MOV Program AOV Program System Affected by EPU MSR Vents and Drains (N35) X* X Extraction Steam System (N36) X* X Condenser Air Removal System (N62) X* X Offgas System (N64) X* X Condensate and Refueling Water Storage and Transfer System (P11) X X Makeup Water Treatment System (P21) X X X X X Standby Service Water System (P41) X X X X X X Component Cooling Water System (P42) X X X X* X Turbine Building Cooling Water System (P43) X* X Plant Service Water System (P44) X X X X X Floor and Equipment Drains System (P45) X X X Plant Service Water Radial Well System (P47) X X Suspended Floor and Equipment Drain System (P48) X X X Service Air System (P52) X X X X X Instrument Air System (P53) X X X X X Suppression Pool Cleanup System (P60) X X Fire Protection System (P64) X Domestic Water System (P66) X X Plant Chilled Water System (P71) X X Drywell Chilled Water System (P72) X X X X Standby Diesel Generator System (P75) X HPCS Diesel Generator System (P81) X Auxiliary Building Ventilation System (T41) X Fuel Handling Area Ventilation System (T42) X Standby Gas Treatment System (T48) X Control Room HVAC System (Z51) X X Containment Atmospheric Sampling X X Leak Detection System X X Note:

Non-Safety Related Components

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2-126 Table 2.2-14 Ambient Temperature Effects to GGNS Program Valves Valve # Valve Function CLTP Temp (ûF) EPU Temp (ûF) Delta Temp (ûF) EPU Torque Effect CLTP Safety Margin 1E12F004C RHR C Pump SP Suction 143 151 8 0.4% >60% 1E12F008 RHR SDC Suction from RPV 295 304 9 0.3% 30%-60% 1E12F064C RHR C Pump Minimum Flow 143 151 8 0.9% 10%-30% 1E12F094 SSW Flooding of RPV 113 114 1 0.07% 30%-60% 1E12F096 SSW Flooding of RPV 113 114 1 0.07% 30%-60% 1E21F001 LPCS SP Suction 151 157 6 0.7% 60% 1E21F011 LPCS Pump Minimum Flow 151 157 6 0.5% 30%-60% 1E22F001 HPCS CST Suction 113 114 1 0.08% 30%-60% 1E22F012 HPCS Pump Minimum Flow 154 155 1 0.09% 60% 1E22F015 HPCS SP Suction 154 155 1 0.1% 60% 1E30F593A SP Instrument Line Isolation 113 114 1 0.07% >60% 1E30F593B SP Instrument Line Isolation 143 151 8 0.6% >60% 1E30F594A SP Instrument Line Isolation 151 157 6 0.5% >60% 1E30F594B SP Instrument Line Isolation 113 114 1 0.07% >60% 1P41F113 SSW Fill Tank Outlet 113 114 1 0.1% >60%

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2-127 2.3 Electrical Engineering

2.3.1 Environmental

Qualification of Electrical Equipment Regulatory Evaluation EQ of electrical equipment involves demonstrating the equipment is capable of performing its safety function under significant environmental stresses, which could result from DBAs. The review focused on the effects of the proposed EPU on the environmental conditions that the electrical equipment will be exposed to during normal operation, AOOs, and accidents. The review was conducted to ensure the electrical equipment will continue to be capable of performing its safety functions following implem entation of the proposed EPU. The regulatory acceptance criteria for EQ of electrical equipm ent are based on 10 CFR 50.49, which sets forth requirements for the qualification of electrical equipment important to safety that is located in a harsh environment. GGNS Current Licensing Basis EQ of equipment is described in UFSAR Section 3.11, "Environmental Design of Safety-Related Mechanical and Electrical Equipment."

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 10.3.1 of the CLTR addresses the effect of CPPU on the EQ of Electrical Equipment. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Electrical Equipment [[

     ]] Meets CLTR Disposition The CLTR states that the increas e in power level increases the radiation levels experienced by equipment during normal operati on and accident conditions.  [[

     ]] All electrical equipment in the EQ program was evaluated. The equipment was reviewed consistent with the requirements of 10 CFR 50.49 and RG 1.89 (Reference 39) to assure the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-128 existing qualification for the normal and accident conditions expected in the area where the devices are located remains adequate. The 10 CFR 50.49 acceptance criteria, including pressure, temperature, relative humidity (RH), and radiation, were used in making this determination. The EQ program equipment qualification basis was evaluated using the changes to existing normal and accident radiation doses expected when operating at the EPU increased reactor power level. The cumulative effect on the dose applied to equipment is dependent on location, the post-accident operating time, and installed equipment configuration (i.e., completely sealed equipment does not consider beta radiation). The normal and post-LOCA radiation dose value changes are based on dose scaling factors and apply to plant areas inside and outside primary containment. At GGNS, there are approximately 73 normal operation environments and approximately 125 accident environments that are assigned, as applicable, to plant areas that contain safety-related equipment. The normal operation dose multiplying factors are for gamma, beta and neutron radiation, and are, for most areas, the percentage of the uprate. Areas that have normal operation gamma radiation level increases greater than the percen tage of the uprate (e.g., the Turbine Building) do not have any EQ environmental zones. The effect of the EPU on normal operation radiation environments was estimated by utilizing the percentage increase in reactor power including the effect of power uncertainty. The estimated increase took into consideration a GGNS operating period that reflects approximately 18 years of operation at OLTP, 9 years at CLTP, and 13 years at EPU conditions. The accident dose multiplying factors are for gamma and beta radiation, and vary with the contributing liquid and airborne accident radia tion sources in the area (taking into account whether shielded or unshielded), over post-LOCA time intervals from 0 hours to 180 days. These multiplying factors are based on a comparison of the integrated photon source terms and beta average energies developed based on the CLTP core inventory versus the photon source terms and beta average energies developed ba sed on the EPU core inventory. Because the relative abundance of each isotope and the average energy of each isotope are the key parameters that affect direct exposure, having a scaling factor that addresses the change in these parameters, and taking into consideration decay and repres entative shielding, is sufficient to assess the radiological effect of the EPU and change in fuel cycle. The EPU evaluation determined that there was significant conservatism in the development of the CLTP radiation source terms that served to overestimate the CLTP doses at all post-accident time intervals greater than approximately 10 hours. Thus, the EPU versus CLTP source term comparison indicated that the CLTP doses would bound the EPU doses for time intervals greater than the first 10 hours. However, the EPU evaluation also identified seve ral deficiencies in the ODB dose calculations supporting equipment qualification which had widesp read effect and served to increase the CLTP gamma radiation environments to the extent that the EPU was no longer bounded by the CLTP. These deficiencies were evaluated through the GGNS corrective action program. The NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-129 EPU environments used to assess the effect on equipment qualification reflect the combined effect of the EPU and the deficiencies in the ODB dose calculations. Once the effect of EPU radiation dose value was determined, the equipment was categorized into three groups as follows: Group I - Not affected by implementation of EPU. The EPU parameters are bounded by the existing qualification levels of the equipment. This group includes equipment with sufficient life to demonstrate radiation qualif ication for 13 years at EPU operation after consideration of GGNS operation history that reflects approximately 18 years of operation at OLTP and 9 years at CLTP conditions. Group II - Partially affected by implemen tation of EPU due to postulated higher radiological conditions. The EQ equipment is still qualified with reduced radiation life; however, plant modification is required to operate the plant at EPU limits starting February 2012. Table 2.3-1 provides a list of components which require disposition in accordance with the GGNS EQ program due to their limited qualified life. Group III - Evaluation of the existing equipmen t in this group indicated that the EPU total integrated dose (TID) is higher than the currently qualified limits. Table 2.3-2 provides a list of this equipment along with component IDs, Manufacturer, Model, zone/location, location identification for inside or outside containment, qualification doses, EPU TID and if EPU thermal environments need to be addressed for qualification. EQ file updates will be completed as required by 10 CFR 50.49 prior to EPU implementation. Remaining life determinations will be made for all Group II items and any required modifications or replacement of equipment will also be completed prior to EPU implementation. Inside Containment EQ for safety-related electrical equipment located inside the containment is based on MSLB and/or DBA LOCA conditions and their resultant temperature, pressure, humidity, and radiation consequences, and includes the environments expected to exist during normal plant operation. Normal temperatures are essentially unchanged and remain bounded by the normal temperatures used in the EQ analyses. The worst case plant post-LOCA temperature profiles inside the DW were revised due to EPU changes. These changes require the DW EQ profile to have a revised long-term temperature beyond 60 days. The EPU curve at 60 days continues a downward slope (cooling) out to 100 days, but at a slightly slower cooldown rate than the CLTP EQ profile. The worst case plant inside containment (outside the DW) profiles were not revised due to EPU changes. The worst case EQ enveloping temperature profiles are shown in Figure 2.3-1. Generally, the normal operation radiation doses in areas within the containment and DW increased by approximately 6% and 15%, respectively. The increase in the DW reflects the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-130 combination of the dose increase due to the EPU, plus the adjustments made to account for an estimated plant capacity factor post-EPU (expected to be essentially 1.0), while the CLTP doses in the DW are based on a plant capacity factor of 0.8. Capacity factor corrections were not required in the containment because the CLTP dose estimates are conservatively based on a capacity factor of 1.0. The percentage increase in accident doses in the containment and DW vary due to the CLTP design basis calculation deficiencies discussed earlier. The total integrated doses (normal plus accident) for EPU conditions were determined to affect qualification of some equipment located inside containment. Several EQ zones experience an increase in normal gamma dose rates. As indicated in Table 2.3-1, some components within primary containment are expected to exceed EQ limits prior to the end of 13 years of operation at EPU conditions. These components will be addressed within the GGNS EQ program as warranted. The changes to the GGNS EQ program brought about by the implementation of EPU will be documented and administered per Entergy Administrative Procedure "Environmental Qualification (NUREG-0588 / 10 CF R 50.49)," 01-S-06-57, Revision 0. Outside Containment Accident temperature, pressure, and humidity environments were used for qualification of equipment outside containment resulting from an MSLB, or other HELBs, whichever is limiting for each plant area. Accident pressure and humidity are unaffected by EPU. In some areas outside containment, post-accident ambient temper ature increases slightly due to EPU, either from HELBs with changed conditions (see Table 2.2-

1) or due to the loss of non-safety related HVAC post-LOCA (see Section 2.7.6). The exis ting CLTP Auxiliary Building enveloping temperature profile was raised 6°F early in the accident; however, the peak temperature was not increased, as shown in Figure 2.3-1. Zone-specific temperature profiles were used for CLTP qualification in the Auxiliary Building. Two of th ese zone-specific profiles were affected by the HELB analysis for EPU during the first 200 seconds. The Auxiliary Building enveloping qualification temperature profile shown in Figure 2.3-1 is a composite of the zone specific profiles and includes the HELB analysis for EPU. The normal temperature, pressure, and humidity conditions do not change as a result of EPU. The current radiation levels under normal plant conditions were conservatively evaluated to increase in proportion to the increase in reactor thermal power. In general, normal operation

radiation doses in areas within the Auxiliary Building increased by approximately 6%. The percentage increase in accident doses in the A uxiliary Building vary due to the CLTP design basis calculation deficiencies discussed earlier. The qualification basis for the EQ program equipment was evaluated based on the revised EPU normal and accident radiation dose values. This evaluation used environmental area dose values and equipment-specific dose values where necessary. The evaluation determined that the post-EPU radiation dose changes result in changes in existing equipment qualification status as noted below. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-131 Table 2.3-2 lists components for which the existing qualification outside containment is not adequate to support equipment operation at EPU limits. The components listed in Table 2.3-2 are to be replaced with equipment suitably qualified for the EPU environmental conditions. The changes to the GGNS EQ program brought about by the implementation of EPU will be documented and administered per Entergy Administrative Procedure "Environmental Qualification (NUREG-0588 / 10 CF R 50.49)," 01-S-06-57, Revision 0. Therefore, the EQ of electrical equipment meets all CLTR dispositions. Conclusion Entergy has reviewed the effects of the proposed EPU on the EQ of electrical equipment and concludes the effects of the proposed EPU on the environmental conditions and the qualification of electrical equipment have been adequately a ddressed. Entergy further concludes the electrical equipment will continue to meet the relevant requirements of 10 CFR 50.49 following implementation of the proposed EPU, consistent with the current licensing basis. Therefore, Entergy finds the proposed EPU acceptable with respect to the EQ of electrical equipment.

2.3.2 Off-site Power System Regulatory Evaluation The off-site power system includes two or more physically independent circuits capable of operating independently of the on-site sta ndby power sources. The review covered the descriptive information, analyses, and referenced documents for the off-site power system and the stability studies for the electrical transmission grid. The review focused on whether the loss of the nuclear unit, the largest operating unit on the grid, or the most critical transmission line will result in the LOOP to the plant following implementation of the proposed EPU. The regulatory acceptance criteria for off-site power systems are based on GDC-17. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The off-site power system is described in UFSAR Section 8.2, "Offsite Power System." Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.1 of the CLTR a ddresses the effect of CPPU on the AC Power System. The results of this evaluation are described below. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-132 GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result AC Power (Degraded Voltage) [[ Meets CLTR Disposition AC Power (Normal Operation)

     ]] Meets CLTR Disposition 2.3.2.1 AC Power (Degraded Voltage)

As explicitly stated in Section 6.1 of the CLTR, the increase in thermal power translates to an increased electrical output from the station. For the off-site power supply, the equipment is typically adequate for operation with the uprat ed electrical output. Changes in electrical requirements to support normal plant operation are not safety-related. The increased power from the generator will have no adverse effect on the grid stability/reliability. The GGNS main generator is connected to the 20.9/500kV Main Step-up Transformer. The 500kV switchyard consists of buswork, a 500kV di sconnect switch, and the associated control and protection systems. The 500kV off-site power sources originate from the Baxter Wilson and Franklin Substations. The 500kV off-site power sources consist of 500/34.4kV station service transformers, circuit breakers, disconnect switches, and transmission lines. The 115kV off-site power source originates from Port Gibson. The 115kV off-site power source consists of an overhead 115kV line from the Port Gibson Substation terminated near the plant site to an underground 115kV cable. The 115kV cable feeds a 115/4.16kV ESF transformer located adjacent to the plant. The protective relaying schemes are designed to protect the equipment from electrical faults. Electrical ratings and margins associated with major components of the off-site power system are given in Table 2.3-3. A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to GDC 17 (10 CFR 50 Appendix A). Details of these studies are provided in Attachment 12 to the EPU LAR. The analysis has determined that the power uprate will not adversely affect bulk power transmission system steady-state power flow (thermal

ratings and voltage), stability, short circuit duty or power transfer levels. Grid events analyzed included loss of the largest generator, loss of GGNS, and loss of the most critical transmission line due to a fault with the unit operating at full power uprate capacity. Pre-event line outages were also considered. Stability simulations were transiently stable and exhibited positive damping with the power uprate. GGNS off-s ite power steady-state and transient voltages resulting from critical transmission line faults or loss of GGNS generation are adequate to operate loads required for safe shutdown and will preclude the inadvertent separation from the off-site supply. Reactive power will be maintained within acceptable limits analyzed in the grid NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-133 studies. This will be accomplished by utilizing the existing generator-exciter control system, with the addition of 216 MVAR capacitor banks in various 500kV substations, and governed by operational procedures as described in Section 2.5.1.2.2. Therefore, off-site power at degraded voltage meets all CLTR dispositions. 2.3.2.2 AC Power (Normal Operation) As explicitly stated in Section 6.1 of the CLTR, the increase in thermal power translates to an increased electrical output from the station. For the off-site power supply, other than the main generator, the equipment is typically adequate for operation with the uprated electrical output. Changes in electrical requirements to support normal plant operation are not safety-related. The increased power from the generator will have no adverse effect on the grid stability/reliability. The existing off-site electrical equipment, other than the Main Step-Up Transformer was determined to be adequate for operation with the uprated electrical output and increased electrical loading . The review concluded the following: The 500kV switchyard components (i.e. bus, breakers, switches, current transformers (CTs), and lines) are adequate for increased generator output associated with EPU. The 115kV system components (i.e. bus, breakers, switches, circuit switchers, and lines)

are adequate for operation under EPU conditions. The main step-up transformer is being replaced to increase transformer power handling capacity prior to operation at EPU conditions. In add ition, the isolated phase bus (IPB) duct cooling system is being modified to increase the IPB duc t continuous current rating to provide additional margin for operation at EPU output and reduced voltage. The protective relaying review included protective relaying for the main generator, 20.9/500kV main step-up transformer, 500kV transmission line, the Baxter Wilson and Franklin Substation relays and all on-site equipment supplying power to the plant. For 500kV, 34.5kV, 13.8kV, 6.9kV, 4.16kV and 480V equipment supplying power to the plant, the existing protective relay setpoints are developed and validated based on equipment ratings, which are not being changed for EPU. The existing protective relay settings for the main generator will have to be

recalculated due to the increased EPU power output. Therefore, other than those associated with the uprated main generator, the settings are unaffected by operation at EPU conditions and off-site power during normal operation meets all CLTR dispositions. Conclusion Entergy has reviewed the assessment of the eff ects of the proposed EPU on the off-site power system and concludes the off-site power system will continue to meet the requirements of GDC-17 following implementation of the proposed EPU. Adequate physical and electrical NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-134 separation exists and the off-site power system has the capacity and capability to supply power to safety loads and other required equipment. Entergy further concludes the effect of the proposed EPU on grid stability is insignificant. Therefore, Entergy finds the proposed EPU acceptable with respect to the off-site power system. 2.3.3 AC On-Site Power System Regulatory Evaluation The alternating current (AC) on-site power system includes those standby power sources, distribution systems, and auxiliary supporting systems provided to supply power to safety-related equipment. The review covered the descriptive information, analyses, and referenced documents for the AC on-site power system. The regulat ory acceptance criteria for the AC on-site power system are based on GDC-17, insofar as it requires the system to have the capacity and capability to perform its intended functions during AOOs and accident conditions. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The AC on-site power system is described in UFSAR Section 8.3.1, "AC Power Systems."

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.1 of the CLTR a ddresses the effect of CPPU on the AC On-Site Power System. The GGNS AC on-site power distribution system consists of transformers, numerous buses, and switchgear. AC power to the distribution system is provided from the transmission system, 500kV switchyard, and from on-site diesel generators. The AC on-site power system consists of equipment and systems required to provide AC power to and service loads under all conditions of plant operation. This includes 34.5kV switchgear, 13.8kV switchgear, 6.9kV switchgear, 4.16kV switchgear, 480V load centers and motor control centers, 208Y/120V distribution panels, UPS systems, and standby diesel generators. The AC on-site power distribution system loads were reviewed under both normal and emergency operating scenarios. In both cases, loads are computed based on equipment nameplate data, with conservative demand factors a pplied. These loads are used as inputs for the computation of load, voltage drop and short circuit current values in electrical analysis software NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-135 called ETAP (Electrical Transient Analysis Program ). The significant changes in electrical load demand are associated with two non-safety related changes: the Radial Well addition and Auxiliary Cooling Tower (ACT) expansion. The GGNS review covered the AC power components with respect to their functional performance as affected by various configurations and loading conditions including full operation a nd unit trip with LOCA. The GGNS review focused on the additional electric load that would result from the proposed EPU. Sufficient capacity is available so that no electrical distribution system modifications would be required except for local interfaces. There are no changes to the emergency diesel generator loads or load sequencing for EPU. Therefore the fuel oil requirements do not change and the existing supply is adequate. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result AC Power (Degraded Voltage) [[ Meets CLTR Disposition AC Power (Normal Operation)

     ]] Meets CLTR Disposition 2.3.3.1 AC On-Site Power (Degraded Voltage)

As explicitly stated in Section 6.1 of the CLTR, the increase in thermal power translates to an increased electrical output from the station. Table 2.3-4 provides a summary of the loading increases to the on-site power analysis model due to EPU operation.

[[

                                                                   ]]  The increased load on each Service Transformer (No. 11 and No. 21) 34.4 kV winding due to the increase in power generation pump and motor load has an insignificant effect on the safety-related buses supplied from the 4.16 kV winding of the ESF transformers. Starting and running voltages are affected by <0.12% and the existing degraded voltage relay settings are acceptable for EPU. Therefore, the current emergency power system remains adequate. The electrical systems have sufficient capacity to support all required loads for safe shutdown, to maintain a safe shutdown condition, and to operate the ESF equipment following postulated accidents. Therefore, AC on-site power at degraded voltage meets all CLTR dispositions.

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2-136 2.3.3.2 AC On-Site Power (Normal Operation) As explicitly stated in Section 6.1 of the CLTR, the increase in thermal power translates to an increased electrical output from the station. The existing protective relay settings are adequate to accommodate the increased load on the non-safety 34.5 kV, 6.9 kV and 4.16 kV systems. Selective coordination is maintained between the pump motor breakers and the 34.5 kV, 6.9 kV and 4.16 kV switchgear main feeder breakers. The existing protective relay settings for reactor recirculation pump motors were based on a 7,940 hp rating. Because the reactor recirculation pump motor brake horsepower (bhp) required during EPU operation is approximately 7,554 hp, the reactor recirculation pump motor feed cables are adequate for EPU. The existing protective relay settings for condensate booster pump motors were based on a 2,500 hp rating. Because the condensate booster pump motor bhp required during EPU operation is approximately 2,111 hp, the condensate booster pump motor

feed cables are adequate for EPU. The existing protective relay settings for condensate pump motors were based on a 1,750 hp rating. Because the condensate pump motor bhp required during EPU operation is approximately 1,403 hp; the condensate pump motor feed cables are adequate for EPU. Load flow, voltage drop, and short circuit current calculations were performed to verify the adequacy of the AC on-site power system for th e proposed changes. Analyzed EPU bhp loads as discussed above are within the electrical distribution equipment capabilities (i.e., service transformers, BOP transformers, ACT transformers, bus ducts). The running and starting voltages for motors are within the acceptabl e values (i.e., 90% r unning and 80% starting voltages). Therefore, AC on-site power during normal operation meets all CLTR dispositions.

Conclusion Entergy has reviewed the effects of the proposed EPU on the AC on-site power system and concludes the effects of the proposed EPU on the system's functional design have been adequately evaluated. Entergy further concludes the AC on-site power system will continue to meet the requirements of GDC-17 following implem entation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the AC on-site power system.

2.3.4 DC On-Site Power System Regulatory Evaluation The DC on-site power system includes the DC pow er sources and their distribution and auxiliary supporting systems that are provided to supply motive or control power to safety-related equipment. The review covered the information, analyses, and referenced documents for the DC NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-137 on-site power system. The regulatory acceptance criteria are based on GDC-17, insofar as it requires the system to have the capacity and capability to perform its intended functions during

AOOs and accident conditions. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The DC on-site power system is described in UFSAR Section 8.3.2, "DC Power Systems."

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.2 of the CLTR addr esses the effect of C PPU on DC Power. The results of this evaluation are described below.

The GGNS direct current (DC) power distribution system provides control and motive power for various systems/components within the plant. The DC power loading requirements in the

UFSAR were reviewed and no reactor power-dependent loads were identified. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result DC Power Requirements [[

     ]] Meets CLTR Disposition As stated in Section 6.2 of the CLTR, [[

     ]]   The DC power system provides DC power to protective relaying, control, instrumentation and other DC loads. This system includes the stati on batteries, battery chargers, and DC distribution system. The DC power system is divided into two distinct categories. The components of the DC system and the loads that are required to sa fely shutdown the reactor in case of a DBA are designated nuclear safety-related or class 1E; the others are non-safety related or non-Class 1E. In both the normal and emergency operating conditions, the DC loads are summarized in the battery sizing calculations based on the equipment ratings. These loads are used as inputs to determine the batteries and battery chargers sizing, voltage drop and short circuit current calculations.

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2-138 There will be small DC load changes in the non-safety related batteries due to a proposed modification to the Radial Wells. This DC lo ad addition is acceptable relative to the existing approximately 4.2% margin in the non-safety related batteries affected (1G3 and 2G3). Table 2.3-5 provides the battery margins at C LTP and EPU. The expected EPU electrical modification that could affect DC loads is the increase in control power load to the non-Class 1E 125V DC 1G3 and 2G3 batteries and chargers due to the Radial Well addition. The 125V DC battery margin will be reduced by no more than 2.9% of the total battery capacity due to additional loads from EPU modifications. There are no changes to the loads on the safety-related 125V DC batteries. Additionally, no changes to the margin for the non-safety related 250V DC battery loads will result from EPU modifications. The DC on-site power system changes remain bounded by battery capacity. The safety-related DC power system is not affected by EPU conditions and therefore, the DC on-site power system meets all CLTR dispositions. Conclusion Entergy has reviewed the assessment of the e ffects of the proposed EPU on the DC on-site power system and concludes the effects of the proposed EPU on the system's functional design have been adequately evaluated. Entergy further concludes the DC on-site power system will continue to meet the requirements of GDC-17 following implementation of the proposed EPU. Adequate physical and electrical separation exists and the system has the capacity and capability to supply power to all safety loads and other required equipment. Therefore, Entergy finds the proposed EPU acceptable with respect to the DC on-site power system.

2.3.5 Station

Blackout Regulatory Evaluation The term SBO refers to a complete loss of AC electric power to the essential and nonessential switchgear buses in a nuclear power plant. SB O involves the LOOP concurrent with a turbine trip (TT) and failure of the on-site emergency AC power system. SBO does not include the loss of available AC power to buses fed by station batteries through inverters or the loss of power from "alternate AC sources" (AAC). The review focused on the effect of the proposed EPU on the plant's ability to cope with and recover from an SBO event for the period of time established in the plant's licensing basis. The regul atory acceptance criteria for SBO are based on 10 CFR 50.63. GGNS Current Licensing Basis

SBO is described in UFSAR Appendix 8A, "Loss of All Alternating Current Power (Station Blackout)." NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-139 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 9.3.2 of the CLTR addr esses the effect of CPPU on SBO. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topic addressed in this evaluation is: Topic CLTR Disposition GGNS Result Station Blackout [[

     ]] Meets CLTR Disposition The CLTR states that the plant responses to and coping abilities for an SBO event are affected slightly by operation at the power uprate level, due to the increase in the decay heat.

SBO was re-evaluated using the guidelines of NUMARC 87-00, "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors" (Reference 40), and RG 1.155, Station Blackout (Reference 41). [[

     ]] The major characteristics that affect the ability to cope with an SBO event are identified in NUMARC 87-00 Revision 1 as: 1. Condensate inventory for decay heat removal (DHR)

2. Class 1E battery capacity

3. Compressed air capacity

4. Effects of loss of ventilation

5. Containment isolation By satisfying the criteria used in assessing the a bove characteristics, the plant is able to show satisfactory response to an SBO event.

NUMARC 87-00 Revision 1 (Section 7) provides two methods for conducting the assessment. The first method, the AC independent approach, is used in the GGNS SBO assessment. In the AC independent approach, the plant relies on available process steam, DC power and compressed air to operate equipment necessary to achieve and maintain hot shutdown. The NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-140 four-hour coping duration criteria for AC inde pendent plants applies to GGNS. Thus, GGNS must meet the SBO requirements for at least four hours. Condensate Inventory for Decay Heat Removal

Analyses have shown that the GGNS condensate inventory is adequate to meet the SBO coping requirement for EPU conditions. The SBO evalua tion at EPU conditions shows a need for an additional approximately 24.4%, over CLTP, of CST water for RCIC use to ensure that adequate water volume is available to remove decay heat, depressurize the reactor, and maintain reactor vessel level above the top of active fuel (TAF). This increases the total CST volume required to approximately 136,014 gallons, which is within the current CST inventory reserve of 143,000 gallons. Class 1E Battery Capacity

Evaluation of the GGNS Class 1E battery capacity has shown that GGNS has adequate battery capacity to support DHR during an SBO for the required coping duration. The battery capacity remains adequate to support RCIC operation after EPU.

Compressed Air Capacity GGNS meets the requirement for compressed air capacity. An evaluation has shown that the GGNS air operated SRVs required for DHR have sufficient compressed air for the required automatic and manual operation during the SBO ev ent for EPU conditions. Sufficient capacity remains to perform emergency RPV depressu rization in case it is required. Adequate compressed air capacity exists to support the SRV actuations because the maximum number of SRV valve operations is less than the capacity of the pneumatic supply. Effects of Loss of Ventilation Areas containing equipment necessary to cope with an SBO event were evaluated for the effect of loss of ventilation due to an SBO. The evaluation shows that equipment operability is maintained because the SBO environment is m ilder than the existing design and qualification bases. These areas for GGNS included:

1. DW 2. Steam Tunnel

3. RCIC Room

4. Control Room (CR) and Upper Cable Spreading Room NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-141 5. Switchgear Room / Inverter Room Containment Isolation Containment isolation capability is not adversely affected by the SBO event for EPU.

SBO Sequence of Events In order to evaluate the relative changes in key parameters from EPU, the SBO calculations for both CLTP and EPU conditions are performed us ing the NRC-accepted SHEX analysis code and nominal ANSI/American Nuclear Society (ANS) 5.1-1979 Decay Heat Source Term consistent with the recommendations of SIL 636 at 100% rated power for Containment Long-Term Pressure and Temperature Analysis (Reference 8). The SBO sequence of events is given in

Table 2.3-6. Key Parameters

For the evaluation of the effect of EPU on the SB O results, the event was analyzed at both CLTP and EPU conditions using SHEX and a more conservative scenario than was originally used. The SHEX analyses model a scenario that incl udes a partial depressurization of the reactor vessel near the end of the event; the current licensing basis approach maintains the reactor pressurized. The SHEX analyses provide a consiste nt basis for: (1) evaluation of the effects of EPU on the SBO response, and (2) comparison of the results, which are presented in the following table. Key Containment Parameters Comparison Parameter Units CLTP EPU Design Limit DW Pressure psia 24.0 24.6 <44.7 SP Temperature ºF 186.9 200.1 <210 2 Notes: 1. The current licensing basis SBO results are 20.6 psia and 177ºF, for DW pressure and SP temperature, respectively. 2. The design limit for the SP temperatur e has been increased for the EPU project as discussed in Section 2.6.1. The containment response comparison is base d on a scenario that provides conservative containment parameters as compared to the SBO procedures that are used for reactor pressure control. Based on the above evaluations, GGNS continues to meet the requirements of 10 CFR 50.63 after the EPU. Therefore, SBO meets all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-142 Conclusion Entergy has reviewed the assessment of the eff ects of the proposed EPU on the plant's ability to cope with and recover from an SBO event for the period of time established in the plant's licensing basis. Entergy concludes: (1) the effects of the proposed EPU on SBO have been adequately evaluated; and (2) the plant will continue to meet the requirements of 10 CFR 50.63 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to SBO. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-143 Table 2.3-1 Group II Partially Qualified Components Item Equipment ID EQDP 1 Component EQ Zone Test Level (rads) Rad Life (Years) Normal Dose Rate (rads/hr) EPU 100 Day Accident Dose (w/10% Margin) (rads) 1 1B21F505A EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 2 1B21F505B EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 3 1B21F505C EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 4 1B21F505D EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 5 1B21F505E EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 6 1B21F505F EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 7 1B21F505G EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 8 1B21F505H EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 9 1B21F505J EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 10 1B21F505K EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 11 1B21F505L EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 12 1B21F505M EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 13 1B21F505N EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 14 1B21F505P EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 15 1B21F505R EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 16 1B21F505S EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 17 1B21F505T EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 18 1B21F505U EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 19 1B21F505V EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 20 1B21F505W EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 21 1B21F506A EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 22 1B21F506B EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 23 1B21F506C EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 24 1B21F506D EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 25 1B21F506E EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 26 1B21F506F EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 27 1B21F506G EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 28 1B21F506H EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-144 Item Equipment ID EQDP 1 Component EQ Zone Test Level (rads) Rad Life (Years) Normal Dose Rate (rads/hr) EPU 100 Day Accident Dose (w/10% Margin) (rads) 29 1B21F506J EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 30 1B21F506K EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 31 1B21F506L EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 32 1B21F506M EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 33 1B21F506N EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 34 1B21F506P EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 35 1B21F506R EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 36 1B21F506S EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 37 1B21F506T EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 38 1B21F506U EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 39 1B21F506V EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 40 1B21F506W EQ27.1 Silicone Oil 1A112 2.92E+07 5.9 121 2.29E+07 Note: 1. EQDP - Environmental Qualification Documentation Package

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2-145 Table 2.3-2 Group III, Non-Qualified Components Component ID EQDP 1 Manufacturer Model Env Zone IC/OC 2 Test Level (Rads) EPU TID (Rads) Remarks 1E12F003A EQ02.2 Limitorque Corp SMB-4-3001A102 OC 2.00E+07 2.32E+07 RHR HX A Outlet VLV; Modification to

replace Reliance Motor, Durez switches and

Scotch tape splices. 1E12F003B EQ02.2 Limitorque Corp SMB-4-3001A106 OC 2.00E+07 2.37E+07 RHR HX B Outlet VLV; Modification to

replace Reliance Motor and Durez switches. 1E12F006A EQ02.2 Limitorque Corp SMB-2-60 1A203 OC 2.00E+07 2.89E+07 RHR Pump A Suction from SDC;

Modification to repl ace Electric Apparatus Motor and Durez switches. 1E12F006B EQ02.2 Limitorque Corp SMB-2-60 1A205 OC 2.00E+07 2.99E+07 RHR Pump B Suction from SDC;

Modification to repl ace Electric Apparatus Motor and Durez switches. 1E12F024A EQ02.2 Limitorque Corp SB-2-80 1A103 OC 2.00E+07 2.73E+07 RHR A Test Return to SP; Modification to

replace Electric Apparatus Motor, Durez switches and Scotch tape splices. 1E12F024B EQ02.2 Limitorque Corp SB-2-80 1A105 OC 2.00E+07 2.73E+07 RHR B Test Return to SP; Modification to

replace Electric Apparatus Motor, Durez switches and Scotch tape splices. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-146 Component ID EQDP 1 Manufacturer Model Env Zone IC/OC 2 Test Level (Rads) EPU TID (Rads) Remarks 1E12F040 EQ02.2 Limitorque Corp SMB-000-51A128 OC 2.00E+07 4.13E+07 RHR to Radwaste Outboard Shutoff VLV;

Modification to replace Reliance Motor and

Durez switches. 1E12F047A EQ02.2 Limitorque Corp SMB-2-80 1A303 OC 2.00E+07 2.64E+07 RHR HX A Inlet VLV; Modification to

replace Electric Apparatus Motor and Durez

switches. 1E12F047B EQ02.2 Limitorque Corp SMB-2-80 1A307 OC 2.00E+07 2.61E+07 RHR HX B Inlet VLV; Modification to

replace Electric Apparatus Motor and Durez

switches. 1E12F048A EQ02.2 Limitorque Corp SMB-4-3001A128 OC 2.00E+07 4.85E+07 RHR HX A Bypass VLV; Modification to

replace Reliance Motor, Durez switches and

Scotch tape splices. 1E12F048B EQ02.2 Limitorque Corp SMB-4-3001A129 OC 2.00E+07 4.98E+07 RHR HX B Bypass VLV; Modification to

replace Reliance Motor and Durez switches. 1E12F049 EQ02.2 Limitorque Corp SMB-000-51A128 OC 2.00E+07 2.62E+07 RHR to Radwaste Inboard Shutoff VLV;

Modification to replace Reliance Motor and

Durez switches. 1E12F064A EQ02.2 Limitorque Corp SB-00-10 1A103 OC 2.00E+07 2.73E+07 RHR A Minimum Flow to SP; Modification

to replace Reliance Motor and Durez

switches. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-147 Component ID EQDP 1 Manufacturer Model Env Zone IC/OC 2 Test Level (Rads) EPU TID (Rads) Remarks Commodity, Wire EQ13.1 Rockbestos Co Rocktherm 1A420 OC 1.20E+05 1.64E+05 Two (2) devices to be addressed.

Modification to replace DW Hydrogen Analyzer Hot Box Hookup Wire in Panels

1E61J001A-H and 1E61J002A-H Commodity, Electrical Splice EQ18.1 Electro Products Scotch 1A103 1A105 OC 2.05E+07 6.11E+07 Two (2) devices to be addressed.

Modification to re place Scotch taped electrical splices for RHR Jockey Pump motors 1E12C003A (1A103) and B (1A105) Notes: 1. EQDP - Environmental Qualification Documentation Package 2. IC: Equipment located inside containment; OC: Equipment located outside containment

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-148 Table 2.3-3 Electrical Equipment Ratings and Margins Component Component Rating CLTP Duty CLTP Margin (%) EPU Duty EPU Margin(%) Main Generator (MVA Capability / PF) 1,525 / 0.9 (CLTP) 1,600 / 0.9 (EPU) 1,525 / 0.89 (Note 1) 0 (Note 1) 1,600 / 0.95 (Note 1) 0 (Note 1) IPB Duct (Amps) 42,500 (CLTP) (Note 3) (EPU) 42,128 (CLTP) 0.88 44,200 (Note 2) (Note 3) Main Step-Up Transformers (MVA) 1,530 (CLTP) 1,650 (EPU) 1,525 0.3 1,600 3.03 (Note 4) Service Transformer 11 (Maximum MVA) 90 40.54 54.96 42.64 52.6 Service Transformer 21 (Maximum MVA) 90 43.54 51.62 44.15 50.94 Notes: 1. Based on the winter base case (CLTP) and winter EPU case heat balance calculations 2. Current at 95% generator voltage and 1,600 MVA. 3. The IPB duct cooling will be increased to suppor t current > 44,200A when voltages are less than 100%. 4. Margin based on replacement main step-up transformers.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-149 Table 2.3-4 Electrical Distribution System Load Increases Required BHP Analyzed BHP Motor Description Nameplate hp CLTP EPU CLTP EPU Condensate Booster Pump 2,500 1,927 2,111 2,500 2,500 Condensate Pump 1,750 1,316 1,403 1,750 1,750 Reactor Recirculation Pump (100% Core Flow) 7,940 7,511 7,554 7,940 7,940 Load Increases from Modifica tions (Non-Safety Related) Modification Description Units Value hp 2 motors x 600 Radial Well Addition Voltage (V) 4,000 hp 8 motors x 200 ACT Expansion Voltage (V) 460

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2-150 Table 2.3-5 Battery Margin at CLTP and EPU Battery Margin (%) Battery Description CLTP EPU 125V DC Division I Battery Note 1 125V DC Division II Battery Note 1 125V DC Division III Battery Note 1 125V DC Non-Class 1E Batteries 1G3 & 2G3 4.2 Note 2 1.3 Note 2 250V DC Non-Class 1E Battery Note 1 Notes: 1. There are no changes to these battery systems due to EPU. 2. Loading shown is identical for batteries 1G3 and 2G3.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-151 Table 2.3-6 GGNS SBO Sequence of Events Time (sec) Description of Event ~0 LOOP Reactor scram MSIV start to close Loss of Feedwater (LOFW) Loss of Service Water RCIC available to mainta in reactor water level 3.5 MSIV closed 5 Feedwater flow stops ~6.9 to 7.6 SRVs open (relief mode) 105.7 Begin RCIC Injection w/ CST suction ~140 Begin RPV pressure control using subsequent SRV operation 2,315.3 End RCIC Injection 3,342.8 Begin RCIC Injection 4,648.9 End RCIC Injection 6,242.8 Begin RCIC Injection 7,750.9 End RCIC Injection 9,188.6 Begin RCIC Injection 11,568.8 End RCIC Injection 11,771.0 Begin RCIC Injection 14,400 End of Coping Period

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2-152 Figure 2.3-1 Worst Case EQ Enveloping Accident Temperature Profiles (All Plant EQ Zones and Elevations) Note: The Auxiliary Building and Containment enveloping EQ temperature profiles showns are a composite of the zone-specific profil es and includes the HELB analysis for EPU. Zone-specific temperature profiles were used for equipment qualification in the Auxiliary Building. J J V I I ! , ! ,

g > T , 0 §m I I V I -'-_T / i' I z 9 5 0 " > ! " < 0
0 " Q < ! 0 0 * , * , & I I z 0 , ,-! i ! , , , I--f-1----, Q "-0 , '---i I z 0
z g * " 9 > z * * > -i 0 , ! z z i Q Z 0 Q , , --, , ---j u " < Q i 0 ,
Q Q Z ! < < * *
e e e , u u u I I , ! , "

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2-153 2.4 Instrumentation and Controls

2.4.1 Reactor

Protection, Safety Features Actuation, and Control Systems Regulatory Evaluation Instrumentation and control (I&C) systems are pr ovided to: (1) control plant processes having a significant effect on plant safety; (2) initiate the reactivity control system (including control rods); (3) initiate the ESF systems and essential auxiliary supporting systems; and (4) achieve and maintain a safe shutdown condition of the plant. Diverse I&C systems and equipment are provided for the express purpose of protecting against potential common-mode failures of I&C protection systems. Entergy conducted a review of the reactor trip system, engineered safety feature actuation system (ESFAS), safe shutdown systems, control systems, and diverse I&C systems for the proposed EPU to ensure the systems and any changes necessary for the proposed EPU are adequately designed such that the systems continue to meet their safety functions. Entergy's review was also conducted to ensure failures of the systems do not affect safety functions. The regulatory acceptance criteria rela ted to the quality of design of protection and control systems are based on 10 CFR 50.55a(a)(1), 10 CFR 50.55a(h), and GDCs 1, 4, 13, 19, 20, 21, 22, 23, and 24. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The I&C systems are described in UFSAR Chapter 7.0, "Instrumentation and Control Systems." GGNS submitted an LAR regarding implementati on of the Power Range Neutron Monitoring System and Stability Option III. The EPU evaluation considered the new system design and licensing basis. Technical Evaluation The setpoint calculation methodology, safety limit-related LSSS determination, and instrument setpoint controls are discussed in this section. A sample calculation is provided in Section 2.4.2 of this report.

NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 5 of the CLTR addr esses the effect of C PPU on Reactor Protection, Safety Features Actuation, and Control Systems. The results of this evaluation are described

below. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-154 2.4.1.1 Nuclear Steam Supply System Monitoring and Control Instrumentation As stated in Section 5.1 of the CLTR, the instruments and controls used to monitor and directly interact with or control reactor parameters ar e usually within the NSSS. Changes in process variables and their effects on instrument performance and setpoints were evaluated for EPU operation to determine any related changes. Process variable changes are implemented through changes in normal plant operating procedures. TSs address instrument AVs and/or setpoints for those NSSS sensed variables that initiate pr otective actions. The effects of EPU on TS instrument functions are addressed in Section 2.4.1.3. EPU affects the performance of the Neutron Monitoring System. These performance effects are associated with the Average Power Range Monitors (APRMs), IRMs, Local Power Range Monitors (LPRMs), and Rod Control and Information System (RCIS). GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Average Power Range Monitors, Intermediate Range Monitors and Source Range Monitors [[ Meets CLTR Disposition Local Power Range Monitors

Meets CLTR Disposition Rod Control and Information System

     ]] Meets CLTR Disposition 2.4.1.1.1 Average Power Range Monitors, Intermediate Range Monitors and Source Range Monitors The CLTR states that at rated power, the increase in power level increases the average flux in the core and at the in-core detectors. The APRM power signals are calibrated to r ead 100% at the new licensed power (i.e., EPU RTP). EPU has little effect on the IRM overlap with the SRMs and the APRMs. Using normal plant surveillance procedures, the IRMs may be ad justed, as required, so that overlap with the SRMs and APRMs remains adequate.

The SRM, IRM, and APRM systems installed at GGNS are in accordance with the requirements established by the GEH design specifications. The specifications provide confirmation that the APRM, IRM and SRM systems meet all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-155 2.4.1.1.2 Local Power Range Monitors The CLTR states that at rated power, the incr ease in power level increases the flux at the LPRMs. The average flux experienced by the detectors incr eases due to the average power increase in the core. The maximum flux experienced by an LPRM remains approximately the same because the peak bundle power does not appreciably increase. Due to the increase in neutron flux experien ced by the LPRMs and traversing in-core probes (TIPs), on average, the neutronic life of the LPRM detectors will decrease and the radiation levels of the TIPs will increase. LPRMs are designed as replaceable components. The LPRM accuracy at the increased flux is within specified limits, and LPRM lifetime is an operational consideration that is handled by routine replacement. A small increase in radiation levels is accommodated by the radiation protection program for normal plant operation. Reliability of LPRM instrumentation and accurate prediction of in-bundle pin powers typically requires operation with bypass voids lower than 5% at nominal conditions. A discussion of the steady-state 5% bypass voiding evaluation is provided in Section 2.8.2.4.1. The LPRMs and TIPs installed at GGNS are in accordance with the requirements established by the GEH design specifications. The specifications provide confirmation that the LPRMs meet all CLTR dispositions. 2.4.1.1.3 Rod Control and Information System The CLTR states that the incr ease in power level could change the power level at which rod patterns are enforced by the RCIS. The RCIS is a normal operating system that does not perform a safety-related function. The RCIS Rod Pattern Controller (RPC) is a safety-related system that supports the operator by enforcing rod patterns until reactor power has reached appropriate levels. The RCIS also provides rod position information to the operator. The RCIS Rod Withdrawal Limiter prevents excessive control rod withdrawal after reactor power has reached an appropriate level. Therefore, no additional plant-specific information for the performance of this system relative to the normal operational function is required. The power-dependent instrument setpoints for the RCIS RPC are included in the plant TSs (see Section 2.4.1.3). The RCIS installed at GGNS provides the same level of protection as described in the CLTR, because the thermal power increase of 575 MWt from OLTP (3,833 MWt) to EPU (4,408 MWt) is < 20% of OLTP, the Banked Position Withdrawal Sequence (BPWS) is used, and the EPU Low Power Setpoint (LPSP) lower AV is maintained at the same absolute thermal power as the CLTP value (389.8 MWt), corresponding to a lower percent thermal power. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-156 Therefore, the RCIS meets all CLTR dispositions. 2.4.1.2 BOP Monitoring and Control

As stated in Section 5.2 of the CLTR, operation of the plant at EPU conditions has minimal effect on the BOP system I&C devices. Based on EPU operating conditions for the power conversion and auxiliary systems, most process control valves and instrumentation have sufficient range/adjustment capability for use at the EPU conditions. However, some (non-safety) modifications are needed to the power conversion systems to obtain EPU RTP (see Table 2.4-2). GGNS meets all CLTR dispositions. The topics considered in this section are: Topic CLTR Disposition GGNS Result Pressure Control System (PCS) [[ Meets CLTR Disposition Turbine Bypass System (Normal Operation)

Meets CLTR Disposition Turbine Bypass System (Safety Analysis)

Meets CLTR Disposition Feedwater Control System (Normal Operation)

Meets CLTR Disposition Feedwater Control System (Safety Analysis)

Meets CLTR Disposition Leak Detection System

     ]] Meets CLTR Disposition 2.4.1.2.1 Pressure Control System

The CLTR states that the increase in power level increases the steam flow to the turbine. The PCS is a normal operating system to provide fast and stable responses to system disturbances related to steam pressure and flow changes to control reactor pressure within its normal operating range. This system does not perform a safety function. Pressure control operational testing is included in the EPU implem entation plan as described in Section 2.12 to ensure adequate turbine control valve (TCV) pressure control and flow margin is available. The PCS at GGNS meets all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-157 2.4.1.2.2 Turbine Bypass System The CLTR states that the bypass system capacity, in terms of mass flow, is not changed for EPU. As a result, at EPU, the bypass steam flow as a percentage of total steam flow decreases due to the increase in the total steam flow. The turbine bypass system (TBS) is not essential for turbine operation and is not credited in any limiting events analyses. The TBS is a normal operating system that is used to bypass excessive steam flow. This system is non-safety related. The flow capacity of the bypass system, 5.77 Mlbm/hr, is not changed. The bypass flow capacity is included in the rod withdrawal error (RWE) and loss of feedwater heating (LFWH) events. The AOO events are discussed further in Section 2.8.5. The limiting

events at GGNS do not credit operation of the turbine bypass valves. The TBS is being added to

the TSs. The TBS at GGNS meets all CLTR dispositions. 2.4.1.2.3 Feedwater Control System The CLTR states that the increase in power results in an increase in FW flow. The FW Control System is a normally operating system to control and maintain the reactor vessel water level. EPU results in an increase in FW flow. FW control operational testing is included in the EPU implementation plan as described in Section 2.12 to ensure that the FW response is acceptable. Failure of this system is evaluated in the reload analysis for each reload core with the FW Controller Failure - Maximum Demand event. An LOFW event can be caused by failure of the FW controller. The LOFW is discussed in Section 2.8. The FW Control System at GGNS meets all CLTR dispositions. 2.4.1.2.4 Leak Detection System

The CLTR states that the only effect on the LDS due to EPU is a slight increase in the FW system temperature and increase in the steam flow.

[[

       ]]    MST in the Auxiliary Building and in the Turbine Building: The FW temperature at EPU conditions results in no increase in the MST temperature. The LDS temperature and differential temperature setpoints remain unchanged. As a result, the MST temperature setpoint is conservative, because it slightly increases leak detection sensitivity.

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2-158 DW: The normal operating DW area temperature experiences no change for EPU conditions. RWCU: There is no significant change to the RWCU system temperature, pressure, or flow; therefore, the RWCU LDS is not affected. RCIC: There is no change to the RCIC system temperature, pressure, or flow; therefore, the RCIC LDS is not affected. RHR Shutdown Cooling (SDC) Mode: There is no increase in the RHR SDC Mode temperature or pressure; therefore, the RHR system LDS is not affected. The flow-based LDS is not affected by EPU, with the exception of MS L high flow. MSL high flow is discussed in Section 2.4.1.3. The LDS at GGNS meets all CLTR dispositions [[

     ]] 2.4.1.3 Technical Specification Instrument Functions

As stated in Section 5.3 of the CLTR, AVs and/or Nominal Trip Setpoints (NTSPs) are those sensed variables that initiate protective actions and are generally associated with the safety analysis. The safety analysis generally establishes the ALs. The determination of the AVs and other instrument setpoints includes consideration of measurement uncertainties and is derived from the ALs. The settings are selected with sufficient margin to minimize inadvertent initiation of the protective action, while assuring that adequate operating margin is maintained between the system settings and the actual limits. There is typically substantial margin in the safety analysis process that should be considered in establishing the setpoint process used to establish the TS

AVs and other setpoints. Increases in the core thermal power and steam flow affect some instrument setpoints. These setpoints are adjusted to maintain comparable differences between system settings and actual limits, and reviewed to ensure that adequate operational flexibility and necessary safety functions are maintained at the EPU RTP level. Where the power increase results in new instruments being employed, an appropriate setpoint calculation is performed and TS and/or TRM changes are implemented, as required. [[

     ]] [[

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2-159

                                                                                                                                                          ]]  The justification for implementing this simplified pro cess for the individual TS and/or TRM setpoints is provided for each instrument below, as applicable. Implementing the constant maximum operating pressure requirement for EPU [[

     ]]   In addition, the following restrictions are imposed on the use of the simplif ied process to assure its validity. Its use is limited to:  NRC approved GEH or plant-specific methodology.  [[

     ]] These restrictions are satisfied for GGNS, except where instrumentation is changed affecting the instrumentation errors. Modifications to the HP turbine affect the Turbine First-Stage Pressure (TFSP) Scram Bypass Permissive, RCIS RPC LPSP (lower and upper bounds), and RCIS Rod Withdrawal Limiter (RWL) High Power Setpoi nt (HPSP) functions.

The new Power Range Neutron Monitoring (PRNM) equipment affects the APRM Flow Biased STP Scrams and Rod Blocks (TLO and SLO) and APRM Neutron Flux Setdown Scram and Rod Block (in startup mode) functions. For these setpoint functions, new setpoints must be re-established using the GEH methodology per Reference 42. Table 2.4-1 summarizes the current and EPU ALs and AVs for GGNS.

The Setpoint Calculation Methodol ogy and the setpoint value for each topic addressed in this section meet all CLTR dispositions. The topics considered in this section are: Topic CLTR Disposition GGNS Result Main Steam Line High Flow Isolation - Setpoint

Calculation Methodology [[ Meets CLTR Disposition Main Steam Line High Flow Isolation - Setpoint

Value Meets CLTR Disposition Turbine First-Stage Pressure Scram Bypass - Setpoint

Calculation Methodology

Meets CLTR Disposition

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2-160 Topic CLTR Disposition GGNS Result Turbine First-Stage Pressure Scram Bypass - Setpoint Value Meets CLTR Disposition APRM Flow Biased Scram - Setpoint Calculation

Methodology

Meets CLTR Disposition APRM Flow Biased Scram - Setpoint Value

Meets CLTR Disposition RCIS Rod Pattern Controller Low Power Setpoint -

Setpoint Calculation Methodology

Meets CLTR Disposition RCIS Rod Pattern Controller Low Power Setpoint -

Setpoint Value

Meets CLTR Disposition RCIS Rod Withdrawal Limiter High Power Setpoint -

Setpoint Calculation Methodology

Meets CLTR Disposition RCIS Rod Withdrawal Limiter High Power Setpoint -

Setpoint Value

Meets CLTR Disposition APRM Setdown in Startup Mode - Setpoint

Calculation Methodology

Meets CLTR Disposition APRM Setdown in Startup Mode - Setpoint Value

     ]] Meets CLTR Disposition 2.4.1.3.1 Main Steam Line High Flow Isolation The CLTR states that the effect on the MSL High Flow Isolation due to EPU is increased reactor power level and steam flow.

The MSL high flow isolation setpoint is used to initiate the isolation of the Group 1 primary containment isolation valves. The only safety anal ysis event that credits this trip is the Main Steam Line Break Accident (MSLBA). For this accident, there are diverse trips from high area temperature and high area differential temperatur e in the MST. For GGNS, there is sufficient margin to choke flow, so the AL for EPU is mainta ined at the current percent of rated steam flow in each MSL. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-161 For GGNS, the AL of 140% of rated steam flow is not changed and no new instrumentation is required (the existing instrumentation has the required upper range limit (URL) and calibration span that the instrument loops need to accomm odate the new setpoint). A new setpoint was calculated using the GEH methodology per Refere nce 42 and resulted in a change to the associated TS AV in terms of DP at the allowable steam flow. Therefore, the MSL High Flow Isolation setpoint meets all CLTR dispositions. 2.4.1.3.2 Turbine First-Stage Pressure Scram Bypass Permissive The CLTR states that the effect on the TFSP Scram Bypass Permissive due to EPU is increased reactor power level and potential change to TFSP. EPU results in an increased power level and the HP turbine modifications result in a change to the relationship of TFSP to reactor power level. The TFSP setpoint is used to reduce scrams and recirculation pump trips (RPTs) at low power levels where the TBS is effective for TTs and generator load rejections. In the safety analysis, this trip bypass only applies to events at low power levels that result in a TT or load rejection. [[

              ]] Because the HP turbine will be modified to support achieving the EPU RT P level, a new AV (in psig) must be established. The TS applicable condition is being changed consistent with the AL change above. The AV (in psig) for GGNS will be revised prior to EPU implementation.

To assure that the new value is appropriate, an EPU plant ascension startup test or normal plant surveillance is performed to validate that the actua l plant interlock is cleared consistent with the safety analysis. Therefore, the TFSP Scram Bypass Permissive meets all CLTR dispositions. 2.4.1.3.3 APRM Flow Biased Scram

This function is referred to in the GGNS TSs as the APRM Flow Biased Simulated Thermal Power - High function. The CLTR states that the effect on the APRM Flow Biased Scram due to EPU is increased reactor power level. APRM Flow Biased Simulated Thermal Power - High function is not associated with a limiting safety system setting. This operating limit for the operating domain is established to provide a pre-emptive scram and to prevent a gross violation of the licensed domain. Operation within this domain ensures compliance with GDC-12.

The GGNS TS AV for this function is being revised based on the methodology outlined in the CLTR. Therefore, a new setpoint was calculated using the GEH methodology per Reference 42, and the TS AV in percent RTP is being changed. Therefore, APRM Flow-Biased Scram at GGNS meets all of the CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-162 2.4.1.3.4 RCIS Rod Pattern Controller Low Power Setpoint The CLTR states that the effects on the RCIS RP C LPSP due to EPU are increased reactor power level, potential change to TFSP, and increased FW flow.

The RCIS RPC LPSP is used to bypass the rod pattern constraints established for the control rod drop accident (CRDA) at greater than a pre-estab lished low power level. Below this value, only banked position mode withdrawals or insertions ar e allowed as described in the UFSAR Section 15.4.9.2.2 description of the CRDA event. At GGNS, Entergy has elected to rescale the lower bound of the RCIS RPC LPSP to maintain the AL value in terms of ab solute power. The upper bound AL for the LPSP is not being changed for EPU. The power level measurement parameter is derived from TFSP. Because the HP turbine will be modified to support achieving the EPU RTP level, new AVs (both upper bound and lower bound) in units of psig must be established. The AVs (in psig) will be revised prior to EPU implementation. Therefore, the GGNS RCIS RPC LPSP meets all CLTR dispositions. 2.4.1.3.5 RCIS Rod Withdrawal Limiter High Power Setpoint The CLTR states that the effect on the RCIS Rod Withdrawal Limiter HPSP due to EPU is increased reactor power level. The RCIS Rod Withdrawal Limiter (RWL) system is only credited in the control RWE analysis. The TSs require the RWL to be enabled wh en reactor power is above the HPSP. [[

     ]] The RCIS RWL signal comes from TFSP. The HP turbine will be modified to support achieving the EPU RTP level. The RWL HPSP AL (in psig) will be revised prior to EPU implementation. The RCIS RWL setpoint (in psig) will be validat ed during power uprate plant ascension start-up testing to ensure the actual plant interlock is cleared consistent with the safety analysis. Therefore, the RCIS RWL HPSP meets all CLTR dispositions.

2.4.1.3.6 APRM Setdown in Startup Mode

This function is referred to in the GGNS TSs as the APRM Neutron Flux - High, Setdown function. The CLTR states that the effect on the APRM Set down in Startup Mode due to EPU is a reduced TS safety limit for reduced pressure or low core flow conditions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-163 No specific safety analyses take direct credit for this function; thus, there is no AL for this parameter. It indirectly ensures that react or power does not exceed 25% CLTP (21.8% EPU power) before the Mode Switch is placed in "R UN." The APRM setdown in the Startup Mode provides margin to the TS safety limit. Further, critical power tests demonstrated that the TS safety limit is conservative. A diverse trip is provided by the IRMs. The value for the TS safety limit for reduced pressure or low core flow condition is established to satisfy the fuel thermal limits monitoring requirements. Because the GGNS TS AV for this function is less than 21.8% RTP, no change is required. NTSP was calculated using the standard GEH met hodology per Reference 42. Therefore, APRM Setdown in Startup Mode at GGNS meets the CLTR disposition. 2.4.1.3.7 Main Steam Line Low Pressure Isolation in the Run Mode The Main Steam Line Low Pressure Isolation in the Run Mode pressure setpoint does not change in a power uprate. However, the steam line pressure near the turbine, where this sensor is located, is expected to change. The margin assessment confirmed that the remaining margin at EPU conditions is adequate for operation at EPU conditions and does not require limitations on the performance of surveillances with the existing setpoint. 2.4.1.4 Changes to Instrumentation and Controls

In the CLTR SER, the NRC staff stated that the EPU evaluation should go beyond the systems analysis and cover all EPU related changes to instrumentation and controls (e.g., setpoint and scaling changes, changes to upgrade obsolescent instruments, changes to control philosophy). The instrumentation changes identified by the GGNS EPU evaluation are provided in Table 2.4-2. No obsolescent instrumentation change s are required as a result of EPU, and there are no changes to instrument philosophy as a result of EPU.

Conclusion Entergy has reviewed the effects of the propos ed EPU on the functional design of the reactor protection system (RPS), ESFAS, safe shutdown system, and control systems. Entergy concludes that the effects of the proposed EPU on these systems have been adequately addressed and the changes necessary to achieve the proposed EPU are consistent with the plant's design basis. Entergy further concludes that the systems will continue to meet the requirements of 10 CFR 50.55a(a)(1), 10 CFR 50.55a(h), and GDCs 1, 4, 13, 19, 20, 21, 22, 23, and 24. Therefore, Entergy finds the proposed EPU acceptable with respect to instrumentation and controls. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-164 2.4.2 MSL High Flow Group 1 Isolation Instrument Setpoint Sample Calculation (Input/Output Document) 2.4.2.1 Function: MSL High Flow Group 1 Isolation Setpoint Characteristics Definition Comment(s) Event Protection: Limiting event for the setpoint: MSL Flow - High is provided to detect a break of the MSL and to initiate closure of the MSIVs. If the steam were allowed to continue flowing out of

the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Allowable Value is chosen to ensure that off-site dose limits

are not exceeded due to a break. Function After Earthquake Required Not Required Setpoint Direction Increasing Decreasing Single or Multiple Channel Single Multiple LER Calculation Basis if Multiple Channel Standard (Conservative) LER Calculation , or Configuration Specific LER Calculation Trip Logic for Configuration

Specific LER Calculation n/a LER: Licensee Event Report n/a: not applicable NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-165 Value/Equation Current Function Limits: Present Calculation (CLTP Conditions) Power Uprate Conditions Comment(s) Analytical Limit 182.0 psid 140% Rated Steam Flow 276.9 psid 140% flow Tech Spec Allowable Value 176.5 psid Setpoint 169 psid Operational Limit Not required. 202.03 psid 127% flow This present value contains additional margin to the calculated value for LER avoidance. Plant Data: Value Present Calculation Value Power Uprate Condition Sigma if not 2 Comment(s) Primary Element

Accuracy (APEA)
Drift (DPEA) 0.0 psid

negligible 0.75% of point = 7.60 psid

negligible

Drift: Comment

7. Process Measurement

Accuracy (PMA)

[[ ]] Random 0.0 psid

[[

     ]]  None (Use 0.0 psid)

[[ ]] Components (or Devices) in Setpoint Function Instrument Loop:

Venturi Flow Nozzle (Flow Restrictor)
Flow Transmitter
Analog Trip Module (Trip Unit)

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2-166 2.4.2.2 Components 2.4.2.2.1 Flow Transmitter (Differential Pressure Transmitter) Component Information: Value/Equation Comment(s) Plant Instrument ID No. E31-N086C, N089A-D E31-N086A,B,D; NO87A-D; and NO88A-

D Instrument vendor Rosemount Model ID No. (including Range Code) 1153 DB 7 RC (Range Code 7) 1152 DP 7E22T028OPB (Range Code 7) Plant Location(s) Room 1A313, Panels 1H22P015 and 1H22P042; Room 1A311, Panels, 1H22P025 and 1H22P041 Process Element Venturi Inputs: Vendor Specifications: Value / Equation Sigma if not 2 Comment(s) Primary Model ID 1153DB7RC 1152DP7E22T028OPB n/a Top of Scale 300 psid (20 mAdc) 300 psid (20 mAdc) n/a Comment 8. Bottom of Scale 0 psid (4 mAdc) 0 psid (4 mAdc) n/a Comment 8. Upper Range Limit 300 psid 300 psid n/a Comment 8. Accuracy 0.25% of calibration span. Comment 17. Temperature Effect {[(0.75% URL) + (0.5 % span)] / 100 deg F ambient temperature change}*(dT) Seismic Effect 0.5% of URL during and after a seismic disturbance defined by a required response spectrum with a ZPA of 4.0 g's. 0.25% of URL during and after seismic disturbance to 3.0 g over a range of 5-100Hz in three major axes. Comment 22. Radiation Effect 4.0% URL during and after exposure to 2.2x10 7 Rads, TID Gamma radiation (no dose rate specified). 8.0% of URL during and after exposure to 5.0x10 6 Rads, TID Gamma. Comment 23. Humidity Effect 0-100% relative humidity (NEMA 4X). Sealed Unit - no effect Power Supply Effect Less than 0.005% of output span/volt. RFI/EMI Effect Included in accuracy Comment 4. Insulation Resistance

Effect Negligible Comment 4. Over-pressure Effect Maximum zero shift after 2000 psi overpressure: 3% of URL ZPA: Zero Period Acceleration NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-167 Vendor Specifications: Value / Equation Sigma if not 2 Comment(s) Primary Model ID 1153DB7RC 1152DP7E22T028OPB n/a Mounting Position

Effect:

Span Effect
Zero Shift
No span effect.
Up to 1.5 inch H 2 O (372 MPa) (Ranges 3,4,and 5), which

can be calibrated out. Calibrated out. For higher ranges (includes Range Code 7) effect is superseded by accuracy specifications.

No span effect.
Up to 1 inch H 2 O, which can be calibrated out.

Calibrated Out. Static Pressure Effect

Zero Effect
Span Effect
Span Effect - Correction Uncertainty
Per 1000 psi: 0.5% of URL (Ranges 3, 6, 7 , and 8). Calibrated out.
The effect is systematic and can be calibrated out for a particular pressure before installation.

Calibrated out.

Correction uncertainty is 0.5% of input reading /per 1000 psi (6.89 MPa).
Per 2000 psi: 0.5% of URL (Ranges 3, 6, 7 , and 8). Calibrated out.
Effect is systematic and can be calibrated out for a particular

pressure before installation. Calibrated out.

Correction uncertainty is 0.25% of input reading /per 1000 psi (6.89 MPa).

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2-168 Plant Data: Value Comment(s) Calib Temperature Range 60 - 105 F Normal Temperature Range 60 - 105 F Trip Temperature Range 60 - 105 F Plant Seismic Value <3.0g Comment

22. Plant Radiation Value 3.5x10 2 Rads Gamma (40 yr TID); negligible (mild environment, does not need to function during a DBA) Comment 23. Plant Humidity Value 60% RH (normal) 100 % RH (DBE)

Power Supply Variation Value 4.5 Vdc Comment 14. RFI/EMI Value Not required Comment 4. Over-pressure Value 1325 psig Comment 10. Static Pressure Value 1325 psig Drift: Value Sigma if not 2 Comment(s) Current Calib. Interval 18 Mo. Includes extra 25% n/a Desired Calib. Interval 24 Mo. Includes extra 25% n/a EPU Calibration Interval for conservatism is 24 month to satisfy future

change to increase refueling interval from current 18 months. Drift Source Calculated Vendor n/a Drift Value 0.2% URL / 30 months 2 Sigma: Comment 9.

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2-169 Calibration: Value / Equation Sigma if not 3 Comment(s) As Left Tolerance +/-0.04 mA Comment 18. Leave Alone Tolerance +/-0.04 mA Comment 18. Input Calibration Tool: Heise Pressure Gauge Accuracy Accuracy equal to Flow Transmitter Accuracy 2 Comment 13. Resolution / Readability Not required Comment 2. Minor Division Not required n/a Comment 2. Upper Range Not required n/a Comment 2. Temperature Effect Not required Comment 2 and 3. Input Calibration Standard: Mensor Model 100 Calibration System Accuracy Accuracy equal to 1/4 of Flow Transmitter Accuracy. 2 Comment 15. Resolution / Readability Not required Comment 2. Minor Division Not required n/a Comment 2. Upper Range Not required n/a Comment 2. Temperature Effect Not required Comment 2 and 3. Output Calibration Tool: Fluke Model 45 Digital Multimeter Accuracy Accuracy equal to Flow Transmitter Accuracy. 2 Comment 13. Resolution / Readability Not required Comment 2. Minor Division Not required n/a Comment 2. Upper Range Not required n/a Comment 2. Temperature Effect Not required Comment 2 and 3. Output Calibration Standard: Fluke Model 5700A Accuracy Accuracy equal to 1/4 Flow Transmitter Accuracy. 2 Comment 15. Resolution / Readability Not required Comment 2. Minor Division Not required n/a Comment 2. Upper Range Not required n/a Comment 2. Temperature Effect Not required Comment 2 and 3. 2.4.2.2.2 Analog Trip Module (Trip Unit) Component Information: Value/Equation Comment(s) Plant Instrument ID No. 1E31-PDIS-N686A-D thru 689A-D Instrument vendor Rosemount Model ID No. (including Range Code) 510DU Plant Location(s) Room OC703, 1H13P691 (1E31-PDIS-N686A-N689A); Room OC504, 1H13P692 (1E31-PDIS-N686B-N689B); Room OC703, 1H13P693 (1E31-PDIS-N686C-N689C); and Room OC504, 1H13P694 (1E31-PDIS-N686D-N689D) Process Element n/a NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-170 Inputs: Vendor Specifications: Value / Equation Sigma if not 2 Comment(s) Top of Scale 20 mAdc (5 Vdc)/ +300 psid n/a Comment 19. Bottom of Scale 4 mAdc (1 Vdc)/ -0 psid n/a Comment 19. Upper Range Limit 300 psid n/a Accuracy Adjustability Trip Repeatability Analog Output 510DU 0.01 mAdc

+ 0.20% span / 6 months n/a   Temperature Effect Not required. Comment 4. Seismic Effect Not required. Comment 4. Radiation Effect Not required. Comment 4. Humidity Effect Not required. Comment 4. Power Supply Effect Not required. Comment 4. RFI/EMI Effect Not required. Comment 4. Insulation Resistance Effect Not required. Comment 4.

Over-pressure Effect n/a Comment 5. Static Pressure Effect n/a Comment 5. Plant Data: Value Comment(s) Calib Temperature Range 60 - 90 F Normal Temperature Range 60 - 90 F Trip Temperature Range 60 - 90 F Plant Seismic Value <2.0g Plant Radiation Value 1.82E2 Rads (40 yr TID Gamma). Plant Humidity Value 50 % RH Power Supply Variation Value Not required. Comment 4. RFI/EMI Value Not required. (Negligible). (mild environment) Comment 4. Over-pressure Value n/a Comment 5. Static Pressure Value n/a Comment 5. Drift: Value Sigma if not 2 Comment(s) Current Calib. Interval 92 days. Includes extra 25% n/a Comment 11. Desired Calib. Interval 92 days Includes extra 25% n/a Comment 12. Drift Source Vendor Calculated n/a Drift Value + 0.20% span / 6 months 2 Comments 2 and 20.

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2-171 Calibration: Value / Equation Sigma if not 3 Comment(s) As Left Tolerance +0.02 mA (0.125% span) Comment 18. Leave Alone Tolerance +0.02 mA (0.125% span) Comment 18. Input Calibration Tool: Fluke Model 45 Digital Multimeter, Transmation 1040 Digital Calibrator, and Heise Pressure Gauge Accuracy Accuracy equal to Analog Trip Unit (ATU) Accuracy 2 Comment 13. Resolution / Readability Not required Comment 2. Minor Division Not required n/a Comment 2. Upper Range Not required n/a Comment 2. Temperature Effect Not required Comment 2 and 3. Input Calibration Standard: Fluke Model 45 Digital Multimeter, Transmation 1040 Digital Calibrator, and Heise Pressure Gauge Accuracy Accuracy equal to 1/4 ATU Accuracy 2 Comment 15. Resolution / Readability Not required Comment 2. Minor Division Not required n/a Comment 2. Upper Range Not required n/a Comment 2. Temperature Effect Not required Comment 2 and 3. Output Calibration Tool: Light Emitting Diode (LED) Comment 16. Accuracy n/a Resolution / Readability n/a Minor Division n/a Upper Range n/a Temperature Effect n/a Output Calibration Standard: n/a Accuracy n/a Resolution / Readability n/a Minor Division n/a Upper Range n/a Temperature Effect n/a NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-172 2.4.2.3 Summary Results Calculated Values Setpoint Function Analytic Limit (from Section 1) Allowable Value Setpointf Meets LER Avoidance Criteria Meets Spurious Trip Avoidance Criteria MSL High Flow Group 1 Isolation 276.9 psid 140% flow 255.9 psid 254.7 psid Y Y f Excludes head correction. 2.4.2.4 Comments and Recommendations 1. Unless specifically identified as "bias" errors in this document, all instrument uncertainty errors are considered to be random in nature, even when the "+/-" symbol is not shown. 2. Some plant-specific information has not b een provided in the current GGNS setpoint calculation(s) or documents and is consider ed unnecessary because the effect of this information is included within the instrume nt accuracy values within the current GGNS setpoint calculation(s) or documents. 3. Unless specified otherwise, all calibration tool errors are considered to include resolution / readability errors and temperature effect errors. For the errors of assumed calibration standards, the temperature effect term is considered included in the accuracy of the assumed standard. The resolution/readability error for assumed calibration standards does not apply, because it is not actually read. 4. Temperature effect, radiation effect, seismic effect, humidity effect, power supply effect, Radio Frequency Interference/ Electromagne tic Interference (RFI/EMI) effect, and

Insulation Resistance Effect (IRE) errors are marked "negligible," "not provided," or "included in accuracy" and are considered to have negligible or no effect on the manufacturer's accuracy terms when they are not identified separately. 5. Per GE Instrument Setpoint Methodology (NEDC-31336P-A, Class 3, September 1996), overpressure effects are only applicable to certain pressure measurement devices (e.g., DP transmitters), and static pressure effects are only applicable to certain DP measurement devices. These effects are marked "n/a" for other devices. 6. Deleted (Not used). 7. The primary element accuracy (PEA) is assumed independent of time because it is assumed that the flow primary elements do not degrade with time (NEDC-31336P-A). Therefore, the drift component of the PEA is considered negligible. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-173 8. The current calibrated span for the Flow Transmitter is from negative (-) 50 psid to +250 psid, resulting in a Calibrated Span (SP) equal to 300 psid. This calibrated span encompasses the EPU AL of 276.9 psid when rescaled to the capability of the transmitter (0 to 300 psid) with its calibration adjustment needs. Therefore, the EPU AL, AV, and the NTSP are within instrument scale. 9. The current approach in Entergy's current setpoint calculation treats vendor data as 2 sigma values based on the vendors not stating the statistical basis for the numbers. GEH will treat them the same in this evaluation. 10. For conservatism, 1325 psig, the RPV Steam Dome Safety Limit is used as the possible over-pressure value for the Flow Transmitters. 11. The GGNS Surveillance Procedure requires additional functional test and calibration of Trip Units to be performed. Although the Su rveillance Procedure requires specific SRs to be performed every 18 months, 92 days is defined as the current calibration interval for the Trip Units based on the most restrictive portion of the GGNS Surveillance Procedure. 12. The GGNS Surveillance Procedure requires additional functional test and calibration of Trip Units to be performed. Although EPU Calibration Interval is desired for 24 months to satisfy future increase of refueling interval from current 18 months, the Trip Units are evaluated based on 92 days to satisfy the more restrictive SRs within the GGNS Surveillance Procedure. 13. For conservatism when data is not all av ailable, the Accuracy of the Input/Output Calibration Tools is assumed equal to the Flow Transmitter or Trip Unit Accuracy as

applicable. 14. For conservatism, and based on historical observations, a Power Supply voltage variation of +/- 4.5 Vdc is assumed for the Flow Transmitters. 15. Where a specific input was not provided on the Calibration Standards, an assumed inaccuracy ratio of one-quarter (i.e., 1/4) that of the Device inaccuracy is used. Accuracy ratio between the Lab Standards utilized to ca librate the device and the device under test is a minimum of 4 to 1 or, if less than 4 to 1 the actual ratio noted and a reason (basis) indicated as to why the 4 to 1 could not be met. 16. The analog output inaccuracy for the Trip Un it does not apply because there is no Slave Trip Unit to the trip setpoint signal to the Primary Containment Isolation System (PCIS), but instead a digital (discrete) output from the Master Trip Unit sent to a PCIS relay. 17. The accuracy for the Flow Transmitter (DP Transmitter) includes the combined effects of linearity, hysteresis, and repeatability. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-174 18. Because the calibration tolerance terminology at GGNS may differ slightly from GEH Instrument Setpoint Methodology, the definitions for the terms used in this document are provided here. a. As Left Tolerance (ALT): This is the to lerance within which the device calibration reading is left after calibration. b. Leave Alone Tolerance (LAT): This is the tolerance within which calibration need not be performed, and is intended to allow for normal variations in instrument readings due to accuracy and drift. 19. Span is 0-300 psid based on EPU AL value of 276.9 psid. A 250 ohm pr ecision resistor is used to develop 1 to 5 Vdc across this dropping resister.

20. [[

     ]] 21. Transfer functions used in this calculation: Flow Transmitter Output (mA) linearly converted from input (psid). Analog Trip Module Comparison of flow signal (mA, equivalent to psid). 22. GGNS pressure differential transmitter (PDT) 1153 seismic effect is treated the same as PDT 1152 based on plant conditions at their instrument locations. 23. GGNS radiation effect is negligible since the tr ip function of this loop is completed prior to seeing the accident environment condition.

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2-175 Table 2.4-1 Technical Specification Function Information Values Functions Current EPU APRM Calibration Basis (MWt) 3,898 4,408 APRM High Flux Scram (% RTP) AL 122 No Change 2 AV 120 119.3 8 APRM Flow Biased Scram (% RTP) TLO AV 1 0.65W + 62.9 3, 6 0.58W + 59.1 3, 9 SLO AV 1 0.65W + 42.3 3, 6 0.58W + 37.4 3, 5, 9 Clamp AV 1 113 No Change 2, 10 APRM Setdown in Startup Mode Scram AV 1 (%RTP) 20 No Change 2, 9 APRM Setdown Rod Block AV (%RTP) 14 No Change 2, 10 APRM Downscale Rod Block AV (%RTP) 3 No Change 2, 9 Rod Pattern Controller LP SP Lower AL and Upper Bounds AL (%RTP) 9 36 8 7 No Change 7 Rod Withdrawal Limiter HPSP (%RTP) AL 70 No Change 7 MSL High Flow Isolation (% rated steam flow) AL 140 No Change 11 MSL Low Pressure Isolation (in RUN Mode, psig) AL 825 No Change TFSP Scram Bypass Permissive (%RTP) AL 40 35.4 Reactor Vessel Water Level - Low, Level 3 Scram (in. AIZ) AL 10.2 No Change 4 Notes: 1. No credit is taken in any safety analysis for the flow referenced setpoints.

2. The EPU APRM High Flux Scram AL, APRM Flow Biased Scram Clamp AV, APRM Setdown in the Startup Mode Scram AV, APRM Setdown Rod Block AV, and APRM Downscale Rod Block AV remain the same in terms of percent rated power. The AV changes for ARPM High Flux Scram and the NTSP (setpoint) for the balance of these functions also change as noted (See Notes 9, 10, and 11 as applicable).

3. W is the Recirculation Drive Flow in percent of rated flow.

4. The AL, AV and NTSP are not changed for EPU for this setpoint function. EPU satisfies the issue with Steam Flow Induced Error (SFIE; also called "Bernoulli error") in the case that the Steam Dryer skirt becomes uncovered for a LOFW, per the related Safety Communication SC 04-14 (Reference 43). AIZ means Above Instrument Zero. Instrument Zero = 533.0 inches Above Vessel Zero (AVZ). Units used in

the GGNS TSs are "inches", equivalent to "in. H 2 O." 5. The AVs for SLO operation are unchanged in terms of MWt.

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2-176 6. Current values are based on the GGNS PRNM system LAR. (Reference 44)

7. The EPU RPC LPSP (upper bounds) AL (36% RTP) and EPU Rod Withdrawal Limiter HPSP AL (70% RTP) remain the same in terms of percent ra ted power. The EPU RPC LPSP (lower bounds) AL (8% RTP) is rescaled to retain the value in absolute power.

8. The AV and NTSP are changed based on the new PRNM equipment and its setpoint calculation with EPU.

9. The NTSP is changed based on the new PRNM equipment and its setpoint calculation with EPU.

10. The NTSP is unchanged with the new PRNM equipment and its setpoint calculation with EPU.

11. The AV and NTSP changed based on the setpoint calculation with EPU.

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2-177 Table 2.4-2 Changes to Instrumentation and Controls Parameter EPU Change MSL High Flow DP instrument loop rescaled DP NTSP adjusted Turbine First Stage Pressure Recalibrate/rescale transmitters and recorder. NTSPs adjusted then

validated with power ascension data. APRM flow biased STP scram / rod block NTSPs adjusted Feedwater Flow Measurement Instruments Recalibrate transmitters and recalibrate/rescale associated loop instruments, indicators and recorders Steam Flow Measurement Instruments Recalibrate transmitters and recalibrate/rescale associated loop instruments, indicators and recorders HP Turbine Outlet Pressure Recalibrate/rescale transmitters and associated loop instruments LP Turbine Inlet Press Recalibrate/rescale transmitters and associated loop instruments Final Feedwater Pressure Recalibrate/rescale transmitters and associated loop instruments Reactor Feed Pump Turbine (RFPT)

Speed Recalibrate/rescale transmitters, associated loop instruments, and indicators. Reset overspeed trip to higher value. RFPT Reheat Steam Supply

Pressure Recalibrate/rescale transmitters and associated loop instruments Main Steam Reheater 1 s t Stage Heating Steam Flow Recalibrate/rescale transmitters and associated loop instruments FWH 5A/B Pressure Recalibrate/rescale transmitters and associated loop instruments FWH 2A/B/C Drain Flow Recalibrate/rescale transmitters and associated loop instruments FWH Outlet Condensate Flow Recalibrate/rescale transmitters and associated loop instruments RMS Various Radiation Monitors Setpoints are based on b ackground radiation input which will be evaluated and revised as required during EPU power ascension Main Steam Line Radiation NTSP for mechanical vacuum pump and reactor recirculation sample isolation changed based on EPU main steam line full power radiation level. Carbon Bed Vault Radiation NTSPs for High and High High alarms changed based on EPU carbon bed vault full power background radiation value. Main Generator Current Replace curre nt indicators and transducers. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-178 Parameter EPU Change Main Generator VARs Replace r ecorder and transducers. Main Generator Power. Replace recorder, meters, and transducers. FWH 3/4/5/6 Drain Temperature Recalibrate temperature elements. Condensate Booster Pump Suction Pressure Low suction pressure trip reset to a higher value. Main Turbine Pressure Regulator Lower the pressure controller setting Turbine Stress Replace turbine stress evaluator transducer

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2-179 2.5 Plant Systems

2.5.1 Internal

Hazards 2.5.1.1 Flooding 2.5.1.1.1 Flood Protection Regulatory Evaluation Entergy conducted a review in the area of flood protection to ensure that SSCs important to safety are protected from flooding. The review covered flooding of SSCs important to safety from internal sources, such as those caused by fa ilures of tanks and vessels. The review focused on increases of fluid volumes in tanks and vessels assumed in flooding analyses to assess the effect of any additional fluid on the flooding protection that is provided. The regulatory acceptance criteria for flood protection are based on GDC-2. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. High energy and moderate energy piping br eaks are described in UFSAR Section 3.6, "Protection against the Dynamic E ffects Associated with the Postulated Rupture of Piping," and in Appendix 3C, "Failure Mode Analysis for Pipe Breaks and Cracks."

Technical Evaluation 2.5.1.1.1.1 High Energy Line Break NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 10.1 of the CLTR addresses the effect of CPPU on flooding. The

results of this evaluation are described below. HELBs are evaluated for their effects on equipment qualification. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-180 Topic CLTR Disposition GGNS Result Liquid Lines [[

     ]] Meets CLTR Disposition As stated in Section 10.1 of the CLTR, EPU ma y increase subcooling in the reactor vessel, which may lead to increased break flow rates for liquid line breaks.

Components and/or equipment required for safe s hutdown of the reactor were evaluated for the effect of flooding from breaks and cracks in high-energy lines. The evaluations verified that the plant can be safely shut down, assuming a concurrent single active failure in systems necessary to mitigate the consequences of the postulated component failure. Plant flooding due to internal piping failures in the FW system is evaluated for changes due to EPU. Auxiliary Building flooding from the FW header in the steam tunnel at CLTP is limited to postulated cracks. No changes are made to the existing flood barriers. Du e to pipe chases at the floor of the Auxiliary Building steam tunnel, any liquid line break will result in flooding to the RWCU rooms and to the RCIC room. The CLTP and EPU consequences are bounded by assuming the hotwell volume and the CST available make-up to the condensate system being released to the flood zone. Breaks in the RHR SDC return lines and RCIC pump return line that are part of the FW header pressure boundary were also evaluated by assuming that the entire contents of the condenser hotwell and CST available make-up to the condensate system being released to the flood zone. Because these volumes do not change with EPU, the consequences are not changed and remain acceptable. Breaks in the RWCU system outside containment are not affected because the CLTP assumptions are unchanged by EPU. Evaluations of the remaining high energy systems determined that flooding effects from high-energy pipe breaks and cracks outside of containment and failure of non-seismic tanks and vessels remain unchanged by EPU. Therefore, GGNS meets all CLTR dispositions for liquid lines. 2.5.1.1.1.2 Moderate Energy Line Break NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 10.2 of the CLTR addresses the effect of CPPU on flooding. The

results of this evaluation are described below. Moderate Energy Line Breaks (MELBs) are evaluated for their effects on equipment qualification. The EPU effect on MELB spray and subcompartment temperature is addressed in Section 2.5.1.3.2. This section discusses the EPU effect on flooding levels. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-181 Topic CLTR Disposition GGNS Result Flooding [[

]] Acceptable The CLTR states that [[
]] Also, in general, EPU does not introduce new MELB locations and does not introduce or move safety-related equipment. Three modifications to be implemented for the EPU and their potential eff ect on the MELB flooding level evaluations are discussed below. The original GGNS flooding analysis assumes that the entire Circulating Water System (CWS) volume flows into the Turbine Building following a line break. It does not credit tripping the CWS pumps or break isolation. The analysis does not credit any equipment located in the Turbine Building for safe shutdown. A modification of the ACTs to add more tower cells over the existing basin increases the CWS inventory. This change results in a marginal increase (less than 0.5%) in the total CWS volume with no change in consequences. A modification to the Circulating Water pumps to increase the CWS flow rate similarly does not affect the total volume released by a CWS break. Thus, the two CWS modifications are expected to have negligible effect on the calculated flood height and do not affect safety-related equipment. As described in Section 2.5.1.3.2, modifications to the fuel pool cooling system are planned as a result of EPU. The flow rates of the fuel pool cooling and its supporting cooling systems are not changed by the EPU modification and therefor e do not affect the flooding evaluation. The GGNS Design Control and 10 CFR 50.59 processes ensure that the effects of this modification on flooding levels and safety-related equipment are acceptable prior to implementation. Therefore, the CLTR disposition is met. However, a plant-specific evaluation demonstrates that the GGNS EPU does not adversely affect the acceptance criteria for flooding from MELBs.

Conclusion The proposed changes in fluid volumes in tanks and vessels for EPU have been reviewed. Entergy concludes that SSCs important to safety will continue to be protected from flooding and will continue to meet the requirements of GDC-2 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to flood protection. 2.5.1.1.2 Equipment and Floor Drains Regulatory Evaluation The function of the equipment and floor drainage system (EFDS) is to assure that waste liquids, valve and pump leak offs, and tank drains are di rected to the proper area for processing or NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-182 disposal. The EFDS is designed to handle the volume of leakage expected, prevent a backflow of water that might result from maximum flood levels to areas of the plant containing safety-related equipment, and protect against the potential for inadvertent transfer of contaminated fluids to an uncontaminated drainage system. The review of the EFDS included the collection and disposal of liquid effluents outside containment. The review focused on any changes in fluid volumes or pump capacities that are necessary for the proposed EPU and are not consistent with previous assumptions with re spect to floor drainage considerations. The regulatory acceptance criteria for the EFDS are ba sed on GDCs 2 and 4 insofar as they require the EFDS to be designed to withstand the effects of earthquakes and to be compatible with the environmental conditions (flooding) associated with normal operation, maintenance, testing, and postulated accidents (pipe failures and tank ruptures). GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. Floor and Equipment Drainage Systems are described in UFSAR Section 9.3.3. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 8.1 of the CLTR addresses the effect of CPPU on the Equipment and Floor Drain system. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Waste Volumes [[

     ]] Meets CLTR Disposition The CLTR states that power uprate does not affect the floor drain collector subsystem and the waste collector subsystem operation or equipment performance. The floor drain collector subsystem and the waste collector (equipment drain) subsystem both receive periodic inputs from a variety of sources. Equipment and floor dr ain collections of liquid waste are expected to increase by less than 1% due to operation at the EPU condition. The design of the GGNS equipment and floor drains inside and outside of containment has been evaluated to ensure any EPU-related liquid radwaste increases can be pr ocessed. GGNS has sufficient capacity to handle added liquid increases. The drainage systems backflow at maximum flood levels and infiltration NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-183 of radioactive water into non-radioactive water dr ains do not change as a result of EPU. The drainage systems design capability to withstand the effects of earthquakes and to be compatible with environmental conditions does not change as a result of EPU. Therefore, EPU does not affect system operation or equipment performance and meets all CLTR dispositions. Conclusion The assessment of the effects of the proposed EPU on the EFDS has been reviewed. Entergy concludes that the plant changes resulting in increased water volumes and larger capacity pumps or piping systems have been adequately addr essed. Entergy concludes that the EFDS has sufficient capacity to: (1) handle the additional expected leakage resulting from the plant changes; (2) prevent the backflow of water to areas with safety-related equipment; and (3) ensure that contaminated fluids are not transferred to non-contaminated drainage systems. Based on this, the EFDS will continue to meet the requirements of GDCs 2 and 4 following implementation of the proposed EPU. Theref ore, Entergy finds the proposed EPU acceptable with respect to the EFDS. 2.5.1.1.3 Circulating Water System Regulatory Evaluation The CWS provides a continuous supply of cooling water to the main condenser (MC) to remove the heat rejected by the turbine cycle and auxiliary systems. The review of the CWS focused on changes in flooding analyses that are necessary due to increases in fluid volumes or installation of larger capacity pumps or piping needed to accommodate the proposed EPU. The regulatory acceptance criteria for the CWS are based on GDC-4 for the effects of flooding of safety-related areas due to leakage from the CWS and the effects of malfunction or failure of a component or piping of the CWS on the functional performance capabilities of safety-related SSCs. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The CWS is described in UFSAR Section 10.4.5.

Technical Evaluation The CWS is being modified for EPU operation. To increase cooling capacity, eight new cells are being added to the existing 20 cell mechanical draft ACT and the existing hyperbolic cooling tower. The performance of this system was ev aluated for EPU based on the addition of the new cooling tower cells over the actual range of circulating water inlet temperatures, and confirms that the CWS and heat sink (with modification) are adequate for EPU operation. The evaluation NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-184 of the CWS at EPU power indicates sufficient system capacity to ensure that the plant maintains adequate condenser backpressure. Condenser hotwell temperature limitations may require load reductions at the upper range of the anticipated circulating water inlet temperatures. To help lower condenser pressure, increase plant output, and reduce condensate temperatures, the circulating water pumps will be modified to increase circulating water flow rate. The cooling towers are non-safety related and remotely located from safety-related SSCs. The cooling tower cells being added for EPU supply water to the plant through existing piping and the new cooling tower fans are powered from non-safety related electrical busses so no new interactions with safety-related systems or components are created. MELB flooding analysis changes resulting from the cooling tower modification are addressed in Section 2.5.1.1.1.

Conclusion The assessment of the modifications to the CWS has been reviewed. Entergy concludes it has adequately evaluated these modifications. En tergy concludes that, consistent with the requirements of GDC-4, the increased volumes of fluid leakage that could potentially result from these modifications would not result in the failure of safety-related SSCs following implementation of the proposed EPU. Theref ore, Entergy finds the proposed EPU acceptable with respect to the CWS. 2.5.1.2 Missile Protection 2.5.1.2.1 Internally Generated Missiles Regulatory Evaluation The review concerns missiles that could result from in-plant component overspeed failures and high pressure system ruptures. The review of potential missile sources covered pressurized components and systems, and high-speed rotating machinery. The review was conducted to ensure that safety-related SSCs are adequately protected from internally generated missiles. In addition, for cases where safety-related SSCs are located in areas containing non-safety related SSCs, the non-safety related SSCs were reviewed to ensure that their failure will not preclude the intended safety function of the safety-related SSC

s. The review focused on any increases in system pressures or component overspeed cond itions that could result during plant operation, AOOs, or changes in existing system configurations such that missile barrier considerations could be affected. The regulatory acceptance criteria for the protection of SSCs important to safety, against the effects of internally generated missiles that may result from equipment

failures, are based on GDC-4. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-185 GGNS Current Licensing Basis NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. Missile protection is described in UFSAR Section 3.5.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 7.1 of the CLTR addresses the effe ct of CPPU on the T-G. The results of this evaluation regarding turbine missiles are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Turbine-Generator Missile Avoidance [[

     ]] Meets CLTR Disposition As explicitly stated in Section 7.1 of the CLTR, the increase in steam flow can change the previous missile avoidance and protection analysis.

As outlined in UFSAR Section 3.5.1, protection from internally generated missiles has been provided to ensure such missiles do not cause a loss of containment function, loss of safe

shutdown function, loss of spent fu el pool (SFP) integrity, and no off-site exposure in excess of 10 CFR 100 limits. The missiles considered include those from rotating equipment failure and those from pressurized component failure. In support of the EPU, GGNS is replacing the HP turbine. The replacement HP rotor is of a monoblock design. Monoblock rotors cause no increase in missile failure probability due to EPU. The low pressure (LP) turbine rotors at GGNS (for both CLTP and EPU RTP) have shrunk-on wheels w ith several design features to reduce the probability of stress corrosion cracking. The LP turbine rotors are not being replaced due to EPU at GGNS. A missile probability analysis was performed in November 1998 when the LP rotors were replaced. The probability of turbine missile generation for GGNS is 3.00 x 10 -6 per year based on that turbine missile analysis. A review of the 1998 turbine missile analysis determined that its conclusions are still valid. The SFP system is located in the reinforced concrete Auxiliary Building. Dynamic effects and missiles that might result from plant equipment failures have not changed with respect to the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-186 plant's current design basis as discussed in UFSAR Section 3.5. The resulting pressures and temperatures were found to be within current licensing values. The frequency of catastrophic failure of rotating equipment having synchronous motors as discussed in the UFSAR is extremely low. The conclusions in the current UFSAR Sections 3.5.1.2.1 and 3.5.1.2.2 have not changed. This review topic is applicable to EPUs that result in substantially higher system pressures or changes in existing system configuration. The GGNS EPU does not result in any condition (system pressure increase or equipment oversp eed) that could result in an increase in the generation of internally generated missiles. In addition, the GGNS EPU does not entail any changes in equipment configurations that could change the effect of internally generated missiles on safety-related or non-safety related equipment. Therefore, internally-generated missiles meet

all CLTR dispositions. Conclusion Entergy has reviewed the changes in system pre ssures and configurations that are required for the proposed EPU and concludes that SSCs important to safety will continue to be protected from internally generated missiles and will continue to meet the requirements of GDC-4 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to internally generated missiles. 2.5.1.2.2 Turbine-Generator Regulatory Evaluation The turbine control system, steam inlet stop and control valves, LP turbine steam intercept and inlet control valves, and extraction steam contro l valves control the speed of the turbine under normal and abnormal conditions, and are thus related to the overall safe operation of the plant. The review of the T-G focused on the effects of the proposed EPU on the turbine overspeed protection features to ensure that a turbine overspeed condition above the design overspeed is unlikely. The regulatory acceptance criteria for the T-G are based on GDC-4, and relates to protection of SSCs important to safety from the effects of turbine missiles by providing a turbine overspeed protection system (with suitable redundancy) to minimize the probability of generating turbine missiles. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-187 The turbine overspeed protection system is described in UFSAR Section 10.2.2.5.4. Surveillance requirements are included in Technical Requirements Manual (TRM) Section 6.3.8. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 7.1 of the CLTR addresses the eff ect of CPPU on the T-G. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Turbine-Generator Performance [[

     ]] Meets CLTR Disposition The CLTR states that the increase in thermal energy and steam flow from the reactor is translated to an increased electrical output from the station by the T-G.

The T-G is required for normal plant operation a nd is not safety-related. Most plants were originally designed for a maximum steam flow of 105%. Experience with previous power uprate applications indicates that turbine and generator modifications are required to support power uprate. The uprate of the HP turbine includes a complete replacement of the inner parts (inner casing and rotor) with a new blading that has a flow path design featuring advanced blade profiles designed for the increased mass flow. The LP turbines do not require modification as they have sufficient flow capacity, and their inlet temperature increase is only in the range of 1%. The generator rotor and stator also support EPU conditions. The rotor is being rewound to ensure reliability. These modifications are required to support normal operation and are non-safety related. An evaluation of T-G overspeed protection systems indicates that adequate

protection is provided for EPU conditions. The turbine and generator were originally designed with a maximum flow-passing capability and generator output in excess of rated conditions to ensure that the original rated steam-passing

capability and generator output was achieved. Th is excess design capacity ensures that the turbine and generator meet rated conditions fo r continuous operating capability with allowances for variations in flow coefficients from expected values, manufacturing tolerances, and other variables that may adversely affect the flow-passing capability of the units. The difference in the steam-passing capability between the design condition and the rated condition is called the flow margin. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-188 The T-G was originally designed with a flow margin of 5%. The current rated throttle steam flow is 15.97 Mlbm/hr at a throttle pressure of 993 psia. The generator is rated at 1,525 MVA, which results in a rated electrical output (gross) of 1,373 MWe at a PF of 0.9 and a reactive power of 664 MVAR. At the EPU RTP and reactor dome pressure of 1,040 psia, the turbine operates at an increased rated throttle steam flow of 19.00 Mlbm/hr and at a throttle pressure of 951 psia. A flow margin of 4% is used in designing the new HP turbin e section. The design point of the new turbine includes this flow margin in order to ensure that the turbine will be able to pass the rated throttle, as well as to allow sufficient margin for reactor pressure control. In order to meet these design parameters, the HP turbine and inner casing are being replaced with modified components. The valves wide open (VWO) condition therefore refers to the turbine supply steam flow at 4% over rated condition (i.e., rated flow + 4%). For operation at EPU, the HP turbine has been redesigned with new diaphragms and buckets for at least the minimum target throttle flow margin, to increase its flow passing capability. The increased EPU steam flow will have no effect on the structural integrity of the LP turbine blades for the following reasons: 1. For CLTP, all 8 stages (4 in each flow path) of LP blades, including the airfoil and the roots, were manufactured from 12Cr materials. 12Cr materials are resistant to corrosion in this application, including the blade surfaces. 2. The design for both the rotating and the st ationary parts account for stress corrosion cracking (SCC) was required for both the CLTP and EPU steam conditions. These features include disk keyways which are rounded and located on the downstream side to eliminate stagnation, and the lack of keyw ays on the last two stages. Additionally, operating stress levels are reduced by introducing significant compressive stresses through heat treatment of all disks, shot p eening of SCC sensitive disk surfaces (Disks 1 and 2) including the blade attachment area of Disk 2, and rolling to induce compressive residual stresses following honing of the keyways. Hence, the design includes features to account for SCC. 3. Because there is no change in LP blade or rotor mass, there is no change in LP blade or rotor natural frequency. Both the new and the existing turbine designs exclude natural frequencies that are coincident with operating resonance frequencies. 4. Analysis of stress and blade vibration also yielded acceptable results. Analyses of both lateral and torsional vibrations were performed at EPU conditions to demonstrate that the power uprate will not cause excessive vibration on the LP and HP turbines and the generator. For the LP turbine, there is no mass change and therefore no change in natural frequencies. For the HP turbine, there is a minor mass reduction due to the change to a NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-189 9-stage versus a 10-stage design. Both the C LTP and EPU lateral and torsional analyses identified natural frequencies within the operating range but outside of the operating speed. Operating restrictions will be implemented by GGNS to assure operation at speeds other than at speeds within the natural frequency ranges. Generator components were evaluated to identify the effect of the steam turbine uprate on the generator. The generator has sufficient capability for an increase above its present rating of 1,525 MVA to a generator rating of 1,600 MVA with enhanced cooling. This is sufficient to support the steam turbine uprate to 115% OLTP - 1,523.5 MWe at approximately 0.96 PF lagging with a reactive power output of approximately 660 MVAR (Figure 2.5-7). Additionally, the Current Transformers are upgraded from 45,000/5 to 50,000/5 to support the increased generator output. The replacement HP rotor is of a monoblock design. Monoblock rotors cause no increase in missile failure probability due to EPU. The LP turbine rotors at GGNS (for both CLTP and EPU RTP) have shrunk-on wheels with several design f eatures to reduce the probability of SCC. A missile probability analysis was performed in November 1998 when the LP rotors were replaced. The probability of turbine missile generation for GGNS is 3.00 x 10 -6 per year per unit based on that turbine missile analysis. A review of the 1998 turbine missile analysis determined that its conclusions are still valid. Materials used for both the HP and LP rotors are based upon both a FEA and successful OE with the rotor materials. The EPU HP rotors and the CLTP HP rotors use the same material. Fracture toughness is determined using vendor specifications. For both the HP monoblock rotors and the LP shrunk-on disk rotors, Entergy reviewed all disk and rotor properties and confirmed that they were within the vendor's specification limits. Pre-service inspection requirements are detailed in the vendor Quality Steam Turbine (QST). Entergy reviewed and approved the vendor QST. Contained in the CLTP QST are the actual material properties for all rotors and LP disks. The turbine rotor design complies with the vendor's design procedures. The EPU HP turbine is a monoblock rotor design and therefore does not have separate disks. The ISI requirements for EPU will be the same as those for CLTP. Hence, the CLTP ISI requirements currently described in UFSAR Section 10.2.3.6 will not change for EPU. An overspeed calculation was performed to evaluate the entrapped steam energy contained within the turbine and the associated piping after the stop valves trip, and the sensitivity of the rotor train for the capability of overspeeding. The entrapped energy increases slightly for EPU conditions. The hardware modification design and implementation process establishes the overspeed trip settings to provide protection for a TT. The GGNS EPU turbine design does not result in increases in system pressures, configurations, or equipment overspeed that would affect the evaluation of internally generated missiles on safety-related or non-safety related equipment. Therefore, the GGNS T-G meets all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-190 Conclusion Entergy has reviewed the assessment of the effects of the proposed EPU on the T-G and concludes that the effects of changes in pl ant conditions on turbine overspeed have been adequately addressed. Entergy concludes that the T-G will continue to provide adequate turbine overspeed protection to minimize the probability of generating turbine missiles and will continue to meet the requirements of GDC-4 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the T-G. 2.5.1.3 Pipe Failures Regulatory Evaluation A review was conducted of the plant design for protection from piping failures outside containment to ensure that: (1) such failures would not cause the loss of needed functions of safety-related systems; and (2) the plant could be safely shut down in the event of such failures. The review of pipe failures included high and moderate energy fluid system piping located outside of containment. The review focu sed on the effects of pipe failures on plant environmental conditions, control room habitability, and access to areas important to safe control of post-accident operations where the consequen ces are not bounded by previous analyses. The regulatory acceptance criteria for pipe failures are based on GDC-4, which requires, in part, that SSCs important to safety be designed to accommodate the dynamic effects of postulated pipe ruptures, including the effects of pipe whipping and discharging fluids. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. High energy piping and moderate energy piping failures are described in UFSAR Section 3.6. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Powe r Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Sections 9.2.1, 10.1, and 10.2 of the CLTR address the effects of CPPU on Piping Failures. The results of this evaluation are described below. 2.5.1.3.1 High Energy Piping Outside Containment Where EPU resulted in increased piping stresses in high energy piping outside containment, the increased stresses were evaluated against existing line break criteria to identify any potential new NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-191 break locations. The results of that evaluation (see Section 2.2.2) determined that there are no new HELB locations outside containment due to operation at EPU conditions. Pipe break criteria were evaluated based on the requirements of Section 3.6A.2.1 of the UFSAR, which is based on Section 3.6.2 of the NRC Branch Technical Position MEB 3-1 (SRP 3.6.2). Where required, percentage temperature increases were applied to the existing stress levels for the applicable piping systems at all node points due to EPU conditions. The combinations of stresses were evaluated to meet the requirement of pipe break criteria. Based on these criteria, no new postulated pipe break locations were identified. Existing HELB locations outside the DW that are affected by EPU are identified in Section 2.2.1 with the effects summarized in Table 2.2-1. Because the CLTP post-HELB mass and energy releases and environmental conditions (pressures and temperatures) are minimally affected by the EPU (see Table 2.2-1), there is no adverse effect on the post-HELB subcompartment structural integrity and minimal effect on environmental conditions outside the DW. The results of this

evaluation are described below. The ability of the plant to cope with the flooding effects from HELBs outside containment that are affected by EPU is evaluated in Section 2.5.1.1. 2.5.1.3.2 Moderate Energy Piping Outside Containment The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Environmental Qualification [[

     ]] Meets CLTR Disposition GGNS has MELB as part of its licensing basi

s. Typically, as stated in the CLTR, [[

     ]]  Generally, EPU does not affect the ability of the plant to cope with effects of spray from MELBs, introduce new MELB locations, or introduce or move safety-related equipment.

At GGNS, process parameters for moderate energy lines are not affected by EPU, except for two modifications, as discussed below. The effect on subcompartment temperatures and spray is discussed in this section. The effect of MELBs on plant flooding is addressed in Section 2.5.1.1.1.2. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-192 Increases in subcompartment temperatures do not result from postulated MELBs because the temperatures of the systems in which MELBs are postulated change little due to EPU. A modification to increase the flow rate of the CWS only affects the presently postulated Circulating Water break in the Turbine Building. The spray consequences from this break are not quantitatively evaluated because the Turb ine Building does not contain safety-related equipment required to perform a shutdown following a DBA. The modification to replace the FPCCS heat exch angers will not increase the pressure in the CCW system, SSW system or FPCCS. It may result in the relocation of existing safety-related FPCCS equipment, but only within the same local area. These changes may affect the spray consequences. The GGNS Design Control and 10 CFR 50.59 processes ensure that the spray effects on safety-related equipment are acceptable prior to implementation of this modification. An evaluation of the MELB consequences at GGNS considered the increase in RTP and planned EPU modifications. This evaluation demonstrates the CLTR disposition is met. 2.5.1.3.3 Environmental Conditions Accident temperature, pressure, and humidity environments used for qualification of equipment outside containment result from the limiting HELB for each plant area. The HELB pressure profiles for CLTP conditions were determined to be bounding for EPU conditions. The peak HELB temperatures at EPU RTP are bounded by the values used for equipment qualification at CLTP conditions. Details regarding analyses pertaining to the above environmental conditions are addressed in Section 2.3.1. Conclusion Entergy has reviewed the changes that ar e necessary for the proposed EPU and the GGNS-proposed operation of the plant, and concludes that SSCs important to safety will continue to be protected from the dynamic effects of postulated piping failures in fluid systems outside containment and will continue to meet the requirements of GDC-4 following implementation of the proposed EPU. Theref ore, Entergy finds the proposed EPU acceptable with respect to protection against postulated piping failures in fluid systems outside containment. 2.5.1.4 Fire Protection Regulatory Evaluation The purpose of the fire protection program (FPP) is to provide assurance, through a defense-in-depth design, that a fire will not prevent the performance of necessary safe plant shutdown functions and will not significantly increase the risk of radioactive releases to the environment. The review focused on the effects of the incr eased decay heat on the plant's safe shutdown analysis to ensure that SSCs required for the safe shutdown of the plant are protected from the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-193 effects of the fire and will continue to be able to achieve and maintain safe shutdown following a fire. The regulatory acceptance criteria fo r the FPP are based on: (1) 10 CFR 50.48 and associated Appendix R to 10 CFR Part 50, insofar as they require the development of an FPP to ensure, among other things, the capability to safely shut down the plant; (2) GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; and (3) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 is not applicable to GGNS; see UFSAR Section 3.1.2.1.5. The fire protection system is described in UFSAR Section 9.5.1. GGNS meets the guidelines of Appendix A to Branch Technical Position APCS B 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants," and also meets the intent of Appendix R to 10 CFR Part 50.

Technical Evaluation 2.5.1.4.1 Fire Protection Program NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.7 of the CLTR addresses the effe ct of CPPU on the FPP. The results of this evaluation are described below. As explicitly stated in Section 6.7 of the CLTR, [[

     ]]  Therefore, the reactor and containment responses and operator actions will be evaluated [[

     ]] for EPU.

This section addresses the effect of EPU on the FPP, fire suppression and detection systems, and reactor and containment system responses to postulated 10 CFR 50 Appendix R fire events. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-194 Topic CLTR Disposition GGNS Result Fire Suppression and Detection Systems [[ Meets CLTR Disposition Operator Response Time

Meets CLTR Disposition Peak Cladding Temperature

Meets CLTR Disposition Vessel Water Level

Meets CLTR Disposition Suppression Pool Temperature

     ]] Meets CLTR Disposition As explicitly stated in Section 6.7 of the CLTR, the higher decay heat associated with EPU may reduce the time available for the operator to perform the actions necessary to achieve and maintain cold shutdown conditions. The highe r decay heat also results in higher SP temperatures. The higher decay heat may result in lower vessel water levels or higher peak cladding temperatures (PCTs), depending on the plan t-specific analysis basis. As a result of

these effects, fire suppression and detection systems, operator response time, PCT, vessel water level, and SP temperature need to be addressed. [[

              ]]. Any changes in physical plant configuration or combustible loading as a result of modifications to implement the EPU will be evaluated in accordance with the plant modification and FPPs. The safe shutdown systems and equipment used to achieve and maintain cold shutdown conditions do not change, and are adequate for the EPU conditions. The operator actions required to mitigate the consequences of a fire are defined. None of the planned EPU plant modifications represent physical changes to plant fire protection equipment or systems.

GGNS Safe Shutdown Instructions currently require operators to depressurize the reactor within 13 minutes following initiation of the fire even

t. To depressurize the reactor, the GGNS 102.46% CLTP analysis determined that six main steam relief valves (MSRVs) were required to be opened within 18 minutes. At EPU conditions, the time available to the operator to open six MSRVs is 14.3 minutes. Operators have sufficient time, per the current licensing basis, to successfully complete the required actions within the specified time requirements to mitigate the consequences of a fire. Moreover, it is confirme d that additional heat associated with EPU will not interfere with operator success in performing manual actions within the licensing basis action NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-195 time, as currently stipulated. No new operator action has been identified for demonstrating compliance. GGNS does not take credit in any safety analysis for the fire protection system other than for fire protection activities. Procedures are provided under Emergency Procedures (EPs), Severe Accident Procedures (SAPs), and Off-Normal Event Procedures (ONEPs) that provide instructions for utilizing fire protection system pumps to provide water to the reactor, the DW, or the containment if necessary. However, this use of the non-safety related fire protection system is not credited in analyses and EPU operation w ill not require any changes to these procedures regarding the utilization of the fire protection system. With these procedures implemented, the fire protection systems and analyses are sufficient to support EPU. EPU is found to not affect the elements of the fire protection plan related to: (1) administrative controls, (2) fire suppression and detection systems, (3) fire barriers, (4) fire protection responsibilities of plant personnel, and (5) procedures and resources necessary for the repair of systems required to achieve and mainta in cold shutdown. In addition, the increase in decay heat will not result in an increase in the potential for a radiological release resulting from a fire. Administrative controls associated with fire protection in the TSs, the TRM, and the Nuclear Quality Assurance Plan were reviewed, and there are no changes required for EPU. The reactor and containment response to the postulated 10 CFR 50 Appendix R fire event at EPU conditions is evaluated in Section 2.5.1.4.2. Th e results show that the peak fuel cladding temperature, reactor pressure and containment pressures and temperatures are below the acceptance limits and demonstrate that there is sufficient time available for the operators to perform the necessary actions to achieve and main tain cold shutdown conditions. Therefore, the fire protection systems and analyses are not adversely affected by EPU.

Cold shutdown is achieved within 1.5 hours unde r Alternate Shutdown C ooling (ASDC). As such, it can be concluded that the 72-hour cold shutdown as stipulated by Appendix R is met. Therefore, the fire protection systems at GGNS meet all CLTR dispositions. 2.5.1.4.2 10 CFR 50 Appendix R Fire Event

 [[                                                                                                                                                                                     
                                                                                                                                                               ]]  The limiting Appendix R fire event was analyzed under both 102.46% CLTP and EPU conditions. The fuel heatup analysis was performed using the SAFER/GESTR-LOCA analysis model. The containment analysis was performed using the SHEX model. This evaluation determined the effect of EPU on fuel cladding integrity, reactor vessel integrity, and containment integrity as a result of the fire event.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-196 The results of the Appendix R evaluation fo r 102.46% CLTP and EPU provided in Table 2.5-1 and Figures 2.5-1 through 2.5-6 demonstrate that the fuel cladding integrity, reactor vessel integrity and containment integrity are maintained and that sufficient time is available for the operator to perform the necessary actions. No changes are necessary to the equipment required for safe shutdown for the Appendix R event. One train of systems remains available to achieve and maintain safe shutdown conditions from either the main control room or the remote shutdown panel. As indicated by the results of the Appendix R ev aluation, all Appendix R acceptance criteria are met under EPU; therefore, there is no increase in the potential for a radiological release resulting from a fire. Therefore, EPU has no adverse effect on the ability of the systems and personnel to mitigate the effects of an Appendix R fire event, and satisfies the requirements of Appendix R with respect to achieving and maintaining safe shutdown in the event of a fire. Conclusion Entergy has reviewed the fire-related safe shutdown assessment and concludes that the effects of the increased decay heat on the ability of the required systems to achieve and maintain safe shutdown conditions have been adequately evalua ted. Entergy further concludes that the FPP will continue to meet the requirements of 10 CFR 50.48 and GDC-3 and meet the intent of Appendix R to 10 CFR Part 50 and GDC-3 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to fire protection.

2.5.2 Fission

Product Control 2.5.2.1 Fission Product Control Systems and Structures Regulatory Evaluation The review for fission product control systems a nd structures covered the basis for developing the mathematical model for DBLOCA dose computations, the values of key parameters, the applicability of important modeling assumptions, and the functional capability of ventilation systems used to control fission product releases. The review primarily focused on any adverse effects that the proposed EPU may have on the assumptions used in the analyses for control of

fission products. The regulatory acceptance criteri a are based on GDC-41, insofar as it requires that the containment atmosphere cleanup system be provided to reduce the concentration of fission products released to the environment following postulated accidents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-197 The fission production control and removal systems are described in UFSAR Section 6.5.1. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 4.5 of the CLTR a ddresses the effect of CPPU on the SGTS. The assumptions regarding leakage and exhaust paths from the primary and secondary containments and other sources are as described in Alternative Source Term (AST) methodology for GGNS (Reference 45). See UFSAR Sections 15.4.9, 15.6.5, and 15.7.4 for further information. GGNS meets all CLTR dispositions except for th at involving the iodine inventory. Based on changes in source term due to EPU, the new iodine inventory at DBLOCA time 0 is increased over that found in the CLTR. As such, a [[ ]] evaluation has been performed. Also, the GGNS SGTS utilizes a deluge system instead of minimum cooling flow to prevent desorption in the case of increased decay heating. This is considered acceptable for that purpose. The topic addressed in this evaluation is: Topic CLTR Disposition GGNS Result Iodine Removal Capability [[

      ]] [[                    
     ]] The CLTR states that the core inventory of iodine and subsequent loading on the SGTS filters or charcoal adsorbers are affected by EPU. The SGTS is designed to maintain secondary containment at a negative pressure and to filter the exhaust air for removal of fission products potentially present during abnormal conditions. By limiting the release of airborne particulates and halogens, the SGTS limits off-site dose following a postulated DBA. The flow capacity of the SGTS and its ability to maintain a negative pressure in the secondary containment is discussed in Section 2.6.6.

At GGNS, neither the SGTS component design nor the filter materials are being altered due to the EPU. The total (radioactive plus stable) post-LOCA iodine loading on the charcoal adsorbers increases proportionally with the increas e in core iodine inventory, which increases with core thermal power. However, sufficient charcoal mass is present so that the post-LOCA iodine loading on the charcoal remains below the guidance provided by RG 1.52 (Reference 46). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-198 [[

     ]] Two bounding analyses have been performed in the CLTR to evaluate decay heating in the SGTS for: (1) plants that implement Alternate Source Term (AST) in accordance with RG 1.183 (Reference 45), and (2) plants committed to RG 1.3 (Reference 47) for fission product transport.

[[

                                                                                             ]]  The parameters and their bounding values, with a comparison to the GGNS specific values, are shown in Table 2.5-2.

As seen in Table 2.5-2, the GGNS SGTS design was not bounded for two of the parameters: [[ ]] These issues are discussed further below. [[

                                                                                                                                                               ]] the results of the GGNS evaluation show that the actual charcoal loading of 0.00097 mg/gm is much less than the RG 1.52 allowable of 2.5 mg/gm.

[[

                                                                                                      ]]  The maximum component temperature for the GGNS evaluation is 150.5

ûF, which is well below the allowable component temperature. The GGNS SGTS utilizes a deluge system to assu re no desorption of radionuclides in the case of increased decay heating. While decay heat from fission products accumulated within the system filters and charcoal adsorbers increases with the increase in thermal power, the manually operated deluge sub-system of the SGTS will still continue to protect the system from desorption should there be a loss of a system fan. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-199 The remaining parameters used in the C LTR bounding analysis for AST application are confirmed to bound the GGNS plant-specific values. Therefore, and considering the exceptions noted above, the GGNS SGTS design and operation under EPU conditions is consistent with the overall CLTR disposition for the SGTS (that the ability of the SGTS to remove fission products is not adversely affected by EPU) and satis fies applicable regulatory guidance. Conclusion The effects of the proposed EPU on fission product control systems and structures have been reviewed. Entergy concludes that it has adequa tely accounted for the increase in fission products and changes in expected environmental conditions that would result from the proposed EPU. Entergy further concludes that the fission product control systems and structures will continue to provide adequate fission product removal in post-accident environments following implementation of the proposed EPU. Based on this, Entergy also concludes that the fission product control systems and structures will continue to meet the requirements of GDC-41. Therefore, Entergy finds the proposed EPU accepta ble with respect to the fission product control systems and structures. 2.5.2.2 Main Condenser Evacuation System Regulatory Evaluation The main condenser evacuation system (MCES) generally consists of two subsystems: (1) the "hogging" or startup system, which initially establishes MC vacuum; and (2) the steam jet air ejector (SJAE) system, which maintains condenser vacuum once it has been established. The review focused on modifications to the system that may affect gaseous radioactive material handling and release assumptions, and design features to preclude the possibility of an explosion (if the potential for explosive mixtures exists). The regulatory acceptance criteria for the MCES are based on: (1) GDC-60, insofar as it requires that the plant design include means to control the

release of radioactive effluents; and (2) GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including AOOs and postulated accidents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The gaseous radwaste system is described in UFSAR Section 11.3. The condenser air removal system is described in UFSAR Section 10.4.2. The CWS is described in UFSAR Section 10.4.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-200 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 7.2 of the CLTR addr esses the effect of CPPU on the Condenser and SJAEs. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Condenser and SJAE [[

     ]] Meets CLTR Disposition The CLTR states that the increase in steam flow increases the heat removal requirement for the condenser. The additional power level increases the noncondensable gases generated by the reactor. The MC "hogging" (mechanical vacuum pump) and the SJAE functions are required for normal

plant operation and are not safety-related. The design of the condenser air removal system is not adversely affected by EPU and no modification to the system is required. The following aspects of the condenser air removal system were evaluated for this determination: Non-condensable gas flow capacity of the SJAE system; Capability of the SJAEs to operate satisfactorily with available dilution / motive steam flow; and Mechanical vacuum (hogging) pump capability to remove required non-condensable gases from the condenser at EPU start-up c onditions (evacuation of the MC to bring it down to vacuum prior to operation at power) . The physical size of the main condenser and evacuation time are the main factors in establishing the capabilities of the vacuum pumps. These paramete rs do not change. Because flow rates do not change, there is no change to the holdup time in the pump discharge line routed to the Reactor Building vent stack. The capacity of the SJAEs is adequate because they were originally designed for operation at flows greater than those required at EPU conditions. Therefore, the MCES design bases for GGNS are unchanged for EPU and meet all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-201 Conclusion The assessment of the MCES has been review ed. Entergy concludes that the MCES will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment following implementation of the proposed EPU. Entergy also concludes that the MCES will continue to meet the requirements of GDCs 60 and 64. Therefore, Entergy finds the proposed EPU acceptable with respect to the MCES. 2.5.2.3 Turbine Gland Sealing System Regulatory Evaluation The turbine gland sealing system is provided to control the release of radioactive material from steam in the turbine to the environment. Enter gy reviewed changes to the turbine gland sealing system with respect to factors that may affect gaseous radioactive material handling (e.g., source of sealing steam, system interfaces, and poten tial leakage paths). The regulatory acceptance criteria for the turbine gland sealing system are based on: (1) GDC-60, insofar as it requires that the plant design include means to control the re lease of radioactive effluents; and (2) GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, AOOs, and

postulated accidents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The gaseous radwaste system is described in UFSAR Section 11.3. The turbine gland sealing system is described in UFSAR Section 10.4.3. Technical Evaluation Taking into account the modification of the GGNS main turbine to accept the increased steam flow at EPU operating conditions, and new springs for the steam seal regulators as a result of recent uprates, the evaluation of the turbine gland sealing system demonstrated that no hardware changes are required to support operation at EPU conditions. Conclusion The assessment of the turbine gland sealing system has been reviewed and has been adequately evaluated. Entergy concludes that the turbine gland sealing system will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-202 consistent with GDCs 60 and 64. Therefore, Entergy finds the proposed EPU acceptable with respect to the turbine gland sealing system. 2.5.2.4 Main Steam Isolation Valve Leakage Control System Regulatory Evaluation Redundant quick-acting isolation valves are pr ovided on each MSL. The LCS is designed to reduce the amount of direct, untreated leakage from the MSIVs when isolation of the primary system and containment is required. The review of the MSIV LCS focused on the effects of the proposed EPU on the amount of leakage assumed to occur. The regulatory acceptance criteria for the MSIV LCS are based on GDC-54, insofar as it requires that piping systems penetrating containment be provided with leakage detection and isolation capabilities.

GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The MSIV LCS is described in UFSAR Section 6.7.1.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 4.6 of the CLTR addresses the effect of CPPU on the MSIV LCS. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Radiological Effect [[

     ]] Meets CLTR Disposition The CLTR states that the radioisotopes rel eased through the MSIVs during an accident will increase due to EPU.

The GGNS MSIV LCS sends the leakage flow to the secondary containment where the radiation is handled by the SGTS. EPU will not significantly affect the leakage flow rate. Additional discussion of the role of the SGTS in the secondary containment design is provided in Section 2.6.6. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-203 Therefore, the radiological effect meets all CLTR dispositions. Conclusion The MSIV LCS has been assessed and the system adequately accounts for the effects of the proposed EPU on the assumed leakage through the MS IVs. Entergy further concludes that the LCS will continue to reliably detect and isolate the leakage, as required by GDC-54. Therefore, Entergy finds the proposed EPU acceptable with respect to the MSIV LCS.

2.5.3 Component

Cooling and Decay Heat Removal 2.5.3.1 Fuel Pool Cooling and Cleanup System Regulatory Evaluation The SFP provides wet storage of spent fuel (SF) assemblies. The safety function of the FPCCS is to cool the SF assemblies and keep the SF assemblies covered with water during all storage conditions. The review for the proposed EPU fo cused on the effects of the proposed EPU on the capability of the system to provide adequate cooling to the SF during all operating and accident

conditions. The regulatory acceptance criteria fo r the FPCCS are based on: (1) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety

functions; (2) GDC-44, insofar as it requires that a system with the capab ility to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident

conditions be provided; and (3) GDC-61, insofar as it requires that fuel storage systems be designed with RHR capability reflecting the importance to safety of DHR, and measures to

prevent a significant loss of fuel storage coolant inventory under accident conditions. GGNS Current Licensing Basis NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC 5 is not applicable to GGNS; see UFSAR Section 8.3.1.1.2.5. The FPCCS is described in UFSAR Section 9.1.3.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.3 of the CLTR addresses the effect of CPPU on the Fuel Pool. The results of this evaluation are described below. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-204 Due to the increased decay heat associated with EPU, a modification to the FPCCS is planned to restore cooling margin and post outage flexibility. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Fuel Pool Cooling (Normal Core Offload and Full Core Offload) [[ Meets CLTR Disposition Crud Activity and Corrosion Products

Meets CLTR Disposition Radiation Levels

     ]] Meets CLTR Disposition 2.5.3.1.1 Fuel Pool Cooling (Normal and Full Core Offload)

As stated in Section 6.3.1 of the CLTR, the SFP heat load increases due to the decay heat generation as a result of EPU. The SF cooling section of the FPCCS is classified as nuclear safety-related and is redundant. Cooling water to the FPCCS HXs is provided by the CCW or SSW system. Additional SFP cooling is available from the RHR system. The SSW and RHR systems are redundant and classified as nuclear safety-related. EPU does not affect the alignments, availability or safety-related designations of these systems. EPU did not change the trains of cooling used to evaluate the effects of core offload.

EPU will increase the heat load on the FPCCS duri ng and after RFOs because of the increase in decay heat. The decay heat for the EPU was calculated using the formulation and uncertainty factors from ANS/ANSI-5.1-1994 (Reference 48) with two-sigma uncertainty added. The effect of this heat load on the SFP temperature was then evaluated for both batch and full-core offloads. The evaluation of the batch offload credits the CCW system for directly removing the decay heat from the FPCCS HXs. The heat removal by the FPCCS HX is conservatively based on a CCW temperature of 95°F. The result of this cons ervative evaluation shows that, using the FPCCS alone, the SFP temperature would rise above 140°F (to 146°F) for about 4 days. Even with a single failure, the FPCCS would maintain the SFP temperature below 149°F. In practice, using the RHR system in a cooling assist mode could mitigate these elevated temperatures, and thus the SFP temperature could be maintained below 130° F. The full-core offload was evaluated in two scenarios: one using the FPCCS alone and one using the RHR system alone. It was shown that the FPCCS can maintain the pool temperature below boiling. The RHR system, assuming a maximum SSW temperature of 90°F, can maintain the SFP temperature below 140°F. The NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-205 results of this evaluation confirm that the full-core offload heat load can be removed by the FPCCS with elevated SFP temperatures or by the RHR system, consistent with the current licensing basis. Maintaining the SFP temperature within design limits for the full-core offload is accomplished through existing administrative and procedural limitations that require cycle-specific core offload evaluations prior to initiating the core offload. No changes are required to accommodate the full-core offload. Table 2.5-3 summarizes the three cases of the FPCCS evaluations performed: Full Core Offload Following a RFO, Emergency Full Core Offloa d, and Core Shuffle Following a RFO. The predicted boil-off rates remain within the available make-up capability. The worst-case makeup requirement occurs when all cooling is lost after a full-core offload. If this condition occurs, any of the following sources can provide makeup to maintain SFP level: the diesel-driven fire pump, the RHR system, and the Condensate and Refueling Water Storage and Transfer System. The heating rate is sufficiently slow to allow operator actions to initiate a redundant cooling system. The FPCCS remains capable of performing its required safety functions after EPU implementation. Therefore, FPC meets all CLTR dispositions. Although the FPCCS was determined to remain capable of performing its safety functions after EPU, Entergy has chosen to increase the operating margins on this system. As part of EPU, Entergy is implementing a modification to replace the heat exchangers and increase the heat rejection capability in the FPCCS. This change will maintain the pool water temperature within the current licensing basis. 2.5.3.1.2 Crud Activity and Corrosion Products As stated in Section 6.3.2 of the CLTR, crud activ ity and corrosion products associated with SF can increase slightly due to power uprate. The amount of crud activity and pool quality are operational considerations and are unrelated to safety. An evaluation of the capability of the FPCCS to maintain water clarity concludes that water clarity will not be affected by EPU. Therefore, the Crud Activity and Corrosion Products meet all CLTR dispositions. 2.5.3.1.3 Radiation Levels As stated in Section 6.3.3 of the CLTR, the normal radiation levels around the SFP may increase slightly, primarily during fuel handling operations. Radiation levels in those areas of the plant, which are directly affected by the reactor core and SF, increase by the percentage increase in the average power density of the fuel bundles. Theref ore, for an EPU increase of 15%, the radiation dose rates increase by 15%. The radiation leve l around the SFP is an operational consideration and is unrelated to safety. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-206 The design of SFPs is typically conservative from the perspective of radiation exposure such that changes in the fuel inventory/bundle surface dose ra te of 15% results in inconsequential changes in operating dose. The current GGNS radiation procedures and radiation monitoring program would detect any changes in radiation levels and initiate appropriate actions. Therefore, the radiation levels around the SFP meet all CLTR dispositions. Conclusion The FPCCS has been assessed and the effects of the proposed EPU on the SFP cooling function of the system have been adequately evaluated. Based on this review, Entergy concludes that the FPCCS will continue to provide sufficient cooling capability to cool the SFP following implementation of the proposed EPU and will continue to meet the requirements of GDCs 44 and 61. Therefore, Entergy finds the proposed EPU acceptable with respect to the FPCCS.

2.5.3.2 Station Service Water System Regulatory Evaluation At GGNS, service water for cooling safety-related equipment is provided by the SSW. The PSW provides makeup water to the SSW. The review covered the SSW components with respect to their functional performance as affected by adverse operational (i.e., water hammer) conditions, abnormal operational conditions, and accident cond itions (e.g., a LOCA with the LOOP). The review focused on the additional heat load that would result from the proposed EPU. The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, in cluding flow instabilities and loads (e.g., water hammer), maintenance, testing, and postulated accidents; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuc lear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be

provided. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC 5 is not applicable to GGNS; see UFSAR Section 8.3.1.1.2.5. The SSW system is described in UFSAR Section 9.2.1. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-207 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.4 of the CLTR addresses the effect of CPPU on Water System Performance. The results of this evaluation are described below. The SWS is comprised of two systems at GGNS, the non-safety related PSW system and the safety-related SSW system. The PSW system provides once through cooling water from radial wells to various non-safety related plant systems and components. The PSW system is designed to operate during normal conditions. The safety-related portion of the PSW system includes some piping and (safety-related to non-safety rela ted) isolation valves. The non-safety related portions of the PSW system include pumps, valves, piping and instrumentation that provide cooling and makeup water to various non-safety related systems and components, including the TBCW HXs and CCW HXs, as well as providing make-up water to the CWS. The SSW system includes pumps, valves, piping and instrumentation to provide cooling water from the Standby Cooling Tower (Section 2.5.3.4) to various safety-related plant systems and components. The SSW is safety-related and is designed to operate during normal shutdown, LOOP, transient, and post-accident conditions. The SSW system normally services the RHR HXs and picks up cooling loads normally handled by PSW and the CCW system during periods of LOOP and/or accident conditions. In order to transfer loads from these non-safety related systems, safety-related isolation valves are provi ded to isolate the non-safety related portion of the systems. There are three divisions of SSW pumps and cooling water distribution systems. Division A and Division B both provide cooling water to Containment and the Auxiliary, Control, and Emergency Diesel Generator Buildings, as well as the SSW Pumps located at the

Standby Cooling Towers (UHS). The third di vision (Division C) is HPCS which provides cooling water to the HPCS Emergency Diesel Generator and the HPCS pump room cooler. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Water Systems Performance (Safety-Related) [[

Meets CLTR Disposition Water Systems Performance (Normal Operation)

Meets CLTR Disposition Suppression Pool Cooling (RHR Service Operation)

     ]] Meets CLTR Disposition

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-208 2.5.3.2.1 Water System Performance (Safety-Related) As explicitly stated in Section 6.4 of the CLTR, EPU results in a greater decay heat rate which increases the safety-related water systems cooling requirement during accident conditions. The performance of the SSW system during and immediately following the most limiting design basis event, the LOCA, [[

     ]] For GGNS, the performance of the SSW system during and immediately following the most limiting design basis event, the LOCA, is not dependent on reactor power. The longer term performance results in an increase heat re jected to the Standby Cooling Tower and the subsequent cooling tower evaporation due to power uprate. The SSW system is designed to provide a relia ble supply of cooling water during and following a DBA for the following essential equipment and systems:

Services which have increased heat loads with EPU: RHR System HXs SFP Cooling Water HXs (safety-related backup to normal non-safety related CCW supply) RHR Pumps A/B/C Pump Room Coolers RCIC Pump Room Cooler HPCS Pump Room Cooler LPCS Pump Room Cooler SFP Cooling Water Pump Room Coolers ESF Electrical Switchgear Room Coolers Services for which heat loads are not dependent on RTP: Division I, II and HPCS Standby Diesel Generator Jacket Water Coolers RHR Pumps Seal Coolers (safety-related backup to normal non-safety related CCW supply) SSW Pumps A/B Motor Bearing Coolers NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-209 DW Purge Compressor Coolers Control Room Air Conditioners The capacity of the UHS (Standby Cooling Tower) is being increased by 15% for EPU operation. The capacity increase will be accomplished by an EPU modification which replaces the fill and increases the air flow. The SSW system was evaluated for changes due to EPU and

is adequate based on the 15% UHS capacity increase. Therefore, the SSW system meets all CLTR dispositions. 2.5.3.2.2 Water System Performance (Normal Operation) As stated in Section 6.4 of the CLTR, EPU results in an increased heat load during normal

operation. The increased heat load at GGNS is due to the increased SFP cooling decay heat load transferred to the CCW system. To accommodate the added heat load, three CCW HXs will be used for a period following RFOs, which will require additional PSW flow. Also, due to higher condenser heat load and subsequent cooling tower evaporation, the CWS requires additional makeup water flow from PSW. The temperature control valve that regulates PSW flow rate to the CCW HXs is not adequately sized for operation with 3 CCW HXs in operation, and a modification will be made to better control the flow through the CCW HXs at EPU conditions. Although the increased flow requirements are within the capacity of the existing PSW pumps, a planned modification will increase the system flow margin by adding a radial well with two PSW pumps, similar to the four existing radial wells. Requirements for system isolation when the SSW system picks up the serviced loads are not changed by EPU. Therefore, Water System performance during normal operation meets all CLTR dispositions. 2.5.3.2.3 Suppression Pool Cooling (RHR Service Operation) As stated in Section 6.4 of the CLTR, EPU results in a greater decay heat rate. The containment cooling analysis in Section 2.6.5 indicates that the post-LOCA RHR system heat load increases due to an increase in the maximum SP temperat ure that occurs following a LOCA. The post-LOCA containment and SP responses have been calculated based on an energy balance between

the post-LOCA heat loads and the existing heat removal capacity of the RHR system and SSW. The containment cooling analysis and equipment review demonstrate that the SP temperature can be maintained within acceptable limits in the post-accident condition at EPU based on the existing capability of the SSW system. As discussed in Section 2.5.3.2.1, the SSW system

transfers heat to the UHS, which is addressed in Section 2.5.3.4. The SSW flow rate is adequate for Suppression Pool Cooling (SPC) at EPU conditions. Therefore, SPC meets all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-210 Conclusion The effects of the proposed EPU on the standby SW S have been adequately evaluated for the increased heat loads on system performance that would result from the proposed EPU. Entergy concludes the station SWS will continue to be protected from the dynamic effects associated with flow instabilities and provide sufficient cooling for SSCs important to safety following implementation of the proposed EPU. Therefore, Entergy has determined that the standby SWS will continue to meet the requirements of GDCs 4 and 44. Based on the above, Entergy finds the proposed EPU acceptable with respect to the standby SWS. 2.5.3.3 Reactor Auxiliary Cooling Water Systems Regulatory Evaluation The review covered reactor auxiliary cooling water systems that are required for: (1) safe shutdown during normal operations, AOOs, and m itigating the consequences of accident conditions; or (2) preventing the occurrence of an accident. These systems include closed-loop auxiliary cooling water systems for reactor system components, reactor shutdown equipment, ventilation equipment, and components of the ECCS. The review covered the capability of the auxiliary cooling water systems to provide adequate cooling water to safety-related ECCS components and reactor auxiliary equipment for all planned operating conditions. Emphasis was placed on the cooling water systems for safety-related components (e.g., ECCS equipment, ventilation equipment, and reactor shutdown equipment). The review focused on the additional heat load that would result from the proposed EPU. The regulatory acceptance criteria for the reactor auxiliary cooling water system are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, including flow instabilities and attendant loads (i.e., water hammer), maintenan ce, testing, and postulated accidents; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their

safety functions; and (3) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and

accident conditions be provided. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC 5 is not applicable to GGNS; see UFSAR Section 8.3.1.1.2.5. The CCW system is described in UFSAR Section 9.2.2. The TBCW system is described in UFSAR Section 9.2.9. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-211 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.4 of the CLTR addr esses the effect of CPPU on Water Systems. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Water Systems Performance (Safety Related) [[

Meets CLTR Disposition Water Systems Performance (Normal Operation)

     ]] Meets CLTR Disposition As explicitly stated in Section 6.4 of the CLTR, EPU results in a greater decay heat rate which increases the safety-related water systems cooling requirement during accident conditions.

The safety-related Reactor Auxiliary Cooling Water systems include the CCW piping and valves associated with Spent FPCCS HXs and piping and valves that form part of the containment boundary. The non-safety related Reactor Auxiliary Cooling Water systems include the TBCW system, the DW Chilled Water System and the Plant Chilled Water System.

Component Cooling Water System CCW is designed to cool auxiliary plant equipment during normal operating and normal shutdown conditions. Cooling water is also provided to some components during a LOOP; during a LOOP the RWCU HX and SFP HXs automatically isolate from the rest of the CCW. CCW piping and valves associated with SFP HXs and piping and valves that form part of the containment boundary are safety-related. EPU does not affect the ability of these pipes and valves to perform their intended safety-related function. The rest of the CCW system is non-safety related and is not intended to

operate during accident conditions. Safety-related cooling to SFP HXs will be provided by SSW and is addressed in Section 2.5.3.2. The CCW heat loads are mainly dependent on the reactor vessel temperature, the decay heat in the SFP removed by the spent FPCCS HXs, and/or flow rates in the systems cooled by the CCW. The flow rates in the systems cooled by the CCW (e.g., Recirculation pump motor and bearing coolers, RWCU pump coolers and non-regenerative HX) do not change due to power uprate and therefore, are not affected by power uprate. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-212 The only significant increase in heat load due to EPU is an increase in FPCCS heat load. Safety-related cooling for the SFP is provided by the SSW and not the CCW system. The peak SFP heat load during refueling currently is 18.34 MBTU/hr. This would increase to 27.4 MBTU/hr at EPU. The maximum normal EPU heat load is 22.08 MBTU/hr at the end of a RFO, which is assumed to be 15 days after shutdown. The maximum FPCCS HX heat load occurs during refueling when other CCW loads are offline or significantly re duced. Therefore, the increase in SFP heat load does not increase CCW system heat loads beyond system design. The operation of the remaining equipment cooled by the CCW (Recirculation pump and motor coolers, CRD pump coolers, RWCU pump cooling, RWCU non-regenerative HX, DW equipment drain sump cooler, and post-accident sample coolers) is not affected by power uprate. The CCW system contains sufficient redundancy in pumps and HXs to ensure that adequate heat removal capability is available using two CCW HXs during refueling and LOOP conditions and three CCW HXs for some period after restart (approximately four months). Two CCW HXs are sufficient for normal operation after that period. Sufficient heat removal capacity is available to accommodate the increase in SFP heat load due to EPU. Therefore, CCW meets all CLTR dispositions. Section 2.6.1.4 evaluates issues related to GL 96-06 and GL 96-06 Supplement 1.

Turbine Building Cooling Water System The TBCW system is a non-safety related closed cooling water system which cools auxiliary plant equipment during normal plant operation. Failure of the system will not compromise any safety-related system or component and will not prevent safe reactor shutdown. The design basis cooling water supply temperature is 95°F. The operating supply temperature of the TBCW system is dependent on the heat rejected to the TBCW system via components cooled by the system, as removed by the TBCW HXs and controlled by the system temperature control valve(s). Some heat loads on the TBCW system are power-dependent and are increased by power uprate, such as those related to the T-G Auxiliaries (generator hydrogen coolers, primary water coolers, generator leads coolers (bus duct cooling). Examination of Table 2.5-4 shows the heat load of the total system is 96.7 MBTU/hr due to EPU. The system normal capacity is 141.4 MBTU/hr, which is unaffected by EPU. The increase in heat load of the TBCW system can be accommodated by the margin in the system HXs, and the system pumps have sufficient capacity to accommodate any minor flow increases from potential changes in localized flows to affected components, as required. Therefore, TBCW meets all CLTR dispositions. Drywell Chilled Water System Some heat loads on the DW Chilled Water System are power-dependent and are potentially increased by power uprate, such as piping temperatures. However, it was determined by evaluation of all serviced heat loads at EPU conditions that the DW Chilled Water System is adequate for operation at EPU. The DW Chilled Water System contains sufficient redundancy in NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-213 pumps and chillers to ensure that adequate heat removal capability is available during normal operation. Therefore, the DW Chilled Water System meets all CLTR dispositions. Plant Chilled Water System Some heat loads on the Plant Chilled Water System are power-dependent and are potentially increased by power uprate, such as piping temperatures and pump motor heat loads. However, it was determined by evaluation of all serviced heat loads at EPU conditions that the Plant Chilled Water System is adequate for operation at EPU. The Plant Chilled Water System contains sufficient redundancy in pumps and chillers to ensure that adequate heat removal capability is available during normal operation. Therefore, the Plant Chilled Water System meets all CLTR dispositions. Conclusion The effects of the proposed EPU on the reactor auxiliary cooling water systems have been adequately evaluated for the increased heat loads from the proposed EPU on system performance. Entergy concludes that the reactor auxiliary cooling water systems will continue to be protected from the dynamic effects associated with flow instabilities and provide sufficient cooling for SSCs important to safety following implementation of the proposed EPU. Therefore, Entergy has determined that the reactor auxiliary cooling water systems will continue to meet the requirements of GDCs 4 and 44. Based on th e above, Entergy finds the proposed EPU acceptable with respect to the reactor auxiliary cooling water systems.

2.5.3.4 Ultimate Heat Sink Regulatory Evaluation The ultimate heat sink (UHS) is the source of c ooling water provided to dissipate reactor decay heat and essential cooling system heat loads after a normal reactor shutdown or a shutdown following an accident. The review focused on th e effect that the proposed EPU has on the DHR capability of the UHS. Additionally, the review included evaluation of the design-basis UHS temperature limit determination to confirm that post-licensing data trends (e.g., air and water temperatures, humidity, wind speed, water volume) do not establish more severe conditions than previously assumed. The regulatory acceptance criteria for the UHS are based on: (1) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their

safety functions; and (2) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and

accident conditions be provided. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-214 GGNS Current Licensing Basis NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 is not applicable to GGNS; see UFSAR Section 3.1.2.1.5. The UHS is described in UFSAR Section 9.2.5.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.4 of the CLTR a ddresses the effect of CPPU on the UHS. The results of this evaluation are described below. The UHS temperature may be affected by the increase in normal operating heat load. For most plants, the environmental effects of uprate are controlled at the same level as is presently in place. That is, the plant operation is managed such that none of the present limits such as maximum allowed UHS temperature is increased as a result of uprate. However, for some plants, there may be a small change in UHS temperature. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Ultimate Heat Sink [[

     ]] Meets CLTR Disposition The GGNS UHS (SSW Cooling Towers) consists of two mechanical draft cooling towers, each with four independent cells, and two concrete makeup water basins of the SSW system. One tower services one RHR train, other safety-related loads on the same electrical division, and the HPCS cooling loads. The other tower services the second RHR train and safety-related cooling loads on the same electrical division. The SSW Cooling Towers are not used during normal plant operation, except during reactor shutdown fo r RHR cooling. During accident and LOOP conditions, the SSW Cooling Towers act as the safety-related source of cooling water. The SSW Cooling Water system (refer to Section 2.5.3.2, Station Service Water System) circulates water from the SSW Cooling Tower to serviced HXs and returns the water to the SSW Cooling Tower for re-use. The UHS for GGNS is designed to supply water at 90°F maximum. The SSW Cooling Tower basins are required to maintain a reserve water supply for 30 days post-accident operation without replenishment. As a result of operation at the EPU Reactor Thermal Power level, heat loads will increase.

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2-215 As explicitly stated in Secti on 6.4 of the CLTR, EPU results in increased heat loads during normal operation and in a greater decay heat ra te, which increases the safety-related water systems cooling requirements during accident cond itions. This increases the RHR HX heat loads as a result of EPU. The SFP cooling heat loads, which are serviced by the SSW system following a LOOP, also increase due to EPU. These two major heat load increases, along with other smaller increases discussed in Section 2.5.3.2, must be accommodated by the UHS during accident conditions at EPU. The UHS is operated so that none of the present limits (e.g., minimum cooling tower basin water level during normal operation) are changed as a result of EPU. The SSW Cooling Towers were evaluated for their capability to handle the increased EPU heat load for a 30-day period. The cooling capacity of the cooling towers is increased by plant modification to ensure that the towers can maintain the temperature of the water supplied within the maximum design basis temperature of 90°F during all modes of required operation, including consideration of single failure. The usable water inventory in the basins for EPU is increased by plant modification to ensure a 30-day post-accident operating period without makeup. The GGNS Design Control and 10 CFR 50.59 processes ensu re that the cooling capacity is adequate and the 30-day water inventory is available prior to implementation of EPU. Therefore, the UHS meets all CLTR dispositions. Conclusion The effects that the proposed EPU would ha ve on the UHS safety function, including the validation of the design-basis UHS temperature limit based on post-licensing data, have been reviewed. Entergy concludes that the proposed EPU will not compromise the design-basis safety function of the UHS, and that the UHS will continue to satisfy the requirements of GDC-44 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the UHS. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-216 2.5.4 Balance-of-Plant Systems 2.5.4.1 Main Steam Regulatory Evaluation The Main Steam and Reheat System (MSRS) transports steam from the NSSS to the power conversion system and various safety-related a nd non-safety-related auxiliaries. The review focused on the effects of the proposed EPU on the system's capability to transport steam to the power conversion system, provide heat sink capacity, supply steam to drive safety system pumps, and withstand adverse dynamic loads (e.g., water steam hammer resulting from rapid valve closure and relief valve fluid discharge lo ads). The regulatory acceptance criteria for the MSRS are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects, including the effects of mi ssiles, pipe whip, and JI forces associated with pipe breaks; and (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 is not applicable to GGNS; see UFSAR Section 3.1.2.1.5. The MSRS system is described in UFSAR Secti on 10.3. The MSL flow restrictors are described in UFSAR Section 5.4.4.

Technical Evaluation The heat balance for the EPU conditions is provi ded in Section 1.3. The heat balance shows the transport of steam to the power conversion equipment, the heat sink, and to steam driven components. FIV and structural loading of the MS system piping and supports is addressed in Sections 2.2.2. Dynamic loading from water hammer is discussed below. SRV dynamic loads are discussed in Sections 2.2.2 and 2.2.3. Th e function and capability of the MSIVs are

discussed in Section 2.2.2. SRV setpoint tolera nce and FIV effects are discussed below. 2.5.4.1.1 Structural Evaluation of Main Steam Piping NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also

referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.4.1 of the CLTR addresses the effect of CPPU on FIV in the MSL. The results of this evaluation are described below. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-217 GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Structural Evaluation of Main Steam Piping [[

     ]] Meets CLTR Disposition The CLTR states that because the MS piping pressures and temperatures are not affected by EPU, there is no effect on the analyses for these parameters. Seismic inertia loads, seismic building displacement loads, and SRV discharge lo ads are not affected by EPU, thus, there is no effect on the analyses for these load cases. The increase in MS flow results in increased forces from the TSVC transient. The TSVC loads bound the MSIVC loads because the MSIVC time is significantly longer than the stop valve closure time.

MS piping supports and other critical piping components were evaluated for the TSVC loads. The TSVC loads diminish rapidly for branch pi ping. These loads are considered insignificant after the second pipe support away from the location of an attachment to the MS piping. The MS system outside containment was reanaly zed for EPU conditions as described in Section 2.2.2.2.2.2. Table 2.2-5a and Table 2.2-5b provi de the stress data at critical locations and allowable vs. calculated values for EPU conditions. The capability of the RCPB system to withstand adverse dynamic loads (e.g., water or steam hammer resulting from rapid valve closure) was evaluated. A summary of the results of the MS piping system evaluation, including pipe support loads, that contains the increased loading associated with EPU conditions (i.e., temperatur e, pressure, and flow, including the effects of MS flow induced transient loads at EPU conditions) along with a comparison to the code allowable limits is provided in Section 2.2.2. This section contains summary level tables that demonstrate the rigor of the analytical effort and provide an indication of the magnitude of the highest stress and load ratios found in the analysis. SRV setpoint tolerance is independent of an EPU. EPU evaluations are performed using the existing SRV setpoint tolerance ALs as a basis. Actual historical in-service surveillance of SRV setpoint performance test results are monitored separately for compliance to the TSs and IST program. The in-service surveillance testing of GGNS's SR Vs has not shown a signi ficant propensity for high setpoint drift greater than 1.8%. Increased MSL flow may affect vibration of the piping during normal ope ration. The vibration frequency, extent, and magnitude depend upon plant-specific parameters, valve locations, the valve design, and piping support arrangements. Th e effects of EPU on FIV of the piping will be assessed by vibration testing during initial plant operation at the higher steam flow rates. This NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-218 topic is addressed in Section 2.2.2.1.2. Attachment 10 to the EPU LAR contains details of the vibration monitoring program. FIV may increase incidents of SRV leakage. GGNS currently has a program for monitoring SRV leakage. A monthly procedure is performed to trend SRV tail pipe temperatures. GGNS has performed analyses and testing which inves tigated and addressed the potential for acoustic resonance due to the increased steam flow past the SRV standpipes, as well as other branch connections, and concluded that the onset of SRV standpipe vortex shedding acoustic resonance could be expected beyond EPU power steam flow rates. Therefore, SRV vibration resulting from acoustic resonance is not expected at EPU operating conditions. For this reason, the existing SRV leakage monitoring instrumentation should be sufficient to detect any increased SRV leakage. Therefore, the structural evaluation of MS piping meets all CLTR dispositions. 2.5.4.1.2 Main Steam Line Flow Restrictors

NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.7 of the CLTR addresses the eff ect of CPPU on the MSL flow restrictors. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Structural Integrity [[

     ]] Meets CLTR Disposition The CLTR states that at uprated power, the flow restrictors are required to pass a higher flow rate, which will result in an increased pressure drop.

The increase in steam flow rate has no significant effect on flow restrictor erosion. There is no effect on the structural integrity of the MSL flow element (restrictor) due to the increased DP because the restrictors were designed and analyzed for the choke flow condition. After a postulated steam line break outside containment, the fluid flow in the broken steam line increases until it is limited by the MSL flow restrictor. [[

]]

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2-219 The GGNS restrictors were originally analyzed for these flow conditions and therefore the restrictors remain within the acceptable calculated DP drop and choke flow limits under EPU conditions. Therefore, the flow restrictors meet all CLTR dispositions. Conclusion The effects of the proposed EPU on the MSRS have been reviewed and the effects of changes in plant conditions on the design of the MSRS have been adequately evaluated. Entergy concludes that the MSRS will maintain its ability to transport steam to the power conversion system, provide heat sink capacity, supply steam to steam-driven safety pumps, and withstand steam hammer. Entergy further concludes that the MSRS will continue to meet the requirements of GDC-4. Therefore, Entergy finds the proposed EPU acceptable with respect to the MSRS. 2.5.4.2 Main Condenser Regulatory Evaluation The MC system is designed to condense and deaerate the exhaust steam from the main turbine and provide a heat sink for the TBS. The review focused on the effects of the proposed EPU on the steam bypass capability with respect to load rejection assumptions, and on the ability of the MC system to withstand the blowdown effects of steam from the TBS. The regulatory acceptance criteria for the MC system are based on GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents.

GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The MC is described in UFSAR Section 10.4.2. The CWS is described in UFSAR Section 10.4.5. The condenser offgas system is described in UFSAR Section 11.3.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Powe r Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 7.2 of the CLTR addresses the e ffect of CPPU on the Condenser and SJAEs. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-220 Topic CLTR Disposition GGNS Result Condenser and SJAE [[

     ]] Meets CLTR Disposition As stated in the CLTR, the increase in steam flow increases the heat removal requirement for the condenser. The additional power level increas es the non-condensable gases generated by the reactor. The MC is designed to reject heat to the CWS and thereby maintain adequately low condenser pressure as recommended by the turbine vendor. Maintaining adequately low condenser pressure assures the efficient operation of the T-G and minimizes wear on the turbine last stage

buckets. The MC is not being modified to improve thermal performance for EPU operation. The performance of the condenser was evaluated fo r EPU. This evaluation was based on a design duty over the actual range of circulating water inlet temperatures, and confirms that the condenser is adequate for EPU operation. Condenser hotwell temperature limitations may require load reductions at the upper range of the anticipated circulating water inlet temperatures. EPU operation decreases the margin for the MC storage capacity from approximately 92 seconds at CLTP to 79 seconds at EPU. MC storage capac ity is less than the original design objective of 90 second holdup time for the decay of short-lived ra dioactive isotopes. However, this reduction of condensate retention time will have no significant effect on the radiati on level in the MC area because the major source is the N-16 activity in the MC exhaust steam. In addition, the small reduction in the condensate retention time will not significantly affect the radiation source in the Condensate Demineralizers and the FW system, because the major source in these systems is radioiodines, which have half-lives much longer than the reduction in the condensate retention time in the hotwell. The absolute value in lbm/hr of the steam bypasse d to the MC during a load rejection event is not increased for EPU as discussed in Section 2.5.4.3. Because the mass flow and pressure of the bypass steam do not increase due to EPU, the condenser remains adequate for the bypass blowdown conditions. Therefore, the Condenser and SJAEs for GGNS meet all CLTR dispositions.

Conclusion The effects of the proposed EPU on the MC system have been considered and the effects of changes in plant conditions on the design of the MC system have been adequately addressed. Entergy concludes that the MC system will continue to maintain its ability to withstand the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-221 blowdown effects of the steam from the TBS and thereby continue to meet GDC-60 with respect to controlling releases of radioactive effluent

s. Therefore, Entergy finds the proposed EPU acceptable with respect to the MC system.

2.5.4.3 Turbine Bypass Regulatory Evaluation The TBS is designed to discharge a stated percentage of rated MS flow directly to the MC system, bypassing the turbine. This steam bypass en ables the plant to take step-load reductions up to the TBS capacity without the reactor or turbine tripping. The system is also used during startup and shutdown to control reactor pressu re. The review of th e TBS focused on how it affects the load rejection capability at EPU, analysis of postulated system piping failures, and the consequences of inadvertent TBS operation. Th e regulatory acceptance criteria for the TBS are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents (including pipe breaks or malfunctions of the TBS); and (2) GDC-34, insofar as it requires that a RHR system be provided to transfer fission product decay heat and other residual heat from the reactor core at a rate such that SAFDLs and the design conditions of the RCPB are not exceeded.

GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The TBS is described in UFSAR Section 10.4.4.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 7.3 of the CLTR a ddresses the effect of CPPU on the TBS. The results of this evaluation are described below. The TBS provides a means of accommodating excess steam generated during normal plant maneuvers and transients. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-222 Topic CLTR Disposition GGNS Result Turbine Steam Bypass (Safety Analysis) [[

     ]] Meets CLTR Disposition The CLTR states that the increase in steam flow reduces the relative capacity of the TBS.

[[

          ]] Each of 3 bypass valves is designed to pass a steam flow of 1.92 Mlbm/hr, resulting in a system bypass capacity of 5.77 Mlbm/hr. The bypass capacity in terms of mass flow is not changed due to EPU. At the GGNS EPU conditions, rated steam flow is 18.968 Mlbm/hr; the system bypass capability in terms of rated steam flow is 30.4%. The bypass capacity at GGNS remains adequate for normal operational flexibility at EPU RTP. Therefore, the GGNS steam bypass capacity used in the turbine steam bypass safety analysis meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the TBS have been reviewed. Entergy concludes that the effects of changes in plant conditions on the de sign of the TBS have been adequately accounted for. Entergy concludes that the TBS will continue to mitigate the effects of MSIV leakage during a LOCA and provide a means for shutting down the plant during normal operations. Entergy further concludes that TBS failures will not adversely affect essential SSCs. Based on this, Entergy concludes that the TBS will continue to meet GDCs 4 and

34. Therefore, Entergy finds the proposed EPU acceptable with respect to the TBS.

2.5.4.4 Condensate and Feedwater Regulatory Evaluation The condensate and feedwater system (CFS) provides FW at a particular temperature, pressure, and flow rate to the reactor. The review focused on how the proposed EPU affects previous analyses and considerations with respect to the capability of the CFS to supply adequate FW during plant operation and shutdown, and isolate components, subsystems, and piping in order to preserve the system's safety function. The re gulatory acceptance criteria for the CFS are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-223 operation including possible fluid flow instabilities (e.g., water hammer), maintenance, testing, and postulated accidents; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform th eir safety functions; and (3) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided, and that the system be provided with suitable isolation capabilities to assure the safety function can be accomplished with electric power available from only the on-site system or only the off-site system, assuming a single

failure. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 is not applicable to GGNS; see UFSAR Section 3.1.2.1.5. The CFS is described in UFSAR Section 10.4.7.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 7.4 of the CLTR addr esses the effect of CPPU on the Condensate and Feedwater Systems. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Feedwater and Condensate Systems [[

     ]] Meets CLTR Disposition The CLTR states that the increase in power level increases the FW requirements of the reactor.

The FW and Condensate systems are required for normal plant operation and are not safety-related. The FW and Condensate systems do not perform a system level safety-related function.

They are designed to provide a reliable supply of FW at the temperature, pressure, quality, and flow rate as required by the reactor. Their performance has a major effect on plant availability and

capability to operate at EPU conditions. Evaluations of operation at EPU conditions have confirmed that BOP system modifications are required to support power uprate. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-224 For EPU, the following FW and Condensate system modifications are being implemented to improve plant and system reliability and margins:

1. RFP turbines - provides increased flow at increased speed

2. Condensate full flow filtration (CFFF) with automatic bypass capability -eliminates crud loadings on the condensate demineralizers and reduces iron depositions to reactor

3. Low pressure FW heaters replacement - corrects existing degradation or size restrictions

4. Related instrumentation changes (see Table 2.4-2)

Normal Operation System operating flows at EPU increase to approximately 114% of rated flow at CLTP. The system flow margin is 14% at EPU. The FW and Condensate system modifications assure acceptable performance with the new system ope rating conditions, provided that 3 condensate, 3 condensate booster, 2 heater drain, and 2 RFPs are in operation. The FW heater design has been analyzed and the 1 st , 5 th and 6 th stage FW heaters have been verified to be acceptable for the higher FW heater flows, temperatures, and pressures for EPU. The 2 nd , 3 rd , and 4 th stage FW heaters are being replaced before the plant operates under EPU conditions. The 3 rd and 4 th stage FW heaters have deteri orated due to wear. The 2 nd stage FW heater is undersized at EPU conditions.

The overspeed setpoint on the RFP turbines will be increased to accommodate the increased speed demand for normal operations.

The RFP at GGNS is tripped on high discharge pressure, which prevents pump operation at shut-off-head conditions. This is a two-out-of-three action, which is unchanged from CLTP.

Transient Operation To account for FW demand transients, the FW system was evaluated and determined to have approximately 14% margin at EPU FW flow. For system operation with all system pumps available, the predicted operating parameters were acceptable and within the component capabilities. The evaluation was performed for the trip of a condensate, condensate booster, and feed pump to ensure system capacity is adequate. This evaluation confirmed that the condensate pump has the capability to supply the transient flow requirements. A CFFF system is being installed upstream of the condensate booster pumps, resulting in reduced available pressure at the pump suction. An automatic bypass around the CFFF system is included in the modification to ensure that adequate net positive suction head (NPSH) is maintained to the remaining condensate booster pumps in the event that one of the pumps trips. When only two of the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-225 three condensate booster pumps operate they will be at run-out conditions. This facilitates full feed flow and is acceptable for a short period of time. Pump bearing temperature and vibration are monitored and appropriate actions, including reducing power, are taken to prevent pump damage in this situation. A trip of an RFP at 100% power results in a transient power level of 81%. No system bypass valves are required to open on an RFP trip. This meets the transient analysis requirements for loss of a single feed pump. The single pump operation at 81% power is limited by the pump original equipment manufacturer (OEM) to 1 to 3 hours for 3 to 4 occurrences per year to prevent pump deterioration. After 1 to 3 hours, the tripped pump must be returned to service or the unit must be

de-rated to 78.5% power. Condensate Demineralizers The effect of EPU on the condensate demineralizer system was reviewed. The condensate demineralizers can support full flow operation without modification. A full flow iron filtration system is being installed to eliminate crud loadings on the condensate demineralizers. The condensate demineralizers will experience reduced loadings resulting in unchanged or increased run times. The precoat filters were designed for condensate cleanup at approximately 33% full condensate flow during plant startup only and are not affected by EPU. Therefore, the feedwater and condensate systems meet all CLTR dispositions. Conclusion The effects of the proposed EPU on the CFS have been reviewed and adequately accounted for in the CFS design. Entergy concludes that the CFS will continue to maintain its ability to satisfy FW requirements for normal operation and shutdown, withstand water hammer, maintain isolation capability in order to preserve the system safety function, and not cause failure of safety-related SSCs. Entergy further concludes that the CFS will continue to meet the requirements of GDCs 4 and 44. Therefore, En tergy finds the proposed EPU acceptable with respect to the CFS.

2.5.5 Waste

Management Systems 2.5.5.1 Gaseous Waste Management Systems Regulatory Evaluation The GWMSs involve the gaseous radwaste system, which deals with the management of radioactive gases collected in the offgas system or the waste gas storage and decay tanks. In addition, it involves the management of the condenser air removal system; the gland seal exhaust and the mechanical vacuum pump operation exhaust; and the building ventilation system exhausts. The review focused on the effects that the proposed EPU may have on: (1) the design NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-226 criteria of the GWMSs; (2) methods of treatment; (3) expected releases; (4) principal parameters used in calculating the releases of radioactive materials in gaseous effluents; and (5) design features for precluding the possibility of an explosion if the potential for explosive mixtures exists. The regulatory acceptance criteria for GWMSs are based on: (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials

released at the boundary of the unrestricted area do not exceed specified values; (2) GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; (3) GDC-60, insofar as it requires that the plant design include means to control the release of radioac tive effluents; (4) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate confinement; and (5) 10 CFR 50, Appendix I, Sections II.B, II.C, and II.D, which set numerical guides for design objectives and limiting conditions for operation (LCOs) to meet the "as low as is reasonably

achievable" (ALARA) criterion. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The gaseous radwaste system is described in UFSAR Section 11.3.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 8.2 of the CLTR addresses the effect of CPPU on Gaseous Waste Management. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Off-Site Release Rate [[ Meets CLTR Disposition Recombiner Performance

     ]] Meets CLTR Disposition NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-227 2.5.5.1.1 Off-Site Release Rate The CLTR states that under EPU conditions, Offgas System functions other than the recombiner and related components are not significantly affected by power uprate. The CLTP design basis radiolytic gas production rate is 0.067 cfm/MWt. Based on the CLTR, this rate is conservative. The actual radiolytic gas production rate is 0.044 cfm/MWt. This is a constant rate. As this rate is proportional to reactor power, the radiolytic gas flowrate is expected to increase in proportion to the change in power, approximately 13% under EPU conditions. Because the actual radiolytic gas flowrate at EPU conditions is within the design basis (radiolytic gas) flowrate at CLTP, the design basis production value is retained at EPU conditions. As such, the CLTP design basis is maintained at EPU conditions and an evaluation

was conducted. This evaluation verified that all SSCs of the offgas system were acceptable for EPU operation. The primary function of the GWMS is to process and control the release of gaseous radioactive effluents to the site environs so that the total radiation exposure of persons in off-site areas is within the guideline values of 10 CFR 50, Appendix I. The GWMS involves the management of the condenser air removal system, gland seal exhaust and mechanical vacuum pump operation exhaust. The condenser offgas system radiological release rate is administratively controlled to remain within existing site release rate limits and is a function of fuel cladding performance, MC air inleakage, charcoal adsorber inlet dew point, and charcoal adsorber temperature. GGNS has TS requirements and administrative controls to limit fission gas releases to the environment. Plant procedures for reducing power, identifying a nd suppressing power near leaking fuel, and repairing condenser air inleakage exist and have been used at GGNS to maintain the offgas limits. These procedures are not affected by EPU. Further information regarding the production of noble gases at EPU conditions is found in Section 2.9.1.2. The GWMS (Offgas System) design criteria ensure that it will meet the plant licensing basis for controlling gaseous waste such that the total radi ation exposure of persons in off-site areas will be within the applicable guideline values of 10 CFR 20.1302 and 10 CFR 50, Appendix I. The plant gaseous waste licensing basis and the GWMS design criteria (for the Offgas portion) that support the licensing basis are unchanged by EPU. The GWMS will continue to satisfy this licensing basis under EPU operating conditions. The GWMS methods of treatment for radiological releases from the Offgas System consist of holdup and filtration to reduce the gaseous radioactiv ity that could be potentially released to off-site areas. The capacity and capability of the condenser offgas holdup and filtration system to adequately perform its design function are uncha nged by EPU. The evaluation of the turbine gland sealing system and mechanical vacuum pump system exhaust paths (as applicable to the GWMS) is contained in Section 2.7.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-228 The off-site release rate at GGNS meets all CLTR dispositions. 2.5.5.1.2 Recombiner Performance The CLTR states that under EPU conditions, core radiolysis increases linearly with reactor thermal power, thus increasing the heat load on the offgas recombiner and related components. The design features for precluding the possibility of an explosion include (a) dilution to control the concentration of hydrogen (steam is used upstream of the recombiner and air is used downstream of the recombiner) and (b) catalytic recombination to remove the combustible gas. The GWMS at GGNS is consistent with GEH desi gn specifications for radiolytic flowrate, and the GGNS-specific value for radiolytic gas flowrate is 0.044 cfm/MWt, which is well below the GGNS site-specific design value of 0.067 cfm/MWt (130ºF and 1 atm.). Therefore, the recombiner and condenser, as well as downstream system components, are designed to handle the increase in thermal power of the EPU. The combustible gas control component design requirements are determined by the quantity of radiolytic hydrogen and oxygen, which is expected to increase in proportion to the EPU power increase. The additional radiolytic hydrogen will also increase the catalytic recombiner temperature and offgas condenser heat load. These increases have been evaluated and it has been confirmed that sufficient margin remains in the GGNS offgas system component design to ensure that the gaseous radwaste system will continue to satisfy the plant licensing basis. The recombiner performance at GGNS meets all CLTR dispositions.

Conclusion The GWMSs have been reviewed with respect to the effects of the EPU. Entergy concludes that it has adequately accounted for the effects of the increase in fission product and amount of gaseous waste on the abilities of the systems to control releases of radioactive materials and preclude the possibility of an explosion, if the potential for explosive mixtures exists. Entergy finds that the GWMSs will continue to meet their design functions following implementation of the EPU. Entergy further concludes that the GWMSs will continue to meet the requirements of 10 CFR 20.1302; GDCs 3, 60, and 61; and 10 CFR 50, Appendix I, Sections II.B, II.C, and II.D. Therefore, Entergy finds the proposed EPU acceptable with respect to the GWMSs. 2.5.5.2 Liquid Waste Management Systems Regulatory Evaluation The review of liquid waste management systems (LWMSs) focused on the effects that the proposed EPU may have on previous analyses and considerations related to the LWMSs' design, design objectives, design criteria, methods of treatment, expected releases, and principal parameters used in calculating the releases of radioactive materials in liquid effluents. The NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-229 regulatory acceptance criteria for the LWMSs are based on: (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not ex ceed specified values; (2) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; (3) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate confinement; and (4) 10 CFR 50, Appendix I, Sections II.A and II.D, which set numerical guides for dose design objectives and LCOs to meet the ALARA criterion. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The LWMS is described in UFSAR Section 11.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 8.1 of the CLTR addresses the effect of CPPU on Liquid Waste Management. The results of this evaluation are described below. As stated in Section 8.1 of the CLTR, the Liquid Radwaste System collects, monitors, processes, stores and returns processed radioactive waste to the plant for reuse or for discharge. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Waste Volumes [[ Meets CLTR Disposition Coolant Fission and Corrosion Product Levels

     ]] Meets CLTR Disposition 2.5.5.2.1 Waste Volumes

The CLTR states that increased power levels and steam flow result in the generation of slightly higher levels of liquid radwaste. [[

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2-230

     ]] For GGNS, the single largest change in liquid waste volume due to EPU is from the backwash of condensate demineralizers. The increased flow rate through the condensate demineralizers results in a reduction in the average time between backwashes. This reduction does not affect plant safety. The only other consequential effect of EPU on the LWMS is primarily a result of the increased load on RWCU filter/demineralizers. The RWCU filter/demineralizers require more frequent backwashes due to slightly highe r levels of activation and fission products. Other increases in the LWMS load, such as increased leakage due to system condition changes, are minimal. The increased condensate demineralizer backwashes and RWCU filter/demineralizer loads are expected to increase the volume of liquid waste processed by the LWMS due to EPU by less

than 1%. The existing confined liquid storage and waste processing capacity can accommodate this increase with no changes. Additionally, a CFFF - iron control modification is being installed upstream of the condensate demineralizers to reduce the corrosion product loading on the demineralizer resins. The condensate demineralizers will require less frequent backwashes at EPU as a result. The effect on waste volumes as a result of the CFFF modification being installed upstream of the condensate demineralizers is expected to be a net reduction in liquid radwaste processed by the LWMS. Because the liquid volume increases less than 1%, the current design and operation of the LWMS will accommodate the effects of EPU with no changes. The existing equipment and procedures that control releases to the environment will continue to meet GDC-60 and GDC-61 requirements and ensure that releases remain within the applicable guideline values of 10 CFR 20.1302, 10 CFR 50 Appendix I, and 40 CFR 190. Therefore, the waste volumes meet all CLTR dispositions. 2.5.5.2.2 Coolant Fission and Corrosion Product Levels

The CLTR states that increased power levels and steam flow result in the generation of slightly higher levels of coolant concentrations of fission and corrosion products.

The coolant activation and corrosion products are sli ghtly increased as a result of EPU as discussed Section 8.4 of the CLTR. For the purpose of evaluating the radiological effects of the EPU, it was assumed that the operational radiological sources increased by up to 32% relative to CLTP. The design basis NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-231 fission and corrosion product source term is greater than the estimated EPU source term by a factor of approximately 7. Therefore, the current design basis sources remain bounding. Because the radiological sources remain bounded by the existing design basis, the current design and operation of the LWMS will accommodate the e ffects of the EPU with no changes, and the existing equipment and procedures that control releases to the environment will continue to meet GDC-60 and GDC-61 requirements. Releases remain within the applicable guideline values of 10 CFR 20.1302, 10 CFR 50 Appendix I, and 40 CFR 190. Therefore, the coolant fission and corrosion product levels meet all CLTR dispositions.

Conclusion The effects of the EPU on the LWMSs have been evaluated. Entergy concludes that it has adequately accounted for the effects of the increase in fission product and amount of liquid waste on the ability of the LWMSs to control releases of radioactive materials. Entergy finds that the LWMSs will continue to meet their design functions following implementation of the proposed EPU. Entergy further concludes that the LWMSs will continue to meet the requirements of 10 CFR 20.1302; GDCs 60 and 61; and 10 CFR 50, Appendix I, Sections II.A and II.D. Therefore, Entergy finds the proposed EPU acceptable with respect to the LWMSs. 2.5.5.3 Solid Waste Management Systems Regulatory Evaluation The review of the solid waste management systems (SWMSs) focused on the effects that the proposed EPU may have on previous analyses and considerations related to the design objectives in terms of expected volumes of waste to be processed and handled, th e wet and dry types of waste to be processed, the activity and expected radionuclide distribution contained in the waste, equipment design capacities, and the principal parameters employed in the design of the SWMS. The regulatory acceptance criteria for the SWMS are based on: (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not ex ceed specified values; (2) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; (3) GDC-63, insofar as it requires that systems be provided in waste handling areas to detect conditions that may result in excessive radiati on levels; (4) GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including AOOs, and postulated accidents; and (5) 10 CFR 71, which states requirements for radioactive material packaging. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-232 GGNS Current Licensing Basis NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The solid radwaste system is described in UFSAR Section 11.4 and the process and effluent radiation monitoring systems are described in UFSAR Section 11.5. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 8.1 of the CLTR addresses the effect of CPPU on Solid Waste Management. The results of this evaluation are described below. The Solid Radwaste System collects, monitors, pr ocesses, and stores processed radioactive waste prior to off-site disposal. GGNS meets all CLTR dispositions. The topics considered in this section are: Topic CLTR Disposition GGNS Result Coolant Fission and Corrosion Product Levels [[

Meets CLTR Disposition Waste Volumes

     ]] Meets CLTR Disposition 2.5.5.3.1 Coolant Fission and Corrosion Product Levels The CLTR states that increased power levels and steam flow result in the generation of slightly higher levels of coolant concentrations of fission and corrosion products.

For the purpose of evaluating the radiological effects of the EPU, it was assumed that the operational radiological sources increased by up to 32%. The design ba sis fission and corrosion product source term is greater than the estimated EPU source term by a factor of approximately 7. Therefore, the current design basis sources remain bounding. Because the radiological sources remain bounded by the existing design basis, the current design and operation of the SWMS will accommodate the effects of the EPU with no changes. The existing equipment and procedures that control waste shipments and releases to the environment will continue to meet GDC-60 and 10 CFR 71 requirements and ensure that releases remain within the applicable regulatory guidance of 10 CFR 20.1302. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-233 Radiation effluent limits of 10 CFR 20.1302 and monitoring requirements of GDC-63 and GDC-64 are independent of reactor thermal pow er and therefore are not affected by EPU . Therefore, the coolant fission and corrosion product levels meet all CLTR dispositions. 2.5.5.3.2 Waste Volumes The CLTR states that increased power levels and steam flow result in the generation of slightly higher levels of liquid and solid radwaste. The waste streams for the SWMS are: (1) dry ac tive waste, (2) spent ion exchange resin, and (3) filter sludge. EPU does not affect dry active waste so the volume and mix of dry active waste is unchanged. The effect of EPU on the SWMS is primarily a result of the increased load on the RWCU and condensate demineralizers. The increased demineralizer loads are expected to increase the volumes of spent ion exchange resin and filter sludge. Condensate demineralizer resin is no longer regenerated; fouled resin is cleaned and eventually depleted resin is replaced. Conservatively assuming that the resin replacement frequency increases due to higher condensate flowrate at EPU, the result is that the anticipated increase in solid radwaste volume is less than 4%, which is within the capacity of the collection and processing system. EPU does not generate a new type of waste or create a new waste stream. Therefore, the types of waste that require shipment are unchanged. A CFFF - iron control modification is being installed upstream of the condensate demineralizers to reduce the corrosion product loading on the demi neralizer resins. The installed CFFF filters are expected to reduce the total quantity of depleted ion exchange resins. Because the solid volume increases less than 4%, the current design and operation of the SWMS will accommodate the effects of EPU with no changes. The existing equipment, instrumentation, and procedures that control waste shipments and releases to the environment will continue to meet GDC-60, 63, and 64 and 10 CFR 71 requirements and ensure that releases remain within the applicable regulatory guidance of 10 CFR 20.1302. Therefore, the waste volumes meet all CLTR dispositions. Conclusion The effects of the EPU on the SWMSs have been evaluated. Entergy concludes that it has adequately accounted for the effects of the increase in fission product and amount of solid waste on the ability of the SWMS to process the waste. Entergy finds that the SWMS will continue to meet its design functions following implementa tion of the proposed EPU. Entergy further concludes that it has demonstrated that the SWMS will continue to meet the requirements of 10 CFR 20.1302; GDCs 60, 63, and 64; and 10 CFR 71. Therefore, Entergy finds the proposed EPU acceptable with respect to the SWMS. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-234 2.5.6 Additional Considerations 2.5.6.1 Emergency Diesel Engine Fuel Oil Storage and Transfer System Regulatory Evaluation Nuclear power plants are required to have redundant on-site emergency power supplies of sufficient capacity to perform their safety func tions (e.g., power diesel engine-driven generator sets), assuming a single failure. The review focused on increases in emergency diesel generator electrical demand and the resulting increase in the amount of fuel oil necessary for the system to perform its safety function. The regulatory acceptance criteria for the emergency diesel engine fuel oil storage and transfer system are base d on: (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects, including missiles, pipe whip, and JI forces associated with pipe breaks; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform thei r safety functions; and (3) GDC-17, insofar as it requires on-site power supplies to have sufficient independence and redundancy to perform their safety functions, assuming a single failure. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 does not apply to GGNS; see UFSAR Section 8.3.1.1.2.5. The Emergency Diesel Engine Fuel Oil Storage and Transfer System is described in UFSAR Section 9.5.4.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 6.8 of the CLTR addr esses the effect of CPPU on other systems not addressed in the CLTR. It concludes that system s not specifically addressed in the CLTR are not significantly affected by the power uprate. The emergency diesel engine fuel oil storage and transfer system is not addressed in the CLTR, and this disposition applies to GGNS. Emergency loads are computed based on equipment nameplate data or bhp with conservative demand factors applied. EPU conditions are achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated bhp for the required pump motors. In addition, UFSAR Sections 8.3.1.1.4.1 and 9.5.4.2 define the mission time as it relates to the diesel fuel oil system as each storage tank cont aining sufficient fuel oil for continuous operation NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-235 of its respective diesel generator for 7 days. EPU does not affect this mission time. 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 EPU. Therefore, the amount of power required to perform safety-related functions (pump and valve loads) is not increased with EPU, and the current emergency power system remains adequate. The systems have sufficient capacity to support all required loads to achieve and maintain safe shutdown conditions and to operate the ECCS equipment following postulated accidents and transients. Because the loads and mission times are not changed for EPU, no changes to the emergency diesel engine fuel oil storage and transfer system are necessary. Conclusion The amount of required fuel oil for the emergenc y diesel generators has been reviewed and found to adequately account for the effects of EPU on fuel oil consumption. Entergy concludes that the fuel oil storage and transfer system will continue to provide an adequate amount of fuel oil to allow the diesel generators to meet the on-site power requirements of GDCs 4 and 17. Therefore, Entergy finds the proposed EPU acceptabl e with respect to the fuel oil storage and transfer system.

2.5.6.2 Light Load Handling System (Related to Refueling) Regulatory Evaluation The light load handling system (LLHS) includes components and equipment used in handling new fuel at the receiving station and the loading of SF into shipping casks. The review covered the avoidance of criticality accidents, radioactivity releases resulting from damage to irradiated fuel, and unacceptable personnel radiation exposures. The review focused on the effects of the new fuel on system performance and related anal yses. The regulatory acceptance criteria for the LLHS are based on: (1) GDC-61, insofar as it requires systems containing radioactivity be designed with appropriate confinement and with suitable shielding for radiation protection; and (2) GDC-62, insofar as it requires that criticality be prevented. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The fuel handling systems are described in UFSAR Section 9.4.1. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-236 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.8 of the CLTR addr esses the evaluation of the effect of the EPU on several plant systems that were not addressed elsewhere in that report. The LLHS (related to Fuel Handling) is one of the systems so eval uated (see Table 2.5-5, Item 18). CLTR Section 6.8 is supported by ELTR1 (Reference 2), Section 5.12 a nd Appendix J, also previously approved by the NRC for use as guidelines for EPUs. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Other Systems [[

     ]] Meets CLTR Disposition The EPU has been found to not have any significant effect on the LLHS.

The LLHS meets the CLTR disposition.

Conclusion The effects of the new fuel on the ability of th e LLHS to avoid criticality accidents have been reviewed. Entergy concludes that GGNS has co rrectly applied the conclusion of the CLTR Section 6.8 as having no effect. Based on this review, Entergy further concludes that the LLHS will continue to meet the requirements of GDCs 61 and 62 for radioactivity releases and prevention of criticality accidents. Therefore, Entergy finds the proposed EPU acceptable with

respect to the LLHS.

2.5.7 Additional

Review Areas (Plant Systems) NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.8 of the CLTR addr esses the evaluation of the effect of the EPU on several plant systems that were not addressed elsewhere in that report. The systems included in this evaluation are listed in Table 2.5-

5. CLTR Section 6.8 is supported by ELTR1 (Reference 2), Section 5.12 and Appendix J, also previously approved by the NRC for use as guidelines for EPUs. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are:

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2-237 Topic CLTR Disposition GGNS Result Other Systems [[

     ]] Meets CLTR Disposition The EPU has been found to not have any significant effect on the systems in Table 2.5-5. The assessment of other systems meets the CLTR disposition.

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2-238 Table 2.5-1 Appendix R Fire Event Evaluation Results Parameter 102.46% CLTP 1, 4 EPU 1, 4 App. R Criteria PCT ( F) 595 2 597 2 1500ºF Maximum Operator Action Time to Open ADS valves (minute) 18 14.3 See Note 3 Peak DW Pressure (psia) 19.5 19.5 < 44.7 Peak WW Airspace Pressure (psia) 19.0 19.3 29.7 Peak DW Temperature ( F) 223.8 234.0 < 330 Peak WW Airspace Temperature ( F) 140.9 149.2 185 SP Bulk Temperature ( F) 173 5 / 173.9 181.4 185 Notes: 1. Using SAFER/GESTR-LOCA and SHEX methodologies.

2. Initial steady-state fuel cladding temperature.

3. The maximum ADS actuation time should allow the core to remain covered with a short fuel uncovery period permitted, providing the PCT acceptance criterion is met.

4. Reactor vessel pressure remains low enough to en sure no risk of reactor vessel overpressure.

5. Original analysis based on simplified SP model.

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2-239 Table 2.5-2 SGTS Iodine Removal Capacity Parameters [[

]]

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2-240 Table 2.5-3 Spent Fuel Pool Response Normal Offload, Full Cooling Capability 2 FPCCS pumps, 2 FPCCS HXs. 2,200 gpm total SFP flow, 2,130 gpm total CCW flow. CCW Temperature ( F) Start of Offload (hrs after S/D) Max. SFP Temperature ( F) Time to Boil from Max. Temperature (hrs) Makeup Flow Required at Boiling (gpm) 95 110 146 6.3 58 95 220 140 7.8 50 Full-Core Offload, Full Cooling Capability 1 RHR pump, 1 RHR HX. 3,000 gpm SFP flow, 7,900 gpm SSW flow. SSW Temperature ( F) Start of Offload (hrs after S/D) Max. SFP Temperature ( F) Time to Boil from Max. Temperature (hrs) Makeup Flow Required at Boiling (gpm) 90 24 134 3.3 119 Normal Offload, Single Failure 1 FPCCS pump, 2 FPCCS HXs. 1,600 gpm total SFP flow, 2,508 gpm total SSW flow. SSW Temperature ( F) Start of Offload (hrs after S/D) Max. SFP Temperature ( F) Time to Boil from Max. Temperature (hrs) Makeup Flow Required at Boiling (gpm) 86 110 149 6.3 58

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2-241 Table 2.5-4 EPU TBCW Effect Power Level Total System Heat Load MBTU/hr Total System Capa city MBTU/hr CLTP 92.1 141.4 EPU 96.7 141.4

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2-242 Table 2.5-5 Basis for Classification of No Significant Effect [[

1

2

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2-243 [[

3

4

5

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-244 [[

6

7

8

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-245 [[

9

10

11

12

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-246 [[

13

14

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-247 [[

15

16

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-248 [[

17

18

19

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-249 [[

20

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-250 [[

21

22

]]

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2-251 Figure 2.5-1 Appendix R Evaluation Results EPU Containment (SHEX) - SP Temperature LL (-') L1J o 300. 200. 100. =::) f----<C Q" L1J CL 2:: L1J f----GLI 021010 -----------------

4124B39F 1403.9 1 GGNS EPU Tn 1 1 ---------10 1 TIME -SEC 1 SP TEMP lTV P 1 1 1 10 4 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-252 Figure 2.5-2 Appendix R Evaluation Results 102.46% CLTP Containment (SHEX) - SP Temperature

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2-253 Figure 2.5-3 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) - Hot and Average Channel Water Level

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-254 Figure 2.5-4 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) - Reactor Vessel Pressure uJ 0::: =:) CD CD LlJ 0::: GRAND GULF 1 1 VESSEL PRESSURE CASE 1 (6 SRV/l LPC ) EPU GGNS EPU APPENDIX-R SAFER ANALYSIS xl0' ,1""1,,,,1 O. 1.5 2. xl 0' LANDER 2C626 0.5 TIME 1. (SECOND) 20100107 0949.4 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-255 Figure 2.5-5 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) - Peak Cladding Temperature

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-256 Figure 2.5-6 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) - Water Level Outside the Shroud

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2-257 Figure 2.5-7 Reactive Capability Curve

  ,,_,.tOf TYPO 180110*18 upr"'(_1 P"'* (---I ,o,pp..ftO

' PO,..' , 1 600.00 MV" 1 525,00 MV" Ar matur. V ol laij. , 22.00 k V Ar m.tuft CUITOI1' 41.989 "'" 40.0 2 1 F.-.q,,"ocy 60.0 " , P ow.< FoctOf P.f. 0 , 9 00 .. " 5. 110 4.1 40 Cold G .. n 4 0.0 'C 40,0 'C * " "' "' " ** M v ar '00 ! I I V * --, --V /. /. / / --r-/ /. -* I / V --/ , '00 V /. V \ , I V /. ./ / , * / V /. V , , '00 * , I / / /. /. ./ , -. w /. / I-w , , 00 -,

1.0 M '00 JOO 400 500 IlQO 1 00 !IOO 000 t OOl "00 1100 1)00 "00 ,", , " 00 '00 . '-. -, -, " '-'-. . -. -w . ,00 \ \ '-... '-r-:::> -* ,,, \ '-... '-r'-\ \ '-... ." "-'-. * , I'-'-'00 I \ \ " "-'-* , \ '-... ..., ...
M v a r
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2-258 2.6 Containment Review Considerations

2.6.1 Primary

Containment Functional Design Regulatory Evaluation The containment encloses the reactor system and is the final barrier against the release of significant amounts of radioactive fission products in the event of an accide nt. The review for the primary containment functional design covered: (1) the temperature and pressure conditions in the DW and WW due to a spectrum of postu lated LOCAs; (2) the DP across the operating deck for a spectrum of LOCAs; (3) SP dynami c effects during a LOCA or following the actuation of one or more RCS safety/relief valv es; (4) the consequences of a LOCA occurring within the containment (WW); (5) the capability of the containment to withstand the effects of steam bypassing the SP; (6) the SP temperature limit during RCS safety/relief valve operation; and (7) the analytical models used for containment analysis. The regulatory acceptance criteria for the primary containment functional design are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and that such SSCs be protected against dynamic effects; (2) GDC-16, insofar as it requires that reactor containment be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment; (3) GDC-50, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated temperature and pressure conditions resulting from any LOCA; (4) GDC-13, insofar as it requires that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation and for accident conditions, as appropriate, to assure adequate safety; and (5) GDC-64, insofar as it requires that means be provided to monitor the reactor containment atmosphere for radioactivity that may be released from normal operations and from postulated accidents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The GGNS primary containment functional desi gn is described in UFSAR Section 6.2.1. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-259 effects of CPPUs. Section 4.1 of the CLTR addresses the effect of CPPU on Primary Containment Functional Design. The results of this evaluation are described below. The UFSAR provides the containment responses to various postulated accidents that validate the design basis for the containment. EPU operation changes some of the conditions for the containment analyses. For example, the short-term DBA LOCA containment response during the blowdown is governed by the bl owdown flow rate. This blowdown flow rate is dependent on the reactor initial thermal-hydraulic conditions, such as vessel pressure and the mass and energy of the vessel fluid inventory, which change slightly with EPU. Also, the long-term heat-up of the SP following a LOCA or a transient is governed by the ability of the RHR to remove decay heat. Because the decay heat depends on the initial reactor power level, the long-term containment response is affected by EPU. The containment response was reanalyzed to demonstrate the plant's capability to operate with a rated power increase to 4,408 MWt. The key plant parameters used to model and analy ze the plant response at EPU are provided in Table 2.6-2. The analyses of containment pressure and temp erature responses, as described in Section 2.6.1.1, were performed in accordance with RG 1.49 and ELTR1 using GEH codes and models. The M3CPT code was used to model the short-term containment pressure and temperature response. The modeling used in the M3CPT analyses is de scribed in References 49 and 50. References 49 and 50 describe the basic containment analytical models used in GEH codes. Reference 51 describes the more detailed RPV model (LAMB) used for determining the vessel break flow in the containment analyses for EPU at off-rated reactor conditions. The LAMB code models the recirculation loop as a separate pressure node. It also allows for inclusion of flashing in the pipe and vessel during the blowdown and flow choking at the jet pump nozzles when the conditions warrant. The use of the LAMB bl owdown flow in M3CPT was identified in ELTR1 by reference to the LAMB code qualification in Reference 51. The SHEX code was used to model the long-term containment pressure and temperature response. The key models in SHEX are based on models described in Reference 50. The GEH containment analysis methodologies have been applied to all BWR power uprate projects performed by GEH and accepted by the NRC. The effects of EPU on the containment dynamic lo ads due to a LOCA have been evaluated as described in Section 2.6.1.2. The LOCA hydrodynamic loads are defined in the GGNS Containment Load Report (CLR). UFSAR Appendi x 6A and 6D represent the GGNS CLR. The hydrodynamic loads defined for GGNS are based on the methods and assumptions recommended in Appendix 3B of GESSAR II. The LOCA hydrodynamic loads evaluated for EPU include loads during pool swell, condensation oscillation, and chugging. Pool swell loads include water jet and LOCA bubble drag loads on submerged st ructures, LOCA air bubble pressure loads on the DW wall (exterior), basemat, and containmen t wall (interior), drag loads and bulk and froth affect loads on equipment above the SP, and fallb ack loads. The specific application of these NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-260 loads to GGNS is described in Section 6A.4 of the UFSAR. The ev aluation of the LOCA containment dynamic loads is based primarily on the results of the short-term analysis described in Section 2.6.1.1. The SRV discharge load evaluation would normally consider any increases in the SRV opening setpoints for EPU. Because EP U does not change the SRV setpoints, the pressure related SRV loads do not change. The metal-water reaction energy vs. time relationship is calculated using the method described in RG 1.7 (Reference 52) as a normalized value (fraction of reactor thermal power). All of the energy from the metal-water reaction is assumed tr ansferred to the reactor coolant in the first 120 seconds into the LOCA. The metal-water reaction energy represents a small fraction of the

total shutdown energy transferred to the coolant. Therefore, it is concluded that the effect of an increase in metal-water reaction energy on containment response as a result of the EPU is negligible. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Pool Temperature Response [[ Meets CLTR Disposition Wetwell Pressure

Meets CLTR Disposition Drywell Temperature

Meets CLTR Disposition Drywell Pressure

Meets CLTR Disposition Containment Dynamic Loads

Meets CLTR Disposition Containment Isolation

Meets CLTR Disposition Motor-Operated Valves

Meets CLTR Disposition

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2-261 Topic CLTR Disposition GGNS Result Hardened Wetwell Vent System

Meets CLTR Disposition Equipment Operability

     ]] Meets CLTR Disposition 2.6.1.1 Containment Pressure and Temperature Response   The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. As a result of this, the pool temperature response, WW pressure, DW temperature, and DW pressure need to be addressed. Short-term and long-term containment analysis results are reported in the UFSAR. The short-term analysis is directed primarily at determining the containment pressure response during the initial bl owdown of the reactor vessel inventory to the containment following a large break inside the DW. Short-term containment response analyses were performed for the limiting DBA LOCA that assumes a double-ended guillotine break of a recirculation suction line or a MSL to demons trate that EPU does not result in exceeding the containment design limits. The long-term analyses are directed primarily at the pool temperature response, considering the decay heat addition to the SP. The DBA LOCA a nd ASDC event were both reanalyzed for EPU. Peak values of the containment pressure and temperature responses to the DBA LOCA are given in Table 2.6-1; however, it should be noted that the ASDC results are limiting for peak SP and WW temperature. Peak SP temperatures resulting from the postulated ATWS, SBO, and 10 CFR Part 50, Appendix R Fire events are given in Table 2.6-3.

The steam bypass capability for the GGNS containment was re-analyzed for EPU and addressed in Section 2.6.1.1.2. The effect of EPU on the events which yield the limiting containment pressure and temperature response is provided below. 2.6.1.1.1 Long-Term Suppression Pool Temperature Response 2.6.1.1.1.1 Bulk Pool Temperature The long-term bulk pool temperature response for EPU is evaluated for the limiting DBA LOCA in Section 6.2 of the UFSAR and the limiting Alternate Shutdown activity in Section 15.2 of the UFSAR. Per GE Safety Communication SC 06-01 (Reference 53), the potential was identified that a single failure that eliminated only the RHR HX could prove more limiting than the typically analyzed scenario of the single failure of an entire electrical bus. The SP temperature response evaluations discussed below have considered this single failure. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-262 The current design limit for the bulk SP temperature is 185°F. For EPU implementation, this design limit is increased to 210°F. The analysis of the DBA LOCA was performed at 102% of EPU RTP. The time-dependent SP and WW temperature response is presented in Fi gure 2.6-1 and the calculated peak values for LOCA bulk pool temperature for the current Anal ysis of Record (AOR) and the EPU RTP case are compared in Table 2.6-1. The EPU analyses were performed using a decay heat table applicable to GGNS and based on ANS/ANSI 5.1-1979 with 2-sigma adders with additional actinides and activation products per GE SIL 636 (Reference 54). The analysis assumed the single failure of one of the two RHR HXs. The resulting calculated peak bulk SP temperature is 189°F. This temperature is within the ECCS NPSH pump limit of 194°F. The highest bulk pool temperature response from a non-LOCA event results from an ASDC event. This event was also analyzed at 102% of EPU RTP. The limiting alternate shutdown activity assumes reactor isolation with availability of one RHR HX. The resulting time-dependent SP and WW temperature response is presented in Figures 2.6-2 and 2.6-3 and the peak bulk pool temperature is 198°F which also is within the ECCS NPSH pump limit of 212°F. 2.6.1.1.1.2 Local Pool Temperature with SRV Discharge The local pool temperature limit for SRV discha rge was originally specified in NUREG-0783 (Reference 55) because of concerns resulting from unstable condensation observed at high pool temperatures in plants without quenchers. Quen cher devices such as the X-quenchers used in GGNS mitigate these loads. Reference 56 provides justification for the elimination of this limit for plants where the ECCS suction strainers ar e located below the SRV quenchers. Because the suction strainers are below the SRV quenchers at GGNS, a local pool temperature analysis is not required. Therefore, the peak local SP temperature at GGNS is acceptable for EPU conditions. 2.6.1.1.2 Steam Bypass Capability The current steam bypass effective area capability, A K, which was established from the UFSAR analysis is 0.9 ft

2. Use of this effective steam bypa ss area at EPU conditions resulted in a containment pressure that exceeded the containment design pressure. For this reason, the steam bypass analyses were performed to establish the maximum allowable effective steam

bypass area with EPU conditions. These analyses determined that an effective steam bypass area of 0.8 ft 2 would maintain the peak calculated containment pressure within the design limit with EPU conditions. 2.6.1.2 Containment Dynamic Loads The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. Containment dynamic loads are addressed in the following sections. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-263 2.6.1.2.1 Loss-of-Coolant Accident Loads The LOCA containment dynamic loads analysis for EPU is primarily based on the short-term Recirculation Suction Line Break (RSLB) LOCA analyses and compliance with generic criteria developed through testing programs. The analyses were performed as described in Section 2.6.1.1 with break flows calculated using both a simple RPV model (Reference 50) and a more detailed RPV model (Reference 51). The NRC approved use of this latter model for the EPU containment evaluations in Reference 2. These analyses also provide calculated values for the controlling parameters for the dynamic loads throughout the blowdown. The key parameters are DW and WW pressures, vent flow rates, and SP temperature. The LOCA dynamic loads considered in the EPU evaluations include pool swell, CO, and chugging. The results of the EPU pool swell evaluations confirmed that the current pool swell load definition remains bounding. These loads occur within the first 5 seconds. The containment response conditions for EPU are within the range of test conditions used to define CO loads for the plant. The containment response conditions fo r EPU are within the conditions used to define the chugging loads. Therefore, the LOCA dynamic loads are not affected by EPU. 2.6.1.2.2 Safety Relief Valve Loads The SRVs provide pressure relief during reactor transients. Steam discharged from the SRVs is routed through the SRVDLs and through the SRVDL quencher into the SP. The SRV loads resulting from SRV operation include the reacti on and thrust loads acting on the SRVDL and quencher and the air-bubble loads which are transmitted to the submerged boundaries and structures. These loads and the basis for these loads as applied to GGNS are summarized in Appendix 6A of the UFSAR. The SRV dynamic loads may be affected by an increase in the actual SRV setpoint pressure. Because there is no increase to any of the SRV setpoints or setpoint tolerances for EPU, initial SRV actuation loads remain bounded at EPU. The SRV load definition for subsequent actuations for GGNS assumes that only one SRV opens during subsequent actuations. GGNS uses the low-low set (LLS) SRV setpoint logic to implement this assumption. Analysis performed at EPU conditions demonstrated that LLS successfully prevented subsequent actuations of multiple valves. If the time between SRV closure and SRV re-opening is reduced by EPU, the potential exists for subsequent actuation with an elevated water leg, wh ich could affect the SRV loads. A review of

the SRV load test data and the analyses which evaluated LLS SRV performance for GGNS EPU shows that the water leg inside the SRVDL returns to the initial (pre-actuation) water level or

lower level (depressed water leg condition) prior to subsequent actuations. Given the fact that containment dynamic loads ar e not affected by EPU and the current SRV load definition is still applicable, all CLTR dispositions are met.

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2-264 2.6.1.3 Containment Isolation The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. However, the system designs for containment isolation are not affected by EPU. The capabilities of isolation actuation devices to perform during normal operations and under post-accident conditions have been determined to be acceptable. Therefore, the GGNS containment isolation capabilities are not adversely affected by the EPU and all CLTR dispositions are met. 2.6.1.4 Generic Letter 96-06 The GGNS response to GL 96-06 was performed based on the post-accident containment pressure and temperature response for CLTP conditions. As the containment analysis presented within this section demonstrates, the containment pressure and suppression pool and containment temperatures increase as a result of the additi onal heat load associated with EPU. These increases were reviewed relative to the assumpti ons and analyses in the CLTP analyses, and it was concluded that the existing GGNS response remains valid for EPU and all CLTR dispositions are met. 2.6.1.5 Generic Letter 89-10 The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. The GGNS response to GL 89-10 was accomplished based on the post-accident containment pressure and temperature response for CLTP conditions. As the containment analysis presented within this section demonstr ates, the CLTP conditions are not significantly affected by EPU conditions. Therefore, the existing GGNS response remains valid for EPU and all CLTR dispositions are met. Conclusion The containment temperature and pressure tr ansient has been reviewed and was found to adequately account for the increase of mass and energy resulting from the proposed EPU. The review also demonstrated that containment systems will continue to provide sufficient pressure and temperature mitigation capability to ensure that containment integrity is maintained. Entergy concludes that containment systems and instrumentation will continue to be adequate for monitoring containment parameters and release of radioactivity during normal and accident conditions and will continue to meet the requirements of GDCs 4, 13, 16, 50, and 64 following implementation of the proposed EPU. Theref ore, Entergy finds the proposed EPU acceptable with respect to primary containment functional design. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-265 2.6.2 Subcompartment Analyses Regulatory Evaluation A subcompartment is defined as any fully or partially enclosed volume within the primary containment that houses high-energy piping and would limit the flow of fluid to the main containment volume in the event of a postulated pipe rupture within the volume. The review for subcompartment analyses covered the determination of the design DP values for containment subcompartments. The review focused on the effects of the increase in mass and energy release into the containment due to operation at EPU conditions, and the resulting increase in pressurization. The regulatory acceptance criteria for subcompartment analyses are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and that such SSCs be protected against dynamic effects; and (2) GDC-50, insofar as it requires that containment subcompartments be designed with sufficient margin to prevent fracture of the structure due to the calculated pressure differential conditions across the walls of the subcompartments. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The GGNS subcompartment analyses are described in UFSAR Section 6.2.1.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 10.1 of the CLTR addresses the effect of CPPU on Subcompartment Analyses. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Liquid Lines [[

     ]] Meets CLTR Disposition As stated in Section 10.1 of the CLTR, EPU ma y increase subcooling in the reactor vessel, which may lead to increased break flow rates for liquid line breaks. EPU may also result in an increase in the mass and energy release for liquid line breaks.

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2-266 An annular structure of reinforced concrete is located inside the DW around the RPV in order to provide thermal and radiation shielding, and is called the Biological Shield Wall (BSW). The BSW is designed to withstand the DP that woul d develop across the wall as a result of a high pressure pipe break within the annulus (i.e., between the RPV and the BSW). A pipe break in this region results in a combination of four dynamic loads, collectively referred to as AP loads. These four dynamic loads consist of: (1) the asymmetric pressurization of the annular area between the BSW and RPV, (2) the JR resulting from the break flow through the vessel nozzle, (3) the JI on the vessel of the break flow from the broken pipe, and (4) the effect load absorbed by the PWR. These loads are a function of the break size, location, fluid thermal-hydraulic conditions, and the annular vent area to the rest of the DW. The second subcompartment region analyzed is the volume between the RPV head and the DW head. Because of the limited vent area between the DW head compartment and the DW, a pipe break in either region can cause a DP load across the DW head refueling bulkhead plate. Section 2.5.1.3.1 provides further details on the discussion of additional containment subcompartment regions. Subcompartment Pressurization Evaluation The ODB mass and energy release rate profiles used in developing the asymmetric loads were calculated using the methods from NEDO-24548, "Annulus Pressurization Load Adequacy Evaluation" (Reference 57). Due to GEH's Safety Communication (SC) 09-01 (Reference 58), the large pipe break mass and energy release rates and annulus pressurization time histories at OLTP, CLTP, and EPU conditions were recalcu lated using TRACG (References 59 and 60). Because of the issues identified in SC 09-01, the simplistic instantaneous break NEDO-24548 mass and energy release methodology was judged to be potentially non-conservative as the method could potentially result in artificial shifts of the pressure response frequency content. Evaluations were performed to determine the effect of the AP load methodology change and EPU operation on the dynamic structural response of the RPV, reactor internals, piping and containment structures. Table 2.6-4 shows the ODB and EPU analysis methodologies. Original Design Basis The ODB annulus pressurization analysis is based on mass and energy release rates generated using the instantaneous break NEDO-24548 methodology. This methodology generates a simple step-function mass and energy release profile based on critical mass flux correlations, the initial thermodynamic conditions in the piping and fluid reservoir, and the break and minimum flow areas in the piping. The idealized NEDO-24548 profile does not account for fluid acceleration, flow losses in the piping, or the effects of flashing in the piping system. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-267 The ODB response spectra were used in the comparison with the new EPU analysis methodology. The review of the ODB exposed some inconsiste ncies with the current plant configuration as well as an error in the application of the MSL vectors. The ODB did not account for horizontal insulation panels that exist in the current plant. It was found that one of the force vectors from the JI, JR, and PWR loads was applied in th e incorrect direction resulting in a highly conservative analysis for most components. These inconsistencies and errors have been addressed in the GGNS Corrective Action Program and corrected in the EPU evaluations. The ODB breaks included the reactor recircula tion discharge line break (RDLB), the reactor RSLB, the feedwater line break (FWLB), and MSLB.

EPU Analysis Analyses were performed for the large piping segment breaks within the annulus for effects including the structural dynamic response of th e reactor vessel, reactor vessel internals, attachments to the vessel, and attachments to the BSW. Conditions analyzed include OLTP, CLTP, and EPU conditions, several power flow points along the MELLLA and the increased core flow (ICF) boundary from minimum core flow to maximum core flow through maximum EPU power. The breaks analyzed include the reactor recirculation discharge, reactor recirculation suction, and FW lines. The EPU effect on the corrected loads resulting from a MSLB was dispositioned according to the CLTR because the MS system's existing pipe rupture design bases evaluations remain the same as the OLTP conditions.

The AP loads analysis process starts with a break flow analysis that generates mass and energy release rates from the pipe break. The mass and energy release rates are then propagated throughout the annulus resulting in AP loads. The break flow calculations are also used as input to the JI and JR loads. JR loads from the pipe end of the break are input into PWR loads. Finally, AP, JI, JR, and PWR loads are input into a structural model. Results from the structural model are then compared to the ODB. Break Flow Analysis To minimize concerns that the mass and energy release methodology could result in non-conservative load estimates, a best-estimate mass and energy release methodology was employed for the GGNS EPU analysis. The EPU analysis uses a detailed TRACG model to generate accurate mass and energy release rates. The TRACG model consists of a full nuclear steam supply model that has been modified to model break s at the nozzle safe end to pipe welds for the recirculation suction, recirculation discharge and FW nozzles. To best mimic the actual geometry of the postulated doubled-ended guillotine break, two valve-type components are used to model the break region. The valves are connected under steady-state conditions and are NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-268 modified to connect to pressure boundary conditions during the transient computation, simulating an instantaneous break. The TRACG model used for the EPU evaluation provides a better estimate of the mass and energy releases resulting from breaks in the reci rculation suction, recirculation discharge, and FW lines. The use of TRACG mass and energy rel ease allows the effect of alternate operating conditions to be realistically predicted. Best-estimate TRACG computations modeling the entire reactor system produce more accurate pressure and enthalpy conditions for the break and also

account for the flow inertia in the piping. Results of the mass and energy release analysis are used in the AP analysis to determine the time-dependent AP response profiles. Annulus Pressurization Analysis

The original AP response analysis is performed using a coarse nodal model of the annulus (25 nodes). The EPU analysis utilizes the TRACG code with a fine nodal model of the annulus (384 nodes). The acoustic propagation is important in determining the forces acting on the vessel and shield wall for the subcompartment pressurization model. While the acoustic response is modeled in the AOR through the inertia inputs (length and area inputs) to the original model, the AOR calculation is crude due to coarse annulus nodalization. The TRACG model used in the EPU evaluation provides a better pr ediction of the acoustics through the use of a more accurate thermal-hydraulic model combined with an order of magnitude finer mesh used in the nodalization of the annulus. The use of TRACG with a fine nodal model allows a more accurate modeling of the pressure waves within the annulus. This leads to more accurate estimates of the time-dependent forces and moments on the shield wall, reactor vessel, and

attached piping and structures. The TRACG vessel component was used to evaluate the time-dependent distributed pressure response in the annulus based on a nodal model of the annulus. The Vessel component models the inertia and flow losses associated with th e annulus flow. The use of actual annulus dimensions in the definition of the TRACG annulus model ensures that the inertia effects are correctly modeled. Volume and area reductions are modeled for all significant blockages within the annulus. Loss coefficients associated with flow past blockages are based on standard loss coefficient formulations and the specific geom etry of each blockage. Blockages are also modeled in the definition of the junction hydraulic diameter and junction relative roughness terms. The vessel component model is a cylindrical mode l of the annulus that is based on a nodalization of the annulus in the z and directions. The model is a full 360° model of the annulus with 16 levels and 24 theta sectors. The nodalization scheme is based on: (1) uniform theta sectors, (2) cell height-to-width aspect ratios of approximately one, and (3) the modeling of all blockages at cell boundaries. Maintaining the aspect ratio of the cells close to one ensures that the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-269 nodalization does not distort the acoustic wave propagation. The model accommodates all break locations without necessitating a re-nodalization for each break. The TRACG model uses fill components to add the mass and energy from the break directly to the respective break cells. The input mass and energy release rates are generated with the TRACG mass and energy release models described in the previous section. The TRACG model employs valve components and associated control logic to model the dynamic vent paths associated with the release of the horizontal insulation panels within the annulus. The associated control logic simulates th e panel release pressure, the pairing of panels that are assumed to release together, the vent area as a function of time after opening, and the vent loss coefficient as a function of vent area. The initial conditions in the annulus are assumed to be the same as those modeled in the AOR. The results of the AP analysis are used in downstream analyses to develop time-dependent forces and moments on the reactor shield wall and reactor vessel.

JI, JR and PWR Analyses The ODB methodology for calculating the RSLB, RDLB and FWLB JR/JI loads on the RPV and the BSW is documented in UFSAR Chapter 3.6. In the ODB methodology, for the breaks with flow diverters (RSLB and FWLB), the nozzle thru sts were not taken into consideration and the diverter thrust was applied to the RPV instead of the BSW. In addition, the methodology utilizes a simplistic approach to calculate the mass and energy release time history. Because of the issues raised in SC 09-01, more realistic analysis methodologies were employed in order to

avoid introducing unrealistic and potentially non-conservative biases into the load definitions applied to the structural model. To address these inconsistencies, a refined methodology was employed for the calculation of the JR and JI loads. The JR and JI load on the RPV has a sudden rise at the moment of the RSLB, because the RPV is the source of the high pressure fluid. However, the BSW does not experience the JR and JI load until the flow diverter is pressurized by the break flow. Therefore, the load on the BSW increases in a ramp fashion as the flow diverter is pressurized. In order to capture this effect, the mass and momentum balances were made on the RPV and the flow

diverter as the break gap opens up and the break flow increases as a function of time. The loads were calculated at each moment using Equati on 6-2 of ANSI/ANS-58.2-1988. The flow rates in the break itself, and in the flow diverter, we re calculated by the critical flow using the Moody Homogeneous Equilibrium Model (HEM). This calculation assumed that the flows in the

nozzle, pipe, and the flow diverter are establis hed instantaneously (i.e., reach the quasi-steady state values). The calculation procedur e followed ANSI/ANS-58.2-1988, and the guidelines

provided in paragraphs (2) and (3) of Section III.2.C of NRC Standard Review Plan (SRP) 3.6.2. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-270 GEH uses the Pipe Dynamic Analysis (PDA) code (Reference 61) to determine PWR loads. The analytical approach is used to determine the re sponse of a pipe subjected to the thrust force occurring after a pipe break. To limit the pipe movement, a PWR is added to inhibit the motion following the break. Nonlinear differential equati ons are used by the PDA code to describe the motion of the pipe whip and the restraining device. The break flow time histories are used as input to the PDA code to define the pipe whip force characteristics with respect to the mass and energy release rate from the broken pipe end. The code sums the kinetic energy after each increment of the pipe angular displacement and uses the energy equations to solve for the PWR force and displacements. The time histories of the JI and JR forces were averaged within a specific time step, and the force versus time identified as step functions were then used in the PDA input. Only the positive forces from the time history data were used to perform the averaging of forces such that it resulted in the maximum load magnitudes for each step function. Structural Analysis The EPU structural evaluations employed two models: (1) the ODB 2D lumped mass centerline beam model with updated fuel mass and stiffness properties, and (2) a 3D shell model which replaced the design basis axi-symmetric shell model. The ODB shell model was an axi-symmetric model capable of simulating non axi-symmetric loading by decomposing the loading terms and modal displacements into their Fourier components. This model has been replaced with a full 3D shell model with the same mass and stiffness properties. The EPU centerline beam model was identical to the ODB model, with the exception that the core was updated to represent GNF2 fuel. The centerline beam model was used to generate RPV and internals forces and moments as well as re sponse spectra at the interface locations for the reactor internals. Results from the centerline beam model are directly comparable to the design basis results. The 3D shell model was created using ANSYS (Reference 62). The model consisted of the same geometry, mass, and stiffness as the ODB model. The 3D shell model was used to calculate the Response Spectra on the RPV and BSW. The modeling assumptions between the design basis model and the new 3D shell model remain unchanged. The main difference is that the EPU model is a full 3D model rather than an axi-symmetric model; therefore, the loading terms and mode shapes do not need to be decomposed into their Fourier representations. Eigenvalues and vectors from the two models were compared and found to be within 1 Hz for the major structural frequencies. Analysis Results

The structural analysis results showed that the effects of changing to the more physics based methodology typically resulted in a reduction of the forces and moments. On average, the peak NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-271 forces and moments have been reduced by about half when comparing the CLTP case results calculated using the new methodology against the CLTP design basis forces and moments. Although, in general, the results showed a decrease in the peak forces and moments for most components, some components did experience more severe loading as a result of the change in methodology. The shroud, guide tubes, and CRD housings experienced loads of up to 1.27 times

the design basis loads. This increase in load ing was included in the EPU reactor internals evaluation in Sections 2.2.2.3 and 2.2.3.2. The results of that evaluation showed that the affected components had sufficient margin to accommodate the increase in loading. When using the more realistic physics based methodology, the change from CLTP to EPU operating conditions had a generally negligible effect on the peak force and moment results. The variation in component loading between the two cases was typically less than 2%. The maximum increase in loading observed was 6.5% for the CRD housing locations for the RSLB case. The EPU reactor internals evaluati on in Sections 2.2.2.3 and 2.2.3.2 shows that the affected components had sufficient margin to accommodate the loading. For most locations, the response spectra were not significantly affected by the change in operating conditions from CLTP to EPU. However, due to the new methodology and the increased number of considered power flow points, jet pump and piping attachment points were evaluated. The evaluations based on change in response spectra were included in the EPU evaluations in Sections 2.2.2.2, 2.2.2.3, and 2.2.3.2. The results of those evaluations showed that the affected components had sufficient margin to accommodate the loading. The peak DP loading on the BSW is not significantly affected by the EPU, which remains below the BSW design DP of 56 psid. The results of this analysis show that the containment SSCs important to safety will continue to be protected from the dynamic effects resulting from pipe breaks and that the subcompartments will continue to have sufficient margins to prevent fracture of the structure due to pressure differences across the walls following implementation of the proposed EPU. Drywell Head Region An evaluation was performed for pipe breaks in the DW head region that may result in a DP load across the DW head refueling bulkhead plate. The EPU evaluation confirms that the CLTR disposition for high-energy line breaks of steam lines is applicable to the two governing breaks (RPV head spray line and MSL) modeled in the licensing basis DW head region DP analysis. The current design basis DW head region DP analysis is unaffected by EPU because the steam dome pressure does not change. Therefore, the DW head refueling bulkhead plate design remains adequate. Therefore, the subcompartment analyses meet all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-272 Conclusion The change in predicted pressurization resulting from the increased mass and energy release following the proposed EPU has been reviewed. It was found that the containment SSCs important to safety will continue to be protected from the dynamic effects resulting from pipe breaks and that the subcompartments will continue to have sufficient margins to prevent fracture

of the structure due to pressure difference across the walls following implementation of the proposed EPU. Entergy concludes that the plant will continue to meet GDCs 4 and 50 for the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to subcompartment analyses. 2.6.3 Mass and Energy Release 2.6.3.1 Mass and Energy Release Analysis for Postulated Loss of Coolant Regulatory Evaluation The release of high-energy fluid into containment from pipe breaks could challenge the structural integrity of the containment, including subcompartments and systems within the containment. The review covered the energy s ources that are available for release to the containment and the mass and energy release rate calculations for the initial blowdown phase of the accident. The regulatory acceptance criteria for mass and energy release analyses for postulated LOCAs are based on: (1) GDC-50, insofar as it requires that sufficient conservatism be provided in the mass and energy release analysis to assure that containment design margin is maintained; and (2) 10 CFR 50, Appendix K, insofa r as it identifies sources of energy during a LOCA. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The GGNS mass and energy release for the pos tulated LOCA is described in UFSAR Section 6.2.1.3.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 4.1 of the CLTR addresses the effect of CPPU on Containment System Performance. The results of this evaluation are described below GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-273 Topic CLTR Disposition GGNS Result Drywell Temperature [[ Meets CLTR Disposition Drywell Pressure

     ]] Meets CLTR Disposition 2.6.3.1.1 Containment Temperature Response The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. The assumption of constant pressure minimizes the effect on other aspects of the containment evaluation. The DW design temperature (330°F) has been determined based on an analysis of the superheated gas temperature caused by a blowdown of steam to the DW during a small break LOCA. This analysis conservatively determined a combination of vessel pressure and DW pressure that produces a maximum calculated DW temperature. The UFSAR reported that expansion of reactor steam under these conditions will result in a calculated peak DW temperature of 330°F. These conditions, wh ich are described in Section 6.2.1.1.3.3.5.4 of the UFSAR, are derived independent of the initial reactor power. Therefore, EPU has no effect on the peak DW temperature and all CLTR dispositions are met. The WW gas space peak temperature response was calculated using a mechanistic model for heat and mass transfer from the SP to the WW gas space. Table 2.6-1 shows the calculated WW gas space peak temperature of 142°F for the DBA LOCA at EPU. The WW gas space peak temperature for the ASDC event was calculated to be 154°F at EPU. The WW gas space peak temperatures are less than the WW design temperature of 185°F.

2.6.3.1.2 Short-Term Containment Pressure Response The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. The assumption of constant pressure minimizes the effect on other aspects of the containment evaluation. Short-term containment response analyses were performed for the limiting DBA LOCA that assumes a double-ended gu illotine break of a recirculation suction line or a MSL to demonstrate that EPU does not result in exceeding the containment design limits. The short-term analysis covers the blowdown period during which the maximum DW pressure, WW pressure, and DP between the DW and WW occur. These analyses were performed at 102% of EPU RTP level. For GGNS, the MSLB remained limiting for peak pressure at EPU. The time-dependent results of the limiting short-term analyses are presented in Figures 2.6-4 and 2.6-5 and are summarized in Table 2.6-1. Table 2.6-1 also includes comparisons of the pressure

values calculated for EPU to the design pressures and to pressure values from previous calculations based on the current power. The maximum calculated DW and containment NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-274 pressures for EPU remain within their design values, and thus, are acceptable. The peak calculated DW-to-WW pressure also remains with in its design value and all CLTR dispositions are met. Conclusion The mass and energy release has been reviewed a nd found to adequately address the effects of the proposed EPU and appropriately accounts for the sources of energy identified in 10 CFR 50, Appendix K. Based on this, Entergy finds that the mass and energy release analysis meets the requirements in GDC-50 for ensuring that the analys is is conservative. Therefore, Entergy finds the proposed EPU acceptable with respect to mass and energy release for a postulated LOCA.

2.6.4 Combustible

Gas Control in Containment Regulatory Evaluation Following a LOCA, hydrogen and oxygen may accumulate inside the containment due to chemical reactions between the fuel rod cladding and steam, corrosion of aluminum and other materials, and radiolytic decomposition of water. If excessive hydrogen is generated, it may form a combustible mixture in the containment atmosphere. The review covered: (1) the production and accumulation of combustible ga ses; (2) the capability to prevent high concentrations of combustible gases in local areas; (3) the capability to monitor combustible gas concentrations; and (4) the capability to reduce combustible gas concentrations. The review primarily focused on any effect that the proposed EPU may have on hydrogen release assumptions, and how increases in hydrogen release are mitigated. The regulatory acceptance criteria for combustible gas control in containment are based on: (1) 10 CFR 50.44, insofar as it requires that plants be provided with the capability for controlling combustible gas concentrations in the containment atmosphere; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; (3) GDC-41, insofar as it requires that systems be provided to cont rol the concentration of hydrogen or oxygen that may be released into the reactor containmen t following postulated accidents to ensure that containment integrity is maintained; (4) GDC-42, insofar as it requires that systems required by GDC-41 be designed to permit appropriate peri odic inspection; and (5) GDC-43, insofar as it requires that systems required by GDC-41 be designed to permit appropriate periodic testing. Additional requirements based on 10 CFR 50.44 for control of combustible gas apply to plants with a Mark III type of containment that do not rely on an inerted atmosphere to control hydrogen inside the containment. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-275 may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 is not applicable to GGNS; see UFSAR Section 3.1.2.1.5. The GGNS containment CGCS is described in UFSAR Section 6.2.5.

Technical Evaluation The revised 10 CFR 50.44 (68FR54123, dated September 16, 2003) does not define a design basis LOCA hydrogen release and eliminates the requirements for hydrogen control systems to mitigate such releases. GGNS License Amendment Number 166, issued on June 16, 2004 (Reference 63), eliminated the requirements associated with hydrogen recombiners and hydrogen monitors. The performance of the CGCS and th e hydrogen analyzers is not adversely affected by EPU. Conclusion The containment CGCS was reviewed and it was found that the effects of the proposed EPU have been adequately addressed. The system will continue to have sufficient capability following the implementation of the proposed EPU. Entergy concludes that the containment CGCS will continue to meet the requirements of GDCs 5, 41, 42, and 43, as well as 10 CFR 50.44. Therefore, Entergy finds the proposed EPU acceptable with respect to combustible gas control in containment.

2.6.5 Containment

Heat Removal Regulatory Evaluation Fan cooler systems, spray systems, and RHR systems are provided to remove heat from the containment atmosphere and from the water in the containment WW. The review in this area focused on: (1) the effects of the proposed EPU on the analyses of the available NPSH to the containment heat removal system pumps; and (2) the analyses of the heat removal capabilities of the spray water system and the fan cooler HX. The regulatory acceptance criteria for containment heat removal are based on GDC-38, insofar as it requires that a containment heat removal system be provided, and that its function shall be to rapidly reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels.

GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The containment heat removal systems are described in UFSAR Section 6.2.2.

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2-276 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Sections 4.1 and 4.2 of the CLTR address the effect of CPPU on Containment Heat Removal. The results of this evaluation are described below: GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Pool Temperature Response [[ Meets CLTR Disposition ECCS Net Positive Suction Head

     ]] Meets CLTR Disposition 2.6.5.1 Pool Temperature Response The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. The long-term bulk pool temperature response for EPU is evaluated for the limiting DBA LOCA in Section 6.2 of the UFSAR and the limiting Alternate Shutdown activity in Section 15.2.9 of the UFSAR.

The current design limit for the bulk SP temperature is 185°F. For EPU implementation, this design limit is increased to 210°F. The analysis of the DBA LOCA was performed at 102% of EPU RTP. The calculated time-dependent SP and WW temperature response is presented in Figure 2.6-1 and the peak values for LOCA bulk pool temperature for the current AOR and the EPU RTP case are compared in Table 2.6-1. The EPU analyses were performed using a decay heat table applicable to GGNS and based on ANS/ANSI 5.1-1979 with 2-sigma adders. The analysis assumed the single failure of one of the two RHR HXs. The resulting calculated peak bulk SP temperature for the bounding DBA LOCA is 189F. This temperature is within the ECCS NPSH pump limit of 194 F. The highest bulk pool temperature response from a non-LOCA event results from an ASDC event. This event was also analyzed at 102% of EPU RTP and ANS/ANSI 5.1-1979 with 2-sigma adders decay heat. The limiting alternate shutdown activity assumes reactor isolation with availability of one RHR HX. The resulting time-dependent SP and WW temperature response is presented in Figures 2.6-2 and 2.6-3 and the peak bulk pool temperature at 102% of

EPU RTP is 198F, which also is within the design limit of 210 F. Therefore, all pool NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-277 temperatures have been found to be within design limits and the pool temperature response meets all CLTR dispositions. 2.6.5.2 ECCS Net Positive Suction Head The CLTR states that EPU RTP operation increases the reactor decay heat, which increases the heat addition to the SP following LOCA, SBO, and Appendix R events. Following a LOCA, the RHR, LPCS and HPCS pumps operate to provide the required core and containment cooling. Adequate NPSH margin (NPSH available (NPSH A) minus NPSH required (NPSH R)) is required during this period to assure the essential pump operation. The NPSH margins for the ECCS pumps were evaluated for the limiting conditions following a DBA LOCA. The limiting NPSH conditions depend on the pump flow rates, debris lo ading on the suction strainers, pipe frictional losses, SP level, and SP temperature. No changes to any of these parameters result from the implementation of EPU. Existing calculations for ECCS NPSH are consistent with RG 1.82 (Reference 64). The NPSH margins were calculated assuming system runout flow rates that exceed ECCS pump operational requirements for a DBA. The pump flow rates are 8,940 gpm flow rate per RHR pump, 9,100 gpm LPCS pump flow rate, and 8,175 gpm HPCS pump flow rate. EPU does not change the types or the quantities of material present in, or transferred to, the SP. Because ECCS system flows remain unchanged under EPU conditions, the assumptions for friction loss through piping and fittings and pump NPSH R are unchanged from pre-EPU calculations. The elevation head is based on the TS minimum SP level corresponding to an elevation of 107.5 feet, which is adjusted for the minimum drawdown level consistent with RG 1.82 requirements. EPU RTP operation increases the reactor decay heat , which increases the heat addition to the SP following a LOCA; however, the analyzed SP water temperature and containment pressure remain essentially unchanged because of original analysis conservatisms as described in Section 2.6.1.1.1. The SP temperature continues to be within the design limit of 194°F for debris generating events and 210°F for non-debris generating events used in the NPSH calculations. Adequate NPSH is demonstrated, and no credit for an increase in containment pressure is needed during a design or licensing basis event. The methodology used by GGNS to determine the amount of debris generated and transported to the strainers is based on NEDO-32686-A, the BWR Owners' Group (BWROG) Utility Resolution Guidance for ECCS Suction Strainer Blockage (Reference 65). The assumptions NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-278 used for debris generation from protective coatings are in accordance with Section 3.2.2.2.2.1.1 of Reference 65. The amount of debris generation is not affected by EPU. The GGNS suction strainer design was tested in th e 1/4-Scale Test Facility at the factory Mutual Research Corporation Test Center. This test facility provides a three-dimensional model of the Mark III containment pressure SP and was originally constructed to conduct large scale hydrogen combustion tests for the Mark III H ydrogen Control Owners Group. Additionally, small-scale modeling and conservative performance testing was performed in a Small-Scale Test Apparatus. These tests modeled conservative amounts of fiber, iron oxide, epoxy paint, and other potential LOCA-generated debris. The larg e toroidal passive ECCS/RCIC suction strainer at GGNS results in a low approach velocity fo r water entering the strainer. The strainer approach velocity is unchanged because the ECCS and RCIC systems are not modified by EPU. The debris generation is not affected becau se the reactor pressure is unchanged and no modifications are made to the current insulation c onfigurations. Therefore, the suction strainer performance testing remains applicable for EPU. The above results support the above conclusion that the debris loading on the suction strainers and the methodology used to calculate available ECCS NPSH for EPU are the same as the pre-EPU conditions. Therefore, all CLTR dispositions are met for ECCS NPSH at EPU.

Conclusion The containment heat removal systems were revi ewed and it was found that the effects of the proposed EPU have been adequately addressed. The systems will continue to be able to rapidly reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels. Entergy concludes that the containment heat removal systems will continue to meet the requirements of GDC-38. Therefore, Entergy finds the proposed EPU acceptable with respect to containment heat removal systems.

2.6.6 Secondary

Containment Functional Design Regulatory Evaluation The secondary containment structure and supporting systems of dual containment plants are provided to collect and process radioactive material that may leak from the primary containment following an accident. The supporting systems maintain a negative pressure within the secondary containment and process this leakage. The review covered: (1) analyses of the pressure and temperature response of the secondary containment following accidents within the primary and secondary containments; (2) analyses of the effects of ope nings in the secondary containment on the capability of the depressurization and filtration system to establish a negative pressure in a prescribed time; (3) analyses of any primary containment leakage paths that bypass the secondary containment; (4) analyses of the pressure response of the secondary containment resulting from inadvertent depressurization of the primary containment when there is vacuum NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-279 relief from the secondary containment; and (5) the acceptability of the mass and energy release data used in the analysis. The review primar ily focused on the effects that the proposed EPU may have on the pressure and temperature response and drawdown time of the secondary containment, and the effect this may have on off-site dose. The regulatory acceptance criteria for secondary containment functional design are ba sed on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and be protected from dynamic effects (e.g., the effects of missiles, pipe whipping, and discharging fluids) that may result from equipment failure s; and (2) GDC-16, insofar as it requires that reactor containment and associated systems be pr ovided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The functional design of secondary containment is described in UFSAR Section 6.2.3.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 4.5 of the CLTR addresses the effect of CPPU on the SGTS. The SGTS is designed to maintain secondary containment at a negative pressure and to filter the exhaust air for removal of fission products potentially present during abnormal conditions. By limiting the release of airborne particulates and halogens, the SGTS limits off-site dose following a postulated DBA. The SGTS fission product control and removal function evaluation is described in Section 2.5.2.1. Generic bounding analyses have been performed with results

located in Section 4.5 of the CLTR. The results of this evaluation are given below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Flow Capacity [[

     ]] Meets CLTR Disposition The CLTR states that the core inventory of iodine and subsequent loading on the SGTS filter or charcoal adsorbers are affected by EPU.

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2-280 The design flow capacity of the SGTS was selected to maintain the secondary containment at the required negative pressure to minimize the potential for exfiltration of air from the reactor building. [[

     ]]  The GGNS HEPA filters are satisfactory for EPU operation. The secondary containment structure, openings, pathways, and drawdown time are unaffected by EPU. Because the maximum dome pressure is also not changed for EPU, there is no effect to the ability of secondary containment to contain mass a nd energy released to it. There is no increase in mass and energy released to secondary containment for EPU. The secondary containment temperature and pressure are not evaluated furthe r in the CLTR because th ere is no effect as a result of EPU. Therefore, the evaluation of the SGTS ability to maintain secondary containment at a negative pressure and contain radionuclides is adequate for this topic. Therefore, the flow capacity meets all CLTR dispositions.

Conclusion A review of the secondary containment pressure and temperature transient has confirmed the ability of the secondary containment to provide an essentially leak-tight barrier against uncontrolled release of radioactivity to the environment. The review determined: (1) there is no significant increase of mass and energy that would result from the implementation of the proposed EPU; and (2) the secondary containment will continue to provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment. Entergy concludes that the secondary containment and associated systems will continue to meet the requirements of GDCs 4 and 16. Therefore, En tergy finds the proposed EPU acceptable with respect to secondary containment functional design.

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2-281 Table 2.6-1 GGNS Containment Performance Results Parameter DBA LOCA CLTP from AOR DBA LOCA CLTP -with EPU Model 1 DBA LOCA EPU -with EPU Model Limit Peak DW Pressure (psig) 22.0 3 26.6 3 26.7 3 30 Peak DW Temperature (°F) 330 3 330 3 330 3 330 Peak Bulk Pool Temperature (°F) 181 184 5 189 5 210 7 Peak Containment Pressure (psig) 2 NR / 11.5 14.7 / 11.3 14.8 / 11.9 8 15 Peak Containment Temperature (°F) 181 138 6 142 6 185 Peak DW to Containment

delta-P (psi) 22.0 2 24.2 4 24.2 4 30 Notes: 1. Containment analyses performed for the EPU uses methods that are similar to the methods used for the AOR at CLTP. The analysis at CLTP with the EPU model uses the plant inputs defined for the EPU as discussed in Section 2.6.1.1.1. 2. Peak containment pressures are given for the short-term / long-term analysis. The limiting event is the steam line break. The short-term peak pressure is increased over the C LTP result as reflected in UFSAR Figure 6.2-10 due primarily to an assumption of more conservative initial conditions. The short-term peak containment pressure is only applicable to the portion of the containment below the Hydraulic Control Unit (HCU) floor, while the long-term peak containment pressure is applicable to the entire containment. 3. Most limiting values obtained from short-term MSLB analysis. 4. Most limiting values obtained from short-term RSLB analysis.

5. Peak temperature values shown are from long-term DBA LOCA analysis. Peak pool temperatures of 191°F and 198°F were calculated for the ASDC event for CLTP and EPU, respectively. 6. The peak containment temperatures reported here are from the long-term LOCA case developed to maximize containment pressures and pool temperature. Containment temperatures are reduced with the EPU model because the original methodology assumed thermal equilibrium between the SP and the containment airspace.

The EPU model uses a mechanistic model to calculate the mass and heat transf er between the SP and the containment airspace. 7. The current design limit for the bulk SP temperature is 185°F. For EPU implementation, this design limit has been increased to 210°F. 8. Value exceeds the current Appendi x J testing criteria of 11.5 psig. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-282 Table 2.6-2 Long-Term Containment Response Key Analysis Input Values No Parameter Unit Analysis Value 1. Reactor

a. Initial power level

1. 102% current rated power MWt 3,976 2. 102% uprated power MWt 4,496 b. Initial vessel dome pressure

1. At 102% current rated power psia 1066 2. At 102% uprated power psia 1066 c. Decay heat model

1. Short-term DBA LOCA ANS 5-1971 + 20% 2. Long-term ANS 5.1-1979 + 2 d. Vessel volumes

1. Total vessel free volume ft 3 21,182 2. Liquid vessel volume Subcooled ft 3 8,294 Saturated ft 3 3,896 e. Vessel related masses

1. Liquid mass per recirculation loop lbm 26,796 2. Liquid mass in the HPCS piping between the RPV nozzle and first normally closed valve lbm 4,931 3. Liquid mass in the RCIC piping between the RPV nozzle and first normally closed valve lbm 0 4. Liquid mass in the LPCI piping between the RPV and the first normally closed valve lbm 12,540 5. Liquid mass in the RHR piping between the RPV nozzle and the first normally closed valve lbm 32,487 6. Liquid mass in the CS piping between the RPV nozzle and the first normally closed valve lbm 5,644 7. Steam mass in the MSL to the first isolation valve. One inner steam line One outer steam line lbm 487 650 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-283 No Parameter Unit Analysis Value f. Time at which MSIVs start to close / fully closed RSLB MSLB sec 0.0 / 3.0 0.5 / 5.5 2. DW/Vent System

a. Total DW free volume (including vent system) ft 3 270,000 b. Initial DW pressure (range) psig -0.40 to 3.50 c. Initial DW temperature (range)

°F 65 to 140 d. Initial DW RH (range) % 20 to 90 e. Vent System

1. Number of horizontal vents per row 45 2. Diameter of each horizontal vent ft 2.33 3. DW weir annulus pool volume, including horizontal vents High water level (HWL)

Low water level (LWL) ft 3 13,303 13,041 4. Inside weir wall diameter ft 65.00 5. Outside weir wall diameter ft 68.00 6. Weir annulus surface area ft 2 533.70 7. Height of weir wall from SP bottom ft 24.31 8. Length of each horizontal vent ft 5.00 9. Centerline elevation of top-row vents from pool bottom ft 11.33 10. Centerline elevation of mid-row vents from pool bottom ft 7.17 11. Centerline elevation of bottom vents from pool bottom ft 3.00 12. Loss coefficient for vent system (including entrance and exit losses) 3.50 f. DW Pool

1. DW Pool Holdup Volume ft 3 49,723 2. DW Pool Surface Area ft 2 2,480 3. WW / SP a. Initial SP volume (including water in vents)

1. HWL ft 3 137,200 2. LWL ft 3 133,750 b. Initial SP temperature (max)

°F 95 c. Initial WW airspace volume

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2-284 No Parameter Unit Analysis Value 1. HWL ft 3 136,786 2. LWL ft 3 139,933 d. Initial WW/containment airspace pr essure (range) psig -0.10 to 1.50 e. Initial WW/containment airspace temperature (range) °F 40 to 95 f. Initial WW/containment airspace RH (range) % 20 to 100 g. Initial upper pool water volume ft 3 36,163 h. Initial upper pool water temperature °F 125 i. DW pressure above which upper pool dump may be initiated psig 2.0 j. Upper pool dump time delay after LOCA sec 1,800 k. Upper pool surface area ft 2 3,383 4. LPCI/LPCS/HPCS

a. LPCI runout flow rate gpm 6,600 b. LPCI pump shutoff head psid 210.8 c. LPCI pump horsepower hp 1000 d. LPCS runout flow rate gpm 7,000 e. LPCS pump shutoff head psid 280.9 f. LPCS pump horsepower hp 2,000 g. HPCS runout flow rate gpm 8,175 h. HPCS pump shutoff head psid 1,177 i. HPCS pump horsepower hp 3,500 5. RHR a. Heat exchanger K-value Pool Cooling Mode Containment Spray Mode LPCI Cooling Mode Normal SDC Mode BTU/sec-°F/HX 540.00 453.75 540.00 557.72 b. Service water temperature

°F 90 c. DW spray flow rate (1 RHR pump) gpm 0 d. WW spray flow rate (1 RHR pump) gpm 5,085 e. RHR flow rate in pool cooling mode (1 RHR pump) gpm 6,600 6. WW-to-DW Vacuum Breakers (VBs)

a. Pressure difference WW-to-DW for VBs to be fully open psid 1.0 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-285 No Parameter Unit Analysis Value b. Number of VB assemblies (multiple valve systems) 2 c. Flow area of each VB assembly for loss coefficient ft 2 0.548 d. Total loss coefficient of each VB assembly 6.000 7. Containment Heat Sink Data

a. DW steel mass lbm 159,691 b. DW steel surface area ft 2 15,670 c. DW concrete mass lbm 15,855,390 d. DW concrete surface area ft 2 14,620 e. WW airspace steel mass lbm 1,015,231 f. WW airspace steel surface area ft 2 96,010 g. WW airspace concrete mass lbm 28,910,655 h. WW airspace concrete surface area ft 2 52,190 8. FW Input - Long-Term Analysis 1a. Time sec 0.0 1b. Flow rate lbm/sec 4,948 1c. Energy rate MBTU/sec 4.1004 2a. Time sec 27.0 2b. Flow rate lbm/sec 4,948 2c. Energy rate MBTU/sec 2.5775 3a. Time sec 30.0 3b. Flow rate lbm/sec 4,948 3c. Energy rate MBTU/sec 2.1593 4a. Time sec 40.0 4b. Flow rate lbm/sec 4,505 4c. Energy rate MBTU/sec 2.0675 5a. Time sec 58.7 5b. Flow rate lbm/sec 4,505 5c. Energy rate MBTU/sec 0.7474 6a. Time sec 202.0 6b. Flow rate lbm/sec 4,505 6c. Energy rate MBTU/sec 0.7474

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2-286 Table 2.6-3 GGNS Peak SP Temperatures for Postulated ATWS, SBO, and 10 CFR 50 Appendix R Events Event Peak Suppression Pool Temperature Limiting ATWS (MSIVC) 165.3ºF Station Blackout 200.1ºF Appendix R Fire 181.4ºF NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-287 Table 2.6-4 Overview of AP Load Evaluation Methodologies Method Parameter Current Design Basis EPU Break Flow Method Model Type NEDO-24548 TRACG AP Method Model Type NEDO-24548 TRACG Dimensionality 1D (Lumped Parameter) 2D (Z, Theta) Model Physics HEM Non-Equilibrium Two-Phase Flow Model Annulus Model Cells 25 384 JI, JR, PWR RSLB FWLB NEDO-24548 Revised Method RDLB NEDO-24548 TRACG JI JR Inconsistencies Yes No PWR Method Peak Value Integrated Energy Structural Method Model Type: Structural and Internals 2D Beam 2D Beam Model Type: Attachment (Piping, Spargers) Asymmetrically Loaded Axi-Symmetric Shell 3D Finite Element Shell JI JR PWR Peak Value Actual Time History Steam Line Vectors PWR Load Reserved (Typically Conservative) Correct NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-288 Figure 2.6-1 Long-Term DBA LOCA Temperature Response at EPU GGNS DBA LOCA RSLB CASE AT EPU POWER - RSLB31 0 50100150200 250300012345678Time (hours)Temperature (°F)TspLong-Term Peak SP Temperature is 189°F @ 3.07 hours4125816702/10/1010:27TwwLong-Term Peak Wetwell Temperature is 142°F @ 6.13 hours

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2-289 Figure 2.6-2 Long-Term ASDC SP and WW Temperature Response at EPU GGNS ASDC AT EPU POWER - ASDC10 0 50100150 200 2500369121518212427303336Time (hours)Temperature (°F)TspLong-Term Peak SP Temperature is 198°F @ 3.83 hours3A451D5703/05/1014:44TwwLong-Term Peak Wetwell Temperature is 154°F @ 7.21 hours

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2-290 Figure 2.6-3 Long-Term ASDC DW and WW Temperature Response at EPU GGNS ASDC AT EPU POWER - ASDC10 0 501001502002500369121518212427303336Time (hours)Temperature (°F)TdwLong-Term Peak Drywell Temperature is 139°F @ 7.21 hours3A451D5703/05/10 14:44TwwLong-Term Peak Wetwell Temperature is 154°F @ 7.21 hours

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2-291 Figure 2.6-4 Short-Term DBA LOCA MSLB Pressure Response at EPU GGNS EPUMPMSLB90 - 102P/100F - Rated ( D ) 0 5 10 15 20 25 30051015202530Time (seconds)Pressure (psig)414B519B 12/17/09 17:57:0 2Peak Drywell Pressure of 26.7 psig occurs at 3.13 seconds. Peak Wetwell Pressure of 14.8 psig occurs at 3.00 seconds.Bubble Pressure at vent clearing is 20.1 psid at 0.86 seconds.DrywellWetwellContainment

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2-292 Figure 2.6-5 Short-Term DBA LOCA MSLB Diff erential Pressure Response at EPU GGNS EPUMPMSLB90 - 102P/100F - Rated ( D ) 0 5 10 15 20 25 30051015202530Time (seconds)Differential Pressure [DW-Cont] (psid)414B519B 12/17/09 17:57:0 2Peak Pdw-Pcont Pressure of 23.7 psid occurs at 1.74 seconds.

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2-293 2.7 Habitability, Filtration, and Ventilation

2.7.1 Control

Room Habitability System Regulatory Evaluation Entergy reviewed the control room habitability system and control building layout and structures to ensure that plant operators are adequately protected from the effects of accidental releases of toxic and radioactive gases. A further objective of the review was to ensure that the control room can be maintained as the backup center from which technical support center (TSC) personnel can safely operate in the case of an acci dent. The review focused on the effects of the proposed EPU on radiation doses, toxic gas concentrations, and estimates of dispersion of airborne contamination. The regulatory acceptance criteria for the control room habitability system are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with postulated accidents, including the effects of the release of toxic gases; and (2) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dos e equivalent (TEDE) for the duration of the accident. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The control room habitability system is described in UFSAR Section 6.4.1, "Habitability Systems." Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 4.4 of the CLTR addresses the effect of CPPU on Main Control Room Atmosphere Control System. The control room HVAC system standby fresh ai r units functions during a DBA or an AOO to provide filtered air for personnel ventilation and pressurization of the control room envelope. Redundant radiation detectors are provided at the outside air intakes to automatically recirculate the air in the control room or outside supply air through the standby fresh air units. With no change to the detection and controls, the operation of the control room HVAC system is not affected. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-294 GGNS meets all CLTR dispositions. The topic addressed in this evaluation is: Topic CLTR Disposition GGNS Result Iodine Intake [[

     ]] Meets CLTR Disposition The CLTR states that EPU increases the radioisotopes seen by the control room atmosphere control system following an accident.

The radiological effect of EPU on the CREF system (which includes HEPA filtration capability only) is due to an increase in the particulates, including particulate iodines, released during an accident. GGNS has implemented the AST met hodology which affects the DBA iodine release model (Reference 66). The EPU analyses were performed for 102% of the EPU power level (i.e., 4,496 MWt), and thus incorporate the increased EPU iodine release as well as the effects of the AST iodine release model. These analyses included the radiological consequences of the DBAs documented in Chapter 15 of the UFSAR that potentially result in the most significant control room exposures. In all cases, the control room doses were within regulatory limits. The quantities and locations of gases and hazardous chemicals that could affect the control room are unaffected by EPU. Therefore, EPU has no effect on the design basis potential toxic gas concentrations that are mitigated by the control room habitability system, and the current analyses remain bounding. Therefore, the iodine intake meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the ability of the control room habitability system to protect plant operators against the effects of accidental releases of toxic and radioactive gases have been reviewed. Entergy concludes the increase of toxic and radioactive gases that would result from the proposed EPU has been adequately evaluated. Entergy further concludes the control room habitability system will continue to provide the required protection following implementation of the proposed EPU. Based on this, Entergy concludes the control room habitability system will continue to meet the requirements of GDCs 4 and 19. Therefore, Entergy finds the proposed EPU acceptable with respect to the control room habitability system.

2.7.2 Engineered

Safety Feature Atmosphere Cleanup Regulatory Evaluation ESF atmosphere cleanup systems are designed for fission product removal in post-accident environments. These systems generally include primary systems (e.g., in-containment NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-295 recirculation) and secondary systems (e.g., SGTSs) for the fuel-handling building, control room, shield building, and areas containing ESF component

s. The review focused on the effects of the proposed EPU on system functional design, environmental design, and provisions to preclude temperatures in the adsorber section from exceeding design limits. The regulatory acceptance criteria for ESF atmosphere cleanup systems are based on: (1) GDC-19, insofar as it requires adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE (GGNS has implemented the alternative source term), for the duration of the accident; (2) GDC-41, insofar as it requires systems to cont rol fission products released into the reactor containment be provided to reduce the concentra tion and quality of fission products released to the environment following postulated accidents; (3) GDC-61, insofar as it requires systems that may contain radioactivity be designed to assure adequate safety under normal and postulated accident conditions; and (4) GDC-64, insofar as it requires means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including AOOs, and postulated accidents.

GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. At GGNS, the function of the ESF atmospheric cleanup systems is performed by the SGTS and the Control Room HVAC System. These system s are described in UFSAR Sections 6.5.3 and 9.4.1, respectively.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 4.5 of the CLTR a ddresses the effect of CPPU on the SGTS. The results of this evaluation are described below. One of the two ESF atmosphere cleanup systems at GGNS is the Control Room Emergency Filtration System (CREFS). The acceptability of this system under EPU conditions is addressed in Section 2.7.1. The second ESF atmosphere cleanup system is the SGTS. The SGTS is designed to maintain secondary containment at a negative pressure and to filter the exhaust air for removal of fission products potentially present during abnormal conditions. By limiting the release of airborne particulates and halogens, the SGTS limits off-site dose following a postulated DBA. The effect of a CPPU on the performa nce of the SGTS was evaluated in the CLTR based on two bounding analys es. CLTR dispositions regarding the flow NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-296 capacity and iodine removal capability of the SG TS have been addressed in Sections 2.6.6 and 2.5.2.1, respectively. Details regarding the SGTS evaluation based on post-LOCA operation after EPU implementation are described in Section 2.5.2. The SGTS at GGNS meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the SGTS have been reviewed and it was found the system design has adequately accounted for the increase of fission products and changes in expected environmental conditions that would result from the proposed EPU. Entergy concludes the SGTS will continue to provide adequate fission product removal in post accident environments following implementation of the proposed EPU. Based on this, Entergy concludes the SGTS will continue to meet the requirements of GDC s 19, 41, 61, and 64. Therefore, Entergy finds the proposed EPU acceptable with respect to the SGTS.

2.7.3 Control

Room Area Ventilation System Regulatory Evaluation The function of the control room area ventilation system (CRAVS) is to provide a controlled environment for the comfort and safety of control room personnel and to support the operability of control room components during normal ope ration, AOOs, and DBA conditions. The review of the CRAVS focused on the effects that the proposed EPU will have on the functional performance of safety-related portions of the system. The review included: (1) the effects of radiation, combustion, and other toxic products; and (2) the expected environmental conditions in areas served by the CRAVS. The regulatory acceptance criteria for the CRAVS are based on: (1) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate the effects of, and to be compatible with, the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (2) GDC-19, insofar as it requires

adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE (GGNS has implemented the alternative source term) for the duration of the accident; and (3) GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-297 At GGNS, the function of the CRAVS is performed by the control room HVAC system. The control room habitability system is described in UFSAR Section 6.4.1. The control room HVAC system is described in UFSAR Section 9.4.1. Technical Evaluation The HVAC systems discussed in the CLTR are only t hose that have power-dependent heat loads. Power-dependent HVAC systems require [[

     ]]  The control room HVAC system maintains temperature and humidity conditions suitable for personnel comfort and for equipment reliable operation inside the control room envelope. The control room HVAC system also maintains the control room envelope at positive pressure to inhibit air infiltration (see Section 2.7.1). Heat loads for the control room area envelope include boundary transmission, lighting, and equipment such as control room panels. These heat loads are not affected by the slightly higher process temperatures that may result from EPU; thus, they are not power-dependent. EPU does not add any electrical or electronic equipment to the control room. EPU may add some amperage for control and i ndication signals, but the resulting changes in temperatures are considered negligible. The rooms that are adjacent to the control room contain ventilation equipment or electrical and electronic equipment. These rooms are not affected by EPU, and therefore, there are no temperature changes in these areas that can affect main control room temperatures.

There is no increase in toxic gases release that may result from EPU. The control of the concentration of airborne radioactive material in the control room envelope during AOOs and after postulated accidents is accomplished by the control room HVAC system described in

Section 2.7.1. There is no change to the control room HVAC system configuration or system parameters as a result of EPU. Conclusion The effects of the proposed EPU on the ability of the CRAVS to provide a controlled environment for the comfort and safety of control room personnel and to support operability of control room components have been reviewed. Entergy concludes that the review has

adequately accounted for the increase of toxic and radioactive gases that would result from a DBA under the conditions of the proposed EPU, and associated changes to parameters affecting environmental conditions for control room personnel and equipment. Accordingly, Entergy

concludes the CRAVS will continue to provide an acceptable control room environment for safe operation of the plant following implementation of the proposed EPU. Entergy also concludes the system will continue to suitably control the release of gaseous radioactive effluents to the environment. Based on this, Entergy concludes the CRAVS will continue to meet the requirements of GDCs 4, 19, and 60. Therefore, Entergy finds the proposed EPU acceptable with respect to the CRAVS.

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2-298 2.7.4 Spent Fuel Pool Area Ventilation System Regulatory Evaluation The function of the spent fuel pool area ventilation system (SFPAVS) is to maintain ventilation in the SFP equipment areas, permit personnel access, and control airborne radioactivity in the area during normal operation, AOOs, and following postulated fuel handling accidents (FHAs). The review focused on the effects of the proposed EPU on the functional performance of the safety-related portions of the system. The regulatory acceptance criteria for the SFPAVS are based on: (1) GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents; and (2) GDC-61, insofar as it requires systems which contain radioactivity be designed with appropriate confinement and containment. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. At GGNS, the normal ventilation function of the SFPAVS is performed by the Auxiliary Building ventilation system (ABVS) while the SGTS performs the radionuclide control function. The ABVS and the SGTS are described in UFSAR Sections 9.4.6 and 6.5.3, respectively.

Technical Evaluation The fuel handling area ventilation system at GGNS is a sub-system of the ABVS. The function of the fuel handling area ventilation system is to provide ventilation for and maintain the fuel handling area at a slightly negative pressure with respect to surrounding areas during normal operation to ensure that airborne radiation is collected by the system. The system includes fuel pool sweep and exhaust fans to provide circul ation across the pool area during fuel handling operations. This system (fuel handling area ventilation system) is further described in UFSAR Section 9.4.2. The fuel handling area ventilation system assures that no ambient air escapes from the fuel-handling area during fuel handling operations without first being monitored for airborne radioactivity. Upon detection of high radioac tivity, the SGTS is in itiated. The SGTS will maintain a 1/4" water gauge negative pressure in the area and provide filtration of the exhaust air through ESF filters. As noted above, the ABVS provides normal ventila tion to the SFP area; other functions of this system are described in Section 2.7.5. When required, the SGTS functions to control radionuclide inventory in the SFP area, and its EPU evaluation is described in Sections 2.5.2.1 and 2.6.6. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-299 Conclusion The effects of the proposed EPU on the SFPAVS ha ve been reviewed. Entergy concludes the review has adequately accounted for the effects of the proposed EPU on the system's capability to: (1) maintain ventilation in the SFP equipment areas; (2) permit personnel access; (3) control

airborne radioactivity in the area; (4) control release of gaseous radioactive effluents to the environment; and (5) provide appropriate containment. Based on this, Entergy concludes the SFPAVS will continue to meet the requirements of GDCs 60 and 61. Ther efore, Entergy finds the proposed EPU acceptable with respect to the SFPAVS.

2.7.5 Auxiliary

and Radwaste Area and Turbine Areas Ventilation Systems Regulatory Evaluation The function of the auxiliary and radwaste area ventilation system (ARAVS) and the turbine area ventilation system (TAVS) is to maintain ventilation in the auxiliary and radwaste equipment and turbine areas, permit personnel access, and contro l the concentration of airborne radioactive material in these areas during normal operation, during AOOs, and after postulated accidents. The review focused on the effects of the proposed EPU on the functional performance of the safety-related portions of these systems. Th e regulatory acceptance criteria for the ARAVS and TAVS are based on GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. At GGNS, the functions of the ARAVS and TAVS are performed by the ABVS, the SGTS, the Radwaste Building Ventilation System, and Turbine Building Ventilation System. Neither the Radwaste Building Ventilation System nor the Turbine Building Ventilation System has a safety-related function. The Turbine Building Ventilation System is equipped with filter trains; however, no credit is taken for their operation. The SGTS provides control of radionuclide concentration in the Auxiliary Building. The ABVS, the SGTS, Radwaste Building Ventilation System, and Turbine Building Ventilation System are described in UFSAR Sections 9.4.6, 6.5.3, 9.4.3, and 9.4.4, respectively. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-300 effects of CPPUs. Section 6.6 of the CLTR addresses the effect of CPPU on CLTR power-dependent HVAC. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Power-Dependent HVAC Performance [[

     ]]Meets CLTR Disposition The CLTR states that EPU results in slightly higher process temperatures and electrical loads on the HVAC system.

The ARAVS and the TAVS evaluated in the CLTR are only those that are power-dependent. The power-dependent HVAC systems consist mainly of heating, cooling supply, exhaust, and recirculation units in the Turbine Building, Auxiliary Building, Radwaste Building, and the DW. These systems do not function to control the concentration of airborne radioactive material in these areas during normal operation, during AOOs, and after postulated accidents. The control of the concentration of airborne radioactive material in the Auxiliary Building during normal operation, during AOOs, and after postulated accidents is accomplished using the SGTS

described in Sections 2.5.2 and 2.6.6. Monitoring of the Radwaste Building exhaust, and the Turbine Building exhaust, including the exhaust from the Turbine Gland Sealing System and the Mechanical Vacuum Pump System, is not affected by EPU. At GGNS, the normal operating EPU process temperatures affecting the normal HVAC loads are bounded by the CLTP analysis with the exception of FW. Although the FW temperature increases in some portions of the system, the temperature used in the existing analysis to determine HVAC heat loads bounds the EPU FW temperature. There are no modifications affecting pump motors that result in increased heat loads. Therefore, the ventilation in these areas is unaffected by EPU. No modifications are planned to any HVAC or atmospheric clean-up system and there is no EPU effect on HVAC systems during normal operation.

Based on a review of design basis calculations and proposed modifications, the design of the HVAC is adequate for EPU. Therefore, the power-dependent HVAC performance meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the ARAVS and TAVS have been reviewed. Entergy concludes the review has adequately accounted for the effects of the proposed EPU on the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-301 capability of these systems to: (1) maintain ventilation in the auxiliary and radwaste equipment areas and in the turbine area; (2) permit pers onnel access; (3) control the concentration of airborne radioactive material in these areas; and (4) control release of gaseous radioactive effluents to the environment. Based on this , Entergy concludes the ARAVS and TAVS will continue to meet the requirements of GDC-60. Therefore, Entergy finds the proposed EPU acceptable with respect to the ARAVS and TAVS.

2.7.6 Engineered

Safety Feature Ventilation System Regulatory Evaluation The function of the engineered safety feature ventilation system (ESFVS) is to provide a suitable and controlled environment for ESF components following certain anticipated transients and DBAs. The review for the ESFVS focused on th e effects of the proposed EPU on the functional performance of the safety-related portions of the system. The review also covered: (1) the ability of the ESF equipment in the areas being serviced by the ventilation system to function under degraded ESFVS performance; (2) the capability of the ESFVS to circulate sufficient air to prevent accumulation of flammable or explosive gas or fuel-vapor mixtures from components (e.g., storage batteries and stored fuel); and (3) the capability of the ESFVS to control airborne particulate material (dust) accumulation. The regulatory acceptance criteria for the ESFVS are based on: (1) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (2) GDC-17, insofar as it requires on-site and off-site electric power systems be provided to permit functioning of SSCs important to safety; and (3) GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. At GGNS, this function is performed by the following ventilation and cooling systems: (1) Safeguard Switchgear and Battery Rooms; (2) Diesel Generator Rooms; (3) SSW Pumphouse; (4) ECCS Pump Rooms; (5) ESF Electrical Switchgear Rooms; (6) FPCCS Pump Room; (7) Containment Cooling; (8) SGTS; a nd (9) Auxiliary Building. These ESFVSs are described in UFSAR Sections 9.4.5, "Miscellane ous Safety-Related Ventilation and Cooling Systems," 9.4.7, "Containment Cooling System," 6.5.3, "Standby Gas Treatment System," and 9.4.6, "Auxiliary Building Ventilation System." NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-302 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.6 of the CLTR addresses the effect of CPPU on CLTR Power-Dependent HVAC. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Power-Dependent HVAC Performance [[

     ]] Meets CLTR Disposition The CLTR states that slightly higher process temperatures and electrical loads occur as a result of EPU.

The ESF HVAC systems consist mainly of heati ng, cooling supply, exhaust, and recirculation units serving the Safeguard Switchgear and Battery Rooms, Diesel Generator Rooms, SSW system pumphouses, and the following areas of the Auxiliary Building: ECCS Pump Rooms (RHR, HPCS, LPCS, and RCIC), ESF Electrical Switchgear Rooms, and the FPCCS Pump Room. These systems do not function to contro l the concentration of airborne radioactive material in these areas during normal operation, during AOOs, and after postulated accidents. The control of the concentration of airborne radioactive material in the secondary containment during normal operation, during AOOs, and after postulated accidents is accomplished using the SGTS described in Sections 2.5.2 and 2.6.6. During normal operation, the HVAC systems serving these areas are unaffected by the EPU because the process temperatures remain bounded by CLTP conditions. The post-LOCA Auxiliary Building room temperatures (including the effects of LOOP and loss of non-safety related HVAC systems) are affected by EPU as a result of an increase in the SP accident temperature. The increased heat loads due to the higher SP temperature has the most effect on the ECCS Pump Rooms where the room temperatures will increase between 2ºF and 9ºF. The room temperatures of the ESF Electrical Switchgear Rooms increase by 2ºF or less and the room temperature of the FPCCS Pump Room is not affected by EPU. The resultant EPU Auxiliary Building post-LOCA room temperatures have been evaluated and accounted for in the GGNS environmental qualification program. Therefore, the power-dependent HVAC performance meets all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-303 Conclusion The effects of the proposed EPU on the ESFVS have been evaluated. Entergy has determined the evaluation adequately accounted for the eff ects of the proposed EPU on the ability of the ESFVS to provide a suitable and controlled environment for ESF components. Entergy further concludes the ESFVS will continue to assure a suitable environment for the ESF components following implementation of the proposed EPU. Entergy also concludes the ESFVS will continue to suitably control the release of gaseous radioactive effluents to the environment following implementation of the proposed EPU. Based on this, Entergy concludes the ESFVS will continue to meet the requirements of GDC s 4, 17, and 60. Therefore, Entergy finds the proposed EPU acceptable with respect to the ESFVS.

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2-304 2.8 Reactor Systems 2.8.1 Fuel System Design Regulatory Evaluation The fuel system consists of arrays of fuel r ods, burnable poison rods, spacer grids and springs, end plates, channel boxes, and reactivity control rods. Entergy reviewed the fuel system to ensure that: (1) the fuel system is not damaged as a result of normal operation and AOOs; (2) fuel system damage is never so severe as to prevent control rod insertion when it is required; (3) the number of fuel rod failures is not underestimated for postulated accidents; and (4) the fuel is adequately cooled during all operational modes. The review covered fuel system damage mechanisms, limiting values for important parameters, and performance of the fuel system during normal operation, AOOs, and postulated accide nts. The regulatory acceptance criteria are

based on: (1) 10 CFR 50.46, insofar as it establis hes standards for the calculation of ECCS performance and acceptance criteria for that calculated performance; (2) GDC-10, insofar as it requires that the reactor core be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operati on, including the effects of AOOs; (3) GDC-27, insofar as it requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; and (4) GDC-35, insofar as it requires that a system to provide abundant emergency core cooling be provided to transfer heat from the reactor core

following any LOCA. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The fuel system design is described in UFSAR Section 4.2.

Technical Evaluation GGNS transitioned to GNF2 fuel in Cycle 18 and the core design for EPU implementation (Cycle 19) includes only GE fu el types. Because GGNS uses GNF2 fuel, the CLTR is not applicable for fuel design dependent evaluations. Therefore, the fuel product line design is evaluated on a plant-specific basis in accordance with ELTR1. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-305 The EPU evaluations assume a reference equilibrium core of GNF2 fuel. GNF2 fuel is resident in the GGNS core. The fuel design limits are esta blished for all new fuel product line designs as a part of the fuel introduction and reload analyses. [[

     ]] At the OLTP as well as at the EPU RTP conditions, all fuel design limits continue to be met by planned deployment of fuel enrichment and burna ble poison. However, revised loading patterns, larger batch sizes and potentially updated bundle designs may be used to provide additional operating flexibility and maintain fuel cycle length. Therefore, because the fuel design limits are evaluated in accordance with approved methodology for each core reload, the assessmen t of the GGNS fuel product line design is acceptable.

Conclusion The effects of the proposed EPU on the fuel system design of the fuel assemblies, control systems, and reactor core have been reviewed. Entergy concludes the review has adequately accounted for the effects of the proposed EPU on the fuel system and demonstrated: (1) the fuel system will not be damaged as a result of normal operation and AOOs; (2) the fuel system damage will never be so severe as to prevent c ontrol rod insertion when it is required; (3) the number of fuel rod failures will not be underestimated for postulated accidents; and (4) the fuel is adequately cooled during all operational mode

s. Based on this, Entergy concludes the fuel system and associated analyses will continue to meet the requirements of 10 CFR 50.46, GDC-10, GDC-27, and GDC-35 following implementa tion of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the fuel system design.

2.8.2 Nuclear

Design Regulatory Evaluation Entergy reviewed the nuclear design of the fuel assemblies, control systems, and reactor core to ensure fuel design limits will not be exceeded during normal operation and anticipated operational transients, and the effects of postulated reactivity accidents will not cause significant damage to the RCPB or impair the capability to c ool the core. The review covered core power distribution, reactivity coefficients, reactivity control requirements and control provisions, control rod patterns and reactivity worths, criticality, burnup, and vessel irradiation. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires the reactor core be designed with appropriate margin to assure th at SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs; (2) GDC-11, insofar as it requires the reactor core be designed so that the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity; (3) GDC-12, insofar as it NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-306 requires the reactor core be designed to assure that power oscillations, which can result in conditions exceeding SAFDLs, are not possible or can be reliably and readily detected and suppressed; (4) GDC-13, insofar as it requires instrumentation and controls be provided to monitor variables and systems affecting the fission process over anticipated ranges for normal operation, AOOs and accident conditions, and to maintain the variables and systems within prescribed operating ranges; (5) GDC-20, insofar as it requires the protection system be designed to initiate the reactivity control systems automatically to assure that acceptable fuel design limits are not exceeded as a result of AOOs and to automatically initiate operation of systems and components important to safety under accident c onditions; (6) GDC-25, insofar as it requires the protection system be designed to assure th at SAFDLs are not exceeded for any single malfunction of the reactivity control systems; (7) GDC-26, insofar as it requires two independent reactivity control systems be provided, with both systems capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes; (8) GDC-27, insofar as it requires the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for st uck rods, to assure the capability to cool the core is maintained; and (9) GDC-28, insofar as it requires the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The nuclear design of the fuel assemblies is described in UFSAR Section 4.3.

Technical Evaluation 2.8.2.1 Core Operation EPU increases the average power density proportional to the power increase and has some effects on operating flexibility, reactivity characteristics and energy requirements. The additional energy requirements for EPU are met by an increase in bundle enrichment, an increase in the reload fuel batch size, and/or changes in fuel loading pattern to maintain the desired plant

operating cycle length. Because GGNS uses GNF2 fuel, the CLTR is not applicable for fuel design dependent evaluations. Therefore, the core design and fuel thermal margin monitoring threshold are evaluated on a plant-specific basis in accordance with ELTR1. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-307 2.8.2.1.1 Core Design GGNS is currently licensed with an average bundle power of 4.87 MW/bundle. The average bundle power for EPU is 5.51 MW/bundle. The EPU average bundle power is comparable to the range of other operating BWRs. The maximum allowable peak bundle power is not increased by power uprate. The additional energy requirements for power uprate are met by an increase in bundle enrichment, an increase in the reload fuel batch size, and/or changes in fuel loading pattern to maintain the desired plant operating cycle length. The power distribution in the core is changed to achieve increased core power, while limiting the minimum critical power ratio (MCPR), linear heat generation rate (LHGR), and maximum average planar linear heat generation rate (MAPLHGR) in any individual fuel bundle to be within limits as defined in the COLR. The reactor core design power distribution represents a limiting thermal operating state at design conditions. It includes allowances for the combin ed effects on the fuel heat flux and temperature of the gross and local power density distributi ons, control rod pattern, and reactor power level adjustments during plant operation. NRC approved core design methods were used to analyze core performance at the EPU RTP level. Deta iled fuel cycle calculations of a representative equilibrium core design for this plant demonstrate the feasibility of EPU RTP operation while maintaining fuel design limits. Thermal-hydraulic design and operating limits ensure an

acceptably low probability of boiling transition-i nduced fuel cladding failure occurring in the core, even for the most severe postulated operational transients. As needed, limits are also placed on fuel average planar linear heat genera tion rate (APLHGR) and/or fuel rod LHGRs in order to meet both PCT limits for the limiting LOCA and fuel mechanical design bases. The subsequent reload core designs for operation at the EPU RTP level will take into account the above limits to ensure acceptable differences between the licensing limits and their corresponding operating values. EPU may result in a small change in fuel burnup, the amount of fuel to be used, and isotopic con centrations of the radionuclides in the irradiated fuel relative to the original level of burnup. NRC-approved limits for burnup on the fuel designs are not exceeded (Reference 67). For an example CLTP condition, the End of Cycle (EOC) 18 peak bundle average discharge exposure is predicted to be [[ ]]. The EPU equilibrium cycle (full core of GNF2) peak bundle average discharge exposure is predicted to be [[ ]]. GNF2 fuel is required to have Bundle Average Discharge Exposure less than [[ ]]. This is in compliance with the fuel-dependent limitations on discharge burnup. For an example CLTP conditi on, the Cycle 18 weighted average fresh bundle enrichment is predicted to be [[ ]]. The EPU equilibrium cycle (full core of GNF2) core weighted average fresh bundle enrichment is predicted to be [[ ]]. There is no fuel product design basis maximum licensed GNF2 bundle enrichment. The maximum licensed fuel product design basis pellet enrichment is [[ ]]. This is in compliance with fuel-dependent design basis limits on U-235 isotopic enrichment. Also, due to the higher steady-state operating NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-308 power associated with the EPU, the short-term cu rie content of the reactor fuel increases. The effects of higher power operation on radiati on sources and DBA doses are discussed in Section 2.9. Therefore, because the core design is established in accordance with approved methodology for each core reload, the assessment of this topic for GGNS is acceptable. 2.8.2.1.2 Fuel Thermal Margin Monitoring Threshold The CLTR states that the percent power level above which fuel thermal margin monitoring is required may change with EPU. The original plant operating licenses set this monitoring threshold at a typical value of 25% of RTP. [[

]] For EPU, as specified in the CLTR, the fuel thermal margin monitoring threshold is scaled down, if necessary, to ensure that monitoring is initiated [[

                                   ]] then the existing power threshold value must be lowered by a factor of 1.2/P 25. For GGNS, the fuel thermal monitoring thres hold is established at 21.8% of EPU RTP (1.2 / (4,408 MWt / 800 bundles) = 21.8%). A change in the fuel thermal monitoring threshold also requires a corresponding change to the TS reactor core safety limit for reduced pressure or low core flow.

The basis for not monitoring thermal limits below this threshold is the large margin to critical power as described in the TS bases, Section 2.0 Safety Limits. Therefore, with these large margins, there are no transients that have limiting consequences when initiated from the 0 - 21.8 percent power range. Therefore, the fuel thermal margin monitoring threshold meets all CLTR dispositions. 2.8.2.2 Thermal Limits Assessment The effect of EPU on the MCPR safety and operating limits and on the MAPLHGR and LHGR limits for GGNS varies from no effect to a slight effect. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-309 Operating 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 EPU on thermal limits. A representative equilibrium core is used for the EP U evaluation. Cycle-specific core configurations, evaluated for each reload, confirm EPU capability, and establish or confirm cycle-specific limits, as is currently the practice. 2.8.2.2.1 Safety Limit MCPR The Safety Limit MCPR (SLMCPR) can be affected slightly by EPU due to the flatter power distribution inherent in the increased power level. Experience has shown that the power uprate flatte r power distribution results in an increase in the SLMCPR of less than 0.01. A SLMCPR corresponding to the representative equilibrium core is used for the EPU evaluation. Cycle-specific SLMCPR calculations are evaluated for each reload, and establish or confirm cycle-specific limits, as is currently the practice. The SLMCPR for SLO will normally be 0.01 or 0.02 greater than the SLMCPR for two loop operation. A 0.02 value shall be added to the cal culated cycle-specific SLMCPR value for both the single-loop and two-loop SLMCPR in accordance with Limitation and Condition 9.4 of Reference 7. The SLMCPR will be evaluated for the uprated reload core prior to EPU implementation. The calculated values will be reported in the SRLR for the EPU core.

Therefore, because the SLMCPR is established in accordance with approved methodology for each core reload, the assessment of this topic for GGNS is acceptable. 2.8.2.2.2 MCPR Operating Limit Consistent with Sections 5.3.2 and 5.7.2.1 of ELTR1 (Reference 2) and Section 3.4 of ELTR2 (Reference 4), the Operating Limit MCPR (OLM CPR) is calculated by adding the change in MCPR due to the limiting AOO event to the SLMCPR and is determined on a cycle-specific basis from the results of the reload transient analysis. This approach does not change for EPU. The effect of EPU on the AOO events is addressed in Section 2.8.5. The required OLMCPR is not expected to significantly change (<0.03) as shown in Table 3-1 of ELTR1 and Figure 5-3 of ELTR2 and from experience with other uprated BWRs. This small effect is due to the small changes in transient void and scram reactivity response and the flatter radial power distribution at EPU RTP. GEH BWR experience to date for power uprates up to 120% of OLTP confirms this assessment with changes in the operating limit MCPR of +0.018 to -0.013. Limitations and Conditions 9.4 and 9.19 of Reference 7 that require a 0.02 SLMCPR adder and the 0.01 OLMCPR adder ar e well within the experience base for the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-310 effect EPU has on the OLMCPR. For the reference equilibrium core of GNF2 fuel, the OLMCPR for EPU RTP operation is discussed in Section 2.8.5. Therefore, because the OLMCPR is established in accordance with approved methodology for each core reload, the assessment of this topic for GGNS is acceptable. 2.8.2.2.3 MAPLHGR Limit Consistent with Section 5.7.2.2 of ELTR1, EPU ope rating conditions do not usually affect the MAPLHGR Operating Limit. The MAPLHGR Operating Limit ensures that the plant does not exceed regulatory limits established in 10 CFR 50.46 or by the fuel design limits. The MAPLHGR Operating Limit is determined by analyzing the limiting LOCA for the plant. No significant change in operation is anticipated due to the EPU based on experience from other BWR uprates. The ECCS performance is addressed in Section 2.8.5.6.2, and uses a reference equilibrium core of GNF2 fuel for EPU. Compared to CLTP, this evaluation shows that no change in the MAPLHGR limit is required for EPU for SLO or dual recirculation loop operation (DLO).

Therefore, because the MAPLHGR Operating Lim it is established in accordance with approved methodology for each core reload, the assessment of this topic for GGNS is acceptable. 2.8.2.2.4 LHGR Operating Limit Consistent with Section 5.7.2.3 of ELTR1, EPU ope rating conditions do not usually affect the LHGR Operating Limit. The LHGR Operating Limit is determined by the fuel rod thermal-mechanical design and is not affected by EPU. Therefore, because the Maximum LHGR Operating Limit is established in accordance with approved methodology for each core reload, the assessment of this topic for GGNS is acceptable. 2.8.2.2.5 Power and Flow Dependent Limits NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 9.1.2 of the CLTR addresses the effect of CPPU on Power and Flow Dependent Limits. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-311 Topic CLTR Disposition GGNS Result Power and Flow Dependent Limits [[

     ]] Meets CLTR Disposition The CLTR states that power and flow dependent limits are not affected by EPU.

The operating MCPR and LHGR thermal limits are modified when the plant is operating at reduced core flow. This modification is primarily based upon an evaluation of the slow recirculation increase event. The current GGNS analysis is based upon a conservative flow runup rod line that bounds operation to the rod line documented in Section 1.2. [[

     ]] Similarly, the thermal limits are modified when the plant is operating at less than 100% power.

[[

     ]]  The power and flow dependent limits at GGNS meet all CLTR dispositions.

2.8.2.3 Reactivity Characteristics As noted in Section 2.3 of the CLTR, the higher core energy requirements of power uprate may reduce the hot excess reactivity and reduce operating shutdown margin. Because GGNS uses GNF2 fuel, the effect of power uprate on core reactivity is evaluated as described in Section 5.7.1 of ELTR1 (Reference 2). The higher core energy requirements of power uprate may reduce the hot excess reactivity and reduce operating shutdown margin during the cycle.

Based on experience with previous plant-specific power uprate submittals, the required hot excess reactivity and shutdown margin can be achieved for EPU through appropriate fuel and core design. These parameters must meet the approved limits established in Reference 5 on a cycle-specific basis. Therefore, because plant reactivity margins are established in accordance with approved methodology for each core reload, the assessment of these topics for GGNS is acceptable. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-312 2.8.2.4 Additional Topics from GEH Li censing Topical Report NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" GEH LTR NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains," (Reference 7) was approved by the NRC in July 2009. Section 9.0 of the NRC SER contains several limitations and conditions that require additional topics to be addressed in EPU applications. Appendix A provides a summary disposition of these limitations and conditions. Additional information for those topics related to Nuclear Design is provided below. This topical report is applicable to fuel designs through GE14. The GGNS EPU core design includes the GNF2 fuel product line. All of the GE14 restrictions have been applied to the GNF2 fuel. 2.8.2.4.1 Steady-State 5 Percent Bypass Voiding Evaluations Limitation and Condition 9.17 of Reference 7 re quires the bypass voiding to be evaluated on a cycle-specific basis to confirm that the void fraction remains below 5 percent. Limitation and Condition 9.17 is applicable to EPU conditions consistent with Reference 68. The best-estimate means of determining 4-channel bypass void fraction is with TRACG. TRACG was applied in response to Methods RA I 14 (Reference 69). TR ACG is capable of accurately modeling bypass heating and cross flow.

A conservative approach (ISCOR) was disc ussed in References 70 and 71. ISCOR conservatively calculates hot bypass channel voiding using its direct moderator-heating model and providing no credit for cross flow while applying additional conservatism with bounding 4-bundle peaking. The use of ISCOR is a more simplified and efficient process to implement compared to the use of TRACG and typically demonstrates margin to the 5% bypass void fraction requirement at the LPRM D Level. For the GGNS reload core prior to EPU implementation, licensing evaluations are performed with the conservative ISCOR process at licensed EPU core power and minimum core flow (e.g., 115 % OLTP, 92.8% flow). The purpose of the calculation is to confirm that the bypass void fraction remains below 5 percent at all LPRM levels when operating at steady-state conditions within the licensed operating domain consistent with Reference 68.

If the resulting bypass void fraction is found to exceed the 5% requirement, it is acceptable to relax the conservative ISCOR input assumptions as long as the overall approach can be demonstrated to remain conservative relative to TRACG. It is also acceptable to perform a cycle-specific TRACG analysis with consideration of assumptions that tend to maximize bypass void fraction (e.g. bypass flow and 4-bundle peaking). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-313 The highest calculated bypass voiding at any LP RM level will be provided with the plant-specific SRLR. For the representative equilibrium core used in EPU evaluation, the steady-state 5% bypass voiding evaluation is provided in the following table: % of Rated Core Power (EPU) % of Rated Core Flow (EPU) Hot Channel Void Fraction in Bypass Region at Instrumentation D Level 100 100 0.000 100 92.8 0.000 2.8.2.4.2 Power-to-Flow Ratio Limitation and Condition 9.3 and Reference 68 re quire plant-specific EPU applications to confirm that the core thermal power to core flow ratio will not exceed 50 MWt/Mlbm/hr at the low flow point at rated power (e.g., EPU: 100% Power / 92.8% Flow) state point in the allowed operating domain. The core thermal power to total core flow ratio is reported in the following table: % of Rated Core Power (EPU) % of Rated Core Flow (EPU) Power-to-Flow Ratio (MWt/Mlbm/hr) 100 92.8 42.22 The 115% OLTP power (i.e. the EPU power) is 4,408 MWt and the minimum flow at this power is 92.8% rated flow, or 104.4 Mlbm/hr. Therefore, the resulting core thermal power to total core flow ratio at the 100% EPU RTP is 42.22 MWt/Mlbm/hr and does not exceed 50 MWt/Mlbm/hr.

This power-to-flow ratio analysis is cons istent with the GEH letter discussing the implementation of methods limitations from NEDC-33173P (Reference 68). The power-flow map is independent of fuel de sign and does not change cycle to cycle. Therefore, the power-to-flow ratio for GGNS's future EPU cycles will also remain below 50 MWt/Mlbm/hr at this state point. 2.8.2.4.3 R-Factor Limitation and Condition 9.6 requires the plant-speci fic R-factor calculation at a bundle level be consistent with lattice axial void conditions expected for the hot channel operating state. The GNF2 bundle R-factors generated for this task are consistent with Global Nuclear Fuel LLC (GNF) standard design procedures which use an axial void profile shape with 60% average in-NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-314 channel voids. This is consistent with lattice axial void conditions expected for the hot channel operating state. Figure 2.8-18 shows bundle average void fractions corresponding to hot channels with low critical power ratios (MCPRs) from the GGNS EPU equilibrium core. The figure demonstrates that the generic R-factor profile, with an aver age void fraction of 0.60, is representative of the MCPR-limiting void conditions predicted by PANAC11. 2.8.2.4.4 Plant-Specific Application Limitation and Condition 9.24 requires plant-specifi c EPU applications to provide a prediction of key parameters for cycle exposures for operation at EPU. The following parameters: (1) Power of Peak Bundle; (2) Coolant Flow for Peak Bundle; (3) Exit Void Fraction for Peak Power Bundle; (4) Maximum Channel Exit Void Fr action; (5) Core Average Exit Void Fraction; (6) Peak LHGR; and (7) Peak Nodal Exposures are shown in Figures 2.8-1 through 2.8-6 and Table 2.8-1. The GGNS data are plotted with th e available EPU experience base as required by Limitation and Condition 9.24. Quarter core maps with mirror symmetry are plotted in Figures 2.8-7 through 2.8-15 showing bundle power, bundle operating MCPR , and LHGR for Beginning of Cycle (BOC), Middle of Cycle (MOC), and EOC. Because the minimum margins to specific limits occur at exposures other than the traditional BOC, MOC, and EOC, the data are provided at these other exposures as applicable (Figures 2.8-16 and 2.8-17). Note that the bundle powers in Figures 2.8-7 through 2.8-9 are dimensionless. To obtain the bundle powers in MWt, multiply each value by the average bundle power of 5.51 MWt. The aver age bundle power is equal to 4,408/800, where 4,408 MWt is the EPU RTP and 800 is the total number of bundles in the core. 2.8.2.4.5 Application of 10 Weight Percent Gd Limitation and Condition 9.13 requires review and approval of 10 weight percent Gd to EPU applications. For GGNS, the maximum burnable poison concentration used is 8.0 Weight Percent Gd 2 O 3; therefore, Limitation and Condition 9.13 is not applicable. 2.8.2.4.6 Mixed Core Method 1 Limitation and Condition 9.21 requires that plants implementing EPU or MELLLA+ with mixed fuel vendor cores provide plant-specific justification for extension of GE's analytical methods or codes. The content of the plant-specific applicati on will cover the topics addressed in this SE as well as subjects relevant to application of GE's methods to legacy fuel. Alternatively, GE may supplement or revise LTR NEDC-33173P for mixed core application. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-315 For GGNS, there is no mixed vendor core; therefore, the mixed core is not evaluated and Limitation and Condition 9.21 is not applicable. 2.8.2.4.7 Mixed Core Method 2 Limitation and Condition 9.22 requires that for a ny plant-specific applications of TGBLA06 with fuel type characteristics not covered in the Interim Methods review, GE needs to provide assessment data similar to that provided for the GE fuels. The Interim Methods review is applicable to all GE lattices up to GE14. Fuel lattice designs, other than GE lattices up to GE14, with the following characteristics are not covered by this review: square internal water channels water crosses Gd rods simultaneously adjacent to water and vanished rods 11x11 lattices MOX fuel The acceptability of the modified epithermal slowing down models in TGBLA06 has not been demonstrated for application to these or other geometries for expanded operating domains. Significant changes in the Gd rod optical thickness will require an evaluation of the TGBLA06 radial flux and Gd depletion modeling before bei ng applied. Increases in the lattice Gd loading that result in nodal reactivity biases beyond thos e previously established will require review before the GE methods may be applied. For GGNS, there is no mixed vendor core; therefore, the mixed core is not evaluated and Limitation and Condition 9.22 is not applicable.

Conclusion The effects of the proposed EPU on the nuclear design of the fuel assemblies, control systems, and reactor core have been reviewed. Entergy concludes the review has adequately accounted for the effects of the proposed EPU on the nuclear design and has demonstrated the fuel design limits will not be exceeded during normal or anticip ated operational transients, and the effects of postulated reactivity accidents will not cause significant damage to the RCPB or impair the capability to cool the core. Based on this evaluation and in coordination with the reviews of the fuel system design, thermal and hydraulic design, and transient and accident analyses, Entergy concludes the nuclear design of the fuel assemblies, control systems, and reactor core will continue to meet the applicable requirements of GDCs 10, 11, 12, 13, 20, 25, 26, 27, and 28. Therefore, Entergy finds the proposed EPU acceptable with respect to the nuclear design. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-316 2.8.3 Thermal and Hydraulic Design Regulatory Evaluation Entergy reviewed the thermal and hydraulic design of the core and the RCS to confirm the design: (1) has been accomplished using acceptable analytical methods; (2) is equivalent to or a justified extrapolation from proven designs; (3) provides acceptable margins of safety from conditions which would lead to fuel damage during normal reactor operation and AOOs; and (4) is not susceptible to thermal-hydraulic inst ability. The review also covered hydraulic loads on the core and RCS components during norma l operation and DBA conditions and core thermal-hydraulic stability under normal operation and anticipated transients without scram (ATWS) events. The regulatory acceptance crite ria are based on: (1) GDC-10, insofar as it requires the reactor core be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal ope ration, including the effects of AOOs; and (2) GDC-12, insofar as it requires the reactor core and associated coolant, control, and protection systems be designed to assure power oscillations, which can result in conditions exceeding SAFDLs, are not possible or can reliably and readily be detected and suppressed. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. Core thermal and hydraulic design is described in UFSAR Section 4.4.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 2.4.4 of the CLTR a ddresses the effect of CPPU on Plants with Option III. Section 9.3.3 of the CLTR addresses the effect of CPPU on ATWS with Core Instability. The results of this evaluation are described below. Section 3.2 of ELTR2 documents interim corrective actions (ICAs) and four long-term stability options. The evaluation of thermal stability for GGNS is based on the implementation of the Option III solution. GGNS is implementing the PRNM system and has proposed a license condition to ensure approval of Option III prior to operation at EPU power levels. A generic evaluation was performed for the ICAs in Secti on 3.2.1 of Reference 4. This generic evaluation continues to be applicable for EPU. GGNS implements the Backup Stability Protection (BSP) measures as the stability licensing basis. These measures are discussed in Section 2.8.3.1.3. For the long-term options, evaluations are core reload dependent and are performed for each reload

fuel cycle. The analyses of Option III are addressed below. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-317 GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Option III (OPRM Trip-Enabled Region and Trip Setpoint) [[ Meets CLTR Disposition Option III (Hot Channel Oscillation Magnitude)

Meets CLTR Disposition ATWS with Core Instability

     ]] Meets CLTR Disposition 2.8.3.1 Option III 2.8.3.1.1 Option III - OPRM Trip Enabled Region and Trip Setpoint The CLTR states that the Option III trip setpoint may be affected by EPU operating conditions.

The OPRM trip-enabled region will be rescaled with EPU. Option III is a detect-and-suppress solution, which combines closely spaced LPRM detectors into "cells" to effectively detect any mode of reactor instability. Plants implementing Option III must demonstrate that the Option III trip setpoint is adequate to provide SLMCPR protection for anticipated reactor instability. This evaluation is dependent upon the core and fuel design and is performed for each reload. [[

     ]] GGNS has requested to be licensed for the Option III long-term stability solution. This feature is a part of the PRNM system. Option III evaluations are core reload dependent and are performed for each reload fuel cycle (Reference 72). The GGNS Option III hardware is being installed during RF18 and is required to be implemented prior to ascension to EPU power. The first 90 days of operation during Cycle 19 will be an OPRM monitoring period during which the OPRM trip function is not enabled. In the event that the Oscillation Power Range Monitor (OPRM) system is declared inoperable, GGNS will operate under the BWROG Guidelines for BSP as

described in Reference 73. Cycle-specific setpoints and BSP regions are determined and documented in the same SRLR. The BSP region in tercepts with the Natural Circulation Line (NCL) and with the High Flow Control Line (HFCL) are the same as or more conservative than the ICA region intercepts in absolute power and core flow. The Option III solution combines closely spaced LPRM detectors into "cells" to effectively detect either core-wide or regional modes of reactor instability. These cells are termed OPRM cells and are configured to provide local area coverage with multiple channels. The GGNS NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-318 Option III hardware combines the LPRM signals and evaluates the cell signals with instability detection algorithms. The Period Based Detection Algorithm (PBDA) is the only algorithm credited in the Option III licensing basis. Two defense-in-depth algorithms, referred to as the Amplitude Based Algorithm (ABA) and the Growth Rate Based Algorithm (GRA), offer a high degree of assurance that fuel failure will not occur as a consequence of stability related

oscillations. The OPRM trip is armed only when plant operation is within the OPRM trip-enabled region. The OPRM trip-enabled region is generically defined as the region on the power/flow map with power 30% of RTP and core flow < 60% of rated core flow. For EPU, the GGNS OPRM trip-enabled region is rescaled to maintain the same absolute power/flow region boundaries. Because the rated core flow is not changed, the 60% core flow boundary is not rescaled. The 29% CLTP boundary is rescaled to the 26% EPU thermal power limit using the CLTP/EPU ratio.

The GGNS OPRM trip-enabled region is shown in Figure 2.8-19. The BSP evaluation described in Section 2.8.3.1.3 shows that the generic Option III trip-enabled region is adequate. The adequacy of the OPRM trip-enabled region will be confirmed for each fuel reload.

Stability Option III provides SLMCPR protection by generating a reactor scram if a reactor instability that exceeds the specified OPRM trip se tpoint is detected. The OPRM setpoint is determined per an NRC approved methodology (References 72 and 74).

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. These OLMCPRs are calcula ted for a range of OPRM setpoints for EPU operation. Selection of an appropriate OPRM trip setpoint is then based upon the OLMCPR required to provide adequate SLMCPR protection. This determination relies on the DIVOM curve (Delta Critical Power Ratio Over Initial Critical Power Ratio Versus Oscillation Magnitude) to determine an OPRM setpoint th at protects the SLMCPR during an anticipated instability event. The DIVOM slope was developed based on a TRACG evaluation in accordance with the BWROG Regional Mode DIVOM Guideline (Reference 74). Option III (OPRM trip-enabled region and trip setpoint) meets all CLTR dispositions. 2.8.3.1.2 Option III - Hot Channel Oscillation Magnitude The CLTR states that the Option III trip setpoint may be affected by EPU operating conditions. The OPRM trip-enabled region will be rescaled with EPU. The Option III automatic scram is provided by the OPRM system. The generic analyses for the Option III hot channel oscillation magnitude and the OPRM hardware were designed to be independent of core power. [[

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2-319

     ]] Although the Option III solution requires cycle-specific evaluations, a demonstration analysis was performed based on an equilibrium GNF2 core design. For this analysis, the calculated DIVOM slope was used. As shown in Table 2.8-2, with an assumed SLMCPR of 1.10, and an assumed rated OLMCPR of 1.39, an OPRM amplitude setpoint of 1.08 is the highest setpoint that may be used. The actual setpoint will be established at the time of each fuel reload based on

cycle-specific core design. Consistent with Limitation and Condition 9.18 per Reference 7, the OPRM system will incorporate a 5% calibration error on the OP RM setpoints to address the bypass voiding uncertainty at low-flow conditions. This ca libration error has been included in the OPRM amplitude setpoints shown in Table 2.8-2. In addition, consistent with Limitation and Condition 9.19 of Reference 7, Table 2.8-2 incl udes a 0.01 adder so that the bypass voiding penalty incorporated in the transient-based OLMCPR is not used as a credit in stability analysis, consistent with Reference 68. Based on the us e of the Option III stability solution, the APRM calibration error required by Limitation and Condition 9.18 is not applicable (see Section 6.2 of

Reference 7). Option III (hot channel oscillation magnitude) meets all CLTR dispositions. 2.8.3.1.3 BSP Evaluation GGNS implements BSP (Reference 73) as the stability licensing basis should the Option III OPRM system be declared inoperable. The BSP evolved from the stability ICAs (Reference 75), which restrict plant operation in the high power, low core flow region of the BWR power/flow operating map. The ICAs provide guidance that reduces the likelihood of an instability event by limiting the period of operation in regions of the power/flow map most susceptible to thermal hydraulic instability. If the Option III OPRM system is declared inoperable, implementation of the associated BSP regions will constitute the stability licensing ba sis for GGNS (Reference 73). The BSP regions consist of two regions (I-Scram and II-Controlled Entry), which are reduced from the three ICA regions (I-Scram, II-Exit and III-Controlled Entry) in Reference 75. The standard ICA region endpoints on the HFCL and on the NCL define th e base BSP region endpoints on the HFCL and on the NCL. The bounding plant-and-cycle-specific BSP region endpoints must enclose the corresponding base BSP region endpoints on the HFCL and the NCL. If a calculated BSP region endpoint is located inside the corresponding ba se BSP region endpoint, the corresponding base BSP region endpoint must replace it. That is, the selected points will result in the largest, or most conservative, region sizes. The proposed BSP Scram and Controlled Entry region boundaries may be constructed by connec ting the corresponding bounding endpoints on the HFCL and the NCL using the Modified Shape Function (MSF) (Reference 76). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-320 The GNF2 equilibrium demonstration analysis was used to determine the ODYSY calculated BSP boundaries as shown in Table 2.8-3. Th ese ODYSY-calculated BSP boundaries are all smaller than the corresponding base BSP boundari es and hence the base BSP boundaries are adopted for the demonstration analysis as shown in Figure 2.8-19. 2.8.3.2 ATWS with Core Instability The CLTR states that the ATWS with core in stability event occurs at natural circulation following a RPT. Therefore, it is initiated at approximately the same power level as a result of EPU operation because the MELLLA upper boundary is not increased. The core design necessary to achieve EPU operations ma y affect the susceptibility to coupled thermal-hydraulic/neutronic core oscillations at the natural circulation condition, but would not

significantly affect the event progression. The analysis and results of the limiting ATWS events are presented in Section 2.8.5.7. 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 77 and 78 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum flow point. [[

     ]] EPU allows plants to increase their operating thermal power but does not allow an increase in control rod line.  [[

       ]] Feedwater heater out-of-service (FWHOOS) and final feedwater temperature reduction (FFWTR) are operational flexibility options that allow continued operation with RFWT. GGNS is not currently licensed to operate with FFWTR. Initial operating conditions of FWHOOS and FFWTR do not significantly affect the ATWS instability response reported in References 77 and 78. The limiting ATWS evaluation assumes that all FW heating is lost during the event and the injected FW temperature approaches the lowest achievable MC hot well temperature.  [[

                                                                    ]]  Initial power is unchanged for both the FWHOOS and FFWTR conditions - the additional reactivity associated with the RFWT is typically offset with control rods, as needed. For both the FWHOOS and FFWTR conditions, an ATWS event analysis would be initiated from the same limiting power/flow state point assumed for the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-321 normal FW temperature case (state point 'D' in Figure 1-1) and transition to essentially the same natural circulation state point (state point 'A' in Figure 1-1) prior to the onset of power oscillations. [[

     ]] Operator actions will mitigate an ATWS inst ability event. The actions contained in References 77 and 78 bound the entire BWR fleet a nd are applicable to GGNS. The conclusion of Reference 78 and the associated NRC SER th at the analyzed operator actions effectively mitigate an ATWS instability even t are applicable to the operating conditions expected for EPU at GGNS. Therefore, the EPU effect on ATWS with core instability at GGNS meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the thermal and hydraulic design of the core and the RCS have been reviewed. Entergy concludes the revi ew has adequately accounted for the effects of the proposed EPU on the thermal and hydraulic design and demonstrated the design: (1) has been accomplished using acceptable analytical methods; (2) is a proven design; (3) provides acceptable margins of safety from conditions that would lead to fuel damage during normal reactor operation and AOOs; and (4) is not susceptible to thermal-hydraulic instability. Entergy further concludes it has adequately accounted for the effects of the proposed EPU on the hydraulic loads on the core and RCS components. Based on this, Entergy concludes the thermal and hydraulic design will continue to meet the requirements of GDCs 10 and 12 following implementation of the proposed EPU. Theref ore, Entergy finds the proposed EPU acceptable with respect to thermal and hydraulic design.

2.8.4 Emergency

Systems 2.8.4.1 Functional Design of Control Rod Drive System Regulatory Evaluation Entergy reviewed the functional performance of the CRD system to confirm the system can, after the implementation of the proposed EPU: (1) e ffect a safe shutdown, respond within acceptable limits during AOOs; and (2) prevent or mitigate th e consequences of postulated accidents. The review also covered the CRD system cooling system to ensure that it will continue to meet its design requirements. The regulatory acceptance cr iteria are based on: (1) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (2) GDC-23, insofar as it requires the protection system be designed to fail into a safe state; (3) GDC-25, insofar as it requires the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems; (4) GDC-26, insofar as it requires two independent reactivity control systems be NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-322 provided, with both systems capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes; (5) GDC-27, insofar as it requires the reactivity control systems be designed to have a combin ed capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; (6) GDC-28, insofar as it requires the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its suppor t structures, or other reactor vessel internals so as to significantly impair the capability to c ool the core; (7) GDC-29, insofar as it requires the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in event of AOOs; and (8) 10 CFR 50.62(c)(3), insofar as it requires BWRs have an alternate rod insertion (ARI) system diverse from the reactor trip system, and the ARI system have redundant scram air header exhaust valves. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The CRD System is described in UFSAR Section 4.6.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 2.5 of the CLTR a ddresses the effect of CPPU on the functional design of the CRD system. The results of this evaluation are described below. As stated in Section 2.5 of the CLTR, the CRD system is used to control core reactivity by positioning neutron absorbing control rods within the reactor and to scram the reactor by rapidly inserting withdrawn control rods into the core. The GGNS ARI System is not affected by EPU because it has no thermal power dependency. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-323 Topic CLTR Disposition GGNS Result Scram Time Response [[ Meets CLTR Disposition CRD Positioning

Meets CLTR Disposition CRD Cooling

Meets CLTR Disposition CRD Integrity

     ]] Meets CLTR Disposition 2.8.4.1.1 Scram Time Response The CLTR states that for BWR/6 plants, the incr ease in the transient pressure response due to EPU increases the scram time.

At normal operating conditions, the CRD HCU accumulator supplies all of the pressure to complete the scram. Because the normal reactor dome pressure for EPU does not change, the scram time performance relative to current plant operation is essentially the same. Therefore, BWR/6 plants will retain their current TS scram requirements. [[

     ]] For the ASME overpressure protection analyses, the EPU transient reactor pressure remains within the generic envelope. For the AOO analys es, the EPU transient reactor pressure is not bounded by the generic envelope; therefore, revised scram times have been developed for use in the plant specific reload analysis core design. To address the expected increase in scram times, GGNS is implementing Option B scram times to maintain operating margin. 2.8.4.1.2 Control Rod Drive Positioning and Cooling

As stated in Section 2.5 of the CLTR, the increase in reactor power at the EPU operating condition results in [[

                                                                                                              ]] from the CRD System to the CRDs during normal plant operation.

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2-324 2.8.4.1.2.1 Control Rod Drive Positioning The CLTR states that, with reactor dome pressure unchanged, there is [[

       ]], and the automatic operation of the system flow control valve maintains the requi red drive water pressure. Therefore, the CRD positioning function is not affected. The normal CRD positioning function is an operational consideration, not a safety-related function, a nd is not affected by EPU operating conditions. Plant operating data has confirmed that the CRD system flow control valve does not operate near full open under normal reactor operating conditions with the CRD system in the cooling mode and has sufficient operating margin to compensate for the anticipated pressure changes. The CRD system flow control valve is approximately 50% open at CLTP. Therefore, the valve will maintain the required system pressure. Therefore, the CRD positioning meets all CLTR dispositions.

2.8.4.1.2.2 Control Rod Drive Cooling The CLTR states that, with reactor dome pressure unchanged, there is [[

       ]], and the automatic operation of the system flow control valve maintains the requi red cooling water flow rate. Therefore, the CRD cooling function is not affected. The CRD cooling function is an operational consideration, not a safety-related function, a nd is not affected by EPU operating conditions. Plant operating data has confirmed that the CRD system flow control valve does not operate near full open under normal reactor operating conditions with the CRD system in the cooling mode and has sufficient operating margin to compensate for the anticipated pressure changes. The CRD system flow control valve is approximately 50% open at CLTP. The valve will maintain the design system pressure. Therefore, the CRD cooling meets all CLTR dispositions.

2.8.4.1.3 Control Rod Drive Integrity Assessment

The CLTR states that [[

     ]] on CRD integrity. The transient pressures due to uprated power may create higher pressure loadings.

The postulated abnormal operating condition for the CRD design assumes a failure of the CRD System pressure-regulating valve that applies the maximum pump discharge pressure to the CRD mechanism internal components. For the AS ME RPV overpressure condition, the postulated abnormal peak RPV bottom head pressure due to EPU (1334 psig) is bounded by the ASME reactor overpressure limit of 1375 psig. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-325 The CRD mechanism has been analyzed for an abnormal pressure operation (the application of the maximum CRD pump discharge pressure) that bounds the ASME RPV overpressure condition. The maximum CRD pump discharge pressure is not changed. [[

                                                                                                                                                  ]]  Other mechanical loadings are [[                                                                                          ]] addressed in Sections 2.2.2 and 2.2.3 of this report.

Therefore, the CRD integrity meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the functional design of the CRD system have been evaluated. The ability of the CRD system to effect a safe shutdown, respond within acceptable limits, and prevent or mitigate the conseque nces of postulated accidents following the implementation of the proposed EPU has been demonstrated. In addition, it has been determined that sufficient cooling exists to ensure the system's design bases will continue to be met upon implementation of the proposed EPU. Based on this, Entergy concludes that the fuel system and associated analyses will continue to meet the requirements of GDCs 4, 23, 25, 26, 27, 28, and 29, and 10 CFR 50.62(c)(3) following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the functional design of the CRD system.

2.8.4.2 Overpressure Protection During Power Operation Regulatory Evaluation Relief and safety valves and the RPS provide overpressure protection for the RCPB during power operation. The effect of the proposed EPU on the performance of the relief and safety valves on the MSLs and the piping from these valves to the SP has been reviewed. The regulatory acceptance criteria are based on: (1) GDC -15, insofar as it requires that the RCS and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the RCPB are not exceeded during any condition of normal

operation, including AOOs; and (2) GDC-31, insofar as it requires that the RCPB be designed with sufficient margin to assure that it behaves in a non-brittle manner and that the probability of rapidly propagating fracture is minimized. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The reactor vessel overpressure protection system is described in UFSAR Section 5.2.2. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-326 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.1 of the CLTR addresses the effect of CPPU on Nuclear System Pressure Relief/Overpressure Protection. The re sults of this evaluation are described below. As stated in Section 3.1 of the CLTR, the system operating pressure does not change but the steam flow rate increases. The increased steam flow rate associated with uprated power may increase steam line vibration. The increased core steam generation also causes an increase in the pressurization during some transient events. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Overpressure Capacity [[

     ]] Meets CLTR Disposition The nuclear system pressure relief system prevents overpressurization of the nuclear system during AOOs, the plant ASME Upset overpressure protection event, and postulated ATWS events. The plant SRVs, along with other functi ons, provide this protection. An evaluation was performed in order to confirm the adequacy of the pressure relief system for EPU conditions.

The SRV discharge lines were designed and configur ed so that the discharge backpressure at the valve outlet is not greater than 40% of the inlet pressure. The valves were designed to achieve sonic (choked) flow conditions through the valve up to this backpressure ratio to provide flow independence to the discharge piping losses and backpressure. The backpressure to inlet pressure ratio is a function of discharge line geometry, which will not change with EPU. Therefore, SRV capacity will not be affected by the EPU discharge line backpressure. The adequacy of the pressure relief system is also demonstrated by the overpressure protection evaluation performed for each reload core and by the ATWS evaluation performed for EPU (Section 2.8.5.7). For GGNS, no SRV setpoint increase is needed because there is no change in the dome pressure or simmer margin. Therefore, there is no effect on valve functionality (opening/closing). Two potentially limiting overpressure protection events are typically analyzed for EPU: (1) Main Steam Isolation Valve Closure with Scram on Hi gh Flux (MSIVF), and (2) Turbine Trip with Bypass Failure and Scram on High Flux (ELTR 1, Section 5.5.1.4). However, based on both plant initial core analyses and subsequent power uprate evaluations, the MSIVF is more limiting NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-327 than the TT event with respect to reactor overpre ssure. The EPU evaluations show a 40 to 50 psi difference between these two events. In a ddition, an evaluation of the MSIVF event is performed with each reload analysis. The design pressure of the reactor vessel and RCPB remains at 1,250 psig. The acceptance limit for pressurization events is the ASME code allowable peak pressure of 1,375 psig (110% of design value). The overpressure protection analysis description and analysis method are provided in ELTR1. The MSIVF event is conservatively analyzed assuming a failure of the valve position scram. The analyses also assume that the event initiates at a reactor dome pressure of 1,060 psia (which is higher than the nominal EPU dome pressure), and seven SRVs OOS. There is no change from th e current analysis basis for GGNS as the current licensing basis considers seven SRVs OOS (UFSAR Section 5.2.2, Overpressure Protection). Starting from 102% of EPU RTP, the calculated peak RPV pressu re, located at the bottom of the vessel, is 1,334 psig. The corresponding calculated maximum reactor dome pressure is 1,302 psig. The peak calculated RPV pressure remains below the 1,375 psig ASME limit, and the maximum calculated dome pressure remains below the 1,325 psig TS safety limit. Therefore, the results are acceptable and within the applicable limits. The results of the EPU overpressure protection analysis for the GGNS MSIVF event are consiste nt with the generic analysis in ELTR2. The GGNS response to the MSIVF event is provided as Figure 2.8-20. The MSIVF event is performed using the NRC approved code ODYN (Reference 79) (see Table 1-1). Therefore, the overpressure capacity meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the overpre ssure protection capability of the plant during power operation have been reviewed. The results of that review demonstrate that: (1) pressurization events and overpressure protec tion features adequately account for the effects of the proposed EPU; and (2) the plant will continue to have sufficient pressure relief capacity to ensure that pressure limits are not exceeded. Based on this, Entergy concludes that the overpressure protection features will continue to meet GDCs 15 and 31 following implementation of the proposed EPU. Theref ore, Entergy finds the proposed EPU acceptable with respect to overpressure protection during power operation.

2.8.4.3 Reactor Core Isolation Cooling System Regulatory Evaluation The RCIC system serves as a standby source of cooling water to provide a limited DHR capability whenever the main FW system is isolated from the reactor vessel. In addition, the RCIC system may provide DHR necessary for coping with an SBO. The primary water supply NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-328 for the RCIC system comes from the CST, with a secondary supply from the SP. The effect of the proposed EPU on the functional capability of the system has been reviewed. The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be demonstrated that sharing will not impair its ability to perform its safety func tion; (3) GDC-29, insofar as it requires that the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in event of AOOs; (4) GDC-33, insofar as it requires that a system to provide reactor coolant makeup for protection against small breaks in the RCPB be provided so the fuel design limits are not exceeded; (5) GDC-34, insofar as it requires that a RHR system be provided to transfer fission product decay heat and other residual heat from the

reactor core at a rate such th at SAFDLs and the design conditions of the RCPB are not exceeded; (6) GDC-54, insofar as it requires that piping systems penetrating containment be designed with the capability to periodically test the operability of the isolation valves to determine if valve leakage is within acceptable limits; and (7) 10 CFR 50.63, insofar as it requires that the plant withstand and recover from an SBO of a specified duration. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 is not applicable to GGNS; see UFSAR Section 3.1.2.1.5. The RCIC system is described in UFSAR Section 5.4.6.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 3.9 of the CLTR addresses the effect of CPPU on the RCIC System. The results of this evaluation are described below. The RCIC system evaluation for EPU at GGNS addressed the following topics: System performance and hardware NPSH Adequate core cooling for limiting LOFW events (Addressed in Section 2.8.5.2.3) Inventory makeup - Operational Level 1 avoidance (Addressed in Section 2.8.5.2.3) GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-329 Topic CLTR Disposition GGNS Result System Performance and Hardware (RCIC) [[ Meets CLTR Disposition Net Positive Suction Head (RCIC)

     ]] Meets CLTR Disposition 2.8.4.3.1 System Performance and Hardware The CLTR states that there is no effect on RCIC system performance and hardware due to EPU. The RCIC system is required to maintain sufficient water inventory in the reactor to permit adequate core cooling following a reactor vessel isolation event accompanied by loss of flow from the FW system. The system design injection rate must be sufficient for compliance with the system limiting criteria to maintain the reactor water level above TAF at EPU conditions. The RCIC system is designed to pump water in to the reactor vessel over a wide range of operating pressures. The results of the GGNS pl ant-specific evaluation indicate adequate water level margin above TAF at EPU conditions. Thus, the RCIC injection rate is adequate to meet this design basis event. An operational requirement is that the RCIC system can restore the reactor water level while avoiding Automatic Depressurization System (ADS) timer initiation and MSIVC activation functions associated with the low-low-low react or water level setpoint (Level 1). This requirement is intended to avoid unnecessary initiations of safety systems. The results of the

GGNS plant-specific evaluation indicates that the RCIC system is capable of maintaining the water level outside the shroud above nominal Level 1 setpoint through a limiting LOFW event at EPU conditions. Thus, the RCIC injection rate is adequate to meet the requirements for inventory makeup. The reactor system response to a LOFW transient with RCIC is discussed in Section 2.8.5.2.3. For EPU, there is no change to the normal reactor operating dome pressure (1,040 psia for both CLTP and EPU conditions) and the SRV setpoints remain the same. There is no change to the maximum specified reactor pressure for RCIC system operation, [[

                                                        ]]  The GGNS RCIC pump is adequate to support EPU.

[[

     ]] The system performance and hardware for RCIC meets all CLTR dispositions.

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2-330 2.8.4.3.2 Net Positive Suction Head The CLTR states that there is no effect on RCIC NPSH due to EPU. The GGNS minimum NPSH A for the GGNS RCIC pump does not change because there are no physical changes to the pump suction configuration, and no changes to the system flow rate or minimum atmospheric pressure in the SP or CST. EPU does not affect the capability to transfer the RCIC pump suction on high SP level or low CST level from its normal alignment, the CST, to the SP, and does not change the existing requirements for the transfer. For certain events (e.g., ATWS, SBO, and fire protection), operation of the RCIC system may be accomplished by using the dedicated CST volume as the water source when the SP temperature exceeds the operational limit. Therefore, the specified operational temperature limit for the process water does not change with EPU. Because GGNS is not changing the RCIC pump or its operating parameters, the required NPSH does not change. The effect of EPU on the operation of the RCIC system during SBO events is discussed in Section 2.3.5. Maximum pump speed and maximum pump flow (800 gpm) are unchanged for EPU and there is no change in normal operating dome pressure (1040 psia at CLTP and EPU conditions). The SRV setpoints remain the same. GGNS plant procedures caution operators that damage may occur if the SP temperature exceeds 140 ûF with the RCIC suction aligned to the SP. No RCIC system power-dependent functions or operating requirements (flows, pressure, temperature, and NPSH) are added or changed from the original design or licensing bases. The RCIC NPSH at GGNS meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the ability of the RCIC system to provide DHR following an isolation of main FW event and an SBO event and the ability of the system to provide makeup to the core following a small break in the RCPB have been analyzed. Accounting for the effects of the proposed EPU on these events, it has been demonstrated that the RCIC system will continue to provide sufficient DHR and makeup for these events following implementation of the proposed EPU. Based on this, Entergy concludes that the RCIC system will continue to meet the requirements of GDCs 4, 29, 33, 34, and 54, and 10 CFR 50.63 following implementation of the

proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the RCIC system. 2.8.4.4 Residual Heat Removal System Regulatory Evaluation The RHR system is used to cool down the RCS following shutdown. The RHR system is a low pressure system that takes over the SDC function when the RCS temperature is reduced. The NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-331 effect of the proposed EPU on the functional capability of the RHR system to cool the RCS following shutdown and provide DHR has been revi ewed. The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-34, which specifies requirements for an RHR system. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. GDC-5 is not applicable to GGNS; see UFSAR Section 3.1.2.1.5. The RHR system is described in UFSAR Section 5.4.7.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 3.10 of the CLTR addresses the effect of CPPU on the RHR system. The results of this evaluation are described below. As explicitly stated in Section 3.10 of the CLTR, the RHR system is designed to restore and maintain the reactor coolant inventory following a LOCA and remove reactor decay heat following reactor shutdown for normal, transient, and accident conditions. The EPU effect on the RHR system is a result of the higher decay heat in the core corresponding to the uprated power and the increased amount of reactor heat discharged into the containment during a LOCA. For GGNS, the RHR system is designed to operate in the LPCI mode, SDC mode, SPC mode, Containment Spray Cooling (CSC) mode, and FPC (Supplemental SFP Cooling) assist. GGNS meets all CLTR dispositions. The GGNS RHR system also supports the containment isolation function of the FWLC system. This function wa s not evaluated in the CLTR; it is evaluated herein for GGNS. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-332 Topic CLTR Disposition GGNS Result LPCI Mode [[ Addressed in Section 2.8.5.6.2 Suppression Pool and Containment Spray Cooling Modes

Addressed in Section

2.6.5 Shutdown

Cooling Mode

Meets CLTR Disposition Fuel Pool Cooling Assist

Addressed in Section 2.5.3.1 Feedwater Leakage Control System

     ]] Addressed in Section 2.8.4.4.5 2.8.4.4.1 LPCI Mode The CLTR states that there is no change in the reactor pressures at which the LPCI mode of the RHR system is required. Following an accident, the RHR system operating in the LPCI mode is one of the ECCS. The LPCI mode performan ce is discussed in Section 2.8.5.6.2. The changes in SP temperature and containment pressure do not adversely affect the capability of the RHR system in the LPCI mode. The LPCI mode of the RHR system at GGNS meets all CLTR dispositions.

2.8.4.4.2 Suppression Pool and Containment Spray Cooling The CLTR states that the SP temperature increases as a result of the higher decay heat associated with EPU. The SPC mode is manually initiated following isolation transients and a postulated LOCA to maintain the containment pressure and SP temperature within design limits. The CSC mode reduces containment temperature and pressu re during an accident. The adequacy of these operating modes is demonstrated by the containment analysis (Section 2.6.5). SP temperatures for all EPU events remain within the RHR design limits. Therefore, the SP temperature during a postulated LOCA at EPU conditions does not change the capabilities of RHR system equipment to perform the SPC and CSC functions. Containment pressures for EPU events increased slightly above the CLTP analyzed pressures, but remained below the existing peak containment internal pressure limit. The slight increase in the predicted containment pressure during a postulated LOCA at EPU conditions (See Table 2.6-1) remains within the equipment design parameters and thus does not adversely affect the hardware capabilities of NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-333 RHR system equipment to perform the SPC a nd CSC functions. Therefore, the SP and CSC system meets all CLTR dispositions. 2.8.4.4.3 Shutdown Cooling Mode The CLTR states that a longer time is required for reactor cool down as a result of the higher decay heat associated with EPU. The SDC mode is designed to remove the sensible and decay heat from the reactor primary system during a normal reactor shutdown. This non-safety operational mode allows the reactor to be cooled down within a certain time objective, so that the SDC mode of operation will not become critical path during refueling operations. EPU increases the reactor decay heat, which requires a longer time for cooling down the reactor. The SDC analysis for the EPU determined that the time needed for cooling the reactor to 125 F during normal reactor shutdown, with two SDC loops in service, is increased from 10.6 hours at CLTP conditions to approximately 16.4 hours at EPU conditions. This calculated normal reactor shutdown time satisfies the UFSAR Section 5.4.7.1.1.1 time criterion of 20 hours, which was selected based on engineering judgment to ensure that SDC operation effect on a normal reactor shutdown schedule is minimized. The RHR HXs were sized to provide the heat transfer capability required for the SPC mode. Therefore, excess capacity exists for the SDC mode. The increase in the normal reactor shutdown time for EPU could affect outage schedules and plant availability. However, this has no effect on plant safety or the design operating margins and no change to the RHR system is required. Additionally, for EPU, the SDC analysis shows that the

reactor can be cooled to 212°F in 8.2 hours at EPU conditions using one loop of RHR. Therefore, the SDC mode meets all CLTR dispositions. 2.8.4.4.4 Fuel Pool Cooling Assist The CLTR states that the SFP heat load increases due to the decay heat generation as a result of the EPU. The FPC Assist (SFP Cooling) mode, using existing RHR system heat removal capacity, provides supplemental fuel pool cooling cap ability in the event that the fuel pool heat load exceeds the heat removal capability of the FPCCS. The adequacy of fuel pool cooling, including use of the SFP Cooling mode, is discussed in Section 2.5.3.1, which confirms that EPU does not affect the ability of the FPCC to perform this function. Therefore, the FPC Assist mode meets all CLTR dispositions. 2.8.4.4.5 Feedwater Leakage Control System The CLTR does not specifically address the FWLC function. The FWLC system is designed to minimize the release of fission products that could bypass the SGTS during a LOCA. The RHR jockey (keep fill) pumps can be used to fill the FW line volume between the containment isolation valves. Analysis results (Section 2.6.5) show that peak containment temperature and pressure are within RHR design limits following a LOCA with no change required to the RHR system. EPU does not change the RHR system capability to support the FWLC function.

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2-334 Conclusion The effects of the proposed EPU on the RHR system have been reviewed. The results of that review demonstrate that the RHR system will maintain its ability to cool the RCS following shutdown and provide DHR after the proposed EPU. Based on this, Entergy concludes that the RHR system will continue to meet the requirements of GDCs 4 and 34 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the RHR system. 2.8.4.5 Standby Liquid Control System Regulatory Evaluation The SLCS provides backup capability for reactivity control independent of the control rod system. The SLCS functions by injecting a boron solution into the reactor to effect shutdown. The effect of the proposed EPU on the functional capability of the system to deliver the required amount of boron solution into the reactor has b een reviewed. The regulatory acceptance criteria are based on: (1) GDC-26, insofar as it requires that two independent reactivity control systems of different design principles be provided, and that one of the systems be capable of holding the

reactor subcritical in the cold condition; (2) GDC-27, insofar as it requires that the reactivity control systems have a combined capability, in conjunction with poison addition by the ECCS, to reliably control reactivity changes under postulated accident conditions; and (3) 10 CFR 50.62(c)(4), insofar as it requires that the SLCS be capable of reliably injecting a borated water solution into the RPV at a boron concentration, boron enrichment, and flow rate that provides a set level of reactivity control. As stated in GGNS TS Bases B3.1.7, the SLCS is manually initiated from the main control room, as directed by the emergency operating procedures (EOPs), if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The SLCS is (are) described in UFSAR Section 9.3.5.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 6.5 of the CLTR addr esses the effect of CPPU on SLCS. The results of this evaluation are described below. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-335 The SLCS is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that some or all of the control rods cannot be inserted. This manually operated system pumps a sodium pentaborate solution into the vessel to provide neutron absorption and achieve a subcritical reactor cond ition. GGNS intends to load an enriched boron solution during the Cycle 18 RFO; however, that modification was conservatively not considered as part of the EPU. SLCS is designed to inject over a wide range of reactor operating pressures. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Core Shutdown Margin [[ Meets CLTR Disposition System Performance and Hardware

Meets CLTR Disposition Suppression Pool Temperature Following Limiting ATWS Event

     ]] Meets CLTR Disposition 2.8.4.5.1 Core Shutdown Margin The CLTR states that the ability of the SLCS boron solution to achieve and maintain safe shutdown is not a direct function of core thermal power, and therefore, is not affected by EPU. SLCS shutdown capability (in terms of the required reactor boron concentration) is reevaluated

for each fuel reload. The EPU evaluations assumed GNF2 fuel. The shutdown concentration of 660 ppm of naturally enriched boron does not change for EPU. No changes are necessary to the solution volume / concentration or to the boron-10 enrichment for EPU to achieve the required reactor boron concentration for shutdown. GGNS is increasing the enrichment of the solution in order to maintain margin for core design. This change will also enable a reduction in the boron concentration as described in Section 2.8.5.7. Therefore, the SLCS shutdown margin capability meets all CLTR dispositions. 2.8.4.5.2 System Performance and Hardware

As stated in Section 6.5 of the CLTR, the effect of EPU on system performance and hardware is increased heat load and potential increase in tran sient reactor pressure. The SLCS is designed for injection at a maximum reactor pressure e qual to the upper AV for the lowest group of SRVs operating in the safety relief mode. At GGNS, the nominal reactor dome pressure and the SRV setpoints are unchanged for EPU. Consequentl y, the capability of the GGNS SLCS to provide NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-336 its backup shutdown function is not affected by EPU. The SLCS is not dependent upon any other SRV operating modes. Based on the results of the GGNS EPU ATWS analysis, the maximum reactor upper plenum pressure following the limiting ATWS event reaches 1190.3 psig (1205 psia) during the time the SLCS is analyzed to be in operation. Conse quently, there is a corresponding increase in the maximum pump discharge pressure to 1337.3 psig and a decrease in the operating pressure margin for the pump discharge relief valves. 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 setpoint for the pump discharge relief valves. The relief valve setpoint margin is 305.9 psi. This margin is based on a SLCS pump relief valve setpoint of 1643.2 psig (cold differential set pressure of 1694.0 psig minus 3% tolerance). The pump discharge relief valves are periodically tested to confirm the valves function within this tolerance. The operation of the pump discharge system was analyzed to confirm that the loss of flow through an open relief valve would not compromise the required boron injection function. The evaluation compared the open/close setpoint of the pump discharge relief valves with the calculated maximum SLCS pump discharge pressure expected during the most limiting ATWS transient. It was confirmed that the SLCS relief valves would not open even if system initiation were to occur prior to reactor pressure recovering from the initial transient peak. Therefore, the current SLCS process parameters associated with the minimum boron injection rate are not changed. The SLCS ATWS performance is evaluated in Section 2.8.5.7 for a representative core design for EPU. The evaluation confirmed acceptable results and demonstrates that EPU has no adverse effect on the ability of the SLCS to mitigate an ATWS. Therefore, GGNS SLCS performance and hardware meet all CLTR dispositions. 2.8.4.5.3 Suppression Pool Temperature Following ATWS Event

As stated in Section 6.5 of the CLTR, changes in the fuel design for EPU may require modifications to the SLCS as a result of the increase in the SP temperature for the limiting ATWS event. The boron injection rate requirement for maintaining the peak SP water temperature limits, following the limiting ATWS event with SLCS injec tion, is not increased for EPU. Therefore, the SP temperature following an ATWS event at GGNS meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the SLCS have been reviewed and it was found the SLCS adequately accounts for the EPU. It was demonstrated that the system will continue to provide the function of reactivity control independent of the CRD system following implementation of the proposed EPU. Based on this, Entergy concl udes that the SLCS will continue to meet the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-337 requirements of GDCs 26 and 27, and 10 CFR 50.62(c)(4) following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the SLCS. 2.8.4.6 Reactor Recirculation System Performance NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 3.6 of the CLTR a ddresses the effect of CPPU on the RRS. The results of this evaluation are described below. RRS performance is not specifically addressed in NRC "Review Standard for Extended Power Uprates," RS-001. The EPU power condition is accomplished by operati ng along extensions of current rod lines on the power/flow map with no increase in the maximum core flow. The core reload analyses are performed with the most conservative allowa ble core flow. The evaluation of the RRS performance at EPU power determines that adequate core flow can be maintained. The cavitation protection interlock remains the same in terms of absolute flow rates. This interlock is based on subcooling in the external recirculation loop and thus is a function of absolute FW flow rate and FW temperature at less than full thermal power operating conditions.

Therefore, the interlock is not changed by EPU. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Net Positive Suction Head [[ Meets CLTR Disposition Flow Mismatch

Meets CLTR Disposition Single Loop Operation

     ]] Meets CLTR Disposition 2.8.4.6.1 Net Positive Suction Head The CLTR states that increased voids in the core during normal uprated power operation requires a slight increase in the recirculation drive flow to achieve the same core flow.

The CLTR shows that recirculation pump N PSH at full EPU power does not significantly increase the NPSH R or significantly reduce the NPSH margin. There is no change in maximum core flow. The maximum core flow (105% rated core flow) at CLTP and EPU is 118.1 Mlb/hr. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-338 Based on past uprate analyses, the NPSH R at full power does not significantly increase or reduce the NPSH margin because the required increase in recirculation flow is small. Therefore, the effects of EPU on NPSH meets all CLTR dispositions. 2.8.4.6.2 Flow Mismatch The GGNS recirculation loop jet pump flow mismatch TS limits do not change because these limits are based on rated core flow, which is not affected by EPU, and the flow mismatch limits are not affected because a detailed ECCS evaluation was not required for GGNS at EPU conditions. Section 4.3 of the CLTR summarizes the effects of EPU on the ECCS analysis, and confirms the limited ECCS evaluation that is required to support EPU. EPU does not result in a change to the core flow. Therefore, the effects of EPU on flow mismatch from recirculation pumps or loops meet all CLTR dispositions. 2.8.4.6.3 Single Loop Operation The CLTR states that increased voids in the core during normal uprated power operation requires a slight increase in the recirculation drive flow to achieve the same core flow. SLO is limited to off-rated conditions and is not affected by EPU. SLO operation at GGNS is restricted to a reactor power of 2,705 MWt and a flow of 60.9 Mlb/hr.

[[

                                                                                                                                          ]]  The absolute power limit for SLO stays the same, requiring a proportiona l reduction in the percent of rated power at the uprate power level (i.e., 69.4% CLTP or 61.4% EPU power). Therefore, the effects of EPU on SLO meet all CLTR dispositions.

2.8.5 Accident

and Transient Analyses 2.8.5.1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Main Steam Relief or Safety Valve Regulatory Evaluation Excessive heat removal causes a decrease in moderator temperature, which increases core

reactivity and can lead to a power level increase and a decrease in shutdown margin. Any unplanned power level increase may result in fuel damage or excessive reactor system pressure. Reactor protection and safety systems are actuated to mitigate the transient. A review has been NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-339 performed of the effects of the proposed EPU on: (1) postulated initial core and reactor conditions; (2) methods of thermal and hydrau lic analyses; (3) the sequence of events; (4) assumed reactions of reactor system components; (5) functional and operational characteristics of the RPS; (6) operator actions; and (7) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin suffici ent to ensure that the design condition of the RCPB are not exceeded during any condition of normal operation; (3) GDC-20, insofar as it requires that the RPS be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to en sure that SAFDLs are not exceeded during any condition of normal operation, including AOOs; and (4) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The events resulting in a decrease in reactor coolant temperature are described in UFSAR Section 15.1.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the eff ects of EPUs. Although ELTR1 is the licensing basis for the GGNS AOO events, for the events disc ussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-5 is consistent with the sensitivity that established the scope approved in the CLTR. [[

                                ]]  The following is a summary of the ev aluation provided for the excessive heat removal events:

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-340 Consistent with the CLTR, the Decrease in Feedwater Temperature limiting event (LFWH with manual flow control) and the Increase in Feedwater Flow limiting event (Feedwater Controller Failure Maximum Demand (FWCF)) are confirmed to be within the GGNS reload evaluation scope. The LFWH event is performed with the NRC approved methods described in GESTAR II (Reference 5). The computer code used to evaluate the LFWH event is PANACEA. The transient evaluation initial conditions are provide d in Table 2.8-4, and the results of the EPU evaluations are reported in Table 2.8-5. The Increase in Steam Flow event and the Inadvertent Opening of a Safety Relief Valve event are not listed in [[

     ]] to be analyzed for EPU. The Increase in Steam Flow event ([[                                                                                                                 
                        ]]) is [[

     ]]. This event results [[

     ]] The Inadvertent Opening of a Safety Relief Valve event is [[

     ]] Conclusion The analyses of the excess heat removal events described above have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. Based on these analyses, it has been demonstrated that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, 20, and 26 following implementation of the propos ed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the events stated.

2.8.5.2 Decrease in Heat Removal by the Secondary System 2.8.5.2.1 Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) Regulatory Evaluation A number of initiating events may result in unplanned decreases in heat removal by the secondary system. These events result in a sudden reduction in steam flow and, consequently, result in pressurization events. Reactor protection and safety systems are actuated to mitigate the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-341 transient. A review has been performed of th e effects of the proposed EPU on the sequence of events, the analytical models used for analyses, the values of parameters used in the analytical models, and the results of the transient analys es. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditi ons of the RCPB are not exceeded during any condition of normal operation; and (3) GDC-26, inso far as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The events that may result in an increase in reactor pressure are described in UFSAR Section 15.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the eff ects of EPUs. Although ELTR1 is the licensing basis for the GGNS AOO events, for the events disc ussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-5 is consistent with the sensitivity that established the scope approved in the CLTR. [[

                                ]]  The following is a summary of the evaluation provided for the decreased heat removal events:

Consistent with the CLTR, the Loss of External Load limiting event (Generator Load Rejection with Steam Bypass Failure (LRNBP)) and the TT limiting event (Turbine Trip with Steam Bypass Failure (TTNBP)) are confirmed to be w ithin the GGNS reload evaluation scope. The transient evaluation initial conditions are provide d in Table 2.8-4, and the results of the EPU evaluations are reported in Table 2.8-5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-342 The Closure of Main Steam Isolation Valves with Direct Scram (MSIVD) event is not within the GGNS reload evaluation scope. The transient evaluation initial conditions are provided in Table 2.8-4, and the results of the EPU evaluations are reported in Table 2.8-5. Consistent with the CLTR, the Pressure Regulator Failure Downscale (PRFD) event is confirmed to be within the GGNS reload evaluation scope. The transient evaluation initial conditions are provided in Table 2.8-4, and the results of the EPU evaluation are reported in Table 2.8-5. The Loss of Condenser Vacuum (LOCV) event for GEH BWRs is also [[

     ]] MSIVF for GGNS is evaluated in Section 2.8.4.2.

Consistent with Limitations and Conditions 9.9 and 9.11 of Reference 7, acceptable fuel rod thermal-mechanical performance for both UO 2 and GdO 2 fuel rods was demonstrated. Results for all AOO pressurization transient events analyzed, including equipment OOS, showed at least 10 percent margin to the fuel centerline melt and at least 10 percent margin to the one percent cladding circumferential plastic strain acceptance criteria. The minimum calculated margin to the fuel centerline melt criterion was 52.4%. The minimum calculated margin to the cladding strain criterion was 51.5%. Fuel rod thermal-mechanical performance will be evaluated as part of the RLA performed for the cycle-specific core. Documentation of acceptable fuel rod thermal-mechanical response will be included in the SRLR or COLR consistent with Limitation and Condition 9.10 of Reference 7. Conclusion The analyses of the decrease in heat removal (i.e., an increase in reactor pressure) events described above have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The results of those analyses demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, and 26 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the events stated. 2.8.5.2.2 Loss of Non-Emergency AC Power to the Station Auxiliaries Regulatory Evaluation The loss of non-emergency AC power is assumed to result in the loss of all power to the station auxiliaries and the simultaneous tripping of all reactor coolant circulation pumps. This causes a NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-343 flow coastdown as well as a decrease in heat removal by the secondary system, a TT, an increase in pressure and temperature of the coolant, a nd a reactor trip. Reactor protection and safety systems are actuated to mitigate the transient. A review has been performed of the effects of the

proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal

operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin suffici ent to ensure that the design conditions of the RCPB are not exceeded during any condition of normal operation; and (3) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the

rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The loss of AC power event is described in UFSAR Section 15.2.6.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. [[

                                ]]  The Loss of Non-Emergency AC Power to the Station Auxiliaries event is not listed in [[                                          ]] to be analyzed for EPU. The following is a summary of the evaluation provided for the Loss of Non-Emergency AC Power to the Station Auxiliaries event:

Consistent with ELTR1, the Loss of Non-Emergency AC Power to the Station Auxiliaries event is [[

]]

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-344 Conclusion The analysis of the loss of non-emergency AC power to station auxiliaries event has been reviewed to ensure it adequately accounted for ope ration of the plant at the proposed power level and was performed using acceptable analytical models. The results of that analysis demonstrate that the reactor protection and safety systems w ill continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a re sult of this event. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, and 26 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the loss of non-emerge ncy AC power to station auxiliaries event. 2.8.5.2.3 Loss of Normal Feedwater Flow Regulatory Evaluation A loss of normal FW flow could occur from pump failures, valve malfunctions, or a LOOP. LOFW flow results in an increase in reactor coolant temperature and pressure, and eventually requires a reactor trip to prevent fuel damage. Decay heat must be transferred from fuel following a loss of normal FW flow. Reactor protection and safety systems are actuated to provide this function and mitigate other aspects of the transient. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance cr iteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during any condition of normal operation; and (3) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The loss of normal FW flow event is described in UFSAR Section 15.2.7.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-345 effects of CPPUs. Section 9.1.3 of the CLTR a ddresses the effect of CPPU on Loss of Water Level Events. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Loss of Water Level Events (Loss of Feedwater Flow) [[ Meets CLTR Disposition Loss of Water Level Events (Loss of One Feedwater Pump)

     ]] Meets CLTR Disposition 2.8.5.2.3.1 Loss of Feedwater Flow Event As stated in the CLTR, higher decay heat results in a lower reactor water level for loss of water level events. For the LOFW event, adequate transient core cooling is provided by maintaining the water level inside the core shroud above the TAF. A plant-specific analysis was performed for GGNS at EPU conditions. This analysis assumed failure of the HPCS system and used only the RCIC system to restore the reactor water level. Because of the extra decay heat from EPU, slightly more time is required for the automatic systems to restore water level. Operator action is only needed for long-term plant shutdown. The results of the LOFW analysis for GGNS show that the minimum water level inside the shroud is 50 inches above the TAF at EPU conditions. After the water level is restored, the operator manually controls the water level, reduces reactor pressure, and initiates RHR SDC.

This sequence of events does not require any ne w operator actions or shorter operator response times. Therefore, the operator actions for an LOFW transient do not significantly change for EPU. As described in Table 1-1, for Transient Analysis, the modeling tool used is the SAFER04 model, which is the same model used in the ECCS LOCA analysis. The analysis is done consistent with the CLTR. The following is the ge neral sequence of events in the analysis. The reactor is assumed to be at 102% of the EPU power level when the LOFW occurs. The initial level in the model is conservatively set at the low-level scram setpoint and RFW is instantaneously isolated at event initiation. Scram is initiated at the start of the event. The RCIC system is initiated when the level decreases to the low-low level. The MSIVC initiates when the level decreases to low-low-low level. The RCIC flow to the vessel begins at 60 seconds into the event, minimum level is reached at 622 seconds and level is recovered after that point. Only NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-346 RCIC flow is credited to recover the reactor water level. There are no additional failures assumed beyond the failure of the HPCS system. The only other key analysis assumption for the LOFW analysis, discussed in Section 9.1.3 of the CLTR, was the assumed decay heat level of ANS 5.1-1979 with a two-sigma uncertainty. The assumed decay heat level for the EPU anal ysis was ANS 5.1-1979 decay heat +10%, which bounds ANS 5.1-1979 + two sigma. Thus, the key analytical assumptions are the same or conservative relative to the current licensing basis. This LOFW analysis is performed to demonstrate acceptable RCIC system performance. The design basis criterion for the RCIC system is confirmed by demonstrating that it is capable of maintaining the water level inside the shroud above the TAF during the LOFW transient. The minimum level (see Figure 2.8-21) is maintained at least 50 inches above the TAF, thereby demonstrating acceptable RCIC system performance. There are no applicable equipment OOS assumptions for this transient. As discussed in Section 2.8.4.3, an operational requirement is that the RCIC system restores the reactor water level while avoiding ADS timer initiation and MSIVC activation functions associated with the low-low-low reactor water level setpoint (Level 1). This requirement is intended to avoid unnecessary initiations of safety systems. This requirement is not a safety-related function. The results of the LOFW analysis for GGNS show that the nominal Level 1 setpoint trip is avoided. Therefore, the LOFW event meets all CLTR dispositions. 2.8.5.2.3.2 Loss of One Feedwater Pump As stated in the CLTR, higher decay heat results in a lower reactor water level for loss of water level events. The Loss of One Feedwater Pump event was included in ELTR1 only for operational considerations. As stated in the NRC SER, S ection 4.5, to ELTR2, "A plant-specific analysis of the loss of one feedwater pump event will be submitted per Appendix E of ELTR1 to assess the effect of a higher flow control line on scram avoidance." The Loss of One Feedwater Pump event only a ddresses operational considerations to avoid reactor scram on low reactor water level (Level 3). This requirement is intended to avoid unnecessary reactor shutdowns. Because the MELLLA region is extended along the existing upper boundary to the EPU RTP, there is no increase in the highest flow control line for the GGNS EPU. Therefore, the Loss of One Feedwater Pump event meets all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-347 Conclusion The analysis of the Loss of Normal Feedwater Flow event has been reviewed to ensure it adequately accounted for operation of the plant at the proposed power level and was performed using acceptable analytical models. The results of that analysis demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a result of the loss of normal FW flow. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, and 26 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the Loss of Normal Feedwater Flow event. 2.8.5.3 Decrease in Reactor Coolant System Flow 2.8.5.3.1 Loss of Forced Reactor Coolant Flow Regulatory Evaluation A decrease in reactor coolant fl ow occurring while the plant is at power could result in a degradation of core heat transfer. An increase in fuel temperature and accompanying fuel damage could then result if SAFDLs are exceeded during the transient. Reactor protection and safety systems are actuated to mitigate the transient. A review has been performed of the effects of the proposed EPU on: (1) the postulated initial core and reactor conditions; (2) the methods of thermal and hydraulic analyses; (3) the sequence of events; (4) assumed reactions of reactor systems components; (5) the functional and opera tional characteristics of the RPS; (6) operator actions; and (7) the results of th e transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to

ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditi ons of the RCPB are not exceeded during any condition of normal operation; and (3) GDC-26, inso far as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analyses of the RPT event and the recirculation flow control failure events resulting in a loss of forced reactor coolant flow are descri bed in UFSAR Sections 15.3.1 and 15.3.2, respectively. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-348 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. [[

                                ]]  The Loss of Forced Reactor Cool ant Flow event is not listed in [[                         
                  ]] to be analyzed for EPU. The following is a summary of the evaluation provided for the Loss of Forced Reactor Coolant Flow event:

The Loss of Forced Reactor Coolant Flow event, including the Trip of Pump Motor and Flow Controller Malfunction events, results in a decreas e in reactor core coolant flow rate. These events are [[

     ]] Conclusion The analyses of the decrease in reactor coolant fl ow event have been revi ewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a re sult of this event. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, and 26 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the decrease in reactor coolant flow event.

2.8.5.3.2 Reactor Recirculation Pump Rotor Seizure and Reactor Recirculation Pump Shaft Break Regulatory Evaluation The events postulated are an instantaneous seizure of the rotor or break of the shaft of a reactor recirculation pump. Flow through the affected l oop is rapidly reduced, leading to a reactor and

turbine trip. The sudden decrease in core coolant flow while the reactor is at power results in a degradation of core heat transfer that could result in fuel damage. The initial rate of reduction of coolant flow is greater for the rotor seizure event. However, the shaft break event permits a

greater reverse flow through the affected loop late r during the transient and, therefore, results in NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-349 a lower core flow rate at that time. In both events, reactor protection and safety systems are actuated to mitigate the transient. A review has been performed of the effects of the proposed EPU on: (1) the postulated initial and long-term core and reactor conditions; (2) the methods of thermal and hydraulic analyses; (3) the sequence of events; (4) the assumed reactions of reactor system components; (5) the functional and opera tional characteristics of the RPS; (6) operator actions; and (7) the results of th e transient analyses. The regulatory acceptance criteria are based on: (1) GDC-27, insofar as it requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; (2) GDC-28, insofar as it requires that the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core; and (3) GDC-31, insofar as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a non-brittle manner and the probability of a rapidly propagating fracture is minimized. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analyses of the recirculation pump seizure event and the recirculation pump shaft break event are described in UFSAR Sections 15.3.3 and 15.3.4, respectively. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. [[

                                                                                                                   ]]  The following is a summary of the evaluation provided for the Reactor Recirculation Pump Rotor Seizure and Reactor Recirculation Pump Shaft Break events:

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2-350 The Reactor Recirculation Pump Rotor Seizure even t results in a decrease in reactor core coolant flow rate. Events in this category, [[

                     ]]  Therefore, the Reactor Recirculation Pump Rotor Seizure event is not analyzed for EPU. The Reactor Recirculation Pump Shaft Break event results in a decrease in reactor core coolant flow rate. Events in this category, [[                                                                                                                   
                                                      ]]  Therefore, the Reactor Recirculation Pump Shaft Break event is not analyzed for EPU.

Conclusion The analyses of the sudden decrease in core coolan t flow events have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensu re that the ability to insert control rods is maintained, the RCPB pressure limits will not be exceeded, the RCPB will behave in a non-brittle manner, the probability of propagating fracture of the RCPB is minimized, and adequate core cooling will be provided. Based on this, Ente rgy concludes that the plant will continue to meet the requirements of GDCs 27, 28, and 31 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU accepta ble with respect to the sudden decrease in

core coolant flow events. 2.8.5.4 Reactivity and Power Distribution Anomalies 2.8.5.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition Regulatory Evaluation An uncontrolled control rod assembly withdrawal from subcritical or low power startup conditions may be caused by a malfunction of the reactor control or rod control systems. This withdrawal will uncontrollably add positive reactivity to the reactor core, resulting in a power excursion. A review has been performed of the effects of the proposed EPU on: (1) the description of the causes of the transient and the transient itself; (2) the initial conditions; (3) the values of reactor parameters used in the analysis; (4) the analytical methods and computer codes used; and (5) the results of the transient anal yses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to

ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-20, insofar as it requires that the RPS be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that SAFDLs are not NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-351 exceeded as a result of AOOs; and (3) GDC-25, insofar as it requires that the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analysis of the low power RWE event is described in UFSAR Section 15.4.1.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 5.1.2 of the CLTR addresses the e ffect of CPPU on Uncontrolled Control Rod Assembly Withdrawal from a Subc ritical or Low Power Startup Condition. The results of this evaluation are described below. The evaluation of the Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition event for the GGNS EPU is a comparison of the expected maximum increase in peak fuel enthalpy with the acceptance criterion of 170 cal/gram. The CLTP Uncontrolled Control Rod Assembly W ithdrawal analysis for GGNS is based on Reference 80. The GGNS EPU core consists only of GE fuel assemblies and the EPU is limited to 115% of OLTP. There is no change to the reactor manual control system or control rod HCUs for EPU. The RCIS installed at GGNS provides the same level of protection for GNF2 fuel following EPU provided the power increase is 20% and BPWS is used at power levels below the lower LPSP AV. The evaluation of this even t for the GGNS EPU considering these features and GNF2 fuel demonstrates the CLTR disposition is applicable. No change in peak fuel enthalpy is expected due to EPU because an RWE is a localized low-power event. If the peak fuel rod enthalpy is conservatively assumed to increase by a factor of 1.2, the RWE peak fuel enthalpy at EPU will be 72 cal/gram. This en thalpy is well below the acceptance criterion of 170 cal/gram.

Conclusion The analysis of the uncontrolled control rod assembly withdrawal from a subcritical or low power startup condition was reviewed to ensure it has adequately accounted for the changes in core design necessary for operation of the plant at the proposed power level. The review also confirmed the analysis was performed using acceptable analytical models. The analysis demonstrates that the reactor protection and safety systems will continue to ensure the SAFDLs are not exceeded. Based on this, Entergy concludes that the plant will continue to meet the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-352 requirements of GDCs 10, 20, and 25 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU accepta ble with respect to the uncontrolled control rod assembly withdrawal from a subc ritical or low power startup condition. 2.8.5.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power Regulatory Evaluation An uncontrolled control rod assembly withdrawal at power may be caused by a malfunction of the reactor control or rod control systems. This withdrawal will uncontrollably add positive reactivity to the reactor core, resulting in a power excursion. A review has been performed of the effects of the proposed EPU on: (1) the description of the causes of the AOO and the description of the event itself; (2) the initial conditions; (3) the values of reactor parameters used in the analysis; (4) the analytical methods and computer codes used; and (5) the results of the associated analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-20, insofar as it requires that the RPS be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that SA FDLs are not exceeded as a result of AOOs; and (3) GDC-25, insofar as it requires that the protection system be designed to assure that SAFDLs are not exceeded for any single malfunction of the reactivity control systems. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analysis of the RWE at power event is described in UFSAR Section 15.4.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the eff ects of EPUs. Although ELTR1 is the licensing basis for the GGNS AOO events, for the events disc ussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-5 is consistent with the sensitivity that established the scope approved in the CLTR. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-353 [[

                                ]]  The following is a summary of the evaluation provided for the Uncontrolled Control Rod Assembly Withdrawal at Power event.

Consistent with the CLTR, the Uncontrolled Control Rod Assembly Withdrawal at Power (RWE) event is confirmed to be within the GGNS reload evaluation scope. The RWE event is performed with the NRC approved methods de scribed in GESTAR II (Reference 5). The computer code used to evaluate the RWE even t is PANACEA. The transient evaluation initial conditions are provided in Table 2.8-4, and the re sults of the EPU evaluation are reported in Table 2.8-5.

Conclusion The analysis of the uncontrolled control rod assembly withdrawal at power event has been reviewed to ensure that the analysis has ade quately accounted for the changes in core design required for operation of the plant at the proposed power level and was performed using acceptable analytical models. The analysis demons trates that the reactor protection and safety systems will continue to ensure the SAFDLs ar e not exceeded. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 20, and 25 following implementation of the proposed EPU. Therefor e, Entergy finds the proposed EPU acceptable with respect to the Uncontrolled Control Rod Assembly Withdrawal at Power event. 2.8.5.4.3 Startup of a Recirculation Loop at an Incorrect Temperature and Flow Controller Malfunction Causing an Increase in Core Flow Rate Regulatory Evaluation A startup of an inactive loop transient may result in either an ICF or the introduction of cooler water into the core. This event causes an increase in core reactivity due to decreased moderator temperature and core void fraction. A review has been performed of the effects of the proposed

EPU on: (1) the sequence of events; (2) the analytical model; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to assure that SAFDLs are not exceeded during any condition of normal operation, including the effects of AOOs; (2) GDC-20, insofar as it requires that the protection system be designed to initiate automatically the operation of appropriate systems to ensure that

SAFDLs are not exceeded as a result of operational occurrences; (3) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design condition of the RCPB are not exceeded during AOOs; (4) GDC-28, insofar as it requires that the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the co re; and (5) GDC-26, insofar as it requires that a NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-354 reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analyses of the abnormal startup of an idle recirculation loop event and the recirculation flow controller failure - flow increase even t are described in UFSAR Sections 15.4.4 and 15.4.5, respectively.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the eff ects of EPUs. Although ELTR1 is the licensing basis for the GGNS AOO events, for the events disc ussed in this section, the sensitivity of the CPR changes provided as part of the transient analysis results in Table 2.8-5 is consistent with the sensitivity that established the scope approved in the CLTR. [[

                                ]]  The following is a summary of the ev aluation provided for the Startup of a Recirculation Loop at an Incorrect Temperature and the Flow Controller Malfunction Causing an Increase in Core Flow Rate events:

Consistent with the CLTR, [[

     ]] The Failure of the Recirculation Flow Controller can result in either a slow or fast recirculation increase. The disposition of these events for EPU indicates that [[

                                                                                                                              ]]  The transient evaluation initial conditions are provided in Table 2.8-4, and th e results of the EPU evaluations are reported in Table 2.8-5.

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2-355 Per NUREG-0800 Section 15.4.4-15.4.5, "Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate," Revision 2, March 2007 (Reference 81), for a BWR, "Startup of a Recirculation Loop at an Incorrect Temperature event" is called the "Startup of an Idle Recirculation Pump" event. Conclusion The analyses of the increase in core flow even ts have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensure that th e SAFDLs and the RCPB pressure limits will not be exceeded as a result of these events. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, 20, 26, and 28 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the increase in core flow events. 2.8.5.4.4 Spectrum of Rod Drop Accidents Regulatory Evaluation Another type of reactivity or power distribution anomaly event is the CRDA. A review has been performed of the effects of the proposed EPU on the occurrences that lead to the accident, safety features designed to limit the amount of reactivity available and the rate at which reactivity can be added to the core, the analytical model used fo r analyses, and the results of the analyses. The regulatory acceptance criteria are based on GDC-28, insofar as it requires that the reactivity control systems be designed to assure that th e effects of postulated reactivity accidents can neither result in damage to the RCPB greater than limited local yielding, nor disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analysis of the control rod drop event is described in UFSAR Section 15.4.9.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-356 effects of CPPUs. Section 5.1.2 and 5.3.4 of the CLTR addresses the effect of CPPU on the RCIS. The results of this evaluation are described below. The spectrum of CRDAs does not change with EP U. The evaluation of a CRDA for the GGNS EPU is a comparison of the expected maximum incr ease in peak fuel enthalpy with the acceptance criterion of 280 cal/gram. The CLTP CRDA for GGNS is based on Reference 82. The GGNS EPU core consists only of GE fuel assemblies and the EPU is limited to 115% of OLTP. Control Rod Sequencing at GGNS for CLTP and EPU follows the BPWS. There is no change to the GGNS reactor manual control system or control rod HCUs for EPU. The RCIS in stalled at GGNS provides the same level of protection for GNF2 fuel following EPU provided the power increase is 20% and BPWS is used at power levels below the lower LPSP AV. The evaluation of this event for the GGNS EPU considering these features and GNF2 fuel demonstrates the CLTR disposition is

applicable. No change in peak fuel enthalpy is expected due to EPU because with the rod drop accident is a limiting localized low-power event. As noted in UFSAR Section 4.3.2.3, the CRDA is inherently self-limiting for core powers above 10% of CLTP. If the peak fuel rod enthalpy is conservatively assumed to increase by a factor of 1.2, the CRDA peak fuel enthalpy at EPU will be 162 cal/gram. This enthalpy is well below the acceptance criterion of 280 cal/gram. Conclusion The analyses of the rod drop accident have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. The analyses dem onstrate that appropriate reactor protection and safety systems will prevent postulated reactivity accidents that could (1) result in damage to the RCPB greater than limited local yielding, or (2) cause sufficient damage that would significantly impair the capability to cool the core. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDC-28 following implementation of EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the rod drop accident.

2.8.5.5 Inadvertent Operation of ECCS or Malf unction that Increases Reactor Coolant Inventory Regulatory Evaluation Equipment malfunctions, operator errors, and abnormal occurrences could cause unplanned increases in reactor coolant inventory. Depending on the temperature of the injected water and the response of the automatic control systems, a power level increase may result and, without adequate controls, could lead to fuel damage or overpressurization of the RCS. Alternatively, a power level decrease and depressurization may result. Reactor protection and safety systems are actuated to mitigate these events. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-357 designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, insofar as it requires that the RCS and its associated auxiliary systems be designed with margin suffici ent to ensure that the design conditions of the RCPB are not exceeded during AOOs; and (3) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analyses of the events that could result in an increase in reactor coolant inventory are described in UFSAR Section 15.5.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because GGNS is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. [[

                                ]]  The following is a summary of the evaluation provided for the Inadvertent Operation of ECCS or Malfunction that In creases Reactor Coolant Inventory events:

Consistent with the CLTR, the Inadvertent Oper ation of ECCS or Malfunction that Increases Reactor Coolant Inventory events (the Inadvertent HPCS System Start) is not within the GGNS reload evaluation scope.

[[

     ]] Conclusion The analyses of the inadvertent operation of ECCS or malfunction that increases reactor coolant inventory have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and were performed using acceptable analytical models. These analyses demonstrate that the reactor protection and safety systems will continue to ensure that NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-358 the SAFDLs and the RCPB pressure limits will not be exceeded as a result of this event. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, and 26 following implementation of the proposed EP U. Therefore, Entergy finds the proposed EPU acceptable with respect to the Inadvertent Op eration of ECCS or Malfunction that Increases Reactor Coolant Inventory. 2.8.5.6 Decrease in Reactor Coolant Inventory 2.8.5.6.1 Inadvertent Opening of a Pressure Relief Valve Regulatory Evaluation The inadvertent opening of a pressure relief valve results in a reactor coolant inventory decrease and a decrease in RCS pressure. The pressure relief valve discharges into the SP. Normally there is no reactor trip. The pressure regulator senses the RCS pressure decrease and partially closes the TCVs to stabilize the reactor at a lo wer pressure. The reactor power settles out at nearly the initial power level. The coolant inventory is maintained by the FW control system using water from the CST via the condenser hotwell. A review has been performed of the effects of the proposed EPU on: (1) the sequence of events; (2) the analytical model used for analyses; (3) the values of parameters used in the analytical model; and (4) the results of the transient analyses. The regulatory acceptance criteria are based on: (1) GDC-10, insofar as it requires that the RCS be designed with appropriate margin to ensure that SAFDLs are not exceeded during normal operations, including AOOs; (2) GDC-15, inso far as it requires that the RCS and its associated auxiliary systems be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during AOOs; and (3) GDC-26, insofar as it requires that a reactivity control system be provided, and be capable of reliably controlling the rate of reactivity changes to ensure that under conditions of normal operation, including AOOs, SAFDLs are not exceeded.

GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analyses of the inadvertent safety / relief valve opening event and other events that could result in a decrease in reactor coolant inventory are described in UFSAR Section 15.6.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-359 products. Because GGNS is based on GNF2, this section will be based on ELTR1. NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. [[

                                       ]]  The Inadvertent Opening of a Pressure Relief Valve event is not listed in

[[ ]] to be analyzed for EPU. The following is a summary of the evaluation provided for the Inadvertent Opening of a Pressure Relief Valve event:

Consistent with ELTR1, the Inadvertent Op ening of a Safety Valve event is [[

     ]] Conclusion The analysis of the inadvertent opening of a pre ssure relief valve event has been reviewed to ensure it has adequately accounted for operation of the plant at the proposed power level and was performed using acceptable analytical models. This analysis demonstrates that the reactor protection and safety systems will continue to ensure that the SAFDLs and the RCPB pressure limits will not be exceeded as a re sult of this event. Based on this, Entergy concludes that the plant will continue to meet the requirements of GDCs 10, 15, and 26 following implementation of the proposed EPU. Therefore, Entergy fi nds the proposed EPU acceptable with respect to Inadvertent Opening of a Pressure Relief Valve event.

2.8.5.6.2 Emergency Core Cooling System and Loss-of-Coolant Accidents Regulatory Evaluation LOCAs are postulated accidents that would result in the loss of reactor coolant from piping breaks in the RCPB at a rate in excess of the capability of the normal reactor coolant makeup system to replenish it. Loss of significant qua ntities of reactor coolant would prevent heat removal from the reactor core, unless the water is replenished. The reactor protection and ECCS systems are provided to mitigate these accidents. A review has been performed of the effects of the proposed EPU on: (1) the determination of br eak locations and break sizes; (2) postulated initial conditions; (3) the sequence of events; (4) the analytical model used for analyses, and calculations of the reactor power, pressure, flow, and temperature transients; (5) calculations of PCT, total oxidation of the cladding, total hydrogen generation, changes in core geometry, and long-term cooling; (6) functional and operationa l characteristics of the reactor protection and ECCS systems; and (7) operator actions. Th e regulatory acceptance criteria are based on: (1) 10 CFR 50.46, insofar as it establishes standards for the calculation of ECCS performance

and acceptance criteria for that calculated performance; (2) 10 CFR 50, Appendix K, insofar as it NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-360 establishes required and acceptable features of evaluation models for heat removal by the ECCS after the blowdown phase of a LOCA; (3) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects associ ated with flow instab ilities and loads such as those resulting from water hammer; (4) GDC-27, inso far as it requires that the reactivity control systems be designed to have a combined capab ility, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained; and (5) GDC-35, insofar as it requires that a system to provide abundant emergency core cooling be provided to transfer heat from the reactor core following any LOCA at a rate so that fuel clad damage that could interfere with continued effective core cooling will be prevented. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The analyses of the LOCA are described in UFSAR Section 15.6.5.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Sections 4.2.2, 4.2.3, 4.2.4, 4.2.5, a nd 4.3 of the CLTR address the effect of CPPU on the ECCS and LOCAs. The results of this evaluation are described below. The ECCS include the HPCS system, the LPCS system, the LPCI mode of the RHR system, and the ADS.

The GGNS EPU LOCA analyses are based on NRC-approved GEH LOCA analysis methods and are in full compliance with 10 CFR 50.46. No new fuel designs are being introduced. No ECCS changes are required to meet LOCA analysis acceptance criteria. Each ECCS is discussed in the following subsections. The effect on the functional capability of each system due to EPU is addressed. [[

     ]] GGNS meets all CLTR dispositions. The topics addressed in this evaluation are:

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2-361 Topic CLTR Disposition GGNS Result High Pressure Core Spray [[ Meets CLTR Disposition Low Pressure Core Spray

Meets CLTR Disposition Low Pressure Coolant Injection System

Meets CLTR Disposition Automatic Depressurization

     ]] Meets CLTR Disposition 2.8.5.6.2.1 High Pressure Core Spray The CLTR states that there is no change to the normal reactor operating pressure or the SRV setpoints. The HPCS system is designed to spray water into the reactor vessel over a wide range of operating pressures and was evaluated in Section 4.3 of ELTR2. The HPCS system provides reactor vessel coolant inventory makeup in the event of a small break LOCA that does not immediately depressurize the reactor vessel and he lps to depressurize the reactor vessel. This system also provides spray cooling for long-term core cooling after a LOCA. The HPCS system also serves as a backup to the RCIC system to provide makeup water in the event of an LOFW flow transient, as descri bed in Section 2.8.5.2.3. Because the HPCS injection flow is greater than that of the RCIC system, which results in RPV depressurization, and there is no change to the range of pressures over which HP CS is required for injection, the adequacy of the HPCS system to meet the safety requirement following an LOFW event is demonstrated by the discussion in Section 2.8.5.2.3.

There is no change to the maximum specified reactor pressure for HPCS system operation and no change in the HPCS system performance parameters. The maximum injection pressure for the HPCS system is conservatively based on th e upper AL for the lowest available group of SRVs. Because the SRV settings and the normal reactor operating pressure remain the same for EPU, the HPCS system operating conditions and operating functions also remain the same. Therefore, there is no change in the original design pressures or temperatures for the system components. EPU does not change the power required by the pump or the power required by the HPCS diesel generator unit. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-362 Because the maximum normal operating pressure a nd the SRV setpoints do not change for EPU, the HPCS system performance requirements do not change. Therefore, the HPCS system at GGNS meets all CLTR dispositions. 2.8.5.6.2.2 Low Pressure Core Spray The CLTR states that there is no change in the reactor pressures at which the LPCS is required. The LPCS system is automatically initiated in the event of a LOCA. When operating in conjunction with other ECCS, the LPCS system is required to provide adequate core cooling for all LOCA events. There is no change in the reactor pressures at which the LPCS is required. The LPCS system sprays water into the reactor vessel after it is depressurized. The primary purpose of the LPCS system is to provide reactor vessel coolant inventory makeup for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. It also provides long-term core cooling in the event of a LOCA. The LPCS system meets all applicable safety criteria for the EPU. The slight change in the system operating cond ition due to EPU for a postulated LOCA does not affect the hardware capabilities of the LPCS system. The generic CS distribution assessment provided in ELTR2, Section 3.3, continues to be valid for EPU. Core spray distribution is not directly credited in the short-term cooling LOCA analyses. This is consistent with ECCS evaluation models specified in Appendix K to 10 CFR 50. Therefore, the convective heat transfer coefficients used during the short-term spray cooling period are the conservative values specified in Appendix K. The LPCS system at GGNS meets all CLTR dispositions because [[

     ]] 2.8.5.6.2.3 Low Pressure Coolant Injection The CLTR states that there is no change in th e reactor pressures at which the LPCI mode of RHR is required.

The LPCI mode of the RHR system is automatically initiated in the event of a LOCA. The primary purpose of the LPCI mode is to help maintain reactor vessel coolant inventory for a large break LOCA and for any small break LOCA after the reactor vessel has depressurized. The LPCI operating requirements are not affected by EPU and the ECCS performance evaluation demonstrates the adequacy of the LPCI core cooling performance. The LPCI mode of the RHR system at GGNS meets all CLTR dispositions. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-363 2.8.5.6.2.4 Automatic Depressurization System The CLTR states that EPU does not change the conditions at which the ADS must function. The ADS uses SRVs to reduce the reactor pressure following a small break LOCA when it is assumed that the high pressure systems have faile

d. This allows the LPCS and LPCI to inject coolant into the reactor vessel. EPU does not change the conditions at which the ADS must function. The ADS initiation logic and valve control is not affected by EPU conditions.

The adequacy of the ADS is demonstrated by the performance evaluation discussed in Section 2.8.5.6.2.5. The ADS at GGNS meets all CLTR dispositions because the SRV setpoints and functions remain the same, the ADS timers are not changed and the small break LOCA event mitigation is acceptable. 2.8.5.6.2.5 Emergency Core Cooling System Performance NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. However, the CLTR is limited to application of GE14 and earlier fuel products. Because the GGNS EPU is based on GNF2, this section will be based on ELTR1.

NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," Class III, February 1999 (also referred to as ELTR1) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. The GGNS ECCS is designed to provide prot ection against postulated LOCAs caused by ruptures in the primary system piping. The ECCS performance characteristics are not changed for EPU. The effects of EPU on ECCS-LOCA response is evaluated on a plant-specific basis. ECCS-LOCA performance analyses demonstrate that the 10 CFR 50.46 requirements continue to be met at the EPU RTP conditions. The basic break spectrum response is not affected by EPU. There are two limiting points on the break spectrum: the full sized RSLB and the worst small break under the HPCS-Diesel Generator failure scenario. Consistent with Limitation and Condition 9.7 of the Methods LTR SE (Reference 7), both top and mid-peaked power shapes were considered for both large and small break LOCA. [[

     ]]  The break spectrum response is determined by the ECCS network design and is common to all BWRs. GEH BWR power uprate evaluation experience shows that the basic break spectrum response is not affected by changes in core power. For SLO, a multiplier is applied to the Two-Loop LHGR and MAPLHGR Operation limits. The operating conditions for SLO are not changed with EPU; therefore, the current SLO analysis NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-364 remains acceptable for EPU. At EPU power condition, the MELLLA core flow extends to approximately 92.8% of rated core flow. Therefor e, the EPU analysis results at rated power and flow are applied to the MELLLA condition. Also, the effect of ICF on PC T is negligible with EPU. Thus the SLO, MELLLA, and ICF domain remain valid with EPU. The Licensing Basis PCT is based on the most limiting Appendix K case plus a plant variable uncertainty term that accounts statistically for the uncertainty in parameters that are not specifically addressed by 10 CFR 50 Appendix K. The Appendix K results demonstrate that the limiting LOCA is the DBA RSLB under the limiting single failure of HPCS-D/G. The EPU Licensing Basis PCT for GNF2 fuel is less than 1690F, which represents no change from the CLTP Licensing Basis PCT of less than 1690F evaluated at CLTP power and rated core flow. Restrictions imposed by the NRC on Upper Bound PCT have been removed for GGNS (Reference 83). The Upper Bound PCT has been shown to be bounded by the Licensing Basis PCT, consistent with the previous evaluation (References 84 and 85). The results of these analyses are provided in Table 2.8-7.

2.8.5.6.2.5.1 Large Break Peak Cladding Temperature - Limiting Case (ECCS-LOCA) The GGNS break spectrum response is determined by the ECCS network design that is common to all BWRs. The peak cladding temperature (PCT) for the limiting large break LOCA is determined primarily by the hot bundle power, wh ich is unchanged with EPU. In the GGNS analysis, the hot bundle is assumed to be operating at the thermal limits (MCPR, MAPLHGR, and LHGR); these limits are not changed for EPU. Comparison of the GGNS PCT results for CLTP and EPU indicate a minimal change (<20° F), and therefore large break LOCA has a negligible effect on compliance with the other acceptance criteria of 10 CFR 50.46 (local cladding oxidation, core-wide metal-water reaction, coolable geometry and long-term cooling).

The local fuel conditions are not significantly changed with EPU, because the hot bundle operation is still constrained by the same operating thermal limits. Because EPU has such a small effect on the GGNS large break PCT, the system response over the large break spectrum is

not affected. 2.8.5.6.2.5.2 Small Break PCT - Break Spectrum (ECCS-LOCA) The GGNS break spectrum response is determined by the ECCS network design that is common to all BWRs. For GGNS, the indicated decay heat for EPU is higher and results in a longer ADS blowdown and a higher PCT for the small break LOCA Appendix K case. Previous analyses (References 84 and 85) demonstrate that GGNS is a large break Appendix K PCT limited plant. The effect of EPU on the calculated small break PCT is acceptable as long as the effect of the results on the Licensing Basis PCT remains below the 10 CFR 50.46 limits. The current TS values for ECCS initiation were used for the anal ysis; no changes to these values were required for EPU. Plant-specific analyses demonstrate th at there is sufficient ADS capacity, with seven ADS valves in service and one OOS, at EPU conditions, to remain below these limits. Key input parameters to the SAFER/GESTR LOCA evaluation model are provided in Table 2.8-6. Input NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-365 parameters are selected as nominal or representative values. For Appendix K calculations, select inputs are chosen so as to set a bounding condition or to assure conservatism. 2.8.5.6.2.5.3 Local Cladding Oxidation (ECCS-LOCA) EPU has no effect on the Local Cladding Oxidation (ECCS LOCA). [[

                                                                          ]]  This conclusion on local cladding oxidation is confirmed by the unchanged plant Licensing Basis PCT due to EPU, which is sufficiently below the 10 CFR 50.46 limit of 2200°F. The bounding GGNS EPU Licensing Basis PCT of 1690°F ensures this 10 CFR 50.46 requirement is met.

2.8.5.6.2.5.4 Core-Wide Metal-Water Reaction (ECCS-LOCA) EPU has no effect on the Core-Wide Metal-Water Reaction (ECCS LOCA). [[

                                                                                                          ]]  This conclusion on the core-wide metal-water reaction is confirmed by the uncha nged plant Licensing Basis PCT due to EPU, which is sufficiently below the 10 CFR 50.46 limit of 2200°F. The bounding GGNS EPU Licensing Basis PCT of 1690°F ensures that this 10 CFR 50.46 requirement is met.

2.8.5.6.2.5.5 Coolable Geometry (ECCS-LOCA) EPU has no effect on the Coolable Geometry (ECCS LOCA). Coolable geometry has been dispositioned for BWRs per ELTR1. Conformance with coolable geometry requirements is demonstrated by conformance with the 2200°F Licensing Basis PCT limit and local cladding oxidation limit of 17%. The bounding GGNS EPU Li censing Basis PCT of 1690°F ensures this 10 CFR 50.46 requirement is met.

2.8.5.6.2.5.6 Long-Term Cooling (ECCS-LOCA) EPU has no effect on the long-term cooling (ECCS LOCA). Long-term cooling has been dispositioned for BWRs per ELTR1. Long-term coo ling is assured by either: (1) core can be reflooded above TAF, or (2) core can be reflooded to the elevation of the jet pump suction and one CS system can be placed in operation at rated flow. The ECCS system design at GGNS ensures this 10 CFR 50.46 requirement is met. Conclusion The analyses of the LOCA have been reviewed to ensure they have adequately accounted for operation of the plant at the proposed power level and that the analyses were performed using acceptable analytical models. The analyses demonstrate that the RPS and the ECCS will continue to ensure that the PCT, total oxida tion of the cladding, total hydrogen generation, and changes in core geometry, and long-term cooling will remain within acceptable limits. Based on NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-366 this, Entergy concludes that the plant will continue to meet the requirements of GDCs 4, 27, 35, 10 CFR 50.46, and 10 CFR 50 Appendix K following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the LOCA. 2.8.5.7 Anticipated Transients Without Scram Regulatory Evaluation ATWS is defined as an AOO followed by the failure of the reactor trip portion of the protection system specified in GDC-20. 10 CFR 50.62 requires that: Each BWR have an ARI system that is designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device. Each BWR have a SLCS with the capability of injecting into the reactor vessel a borated water solution with reactivity control at l east equivalent to the control obtained by injecting 86 gpm of a 13 weight-percent sodium pentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251-inch inside diameter reactor vessel. The SLCS initiation must be automatic (for plants granted a construction permit after July 26, 1984). Each BWR have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS. A review has been performed of the effects of the proposed EPU on: (1) the ability to meet the above requirements; (2) sufficient margin available in the setpoint for the SLCS pump discharge

relief valve; and (3) operator actions specified in the EOPs (consistent with the generic emergency procedure guidelines/severe accident guidelines (EPGs/SAGs)). The review confirmed that: (1) the peak vessel bottom pressure is less than the ASME Service Level C limit of 1,500 psig; (2) the PCT is within the 10 CFR 50.46 limit of 2,200°F; (3) the peak SP temperature is less than the design limit; and (4) the peak containment pressure is less than the containment design pressure. In addition, the potential for thermal-hydraulic instability in conjunction with ATWS events was also evaluated using the methods and criteria approved by

the NRC staff. This evaluation considered the effects of the proposed EPU on the limiting event determination, the sequence of events, the analytical model and its applicability, the values of parameters used in the analytical model, and the results of the analyses. GGNS Current Licensing Basis The analysis of the ATWS event is described in UFSAR Section 15.8. As described in UFSAR Section 9.3.5, the SLCS is capable of injecting the required boron into the reactor vessel; the system is manually initiated. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-367 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 9.3.1 of the CLTR addresses the effect of CPPU on ATWS. Analysis of ATWS events is required for CLTP and for EPU 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 1,500 psig). Maintain containment integrity (i.e., maximum containment pressure and temperature

less than the design pressure (15 psig) and temperature (210°F) of the containment structure). Maintain coolable core geometry (Coolable core geometry is assured by meeting the 2200°F PCT and the 17% local cladding oxidation acceptance criteria of 10 CFR 50.46). This evaluation reviewed the results of the ATWS analyses considering the limiting cases for RPV overpressure and for SP temperature / containment pressure. Previous evaluations considered four ATWS events. Based on experience and the generic analyses performed for Reference 4 (ELTR2), only two cases need to be further analyzed for GGNS: (1) MSIVC and (2) PRFO. For GGNS, a LOOP doe s not result in a reduction in the RHR pool cooling capability relative to these cases. Thus, with the same RHR pool cooling capability, the containment responses for the MSIVC and PRFO cases bound th e LOOP case. These events have been analyzed and the results are presented below in Sections 2.8.5.7.1 through 2.8.5.7.3. The EPU ATWS analysis is performed using the NRC approved code ODYN (see Table 1-1). The key inputs to the ATWS analysis are provided in Table 2.8-8. The results of the analysis are provided in Table 2.8-9 and discussed below. The results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the GGNS response to an ATWS event at EPU is acceptable. The potential for thermal-hydraulic instability in conjunction with ATWS events is evaluated in Section 2.8.3.2. GGNS meets the ATWS mitigation requirements defined in 10 CFR 50.62: Installation of an ARI system; Boron injection equivalent to 86 gpm of 13 weight percent natural boron; and Installation of automatic RPT logic (i.e., ATWS-RPT). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-368 The 86 gpm boron injection equivalency requirement of 10 CFR 50.62 is satisfied via the following relationship: (Q/86) x (M251/M) x (C/13) x (E/19.8) 1 where: Q = Expected SLCS flow rate (gpm) M251/M = Mass of water in a 251-inch diameter reactor vessel and recirculation system (lbs) / mass of water in the reactor vessel and recirculation system at hot rated condition (lbs) C = Sodium pentaborate solution concentration (weight percent) E = Boron-10 isotope enrichment (atom-percent) For GGNS, Q = 82.4 gpm M251/M = 1 (because GGNS has a 251-inch diameter reactor vessel) C = 13.6 % E = 19.8 % (i.e., natural boron enrichment) Therefore, the 86 gpm equivalency requirement for CLTP and EPU (no modification) is satisfied as follows: (Q/86) x (M251/M) x (C/13) x (E/19.8) 1 (82.4/86) x (1) x (13.6/13) x (19.8/19.8) = 1.0 1 To provide margin to the 86 gpm equivalency requirement, GGNS may modify the SLCS sodium pentaborate solution concentration or the Boron-10 isotope enrichment to continue to satisfy the inequality. Leaving the SLCS flow rate and the mass of water ratios unchanged, any combination of concentration (13.6) and enrichment (19.8) when multiplied is greater than 269.28, or (13.6x19.8). For example, using a concentration of 4.4 w/o and an enrichment of 96%, the (C)x(E) result of 422 would satisfy the 86 gpm equivalency requirement. There are no changes to the assumed operator actions for the EPU ATWS analysis. BWROG "Emergency Procedure and Severe Accident Guidelines (EPGs/SAGs)," Revision 2, March 2001 (Reference 86), is currently implemented at GGNS. EPU implementation does not change operator strategy on ATWS level reduction or early boron injection. EPU may affect some of NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-369 the calculated curves, but does not affect stability mitigation actions. The changes due to EPU do not require modification of operator instructions. When required by changes in plant configurati on (as identified by the design change process), changes to EOPs, including changes to EOP calculations and plant data, are developed and implemented in accordance with plant administrative procedure for EOP program maintenance. GGNS performs EOP calculations consistent with the BWROG EPGs/SAGs Appendix C. Critical software is verified and validated by Engineering to generate EOP results. The EOP calculation input and output data is reviewed a nd verified by Engineeri ng. 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. Fi nally, the EOP flow charts are reviewed and validated by Operations, including trial use in the simulator. GGNS meets all CLTR dispositions and the results in this evaluation are described below. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result ATWS (Overpressure) - Event Selection [[ Meets CLTR Disposition ATWS (Overpressure) - Limiting Events

Meets CLTR Disposition ATWS (Suppression Pool Temperature) - Event Selection

Meets CLTR Disposition ATWS (Suppression Pool Temperature) - Limiting Events

Meets CLTR Disposition ATWS (Peak Cladding Temperature)

     ]] Meets CLTR Disposition 2.8.5.7.1 ATWS (Overpressure)

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 will result in higher SP 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.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-370 The overpressure evaluation includes a review of the results of the analyses of ATWS events to identify the most limiting RPV overpressure cond itions. Two events, MSIVC and PRFO, were further analyzed for GGNS. The ATWS (Overpressure) - Event Selection meets all CLTR dispositions. The MSIVC and PRFO sequence of events are given in Tables 2.8-10 and 2.8-11, respectively. The short-term and long-term transient response to the MSIVC and PRFO ATWS events is presented in Figures 2.8-22 through 2.8-37. The key inputs and limiting results are presented in Tables 2.8-8 and 2.8-9. The limiting ATWS event with respect to RPV overpressure for GGNS is MSIVC. The PRFO event produces the highest peak upper plenum pressure at SLCS initiation (1205 psia). Therefore, ATWS (Overpressure) - Limiting Events meet all CLTR dispositions. 2.8.5.7.2 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 will result in higher SP 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 SP temperature evaluation includes a review of the results of the analyses of ATWS events to identify the most limiting containment res ponse. Two events, MSIVC and PRFO, were further analyzed for GGNS. The ATWS (Suppression Pool Temperature) - Event Selection meets all CLTR dispositions. The MSIVC and PRFO sequence of events are given in Tables 2.8-10 and 2.8-11, respectively. The short-term and long-term transient responses to these events are presented in Figures 2.8-22 through 2.8-37. The key inputs and limiting results are presented in Tables 2.8-8 and 2.8-9. The limiting ATWS event with respect to containmen t response for GGNS is PRFO. Therefore, the ATWS (Suppression Pool Temperature) - Limiting Events meet all CLTR dispositions. 2.8.5.7.3 ATWS (Peak Cladding Temperature) The CLTR states that EPU has a negligible effect on the PCT or local cladding oxidation. [[

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

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-371 Coolable core geometry is assured by meeting the 2200ºF PCT and the 17% local cladding oxidation acceptance criteria stated in 10 CFR 50.46. The ATWS analysis results demonstrate significant margin to the PCT acceptance criteria of 2200ºF. Two events, MSIVC and PRFO, were further analyzed for GGNS. The highest calculated PCT for ATWS events is 1560ºF, which resulted from the PRFO event. Local

cladding oxidation is not explicitly analyzed because, with PCT less than 1600ºF, cladding oxidation has been demonstrated to be insignificant compared to the accep tance criteria of 17% of cladding thickness. Therefore, the local cladding oxidation for the GGNS ATWS events is qualitatively evaluated to demonstrate compliance with the acceptance criteria of 10 CFR 50.46. Therefore, ATWS (Peak Cladding Temperature) is in compliance with th e acceptance criteria of 10 CFR 50.46.

Conclusion The analysis of the ATWS event has been review ed to ensure it has adequately accounted for the effects of the proposed EPU. The analysis demonstrates that ARI, SLCS, and RPT systems have been installed and that they will continue to meet the requirements of 10 CFR 50.62 and the analysis acceptance criteria following implementati on of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to ATWS.

2.8.6 Fuel Storage 2.8.6.1 New Fuel Storage Regulatory Evaluation Nuclear reactor plants include facilities for the stor age of new fuel. The quantity of new fuel to be stored varies from plant to plant, depe nding upon the specific design of the plant and the individual refueling needs. A review has been performed of the effect s of the proposed EPU on the ability of the storage facilities to maintain the new fuel in a subcritical array during all credible storage conditions. The review focused on the effect of changes in fuel design on the analyses for the new fuel storage facilities. The regulatory acceptance criteria are based on GDC-62, insofar as it requires the prevention of criticality in fuel storage systems by physical systems or processes, preferably utilizing geometrically safe configurations.

GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The new fuel storage system is described in UFSAR Section 9.1.1. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-372 Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.3.4 of the CLTR addresses the effect of CPPU on Fuel Racks. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The following topics are addressed in this section: Topic CLTR Disposition GGNS Result Fuel Racks [[

     ]] Meets CLTR Disposition The additional energy requirements for EPU are met by an increase in bundle enrichment, an increase in the fuel reload batch size, and/or ch anges in the fuel loading pattern. There are no changes in the fuel mechanical design associated with EPU.

New fuel assemblies are evaluated every cycle for storage in GEH Low-Density Fuel Storage (LDFS) racks. The evaluation confirms the reload fuel batch is less reactive than the design basis fuel assembly assumed in the LDFS criticality safety analysis. The evaluation considers changes in fuel enrichment, gadolinia, and fuel geometry. The representative fuel designs employed to support the EPU have been confirmed to be less reactive than the design basis fuel assembly. Because this evaluation addresses fres h unirradiated fuel storage, EPU has no affect on the results. Therefore, the new fuel storage requirements are fulfilled using the current design parameters, and no changes are required to ensure that new fuel can be maintained in a

sub-critical array during all credible storage conditions. Therefore, New Fuel Storage meets all CLTR dispositions.

Conclusion The analyses of the new fuel storage facilities c ontinue to apply to fuel designs which support EPU operation. The cycle specific designs will c ontinue to be evaluated for each reload to ensure that the new fuel storage facilities will continue to meet the requirements of GDC-62 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the new fuel storage. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-373 2.8.6.2 Spent Fuel Storage Regulatory Evaluation Nuclear reactor plants include storage facilities for the wet storage of SF assemblies. The safety function of the wet fuel storage facilities (SFP and SF UCP) is to maintain the SF assemblies in a safe and sub-critical array during all credible storage conditions and to provide a safe means of loading the assemblies into shipping casks. A review has been performed of the effects of the proposed EPU on the criticality analysis. The regulatory acceptance criteria are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated acci dents; and (2) GDC-62, insofar as it requires that criticality in the fuel storage systems be prevented by physical systems or processes, preferably by use of geometrically safe configurations. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The SF storage system is described in UFSAR Section 9.1.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.3.4 of the CLTR addresses the effect of CPPU on Fuel Racks. The results of this evaluation are described below. GGNS meets all CLTR dispositions. The following topics are addressed in this section: Topic CLTR Disposition GGNS Result Fuel Racks [[

     ]] Meets CLTR Disposition CLTR Section 6.3.4 states that the increased decay heat from EPU results in a higher heat load in the racks during long-term storage.

The wet fuel storage facilities (SFP and UCP) conti nue to rely on a neutron absorber (poison) to maintain sub criticality. GGNS has Boraflex fuel storage racks in both the SFP and UCP. These racks are monitored and evaluated using RACKLIFE software to verify their acceptability. The NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-374 racks are categorized into two regions based on the program results, which consider dose and boron carbide loss. Each assessment includes projections to confirm acceptable performance through the subsequent evaluation period. This program is described in GGNS UFSAR

Section 9.1.2.3. Fuel assemblies are evaluated every cycle for storage in wet fuel storage racks. The evaluation confirms the reload fuel batch is less reactive than the fuel assembly assumed in the criticality safety analysis of record. The evaluation considers any changes in fuel enrichment, gadolinia, or fuel geometry. The representative fuel designs employed to support the EPU have been confirmed to be less reactive than the design basis fuel assembly. Because the EPU is a constant pressure power uprate and the fuel designs continue to meet the same CLTP reactivity criteria, EPU does not affect the assumptions used in the criticality safety analysis. Although there is an increase in the fuel pool heat load due to higher decay heat, the pool temperature continues to remain below the design temperature for all design basis offload scenarios. The temperature will also remain below the operating limit for normal operating conditions. There is no effect on the SF storage racks from the increased EPU heat load. Therefore, the Spent Fuel Storage meets all CLTR dispositions.

Conclusion The analyses of the wet fuel storage facilities (SFP and UCP) continue to apply to fuel designs which support EPU operation. The cycle specific de signs will continue to be evaluated for each reload to ensure that the wet fuel storage facilities meet the requirements of GDCs 4 and 62 following implementation of the proposed EPU. Therefore, Entergy finds the proposed EPU acceptable with respect to the fuel storage. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-375 Table 2.8-1 Peak Nodal Exposures EOC Peak Nodal Exposures Plant Cycle Peak Nodal Exposure (GWD/ST) A 18 38.849 A 19 43.784 B 9 56.359 B 10 51.544 C 7 53.447 C 8 47.766 D 13 56.660 E 11 55.387 F EQ - 120% 51.174 GGNS EPU 55.264 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-376 Table 2.8-2 Option III Setpoint Demonstration OPRM Amplitude Setpoint 1 OLMCPR(SS) 2 OLMCPR(2PT) 2 1.04 1.234 1.257 1.05 1.260 1.285 1.06 1.288 1.313 1.07 1.317 1.343 1.08 1.348 1.374 1.09 1.380 1.406 1.10 1.412 1.439 1.11 1.445 1.473 1.12 1.480 1.509 1.13 1.517 1.546 1.14 1.556 1.586 Acceptance Criteria Off-rated OLMCPR at 45% Flow Non-limiting Rated Power OLMCPR Estimated to be 1.39 Notes: 1. 5% calibration error is included in the OPRM Amplitude Setpoint. 2. 0.01 bypass voiding penalty was added to OLMCPR(SS) and OLMCPR(2PT) values. Assumed SLMCPR = 1.10. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-377 Table 2.8-3 ODYSY Decay Ratios at BSP Region Boundary Endpoints Point Power (%) Flow (%) Core Decay Ratio Channel Decay Ratio Controlled Entry (Region II), NCL Intercept B2 29.6 24.7 0.800 0.305 B2-ICA 26.4 24.4 Scram (Region I), NCL Intercept B1 40.8 25.6 0.795 0.385 B1-ICA 38.5 25.5 Controlled Entry (Region II), HFCL Intercept A2 54.6 34.8 0.795 0.359 A2-ICA 67.3 50.0 Scram (Region I), HFCL Intercept A1 48.5 27.8 0.777 0.411 A1-ICA 59.0 40.0

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2-378 Table 2.8-4 Parameters Used For Transient Analysis Parameter CLTP EPU Rated Thermal Power (MWt) 3898 4408 Analysis Power (% Rated) 100 / 102 1 100 / 102 1 Analysis Dome Pressure (psia) 1040 1040 Rated Core Flow (Mlb/hr) 112.5 112.5 Rated Power Core Flow Range (% Rated) 77.1 - 105 92.8 - 105 Normal Feedwater Temperature ( F) 420 420 Feedwater Temperature Reduction (T F) 100 100 No. of SRVs assumed in the analysis 13 13 Note: 1. GEMINI MCPR analyses at 100%, non-MCPR analyses at 102%.

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2-379 Table 2.8-5 Transient Analysis MCPR Results Plant CLTP Uncorrected CPR EPU Uncorrected CPR Load Rejection with Bypass Failure 0.22 0.23 Turbine Trip with Bypass Failure 0.20 0.22 Feedwater Controller Failure Max Demand 0.20 0.21 Pressure Regulator Failure - Downscale 0.11 0.13 Loss of Feedwater Heater 0.15 0.12 Rod Withdrawal Error 0.16 0.16 Slow Recirculation In crease Note 1 MCPR f Fast Recirculation Increase Note 2 0.07 3 Load Rejection With Bypass Note 2 0.15 MSIV Closure All Valves Note 2 0.02 MSIV Closure 1 Valve Note 2 0.07 Notes: 1. Event not analyzed at CLTP. 2. Event not analyzed at CLTP. The EPU evaluation confirms the event is non-limiting.

3. The Fast Recirculation Increase event initializes from 55% EPU rated power.

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2-380 Table 2.8-6 Key Input Parameters for SAFER/GESTR LOCA Evaluation Item Parameter** Unit Nominal Value Appendix K Value 1 CLTP MWt 3,898 3,909.7 2 CLTP LOCA Analysis Basis (1) MWt 4,025 4,105.5 3 EPU MWt 4,408 4,496.2 4 Vessel Steam Dome Pressure psia 1095 1100 5 Rated Core Flow Mlb/hr 112.5 112.5 6 Maximum RSLB Area (2) ft 2 3.143 3.143 7 GNF2 Number of Fuel Rods per Bundle NA 92 92 8 GNF2 PLHGR kW/ft 13.80 14.40 9 GNF2 MAPLHGR kW/ft 13.20 13.78 10 GNF2 Worst Pellet Exposure for ECCS Evaluation MWd/MTU 14600 14600 11 Single Failure Input NA HPCS Diesel Generator HPCS Diesel Generator 12 Limiting Large / Small Break Location NA Recirculation Suction Line Recirculation Suction Line Notes: 1. CLTP LOCA Analysis Basis is 103.3 % of CLTP. 2. The maximum RSLB area includes the RSLB area (3.112 ft

2) plus the bottom head drain area (0.031 ft 2).

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2-381 Table 2.8-7 ECCS Conformance Results Parameter CLTP EPU 10 CFR 50.46 Limit Method SAFER/GESTR SAFER/GESTR Thermal Power (MWt) 4,025 4,408 Licensing Basis PCT (o F) < 1690 < 1690 < 2200 Cladding Oxidation (% Original Clad Thickness) < 2 < 2 < 17 Hydrogen Generation, Core-Wide Metal-Water Reaction (%) < 0.1 < 0.1 < 1.0 Coolable Geometry Acceptable Acceptable PCT < 2200 o F, and Local Oxidation <17% Core Long-Term Cooling Acceptabl e Acceptable Core flooded to TAF OR Core flooded to jet pump suction elevation and at least one CS system is operating at rated flow. [[

]]

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2-382 Table 2.8-8 GGNS Key Inputs for ATWS Analysis Input Variable CLTP EPU Reactor power (MWt) 3,898 4,408 Reactor dome pressure (psia) 1040 1040 Each SRV capacity at 1205 psig (Mlbm/hr) 0.925 0.925 High pressure ATWS-RPT setpoint (psig) 1139 1139 Number of SRVs OOS 1 5 Table 2.8-9 GGNS Results for ATWS Analysis Acceptance Criteria CLTP 1 EPU 1 Peak vessel bottom pressure (psig) 1299 1455 Peak SP temperature ( F) 165 165 Peak containment pressure (psig) < 8.2 6.4 PCT ( F) 1509 1560 2 Local cladding oxidation (%) < 17 < 17 Notes: 1. Cladding temperature and oxidation remain below their 10 CFR 50.46 limits.

2. The EPU PCT evaluations consider GNF2 fuel. The GNF2 fuel type is evaluated with a higher LHGR than GE14 fuel, which the CLTP analysis is based on.

The difference in PCT is a result of the change in fuel types. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-383 Table 2.8-10 MSIVC Sequence of Events Item Event EPU BOC Event Time (sec) EPU EOC Event Time (sec) 1 MSIV Isolation Initiated 0.0 0.0 2 MSIVs Fully Closed 4.0 4.0 3 High Pressure ATWS Setpoint 4.1 4.0 4 Peak Neutron Flux 4.7 4.0 5 Recirculation Pumps Trip 4.1 4.0 6 Opening of the First Relief Valve 4.6 4.6 7 Peak Heat Flux 4.9 4.9 8 Peak Vessel Pressure 11.3 10.9 9 FW Reduction Initiated 30.0 30.0 10 BIIT Reached 33.0 33.0 11 SLCS Pumps Start 124.1 1 124.0 1 12 Hot Shutdown Achieved (Neutron Flux Remains < 0.1%) 609 639 13 RHR Cooling Initiated 660 660 14 Peak SP Temperature 1540 1400 Note: 1. SLCS injection is the later time of either: 1) two minutes after the high pressure RPT or 2) when the SP temperature reaches the boron injection initiation temperature (BIIT). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-384 Table 2.8-11 PRFO Sequence of Events Item Event EPU BOC Event Time (sec) EPU EOC Event Time (sec) 1 TCV and Bypass Valves Start Open 0.1 0.1 2 MSIVC Initiated by Low Steamline Pressure 12.1 11.5 3 MSIVs Fully Closed 16.1 15.5 4 Peak Neutron Flux 16.2 16.1 5 High Pressure ATWS Setpoint 18.6 18.0 6 Recirculation Pumps Trip 18.6 18.0 7 Opening of the First Relief Valve 19.1 18.5 8 Peak Heat Flux 19.5 18.8 9 Peak Vessel Pressure 23.7 22.5 10 FW Reduction Initiated 42.9 42.9 11 BIIT Reached 49.0 49.0 12 SLCS Pumps Start 138.6 1 138.0 1 13 Hot Shutdown Achieved (Neutron Flux Remains < 0.1%) 615 675 14 RHR Cooling Initiated 660 660 15 Peak SP Temperature 1510 1436 Note: 1. SLCS injection is the later time of either 1) two minutes after the high pressure RPT or 2) when the SP temperature reaches the BIIT. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-385 Figure 2.8-1 Power of Peak Bundle versus Cycle Exposure 4.04.5 5.0 5.5 6.06.57.07.58.0024681012141618Cycle Exposure (GWD/ST)Maximum Bundle Power (MW)Plant A Cycle 18Plant A Cycle 19Plant B Cycle 9Plant B Cycle 10Plant C Cycle 7Plant C Cycle 8Plant D Cycle 13Plant E Cycle 11Plant FGGNS EPU T0200

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2-386 Figure 2.8-2 Coolant Flow for Peak Bundle versus Cycle Exposure 5 6 7 8 9 10 11 12 13024681012141618Cycle Exposure (GWD/ST)Flow for Peak Bundle (10e4 lbm/hr)Plant A Cycle 18Plant A Cycle 19Plant B Cycle 9Plant B Cycle 10Plant C Cycle 7Plant C Cycle 8Plant D Cycle 13Plant E Cycle 11Plant FGGNS EPU T0200

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2-387 Figure 2.8-3 Exit Void Fraction for Peak Power Bundle versus Cycle Exposure 0.700.750.800.85 0.90024681012141618Cycle Exposure (GWD/ST)Exit Void Fraction for Peak Power BundlePlant A Cycle 18Plant A Cycle 19Plant B Cycle 9Plant B Cycle 10Plant C Cycle 7Plant C Cycle 8Plant D Cycle 13Plant E Cycle 11Plant FGGNS EPU T0200

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2-388 Figure 2.8-4 Maximum Channel Exit Void Fraction versus Cycle Exposure 0.700.750.800.850.90024681012141618Cycle Exposure (GWD/ST)Max Channel Exit Void FractionPlant A Cycle 18Plant A Cycle 19Plant B Cycle 9Plant B Cycle 10Plant C Cycle 7Plant C Cycle 8Plant D Cycle13Plant E Cycle 11Plant FGGNS EPU T0200

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2-389 Figure 2.8-5 Core Average Exit Void Fraction versus Cycle Exposure 0.600.620.640.660.680.700.720.740.760.780.80024681012141618Cycle Exposure (GWD/ST)Core Average Exit Void FractionPlant A Cycle 18Plant A Cycle 19Plant B Cycle 9Plant B Cycle 10Plant C Cycle 7Plant C Cycle 8Plant D Cycle 13Plant E Cycle 11Plant FGGNS EPU T0200

1) The core average exit void fraction for Grand Gulf is slightly higher than the experience base, but this is primarily the result of fuel bundle and core design optimization for improved fuel cycle economy. The average exit void fraction remains at a level below 80%, so the amount of any cross section extrapol ation beyond the standard 0-40-70% reference points is minor. As noted in Figures 2.8-3 and 2.8-4, the maximum exit void fraction for the peak bundle and core maximum exit void fraction remains well within the experience base.

Any applied penalties on the SLMCPR and OL MCPR as dictated by the current NRC approved revision of the Methods LTR will be applied to Grand Gulf and remain sufficient to address NRC concerns on application of the GEH/GNF methodology for EPU. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-390 Figure 2.8-6 Peak LHGR versus Cycle Exposure 0 2 4 6 8 10 12 14 16024681012141618Cycle Exposure (GWD/ST)Peak LHGR (kW/ft)Plant A Cycle 18Plant A Cycle 19Plant B Cycle 9Plant B Cycle 10Plant C Cycle 7Plant C Cycle 8Plant D Cycle 13Plant E Cycle 11Plant FGGNS EPU T0200

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2-391 Figure 2.8-7 Dimensionless Bundle Power at BOC (200 MWd/ST)

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2-392 Figure 2.8-8 Dimensionless Bundle Power at MOC (9000 MWd/ST)

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2-393 Figure 2.8-9 Dimensionless Bundle Power at EOC (18615 MWd/ST)

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2-394 Figure 2.8-10 Bundle Operating LHGR (KW/ft) at BOC (200 MWd/ST)

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2-395 Figure 2.8-11 Bundle Operating LHGR (kw/ft) at MOC (9000 MWd/ST)

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2-396 Figure 2.8-12 Bundle Operating LHGR (kw/ft) at EOC (18615 MWd/ST)

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2-397 Figure 2.8-13 Bundle Operating MCPR at BOC (200 MWd/ST)

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2-398 Figure 2.8-14 Bundle Operating MCPR at MOC (9000 MWd/ST)

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2-399 Figure 2.8-15 Bundle Operating MCPR at EOC (18615 MWd/ST)

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2-400 Figure 2.8-16 Bundle Operating LHGR (kw/ft) at 11000 MWd/ST (peak MFLPD point) Note: MFLPD = Maximum Fraction of Limiting Power Density NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-401 Figure 2.8-17 Bundle Operating MCPR at 2000A MWd/ST (peak MFLCPR point) Note: MFLCPR = Maximum Fraction of Limiting Critical Power Ratio NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-402 Figure 2.8-18 Bundle Average Void Fraction vs. MCPR 1.41.45 1.51.55 1.61.65 1.71.75 1.80.450.470.490.510.530.550.570.590.610.630.65Bundle Average Void Fraction MCPR NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-403 Figure 2.8-19 Illustration of OPRM Trip-Enabled Region and BSP Regions 0 10 20 30 40 50 60 70 80 90 100 110 1200102030405060708090100110120Core Flow (% rated)Core Power (% rated)Scram RegionControlled Entry Region OPRM Trip Enabled Region A2-IC A A1-IC AB1-ICAB2-ICAOPRM TripEnabled Region

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2-404 Figure 2.8-20 Response to MSIVF (MSIV Closure with High Flux Scram) (102% EPU power, 105% core flow, and 1060 psia initial dome pressure) Parameters with unique units are identified in brackets. '" "" "" m '" " iO 1 25 * * "" '" " '" "" m '" m "" " " * '" * " _ Neutron F lux _ Av erage S urf ace Heat F lux Core Inlet F low _ Core Inlet S ubcooli Time (5) Level [in abo v e Sep_ Sk irt) V esse l Stea m F low Feedwaler F low Turbine Stea m F low ,,+-.-o---+--+---+--+-.-,_--+_-+----j , " Time (5) '" V esse l Press ure R i se [p sg) Safe ty V a lve F low " Re lief V a lve F low B pass Stea mflow "" m '" "" "" " iO 175 *

'" m "" '" " '"

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2-405 Figure 2.8-21 Loss of Feedwater Flow Note: The level is in reference to vessel bottom. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-406 Figure 2.8-22 EPU MELLLA BOC MSIVC (Short-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-407 Figure 2.8-23 EPU MELLLA BOC MSIVC (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-408 Figure 2.8-24 EPU MELLLA BOC MSIVC (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-409 Figure 2.8-25 EPU MELLLA BOC MSIVC (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-410 Figure 2.8-26 EPU MELLLA BOC PRFO (Short-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-411 Figure 2.8-27 EPU MELLLA BOC PRFO (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-412 Figure 2.8-28 EPU MELLLA BOC PRFO (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-413 Figure 2.8-29 EPU MELLLA BOC PRFO (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-414 Figure 2.8-30 EPU MELLLA EOC MSIVC (Short-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-415 Figure 2.8-31 EPU MELLLA EOC MSIVC (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-416 Figure 2.8-32 EPU MELLLA EOC MSIVC (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-417 Figure 2.8-33 EPU MELLLA EOC MSIVC (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-418 Figure 2.8-34 EPU MELLLA EOC PRFO (Short-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-419 Figure 2.8-35 EPU MELLLA EOC PRFO (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-420 Figure 2.8-36 EPU MELLLA EOC PRFO (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-421 Figure 2.8-37 EPU MELLLA EOC PRFO (Long-Term) [[ ]] NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-422 2.9 Source Terms and Radiological Consequences Analyses

2.9.1 Source

Terms for Radwaste Systems Analyses Regulatory Evaluation Entergy reviewed the radioactive source term asso ciated with EPU to ensure adequacy of the sources of radioactivity used by GGNS as input to calculations that verify the radioactive waste management systems have adequate capacity for treating radioactive liquid and gaseous wastes. The review included the parameters used to determine: (1) concentration of each radionuclide in the reactor coolant; (2) fraction of fission produc t activity released to the reactor coolant; (3) concentrations of all radionuclides other than fission products in the reactor coolant; (4) leakage rates and associated fluid activity of all potentially radioactive water and steam systems; and (5) potential sources of radioactive materials in effluents not considered in the plant's UFSAR related to LWMSs and GWMSs. The regulatory acceptance criteria for source terms are based on: (1) 10 CFR 20, insofar as it establishes requirements for radioactivity in

liquid and gaseous effluents released to unrestr icted areas; (2) 10 CFR 50, Appendix I, insofar as it establishes numerical guides for design objectives and LCOs to meet the "as low as is

reasonably achievable" (ALARA) criterion; and (3) GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The current core source terms are described in UFSAR Section 15.6.5.5 and Table 15.6-9.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Sections 8.3 and 8.4 of th e CLTR address the effect of CPPU on the Radiation Sources in the Reactor Core and in the Reactor Coolant. The results of this evaluation are described below. 2.9.1.1 Radiation Sources in the Reactor Core As explicitly stated in Section 8.3 of the CLTR, the radiation sources in the core are directly related to the fission rate during power operation. These sources include radiation from the fission process, accumulated fission products and neutron reactions as a secondary result of NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-423 fission. Historically, these sources have been defined in terms of energy or activity released per unit of reactor power. However, because GGNS uses GNF2 fuel a nd is not enveloped by the bounding analysis performed for the CLTR, the CLTR is not applicable for fuel design dependent evaluations. Thus, the methods and assumptions for the CLTR radiological evaluations are not applicable. For the EPU at GGNS, the radiological evaluation has been performed using the methods and assumptions outlined in ELTR1 (Reference 2). A bounding analysis has been performed to enve lop the radiation sources evaluation. The ELTR1 bounding parameters used for core radia tion source calculations, and the corresponding GGNS values, are: ELTR1 Generic Parameter Requirement GGNS Corresponding Values [[ EOC final core average exposure = 30.1 GWd/ST

Initial bundle enrichment of 4.12 wt%

     ]] Bundle average power = 5.62 MWt (at 102%

of EPU RTP) As stated in Appendix H of ELTR1, plants not conforming to the generic parameters require a plant-specific evaluation. Therefore, a plant-specific evaluation is performed using the following parameters: Parameter GGNS EPU Plant-Specific Evaluation Value Core Power 4,496 MWt (102% EPU) Initial Bundle Enrichment 4.0% to 4.5% U-235 EOC Core Average Exposure 36.0 GWd/MTU Individual Bundle Discharge Exposure

Range 15.0 - 62.0 GWd/MTU Fuel Design GNF2 Total core radiation sources increase approximately proportional to the increase in reactor

power. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-424 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 of these 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. The total gamma energy source, therefore, 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 data are n eeded for post-accident and SFP evaluations, which are performed in compliance with regulatory guidan ce that applies different release and transport assumptions to different fission products. Th e core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops "equilibrium"

activities in the fuel (typically 3 years). Most radiologically significant fission products reach equilibrium within a 60-day period. [[

                                                                                                                           ]]  The radionuclide inventories of the full core at EOC and a bundle at end-of-life discharge are calculated in terms of Curies. The calculated source terms resulting from this evaluation have been used in the radiological

consequence evaluation of the transient analyses discussed in Section 2.9.2. In addition, they have been used to evaluate the plant radiation levels as discussed in Section 2.10.1.2. The

results are also used in EQ evaluations discu ssed in Section 2.3.1. As noted in those sections, the radiological consequences and the shielding evaluation results meet applicable regulatory acceptance criteria. 2.9.1.2 Radiation Sources in Reactor Coolant For coolant activation products, the typical margin in the plant design basis for reactor coolant concentrations significantly exceeds the potential in creases due to power uprate and needs to be verified. Also, because the transport time from core exit to downstream points will decrease with increased flow from EPU, the resultant dose rates in the MSLs, turbines, and condenser area will increase roughly proportional to power uprate. In the case of activated corrosion products and fission products, plant-specific analysis is re quired by the CLTR to verify that the corrosion

product concentrations do not exceed the design basis concentrations. Tables 2.9-1 through 2.9-6 contain the activity levels , concentrations, and release rates for these radiation sources for GGNS. The topics addressed in this evaluation are: NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-425 Topic CLTR Disposition GGNS Result Coolant Activation Products [[ Tables 2.9-1 and 2.9-6 Activated Corrosion Products and Fission Products

     ]] Tables 2.9-1, 2.9-4, and 2.9-5 2.9.1.2.1 Coolant Activation Products The CLTR states that increases in reactor power will increase the activity of activation products found in reactor coolant. During reactor opera tion, the coolant passing through the core region becomes radioactive as a result of nuclear reac tions. The coolant activation, especially N-16 activity, is the dominant source in the turbine building and in the lower regions of the DW. The activation of the water in the core region is in approximate proportion to the increase in thermal power.  [[

      ]]  The margin in the GGNS plant design basis for reactor coolant activati on concentrations significantly exceeds potential increases due to EPU. Resultant dose rates in the MSLs, turbines, and condenser area will increase roughly proportional to power uprate due to the fact that the transport time from core exit to downstream points will decrease with increased flow. Therefore, no change is required in the activation design basis reactor coolant concen trations for EPU and all CLTR dispositions are met for coolant activation products.

2.9.1.2.2 Activated Corrosion Products and Fission Products The CLTR states that increases in reactor pow er will increase the activity of corrosion products and fission products found in reactor coolant. The reactor coolant contains activated corrosion products, which are the result of metallic materials entering the water and being activated in the reactor region. Under EPU conditions, the FW flow increases with power, the activation rate in the reactor region increases with power, and the filtration run-lengths of the condensate demineralizers may decrease as a result of the FW fl ow increase. The net re sult is an increase in the activated corrosion product present in the coolant. Fission products in the reactor coolant are separable into the products in the steam and the products in the reactor water. The activity in the steam consists of noble gases released from the core plus carryover activity from the reactor water. This activity is the noble gas offgas that is included in the plant design. The calculated offgas rates for EPU after thirty minutes decay are 0.064 Curies/sec, within the ODB of 0.1 Curies/sec, per Table 2.9-1. Therefore, no change is required in the design basis for offgas activity for EPU. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-426 The fission product activity in the reactor water, like the activity in the steam, is the result of minute releases from the fuel rods. EPU fissi on product activity levels in the reactor water remain a fraction (12%) of the design basis fission product activity, per Table 2.9-1. The total activated corrosion product activity was cal culated to be 1% greater than design basis levels, as can be seen in Table 2.9-1. However, the sum of the activated corrosion product activity and the fission product activity remains a fr action (14%) of the total design basis activity in reactor water, per Table 2.9-1. Therefore, the activated corrosion product and fission product activities design bases for GGNS are unchanged for EPU. For EPU, normal radiation sources are expected to increase slightly. Shielding aspects of the plant were conservatively designed for total norma l radiation sources. Thus, the increase in radiation sources does not affect radiation zoning or shielding and plant radiation area procedural controls will compensate for increased normal radiation sources. Therefore, activated corrosion and fission products meet all CLTR dispositions.

Conclusion The radioactive source term associated with the proposed EPU has been reviewed. Entergy concludes the proposed parameters and resultant composition and quantity of radionuclides are appropriate for evaluating the radioactive waste management systems. Entergy further concludes the proposed radioactive source term meets the requirements of 10 CFR Part 20, 10 CFR 50, Appendix I, and GDC-60. Therefore, Entergy finds the proposed EPU acceptable with respect to source terms.

2.9.2 Radiological

Consequences Analyses Using Alternative Source Terms Regulatory Evaluation Entergy has reviewed the DBA radiological consequences analyses. The radiological consequences analyses reviewed are the LOCA, FHA, CRDA, and MSLB. The review for each accident analysis included: (1) the sequence of events; and (2) models, assumptions, and values of parameter inputs used for calculating the TEDE. The regulatory acceptance criteria for radiological consequences analyses using an alternative source term are based on: (1) 10 CFR 50.67, insofar as it sets standards fo r radiological consequences of a postulated accident; and (2) GDC-19, insofar as it requires ad equate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE, as defined in 10 CFR 50.2, for the duration of the accident. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-427 GGNS Current Licensing Basis NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The current core source terms are described in UFSAR Section 15.6.5.5 and Table 15.6-9.

Technical Evaluation GGNS License Amendment No. 145 (Reference 66) provided NRC approval for a full-scope implementation of an Alternative Source Term (AST) that complies with the guidance given in RG 1.183 (Reference 45) and 10 CFR 50.67. As part of the application that resulted in License Amendment No. 145, GGNS originally evaluated thr ee accidents with the AST. As described in

UFSAR Chapter 15, the accidents are: CRDA (UFSAR Section 15.4.9) LOCA (UFSAR Section 15.6.5) FHA (UFSAR Section 15.7.4) The GGNS licensing basis includes a number of ot her events which are subject to a lower radiological threshold than the accidents lis ted above. GGNS subsequently revised the radiological doses associated with these events with the core source term and/or dose acceptance criteria associated with the AST: Pressure Controller Failure (UFSAR Section 15.2.1) Recirculation Pump Seizure Accident (UFSAR Section 15.3.3) Misplaced Bundle Accident (UFSAR Section 15.4.7) Steam System Piping Break Outside Containment (UFSAR Section 15.6.4) The remaining events within the current licensing basis are documented in UFSAR Chapter 15 and listed below: MSIV Closure (UFSAR Section 15.2.4) Offgas System Leak or Failure (UFSAR Section 15.7.1) Radioactive Liquid Waste System Leak or Failure (release to atmosphere) (UFSAR Section 15.7.2) NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-428 Liquid Radwaste Tank Failure (release to groundwater) (UFSAR Section 15.7.3) The radiological dose consequences of all othe r events are considered to be bounded by the events analyzed and listed above. The magnitude of radiological consequences of a DBA is proportional to the quantity of radioactivity released to the environment. Th is quantity is a function of the fission products released from the core as well as the transport mechanism between the core and the release point.

The effect of the proposed EPU on the radiologi cal consequences of the LOCA, FHA and the CRDA is based on an assessment of the effect of EPU changes on the dose consequence analyses that were evaluated by the NRC in License Amendment No. 145. The referenced amendment is based on 3,910 MWt (corresponding to the CLTP power level of 3,898 MWt with a 0.3% ECCS evaluation uncertainty factor applied). The cal culated AST dose analyses showed that the dose criteria of 10 CFR 50.67 are met at the CLTP power level.

The LOCA, FHA and CRDA were assessed for the EPU reactor operating domain (i.e., the EPU core power level with ECCS evaluation uncertainty factor applied, and other parameter changes that would be affected by EPU operations) to confirm that the EPU doses remained within regulatory limits. Loss-of-Coolant Accident

The post-LOCA doses at the exclusion area bounda ry (EAB), low population zone (LPZ), CR, and TSC were analyzed for EPU conditions. The analysis was performed based on plant operation at 102% of the EPU power level of 4,408 MWt. The EPU core inventory was used. The analysis methods were not changed from those used in License Amendment No. 145. Subsequent to License Amendment No. 145, the LOCA analysis was updated under 10 CFR 50.59 rules to reflect minor changes to de sign inputs. Those updated design inputs were confirmed to remain applicable or minimally affected for EPU conditions. The SP pH response is affected by a modification to the SLCS in conjunction with the EPU. This modification increases the boron-10 enrichme nt of the contents of the SLC tank while reducing the sodium pentaborate concentration and injection rate. The final SLCS design will deliver sufficient sodium pentaborate to the SP to maintain a pH greater than 7.0 thirty days post-LOCA taking into consideration incr eased acid production due to EPU radiation environments. The EPU post-LOCA EAB, LPZ, CR, and TSC doses were determined to be within the applicable regulatory limits. The results and regulatory criteria are summarized in Table 2.9-7. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-429 Fuel Handling Accident The post-FHA EAB and CR doses were analyzed for EPU conditions. The analysis was performed based on plant operation at 102% of the EPU power level of 4,408 MWt. The EPU core inventory was used. The analysis methods were not changed from those used in License Amendment No. 145. Subsequent to License Amendment No. 145, the FHA analysis was updated under 10 CFR 50.59 rules to reflect minor changes to design inputs. Those updated design inputs were confirmed to remain applicable or minimally affected for EPU conditions. The EPU post-FHA EAB and CR doses were determin ed to be within the applicable regulatory limits. The results and regulatory criteria are summarized in Table 2.9-8.

Control Rod Drop Accident

The post-CRDA EAB, LPZ, and CR doses were anal yzed for EPU conditions. The analysis was performed based on plant operation at 102% of the EPU power level of 4,408 MWt. The EPU core inventory was used. The analysis methods were not changed from those used in License Amendment No. 145. Subsequent to License Amendment No. 145, the CRDA analysis was updated under 10 CFR 50.59 rules to reflect minor changes to design inputs. Those updated design inputs were confirmed to remain applicable or minimally affected for EPU conditions. The EPU post-CRDA EAB, LPZ, and CR doses were determined to be within the applicable regulatory limits. The results and regulatory criteria are summarized in Table 2.9-9. Conclusion The Alternative Source Term accident analyses ha ve been reviewed in support of the proposed EPU. Entergy concludes they adequately account for the effects of the proposed EPU. Entergy further concludes that the plant site and the dose-mitigating ESFs remain acceptable with respect

to the radiological consequences of postulated DB As because, as set forth above, the calculated TEDE at the EAB, at the LPZ outer boundary, and in the control room meet the exposure guideline values specified in 10 CFR 50.67 and GDC-19. Therefore, Entergy has determined the Alternative Source Term License Amendment is acceptable with respect to the radiological consequences of DBAs following an EPU.

2.9.3 Additional

Review Areas (Radiological Consequences Analyses) Upon receipt of License Amendment No. 145, the following accidents discussed in UFSAR Chapter 15 were updated to address the AST a nd/or the TEDE acceptance criteria: Pressure Controller Failure, Misplaced Bundle Accident, Recirculation Pump Seizure Event, MSLB, and MSIVC. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-430 The original licensing basis was retained for accidents associated with radwaste system failures or malfunctions, specifically: Offgas System Leak or Failure, Radioactive System Waste System Leak or Failure, and Liquid Radwaste Tank Failure. As discussed in the UFSAR, a temporary departure from nucleate boiling and the subsequent assumption of fuel failure may occur for infrequent events. This fuel failure is currently assumed for the Pressure Controller Failure (conservatively estimated to be core wide), Misplaced Fuel Bundle (5 bundles - the misp laced bundle and the 4 surrounding bundles), and the Recirculation Pump Seizure Event (conservatively estimated to be core wide). The EPU evaluation was performed based on plant opera tion at 102% of the EPU power level of 4,408 MWt. The EPU core inventory was used. The analysis methods were not changed from those outlined in the UFSAR. The design inputs were confirmed to remain applicable or be minimally affected for EPU conditions. The EP U EAB, LPZ, and CR doses for the Pressure Controller Failure and the Recirculation Pump Seizure were determined to be within the applicable regulatory limits. The results and regulatory criteria are summarized in Tables 2.9-12 and 2.9-17. The generic analysis presented in the UFSAR for the Misplaced Fuel Bundle bounds EPU operation. As discussed in the UFSAR, the MSLB and the MSIVC events are based on reactor coolant and steam releases assuming maximum iodine spiking permitted by the plant TSs. The analysis methods were not changed from those outlined in the UFSAR. The design inputs were confirmed to remain applicable or minimally affected for EPU conditions. The EPU will have a minor effect on the relative isotopic composition in the postulated reactor coolant and steam releases; however, the effect on the dose consequences will be insignificant. The EPU post-accident doses for the MSLB and the MSIVC events were determined to be within the applicable regulatory limits. The results and regulatory criteria are summarized in Tables 2.9-10, 2.9-11, and 2.9-13. The post-accident doses were estimated for EPU conditions for Offgas System Leak or Failure and Liquid Radwaste Tank Failure. The evaluation was performed based on plant operation at 102% of the EPU power level of 4,408 MWt. The analysis methods were not changed from those outlined in the UFSAR. The design inputs were confirmed to remain applicable or be minimally affected for EPU conditions. The EPU post-accident doses for Offgas System Leak or Failure and Liquid Radwaste Tank Failure were determined to be within the applicable regulatory limits. The results and regulatory criteria are summarized in Tables 2.9-14 and 2.9-16. Chapter 15.7.2 of the UFSAR addresses the worst case rupture of the Radioactive Liquid Waste System Leak or Failure. The UFSAR reports that the dose consequences for the failure of the Evaporator Bottoms Tank (EAB dose of 1.25 Rem thyroid and negligible whole body (WB); currently abandoned in place), remains bounding for all other radwaste tank failures including the current limiting radwaste tank failure (i.e., the equipment drain collector tank). NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-431 The key bases and assumptions utilized to evaluate the failure of the Equipment Drain Collection Tank at the CLTP are as follows: The radioisotope inventory in liquid radwaste system is based on design primary coolant activity concentrations and a core power level of 3,833 MWt. Only radioiodine isotopes are released because noble gases are not present and particulate radioisotopes will not become airborne. The entire airborne iodine inventory is assumed to be in the elemental chemical species. No operator mitigation is assumed. Instantaneous release is assumed. No credit is taken for partition, filtration, holdup, or dilution of iodine once it is released from the failed tank. The EPU evaluation was performed based on plant operation at 102% of the EPU power level of 4,408 MWt. The analysis methods were not changed from those outlined above. The design inputs were confirmed to remain applicable or minimally affected for EPU conditions. The EPU post-accident doses for the Equipment Drain Collection Tank were determined to be within the applicable regulatory limits and continue to remain within the values reported in the UFSAR for the Radioactive Liquid Waste System Leak or Fa ilure. The results and regulatory criteria are summarized in Table 2.9-15. In conclusion, the dose consequences of the a dditional DBAs applicable to the GGNS licensing design basis remain within applicable regulatory limits following EPU.

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2-432 Table 2.9-1 Total Activity Levels Item Parameter Unit Calculated EPU Value Design Basis Value EPU to Design Basis Value Comparison 1 Activity concentrations of principal radionuclides in fluid streams for normal operation N/A Table 2.9-2 -- Table 2.9-2 contains the EPU calculated

radionuclide concentrations in reactor water and steam for GGNS. 2 Total fission product offgas source term µCi/sec after30 min 6.4E+04 1.0E+05 EPU value is bounded with adequate margin by design basis value. See Table 2.9-3. 3 Total fission product activity concentration in reactor water µCi/g 3.1E-01 2.6E+00 EPU fission product activity levels in the reactor water remain a fraction (12%) of the design basis fission product activity See Table 2.9-4. 4 Total activated corrosion product activity concentration

in reactor water µCi/g 6.21E-02 6.15E-02 The total activated corrosion product activity is 1% greater than design basis levels.* See Table

2.9-5. 5 Total coolant activation product concentration in reactor water µCi/g 4.9E+01 5.0E+01 EPU value is bounded by design basis value. See Table 2.9-6. 6 Total coolant activation product concentration in steam µCi/g 2.5E+02 2.5E+02 EPU value is equal to design basis value. See Table 2.9-6. Note:

Although EPU value is not bounded by the design basis value, the total activated corrosion product activity concentration is much smaller than the fission product concentration. The sum of the activated corrosion product activity and the fission product activity (3.7E-01 Ci/g) remains a fraction (14%) of the total design basi s activity in reactor water (2.6E+00 Ci/g).

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2-433 Table 2.9-2 Activity Concentrations of Principal Radionuclides in Fluid Streams for Normal EPU Operation Reactor Water Steam Reactor Water Steam Ci/g Ci/g Ci/g Ci/g Class 1 Class 6 Kr-83m 9.1E-04 Na-24 9.2E-03 9.2E-06 Kr-85m 1.6E-03 P-32 1.9E-04 1.9E-07 Kr-85 5.0E-06 Cr-51 5.6E-03 5.6E-06 Kr-87 5.5E-03 Mn-54 6.6E-05 6.6E-08 Kr-88 5.5E-03 Mn-56 4.4E-02 4.4E-05 Kr-89 3.4E-02 Fe-55 9.4E-04 9.4E-07 Kr-90 7.5E-02 Fe-59 2.8E-05 2.8E-08 Kr-91 9.1E-02 Co-58 1.9E-04 1.9E-07 Kr-92 9.1E-02 Co-60 3.7E-04 3.7E-07 Kr-93 2.4E-02 Ni-63 9.4E-07 9.4E-10 Kr-94 5.9E-03 Ni-65 2.6E-04 2.6E-07 Kr-95 5.5E-04 Cu-64 2.8E-02 2.8E-05 Kr-97 3.6E-06 Zn-65 1.9E-03 1.9E-06 Xe-131m 3.9E-06 Sr-89 9.4E-05 9.4E-08 Xe-133m 7.5E-05 Sr-90 6.6E-06 6.6E-09 Xe-133 2.1E-03 Y-90 6.6E-06 6.6E-09 Xe-135m 7.0E-03 Sr-91 3.7E-03 3.7E-06 Xe-135 6.0E-03 Sr-92 8.8E-03 8.8E-06 Xe-137 3.9E-02 Y-91 3.7E-05 3.7E-08 Xe-138 2.3E-02 Y-92 5.3E-03 5.3E-06 Xe-139 7.5E-02 Y-93 3.7E-03 3.7E-06 Xe-140 8.0E-02 Zr-95 7.5E-06 7.5E-09 Xe-141 6.5E-02 Zr-97 5.5E-06 5.5E-09 Xe-142 1.9E-02 Nb-95 7.5E-06 7.5E-09 Xe-143 3.2E-03 Mo-99 1.9E-03 1.9E-06 Xe-144 1.5E-04 Tc-99m 1.9E-02 1.9E-05 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-434 Reactor Water Steam Reactor Water Steam Ci/g Ci/g Ci/g Ci/g Class 2 Ru-103 1.9E-05 1.9E-08 I-131 3.5E-03 5.6E-05 Rh-103m1.9E-05 1.9E-08 I-132 5.3E-02 8.0E-04 Ru-106 2.8E-06 2.8E-09 I-133 4.7E-02 7.4E-04 Rh-106 2.8E-06 2.8E-09 I-134 8.6E-02 1.7E-03 Ag-110m9.4E-07 9.4E-10 I-135 4.6E-02 7.3E-04 Te-129m 3.7E-05 3.7E-08 Class 3 Te-131m 9.3E-05 9.3E-08 Rb-89 4.1E-03 4.1E-06 Te-132 9.3E-06 9.3E-09 Cs-134 2.8E-05 2.8E-08 Ba-139 8.6E-03 8.6E-06 Cs-136 1.8E-05 1.8E-08 Ba-140 3.7E-04 3.7E-07 Cs-137 7.4E-05 7.4E-08 Ba-141 8.4E-03 8.4E-06 Cs-138 8.3E-03 8.3E-06 Ba-142 5.0E-03 5.0E-06 Ba-137m 7.4E-05 7.4E-08 La-140 3.7E-04 3.7E-07 Class 4 Ce-141 2.8E-05 2.8E-08 N-13 4.0E-02 3.5E-02 Ce-143 2.8E-05 2.8E-08 N-16 4.8E+01 2.5E+02 Ce-144 2.8E-06 2.8E-09 N-17 7.2E-03 1.0E-01 Pr-143 3.7E-05 3.7E-08 O-19 5.6E-01 1.0E+00 Pr-144 2.8E-06 2.8E-09 F-18 3.2E-03 2.0E-02 Nd-147 2.8E-06 2.8E-09 Class 5 W-187 2.8E-04 2.8E-07 H-3 1.0E-02 1.0E-02 Np-239 7.5E-03 7.5E-06 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-435 Table 2.9-3 EPU Noble Gas Radionuclide Source Term and Design Basis Comparison Isotope Steam, Valid for t=30 min EPU noble gas source term Design basis noble gas source term (UFSAR Table 11.1-1) Class 1 Ci/sec Ci/sec Kr-83m 1.8E+03 2.9E+03 Kr-85m 3.5E+03 5.6E+03 Kr-85 1.2E+01 1.0E+01 Kr-87 1.0E+04 1.5E+04 Kr-88 1.2E+04 1.8E+04 Kr-89 1.1E+02 1.8E+02 Xe-131m 9.3E+00 1.5E+01 Xe-133m 1.8E+02 2.8E+02 Xe-133 5.0E+03 8.2E+03 Xe-135m 4.3E+03 6.9E+03 Xe-135 1.4E+04 2.2E+04 Xe-137 4.1E+02 6.7E+02 Xe-138 1.3E+04 2.1E+04 Total 6.4E+04 1.0E+05

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2-436 Table 2.9-4 EPU Fission Product Reactor Water Comparisons to Design Basis Values Isotope EPU Analysis ValuesCi/gm Design Basis Values (UFSAR Tables 11.1-2 and 11.1-3)Ci/gm I-131 3.5E-03 1.4E-02 I-132 5.3E-02 1.3E-01 I-133 4.7E-02 9.2E-02 I-134 8.6E-02 2.5E-01 I-135 4.6E-02 1.3E-01 Cs-134 2.8E-05 1.4E-04 Cs-136 1.8E-05 9.2E-05 Cs-137 7.4E-05 2.1E-04 Cs-138 8.3E-03 2.1E-01 Sr-89 9.4E-05 2.7E-03 Sr-90 6.6E-06 2.0E-04 Sr-91 3.7E-03 6.5E-02 Sr-92 8.8E-03 1.1E-01 Zr-95 7.5E-06 3.5E-05 Zr-97 5.5E-06 2.9E-05 Nb-95 7.5E-06 3.6E-05 Mo-99 1.9E-03 2.0E-02 Tc-99m 1.9E-02 7.6E-01 Ru-103 1.9E-05 1.7E-05 Ru-106 2.8E-06 2.3E-06 Te-129m 3.7E-05 3.0E-05 Te-132 9.3E-06 1.2E-02 Ba-139 8.6E-03 1.7E-01 Ba-140 3.7E-04 7.7E-03 Ba-141 8.4E-03 2.0E-01 Ba-142 5.0E-03 1.9E-01 Ce-141 2.8E-05 3.4E-05 Ce-143 2.8E-05 3.1E-05 Ce-144 2.8E-06 3.0E-05 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-437 Isotope EPU Analysis ValuesCi/gm Design Basis Values (UFSAR Tables 11.1-2 and 11.1-3)Ci/gm Pr-143 3.7E-05 3.3E-05 Nd-147 2.8E-06 1.2E-05 Np-239 7.5E-03 2.1E-01 Total 3.1E-01 2.6E+00

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2-438 Table 2.9-5 EPU Activated Corrosion Product Reactor Water Comparisons to Design Basis Values Isotope EPU Analysis ValuesCi/gm Design Basis Values (UFSAR Table 11.1-5) Ci/gm Na-24 9.2E-03 2.0E-03 P-32 1.9E-04 2.0E-05 Cr-51 5.6E-03 5.0E-04 Mn-54 6.6E-05 4.0E-05 Mn-56 4.4E-02 5.0E-02 Fe-59 2.8E-05 8.0E-05 Co-58 1.9E-04 5.0E-03 Co-60 3.7E-04 5.0E-04 Ni-65 2.6E-04 3.0E-04 Zn-65 1.9E-03 2.0E-06 Ag-110m 9.4E-07 6.0E-05 W-187 2.8E-04 3.0E-03 Total 6.21E-02 6.15E-02

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2-439 Table 2.9-6 EPU Coolant Activation Product Reactor Water and Steam Comparisons to Design Basis Values Isotope EPU Analysis Values Ci/gm Design Basis Values Ci/gm EPU Analysis Values Ci/gm Design Basis Values Ci/gm Reactor Water Steam N-13 4.0E-02 7.1E-01 3.5E-02 1.5E-03 N-16 4.8E+01 4.8E+01 2.5E+02 2.5E+02 N-17 7.2E-03 1.3E-02 1.0E-01 3.5E-02 O-19 5.6E-01 1.2E+00 1.0E+00 5.9E-01 F-18 3.2E-03 4.8E-02 2.0E-02 4.4E-04 Total 4.9E+01 5.0E+01 2.5E+02 2.5E+02 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-440 Table 2.9-7 LOCA Radiological Consequences TEDE Dose (REM) Receptor Location CR EAB LPZ TSC 1 Calculated Dose CLTP 2 3.69 8.70 5.15 3.69 Calculated Dose EPU 3 4.24 10.01 5.92 4.24 Allowable TEDE Limit 4 5 25 25 5 Table 2.9-8 FHA Radiological Consequences TEDE Dose (REM) Receptor Location CR EAB LPZ Calculated Dose CLTP 2 2.80 2.64 N/A Calculated Dose EPU 3 3.14 3.12 N/A Allowable TEDE Limit 4 5 6.3 N/A

Table 2.9-9 CRDA Radiological Consequences TEDE Dose (REM) Receptor Location CR EAB LPZ Calculated Dose CLTP 2 2.62E-01 1.51E-01 7.23E-02 Calculated Dose EPU 3 2.91E-01 1.51E-01 7.36E-02 Allowable TEDE Limit 4 5 6.3 6.3

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2-441 Table 2.9-10 MSLBA Pre-Incident Iodine Spike Radiological Consequences TEDE Dose (REM) Receptor Location CR EAB LPZ Calculated Dose CLTP 2 3.01 2.39 N/A Calculated Dose EPU 3 3.01 2.39 N/A Allowable TEDE Limit 4 5 25 25 Table 2.9-11 MSLBA Equilibrium Iodine Co ncentration Radiological Consequences TEDE Dose (REM) Receptor Location CR EAB LPZ Calculated Dose CLTP 2 0.153 0.123 N/A Calculated Dose EPU 3 0.153 0.123 N/A Allowable TEDE Limit 4 5 2.5 2.5 Table 2.9-12 Pressure Controller Failure Radiological Consequences TEDE Dose (REM) Receptor Location CR EAB LPZ Calculated Dose CLTP 2 3.39 2.28 0.52 Calculated Dose EPU 3 3.74 2.43 0.56 Allowable TEDE Limit 4 5.0 2.5 2.5

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2-442 Table 2.9-13 MSIVC Radiological Consequences TEDE Dose (MREM) at EAB Calculated Dose CLTP 2 0.083 Calculated Dose EPU 3 0.083 10 CFR 20 TEDE Limit 4 100 Table 2.9-14 Offgas System Leak or Failure Radiological Consequences Calculated Dose (REM) Receptor Location CR EAB LPZ Calculated Dose CLTP 2 0.124 TEDE 1.68 WB 0.384 WB Calculated Dose EPU 3 0.143 TEDE 1.93 WB 0.442 WB Regulatory Limits 4 5 TEDE (GDC-19) 6 (10 CFR 100) 6 (10 CFR 100) Table 2.9-15 Radioactive Liquid Waste System Leak or Failure Radiological Consequences EAB (REM) Thyroid WB Calculated Dose CLTP 2 2.47E-01 4.74E-03 Calculated Dose EPU 3 2.90E-01 5.56E-03 Dose Limit 4 30 (10 CFR 100) 2.5 (10 CFR 100)

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2-443 Table 2.9-16 Liquid Radwaste Tank Failure (Release to Groundwater) Radiological Consequences Radioactive Activity Concentration Reaching the River CLTP 2 Concentration < 10-20 µCi/cc (essentially zero) EPU 3 Concentration < 10-20 µCi/cc (essentially zero), i.e., no significant increase Table 2.9-17 Recirculation Pump Seizure Radiological Consequences TEDE Dose (REM) Receptor Location CR EAB LPZ Calculated Dose CLTP 2 3.72 1.886 0.957 Calculated Dose EPU 3 4.28 2.17 1.10 Allowable TEDE Limit 4 5.0 2.5 2.5 Notes for Tables 2.9-7 through 2.9-17: 1. The TSC at GGNS is located in the CR. Consequen tly, TSC doses will always be identical to CR doses at GGNS. 2. 3,910 MWt (100.3% of rated 3,898 MWt). Doses presented reflect AOR. 3. 4,496 MWt (includes 2% margin of uncertainty)

4. RG 1.183 Table 6 or as identified in accordance with current licensing basis.

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2-444 2.10 Health Physics 2.10.1 Occupational and Public Radiation Doses Regulatory Evaluation Entergy conducted its review in this area to as certain what overall effects the proposed EPU will have on both occupational and public radiation doses and to determine the necessary steps to ensure that any dose increases will be maintained ALARA. The review evaluated any increases in radiation sources and how such increases ma y affect plant area dose rates, plant radiation zones, and plant area accessibility. Entergy ev aluated how personnel doses needed to access plant vital areas following an accident are affect ed. Entergy considered the effects of the proposed EPU on nitrogen-16 levels in the plant and any effects this increase may have on

radiation doses outside the plant and at the site boundary from skyshine. Entergy also considered the effects of the proposed EPU on plan t effluent levels and any effect this increase may have on radiation doses at the site bounda ry. The regulatory acceptance criteria for

occupational and public radiation doses are based on 10 CFR Part 20 and GDC-19. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. Occupational and public radiation doses are di scussed and described in UFSAR Section 12.4, "Dose Assessment."

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Sections 8.5 and 8.6 of th e CLTR address the effect of CPPU on the Radiation Sources in the Reactor Core and in the Reactor Coolant. The results of this evaluation are described below. 2.10.1.1 Increases in Radiation Sources The EPU maximum authorized power level of 4,408 MWt is a 13% increase over the CLTP level of 3,898 MWt. All reported percentage increas es throughout Section 2.10.1 are relative to CLTP unless otherwise specified.

The radiation sources that are affected by EPU are the sources generated inside the reactor during power operation. The production rate of reactor-generated sources is approximately NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-445 proportional to the core power level because the fission rate and neutron flux are proportional to the power level. Therefore, EPU is assumed to increase the production rate of reactor generated sources by approximately 13% from the CLTP pr oduction rates. Specifically, an increase of approximately 13% (from the CLTP values) is assumed for the production rate of fission products, the production rate of activation products (due to interaction of the neutron flux with the reactor water and waterborne corrosion products), and radiation from the fission process itself (neutron and gamma). The effects of EPU on doses were evaluated by assuming that the increase in fuel fission product inventory is proportional to EPU (i.e., 13%), and that the increase in reactor water fission and activation product concentrations is approximate ly proportional to the EPU increased production rate (i.e., 13%). EPU increases the flow rates in the MS system, the Condensate Cleanup system, and the FW system, and also increases the moisture carryover fraction in the reactor steam. However, the effect of the above changes on the reactor water fission and activation product concentrations was found to not be significant. While the increased production rate of react or water activation products (e.g., N-16) is approximately proportional to the increased power, the concentration in the steam at the reactor nozzle remains nearly constant because the increase in activation production is balanced by the increase in steam flow. However, because of the increased steam flow, the transit times to Turbine Building equipment are reduced, which reduces the decay time for short-lived isotopes such as N-16. Due to a greater steam flow rate, the total activity appearance rate (µCi/sec) at a component will increase with EPU. The activity appearance rate (µCi/sec) is defined as the activity mass concentration (µCi/gm) times the mass flow rate (gm/sec). As a result, even though the N-16 concentration in MS does not cha nge appreciably, the total inventory of N-16 does increase in Turbine Building equipment su ch as the turbines, cross-over piping, and condenser. The increase in the N-16 related radiation exposure is applied to all steam and condensate bearing systems/component rooms located in the Turbine Building complex. The limiting increase in the N-16 inventory upstream of the MC is 12% due to the reduction in steam transit time. The N-16 inventory in the MC is e xpected to increase by 27%. This is due to an increase of approximately 12% due to the reduced transit time and 13% due to the increase in steam flow rate to the MC (i.e., 1.12 x 1.13 = 1.27). The N-16 source increase in SJAEs, Offgas Pre-heaters, Catalytic Recombiners, and Offgas Condensers is estimated to be 43% due to a further reduction in transit time. The moisture carryover with EPU is expected to be bounded by 0.069%. This is a factor 2.76 times greater than the CLTP moisture carryove r of 0.025%. The principal effect of this will be to change the distribution of the radionuclide activity in the waste streams. Specifically, a small fraction of the activity that would be accumulated on the RWCU system resin will be accumulated on the condensate cleanup media (i.e., f ilters and resin). As a consequence of the increased moisture carryover, the condensate demi neralizer source is expected to increase by 32% with EPU. The activity inventory in the condensate demineralizers is small compared to NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-446 that in the RWCU demineralizers. The radionuclide inventory on the condensate cleanup media will remain a small fraction of the design basis values. In summary, and except as noted above, EPU is assumed to increase the reactor generated radiation sources by up to 13%, and the N-16 sour ces will increase by up to 27%. As discussed above, the EPU increase factor can be higher for some localized radiation sources such as the upstream components of the Offgas System and the condensate demineralizers. However, as discussed below, because of the margin in the OLTP design basis sources for reactor water and steam at the reactor nozzle, the corresponding EP U sources continue to be bounded by the OLTP design basis. The fission product activity in the reactor water is the result of a combination of tramp uranium and minute releases from the fuel rods. Even with an assumed increase of 13% due to EPU, the fission product concentrations in reactor water remain less than 12% of the OLTP design basis sources, and the OLTP design basis sources remain bounding. The non-noble gas fission product activity in reactor steam is a result of carry-over from the reactor water. The noble gas activity release rate from the reactor vessel to the Offgas System

will increase with EPU. Even with an assumed increase of 13% due to EPU, the noble gas release rate from the condenser offgas after 30 minutes of decay, remains below the OLTP design basis of 0.1 Ci/sec after 30 minutes of decay. Furthermore, the actual releases to the environment are much less than the annual releases estimated using NUREG-0016 (Reference 87) based expected coolant concentrations. For example, for the year 2008, the actual noble gas releases were less than 6% of the noble gas annual release estimated using NUREG-0016 (Reference 87) based expected coolant concentrations. 2.10.1.2 Occupational and On-Site Radiation Exposures The CLTR topics regarding occupational and on-site radiation exposures are listed below. GGNS meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition GGNS Result Normal Operational Radiation Levels [[ Meets CLTR Disposition Post-Operation Radiation Levels

Meets CLTR Disposition Post-Accident Radiation Levels

     ]] Meets CLTR Disposition

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2-447 2.10.1.2.1 Occupational and On-Site Radiation Exposures The CLTR states that normal operational radiation levels will increase slightly as a result of EPU. GGNS shielding has been designed based on extremely conservative design basis source terms for water and steam sources. Calculated sources for the EPU are bounded by the original design. Inside containment, the radiation levels near the reactor vessel are assumed to increase by 13% with respect to CLTP. Due to EPU fuel management schemes, there are localized areas adjacent to the reactor vessel that may experience neutron and gamma flux spikes that are greater than the percentage of the uprate. However, the r eactor vessel is inaccessible during operation, and because of the margin in the sh ielding around the reactor vessel, the increase in radiation level due to EPU will not measurably increase occupational doses during power operation. The radiation levels due to SF are assumed to increase by 13%. Radiation exposures in accessible areas adjacent to the sides or bottom of the SFP are expected to be within the allowable dose rate limit of the existing radiation zone designation. Expected increases in these values will occur primarily during fuel handing operations during the RFO. With respect to areas adjacent to the SFP, it is estimated that a core thermal power increase of 13% would result in a 13% increase of dose rates related to SFP operations. Similarly, the increased dose rates at the SFP could potentially have proportionally increased dose rates in accessible areas adjacent to the SFP. Genera lly, the dose rates on the refuel floor are 1 mrem/hr (1.5 mrem during refueling activities), and the doses rates in the accessible areas adjacent to the SFP are also 1 mrem/hr. Therefore, zoning in these areas is not expected to change as a result of EPU conditions. Dose rates at the surface of the pool are primarily due to the presence of radionuclides suspended in the cooling water. The frequency of the backwash and precoats of the fuel pool demineralizers c ontrol these dose rates. Radiation protection surveys in accordance with the current radiation protection program will ensure that refueling activities will continue to be appropriately monitored during these activities. The N-16 dose rate is expected to increase in proportion to the N-16 inventory in the components. This increase is due to a combination of the increase in steam flow rate and the reduced transit times from the reactor vessel nozzle to the component. The N-16 dose rate is expected to increase by approximately 12% upstream of the MC, and up to 27% in the MC. However, the areas with a significant N-16 inventory are heavily shielded and not routinely

occupied, and the N-16 is only present during operation. Outside containment, radiation shielding was sp ecified using the OLTP design basis radiation sources, which are extremely conservative. Outside containment, it is the actual operating sources, not the design basis sources, which may increase. The actual OLTP N-16 concentration

at the reactor nozzle is less than the OLTP design basis. With the increased steam flow rates for EPU, the dose rates in areas of the plant affected by N-16 may increase by a maximum of 27%. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-448 Other fission and activation product concentrations, and the resultant dose rates, may increase up to 13% under EPU operating conditions. The tota l activity associated with the condensate cleanup system may be greater, but the activity is distributed among the prefilters and demineralizers, so the increase in dose rate is e xpected to be less than 32%. In addition, the dose rate increase is estimated to be approximately 43% adjacent to Offgas system components such

as the SJAEs, Offgas Pre-heaters, Catalytic Recombiners, and Offgas Condensers. There is sufficient margin in the GGNS design to ensure that shielding is adequate to maintain occupational and on-site doses ALARA. Based on the 2008 Edition of NUREG-0713 (Reference 88), the total occupational dose between 2003 and 2008 ranged from 31 to 178 person-rem, so the annual occupational doses can vary by about 600%. For example, for the year 2007, the average dose per exposed worker at GGNS was approximately 100 mrem, which is 2% of the limit allowed by 10 CFR 20.1201. Annual cumulative occupational radiation exposure may increase by as much as 13% due to EPU, which is well within the historical variation in station annual cumulative exposure. Individual worker exposure will continue to be maintained within acceptable limits by the ALARA program, which controls access to radiation areas. Except for some localized areas, the radiation exposure in all affected plant areas is not

expected to increase greater than the propos ed 13% EPU. The post-EPU increase in the radiation exposure in various areas housing the steam components remains within the radiation zone allowable dose rate limits. Therefore, no additional measures are required to maintain the plant exposure ALARA. The post-EPU plant operation and maintenance activities will be controlled by the existing radiation protection and ALARA procedures. When the reactor coolant chemistry is changed from Normal Water Chemistry to moderate HWC, the crud in the RPV (primarily on the fuel) can restructure, causing crud to be released into the water during power changes and upon unit shutdowns. Crud bursts as a result of HWC have been assessed, and steps to reduce personnel radiation dose could include maximizing RWCU and Fuel Pool demineralizer operation, additional temporary filters (Tri-Nuc), bleed-and-feed operations, and temporary shielding. EPU is not expected to exacerbate crud restructuring and release. GGNS has been operating with HWC since 1998 without undue radiation hazard to plant personnel and the public. As a result of HWC operation, some shielding modifications were made to alleviate the increased N-16 source. Also, a number of radiation postings were changed to control access to radiation areas. Other than the 27% increase attributed to the EPU, no additional increase in the N-16 doses are expected due to the continued use of HWC. The existing radiation zoning design (e.g., the maximum designed dose rates for each area of the plant) will not change as a result of the increased dose rates associated with EPU. A review was performed to identify areas where the design basis radiation doses could result in a change in the radiation dose zone designation as a result of EPU (Table 2.10-1). Anticipated EPU radiation levels in various areas of the plant were estimated by multiplying current measured dose rates, shown in Table 2.10-1, by the expected increase in radiation levels following EPU for the NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-449 respective plant areas. Based on this review, it was concluded that no changes in the radiation zone designations or shielding requirements will need to be made as a result of EPU. In summary, although normal operation radiation le vels will increase as a result of EPU, the GGNS shielding design is based on a conservative source term. The changes in actual source concentrations due to EPU are well-bounded by the original design. Analyses have confirmed that operation under EPU conditions will have a ne gligible effect on occupational and on-site radiation exposure. Therefore, occupational and on-site radiation exposures meet all CLTR dispositions. 2.10.1.2.2 Post-Operation Radiation Levels The CLTR states that, for EPU, normal operation radiation levels increase slightly. Radiation levels in most areas of the plant during shutdown operation will increase by no more than the percentage increase in power level. In a few areas near the reactor water piping and liquid radwaste equipment, the increase could be slightly higher. At GGNS, the use of area monitoring, pre-job briefs, and appropriate access c ontrols ensure the protection of personnel and the maintenance of doses ALARA. Therefore, post-operation radiation levels meet all CLTR dispositions. 2.10.1.2.3 Post-Accident Radiation Levels The CLTR states that post-accident radiation leve ls will increase slightly as a result of EPU. However, due to conservative CLTP analyses, the post-accident radiation levels are not expected to increase at EPU RTP compared to the estimated post-accident doses rates for CLTP conditions. The post-accident levels specified for CLTP conditions are bounding for the EPU conditions. The EPU post-accident radiation leve ls have no adverse effect on safety-related plant equipment. An analysis for NUREG-0737, Item II.B.2 (Reference 89), post-accident mission doses has been performed for GGNS. A review of the doses associated with post-accident missions to vital areas was conducted to determine the effect of EPU. The times required for transit to and work in vital access areas are not changed with EPU. Due to conservative CLTP evaluation methods, the current operator dose estimates remain unchanged for the EPU. The dose values associated with vital areas can

be found in Table 2.10-2. All of the doses are within the limits of GDC-19. Therefore, post-accident radiation levels meet all CLTR dispositions. 2.10.1.2.4 Public and Off-Site Radiation Exposures GGNS meets all CLTR dispositions. The CLTR t opics regarding public and off-site radiation exposures are listed below.

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2-450 Topic CLTR Disposition GGNS Result Off-Site Plant Gaseous Emissions [[ Meets CLTR Disposition Plant Skyshine From the Turbine

     ]] Meets CLTR Disposition The CLTR states that for EPU, normal operation gaseous activity levels increase slightly, while the level of N-16 in the turbine increases in proportion to the rated steam flow.

The sources responsible for off-site doses in crease by varying factors depending upon the basis for each source. The EPU increase factor in cludes a power uncertain ty factor of 1.02, in accordance with NUREG-0016, Revision 1 (Reference 87). The primary sources of normal off-site doses at GGNS are airborne releases, liquid effluent releases from the radwaste system, and gamma skyshine from N-16 in the plant turbines and some unshielded MS piping. Implementation of EPU could increase the compone nts of off-site doses due to releases of airborne radioactivity by up to 15% for airborne doses due to noble gases, and 19% for airborne doses in the particulate and iodine category, where the thyroid is the dominant organ relative to dose. The relative activity increase in FW that could leak in the Turbine Building was designated the limiting factor for the EPU liquid effluent release evaluation. For liquid effluents, the major pathway is fish consumption with the gastro-int estinal-lower large intestine (GI-Lli) being the dominant organ relative to dose. The EPU doses for the liquid pathway are estimated to increase by up to 214% primarily due to the increase in the expected moisture carryover fraction from a CLTP value of 0.025% to an EPU value of 0.069%, as well as the source increase due to EPU. In general, the dose changes due to N-16 in the equipment above grade is the most significant factor in skyshine, although radiation scatter from other sources may be present. The equipment above grade at GGNS includes steam piping, turbines, FW heaters, the upper portions of moisture separators, and the transition between the turbines and condenser. The component of the off-site dose due to N-16 skyshine could in crease by 29% (This is based on the increase in steam flow rate and velocity and is described in Section 2.10.1.1 plus the effect of a 2% margin

for power uncertainty, i.e. 1.12 x 1.15 = 1.29). The expected post-EPU increase in the in-plant radiation exposure in the Turbine Building complex has a negligible effect on the estimated doses to members of the public. The Turbine Bu ilding concrete shielding and distance between the Turbine Building and off-site boundary are su ch that the post-EPU direct dose contribution from the steam components in the Turbine Building is negligible. The post-EPU N-16 skyshine NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-451 dose rate at the nearest boundary is expected to be near the background radiation level. Therefore, it is a minor contributor to the total estimated doses to members of the public. Although the post-EPU direct dose and skyshine dose contribution to a member of the public (MOP) are negligibly small, a 29% increase in the combined contributions of direct dose and skyshine dose was applied to the MOP doses. Dose value contributions for the primary sources of normal operation off-site doses (all effluent releases, gamma shine, storage and transfer of radioactive materials) to a MOP at CLTP and EPU are provided in Tables 2.10-3 and 2.10-4. The EPU evaluation utilizes: (a) the plant core power operating history during the years 2004 to 2008, (b) the reported gaseous and liquid effluent and dose data, including direct radia tion data during that period, and (c) NUREG-0016 (Reference 87) equations and assumptions, to estim ate the effect of operation at the analyzed EPU core power level of 4,496 MWt, on radioactiv e gaseous and liquid effluents and consequent normal operation off-site doses including direct radiation. The maximum potential percentage increase in coolant activity levels due to the EPU, for each chemical group identified in NUREG-0016, is estimated using the met hodology and equations found in NUREG-0016, Revision 1 (Reference 87). The evaluation is based on a comparison of the change in power level and in RCS parameters (e.g., reactor coolant mass, steam flow rate, reactor coolant cleanup flow rate, flow rate to the condensate demineralizers and moisture carryover) for CLTP and EPU conditions. To estimate an upper bound effect on o ff-site doses, the highest factor found for any chemical group, pertinent to the release pathway, is applied to the average doses previously determined as representative of operation at CLTP conditions extrapolated to 100% licensed core power. This conservative approach is utilized to estimate the maximum potential increase in effluent doses due to the EPU, and to demonstrate that the estimated off-site doses following EPU continue to remain below the regulatory limits of 10 CFR 50 Appendix I and 40 CFR 190. The transport and storage of radioactive materials pathway is a minor contributor to the off-site dose at GGNS. Examination of the off-site thermal luminescence dosimeter (TLD) readings in the time period 2004 to 2008 indicated that radia tion, both skyshine and direct radiation, is linked to plant operation, and plant stationary sources are an insignificant contributor. Currently, the off-site doses from operation of GGNS are well below 10 CFR 50 Appendix I and 40 CFR 190 regulatory guidelines as indicated in Tables 2.10-3 and 2.10-4. As demonstrated in Table 2.10-3 and discussed below, the estimated doses following EPU, although increased, also remain well below the 10 CFR 50 Appendix I and 40 CFR 190 regulatory guidelines. The post-EPU annual off-site WB dose is less than the allowable annual dose limit of 25 mrem in 40 CFR 190. The 40 CFR 190 WB dose to any MOP includes: (a) contributions from direct radiation (including skyshine) from contained radioactive sources within the facility, (b) the total body dose from liquid release pathways, and (c) the WB dose to an individual via airborne pathways. At GGNS, these contributions are estimated to be 4.38 mrem, 0.4 mrem, and 1.22 mrem, respectively, as shown in Tables 2.10-3 and 2.10-4. Thus, the total body dose to NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-452 members of the public is conservatively estimated (locations of peak dose from each contributor are not co-located) to be 6 mrem. This result demonstrates substantial margin to the 40 CFR 190 dose limit of 25 mrem per year. The environmental monitoring program that is in place will continue to ensure that the off-site doses are within regulatory limits and will provide indication should the doses increase above measured background levels. In summary, operation under EPU conditions will have a negligible effect on public and off-site radiation exposure. Therefore, public and o ff-site radiation exposures including gaseous emissions and turbine skyshine maintain substantial margins to regulatory limits. 2.10.1.3 Operational Radiation Protection Program The increased production rates of fission and activation products could increase dose rates from contained sources, surface contamination, and air borne radioactivity. The current operational programs (pre-job briefings, use of supplemental shielding, pre-job decontamination, contamination control practices) will continue to ensure that, with these increases, the occupational doses will continue to remain ALAR A. Individual worker exposure will continue to be maintained within acceptable limits by the ALARA program, which controls access to radiation areas. In summary, the current operational radiation protection programs are capable of controlling, and compensating for, the potential increases in contained sources and surface contamination. Conclusion The effects of the proposed EPU on radiation source terms and plant radiation levels have been reviewed. Entergy concludes the necessary steps have been taken to ensure any increases in radiation doses will be maintained ALARA. Entergy further concludes the proposed EPU meets the requirements of 10 CFR 20 and GDC-19. Th erefore, Entergy finds the proposed EPU acceptable with respect to radiation protection and ensuring occupational radiation exposures will be maintained ALARA. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-453 Table 2.10-1 Current and Anticipated Measured Radiation Fields in Selected Areas Building & Area Description Radiation Zone (mrem/hr) Measured Survey Results (CLTP) (mrem/hr) Expected EPU Increase (%) Anticipated Survey Results (EPU) (mrem/hr) Turbine Building El. 166' (Open Areas) B ( 2.5) 2.0

12% 2.3 Turbine Building El. 133' (Open Areas)

B ( 2.5) 0.1 - 2 12% 0.1 - 2.2 Turbine Building El. 113' (Open Areas) B ( 2.5) 0.1

12% 0.12 Turbine Building (Near LP Turbine) E (> 100) 500 - 3,500 12% 560 - 3,920 Turbine Building (Near Condenser Bays) E (> 100) 8 - 2,500 27% 10 - 3,200 Turbine Building (Near SJAE) E (> 100) 0.4 - 1,620 43% 0.6 - 2,320 Turbine Building (Near Offgas Condensers) E (> 100) 220 - 4,000 43% 315 - 5,720 Turbine Building (Near Moisture Separators) E (> 100) 500 - 3,500 13% 565 - 3,960 Turbine Building (Near Condensate Demineralizers) E (> 100) 2.5 - 100 32% 3 - 135 Turbine Building (near FW Heaters) E (> 100) 200 - 12,000 19% 240 - 14,300 Auxiliary Building (Open Areas)

B ( 2.5) 2.0 13% 2.3 Containment Building (Open Areas) B ( 2.5) 1.0

13% 1.2 Turbine Building (Near Ventilation Charcoal Filters) C ( 15) < 0.1 - 0.1 17% < 0.1 - 0.12 Turbine Building (Near Ventilation HEPA Filters) C ( 15) < 0.1 - 0.1 212% < 0.4 - 0.4 Radwaste Building (General Areas)

B ( 2.5) < 0.1 - 0.5

32% 0.1 - 0.7 Auxiliary Building (Areas Adjacent to SFP During Normal Operations)

B ( 2.5) 0.5

13% 0.6 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-454 Building & Area Description Radiation Zone (mrem/hr) Measured Survey Results (CLTP) (mrem/hr) Expected EPU Increase (%) Anticipated Survey Results (EPU) (mrem/hr) Auxiliary Building (Areas Adjacent to SFP During Refueling) B ( 2.5) 0.1 - 1.5 13% 0.1 - 1.7 Refuel Floor (Auxiliary Building El. 245' Open Area During Normal Operations) B ( 2.5) < 0.1 - 1 13% < 0.1 - 1.2 Refuel Floor (Auxiliary Building El. 245' Above SFP During Refueling) B ( 2.5) < 0.1 - 1 13% < 0.1 - 1.2 Note:

Representative dose rate; localized hot spots exist in these areas.

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2-455 Table 2.10-2 EPU Vital Area Access Mission Doses Location Type of Access EPU Operator Mission Dose (Rem) Remote Shutdown Panel Extended 4 Diesel Generator Buildings Extended 0.12 SGTS Sampling Station Intermittent 0.44 Laboratories Extended 4.38 ADS Air Supply Makeup Connection Intermittent 3.46 ADS Booster Compressor Area Single Entry 1.13 Note: 1. Each one of these areas represents a vital area as described in UFSAR Table 12.6-2.

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2-456 Table 2.10-3 Estimated Annual Doses to Members of the Public Due to Normal Operation Gaseous and Liquid Radwaste Effluents Type of Dose 10 CFR 50 Appendix I Design Objectives CLTP Doses (Normalized to 100% Capacity) EPU Doses Percentage of 10 CFR 50 Appendix I Design Objectives (EPU) Liquid Effluents Dose to total body from all pathways 3 mrem/yr 0.130 mrem/yr 0.401 mrem/yr 13.4% Dose to any organ from all pathways 2 10 mrem/yr 0.296 mrem/yr (GI-Lli) 0.929 mrem/yr (GI-Lli) 9.3% Gaseous Effluents Gamma dose in air 3 10 mrad/yr 0.356 mrad/yr 0.410 mrad/yr 4.1% Beta dose in air 3 20 mrad/yr 0.194 mrad/yr 0.224 mrad/yr 1.1% Dose to total body of an individual 3 5 mrem/yr 1.06 mrem/yr 1 1.22 mrem/yr 24.4% Dose to skin of an individual 3 15 mrem/yr 1.71 mrem/yr 1 1.97 mrem/yr 13.1% Radioiodines and Particulates Released to the Atmosphere Dose to any organ from all pathways 4 15 mrem/yr 0.327 mrem/yr (Thyroid) 0.390 mrem/yr (Thyroid) 2.6% Notes: 1. CLTP Total Body and Skin Dose Rates (annual average) are calculated by averaging the quarterly dose rates that are projected on operation for the year, which are based on the daily maximum instantaneous dose rates averaged during that quarter. 2. With respect to dose due to the liquid effluent pathway, the GI-Lli is the dominant organ. 3. The gamma and beta dose in air, including the dose to the total body and skin of an individual due to gaseous effluents is based on noble gases. 4. With respect to dose due to iodines and particulates (including tritium) in the gaseous effluent pathway, the thyroid is the dominant organ.

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2-457 Table 2.10-4 Direct Shine Annual Dose to Members of the Public CLTP (mrem) EPU (mrem) Percent Increase 40 CFR 190 Limit (mrem) Direct Radiation 3.40 4.38 29% 25 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-458 2.11 Human Performance 2.11.1 Human Factors Regulatory Evaluation The area of human factors deals with programs, procedures, training, and plant design features related to operator performance during normal and accident conditions. The GGNS human factors evaluation was conducted to ensure that operator performance is not adversely affected as a result of system changes made to implement the proposed EPU. The review covered changes to operator actions, human-system interfaces, a nd procedures and training needed for the proposed EPU. The regulatory acceptance criteria for human factors are based on GDC-19, 10 CFR 50.120, 10 CFR 55, and the guidance in GL 82-33. GGNS Current Licensing Basis

NRC GDC for Nuclear Power Plants, Appendi x A of 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971 is applicable to GGNS. GGNS conformance with the GDCs may be found in UFSAR Sections 3.1 and 7.1.2.5. The human factors program is not described in any GGNS licensing basis document; however, it is governed in accordance with Engineering Standard ES-17, "Human Factors Design Criteria."

Technical Evaluation 2.11.1.1 Changes in Emergency and Abnormal Operating Procedures EOPs and the SAPs are being revised to reflect the increase in power level; however, no changes to accident mitigation philosophy will be made to these procedures. The following EOP parameters have been identified as being affected: Heat Capacity Temperature Limit (HCTL) - The EPU will result in additional heat being added to the SP during certain accident scenar ios. The HCTL curve will be revised as a result of the increase in decay heat rejected to the SP. The change is not significant (approximately 1 ûF). Pressure Suppression Pressure (PSP) - The PSP Curve will be revised as a result of the increase in reactor power and in decay heat loading. The change is not significant (<1 psi). Minimum Debris Retention Injection Rate - The Minimum Debris Retention Injection Rate will be revised as a result of the increas e in decay heat loading. The injection flow will increase by approximately 10% of the CLTP flow. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-459 Cold Shutdown Boron Weight - The Cold Shutdown Boron Weight will increase in the equilibrium core design for EPU by approximately 18%. The Hot Shutdown Boron Weight is expected to be affected by an equivalent amount. The EOP/SAP revisions related to this parameter also consider the SLC system boron enrichment modification which mitigates this EPU effect. The planned changes to abnormal operating pr ocedures (AOPs), called ONEPs at GGNS, are outlined below. The ONEPs listed below will be revised to re scale action points associated with reactor power; however, the event mitigation philo sophy will not be changed. Affected procedures include: o 05-1-02-I-2, Turbine and Generator Trips; o 05-1-02-III-5, Automatic Isolations; o 05-1-02-V-5, Loss of Feedwater Heating; o 05-1-02-V-7, Feedwater System Malfunctions; o 05-1-02-V-8, Loss of Condenser Vacuum; and o 05-1-02-V-11, Loss of Plant Service Water. 05-1-02-III-1, Inadequate Decay Heat Removal - Revise the decay heat curves, heat up rates, and temperature related data sheets to reflect the new EPU values. 05-1-02-V-5, Loss of Feedwater Heating - Revise the FW temperature vs. core power curve, which determines the actions to be ta ken in response to the event, to reflect the new EPU values. 05-1-02-V-7, Feedwater System Malfunctions - Add RFP flow limitations following a trip of the other RFP. 2.11.1.2 Changes to Operator Actions Sensitive to Power Uprate Most abnormal events result in automatic plant shutdown (scram). Some abnormal events result in SRV actuation, ADS actuation and/or automatic ECCS actuation. All analyzed events result in safety-related SSCs remaining within design limits. EPU does not change any automatic safety function. The subsequent operator action for maintaining core cooling, containment cooling and safe shutdown for plant safety remains unchanged. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-460 There are no new credited operator actions required as a result of EPU. In addition, the analysis for EPU credits existing manual actions following the same time limits currently credited for CLTP. Several operator actions are evaluated in the license basis for GGNS. The following actions were considered during the EPU technical evaluation: The ATWS analysis assumes operator action in 120 seconds to initiate the SLCS and 660 seconds to initiate RHR SPC. These times do not change for EPU. Long-term DBA LOCA assumes operators initiate containment cooling 30 minutes from

initiation of the event. (UFSAR Section 6.2.2.3) In addition, manual isolation of the unfiltered outside air intake is credited at 20 minutes. (UFSAR Section 15.6.5.5.2) These times do not change for EPU. For an SBO event, actions to establish reactor water level control and reactor pressure control are initiated 2 minutes into the event. This time does not change for EPU. The other 15-, 60- and 120-minute action times to defeat RCIC trips, reduce control room heat loads and reduce DC battery loads remain unchanged for EPU as well. For Control Room Evacuation, there is no change required to the operator action time

and no new operator actions are required. The shutdown from the remote shutdown panel ONEP states that operators should be available at the remote shutdown panel within 13 minutes to initiate depressurization of the RPV and vessel injection with RHR. To depressurize the reactor, the CLTP analysis determined that 6 SRVs were required to be opened within 18 minutes. At EPU conditions, the time available to the operator to open 6 SRVs is 14.3 minutes. The existing procedural requirement satisfies both of these times. The MSLB outside containment analysis assumes the operator begins depressurizing the RPV at the 10 minute mark. This time does not change for EPU. The CST was designed with a minimum volume to allow at least 8 hours of RCIC system operation at a constant reactor pressure to remove reactor decay heat. At EPU conditions, slightly less CST volume is available resulting in RCIC operation for 7.9 hours. This minor reduction in operating time does not affect operator actions. The EOPs are symptom-based procedures that are used for a wide range of accidents and events that challenge operators. These procedures, as written, do not have specific time constraints or time limits associated with their execution. They address plant symptoms/parameters independent of cause which allows for comprehensive actions to mitigate fuel damage and to

prevent radioactive release. EPU conditions will re sult in greater decay heat loads. The actual EOP actions performed by operators are not changed; however, those actions that remove decay NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-461 heat will be influenced. The GGNS simulator will be modified for EPU changes. Decay heat changes and their effects on EOP execution have b een reviewed and appropriate training will be provided to the operators. UFSAR action times that would be executed under the guidance of

EOPs are included in the evaluation of this section. The EPU analysis shows there are no significant changes to the operator action times for EOPs. The ONEPs are event-based procedures. UFSAR action times for events such as loss of AC power (station blackout) and shutdown from the remote shutdown panel (Appendix R fire) have been evaluated and do not change for EPU. In other cases, the procedures are designed so that the severity of the event dictates the time available for the response, ranging from immediate operator actions to more long range response. As with the EOPs discussed above, there are no significant changes to the operator action times for ONEPs. 2.11.1.3 Changes to Control Room Controls, Displays and Alarms The effect on the control room instruments and controls is minimal. There are no changes to these systems/controls that will affect the operato r's ability to interpret, read or respond to the information provided by the updated systems/controls. Plant process computer system operation is not affected by EPU. Changes to the control room are prepared in accordance with the plant design change process. Under this process, a Human Factors engineering review is performed for changes associated with the GGNS control room. The change process also requires an effect review by Operations and Training personnel. Results of these reviews, including simulator effect and training requirements, are incorporated into the engineering change package and tracked to completion by the design change process.

As outlined in Section 2.4.1.3, Technical Specification Instrument Functions for the following instrument and control systems are affected: For MSL High Flow Group 1 Isolation, the analytical trip value remains the same in terms of percent power. The trip value for MSL High Flow Group 1 Isolation in terms of differential pressure is being revised to refl ect the changes associated with the EPU rated thermal power level increase and steam flow increase. The trip value for the Turbine First Stage Pressure Scram Bypass Permissive is being revised to reflect the changes associated with the HP turbine modification and the EPU rated thermal power level increase. The absolute thermal power associated with the Turbine First Stage Pressure Scram Bypass Permissive remains unchanged. The specific first stage pressure associated with this power is being changed. Trip values for APRMs are being revised to reflect the changes associated with the EPU rated thermal power level increase. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-462 The Rod Worth Minimizer (RWM) and the Rod Block Monitor (RBM) setpoints remain at the same value in terms of percent. The absolute power values are being changed accordingly. The following BOP instrument setpoints/controls are affected: The overspeed setpoint on the reactor feedpump turbines is being increased to accommodate the increased speed demand at normal EPU operations. The condensate booster pump low suction pressure trip setpoint is being increased due to the increased condensate booster pump flow rates at EPU conditions. The pressure control system pressure regulator setting is being lowered to provide for the increased steam line pressure drop at EPU steam flow rates. The following control room instruments are affected by EPU: Reactor FW flow and steam flow control room indicating meters and recorders are being modified to increase the usable range. Main Turbine 1st Stage Steam Flow recorder indication is being rescaled. The Load Set and Load Meters on the EHC panel are being replaced. Reactor Feed Pump Turbine speed meters are being rescaled. The Generator Amperage, MegaVARS, and Megawatt indication is being replaced. 2.11.1.4 Changes to the Safety Parameter Display System The purpose of the GGNS safety parameter display system (SPDS) is to continuously display information from which plant safety status can be readily and reliably assessed. The principal function of the SPDS is to aid control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid a degraded core. The information presented on the SPDS displays and the method of presentation remains unchanged for EPU; therefore, SPDS equipment is not being modified for the EPU. The SPDS system also provides procedure based display concepts to support execution of the GGNS EOPs. In conjunction with the changes required to the EOPs for EPU operation, two EOP curves (PSP and HCTL) are being revised, as described in Section 2.11.1.1. Training for the EOP changes is included in the Operator Training Program described in Section 2.11.1.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-463 2.11.1.5 Changes to the Operator Training Program and the Control Room Simulator In accordance with industry standards, the Operations Training Group develops required training on modifications installed that affect plant operati on. Operator training is typically presented in the classroom and on the simulator, as appropriate. The training will focus on the plant modifications, procedure changes, start-up and test requirements, and other aspects of EPU at GGNS. Training demonstrations will be conducted highlighting the changes that affect EOPs and SOPs. As determined by the training analysis process, appropriate classroom, simulator and in-plant training will be conducted prior to power escalation or as required to operate modified systems for plant start up. The simulator will be modified to maintain the required fidelity in accordance with site procedures and ANSI/ANS 3.5 - 1998 (Reference 90). The simulator changes include hardware changes for new and modified instrumentation and controls, software updates for modeling EPU changes and re-tuning of the core physics model for cycle-specific data. Simulator performance will be validated using design analysis data and startup and test data from the EPU project and implementation program. The operator training is being scheduled to start in the 4 th quarter 2011 and continue through startup from the EPU installation RFO. Details of the training will be developed through the modification process and the training development process. Detailed schedules will be developed in accordance with GGNS training procedures. Conclusion The changes to operator actions, human-system interfaces, procedures and training required for the proposed EPU have been evaluated. Enter gy concludes the effects of the proposed EPU on the available time for operator actions have been appropriately accounted for and appropriate

actions have been taken to ensure that operator performance is not adversely affected by the proposed EPU. Entergy further concludes that the requirements of GDC-19, 10 CFR 50.120, and 10 CFR 55 will continue to be met following implementation of the proposed EPU.

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2-464 2.12 Power Ascension and Testing Plan 2.12.1 Approach to EPU Power Level and Test Plan Regulatory Evaluation The purpose of the EPU test program is to demonstrate SSCs will perform satisfactorily in service at the proposed EPU power level. The test program also provides additional assurance the plant will continue to operate in accordan ce with design criteria at EPU conditions. The review evaluated: (1) plans for the initial approach to the proposed maximum licensed thermal power level, including verification of adequate plant performance; (2) transient testing necessary to demonstrate plant equipment will perform satisfactorily at the proposed increased maximum licensed thermal power level; and (3) the test program's conformance with applicable regulations. The regulatory acceptance criteria for the proposed EPU test program are based on 10 CFR Part 50, Appendix B, Criterion XI, which requires establishment of a test program to demonstrate SSCs will perform satisfactorily in service. Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the

effects of CPPUs. Section 10.4 of the CLTR a ddresses the testing required for the initial power ascension following the implementation of EPU. The results of this evaluation are described below. Testing is required for the initial power ascension following the implementation of EPU. A standard set of tests is established for the initial power ascension steps of EPU, which supplement the normal TS testing requirements. The EPU testing program at GGNS is based on the GGNS-specific initial EPU power ascension, and TSs use the same performance criteria as in the original power ascension tests, unless they have been replaced by updated criteria since the initial test program.

2.12.1.1 Testing Program GGNS meets all CLTR dispositions. The topics addressed in this section are: Topic CLTR Disposition GGNS Result Testing Program [[

     ]] Meets CLTR Disposition The CLTR states that the increase in power level changes plant and system performance.

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2-465 Based on the analyses and GEH BWR experience with uprated plants, a standard set of tests has been established for the initial power ascensi on steps of EPU. Testing will be done in accordance with the TS surveillance requirements on instrumentation that is re-calibrated for EPU conditions. These tests supplement the normal TS testing requirements. Overlap between the IRM and APRM will be assured. Steady-state data will be taken at points from 90% up to 100% of the CLTP RTP, so that system performance parameters can be projected for EPU power before the CLTP RTP is exceeded.

EPU power increases above the 100% CLTP RTP will be made along an established flow control/rod line in increments of equal to or less than 5% power. Steady-state operating data, including fuel thermal margin, will be taken and evaluated at each step. Routine measurements of reactor and system pressures, flows, and vibration will be evaluated from each measurement point, prior to the next power increment. Radiation measurements will be made at selected power levels to ensure the protection of personnel.

Radiation surveys, including specific locations and criteria, are located in Attachment 9 to the EPU LAR. Control system tests will be performed for the RFW/reactor water level controls and pressure controls. These operational tests will be made at the appropriate plant conditions for that test at each of the power increments, to show acceptable adjustments and operational capability. Testing will be done to confirm the power level near the turbine first-stage scram bypass setpoint. Details on vibration monitoring are provided in Attachment 10 to the EPU LAR. The same performance criteria will be used as in the original power ascension tests, unless they have been replaced by updated criteria since the initial test program. [[

     ]]  Vibrational testing is addressed in Attachment 10 to the EPU LAR. The EPU testing program at GGNS, which is ba sed on the specific testing required for the GGNS initial EPU power ascension, supplemented by normal TS testing, meets all CLTR dispositions. Attachment 9 to the EPU LAR contains details of the testing program.

2.12.1.2 Transient Tests and Modifications Large transient testing is normally performed on new plants because experience does not exist to confirm a plant's operation and response to events. However, these tests are not normally performed for plant modifications following initia l startup because of well-established quality assurance and maintenance programs including component and system level post-modification NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-466 testing and extensive experience with general behavior of unmodified equipment. When major modifications are made to the plant, large transient testing may be needed to confirm that the modifications were correctly implemented. However, such testing should only be imposed if it is deemed necessary to demonstrate safe operation of the plant. GGNS does not intend to perform large transient testing involving an automatic scram from a high power. Transient experience at high powers at operating BWR plants has shown a close correlation of the plant transient data to the evaluated events. The operating history of GGNS demonstrates that previous transient events from full power are within expected peak limiting values. The transient analysis performed for the GGNS EPU demonstrates that all safety criteria are met and that this uprate does not cause any previous non-limiting events to become limiting.

[[

                       ]]  No safety-related systems were modified for the EPU, however some instrument setpoints were changed. The instrument setpoi nts that were changed (see Table 2.4-1) do not contribute to the response to large transient events. No physical modification or setpoint changes were made to the SRVs. No new systems or features were installed for mitigation of rapid pressurization AOOs for this EPU.  [[

                                                                                                                                                     ]]  Should any future large transients occur, GGNS procedures re quire verification that the actual plant response is in accordance with the predicted response. Ex isting plant event data recorders are capable of acquiring the necessary data to confirm the actual versus expected response.

Further, [[

                                                  ]]  In addition, the limiting transient analyses are included as part of the RLA.

Conclusion The EPU test program, including plans for the initial approach to the proposed maximum licensed thermal power level, transient testing necessary to demonstrate plant equipment will perform satisfactorily at the proposed increased maximum licensed thermal power level, and the test program's conformance with applicable regulations, has been reviewed. Entergy concludes the proposed EPU test program provides adequate assurance the plant will operate in accordance

with design criteria and SSCs affected by the proposed EPU, or modified to support the proposed EPU, will perform satisfactorily in service. Further, Entergy finds there is reasonable assurance that the EPU testing program satisfies the requirements of 10 CFR Part 50, Appendix B, Criterion XI. Therefore, Entergy finds the proposed EPU test program acceptable. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-467 2.13 Risk Evaluation 2.13.1 Risk Evaluation of EPU Regulatory Evaluation Entergy conducted a risk evaluation to: (1) demonstrate the risks associated with the proposed EPU are acceptable; and (2) determine if "special circumstances" are created by the proposed

EPU. As described in Appendix D of SRP Chapter 19, special circumstances are present if any issue would potentially rebut the presumption of adequate protection provided by Entergy to meet the deterministic requirements and regulations. Entergy's review cove red the effect of the proposed EPU on core damage frequency (CDF) a nd large early release frequency (LERF) for

the plant due to changes in the risks associated with internal events, external events, and shutdown operations. In addition, Entergy's review covered the quality of the risk analyses used by Entergy to support the application for the proposed EPU. This included a review of Entergy's actions to address issues or weaknesses that have been raised in previous NRC staff reviews of the individual plant examinations (IPEs), individual plant examinations of external events (IPEEEs), or by an industry peer review. The NRC's risk acceptability guidelines are contained in RG 1.174 (Reference 91). Technical Evaluation A PRA has been completed for the effect of the EPU on at-power and shutdown conditions. This PRA model includes changes due to the EPU conditions. A combination of quantitative and qualitative methods is used to assess the potential risk effects. The approaches and results of the risk assessment are summarized below. The complete PRA report is included in 3 to the EPU LAR. 2.13.1.1 Probabilistic Risk Assessment Quality The Level 1 and Level 2 GGNS PRA analyses were originally developed and submitted to the NRC in December 1992 as the GGNS IPE submittal. The NRC subsequently provided a safety evaluation of the IPE in March 1996. The GGNS PRA model and documentation has been updated to reflect the current plant configuration and to reflect the accumulation of additional plant operating history and component failure data. The current GGNS PRA model at the time of this analysis is Revision 3 of the GGNS Level 1 and Level 2 PRA models. The GGNS internal events PRA received a formal industry PRA peer review in August 1997; the final report was issued in October 1997. The pur pose of the PRA peer review process is to provide a method for establishing the technical quality of a PRA for the spectrum of potential risk-informed plant licensing applications for which the PRA may be used. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-468 All of the 'A' priority PRA peer review comments have been addressed by GGNS and incorporated into the GGNS PRA model as appropriate. All of the 'B' priority comments have been addressed except for one documentation item for the internal flood analysis and those related to the Level 2 PRA. The LERF model and internal flooding model are being updated at this time and will address these items. These items do not significantly affect the results of this risk assessment or change the conclusions. The GGNS Level 1 and Level 2 (LERF) PRAs pr ovide the necessary and sufficient scope and level of detail to allow the calculation of CD F and LERF changes due to the EPU for the full power internal events challenges. 2.13.1.2 Internal Events Initiating Event Frequencies The EPU study has concluded that no new initiati ng events or an increase in frequency of existing initiating events are anticipated. The EPU does not affect the generic and plant-specific failure data used in initiating event frequency calculations. Future PRA updates will incorporate any future plant initiating events into the initiator frequency calculations. No changes in BOP or support system operating configurations are iden tified that would affect initiator frequencies. Quantitative sensitivity studies were performed to increase the Loss of PCS and Large LOCA initiator frequencies to assess the risk effect of postulated initiator frequency changes due to the changes to BOP equipment and increased RCS flows. Plant Modification Review The plant modifications planned as part of the EPU were reviewed to assess potential effects on the PRA model. Many of the modifications involve replacing equipment with similar type and higher capacity; such changes do not affect PRA m odeling, and no change in risk is due to such replacements. No plant modifications for the EPU (other than the increase in power level) result in any significant effect on the risk profile. Procedure Changes At the time of this report, no procedural change s have been identified that would significantly affect the results of this risk assessment or cha nge the conclusions. Modifications to procedures (e.g., operating procedures, EOPs) will be made to reflect the EPU operating conditions and associated updated analyses. Although symptom-based procedures generally do not require change, certain parameter thresholds and curves are dependent on power and decay heat levels. Any such changes are expected to be minor to maintain margins inherent in the EOPs/SAPs and will not significantly influence the risk profile. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-469 Setpoint Changes At the time of this report, no setpoint changes ha ve been identified that would significantly affect the results of this risk assessment or change the conclusions. Instrument recalibration and/or setpoint may be made to reflect the EPU conditions and to maintain operational flexibility and margin. None of the potential setpoint changes will result in any quantifiable effect to the PRA. An effect on the transient initiating event frequencies may be conservatively postulated due to BOP control modifications (e.g., FW/Condensate pump trip margin) but no significant numerical differences are expected. The SRV setpoints are expected to remain unchanged. As such, the probability of a stuck open relief valve (SORV) is modified in the PRA to account for the postulated increase in SRV cycles. Quantitative sensitivity studies were performed to increase the Loss of PCS initiator frequency to assess the risk effect of postulated initiator frequency changes due to the changes to BOP equipment. Plant Operating Conditions The key plant operation modifications in support of the EPU include the following: Increased reactor thermal power from 3,898 MWt (CLTP) to 4,408 MWt (EPU). Corresponding change in FW flow (increase of approximately 13%) No other plant operating configuration changes of significance to the PRA are identified. RPV operating pressure and temperature are not changed for the EPU. The increase in FW flow supports the EPU. No long-term increase in initiating event frequency is anticipated, and no change was made to the PRA as a result of the EPU. Quantitative sensitivity studies were performed to increase the Loss of PCS and Large LOCA initiator frequencies to assess the risk effect of postulated initiator frequency changes due to the changes to BOP equipment and

increased flows. Component Reliability As discussed previously, the EPU modifications include equipment replacements with similar type and/or higher capacity; such changes do not affect PRA modeling, and no change in risk is due to such replacements. Although equipment re liability can be postulated theoretically to behave as a bathtub curve (i.e., the beginning and end of life phases being associated with higher failure rates than the steady-state period), no significant effect on the long-term average reliability is expected. In addition, there are no plans to operate equipment beyond design ratings for the EPU. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-470 No changes in equipment or system response times of significance to the PRA have been identified. Success Criteria EPU design calculations and GGNS PRA Modular Accident Analysis Program (MAAP) runs were used to confirm the success criteria fo r the PRA. The following PRA success criteria effects were identified due to the EPU: Timing of accident sequence progression affecting allowable time for operator actions. 15 of 20 SRVs required for the EPU condition for RPV initial overpressure protection during an ATWS scenario. No changes in success criteria have been identified with regard to the Level 2 (LERF) containment evaluation. The slight changes in accident progression timing and decay heat load have no direct effects on Level 2 PRA safety functions, such as containment isolation and ex-vessel debris coolability; the only effect is the timing of the mode led accident scenarios. This issue is addressed in the human reliability analysis. Operator Actions Operator actions play a significant role in the PRA model. Operator action success and failure is influenced by the time available to detect, diagnose, and perform the required actions. The increase in thermal power due to the EPU results in higher decay heat levels post-trip, decreasing the allowable time for operator actions. These effects are calculated using the MAAP computer code to determine the change in time. A screening process was first performed to identif y those post-initiator operator actions that have an effect on the PRA results. The operator actions identified for explicit review were selected based on the following criteria: 1. Fussell-Vesely (FV) (with respect to CDF) importance measure 5E-3 2. Risk Achievement Worth (RAW) (with respect to CDF) importance measure 2.0 3. FV (with respect to LERF) importance measure 5E-3 4. RAW (with respect to LERF) importance measure 2.0 5. Time critical ( 30 min. available) action NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-471 These criteria have been used in past EPU risk assessments. If any of the above criteria are met for an operator action, the action is maintained for explicit consideration in the EPU risk assessment. Potential HEP changes for operator actions screened out from explicit assessment in this EPU risk assessment will not have a significant effect on the quantitative results. The non-significant

HEPs, if adjusted, would be expected to affect the risk profile by a fraction of a percent. Operator actions not based on thermal power (i.e., timing based on battery life, compressor failure due to room heatup) remain unchanged in the EPU model. Changes in the allowable timing based on thermal power level were incorporated into the Human Reliability Analysis (HRA) and the human error probabilities (HEPs) were updated using the same methodologies as used in the GGNS base PRA. Dependent HEP combinations and AC power recovery convolution probabilities were also updated. LERF The Level 2 LERF PRA framework, containment event trees, functional fault trees, and Level 2 basic event failure probabilities remain unchanged post-EPU. There are small changes in the accident progression timing, but these small changes do not affect the LERF model. Changes to the LERF frequency are due to changes in the Level 1 model that are passed into the LERF model. Timing changes to RPV breach post core dama ge occur due to the EPU power increase; however, Level 2 PRA HEPs are not affected due to the very high failure probabilities already modeled for such actions (e.g., recover injection prior to vessel breach).

2.13.1.3 External Events Although the frequency of external events is not affected by the EPU, external event hazards were reviewed as part of this risk assessment.

Internal Fires The EPU effect on the GGNS internal fires ri sk profile was assessed using GGNS fire PRA models. Similar changes as the internal event model were made to the fire PRA model. The results of the changes to the GGNS fire PRA due to the reduced timings show a small increase (3%) in the fire CDF. The current GGNS fire PRA model is based on Revision 2 of the Level 1 PRA and the corresponding LERF model. This is consistent with past EPU risk assessments which have reported a smaller percentage change in the spatial hazards models (i.e., seismic, fire, floods) than for the full power internal events PRA. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-472 Seismic GGNS currently does not maintain a seismic PRA. The seismic analysis of the GGNS IPEEEs performed a seismic margins analysis (SMA). The IPEEE SMA analysis does not produce accident sequence frequencies. The EPU does not affect seismic initiating event frequencies. The EPU has little or no effect on the seismic qualifications of SSC. The conc lusions of the GGNS IPEEE SMA are not changed by the EPU. Past industry seismic PRA studies have shown that seismic risk is much less affected by changes in accident sequence timings and HEP changes than are internal events PRAs. Other External Hazards The EPU has no effect on the frequency of othe r external hazards (e.g., high winds, tornadoes, external floods, transportation, or nearby facility events). The EPU also does not involve plant modifications that affect plant protections against such hazards.

The GGNS IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and other external hazards was accomplished by reviewing the plant environs against regulatory requirements regard ing these hazards. Based upon this review, it was concluded that GGNS meets the applicable NRC SRP requirements, and therefore has an acceptably low risk with respect to these hazards. The EPU would not change these conclusions. 2.13.1.4 Shutdown Operations No new initiating events or increased potential for initiating events during shutdown (e.g., loss of RHR train) are identified due to the EPU. Shutdown risk is strongly dependent upon the time available from the start of an event to the onset of core damage. As time elapses after shutdown, accidents leading to boiling of coolant within the RPV and consequential inventory losses take more time to evolve. The burden on plant systems decreases as well. The effect of decreasing decay heat on the times to boil and core damage is accounted for in two ways. The first is the calculation of decay heat pr esent at a particular point in the outage. The second takes into consideration the heat capacity of the water and structures in the system available to absorb decay heat before boiling and core damage occur. Both of these aspects are addressed in Attachment 13 to the EPU LAR to support the assessment of the relationship of decay heat levels and times available in which to perform human actions to prevent core damage during shutdown accident scenarios. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

2-473 The EPU is assessed to have a non-significant (delta CDF of approximately 2%) effect on the GGNS shutdown risk profile. 2.13.1.5 Results The EPU PRA Assessment Report is presented in Attachment 13 to the EPU LAR. The key results from the GGNS EPU PRA are summarized below: Detailed thermal hydraulic analyses of the plant response using the EPU configuration indicate reductions in the operator action "allowable" times for some actions. The reduced operator action "allowable" times resulted in increases in the assessed HEPs for some actions in the PRA model, specifica lly in RPV water level control errors during failure to scram sequences. Only small risk increases were identified for the changes associated with the EPU, those associated with: (1) reduced times available for effective operator actions; and (2) minor changes in some functional success criteria in the PRA (negligible effect on results). The risk effect due to the implementation of the EPU is low and acceptable without the requirement for special compensatory measures. The EPU is estimated to increase the GGNS internal events PRA CDF from the base value of 2.68E-6/yr to 2.91E-6/yr, an increase of 2.3E-7/yr (8.6%). LERF increases from the base value of 1.44E-7/yr to 1.48E-07/yr, an increase of 4.3E-9 /yr (3%). These delta CDF and LERF results are in the "very low" category (i.e., Region III) of the Regulatory Guide 1.174 guidelines for

CDF and LERF. Conclusion An assessment of the risk implications associated with the implementa tion of the proposed EPU has been performed. Entergy concludes it has adequately modeled and/or addressed the potential effects associated with the implementation of th e proposed EPU. Entergy further concludes the results of the risk analysis indicate that th e risks associated with the proposed EPU are acceptable and do not create the "special circumstances" described in Appendix D of SRP Chapter 19. Therefore, Entergy finds the risk implications of the proposed EPU acceptable.

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3-1 3 REFERENCES

1. GE Nuclear Energy, "Constant Pressu re Power Uprate," NEDC-33004P-A, Revision 4, (Proprietary), July 2003; and NEDO-33004-A, Revision 4, (Non-Proprietary), July 2003.

2. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32424P-A, (Proprietary), February 1999; and NEDO-32424, (Non-Proprietary), April 1995.

3. RS-001, US NRC, Office of Nuclear React or Regulation, "Review Standard for Extended Power Uprates," December 2003.

4. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32523P-A , (Proprietary), February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999.

5. GE Nuclear Energy, "General Electric Sta ndard Application for Reactor Fuel,"

NEDE-24011-P-A and NEDE-24011-P-A-US, (Proprietary); and NEDO-24011-P-A and NEDO-24011-P-A-US, (Non-Proprietary ), (latest approved revision).

6. RG 1.49, "Power Levels of Nuclear Power Plants," Revision 1, December 1973. (Withdrawn July 2007)

7. GE Nuclear Energy, "Applicability of GE Methods to Expanded Operating Domains,"

NEDC-33173P, (Proprietary); and NE DO-33713, (Non-Proprietary), Revision 0, February 2006. Thomas B. Blount (NRC) to Jerald G. Head (GEH), "Final Safety Evaluation for GE Hitachi Nuclear Energy Americas, LLC Licensing Topical Report NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" (TAC No. MD0277)," July 21, 2009.

8. Ashok Thadani (NRC) to Gary L. Sozzi (GE), "Use of the SHEX Computer Program

and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.

9. GE Nuclear Energy, "General Electric Met hodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations (TAC N

o. MC3788)," NEDC-32983P-A, Revision 2, (Proprietary), January 2006; and NEDO-32983P-A, Revision 2, (Non-Proprietary), January 2006.

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3-2 10. RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

11. NRC GL 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.

12. BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," EPRI, TR-105697, September 1995.

13. 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 Insp ection 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.

14. BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines,"

EPRI, TR-1009946, November 2004.

15. BWRVIP-76, "BWR Core Shroud Inspection a nd Flaw Evaluation Guidelines," EPRI, TR-114232, November 1999.

16. BWRVIP-25, "BWR Core Plate Inspection a nd Flaw Evaluation Guidelines," EPRI, TR-107284, December 1996.

17. NRC GL 88-01, "NRC Position on IGSCC in BW R Austenitic Stainless Steel Piping,"

January 25, 1988.

18. NUREG-0313, "Technical Report on Material Se lection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," Revision 2, June 1986.

19. BWRVIP-75-A, "BWR Vessel and Internals Pr oject, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules," EPRI, TR-1012621, October 2005.

20. RG 1.54, "Service Level I, II, and III Protec tive Coatings Applied to Nuclear Power Plants," Revision 0, June 1973 and Revision 1, July 2000.

21. NRC GL 98-04, "Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Defi ciencies and Foreign Material in Containment," July 14, 1998.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

3-3 22. NUREG-1801, "Generic Aging Lessons Lear ned (GALL) Report - Tabulation of Results," Volume 2, Revision 1, September 2005.

23. NUREG-1344, "Erosion/Corrosion-Induced Pipe Wall Thinning in US Nuclear Power Plants." 24. NRC GL 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning," May 2, 1989.

25. BWRVIP-135, Revision 1, "BWR Vessel and Inte rnals Project Integrated Surveillance Program (ISP) Data Source Book and Pl ant Evaluations," EPRI, TR-1013400, June 2007. 26. RG 1.99, "Radiation Embrittlement of R eactor Vessel Materials," Revision 2, May 1988.

27. EPRI, "BWR Water Chemistry Guidelines, 2004 Revision," BWRVIP-130, October 2004. 28. ASME B&PV Code, 1998 Edition, Section III, Division 1.

29. GE Hitachi Nuclear Energy, "Grand Gulf Nucl ear Station Reactor Internals Vibration Measurements," NEDE-31148P, Revision 0, (Proprietary), February 1986.

30. RG 1.20, "Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing," Revision 3, March 2007.

31. RG 1.163, "Performance-Based Containment Leak-Test Program," September 1995.

32. Nuclear Energy Institute (NEI) 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," July 26, 1995.

33. ANSI 56.8-1994, "Containment System Leakage Testing Requirements." (Withdrawn, 2004) 34. NRC GL 89-10, "Safety-Related Motor Oper ated Valve Testing and Surveillance,"

June 28, 1989.

35. NRC GL 96-05, "Verification of Design-Basi s Capability of Safety-Related Motor Operated Valves," September 18, 1996.

36. NRC GL 95-07, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," August 17, 1995.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

3-4 37. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," November 1980.

38. W. Glenn Warren (BWROG) to NRC, "'Alternate BWR Feedwater Nozzle Inspection Requirements,' GE-NE-523-A71-0594-A, Revision 1, May 2000," BWROG-00068, June 6, 2000.

39. RG 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants," Revision 1, June 1984.

40. "Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," NUMARC 87-00, Revision 1, August 1991.

41. RG 1.155, "Station Blackout," August 1988.

42. GE Nuclear Energy, "General Electric Instrument Setpoint Methodology," NEDC-31336P-A, (Proprietary), September 1996; and NEDO-31336-A, (Non-Proprietary), September 1996.

43. GE Nuclear Energy Safety Communication, SC 04-14, "Narrow Range Water Level Instrument Level 3 Trip Final Report," October 11, 2004.

44. M. A. Krupa (Entergy) to NRC, "License Amendment Request, Power Range Neutron Monitoring System Upgrade, Grand Gulf Nuclear Station, Unit 1, Docket No. 50-416, License No. NPF-29," November 3, 2009.

45. RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis

Accidents at Nuclear Power Reactors," July 2000.

46. RG 1.52, "Design, Inspection, and Testing Cr iteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," Revision 3, June 2001.

47. RG 1.3, "Assumptions Used for Evaluating th e Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors," Revision 2, June 1974.

48. ANSI/ANS-5.1-1994, "Decay Heat Power in Light Water Reactors."

49. GE Nuclear Energy, "The General Electric Pressure Suppression Containment Analytical Model," NEDM-10320, (Pr oprietary), March 1971; and NEDO-10320, (Non-Proprietary), April 1971.

50. GE Nuclear Energy, "The General El ectric Mark III Pressure Suppression Containment System Analytical Model," NEDO-20533, (Non-Proprietary), June 1974.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

3-5 51. GE Nuclear Energy, "General Electric Model for LOCA Analysis In Accordance With 10 CFR 50 Appendix K," NEDE-20566-P-A, (Proprietary), September 1986; and NEDO-20566A, (Non-Proprietary), September 1986.

52. RG 1.7, "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident," Revision 2, November 1978.

53. GE Nuclear Energy Safety Communication, SC 06-01, "Worst Single Failure for Suppression Pool Temperature Analysis," January 19, 2006.

54. GE Nuclear Energy Service Information Letter No. 636, "Additional Terms included in Reactor Decay Heat Calculations," Revision 1, June 2001.

55. NUREG-0783, "Suppression Pool Temperature Limits for BWR Containment,"

July 1981.

56. GE Nuclear Energy, "Elimination of Limit on Local Suppression Pool Temperatures for SRV Discharge with Quenchers," NEDO-30832-A, (Non-Proprietary), May 1995.

57. GE Nuclear Energy, "Annulus Pressuri zation Load Adequacy Evaluation,"

NEDO-24548, (Non-Proprietary), January 1979.

58. GE Nuclear Energy Safety Communication SC 09-01, "Annulus Pressurization Loads Evaluation," June 8, 2009.

59. GE Hitachi Nuclear Energy, "TRACG ESBWR Safety Analysis - Additional Information," NEDE-33440P, Revision 2, (Proprietary), March 2010; and NEDO-33440, Revision 2, (Non-Proprietary), March 2010.

60. GE Nuclear Energy, "TRACG Application for ESBWR," NEDE-33083P-A, (Proprietary), October 2005; and NEDO-33083-A, (Non-Proprietary), October 2005.

61. GE Nuclear Energy, "PDA - Pipe Dynamic Analysis Program for Pipe Rupture Movement," NEDE-10813A, (Proprietary), February 1976.

62. ANSYS Mechanical Version 11.0, "List of Computer Platforms with ANSYS Installed - Modified August 05 2008," August 5, 2008.

63. N. Kalyanam (NRC) to George A. Williams (GGNS), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment RE: Elimination of Requirements for Hydrogen Recombiners and Hydrogen Monitors (TAC No. MC2177)," June 16, 2004.

64. RG 1.82, "Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident," Revision 3, November 2003.

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3-6 65. Boiling Water Reactor Owners' Group, "Utility Resolution Guide for ECCS Suction Strainer Blockage," NEDO-32686-A, (Non-Proprietary), October 1998.

66. S. Patrick Sekerak (NRC) to William A. Ea ton (GGNS), "Grand Gulf Nuclear Station (GGNS), Unit 1 - Issuance of License Amendment RE: Full-Scope Implementation of

an Alternative Accident Source Term (TAC NO. MA8065)," March 14, 2001 (ML010780172).

67. Global Nuclear Fuel, "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II)," NEDC-33270P, Revision 3, March 2010.

68. James F. Harrison (GEH) to NRC, "Implementation of Methods Limitations - NEDC-33173P (TAC No. MD0277)," September 18, 2008.

69. Louis M. Quintana (GE) to NRC, "Responses to RAIs 1, 13, 14-1, 18 and 22 - Methods Interim Process," May 3, 2005.

70. GE Hitachi Nuclear Energy, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Pl us," NEDC-33006P-A, Revision 3, June 2009.

71. George Stramback (GE) to NRC, "Compl etion of Responses to MELLLA Plus AOO

RAIs," March 4, 2004.

72. GE Nuclear Energy, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications

," NEDO-32465-A, (Non-Proprietary), August 1996.

73. GE Nuclear Energy, "Backup Stability Prot ection (BSP) for Inoperable Option III

Solution," OG 02-0119-260, July 17, 2002.

74. GE Nuclear Energy, "Plant-Specific Regi onal Mode DIVOM Procedure Guideline,"

GE-NE-0000-0028-9714-R1, June 2005.

75. BWROG-94078, "BWR Owners' Group Guidelines for Stability Interim Corrective

Action," June 6, 1994.

76. GE Nuclear Energy, "Licensing Topical Report, ODYSY Application for Stability Licensing Calculations Including Option I-D and II Long Term Solutions,"

NEDE-33213P-A, (Proprietary), April 2009; and NEDO-33213-A, (Non-Proprietary), April 2009.

77. GE Nuclear Energy, "ATWS Rule Issues Relative to BWR Core Thermal-Hydraulic Stability," NEDO-32047-A, (Non-Proprietary), June 1995.

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

3-7 78. GE Nuclear Energy, "Mitigation of BWR Core Thermal-Hydraulic Instabilities in ATWS," NEDO-32164, (Non-Proprietary), December 1992.

79. GE Nuclear Energy, "Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 - Volume 4)," NEDC-24154P-A, Revision 1, Supplement 1, (Proprietary), February 2000.

80. GE Nuclear Energy, "Continuous Control Rod Withdrawal Transient in the Startup Range," NEDO-23842, (Non-Proprietary), April 1978.

81. NUREG-0800, Standard Review Plan, Section 15.4.4-15.4.5, "Startup of an Inactive Loop or Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate," Revision 2, March 2007.

82. GE Nuclear Energy, "Banked Position Withdrawal Sequence," NEDO-21231, (Non-Proprietary), January 1977.

83. General Electric Company, "GESTR-LOCA a nd SAFER Models for the Evaluation of the Loss-of-Coolant Accident Volume III, Supplement 1, Additional Information for Upper Bound PCT Calculation," NEDE-23785P-A, Volume III, Supplement 1, Revision 1, (Proprietary), March 2002; and NEDO-23785-A, Volume III, Supplement 1, Revision 1, (Non-Proprietary), April 2002.

84. GE Hitachi Nuclear Energy, "Grand Gu lf Nuclear Station GNF2 ECCS-LOCA Evaluation," 0000-0100-8822-R1, December 2009.

85. GE Hitachi Nuclear Energy, "Grand Gulf Nuclear Station GE14 ECCS-LOCA

Evaluation," 0000-0075-4473-R0, July 2008.

86. BWROG, "Emergency Procedure and Severe Accident Guidelines (EPGs/SAGs),"

Revision 2, March 2001.

87. NUREG-0016, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Wate r Reactors," Revision 1, January 1979.

88. NUREG-0713, Volume 30, "Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2008," January 2010.

89. NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980.

90. ANSI/ANS-3.5-1998, "Nuclear Power Plant Simu lators for Use in Operator Training and Examination." (Withdrawn 2008)

NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

3-8 91. RG 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-1 Appendix - A Limitations from Safety Evaluation for LTR NEDC-33173P The following table lists the limitations from the Safety Ev aluation for LTR NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains," (Reference A-1) with their location for implementation in the PUSAR. NEDC-33173P is applicable to fuel designs through GE14. The GGNS EPU core design includes the GNF2 fuel product line. All of the GE14 restrictions have been applied to the GNF2 fuel. Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.1 TGBLA/PANAC Version The neutronic methods used to simulate the reactor core response and that feed into the downstream safety analyses supporting operation at EPU/MELLLA+ will apply

TGBLA06/PANAC11 or later NRC-approved version of neutronic method. Comply Note 4 in Table 1-1 9.2 3D Monicore For EPU/M ELLLA+ applications, relying on TGBLA04/PANAC10 methods, the bundle RMS difference uncertainty will be established from plant-specific core-

tracking data, based on TGBL A04/PANAC10. The use of plant-specific trendline based on the neutronic method employed will capture the act ual bundle power uncertainty of the core monitoring system. N/A Note 1 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-2 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.3 Power to Flow Ratio Plant-specific EPU and expanded operating domain applications will confirm that the core thermal power to

core flow ratio will not exceed 50 MWt/Mlbm/hr at any statepoint in the allowed operating domain. For plants that

exceed the power-to-flow value of 50 MWt/Mlbm/hr, the application will provide power distribution assessment to establish that neutronic methods axial and nodal power distribution uncertainties have not increased. Comply Section 2.8.2.4.2 Consistent with Reference A-5 9.4 SLMCPR1 For EPU operation, a 0.02 value shall be added to the cycle-specific SLMCPR value. This adder is applicable to SLO, which is derived from the dual loop SLMCPR value. Comply Section 2.8.2.2.1 9.5 SLMCPR2 For operation at M ELLLA+, including operation at the EPU power levels at the achievable core flow statepoint, a 0.03 value shall be added to the cycle-specific SLMCPR value. N/A Note 2 9.6 R-Factor The plant specific R-f actor calculation at a bundle level will be consistent with lattice axial void conditions expected for the hot channel operating st ate. The plant-specific EPU/MELLLA+ application will confirm that the R-factor calculation is consistent with the hot channel axial void conditions. Comply Section 2.8.2.4.3 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-3 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.7 ECCS-LOCA 1 For applications requesting implementation of EPU or expanded operating domains , including MELLLA+, the small and large break ECCS-LOCA analyses will include top-peaked and mid-peaked pow er shape in establishing the MAPLHGR and determining the PCT. This limitation is applicable to both the licensing bases PCT and the upper

bound PCT. The plant-specific applications will report the limiting small and large break licensing basis and upper bound PCTs. Comply Section 2.8.5.6.2 and Table 2.8-7 9.8 ECCS-LOCA 2 The ECCS-LOCA will be performed for all statepoints in the upper boundary of the expanded operating domain, including the minimum core flow statepoints, the transition statepoint as defined in Reference A-2 and the 55 percent core flow statepoint. The plant-specific application will report the limiting ECCS-LOCA results as well as the rated power and flow results. The SRLR will include both the limiting statepoint ECCS-LOCA results and the rated conditions ECCS-LOCA results. N/A Note 2 9.9 Transient LHGR 1 Plant-specifi c EPU and MELLLA+ applications will demonstrate and document that during normal operation

and core-wide AOOs, the T-M acceptance criteria as specified in Amendment 22 to GESTAR II will be met. Specifically, during an AOO, th e licensing application will demonstrate that the: (1) loss of fuel rod mechanical integrity will not occur due to fuel melting and (2) loss of fuel rod mechanical integrity will not occur due to pellet-cladding mechanical interac tion. The plant-specific application will demonstrat e that the T-M acceptance criteria are met for the both the UO 2 and the limiting GdO 2 rods. Comply Section 2.8.5.2.1 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-4 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.10 Transient LHGR 2 Each EPU and MELLLA+ fuel reload will document the calculation results of the analyses demonstrating compliance to transi ent T-M acceptance crite ria. The plant T-M response will be provided with the SRLR or COLR, or it will be reported directly to the NRC as an attachment to

the SRLR or COLR. Comply Section 2.8.5.2.1 9.11 Transient LHGR 3 To account for the impact of the void history bias, plant-specific EPU and MELLLA+ app lications using either TRACG or ODYN will demonstrate an equivalent to 10 percent margin to the fuel centerline melt and the 1 percent cladding circumferential plas tic strain acceptance criteria due to pellet-cladding mechanical interaction for all of limiting AOO transient events, including equipment out-of-service. Limiting transients in this case, refers to transients

where the void reactivity coeffi cient plays a significant role (such as pressurization events). If the void history bias is incorporated into the transient model within the code, then the additional 10 percent margin to the fuel centerline melt and the 1 percent cladding circumferential plastic strain is no longer required. Comply Section 2.8.5.2.1 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-5 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.12 LHGR and Exposure Qualification In MFN 06-481, GE committed to submit plenum fission gas and fuel exposure gamma scans as part of the revision to the T-M licensing process. The conclusions of the plenum fission gas and fuel exposure gamma scans of GE 10x10 fuel designs as operated will be submitted for NRC staff review and approval. This revision will be accomplished through Amendment to GESTAR II or in a T-M licensing LTR. PRIME (a newly developed T-M code) has been submitted to the NRC staff for review (Reference A-3). Once the PRIME LTR and its application are approved, future license applications for EPU and

MELLLA+ referencing LTR NEDC-33173P must utilize the PRIME T-M methods. N/A Note 3 9.13 Application of 10 Weight Percent Gd Before applying 10 weight percent Gd to licensing applications, including EPU and expanded operating domain, the NRC staff needs to review and approve the T-M LTR demonstrating that the T-M acceptance criteria specified in GESTAR II and Amendment 22 to GESTAR II can be met for steady-state and transient conditions. Specifically, the T-M application must demonstrate that the T-M acceptance criteria can be met for TOP and MOP conditions that bounds the response of plants operating at EPU and expanded operating domains at the most limiting statepoints, considering the operating flexibilities (e.g., equipment out-of-service). Before the use of 10 weight percent Gd for modern fuel designs, NRC must review and approve TGBLA06 qualification submittal. Where a fuel design refers to a

design with Gd-bearing rods ad jacent to vanished or water rods, the submittal should include specific information regarding acceptance criteria for the qualification and N/A Section 2.8.2.4.5. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-6 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation address any downstream impacts in terms of the safety analysis. The 10 weight percent Gd qualifications submittal can supplement this report. 9.14 Part 21 Evaluation of GESTR-M Fuel Temperature Calculation Any conclusions drawn from the NRC staff evaluation of the GE's Part 21 report will be applicable to the GESTR-M T-M assessment of this SE for future license application. GE submitted the T-M Part 21 evaluation, which is currently under NRC staff review. Upon completion of its review, NRC staff will inform GE of its conclusions. Comply Note 4 9.15 Void Reactivity 1 The void reactivity coefficient bias and uncertainties in TRACG for EPU and MELLLA+ must be representative of the lattice designs of the fu el loaded in the core N/A Note 5 9.16 Void Reactivity 2 A supplement to TRACG /PANAC11 for AOO is under NRC staff review (Reference A-4). TRACG internally models the response surface for the void coefficient biases and uncertainties for known dependencies due to the relative moderator density and exposure on nodal basis. Therefore, the void history bias determined through the methods review can be incorporated into the response

surface "known" bias or through changes in lattice physics/core simulator methods for establishing the instantaneous cross-sections. Including the bias in the

calculations negates the need for ensuring that plant-

specific applications show sufficient margin. For application of TRACG to EPU and MELLLA+ applications, the TRACG methodology must incorporate the void history bias. The manner in which this void

history bias is accounted for will be established by the NRC staff SE approving NEDE-32906P, Supplement 3, "Migration to TRACG04/PANAC11 from N/A Note 6 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-7 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation TRACG02/PANAC10," May 2006 (Reference A-4). This limitation applies until the new TRACG/PANAC methodology is approved by the NRC staff. 9.17 Steady-State 5 Percent Bypass Voiding The instrumentation specification design bases limit the presence of bypass voiding to 5 percent (LRPM levels). Limiting the bypass voiding to less than 5 percent for long-term steady operation ensures that instrumentation is

operated within the specification. For EPU and MELLLA+ operation, the bypass voiding will be evaluated on a cycle-specific basis to confirm that the void fraction remains below 5 percent at all LPRM levels when operating at

steady-state conditions within the MELLLA+ upper boundary. The highest calculated bypass voiding at any LPRM level will be provided with the plant-specific SRLR. Comply Section 2.8.2.4.1. 9.18 Stability Setpoints Adjustment The NRC staff concludes that the presence bypass voiding at the low-flow conditions where instabilities are likely can result in calibration errors of less than 5 percent for OPRM

cells and less than 2 percent for APRM signals. These calibration errors must be accounted for while determining the setpoints for any detect and suppress long term methodology. The calibration values for the different long-term solutions are specified in the associated sections of

this SE, discussing the stability methodology. N/A Section 2.8.3.1.2 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-8 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.19 Void-Quality Correlation 1 For applications involving PANCEA/ODYN/ISCOR/TASC

for operation at EPU and M ELLLA+, an additional 0.01 will be added to the OLMCPR, until such time that GE expands the experimental database supporting the Findlay-

Dix void-quality correlation to demonstrate the accuracy and performance of the void-quality correlation based on experimental data representative of the current fuel designs and operating conditions during steady-state, transient, and

accident conditions. Comply Section 2.8.2.2.2 and Section 2.8.3.1.2 9.20 Void-Quality Correlation 2 The NRC staff is currently reviewing Supplement 3 to NEDE-32906P, "Migration to TRACG04/PANAC11 from TRACG02/PANAC10," dated May 2006 (Reference A-4).

The adequacy of the TRACG interfacial shear model qualification for application to EPU and MELLLA+ will be addressed under this review. Any conclusions specified in the NRC staff SE approving Supplement 3 to LTR NEDC-

32906P (Reference A-4) will be applicable as approved. N/A Note 6 9.21 Mixed Core Method 1 Plants implementing EPU or MELLLA+ with mixed fuel

vendor cores will provide plan t-specific justification for extension of GE's analytical methods or codes. The content of the plant-specific application will cover the topics addressed in this SE as well as subjects relevant to application of GE's methods to legacy fuel. Alternatively, GE may supplement or revise LTR NEDC-33173P (Reference A-1) for mixed core application. N/A Section 2.8.2.4.6 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-9 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.22 Mixed Core Method 2 For any plant-specific applications of TGBLA06 with fuel type characteristics not covered in this review, GE needs to provide assessment data similar to that provided for the GE fuels. The Interim Methods review is applicable to all GE lattices up to GE14. Fuel lattice designs, other than GE lattices up to GE14, with the following characteristics are not covered by this review: square internal water channels water crosses Gd rods simultaneously adjacent to water and vanished rods 11x11 lattices MOX fuel The acceptability of the modified epithermal slowing down models in TGBLA06 has not been demonstrated for application to these or other geometries for expanded operating domains. Significant changes in the Gd rod optical thickness will require an evaluation of the TGBLA06 radial flux and Gd depletion modeling before being applied. Increases in the

lattice Gd loading that resu lt in nodal reactivity biases beyond those previously esta blished will require review before the GE methods may be applied. N/A Section 2.8.2.4.7 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-10 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.23 MELLLA+ Eigenvalue

Tracking In the first plant-specific implementation of MELLLA+, the cycle-specific eigenvalue tracking data will be evaluated and submitted to NRC to establish the performance of nuclear methods under the operation in the new operating domain. The following data will be analyzed: Hot critical eigenvalue, Cold critical eigenvalue, Nodal power distribution (measured and calculated TIP comparison), Bundle power distribution (measured and calculated TIP comparison), Thermal margin, Core flow and pressure drop uncertainties, and The MIP Criterion (e.g., determine if core and fuel

design selected is expected to produce a plant response outside the prior experience base). Provision of evaluation of the core-tracking data will provide the NRC staff with bases to establish if operation at the expanded operating domain indicates: (1) changes in the performance of nuclear methods outside the EPU

experience base; (2) changes in the available thermal margins; (3) need for changes in the uncertainties and NRC-approved criterion used in the SLMCPR methodology; or (4) any anomaly that may require

corrective actions. N/A Note 7 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-11 Limitation Number from NRC SER Limitation Title Limitation Description Disposition Section of GGNS PUSAR Which Addresses the Limitation 9.24 Plant Specific Application The plant-specific applications will provide prediction of key parameters for cycle exposures for operation at EPU (and MELLLA+ for MELLLA+ applications). The plant-specific prediction of these key parameters will be

plotted against the EPU Reference Plant experience base

and MELLLA+ operating experien ce, if available. For evaluation of the margins available in the fuel design limits, plant-specific applications will also provide quarter core map (assuming core symmetry) showing bundle power, bundle operating LHGR, and MCPR for BOC, MOC, and EOC. Because the minimum margins to specific limits may occur at exposures other than the traditional BOC, MOC, and EOC, the data will be provided at these exposures. Comply Section 2.8.2.4.4 NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-12 Notes: 1. No reliance on TGBLA04/PANAC10 for GGNS.

2. Not applicable to EPU. 3. PRIME was approved in January 2010. The nuclear analysis and evaluations supporting the PUSAR were based on a PRIME LHGR limit. Nuclear information used in downstream transient and stability analyses is based on this core design. However, as described in NEDC-33173P, Supplement 4, the GSTRM thermal-mechanical properties will be used in the downstream codes until the changes have been implemented and the NRC has performed an audit of that process and published their SE. 4. As a consequence of the NRC staff review of the GE Part 21 report, a modified GNF2 T-M basis using GESTR-M was established by the GNF2 Compliance Report (Reference A-6). 5. The GGNS EPU license application is not based on TRACG. Therefore, this limitation is not applicable. 6. The Limitation is not applicable to the current GGNS EPU li cense application because the a pplication is not based on Supplement 3 to NEDE-32906P. 7. The Limitation is applicable to MELLLA+ applications only. Therefore, this limita tion is not applicable to the EPU application.

References:

A-1 GE Nuclear Energy, "Applicability of GE Methods to Expanded Operating Domains," NEDC-33173P, (Proprietary); and NEDO-33713, (Non-Proprietary), Revision 0, February 2006. Thomas B. Blount (NRC) to Jerald G. Head (GEH), "Final Safety Evaluation for GE Hitachi Nuclear Energy Americas, LLC Licensing Topical Report NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" (TAC No. MD0277)," July 21, 2009. A-2 GE Hitachi Nuclear Energy, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus," NEDC-33006P-A, Revision 3, June 2009. NEDO-33477 - REVISION 0 NON-PROPRIETARY INFORMATION

A-13 A-3 Global Nuclear Fuel, "The PRIME Model for Analysis of Fuel Rod Thermal-Mechanical Performance," NEDC-33256P, NEDC-33257P, and NEDC-33258P, January 2007. Thomas B. Blount (NRC) to Andrew A. Lingenfelter (GNF), "Final Safety Evaluation For Global Nuclear Fuel - Americas Topical Reports NEDC-33256P, NEDC-33257P and NEDC-33258P, 'The PRIME Model For Analysis of Fuel Rod Thermal-Mechanical Performance' (TAC NO. MD4114)," January 22, 2010. A-4 GE Hitachi Nuclear Energy, "Migration to TRACG04/PANAC11 from TRACG02/PANAC10 for TRACG AOO and ATWS Overpressure Transients," NEDE-32906P, Supplement 3-A, Revision 1, April 2010. A-5 James F. Harrison (GEH) to Document Control Desk (NRC), "Implementation of Methods Limitations - NEDC-33173P (TAC No. MD0277)," September 18, 2008. A-6 Global Nuclear Fuel, "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II)," NEDC-33270P, Revision 3, March 2010.}}

v t e Letter Sequence Request
TAC:ME4679, Clarify Application of Setpoint Methodology for LSSS Functions (Open)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting Questions 26 January 2011 26 January 2011 27 January 2011 31 January 2011 8 February 2011 23 February 2011 23 February 2011 1 March 2011 16 March 2011 23 March 2011 23 March 2011 4 April 2011 5 April 2011 5 April 2011 12 April 2011 10 May 2011 20 May 2011 1 June 2011 6 June 2011 22 June 2011 29 June 2011 17 August 2011 6 October 2011 14 December 2011 14 February 2012 21 March 2012 18 April 2012 18 April 2012 Answers 31 March 2011 14 April 2011 5 May 2011 15 June 2011 6 July 2011 12 September 2011 10 October 2011 6 February 2012 15 February 2012 20 February 2012 13 March 2012 21 March 2012 5 April 2012 18 April 2012 Results Approval... Other: GNRO-2012/00150, Final Report for Replacement Steam Dryer, GNRO-2017/00039, Steam Dryer Visual Inspection Results for the First Two Scheduled Refueling Outages, ML102660403, ML111720830, ML112560396, ML12167A257, ML12167A269, ML12177A287, ML13108A217, NRC-2012-0105, Environmental Assessment and Finding of No Significant Impact Related Amendment Request for Extended Power Uprate to Increase the Maximum Reactor Core Power Operating Limit from 3898 to 4408 Mwt
MONTHYEAR2011-03-30 [Table View]

ML102660399

Text

SUMMARY

1.2.2 Computer

1.3.2 Reactor

2.1.3 Reactor

2.1.4 Reactor

2.1.5 Protective

2.1.7 Reactor

2.2.5 Seismic

2.3.5 Station

2.5.1 Internal

2.5.2 Fission

2.5.3 Component

2.5.5 Waste

2.5.6 Additional

2.5.7 Additional

2.6.2 Subcompartment

2.6.4 Combustible

2.6.5 Containment

2.6.6 Secondary

2.7.2 Engineered

2.7.3 Control

2.7.4 Spent

2.7.5 Auxiliary

2.7.6 Engineered

2.8.3 Thermal

2.8.4 Emergency

2.8.5 Accident

2.9.2 Radiological

2.9.3 Additional

SUMMARY

1.1 Report

1.1.1 Generic

1.2 Purpose

1.2.2 Computer

1.2.3 Approach

1.3.1 Reactor

Subject:

Subject:

2.1 Materials

2.1.1 Reactor

2.1.3 Reactor

2.1.4 Reactor

2.1.5 Protective

2.1.7 Reactor

2.2.3 Reactor

2.2.5 Seismic

Conclusion:

2.3.1 Environmental

2.3.5 Station

2.4.1 Reactor

2.5.1 Internal

2.5.2 Fission

2.5.3 Component

2.5.5 Waste

2.5.7 Additional

2.6.1 Primary

2.6.4 Combustible

2.6.5 Containment

2.6.6 Secondary

2.7.1 Control

2.7.2 Engineered

2.7.3 Control

2.7.5 Auxiliary

2.7.6 Engineered

2.8.2 Nuclear

2.8.4 Emergency

2.6.5 Shutdown

2.8.5 Accident

2.9.1 Source

2.9.2 Radiological

2.9.3 Additional

References:

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