ML073300573

NEDO-33385, Rev. 0, Safety Analysis Report for Cooper Nuclear Station Thermal Power Optimization, Enclosure 3

ML073300573
Person / Time
Site:	Cooper
Issue date:	11/30/2007
From:	GE-Hitachi Nuclear Energy Americas
To:	Office of Nuclear Reactor Regulation
References
DRF 0000-0044-9747, NEDO-33385, Rev 0
Download: ML073300573 (103)

Text

NLS2007069 ENCLOSURE 3 GE-HITACHI NUCLEAR ENERGY SAFETY ANALYSIS REPORT FOR COOPER NUCLEAR STATION THERMAL POWER OPTIMIZATION, NEDO-33385 APPENDIX K MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE COOPER NUCLEAR STATION DOCKET NO. 50-298, DPR-46 NON-PROPRIETARY VERSION

GE.-Hitachi Nuclear Energy HITACHI 3901 Castle Hayne Rd Wilmington, NC 28402 NEDO-33385 Revision 0 Class I DRF 0000-0044-9741 November 2007 SAFETY ANALYSIS REPORT FOR COOPER NUCLEAR STATION THERMAL POWER OPTIMIZATION Copyright 2007 GE HitachiNuclear Energy

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION NON-PROPRIETARY NOTICE This is a non-proprietary version of the document NEDC-33385P 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 (( I].

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The only undertakings of the GE Hitachi Nuclear Energy (GEH) respecting information in this document are contained in Agreement 86A-MS2 between Nebraska Public Power District and GE, Task No. 4700000808, effective September 30, 2006, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Nebraska Public Power District, or for any purpose other than that for which it is intended, is not authorized; and, with respect to any unauthorized use, GEH makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION TABLE OF CONTENTS Page Acronynms and Abbreviations ................................................................................................. viii Executive Sum m ary ................................................................................................................... xiii 1.0 Introduction ...................................................................................................................... 1-1 1.1 Overview................ ......................................... 1-1 1.2 Purpose and Approach........................................... .1-2 1.2.1 17PO A nalysis Basis 1-2 1.2.2 Margins, 1-3 1.2.3 Scope of Evaltations 1-3 1.2.4 Lkceptions to the IL 77? 1-4 1.2.5 ConcurrentChanges Unrelatedto 1i-0 1-5 1.3 TPO Plant O perating Conditions ............. ................ .......... ..................................... .1-5 1.3.1 Reactor Heat Balance 1-5 1.3.2 Reactor PerftrmanceImprovemenl Features 1-5 1.4 B asis for T PO U prate ................................ ................ .......................................... 1-5 1.5 Summary and Conclusions ................................ ...... 1-6 2.0 Reactor Core and Fuel Performance ........................................................................ 2-1 2.1 Fuel Design and Operation ........... .............................................. 2-1 2.2 Therm al L im its A ssessm ent...................................................................................... 2-1 2.2.1 Safety Limit MCPR 2-1 2.2.2 MCPR Operating Limit 2-2 2.2.3 MA 1LHGR andMaximum LHGR OperatingLimits 2-2 2.3 R eactivity C haracteristics .... ............ ........... .. ............................ ............ ... 2-2 2.4 Thermal Hydraulic Stability ......................................... 2-2 2.5 Reactivity Control ......................................................... 2-3 3.0 Reactor Coolant and Connected Systems ................................................................. 3-1

3. 1 Nuclear System Pressure Relief!/Overpressure Protection .......... ......... ........ 3-1 3 .2 R eacto r Vessel ............... ............................ ........ ................................. ............. 3-1 3.2.1 tracture Tonghness 3-4 3.2.2 Reactor Vessel StructuralEvaluation 3-2 3 .3 R eactor Internals ............................................................................ ............................ 3-6 3.3.1 Reactor Internal PressureDifference 3-6 3.3.2 Reactor Internals StructuralEvaluation 3-6 3.3.3 Steam SeparatorandDryer Peiformnance 3-8 3.4 Flow-Induced Vibration ..... .................................... 3-8 3.5 PipingEvaluation ....... ..................................... 3-9 3.5.1 Reactor Cookint PressureBoundary Piping 3-9 3.5 2 Balance-of-PlantPipingEvahlation 3-13 iii

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 3.6 Reactor Recirculation System ................ 3-14 3.7 Main Steam Line Flow Restrictors .... ...... .......................... 3-14 3.8 Main Steam Isolation Valves ...................................... ............. ...... ..... 3-14 3.9 Reactor C ore Isolation C ooling ................................................................................. 3-14 3.10 R esidual Heat R em oval System .............................................. ................................. 3-14

3. 1 Reactor W ater Cleanup System .. . .. ......................

. . ....... ......... 3-15 4.0 Engineered Safety Features ............................................................................................. 4-1 4.1 Containm ent System Perform ance ............................................ ............................ .4-1 4.1. 1 Generic Letter 89-10 Program 4-1 4.1.2 Generic Letter 95-07 Program 4-2 4.1,3 GenericLetter 96-06 4-2 4.2 Emergency Core Cooling Systems ....................................... 4-2 4.2.1 High PressureCoolantInjection 4-2 4.2.2 High Pressure (ore Spray 4-2 4.2.3 Core S*pray 4-2 4.2.4 Low PressvureCoolant Injection 4-3 4.2.5 Automatic D)epressurizationSystem 4-3 4.2.6 ECCS Net Positive Suction Head 4-3 4.3 Emergency Core Cooling System Performance .............................. 4-3 4.4 M ain Control Room Atmosphere Control System ...................................................... 4-4 4.5 Standby Gas Treatment System ........... ................................... 4-4 4.6 Main Steam Isolation Valve Leakage Control System ................................... ...... 4-4 4.7 Post-LOCA Combustible Gas Control System .......................... 4-5 5.0 Instrumentation and Control .................................................................................... 5-5.1 N SSS M onitoring and Control .............. . . .............

......................... 5-1 5.1. 1 Neutron MonitoringSystem 5-1 5.1.2 Rod Worth Minimizer 5-2 5.2 BOP Monitoring and Control............. .............................. 5-2 5.2.1 1'ressure Control System 5-2 5.2.2 EHC Turbine ControlSKystem 5-3 5.2.3 l',eedwater Gontrol System 5-3 5.2.4 Leak Detection System 5-3 5.3 Technical Specification Instrum ent Setpoints .................................. ......................... 5-4 5.3.1 High-PressureScram 5-4 5.3.2 HydraulicPres'ureScram 5-4 5.3.3 High-PressureRecirculationPumlp Trip 5-5 5.3.4 Safety Relief Vah'e 5-5 5.3.5 Main Steam Line High F'low Isolation 5-5 5.3.6 ./iedAP1?M Scram 5-5 5.3.7 API1M 17low-Biased Scram 5-5 5.3.8 Rod Worth Minimizer Low Power Setpoint 5-6 5.3.9 Rod Block Monitor 5-6 iv

NEDO-33385 .REVISION 0 NON-PROPRIETARY INFORMATION 5.3.10 Flow-BiasedRod Block Monitor (1%RIP0) 5-6 5.3.11 Main,Steam Line High RIadiation Ivolation 5-6 5.3.12 Low Steam Line PressureMSIV Closure (RUN Mode) 5-6 5.3..13 Reactor Water Level Instrunmewnts 5-6 5.3.14 Main Steam Line Tunnel High Temperature Isolations 5-7 5.3.15 Low Condenser Vacuum 5-7 5.3.16 7SV Closure Scram, "1'7VFast Closure Scr'am Bypasses 5-7 6.0 Electrical Power and Auxiliary Systems ................................................................... 6-1 6 .1 A C .Po wer ............. ............................. ............. ............. .................. ....... 6-1 6.1.1 Off-Site Power 6-1 6.1.2 On-Sile Power 6-1 6.2 DC Power.,................................................. 6-2 6.3 Fuel Pool ... ................................ ............ ... 6-2 6.3.1 Fiel Pool Cooling 6-3 6.3.2 CrudA ctivity and CorrosionProduels 6-3 6.3.3 Radiation Levels 6-3 6.3.4 Fuel Racks 6-3 6.4 W ater Sy stem s .............. ........................................... .............................................. 6-3 6.4. 1 Service Water Systems 6-4 6.4.2 Main Condenset'CircnlatingWater/Normal Heat Sink Petformatce 6-4 6.4.3 Component Cooling Water System 6-5 6.4.4 Turbine Equipment Cooling System 6-5 6.4.5 Ultimate Heat Sink 6-5 6.5 Standby Liquid Control System .................................... .............. 6-5 6.6 Power Dependent Heating, Ventilation and Air Conditioning .................................. 6-6 6.7 Fire Protection .................................................... 6-6 6.7. 10 C('"R 50 Appendix R Plre Event 6-6 6.8 Systems Not Affected By TPO Uprate ................................... 6-7 7.0 Power Conversion Systems ......................................................................................... 7-1 7.1 Turbine-G enerator......................................................................................... .7-1 7.2 Condenser And Steam Jet Air Ejectors ...... _...................... .......... 7-2 7.3 Turbine Steam Bypass ........................................ .7-2 7.4 Feedwater And Condensate Systems .......... . ....................................... 7-2 7.4.,1 Normal Operation 7-3 7.4.2 TransientOperation 7-3 7.4.3 Condensate Demineralizers 7-4 8.0 Radwaste and Radiation Sources .................................................................................... 8-1 8.1 Liquid and Solid W aste M anagem ent .......................................................................... 8-1 8.2 G aseous Waste M anagem ent ....................................................................................... 8-1 8.3 Radiation Sources in the Reactor Core...................................... 8-2 8.4 Radiation Sources in Reactor Coolant ............................... 8-3

8. 4. 1 CoolantActivation Products 8-3

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 8.4.2 Activated CorrosionProducts 8-3 8.4.3 Fission Products 8-4 8.5 R adiation L evels .... ............. ............................................. 8-4 8.6 Normal Operation Off-Site Doses........................................ 8-5 9.0 Reactor Safety Performance Evaluations ...................................................................... 9-1 9.1I Anticipated Operational Occurrences ..................................... 9-1 9.1.1 Alternate Shutdown CoolingEvaluation 9-1 9.2 D esign B asis A ccidents................. .......... ................. .................. ......... ............ 9-1 9.3 Special Events ......................... .......................... 9-2 9.31 Anticipated Transient Without Scram 9-2 9.3.2 Station Blackout 9-3 10.0 O ther Evaluations ........................................................................................................ 10-1 10.1 High Energy Line Break ............................................... 10-1

10. 1.1 Steam Line Breaks 10-1 10.1.2 Liquid Line Breaks 10-I 10.2 M oderate Energy Line Break ............. .......................... .................................... 10-3 10.3 E nvironm ental Q ualifi cation ...................................................................................... 10-3 10, 3.1 ElectricalEquipment 10-3 10.3.2 Mechanical Equipment With Non-Metallic Components 10-4

'10.4 T esting .............. .. ............... ...... ........... ........... .... 10-4 10.5 O perator Training A nd Hum an Factors ..................................................................... 10-5 10.6 Plant Life ...................................... ........... 10-5 10.7 NRC and Industry Communications .................................. _10-6 10.8 Plant Procedures and Programs ................................... 10-6 10.9 Emergency Operating Procedures . ............................................. .10-6 10.10 Individual Plant Examination (IPE) ..... . .............................. 10-6 11.0 References ..................................................................................................................... 11-1 vi

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION LIST OF TABLES Table No. Title 1-1 Computer Codes Used for TPO Analyses 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions 1-3 Summary of Effect of TPO Uprate on Licensing Criteria 3-1 Adjusted Reference Temperatures - 40 Year Life (32 EFPY) 3-2 P+Q Stresses and CUFs of Limiting Components 3-3 Summary of TPO Stresses for RPV Internal Components 4-1 Cooper ECCS-LOCA Analysis Results for GE14 5-1 Analytical Limits that Change due to TPO 6-1 TPO Plant Electrical Characteristics 6-2 Fuel Pool Cooling and Cleanup System Parameters 6-3 Effluent Discharge Comparison LIST OF FIGURES No 1-Figure or. Title 1-1 Power/Flow Map for the TPO (101.7% of CLTP) 1-2 Power/Flow Map for the TPO (101.62% of CLTP) 1-3 Reactor Heat Balance - TPO Power (101.7% of CLTP), 100% Core Flow 1-4 Reactor Heat Balance - TPO Power (101.62% of CLTP), 100% Core Flow vii

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION ACRONYNMS AND ABBREVIATIONS Term Definition AC Alternating Current ADS Automatic Depressurization System AL Analytical Limit ALARA As Low As Reasonably Achievable ANSI American National Standards Institute AOO Anticipated Operational Occurrence APRM Average Power Range Monitor ART Adjusted Reference Temperature ARTS Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program ASOC Alternate Shutdown Cooling ASME American Society Of Mechanical Engineers ATWS Anticipated Transient Without Scram AV Allowable Value B&PV Boiler and Pressure Vessel BHP Brake Horsepower BOP Balance of Plant BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Project CD Condensate Demineralizer CFR Code of Federal Regulations CLTP Current Licensed Thermal Power CNS Cooper Nuclear Station CRD Control Rod Drive CS Core Spray CSC Containment Spray Cooling CSS Core Support Structure CUF Cumulative Usage Factor DBA Design Basis Accident DC Direct Current DG Diesel Generator ECCS Emergency Core Cooling System EFPY Effective Full Power Years ELTR1 NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate ELTR2 NEDC-32523P-A, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate viii

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION EOC End of Cycle EOL End of Ufe EOOS Equipment Out-Of-Service EOP Emergency Operating Procedure EPU Extended Power Uprate EQ Environmental Qualification FAC Flow Accelerated Corrosion FCV Flow Control Valve FIV Flow-Induced Vibration FPCCS Fuel Pool Cooling And Cleanup System fps Feet Per Second FW Feedwater FWHOOS Feedwater Heater(s) Out-Of-Service GDC General Design Criteria GE General Electric Company GEH GE-Hitachi Nuclear Energy GL Generic Letter HELB High Energy Line Break HEPA High Efficiency Particulate Air HP High Pressure HPCI High Pressure Coolant Injection HVAC Heating, Ventilation And Air Conditioning HWC Hydrogen Water Chemistry IASCC Irradiation Assisted Stress Corrosion Cracking ICF Increased Core Flow IEEE Institute Of Electrical And Electronics Engineers IGSCC Intergranular Stress Corrosion Cracking IPE Individual Plant Examination IRM Intermediate Range Monitor kA Kilo-amp ksi Kips Per Square Inch kV Kilovolt kW Kilowatt LCO Limiting Conditions For Operation LHGR Linear Heat Generation Rate LOCA Loss-Of-Coolant-Accident LOOP Loss Of Offsite Power LPRM Local Power Range Monitor ix

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION

~Termn Definitiona LPSP Low Power Setpoint MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCC Motor Control Circuit/Center MCPR Minimum Critical Power Ratio MELD Moderate Energy Line Break MELLLA Maximum Extended Load Line Limit Analysis MeV Million Electron Volts MIb Millions Of Pounds MLHGR Maximum Linear Heat Generation Rate MOV Motor Operated Valve MS Main Steam MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSL Main Steam Line MSLB Main Steam Line Break MSLBA Main Steam Line Break Accident MVA Million Volt Amps MWe Megawatt-Electric MWt Megawatt-Thermal NPDES National Pollutant Discharge Elimination System NPPD Nebraska Public Power District NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NTSP Nominal Trip Setpoint NUMARC Nuclear Utilities Management and Resources Council NUREG Nuclear Regulations (NRC Document)

OLMCPR Operating Limit Minimum Critical Power Ratio COS Out-Of-Service P-T Pressure-Temperature PCS Pressure Control System PCT Peak Clad Temperature PRFO Pressure Regulator Failure Open - Maximum Steam Demand PSA Probabilistic Safety Assessment psi Pounds Per Square Inch psia Pounds Per Square Inch - Absolute psid Pounds Per Square Inch - Differential psig Pounds Per Square Inch - Gauge

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Term Definition RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary REC Reactor Equipment Cooling REM Roentgen Equivalent Man (Radiation Dose Measurement)

RFP Reactor Feed Pump RG Regulatory Guide RHR Residual Heat Removal RIPD Reactor Internal Pressure Difference RIS Regulatory Issue Summary RPS Reactor Protection System RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RTNDT Reference Temperature Of Nil-Ductility Transition RTP Rated Thermal Power RWCU Reactor Water Cleanup RWM Rod Worth Minimizer SAFERtGESTR- A Computer Program, A Model For Analysis Of System Response To LOCA Loss-Of-Coolant Accident SAR Safety Analysis Report SBO Station Blackout SBPCS Steam Bypass Pressure Control System SDC Shutdown Cooling SER Safety Evaluation Report SFP Spent Fuel Pool SGTS Standby Gas Treatment System SJAE Steam Jet Air Ejector SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SPC Suppression Pool Cooling SR Surveillance Requirement SRM Source Range Monitor SRP Standard Review Plan SRV Safety Relief Valve SRVDL Safety Relief Valve Discharge Line TAP Top Of Active Fuel TBCCW Turbine Building Closed Cooling Water System TCV Turbine Control Valve TFSP Turbine First Stage Pressure xi

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION Term Definition T/G Turbine-Generator TIP Traversing In-Core Probe TLO Two (Recirculation) Loop Operation TLTP TPO Licensed Thermal Power TLTR NEDC-32938P-A, Thermal Power Optimization Licensing Topical Report TPO Thermal Power Optimization TS Technical Specification TSAR Thermal Power Optimization Safety Analysis Report TSV Turbine Stop Valve USAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink USE Upper Shelf Energy VWO Valves Wide Open Wd Recirculation Drive Flow xii

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION EXECUTIVE

SUMMARY

This report summarizes the results of all significant safety evaluations performed that justify increasing the licensed thermal power at Cooper Nuclear Station (CNS) to 2419 MWt. The requested license power level is 1.62% above the current licensed thermal power (CLTP) level of 238'1 MWt. The actual power increase is governed by the results of the core thermal power uncertainty calculation, which currently allows for an increase to 2419 MWt, 1.62% above CLTP. However, the significant safety evaluations were performed at 2421 MWt, i.e., 1.7%

above CLTP.

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

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

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

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

Evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, environmental issues, design basis accidents, and previous licensing evaluations were performed. This report demonstrates that CNS can safely operate at a power level of 2421 MWt.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION The following evaluations were conducted in accordance with the criteria of TLTR Appendix B:

I. All safety aspects of the plant that are affected by a 1.7% increase in the thermal power level were evaluated, including the Nuclear Steam Supply System (NSSS) and Balance-of-Plant (13OP) systems.

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

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

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

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

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

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

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

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

10. A review of the latest USAR and of design changes / 10 CFR 50.59 evaluations implemented, but not yet shown in the USAR, ensures adequate evaluation of the licensing basis for the effect of TPO through the date of that evaluation. Additionally, 10 CFR 50.59 evaluations for changes not yet implemented were reviewed for the effects of increased power.

The plant licensing requirements have been reviewed, and it is concluded that this TPO can be accommodated (1) without a significant increase in the probability or consequences of an accident previously evaluated, (2) without creating the possibility of a new or different kind of accident from any accident previously evaluated, and (3) without exceeding any existing regulatory limits applicable to the plant, which might cause a significant reduction in a margin of safety. Therefore, the requested TPO uprate does not involve a significant hazards consideration.

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NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION

1.0 INTRODUCTION

1.1 OVERVIEW This document addresses a Thermal Power Optimization (T.PO) power uprate of 1.62% of the current licensed thermal power (CLTP), consistent with the magnitude of the thermal power uncertainty reduction for the Cooper Nuclear Station (CNS) plant. While the significant safety evaluations performed at 2421 MWt, i.e., 1.7% above CLTP, the actual power increase is governed by the results of the core thermal power uncertainty calculation, which currently allows for an increase to 2419 MWt, 1.62% above CLTP. This report follows the Nuclear Regulatory Commission (NRC)-approved format and content for Boiling Water Reactor (BWR) Thermal Power Optimization (TPO) licensing reports documented in NEDC-32938P-A, "Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization," (TLTR) (Reference 1). Power uprates in GE BWRs of up to 120% of original licensed thermal power (OLTP) are based on the generic guidelines and approach defined in the Safety Evaluation Reports provided in NEDC-32424P-A, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," (ELTRI) (Reference 2) and NEDC-32523P-A, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2) (Reference 3). Since their NRC approval, numerous extended power uprate (EPU) submittals have been based on these reports. The outline for the TPO Safety Analysis Report (TSAR) in TLTR Appendix A follows the same pattern as that used for the EPUs. All of the issues that should be addressed in a plant-specific TPO licensing report are included in this TSAR. For issues that have been evaluated generically, this report references the appropriate evaluation and establishes that it is applicable to CNS.

BWR plants, as currently licensed, have safety systems and component capability for operation at least 1.5% above the CLTP level. The amount of power uprate (< 1.5%) contained in the TLTR was based on the expected reduction in power level uncertainty with the instrumentation technology available in 1999. The present instrumentation technology has evolved to where a power level uncertainty is reduced to as low as 0.3%, thereby supporting the evaluation of a power level increase of 1.7%. Several Pressurized Water Reactor and BWR plants have already been authorized to increase their thermal power above the OLTP based on a reduction in the uncertainty in the determination of the power through improved feedwater (FW) flow rate measurements. When a previous uprate (other than a TPO) has been accomplished, the > 1.02%

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

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 1.2 PURPOSE AND APPROACH 1.2.1 TPO Analysis Basis CNS was originally licensed at 2381 MWt. The current safety analysis basis assumes, where required, that the reactor had been operating continuously at a power level at least 1.02 times the licensed power level. The analyses performed at 102% of CLTP remain applicable at the TPO rated thermal power (RTP), because the 2% factor from Regulatory Guide (RG) 1.49, "Power Levels of Nuclear Power Plants," is effectively reduced by the improvement in the FW flow measurements. Some analyses may be performed at TPO RTP, because the uncertainty factor is accounted for in the methods, or the additional 2% margin is not required (e.g., Anticipated Transient Without Scram (ATWS)). Detailed descriptions of the basis for the TPO analyses are provided in the subsequent sections of this report.

The TPO uprate is based on the evaluation of the improved FW flow rate measurement provided in Section 1.4. Figures 1-1 (101.7% of CLTP) and 1-2 (101.62% of CLTP) illustrate the TPO power/flow-operating map for CNS. Figure 1-1 is the power/flow operating map for the bounding analysis at 101.7% of CLTP and Figure 1-2 is the power/flow operating map for the requested 101.62% of CLTP. The changes to the power/flow operating map are consistent with the generic descriptions given in TLTR Section 5.2. The approach to achieve a higher thermal power level is to increase core flow along the established Maximum Extended Load Line Limit Analysis (MELLLA) rod lines. This strategy allows CNS to maintain most of the existing available core flow operational flexibility while assuring that low power related issues (e.g.,

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

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

With respect to absolute thermal power and flow, there is no change in the extent of the SLO operating domain as a result of the TPO uprate. Therefore, the SLO operating domain is not provided. For CNS the maximum reactor core thermal power for SLO operation remains at 1657.2 MWt.

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

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

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 1.2.2 Margins The TPO analysis basis ensures that the power-dependent instrument error margin identified in RG 1.49 is maintained. NRC-approved or industry-accepted computer codes and calculation techniques are used in the safety analyses for the TPO uprate. A list of the NSSS computer codes used in the evaluations is provided in Table 1-1. Computer codes used in previous analyses (i.e., analyses at 102% of CLTP) are not listed. Similarly, factors and margins specified by the application of design code rules are maintained, as are other margin-assuring acceptance criteria used tojudge the acceptability of the plant.

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

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

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

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

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

Standby Gas Treatment, and other Engineered Safety Features are evaluated for key events. The evaluations include the containment responses during limiting abnormal events, LOCA, and safety relief valve containment dynamic loads.

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

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION changed only to maintain adequate operating margins between plant operating parameters and trip values.

6.0 Electrical Power and Auxiliary Systems Evaluations are performed to establish the operational capability of the plant electrical power and distribution systems and auxiliary systems to ensure that they are capable of supporting safe plant operation at the TPO RTP level.

7.0 Power Conversion Systems Evaluations are performed to establish the operational capability of various (non-safety) balance-of-plant (BOP) systems and components to ensure that they are capable of delivering the increased TPO power output.

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

9.0 Reactor Safety Performance Evaluations

[Ii

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

10.0 Other Evaluations High energy line break and environmental qualification evaluations are performed at bounding conditions for the TPO range to show the continued operability of plant equipment under TPO conditions, The IPE (PRA) will not be updated, because the change in plant risk from the subject power uprate is insignificant. This conclusion is supported by NRC Regulatory Issue Summary (RIS) 2002-03 (Reference 4). In response to feedback received during the public workshop held on August 23, 2001, the Staff wrote, "The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant impact on plant risk.

Therefore, no risk information is requested to support such applications."

1.2.4 Exceptions to the TLTR None.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 1.2.5 Concurrent Changes Unrelated to TPO None.

1.3 TPO PLANT OPERATING CONDITIONS 1.3.1 Reactor Heat Balance The following typical heat balance diagrams at the TPO conditions are presented:

Figure 1-3 Reactor Heat Balance - 2421 M.Wt (101.7% of CLTP), 100% Core Flow Figure 1-4 Reactor Heat Balance - 2419 MWt (101.62% of CLTP), 100% Core Flow Figure 1-3 is the Reactor Heat Balance for the bounding analysis at 101 .7% of CLTP and Figure 1-4 is the Reactor Heat Balance for the requested 101.62% of CLTP.

The small changes in thermal-hydraulic parameters for the T.PO are illustrated in Table 1-2.

These parameters are generated for TPO by performing coordinated reactor and turbine-generator heat balances that relate the reactor thermal-hydraulic parameters to the increased plant FW and steam flow conditions. Input from CNS operation is considered (e~g., steam line pressure drop) to match expected TPO uprate conditions.

1.3.2 Reactor Performance Improvement Features The following performance improvement and equipment out-of-service (OOS) features currently licensed at CNS are acceptable at the TPO RTP level:

Performanc Improvement Feature Single Loop Operation (SLO)

Increased Core Flow (ICF) (105.0% of rated)

Average Power Range Monitor, Rod Block Monitor, Technical Specifications Improvement Program (ARTS) I MELLLA (76.9% of Rated Core Flow at TLTP) 3% Safety Relief Valve (SRV) Setpoint Tolerance Turbine Bypass Valve (TBV) OOS 1.4 BASIS FOR TPO UPRATE The safety analyses in this report are based on a total thermal power measurement uncertainty of 0.3%. This will bound the actual power level requested. The detailed basis value is provided in separate documentation, which addresses the improved feedwater flow measurement accuracy using the Caldon Leading Edge Flow Meter Check-Plus system.

1-5

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 1.5

SUMMARY

AND CONCLUSIONS This evaluation has investigated a TPO uprate to 101.7% of CLTP. The strategy for achieving higher power is to increase core flow along the established MELLLA rod lines. The plant licensing challenges have been reviewed (Table 1-3) to demonstrate how the TPO uprate can be accommodated without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any existing regulatory limits or design allowable limits applicable to the plant which might cause a reduction in a margin of safety. The TPO uprate described herein involves no significant hazards consideration.

[-6

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-4 Computer Codes For TPO Analyses*

• ' *Task Computer Versionor NRC ........

Code Revision Approved ommnt Reactor Heat Balance ISCOR 09 Y (1) NEDE-2401 I-P Rev 0 SER

• The application of these codes to the CNS TPO analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The application of the codes also complies with the SERs for the extended power uprate programs.

Notes For Table 1-1:

(1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P Rev 0 by the May 12, 1978 letter from D.G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.

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NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-2 Thermal-Hydraulic Parameters at TPO Uprate Conditions TPO RTP TPO RTP Parameter CLTP (101.7% of CLTP) (101.62% of CLTP)

Thermal Power (MWt) 2381.0 2421.0 2419.0 (Percent of Current Licensed Power) 100,0 101.7 101.62 Steam Flow (Mlb/hr) 9.556 9,716 9.707 (Percent of Current Rated) 100.0 101.7 101.6 FW Flow (Mlblhr) 9,521 9.681 9.672 (Percent of Current Rated) 100.0 101.7 101.6 Dome Pressure (psia) 1020 1020 1020 Dome Temperature (*F) 547.0 547.0 547.0 FW Temperature (°F) 367.1 367.1 367.1 Full Power Core Flow Range (Mlb/hr) 55.1 to 77.2 56.5 to 77.2 56.4 to 77.2 (Percent of Current Rated) 75.0 to 105.0 76.9 to 105.0 76.8 to 105.0 1-8

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 1-3 Summary of Effect of TPO Uprate on Licensing Criteria Key L Explanation ~of Effect LOCA challenges to fuel No increase in peak clad Previous analysis accounted for Žý102% of (10 CFR 50, Appendix K) temperature (PCT), no change of licensed power, bounding TPO operation. No maximum LHGR required. vessel pressure increase.

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

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

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

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

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

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

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

Heat discharge to -1.1'F temperature increase. Small % power increase.

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

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

10 CFR 50, Appendix G Stability No direct effect of TPO uprate No increase in maximum rod line boundary.

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

current boundaries to preserve MWt-core flow boundaries as applicable for each stability option.

ATWS peak vessel pressure Slight increase (30 psig), must stay Slightly increased power relative to SRV within existing ASME Code capacity.

'Emergency" category stress limit.

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

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-1 Power/Flow Map for the TPO (101.7% of CLTP)

Core Flow (Mlbrn/hr) 0 to 20 30 40 50 60 70 80 130 120 2800 310 300 2400 90 80 20o0 0

L 70 t600 ~

0 60 2

I- 50 1200 C..

I- 40 800 30 20 400 10 0

0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-2 Power/Flow Map for the TPO (101.62% of CLTP)

Core Flow (MIIbmbhr) 0 10 20 30 40 50 60 70 so 130 120 2900 110 600 2400 90 9g0 2000 70 60 so 1200 t 40 30 20 400 10 0

0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

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NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-3 Reactor Heat Balance- TPO Power (101.7% of CLTP), 100% Core Flow Legend I-low, 11b1/h1r 11I l~nli Ihpy. B~t u'1bin 1',= Temperat tire, yF M.Moisture,. %

P Pressure, psia

• Conditions at upsiream side oaf TSV Core Thermal Povxr 2421 M Pump Heating 7A4 Cleanup Losses 3 Other System L.osses -0.6 Turbine Cycle Use 2424.5 MWt 1-12

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Figure 1-4 Reactor Heat Balance- TPO Power (101.62% of CLTP), 100% Core Flow Legend

= Flow, Ibm'lhr 10()2 H = Enthalpy, Btu/Ibm P F = Temperature, OF M = Moisture, % rr1.rri Main Steam Flow I 1.1' IlL' 1 IIrriI 9.707E+06 #

• P Pressure. psia 1191.5 H

• 039 M

• Carryunder = 0.22% 968 p

• 2419 Main Feed Flow A )

MWt Wd = I WK%/ 9.772E+06 # 9.672E+06 #

I 5204 5 H- 341.8 H 341.1 H 527.2 OF Total 367,7 OF 367.1 OF Cole Flow lJ 7.5E+06 Ah=0.7 H L.0OE-+05 #

407.6 H 429.2 OF HH Cleanup Demineralin 2r System 3.5(X)E +04 # Control Itod Driht I .0X0E+05 #

( ).0 H Feed Flow 519.8 H 972 0F 526.6 OF F

• Conditions at upstream side of TSV Core Thermal Power 2419.0 Pump Heating 7.4 Cleanup Losses -33 Other Svstem Losses -0.6 Turbine Cycle Use 2422.5 MWt 1-13

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 2.0 REACTOR CORE AND FUEL PERFORMANCE 2.1 FuEL DESIGN AND OPERATION At the TPO RTP conditions, all fuel and core design limits are met by the deployment of fuel enrichment and burnable poison, control rod pattern management, and core flow adjustments.

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

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

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

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

))

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

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

2-1

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 2.2.2 MCPR Operating Limit TLTR Appendix E shows that the changes in the Operating Limit Minimum Critical Power Ratio (OLMCPR) for a TPO uprate ((

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

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

2.3 REACTIVITY CHARACTERISTICS All minimum shutdown margin requirements apply to cold shutdown (< 212'F) conditions and are maintained without change. Checks of cold shutdown margin based on Standby Liquid Control System (SLCS) boron injection capability and shutdown using control rods with the most reactive control rod stuck out are made for each reload. The TPO uprate has no significant effect on these conditions; the shutdown margin is confirmed in the reload core design.

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

2.4 THERMAL HYDRAULIC STABILITY Option l-D is a solution combining prevention and detect-and-suppress elements (Reference 5).

The prevention portion of the solution is an administratively controlled exclusion region. The exclusion region calculation is a confirmation that regional mode instabilities are not probable.

The flow biased Average Power Range Monitor (APRM) scram provides automatic detection and suppression of core wide instabilities. This scram ensures the fuel cladding integrity safety limit is met for thermal hydraulic oscillations.

Thermal Power Optimization may affect the exclusion region slightly. However, the exclusion region is dependent upon the core loading, then reviewed and adjusted as required, for each reload core. The confirmation that regional mode reactor instability is not probable is also re-evaluated when the exclusion region is recalculated. These features will be analyzed for the first core reload analysis that incorporates the new rated power level.

2-2

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Thermal Power Optimization may also affect the SLMCPR protection confirmation slightly.

Changes to the nominal flow-biased APRM trip setpoint or the rated rod line require the hot channel oscillation magnitude portion of the detect-and-suppress calculation to be recalculated.

This calculation is not dependent upon the core and fuel design. However, the SL..MCPR protection calculation is dependent upon the core and fuel design and is performed for each reload. These features will be analyzed for the first reload analysis that incorporates the new rated power level. Therefore, a separate evaluation for Option I-D plants is not required for Thermal Power Optimization.

2.5 REACTIVITY CONTROL The generic discussion in TLTR Sections 5.6.3 and Appendix J.2.3.3 applies to the CNS plant.

The Control .Rod Drive (CRD) and CRD hydraulic systems and supporting equipment are not affected by the TPO uprate and no further evaluation of CRD performance is necessary.

2-3

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION 3.0 REACTOR COOLANT AND CONNECTED SYSTEMS 3.1 NUCLEAR SYSTEM PRESSURE RELIEF / OVERPRESSURE PROTECTION The pressure relief system prevents over-pressurization of the nuclear system during abnormal operational transients. The safety relief valves (SRVs) along with other functions provide this protection. Evaluations and analyses for the CLTP have been performed at 102% of CLTP to demonstrate that the reactor vessel conformed to ASME B&PV Code and plant Technical Specification requirements. There is no increase in nominal operating pressure for the CNS TPO uprate. There are no changes in the SRV setpoints or valve OOS options. There is no change in the methodology or the limiting overpressure event. Therefore, the generic evaluation contained in the TLTR is applicable.

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

3.2 REACTOR VESSEL The RPV structure and support components form a pressure boundary to contain reactor coolant and moderator, and form a boundary against leakage of radioactive materials into the drywell.

The RPV also provides structural support for the reactor core and internals.

3.2.1 Fracture Toughness The TLTR, Section 5.5.1.5 describes the RPV fracture toughness evaluation process. RPV embrittlement is caused by neutron exposure of the wall adjacent to the core including the regions above and below the core that experience fluence greater than or equal to I x 1017 n/cm2; this region is defined as the "beltline" region. Operation at the TPO conditions results in a higher neutron flux, which increases the integrated fluence over the period of plant life.

The fluence at CLTP was scaled for use at TPO. This fluence is used to evaluate the vessel against the requirements of 10CFR_50, Appendix G. The results of these evaluations indicate that:

(a) The upper shelf energy (USE) for the plate materials remains greater than 50 ft-lb for the design life of the vessel and maintains the margin requirements of IOCFR50 Appendix G.

Equivalent margin for the weld material remains greater than the required 35 ft-lb per Code Case N-512. The maximum % decrease for the plate beltline materials is 21% of 32 EFPY and the USE is 43.67 ft-lb for the limiting weld.

(b) The beltline material reference temperature of the nil-ductility transition (RTNDrT) remains below the 200'F screening criteria as defined in Reference 6.

3-1

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION (c) The current P-T curves bound TPO operation up to 28 EFPY. CNS will revise the curves for operation beyond 28 EFPY.

(d) The 32 effective full power year (EFPY) shift is increased, and consequently, requires a change in the adjusted reference temperature (ART), which is the initial RTNDT plus the shift.

These values are provided in Table 3-1.

(e) The reactor vessel material surveillance program consists of three (3) capsules.

One capsule, at the 300 azimuth, containing Charpy specimens was removed and tested in 1985 after 6.8 EFPY of operation. A second capsule, at the 300' azimuth, was removed and tested in 1991 after 11 .2 EFPY of operation. Specimens from this capsule were reconstituted and installed in the vessel in 1993; this capsule is scheduled to be removed in 2029. The remaining original capsule has been in the reactor vessel since plant startup. This capsule is scheduled to be removed in 2016.

TPO has no effect on the existing surveillance schedule.

The maximum normal operating dome pressure for TPO is unchanged from that for original power operation. Therefore, the hydrostatic and leakage test pressures are acceptable for the TPO.

Because the vessel is still in compliance with the regulatory requirements as demonstrated above, operation with TPO does not have an adverse effect (not exceeding regulatory requirements) on the reactor vessel fracture toughness.

3.2.2 Reactor Vessel Structural Evaluation Certain reactor vessel components are generically dispositioned without detailed structural analysis.

For components with no increase in flow, temperature, pressure, reactor internal pressure difference (R1PD), or other mechanical loads, no further evaluation is required. In addition, previous GE BWR TPO experience has demonstrated that, using the evaluation process documented in Appendix I of the TLTR, numerous components are below the 0.5 CUF criteria, thus requiring no further evaluation.

3-2

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 3-3

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION 1]

The above components are unaffected by TPO operating conditions.

Because the Recirculation Inlet Nozzle (N2), Steam Outlet Nozzle (N3), Core Spray Nozzle (N5),

Top Head Spray (N6A), Top Head Instrumentation Nozzle (N6B), and Top Head Vent Nozzle &

Bolts (N7), Jet Pump Instrumentation Nozzle (NM), CRD-HSR Nozzle (N9); Core AP and Liquid Control Nozzle (N 10), 2" Instrumentation Nozzles (NI 1, N12, N 16), Drain Nozzle (NI 5), Support Skirt, In-Core Housing Penetrations, Vessel Cylindrical Shell, Basin Seal Skirt, Stabilizer Bracket, Insulation Bracket, Top Head Lifting Lugs, Steam Dryer Hold Down Bracket, Guide Rod Bracket, Steam Dryer Bracket, Feedwater Sparger Bracket, Core Spray Bracket, Jet Pump Support Pads, and Surveillance Bracket, and IRM/SRM Dry Tube and Power Range Detector components have fatigue usages less than or equal to 0.5 or were originally exempt from fatigue requirements, these components are confirmed to be consistent with the generic disposition provided in the TLTR, thus requiring no further evaluation for TPO.

Seal Leak Detection Nozzles (N 13, N14) were not considered to be a pressure boundary component at the time that the OLTP evaluation was performed, and have not been evaluated for TPO.

The effect of TPO was evaluated to ensure that the reactor vessel components continue to comply with the existing structural requirements of the ASME Boiler and Pressure Vessel Code. For the components under consideration, the 1965 Code with Addenda to and including Winter 1966, which is the code of construction, was used as the governing code. However, if a component's design has been modified, the governing code for that component was the code used in the stress analysis of the modified component. The following components were modified since the original construction of Cooper Nuclear Station:

3-4

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION

" Recirculation Outlet Nozzle (N]): This nozzle was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section 11, 1983 Edition.

" Recirculation Inlet Nozzle (N2): This nozzle was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section .111, 1983 Edition.

" Feedwater Nozzle (N4): This nozzle was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Summer 1976.

" Core Spray Nozzle (N5): This nozzle was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section HI, 1983 Edition.

" Jet Pump Instrumentation Nozzle (N8): This nozzle was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code, Section 1II, 1980 Edition with Addenda to and including Winter 198 1.

• CRD Hydraulic System Return Nozzle (N9): This nozzle was modified (capped) and the governing Code for the evaluation/modification is the ASSME Boiler and Pressure Vessel Code,Section III, 1974 Edition with Addenda to and including Winter 1975.

• Core AP & Liquid Control Nozzle (N 10): This component was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code,Section III, 1980 Edition with Addenda to and including Winter 1981.

• CRD Penetration: This component was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Addenda to and including Winter 1966.

" IRM/SRM Dry Tube, Power Range Detector: This component was modified and the governing Code for the evaluation/modification is the ASME Boiler and Pressure Vessel Code,Section III, 1977 Edition with Addenda to and including Summer 1977.

Typically, "new" stresses are determined by scaling the "original" stresses based on the TPO conditions (pressure, temperature, and flow). The analyses were performed for the design, normal

& upset, and emergency & faulted conditions. There are no changes in annulus pressurization, jet reaction, pipe restraint or fuel lift loads.

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

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION 3.2.2.2 Normal and Upset Conditions The reactor coolant temperature and flows (except core flow) at TPO conditions are only slightly changed from those at current rated conditions. Evaluations were performed at conditions that bound the slight change in operating conditions. The type of evaluations is mainly reconciliation of the stresses and usage factors to reflect TPO conditions. A primary plus secondary stress analysis was performed showing TPO stresses still meet the requirements of the ASME Code, Section TH1, and Subsection NB for all components. Lastly, the fatigue usage was evaluated for the limiting location of components with a usage factor greater than 0.5. The fatigue analysis results for the limiting components are provided in Table 3-2. The analysis results for TPO show that all components meet their ASME Code requirements.

3.2.2.3 Emergency and Faulted Conditions The stresses due to emergency and faulted conditions are based on loads such as peak dome pressure, which are unchanged. These loads remain unchanged and bound the TPO values.

Therefore, ASME Code Section II. requirements are met for all RPV components.

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

3.3.1 Reactor Internal Pressure Difference The reactor internal pressure differences (RI.PDs) are affected more by the maximum licensed core flow rate than by the power level. The maximum flow rate is not to be changed for the TPO uprate. The effect due to the changes in loads for both Normal and Upset conditions is reported in Section 3.3.2. The Normal and Upset evaluations of RIPDs for the TPO uprate increase slightly or are bounded by the current analyses that assumed an initial power level of 102%

CLTP. The Emergency and Faulted evaluations of RJPDs for the TPO uprate are bounded by the current analyses that assumed an initial power level of 102% CLTP.

Fuel Bundle Lift Margins are calculated for the Emergency and Faulted conditions for 102%

CLTP, to demonstrate that fuel bundles would not be lifted under the worst conditions. The Fuel Lift Margins at Normal and Upset conditions are bounded by Emergency and Faulted conditions.

As an older plant, the Cooper licensing basis does not require the hydraulic lift forces to be combined with seismic loads. Thus, the hydraulic control rod guide tube (CRGT) lift forces are not calculated.

3.3.2 Reactor Internals Structural Evaluation The RPV internals consist of the (Core Support Structures) CSS components and non-CSS components. The RPV Internals are not certified to the ASME Code; however, the requirements of the ASME Code are used as guidelines in their evaluations. The evaluations/stress 3-6

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION reconciliation in support of the TPO was performed consistent with the design basis analysis of the components. The RPV Internals evaluated are:

Core Support Structure Components (CSS)

" Shroud

• Shroud Support

" Core Plate

" Top Guide

" Control Rod Drive Housing

" Control Rod Guide Tube

" Orificed Fuel Support

" Fuel Channel Non-Core Support Structure Components (Non-CSS)

• Steam .Dryer

" Feedwater Sparger

" Jet Pump Assembly

• Core Spray Line and Sparger

• Access Hole Cover

• Shroud Head and Steam Separator Assembly

• In-core Housing and Guide Tube

• Jet Pump Instrument Penetration Seal

• Core DP/Standby Liquid Control System The original configurations of the RPV internals are considered in the TPO evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation. The evaluations are based on GE-14 fuel, The following loads were considered in the evaluation of the RPV internals.

" RPV design pressure

• Deadweight

• Seismic, loads

" RIPD

" Acoustic and flow induced loads

" Thermal effects 3-7

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION RPV design pressure remains unchanged. The temperature in the annulus changes insignificantly (< 0.5 'F). Seismic load remains unchanged. RIPD loads for normal, upset conditions decrease for all components except the shroud head and steam dryer. For the shroud head, they increase with respect to C[,TP. However, remain bounded by the design basis RIPDs.

For the steam dryer adequate stress margins exist in the normal and upset conditions to accommodate the increase in RIPMs. The faulted condition RIPD for all components remain unchanged. Acoustic and flow induced loads remain unaffected.

Based on the above, the governing stresses for all RPV internal components in the TPO condition remain bounded by their CLTP values or bounded by the current design basis (see Table 3-3). Therefore, the qualification of the RPV internals in the CLTP condition remains valid for the TPO condition.

3.3.3 Steam Separator and Dryer Performance The steam separator and dryer performance evaluation is described in TLTR Section 5.5.1.6. As described in the TLTR, no additional evaluation of the steam separator and dryer performance is necessary unless the plant has been ((

)), the generic evaluation in the TLTR is applicable and no further evaluation is needed.

3.4 FLOW-INDUCED VIBRATION The process for the reactor vessel internals vibration assessment is described in TLTR Section 5.5.1.3. An evaluation determined the effects of flow-induced vibration (FIV) on the reactor internals at TPO RTP 101.7% and rated core flow. The vibration levels for the TPO uprate conditions were estimated from vibration data recorded during startup testing of the NRC designated prototype plant (FitzPatrick) at CNS and from operating experience at similar plants.

These expected vibration levels were compared with established vibration acceptance limits.

The following components are evaluated for the TPO uprate:

Comnponent(s)K Process Parameter(s) TPO Evaluation Shroud Steam flow at TPO RTP is -2% Slight increase in FIV. Cooper Shroud Head and Separator greater than CLTP core flow and steam flow at Steam Dryer TPO conditions are less than the prototype plant core flow and steam flow at rated conditions. Prior to exceeding CLTP and ascension to TPO, CNS will ensure compliance with Reg. Guide 1.20.

Core Spray (CS) Line Core flow at TPO RTP is identical to No change Low Pressure Coolant CLTP Injection (LPCI) Coupling Control Rod Guide Tube In-Core Guide Tubes 3-8

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION Component(s)  : Process Paramneter(s) TPO Evaluation.

Fuel Channel Core flow at TPO RTP is identical to No change LPRM/IRM Tubes CLTP No change in core flow distribution Jet Pumps Jet pump flow at TPO RTP is identical No meaningful change to CLTP Fluid temperature increases - 1 *F Jet Pump Sensing Lines Vane passing frequency of No change in possibility of recirculation pumps resonance FW Sparger FW flow at TPO RTP is -2% greater Slight increase in FIV. Cooper than CLTP FW flow at TPO condition is less than the FW flow at the prototype plant at rated conditions.

The calculations for the TPO uprate conditions indicate that vibrations of all safety-related reactor internal components are within the GE acceptance criteria. The analysis is conservative for the following reasons:

" The GE criteria of 10,000 psi peak stress intensity is much more conservative than the ASME allowable peak stress intensity of 13,600 psi for service cycles equal to 10".

• The modes are absolute summed.

• The maximum vibration amplitude in each mode is used in the absolute sum process, whereas in reality the vibration amplitude fluctuates.

Therefore, it is concluded that the flow-induced vibrations of the reactor vessel internals remain within acceptable limits.

The safety-related Main Steam (MS) and FW piping have minor increased flow rates and flow velocities resulting from the TPO uprate. The MS and FW piping experience increased vibration levels, approximately proportional to the square of the flow velocities and also in proportion to any increase in fluid density. The decrease in FW fluid density for TPO uprate conditions as a result of -2 0 F temperature increase is insignificant. Prior to exceeding CLTP and ascension to TPO, CNS will ensure compliance with Reg. Guide 1.20.

The safety-related thermo wells and sample probes in the MS, FW, and Recirculation piping systems are also adequate for the TPO operating condition.

3.5 PIPING EVALUATION 3.5.1 Reactor Coolant Pressure Boundary Piping The methods used for the piping and pipe support evaluations are described in TLTR Appendix K. These approaches are identical to those used in the evaluation of previous BWR power uprates of up to 20% power. The effect of the TPO uprate with no nominal vessel dome pressure increase is negligible for the RCPB portion of all piping except for portions of the FW 3-9

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION lines, MS lines, and piping connected to the FW and MS lines. The following table summarizes the evaluation of the piping inside containment.

Component(s) /lConcern,< Process Parameter(s) TPO. Evalujation Recirculation System Nominal dome pressure at TPO RTP is identical Negligible change in to CLTP pipe stress Pipe Stresses Recirculation flow at TPO RTP is identical to Negligible effect on Pipe Supports CLTP pipe supports Small increase in core pressure drop of < 1 psi Recirculation fluid temperature increases ~-1°F MS and Attached Piping Nominal dome pressure at TPO RTP is identical Plant specific (inside Containment) (e.g., SRV to CLTP evaluation performed Discharge Line (SRVDL) piping up to Steam flow at TPO RTP is -2% greater than first anchor, Reactor Core Isolation CLTP Cooling (RCIC) / High Pressure Minor decrease in main steam line (MSL) Negligible change in Coolant Injection (HPCI) piping pressure < 2 psi pipe stress (Steam Side), MS drain lines, RPV pipe supports head vent line piping located inside containment) Minor increase inthe Pipe Stresses potential for FAC Pipe Supports (FAC concerns are Flow-accelerated erosion/corrosion covered by existing (FAC) piping monitoring program)

FW and Attached Piping Nominal dome pressure at TPO RTP is identical Plant specific (Inside Containment) to CLTP evaluation performed FW flow at TPO RTP is -2% greater than CLTP Minor change in FW line pressure Pipe Stresses Fluid temperature remains the same pipe stress Pipe Supports Negligible effect on pipe supports FAC Minor increase in the potential for FAC (FAC concerns are covered by existing piping monitoring program)

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

For the MS and FW lines, supports, and connected lines, the methodologies as described in T

LTR Section 5.5.2 and Appendix K were used to determine the percent increases in applicable ASME Code stresses, displacements, cumulative usage factors (CUF), and pipe interface component loads (including supports) as a function of percentage increase in pressure (where applicable), temperature, and flow due to TPO conditions. The percentage increases were 3-10

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION applied to the highest calculated stresses, displacements, and the CUF at applicable piping system node points to conservatively determine the maximum TPO calculated stresses, displacements and usage factors. This approach is conservative because the TPO does not affect weight and all building filtered loads (e.g., seismic loads are not affected by the TPO). The factors were also applied to nozzle load, support loads, penetration loads, valves, pumps, heat exchangers and anchors so that these components could be evaluated for acceptability, where required. No new computer codes were used or new assumptions introduced for this evaluation.

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

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

The design adequacy evaluation results show that the requirements of ANSI (USAS) B331.1, B301.7 Power Piping and ASME,Section III, Subsection ND (as applicable) requirements are satisfied for the evaluated piping systems. Therefore, the TPO does not have an adverse effect on the MS piping design.

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

Irosion/ Corrosion The carbon steel MS piping can be affected by FAC. FAC is affected by changes in fluid velocity, temperature and moisture content. CNS has an established program for monitoring pipe wall thinning in single and two-phase high-energy carbon steel piping. The variation in velocity, temperature, and moisture content resulting from the TPO uprate are minor changes to parameters affecting FAC.

No changes to piping inspection scope and frequency are required to ensure adequate margin for the changing process conditions. The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that 3-11

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION the TPO uprate has no adverse effect on high-energy piping systems potentially susceptible to pipe wall thinning due to FAC.

FW Piping System Evaluation The FW piping system (inside containment) was evaluated for compliance with the ASME Section 1.11 Code stress criteria, and for the effects of thermal expansion displacements on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, and valves were also evaluated.

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

The design adequacy evaluation shows that the requirements of ANSI (USAS) B31.1, B334.7 Power Piping and ASME, Section 111, Subsection ND-3600 requirements remain satisfied.

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

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

Erosion / ('OTOSI Oi The carbon steel FW piping can be affected by FAC. FAC in the FW piping is affected by changes in fluid velocity and temperature. CNS has an established program for monitoring pipe wall thinning in single and two-phase high-energy carbon steel piping. The variation in velocity and temperature resulting from the TPO uprate are minor changes to parameters affecting FAC, No changes to piping inspection scope and frequency are required to ensure adequate margin exists for the TPO process conditions. The continuing inspection program will take into consideration adjustments to predicted material loss rates used to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. This program provides assurance that the TPO uprate has no adverse effect on high energy piping systems potentially susceptible to pipe wall thinning due to FAC.

3-12

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 3.5.2 Balance-of-Plant Piping Evaluation This section addresses the adequacy of the BOP piping design (outside of the RCPB) for operation at the TPO conditions. The evaluation of the BOP piping and supports was performed in a manner similar to the evaluation of RCPB piping systems and supports (Section 3.5.1).

Pipe Sulpporfs Because there is no change in the MS temperature, there is no change in the MS pipe support loads. The supports for piping that contains fluid that increases in temperature (e.g. the FW piping) have slightly increased pipe support loadings. However, when considering the loading combination with other loads that are not affected by the TPO uprate, such as seismic and deadweight, the combined support load increase is insignificant.

For the MS system piping outside containment, the turbine stop valve closure transient was reviewed and determined to be bounded by the TPO uprate and the no new piping analysis was required.

For the FW system piping outside containment, no existing transient analysis that is impacted by the TPO was identified. A review of the increases in FW flow associated with the TPO indicates that piping load changes do not result in any load limit being exceeded.

Erosion / Corrosion The integrity of high-energy piping systems is assured by proper design in accordance with the applicable codes and standards. Piping thickness of carbon steel components can be affected by FAC. CNS has an established program for monitoring pipe wall thinning in single phase and two-phase high-energy carbon steel piping. FAC rates may be influenced by changes in fluid velocity, temperature, and moisture content.

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

3-13

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 3.6 REACTOR RECIRCULATION SYSTEM The Reactor Recirculation System (RRS) evaluation process is described in TLTR Section 5.6.2.

The TPO uprate has a minor effect on the RRS and its components. The TPO uprate does not require an increase in the maximum core flow. No significant reduction of the maximum flow capability occurs due to the TPO uprate because of the small increase in core pressure drop (< 1 psi). The effect on pump NPSH at TPO conditions is negligible. An evaluation has confirmed that no significant increase in RRS vibration occurs from the TPO operating conditions.

The cavitation protection interlock for the recirculation pumps and jet pumps is expressed in terms of feedwater flow. This interlock is based on sub-cooling and thus is a function of absolute feedwater flow rate and feedwater temperature at less than full thermal power operating conditions. Therefore, the interlock is not changed by TPO.

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

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

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

3.1.0 RESIDUAL HEAT REMOVAL SYSTEM.

The RHR system is designed to restore and maintain the coolant inventory in the reactor vessel and to remove sensible and decay heat from the primary system and containment following reactor shutdown for both normal and post accident conditions. The RHR system is designed to function in several operating modes- The generic evaluation provided in TLTR Sections 5.6.4 and J.2.3.13 is applicable to CNS.

3-14

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION The following table summarizes the effect of the TPO on the design basis of the RHR system.

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

SPC mode maintains the long-term pool temperature below the design limit.

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

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

Steam Condensing Mode Decay Heat removal Cooper does not have a Steam Condensing Mode of RHR Fuel Pool Cooling Assist Supplemental fuel pool cooling inthe See Section 6.3.1 event that the fuel pool heat load exceeds the heat removal capability of the Fuel Pool Cooling system.

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

3.11 REACTOR WATER CLEANUP SYSTEM The generic evaluation of the Reactor Water Cleanup (RWCU) system provided in TLTR Sections 5.6.6 and 12.3.4 is applicable to CNS. The performance requirements of the RWCU system are negligibly affected by TPO uprate. There is no significant effect on operating temperature and pressure conditions in the high-pressure portion of the system. Steady power level changes for much larger power uprates have shown no effect on reactor water chemistry and the performance of the RWCU system. Power transients are the primary source of challenge to the system, so safety and operational aspects of water chemistry performance are not affected by the TPO.

3-15

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-1 Adjusted Reference Temperatures - 40 Year Life (32 EFPY)

Lower Shell Thickness in inches = 6.38 32 EFPY Peak 1.0,tluencez 1,23E-16 n/cmr'2 32 EFPYPeakl14Ttluence= 8.39E+17 n/cm^2 Lower-ntermedlate S1helland All Welds Thiclsiess inninches 5.38 32 EFPY Peak ID, fluence= 1,68E+18 nlcm"2 32 EFPYPeakl/4Tfluence= 1.22E+18 n/cm'2 Adjusted Inbital 114T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT %Cu %Hi CF CF RTndt Fluence A RTUndl a, r'7 Margin Shift ART (1) 'F nlcm`2 'F 'F ' F PLATES:

Lower G2803-1 C2274-1 0.20 0,68 153.0 2417 14 8,39E+17 92.5 0 17 17 110 1215 G2803-2 C2307.1 0,21 0,73 1628 25712 0 839E*17 98.4 0 17 17 115 115.4 G2803-3 (2) C2274-2 0,20 0,68 153.0 241.7 -8 8.39E+17 92.5 0 17 17 110 101.5 Lower.Intermediate G2801-7 C2407-1 0,13 0.65 923 145.8 -10 1.22E'18 66.7 0 17 17 64 73.7 (2802,1 C2331-2 0,17 0,58 125,3 190,0 10 1.22E+18 90.5 0 17 17 108 117.5 G2802-2 C2307-2 0.21 0,73 1626 257,2 -20 1.22E018 117,6 0 17 17 135 114,6 WELDS:

Lower Longitudinal 2-233(3) 12420-37083724 0.22 1.02 234.5 375.2 -50 8.39E+17 143.6 0 28 28 172 121,6 Lower-Intermediate Longitudinal 1-233(3) 27204 12008 0.19 0.97 215.7 345.0 -50 1.22E+18 157.8 0 28 28 180 135.8 Lower to Lower-Intermediate Circumferential 1-240 (3)(4) 21935 0,20 0,69 175,3 280.5 -50 8&39E+17 107.3 0 26 28 135 85.3 INTEGRATED SURVEILLANCE PROGRAM (ISPX6):

Plate(6) C2307-2 0,21 0.76 114.6 2555 -20 1,22E+16 118.8 0 17 17 134 113.8 Plate (7) C2331-2 0.16 0,62 118.5 158.1 10 1.22E+t18 72.3 0 17 17 s0 99,3 Weld M 0291 0.23 0.76 194.6 -50 1.22E.18 88,9 0 28 28 117 86.9 (1) CNS has removed 2 surveillance capsules, The adjusted CFs include factois determined per RG1.99 Position 2.1. Position 21 all*ks for the use ot 0.5*oA in the Margin Term (2) CNS adopted the ISP best estimate chemistry for this mateial. which Is evaluated in the ISP section below, The upper portion of this table evaluates the vessel materials using the original mterial properties. The bottom portin of this table evaluates all surveillance materials using the ISP properties provided by NPPD (3) The adjusted CF is conservatively applied to all materiais in order to remain consistent with the previously licensed approach..

(4) CNS previously used the knver-inlermediate shell flhence, this evaluation uses the more-representative tower shell fluence for this location, (5) Ptocedures defined in RG1.,99 are applied to determine the ART considering the Integrated Surveillance Program, For the weld material, the higher fluence is used to be applicable to all weld locations.

(6) The ISP plate is the identical heat to that in the vessel and is ptesented using the ISP chenistry with the vessel plate Initial RT,, and fluence. As defined by ISP, as there are two (2) sets of credible capsule data available for this material, the OF is obtained from RG1.99 Position 21 and equals 25&35'(162.11164.8) = 255,5, (7) This ISP plate is the identical heat to a second plate in the vessel (although not one of the representative ISP materials) and is presented using the ISPchemistrlrvat the vessel plate Initial RT,,,1 and fluence. As defined in ISP, as there are six (68sets of credible capsule data available for this material, the CF is obtained trom RG1,99 Position 2.1, This material was Included in SSP capsules. and not in the Cooper capsule or representative ISP capsule, The adjusted CF is equal to 149,52 *(125,311185) = 158.1, (6) The ISP weld is not the identical heat to th*t in the vessel and is presented using the ISP chemistry and CF and applied to the initial RT*,ý for the limiting CNS weld, which is Heat 27204 12008.

3-16

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-2 P+Q Stresses and CUFs of Limiting Components [

P + Q Stress CUF (ksi)i~ (ASME Code Allowable = 1.0)

Component CLTP TPO121 Allowable CLTP TPO'..

(2381 Mt)ý (2428 MWt) '(ASME Code =(238iMWt) (2428 MWt)

__________________ -Limit)

Recirculation Outlet Nozzle 40.5 40.5 47.4 0.839 0.02141 (Bimetal Weld)

Feedwater Nozzle 191 Nozzle 48.9110,81:] 49.2/10.9[3] 40,05 0,63 0.71 Safe End 76.1/21.2M3 76.6121.3P 55.94 0.98 0.997 Nozzle/Shell Junction 59.7 60.1 80.1 0.87 0 .3 7 16[15 CRD Penetration Stub Tube to Bottom Head 54,5 54.8 6949" 0.92161 0.87M Stub Tube Housing Junction 38.1 38.3 47.4 Main Closure Region - Flange 78,5 78.5 80.0 1.008 0.10 141 Main Closure Region - Bolts 129.2 129.2 129.9 0.82 0.98 Shroud Support 50.84" 50.8 69.9 0.567 0.0214]

Notes:

I. Only the limiting components with CUF values > 0.5 arc provided.

2. TPO was conservatively evaluated for 102% of OLTP, which bounds TPO operating conditions.

3. Thennal Bending includedflhermal bending removed. P + Q stresses are acceptable per CLTP elastic-plastic analysis, which is valid for TPO conditions.

4. This component was re-evaluated considering a more representative and less conservative treatment of the duty cycles.

5. The CUF values for the nozzle/shell junction include system cycling only. The total CUF (system plus rapid cycling) is greater than 1.0. Rapid cycling effects are managed using the inspection requirements defined in BWROG Letter BWROG-00068, "Alternate BWR Feedwater Nozzle Inspection Requirements", GE-NE-523-A71-0594-A, Revision 1, June 6, 2000, as approved by the NRC, consistent with a plant-specific fracture mechanics evaluation..

6. Previously listed as 60.0 ksi.

7. Only the limiting CUF is provided,

8. Previously listed as 26.2 ksi.

9. Values presented are the maximum CUF (total) for all nodes in that section (i.e. Nozzle, Safe End, Nozzle-Shell Junction).

3-17

NEDO-.33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 3-3 Summary of TPO Stresses for RPV Internal Components Itemn Evaluation Load Level TPO Result 1 Shroud N, U, E, &F Bounded by CLTP 2 Shroud Support N, U, E, &F Bounded by CLTP Core Plate N, U, E, &F Bounded by CLTP 4 Top Guide N, U, E, &F Bounded by CLTP 5 Control Rod Drive Housing (CRDH) N, U, E, &F Bounded by CLTP 6 Control Rod Guide Tube (CRGT) N, U, E, &F Bounded by CLTP 7

Orificed Fuel Support (OFS).

N, U, E, &F Bounded by CLTP 8 Fuel Channel N, U, E, &F Bounded by CLTP

&F Bounded by current N, U, E, &Fbysi e 9 Steam Dryer design basis 1

10. . . .... . .. Feedwater -

(FW) Sparger N, U, E, &F Bounded ..by CLTP 11 Jet Pump Assembly N, U, E, &F Bounded by CLTP 12 Core Spray Line and Sparger N, U, E, & F Bounded by CLTP 13 Access Hole Cover (AHC). N, U, E, &F Bounded by CLTP Bounded by current 14 Shroud head and Separator Assembly N, U, E, &F design basis 15 In-core Housing and Guide Tube (ICHGT). N, U, E, &F Bounded by CLTP 16 Jet Pump Instrument Penetration Seal N, U, E, &F Bounded by CLTP 17 jCore DPiStand by Liquid Control System. N, U, E, & ou nd~ed-by CLTP Where: N -Normal, U - Upset, E - Emergency, F - Faulted 3-18

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 4.0 ENGINEERED SAFETY FEATURES 4.1 CONTAINMENT SYSTEM PERFORMANCE TLTR Appendix G presents the methods, approach, and scope for the TPO uprate containment evaluation for LOCA. The current containment evaluations were performed at 102% of CLTP.

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

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

~Topic Ke Parameters

.**Key aaotr , :* TPO Effect~ i:

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

conditions, 4.1.1 Generic Letter 89-10 Program The motor-operated (MOV) requirements in the USAR were reviewed, and no changes to the functional requirements of the GL 89-10 MOVs are identified as a result of operating at the TPO RTP level. Because the previous analyses were based on 102% of CLTP, there are no increases in the pressure or temperature at which MOVs are required to operate. Therefore, the GL 89-10 MOVs remain capable of performing their design basis function.

4-1

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 4.1.2 Generic Letter 95-07 Program The commitments relating to the GL 95-07, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," have been reviewed and no changes are identified as a result of operating at the TPO RTP level. Valves in the RHR, RCIC, HPCI, and CS were included in the evaluation. Because the previous analyses were based on 102% of CLTP, there is no change in the environmental conditions at which the valves are required to operate. The process parameters for these systems do not change as a result of the TPO uprate. Therefore, the valves remain capable of performing their design basis function.

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

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

4.2.2 High Pressure Core Spray The High Pressure Core Spray system is not applicable to CNS 4.2.3 Core Spray The Core Spray (CS) system sprays water into the reactor vessel after it is depressurized. The primary purpose of the CS system is to provide reactor vessel coolant makeup for a large break LOCA and for any small break LOCA after the RPV has depressurized. It also provides spray cooling for long-term core cooling in the event of a LOCA. The generic evaluation of the CS system provided in TLTR Section 5.6.10 is applicable to CNS. The ability of the CS system to perform required safety functions is demonstrated with previous analyses based on 102% of 4-2

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION CLTP. Therefore, all safety aspects of the CS system are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

4.2.4 Low Pressure Coolant Injection The LPCI mode of the RH.R system is automatically initiated in the event of a LOCA. The primary purpose of the LPC1 mode is to provide reactor vessel coolant makeup during a large break LOCA or small break LOCA after the RPV has depressurized. The generic evaluation of the LPCI mode provided in TLTR Section 5.6.4 is applicable to CNS. The ability of the RHR system to perform required safety functions required by the LPCI mode is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the RHI- system LPCI mode are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

4.2.5 Automatic Depressurization System The Automatic Depressurization System (ADS) uses relief valves or SRVs to reduce the reactor pressure following a small break LOCA when it is assumed that the high-pressure systems have failed, This allows the CS and LPCI to inject coolant into the RPV. The ADS initiation logic and valve control is not affected by the TPO uprate. The generic evaluation of the ADS provided in TLTR Section 5.6.8 is applicable to CNS. The ability of the ADS system to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the ADS are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

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

4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE The Emergency Core Cooling System (ECCS) is designed to provide protection against postulated Loss-of-Coolant Accidents (LOCA) caused by ruptures in the primary system piping.

The current 10 CFR 50.46, or LOCA, analyses for the CNS plant has been performed at 102% of CLTP, consistent with Appendix K. Experience with power uprates up to 20% has, shown that there is substantial margin to the IOCFR50.46 criteria, including PCT, local cladding oxidation, core wide metal water reaction, cool able geometry and long-term core cooling.

Reference 7 provides justification for the elimination of the 1600'F Upper Bound PCT limit and generic justification that the Licensing Basis PCT will be conservative with respect to the Upper Bound PCT. The NRC SER for Reference 7 accepted this position, noting that because plant-4-3

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION specific Upper Bound PCT calculations have been performed for all plants, other means may be used to demonstrate compliance with the original SER requirements.

These other means are acceptable provided there are no significant changes to a plant's configuration that would invalidate the existing Upper Bound PCT calculations. This is the case for the CNS TPO. There are no changes in the plant configuration that would affect the PCT and there is no change in the Licensing Basis PCT. Therefore, the requirements of Reference 7 are satisfied to disregard the Upper Bound PCT limit.

The analysis presented in the relevant report (Reference 8) demonstrates that for GEI4 fuel, the limiting break and single failure combination is the maximum recirculation line break with battery failure for both Nominal and Appendix K assumptions. Based on the limiting large break and applying the SAFER/GESTR-LOCA methodology, the CNS ECCS-LOCA analysis was performed for the limiting LOCA event for GEl4 fuel. The results are summarized in Table 4-1.

These results meet all licensing limits of IOCFR50.46 and SAFER/GESTR methodology conditions in Reference 7.

Therefore, the pre-TPO SAFER/GESTR LOCA analysis for GE-supplied fuel bounds the 1.7%

TPO uprate for CNS.

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

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

4.6 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM Cooper Nuclear Station does not have an MSIV Leakage Control system.

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NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION 4.7 POsT-LOCA COMBUSTIBLE GAS CONTROL SYSTEM The Combustible Gas Control System (CGCS) maintains the post-LOCA concentration of oxygen or hydrogen in the containment atmosphere below the flammability limit. The generic evaluation of the CGCS provided in TLTR Section J.23.10 is applicable to CNS. The metal available for reaction is unchanged by the TPO uprate and the hydrogen production due to radiolytic decomposition is unchanged because the system was previously evaluated for accident conditions from 102% of CLTP. Therefore, the current evaluation is valid for the TPO uprate.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 4-1 Cooper ECCS-LOCA Analysis Results for GE14 Parameter MELLLA Analysis Limit 0

Nominal PCT 1359 F Appendix K PCT 2027°F Licensing Basis 204OF <22OQ0*F PCT Maximum Local Oxidation Core-Wide Metal- <0.1% <1.0%*

Water Reaction

• IOCFR5046 ECCS-LOCA Analysis Acceptance Criteria 4-6

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 5.0 INSTRUMENTATION AND CONTROL 5.1 NSSS MONITORING AND CONTROL The instruments and controls that directly interact with or control the reactor are usually considered within the NSSS. The NSSS process variables and instrument setpoints that could be affected by the TPO uprate were evaluated.

5.1.1 Neutron Monitoring System 5.1.1.1 Average Power Range Monitors, Intermediate Range Monitors, and Source Range Monitors The APRMs are re-calibrated to indicate 100% at the TPO RTP level of 2421.0 MWt. The APRM high flux scram and the upper limit of the rod block setpoints, expressed in units of percent of licensed power, are not changed. The flow-biased APRM trips, expressed in units of absolute thermal power (i.e., MWt), remain the same however, in order to accommodate limits in the Stability Region, new flow-biased APRM Analytical Limits (ALs) were established that conservatively bound the entire operating envelope. This approach for the CNS TPO uprate follows the guidelines of TLTR Section 5.6.1 and Appendix F, which is consistent with the practice approved for GE BWR uprates in ELTRI (Reference 2).

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

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

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

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION the higher average local flux. The RBM instrumentation is not significantly affected by the TPO uprate conditions, and no change is needed.

5.1.2 Rod Worth Minimizer The Rod Worth Minimizer (RWM) does not perform a safety-related function. The function of the RWM is to support the operator by enforcing rod patterns until reactor power has reached appropriate levels. The power-dependent setpoints for the RWM are included in Section 5.3.8.

5.2 BOP MONITORING AND CONTROL Operation of the plant at the TPO RTP level has minimal effect on the BOP system instrumentation and control devices. The improved FW flow measurement, which is the basis for the reduction in power uncertainty, is addressed in Section 1.4. All of the control systems and instrumentation have sufficient range/adjustment capability for use at the TPO uprate conditions. No safety-related BOP system setpoint changes are required as a result of the TPO uprate. The plant-specific instrumentation and control design and operating conditions are bounded by those used in the evaluations contained in the TLTR.

5.2.1 Pressure Control System The .Pressure Control System (PCS) provides a fast and stable response to steam flow changes so that reactor pressure is controlled within allowable values. The turbine utilizes a Digital Electro-Hydraulic (DEH) control system consisting of solid state governing devices, governor startup control devices, emergency devices for turbine and plant protection (over speed governor, master trip, vacuum trip, motoring protection, thrust bearing wear trip, low bearing oil pressure trip) and special control and test devices. The system operates the main stop valves, governor valves, bypass valves, reheat stop and interceptor valves, and other protective devices.

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

CNS has demonstrated acceptable pressure control performance at current rated conditions and has in excess of the -2% steam flow margin needed for the TPO uprate. The existing DEH electronic controls as designed for the current 100% CLTP conditions are adequate and require no electronic component changes for the TPO uprate conditions.

No modification is required to the steam bypass valves. No modifications are required to the operator interface indications, controls or alann annunciators provided in the main control room.

The required adjustments are limited to "tuning" of the control settings that may be required to operate optimally at the TPO uprate power level.

PCS tests, consistent withthe guidelines in TLTR Appendix L, will be performed during the power ascension phase.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 5.2.2 EHC Turbine Control System CNS does not have an EHC system. The DEH system was discussed in Section 5.2.1. The existing DEH electronic controls as designed for the current 100% CLTP conditions are adequate and require no electronic component changes for the TPO uprate conditions.

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

5.2.4 Leak Detection System The setpoints associated with leak detection have been evaluated with respect to the -2% higher steam flow and --20 F increase in FW temperature for the TPO uprate. Each of the systems, where leak detection potentially could be affected, is addressed below.

Main Steam Tunnel Temperature Based Leak Detection The -21F increase in FW temperature for the TPO uprate decreases leak detection trip avoidance margin. As described in TLTR Section F.4.2.8, the high steam tunnel temperature setpoint remain unchanged.

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

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

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION HPCI System Temperature Based Leak Detection The TPO uprate does not increase the nominal vessel dome pressure or temperature. Therefore, there is no change to the HPCI system temperature or pressure, and thus, the HPCI temperature based leak detection system is not affected.

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

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

5.3 TECHNICAL SPECIFICATION INSTRUMENT SETPOINTS The determination of instrument setpoints is based on plant operating experience, conservative licensing analyses or limiting design/operating values. Standard GE setpoint methodologies (References 9 and 10) are used to generate the allowable values (AV) and nominal trip setpoints (NTSP) related to any AL change, as applicable. Each actual trip setting is established to preclude inadvertent initiation of the protective action, while assuring adequate allowances for instrument accuracy, calibration, drift and applicable normal and accident design basis events.

Table 5-1 lists the ALs that change based on results from the TPO evaluations and safety analyses. In general, if the AL does not change in the units shown in the Technical Specifications, then no change in its associated plant AV and NTSP is required, as shown in the Technical Specifications. Changes in the setpoint margins due to changes in instrument accuracy and calibration errors caused by the change in environmental conditions around the instrument due to the TPO uprate are negligible. Maintaining constant nominal dome pressure for the TPO uprate minimizes the potential effect on these instruments by maintaining the same fluid properties at the instruments. The setpoint evaluations are based on the guidelines in TLTR Sections 5.8 and F.4 and on Section 5.3 of Reference 9.

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

5.3.2 Hydraulic Pressure Scram The AL for the turbine hydraulic pressure that initiates the T/G trip scram at high power remains the same as for the CLTP. No modifications are being made to the turbine hydraulic control 5-4

NEDO-.33385 REVISION 0 NON-PROPRIETARY INFORMATION systems for TPO; actuation of these safety functions remains unchanged from the current operation, 5.3.3 High-Pressure Recirculation Pump Trip The anticipated transient without scram recirculation pump trip (ATWS-RPT) trips the pumps during plant transients with increases in reactor vessel dome pressure. The ATWS-RPT provides negative reactivity by reducing core flow during the initial part of an ATWS. The evaluation in Section 9.3.1 demonstrates that the current high pressure ATWS-RPT AL is acceptable for the TPO uprate, 5.3.4 Safety Relief Valve Because there is no increase in reactor operating dome pressure, the SRV ALs are not changed.

5.3.5 Main Steam Line High Flow Isolation The Technical Specification AV of this function is expressed in terms of percent steam flow rate.

The corresponding differential pressure, and therefore absolute steam flow rate is not changed.

Therefore this AV is decreased for the TPO uprate. Although the MS flow increases by -2%,

the main steam line (.MSL) flowelement AL AP setpoint is not changed for the TPO uprate. The corresponding setpoint AV in terms of steam flow is decreased to approximately

• 141.6% (better estimate 142.1%) of the TPO rated steam flow at 101.7% CLTP

• 141.7% (better estimate 142.2%) of the TPO rated steam flow at 101.62% CLTP Because of the large spurious trip margin, sufficient margin to the trip setpoint exists to allow for normal plant testing of the MSIVs and turbine stop and control valves. This is consistent with TLTR Section F.4.2.5.

5.3.6 Fixed APRM Scram The fixed APRM ALs, for both two (recirculation) loop (TLO) and SLO, expressed in percent of RTP do not change for the TPO uprate. The generic evaluation and guidelines presented in TLTR Section F.4.2.2 are applicable to CNS* The limiting transient that relies on the fixed APRM trip is the MSIV closure transient with indirect scram. This event has been analyzed assuming 102% of CLTP and is reanalyzed on a cycle specific basis.

5.3.7 APRM Flow-Biased Scram The flow-referenced APRM ALs, for both TLO and SLO, are unchanged in units of absolute core thermal power versus recirculation drive flow. Because the setpoints are expressed in percent of RTP, they decrease in proportion to the power uprate or CLTP RTP / TPO RTP. This is the same approach taken for generic BWR uprates described in ELTRI (Reference 2). In

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION order to accommodate limits in the Stability Region, new APRM ALs were established that conservatively bound the entire operating envelope. There is no significant effect on the instrument errors or uncertainties from the TPO uprate. Therefore, the AV and NTSP are established by directly incorporating the change in the AL.

5.3.8 Rod Worth Minimizer Low Power Setpoint The Rod Worth Minimizer (RWM) Low Power Setpoint (LPSP) is used to enforce the rod patterns established for the control rod drop accident at low power levels. The generic guidelines in TLTR Section F.4.2.9 are applicable to CNS. The RWM LPSP AL is rescaled by the CLTP RTP / TPO RTP ratio to maintain the AL in terms of absolute power, 5.3.9 Rod Block Monitor The severity of the Rod Withdrawal Error (RWE) during power operation event is dependent upon the RBM rod block setpoint. The power-dependent ALs are maintained at the same percent power. The cycle specific reload analysis is used to determine any changes in the rod block setpoint.

5.3.10 Flow-Biased Rod Block Monitor (%RTP)

CNS does not have a flow-biased RBM system.

5.3.11 Main Steam Line High Radiation Isolation The MSL normal radiation level increases approximately proportionally to power. The setpoint is based on normal operating background radiation level, and will be adjusted to provide the same level of protection at the TPO uprate conditions with no appreciable increase in spurious trip frequency. No change in the Technical Specifications is required. This approach is consistent with TLTR Section F.4,2,8.

5.3.12 Low Steam Line Pressure MSIV Closure (RUN Mode)

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

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

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Water level change during operational transients (e.g., trip of a recirculation pump, FW controller failure, loss of one FW pump) is slightly affected by the TPO uprate. The plant response following the trip of one FW pump does not change significantly, because the maximum operating rod line is not being increased, Therefore, the final power level following a single FW pump trip at TPO uprate conditions would remain the same relative to the remaining FW flow as exists at CLTP.

5.3.14 Main Steam Line Tunnel High Temperature Isolations As noted in Section 5.2.4 above, the high steam tunnel temperature AL remains unchanged for the TPO uprate.

5.3.15 Low Condenser Vacuum In order to produce more electrical power, the amount of heat discharged to the main condenser increases slightly. This added heat load may slightly increase condenser backpressure, but the increase would be insignificant (less than 0.1 in. HgA). The slight change in condenser vacuum after implementation of TPO will not adversely affect any trip signals associated with low condenser vacuum (turbine trip / MSIV closure).

5.3.16 TSV Closure Scram, TCV Fast Closure Scram Bypasses The Turbine Stop Valve (TSV) closure scram and Turbine Control Valve (TCV) fast closure scram allow this scram to be bypassed, when reactor power is sufficiently low, such that the scram functions are not needed to mitigate a T/G trip. This power level is the AL for determining the actual trip setpoint, which comes from the turbine first-stage pressure (TFSP).

The TFSP setpoint is chosen to allow operational margin so that scrams can be avoided, by transferring steam to the turbine bypass system during T/G trips at low power.

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

3] The maneuvering range for plant startup is maximized.

No modifications to the CNS turbine are made for the TPO uprate, so there is no change in the first-stage pressure/steam flow relationship from previous operation.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 5-1 Analytical Limits that Change due to TPO APRM High Neutron Flux Scram 123.0 No change APRM Flow Biased Scram TLO Fixed (%RTP) not applicable not applicable SLO Fixed (%RTP) not applicable not applicable TLO Flow Biased (%RTP)(1)(2 0.66Wd + 74.8 0.75Wd + 65.6 1t SLO Flow Biased (%RTP)(' 0,66(Wd-AW) + 74.8 0.75(Wd-AW) + 65.6 APRM Flow Biased Rod Block TLO Fixed (%RTP) 112.2 No change SLO Fixed (%RTP) 112.2 No change 1

TLO Flow Biased (%RTP)( )(2) 0.66Wd + 64.0 0.75Wd + 54.8 1t SLO Flow Biased (%RTP)( ) ' 0.66(Wd-A\W) + 64.0 0.75(Wd-AW) + 54.8 TSV & TCV Scram & RPT Bypasses (%RTP) 30 29.5 MSL High Flow Isolation TPO = 101.7% CLTP:

% rted steam flow 147.5 (better estinmate 148) psid 150 121.24 121.24 TPO = 101.6% CLTP:

147.6 (belicr estimate 148.2) 121.24 Rod Worth Minimizer TPO = 101.7% CLTP:

LPSP (%RTP) 10 9.84 TPO =101.6% CLTP:

9.85 Notes:

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

(2) Wd is % recireuhation drive flow where 100% drive flow is that required to achieve 100% core flow at 100% power, and AW is the diflerence between the TLO and SLO drive flow at the same core flow. The current value of AW is 6.73% and is not changed.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 6.0 ELECTRICAL POWER AND AUXILIARY SYSTEMS 6.1 AC POWER Plant electrical characteristics are given in Table 6-1.

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

Generator: 983 MVA, 22 kV, 0.85 power factor, 0.58 short-circuit ratio, 60 psig hydrogen pressure (maximum), 1800 rpm three-phase, 60 Hertz Main Transformer: 900 MVA/1008 MVA, 3-300/336 MVA, FOA, 55/65'C, 345-20.9 kV, 60 Hz (3 normally in service - I spare)

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

" The isolated phase bus duct is adequate for both rated voltage and low voltage current output.

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

A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criteria 17 (10 CFR 50, Appendix A). GDC 17 addresses on-site and off-site electrical supply and distribution systems for safety-related components. There is no significant effect on grid stability or reliability. There are no modifications associated with the T.PO uprate, which would increase electrical loads beyond those levels previously included or revise the logic of the distribution systems.

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

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

Station loads under normal operation/distribution conditions are computed based on equipment nameplate data with conservative demand factors applied. The only identifiable change in 6-1

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION electrical load demand is associated with condensate and condensate booster pumps. These pumps experience increased flow and pressure due to the TPO uprate conditions. Because these changes are small, the motor demand for each of these loads remains bounded by the existing design. Accordingly, there are negligible changes in the on-site distribution system design basis loads or voltages due to the TPO conditions. The system environmental design bases are unchanged. Operation at the TPO RTP level is achieved by utilizing existing equipment operating at or below the nameplate rating; therefore, under normal conditions, the electrical supply and distribution components (e.g., switchgear, motor control centers (MCCs), cables) are adequate.

Station loads under emergency operation and distribution conditions (emergency diesel generators) are based on BHP or running KW. The ECCS pumps use a conservatively high flow BHP. Emergency operation at the TPO RTP level is achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the stated pumps; therefore, under emergency conditions the electrical supply and distribution components are adequate.

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

6.2 DC POWER The direct current (DC) loading requirements in the USAR were reviewed, and no reactor power-dependent loads were identified. The DC power distribution system provides control and motive power for various systems and components. In both normal and emergency operating scenarios, loads are computed based on equipment nameplate data or BH-P. These loads are used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP level is achieved in both normal and emergency conditions by operating equipment at or below the nameplate rating running kW or BIP. Additionally, operation at the TPO RTP level does not increase any loads or revise control logic. Therefore, there are no changes to the load, voltage drop or short circuit current values.

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

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 6.3.1 Fuel Pool Cooling The Spent Fuel Pool (SFP) heat load increases slightly as a result of operation at the TPO condition. The TPO uprate does not affect the heat removal capability of the Fuel Pool Cooling and Cleanup System (FPCCS) as shown in Table 6-2. The TPO heat load is within the design basis heat load for the FPCCS, and does not result in a delay in removing the RHR system from service (i.e., the outage day the FPC System can maintain the SFP temperature below 150'F such that the Fuel Pool Assist Mode of the RHR system is not required).

The SFP cooling adequacy is determined by calculating the heat load generated by a full core discharge plus remaining spaces filled with used fuel discharged at regular intervals. The analysis assumes 18-month fuel cycle lengths as the basis. The existing analyses and continuing compliance with the commitment to maintain the pool design limits (i.e., maximum temperature and corresponding heat removal capacity) by controlling the rate of the discharge (fuel offload) to the spent fuel pool confirm the capability of the FPC System to maintain adequate fuel pool cooling for the TPO uprate.

The FPC heat exchangers are sufficient to remove the decay heat during normal refueling and under full core off-load conditions following operation at the TPO uprate conditions. The RHR system in Fuel Pool Cooling Assist mode is available, if needed, to maintain the SFP water temperature below design limit 6.3.2 Crud Activity and Corrosion Products The crud activity and corrosion products associated with spent fuel can increase very slightly due to the TPO. The increase is insignificant and SFP water quality is maintained by the FPCCS.

6.3.3 Radiation Levels The normal radiation levels around the SFP may increase slightly during fuel handling operation.

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

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

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

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 6.4.1 Service Water Systems 6.4.1.1 Safety-Related Loads The safety-related Service Water (SW) system provides cooling water during and following a design basis accident. The safety-related performance of the SW system during and following the most demanding design basis event Loss of Coolant Accident (LOCA) does not change because the original LOCA analysis was based on 102% of CLTP (Section 4.3). Similarly, the containment response analysis in Section 4.1 is also based on 102% of CLTP. There is no change in the safety-related heat loads and the requirements are within the existing capacity of the R1HR and associated SW system.

6.4.1.2 Nonsafety-Related Loads The major service water heat load increases from the TPO reflect an increase in main generator losses rejected to the generator hydrogen coolers and the Turbine Equipment Cooling (TEC)

System. The thermal efficiency of the power generation cycle is not expected to change.

Therefore, the increase in service water heat loads from these sources due to the TPO uprate operation is approximately proportional to the TPO (-1.7%). The design of these systems is adequate to handle the TPO uprate.

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

6.4.2.1 Discharge Limits The CNS Nebraska Department of Environmental Quality National Pollutant Discharge Elimination System (NPDES) Permit provides the effluent limitations and monitoring requirements for discharge wastewater at the site. The discharge limits on free available chlorine are 0.01 mg/l for the monthly average and 0.02 mg/I for the maximum. The discharge water temperature shall not exceed a maximum of 109.4°F. Frequent monitoring of these parameters ensures that permit limits are not exceeded. The TPO uprate has minimal effect on the parameters, and no changes to NPDES permit requirements are needed.

The state thermal discharge limits, the current discharges, and bounding analysis discharges for the TPO uprate are shown in Table 6-3. This comparison demonstrates that the plant remains within the state discharge limits, during operation at TPO conditions.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 6.4.3 Component Cooling Water System The heat loads on the Reactor Equipment Cooling (REC) system do not increase significantly due to TPO because they depend on either reactor vessel water temperature or flow rates in the systems cooled by the REC, The change in reactor vessel water temperature is minimal and there is no change in the nominal reactor operating pressure. The REC system experiences a slight heat load increase in the Fuel Pool Coolers heat exchangers. However, the system has adequate design margin to remove the additional heat. Therefore, the REC system is acceptable for the TPO uprate.

6.4.4 Turbine Equipment Cooling System The power-dependent heat loads on the Turbine Equipment Cooling (TEC) system that are increased by the TPO, are those related to the operation of the generator hydrogen coolers, the bus duct cooler and exciter coolers. The remaining TEC heat loads are not strongly dependent upon reactor power and do not significantly increase. The TEC system has sufficient capacity to assure that adequate heat removal capability is available for TPO operation.

6.4.5 Ultimate Heat Sink The ultimate heat sink (UHS) for Cooper Nuclear Station is the Missouri River. The Service Water System provides the ultimate heat sink for equipment cooling throughout the plant. As a result of operation at the TPO RTP level, the post-LOCA heat load increases slightly, primarily due to higher reactor decay heat. However, the ability of the UH.S to perform required safety functions is demonstrated with previous analyses based on 102% of CLTP. Therefore, all safety aspects of the UHS are within previous evaluations and the requirements are unchanged for TPO uprate conditions. The current Technical Specifications for UHS limits are adequate due to conservatism in the current design.

6.5 STANDBY LIQUID CONTROL SYSTEM The SLCS is designed to shut down the reactor from rated power conditions to cold shutdown in the postulated situation that all or some of the control rods cannot be inserted. It is a manually operated system that pumps a highly enriched sodium pentaborate solution into the vessel to achieve a sub critical condition. The generic evaluation presented in TLTR Sections 5.6.5 (SLCS) and L.3 (ATWS Evaluation) is applicable to the CNS TPO uprate. The TPO uprate does not affect shutdown or injection capability of the SLCS. Because the shutdown margin is reload dependent, the shutdown margin and the required reactor boron concentration are confirmed for each reload core.

The SLCS relief valve margin is adequate for the TPO uprate because the SLCS system prior to the TPO uprate has a confirmed minimum relief valve margin of 70 psi (measured between the inlet to the SLCS relief valve and the minimum SLCS relief valve opening setpoint accounting for setpoint tolerance).

6-5

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION The SLCS ATWS performance is evaluated in TSAR Section9.3.1. The evaluation shows that the TPO has no adverse effect on the ability of the SLCS to mitigate an ATWS.

6.6 POWER DEPENDENT HEATING, VENTILATION AND AIR CONDITIONING The Heating Ventilation and Air Conditioning (HVAC) system that are potentially affected by the TPO uprate consist mainly of heating, cooling supply, exhaust, and recirculation units in the turbine building, reactor building (including steam tunnel and drywell), and control building.

TPO results in a minor increase in the heat load caused by the slightly higher FW process temperature (-2°F). The increased heat load is within the margin of the steam tunnel area coolers. In the drywell, the increase in heat load due to the FW process temperature is within the system capacity. In the turbine building, the maximum temperature increases are <20 F due to the increase in the FW process temperatures. In the reactor building, the increase in heat load due to a slight SFP cooling process temperature increase is within the margin of the area coolers. Other areas are unaffected by the TPO because the process temperatures and electrical heat loads remain constant.

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

6.7 FIRE PROTECTION Operation of the plant at the TPO RTP level does not affect the fire suppression or detection systems. There are no changes in physical plant configuration or combustible loading as a result of the TPO uprate. The safe shutdown systems and equipment used to achieve and maintain cold shutdown conditions do not change, and are adequate for the TPO uprate conditions. The operator actions required to mitigate the consequences of a fire are not affected. Therefore, the fire protection systems and analyses are not affected by the TPO uprate.

6.7.1 10 CFR 50 Appendix R Fire Event TLTR Section L.4 presents a generic evaluation of Appendix R events for an increase of 1.5% of CLTP. ((

)) The current analysis based on CLTP has an available margin of 9°F to the clad temperature limit and 45.2 psi to the containment pressure limit. The current PCT of 1491'F was based on a conservative decay heat table that contains less data points for the period when the PCT occurs.

The newly updated decay heat table with more data points (applicable for all BWRs) along with some SAFER model corrections would reduce the PCT by 50-150°F. Therefore, there is a significant margin (greater than 28'F as specified in TLTR) for Cooper TPO and the current .PCT of 1491'F remains applicable for Cooper TPO. Therefore, the generic results are clearly applicable and no further plant specific Appendix R analysis is necessary for the TPO uprate.

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NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION 6.8 SYSTEMS NOT AFFECTED BY TPO UPRATE Based on experience and previous NRC reviews, all systems that are significantly affected by TPO are addressed in this report. Other systems not addressed by this report are not significantly affected by TPO. The systems unaffected by TPO at CNS are confirmed to be consistent with the generic description provided in the TLTR.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Table 6-1 TPO Plant Electrical Characteristics

• +p;Parameter** e*+u+*

Guaranteed Generator Output (MWe) 830A4 Rated Voltage (kV) 22 Power Factor 0, 58 Guaranteed Generator Output (MVA) 983.0 Current Output (kA) 25.797 Isolated Phase Bus Duct Rating:

Main Section (kA) 28 Branch Section (kA) 1.2 Main Transformers Rating (MVA) 900 Table 6-2 Fuel Pool Cooling and Cleanup System Parameters Parameter Pre-TPO TPO System Component Design Temperature (OF) 150 150 System Component Design Pressure (psig) 150 150 Number of Fuel Pool Cooling Loops 2 2 Structural Design Temperature (CF) 150 150 Fuel Cycle (months) 18 18 Bulk Pool Temperature for a Full Core Off-load, Fuel Pool With < 150 < 150 Maximum Capacity, with Supplemental RHR Cooling (°F), if required.

Table 6-3 Effluent Discharge Comparison Parameter Discharge Temperature not to exceed (OF)

Chlorine mglL (Maximum TRC) a

(')-Monitoring for TRC is required only when chlorine is introduced into any waste-streams.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 7.0 POWER CONVERSION SYSTEMS 7.1 TURBINE-GENERATOR The CNS main T/G is designed with a maximum flow passing and generator capability in excess of rated conditions to ensure that the design rated output is achieved. The excess capacity ensures that the T/G can meet rated conditions for continuous operating capability with allowances for variations in flow coefficients from expected values, manufacturing tolerances, and other variables that may affect the flow-passing capability of the unit. The difference in the steam-passing capability between the current analyzed and rated conditions is called the flow margin.

The CNS turbine-generator has a flow margin of 10% at the rated throttle steam flow of 9,510,617 lb/hr at a throttle pressure of 970 psia and rated electrical power output of 828.971 MW at a generator capability of 983.000 MVA.

For the TPO uprate RTP of 2429 MWt (-102% of CLTP), the rated throttle steam flow is increased to 9,703,000 lb/hr at a throttle pressure of 968.0 psia. The evaluated increased throttle flow is approximately 102.0% of current rated. The evaluated increased throttle flow is due to the steam flow increase associated with operation at 102% CLTP conditions (-2%). The maximum uprated electrical output is 835.550 MW. Maximum expected reactive power at TLTP conditions is expected to be less than 300 MVAR. These conditions result in a maximum generator load (capability)of 887.770 MVA.

Steam specification calculations were performed to determine the TPO uprate turbine steam path conditions. These TPO uprate operating conditions are bounded by the previous analysis of the turbine and generator stationary and rotating components. Thus, the increased loadings, pressure drops, thrusts, stresses, overspeed capability and other design considerations resulting from operation at TLTP conditions are within existing design limits and operation therefore is acceptable at the TPO uprate condition. In addition, valves, control systems and other support systems were evaluated and TPO operating conditions are bounded by the existing analyses. The results of these evaluations show that no modifications are needed to support operation at the TPO uprate condition.

The existing rotor missile analysis was performed at 120% design overspeed conditions. The low-pressure turbine casing is designed to prevent rupture due.to disc failure at 120% design overspeed conditions. The TPO uprate does not change turbine rated speed. Therefore, there is no change in the missile generation probability (a missile does not escape from the turbine casing) and thus, the missile generation probability remains unchanged and is therefore acceptable.

The overspeed evaluation addressed the sensitivity of the rotor train for the capability of overspeeding. Although the entrapped energy increases slightly for the TPO uprate conditions, 7-1

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION no change in the overspeed trip settings is required because the existing analysis bounds the TPO uprate conditions.

7.2 CONDENSER AND STEAM JET AIR EJECTORS The main condenser capability was evaluated for performance at the TPO uprate conditions in section 6.4.2. The design margin in the condenser heat removal capability can accommodate the additional heat rejected for operation at the TPO uprate conditions. Air leakage into the condenser does not increase as a result of the TPO uprate. The small increase in hydrogen and oxygen flows from the reactor does not affect the SJAE capacity because the design was based on operation at greater than required flows at uprate conditions. Therefore, the condenser air removal system is not affected by the TPO uprate and the mechanical vacuum pumps and SJAEs are adequate for operation at the TPO uprate conditions.

7.3 TURBINE STEAM BYPASS The Steam Bypass Pressure Control System (SBPCS) was originally designed for a steam flow capacity of approximately 33% of the 100% rated flow at CLTP. The steam bypass capacity at the TPO RTP is approximately 32% of the 100% TPO RTP steam flow rate. The steam bypass system is non-safety-related. While the bypass capacity as a percent of rated steam flow is reduced, the actual steam bypass capacity is unchanged. The transient analyses that credit the turbine bypass system use a bypass capacity that is less than the actual capacity. Therefore, the turbine bypass capacity remains adequate for TPO. operation since the actual capacity (unchanged) continues to bound the value used in the analyses.

7.4 FEEDWATER AND CONDENSATE SYSTEMS The FW and condensate systems are designed to provide FW at the temperature, pressure, quality, and flow rate required by the reactor. These systems are not safety-related; however, their performance may have an effect on plant availability and the capability to operate reliably at the TPO uprate condition, A review of the CNS FW heaters, heater drains, condensate demineralizers, and the pumps (FW and condensate) demonstrated that the components are capable of performing in the proper design range to provide the slightly higher TPO uprate FW flow rate at the desired temperature and pressure. The review also concluded that the FW control valves and FW turbine controls can maintain water level control at the TPO uprate conditions. A review of the CNS heater drains demonstrated that the components are capable of supporting the slightly higher TPO uprate extraction flow rates. The plant will continue to operate with a reduction in cycle efficiency due to inadequate drain valve sizing on the No. 2 and No. 3 heater drain valves.

However, there is no adverse impact on plant operation or safety.

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

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION

" Pump NPSH

• Ability to avoid suction pressure trip

" Flow capacity

" Bearing cooling capability

" Rated motor horsepower

• Full load motor amps

" Vibration The FW system run-out and loss of FW heating events would see very small changes from the TPO uprate as shown by the experience with substantially larger power uprates.

7.4.1 Normal Operation System operating flows for the TPO uprate increase approximately 1.8%. The condensate and FW systems were originally designed for approximately 105% warranted steam flows.

Operation at the TPO RTP level does not significantly affect operating conditions of these systems. Discharge pressure of the condensate pumps decreases due to the pump head characteristics at increased flows. Discharge pressure of the FW pumps will increase to compensate for the increase in FW friction losses due to higher flow. To accomplish this function, opening the flow control valves to the feed pump turbine increases the feed pump speed, During steady-state conditions, the condensate and FW systems have available NPSH for all of the pumps to operate without cavitation at the TPO uprate conditions. Adequate margin during steady-state conditions exists between the calculated minimum pump suction pressure and the minimum pump suction pressure trip set points.

The existing FW design pressure and temperature requirements are adequate. The FW heaters and associated regulating valves were originally designed for the original guarantee (100%) heat balance conditions. The FW heaters are ASME Section VI1i pressure vessels. The heaters were analyzed and verified to be acceptable for the slightly higher FW heater temperatures and pressure for the TPO uprate. There is no need for recertification of the feedwater heaters.

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

Following a single FW pump trip, the reactor recirculation system would runback recirculation flow, such that the steam production rate is within the flow capacity of the remaining FW pumps.

The runback setting prevents a reactor low water level scram, and is sufficient to maintain 7-3

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION adequate margin to the potential power/flow instability regions. Operation at the TPO condition does not degrade this capability.

7.4.3 Condensate Demineralizers The effect of the TPO uprate on the condensate demineralizers (CDs) was reviewed. The CDs experience slightly higher loadings at the TPO RTP level which result in slightly reduced resin life. However, the reduced resin life is acceptable (refer to Section 8.0 for the effect on the radwaste systems). Because the system can accommodate (without bypass) TPO uprate operation with one vessel removed from service (when backwash/precoat is required), reduced resin life (more frequent backwasldprecoat) of the units does not adversely affect CD operation.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 8.0 RADWASTE AND RADIATION SOURCES 8.1 LIQUID AND SOLID WASTE MANAGEMENT The liquid radwaste system collects monitors, processes, stores, and returns processed radioactive waste to the plant for reuse, discharge, or shipment.

The single largest source of liquid and wet solid waste is from backwash of the condensate CD and pre-coat (replacement of CD resins). The TPO uprate results in a -2% increased flow rate through the CDs, resulting in a reduction in the average time between backwashes and pre-coating. The reduction of CD service time does not affect plant safety. The RWCU filter demineralizer may also require more frequent backwashes due to slightly higher levels of activation and fission products.

The floor drain collector subsystem and the waste collector subsystem both receive periodic inputs from a variety of sources. Neither subsystem experiences a significant increase in volume due to operation at the TPO uprate condition.

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

8.2 GASEOUS WASTE MANAGEMENT The gaseous waste systems collect, control, process, and dispose of gaseous radioactive waste generated during normal operation and abnormal operational occurrences. The gaseous waste management systems include the offgas system and various building ventilation systems. The systems are designed to meet the requirements of 10 CFR 20 and 10 CFR 50, Appendix 1.

Non-condensable radioactive gas from the main condenser normally contains activation gases (principally N-16, 0-19 and N-13) and fission product radioactive noble gas parents. This is the major source of radioactive gas, which is greater than all other sources combined. These non-condensable gases, along with non-radioactive air in leakage, are continuously removed from the main condensers by the steam jet air ejectors (SJAE) that discharge into the offgas system.

Building ventilation systems control airborne radioactive gases by using devices such as HEPA and charcoal filters, and radiation monitors that activate isolation dampers or trip supply and exhaust fans, or by maintaining negative or positive air pressure to limit migration of gases. The 8-1

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION activity of airborne effluents released through building vents does not increase significantly due to the TPO uprate because:

• The amount of fission products released into the coolant depends on the number and nature of the fuel rod defects and is not dependent on reactor power; and

" The concentration of coolant activation products remains unchanged because the increase in production of these products is offset by the increase in the steaming rate.

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

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

The off gas system was evaluated for the TPO uprate, including the effects of hydrogen water chemistry (HWC) and noble metal injection. Radiolysis of water in the core region, which forms H2 and 02, increases linearly with core power, thus increasing the heat load on the recombiner and related components. The Offgas system design basis H2 is 3 1.8 lbs/hr (with a corresponding stoichiometric 02 of 254 lbs/hr). The expected H2 flow rate for the TPO uprate is 13.27 lbs/hr (106.20 lbs/hr of 02). The increase in H2 and 02 due to the TPO uprate remains well with the capacity of the system. The system radiological release rate is administratively controlled, and is not changed with operation power. Therefore, the TPO uprate does not affect the offgas system design or operation.

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

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

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

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

The second set of post-operation source data consists primarily of nuclide activity inventories for fission products in the fuel. These are needed for post-accident and spent fuel pool evaluations, which are performed in compliance with regulatory guidance that applies different release and 8-2

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION transport assumptions to different fission products. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops "equilibrium" activities in the fuel (typically three years). Most radiologically significant fission products reach equilibrium within a 60-day period. The calculated inventories are approximately proportional to core thermal power. Consequently, for TPO, the inventories of those radionuclides, which reached or approached equilibrium, are expected to increase in proportion to the thermal power increase. The inventories of the very long-lived radionuclides, which did not approach equilibrium, are both power and exposure dependent. They are expected to increase proportionally with power if the fuel irradiation time remains within the current basis.

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

))

8.4 RADIATION SOURCES IN REACTOR COOLANT 8.4.1 Coolant Activation Products During reactor operation, the coolant passing through the core region becomes radioactive as a result of nuclear reactions. The coolant activation is the dominant source in the turbine building and in the lower regions of the drywell. Because these sources are produced by interactions in the core region, their rates of production are proportional to power. However, the concentration in the steam remains nearly constant, because the increase in activation production is balanced by the increase in steam flow. As a result, the activation products, observed in the reactor water and steam, increase in approximate proportion to the increase in thermal power. The activation products in the steam and coolant are bounded by the existing design basis concentration.

8.4.2 Activated Corrosion Products The reactor coolant.contains activated corrosion products from metallic materials entering the water and being activated in the reactor region. Under the TPO uprate conditions, the FW flow 8-3

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION increases with power, the activation rate in the reactor region increases with power, and the filter efficiency of the condensate demineralizers may decrease as a result of the FW flow increase.

The net result may be an increase in the activated corrosion product production. However, the TPO uprate corrosion product concentrations are not expected to exceed the design basis concentrations. Therefore, no change is required in the design basis activated corrosion product concentrations for the TPO uprate.

8.4.3 Fission Products Fission products in the reactor coolant are separable into the products in the steam and the products in the reactor water. The activity in the steam consists of noble gases released from the core plus carryover activity from the reactor water. The noble gases released during plant operation result from the escape of minute fractions of the fission products from the fuel rods.

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

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

8.5 RADIATION LEVELS Normal operation radiation levels increase slightly for the TPO uprate. CNS was designed with substantial conservatism for higher-than-expected radiation sources. Thus, the increase in radiation levels does not affect radiation zoning or shielding in the various areas of the plant because it is offset by conservatism in the design, source terms, and analytical techniques.

Post-operation radiation levels in most areas of the plant increase by no more than the percentage increase in power level. In a few areas near the SFP cooling system piping and the reactor water piping, where accumulation of corrosion product crud is expected, as well as near some liquid radwaste equipment, the increase could be slightly higher. Regardless, individual worker exposures will be maintained within acceptable limits by the site As Low As is Reasonably Achievable (ALARA) program, which controls access to radiation areas. Procedural controls compensate for increased radiation levels.

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

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

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION accident occupancy concluded that access needed for accident mitigation is not significantly affected by the TPO uprate.

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

8.6 NORMAL OPERATION OFF-SITE DOSES The Technical Specification limits implement the guidelines of 10 CFR 50, Appendix 1. A review of the normal radiological effluent doses shows that at CLTP, the annual doses are less than 1% of the doses allowed by Technical Specification limits with the exception of the Site Boundary for maximum organ dose due to 1-131, 1-133, and particulates. The largest reported value during 2004 and 2005 for 1-131, 1-133 and particulates was 2.18% dose to bone (child) at the site boundary. The TPO uprate does not involve significant increases in the offsite dose from noble gases, airborne particulates, iodine, tritium or liquid effluents. In addition, radiation from shine is not a significant exposure pathway. Present offsite radiation levels are negligible portion of background radiation. Therefore, the normal offsite doses are not significantly affected by operation at the TPO RTP level and remain below the limits of 10 CFR 20 and 10 CER 50, Appendix 1.

8-5

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 9.0 REACTOR SAFETY PERFORMANCE EVALUATIONS 9.1 ANTICIPATED OPERATIONAL OCCURRENCES TLTR Appendix E provides a generic evaluation of the ADOs for TPO uprate plants. ((

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

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

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

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

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

The evaluations and conclusions of TLTR Appendix E are applicable to the CNS TPO uprate.

Therefore, it is sufficient for the plant to perform the standard reload analyses at the first fuel cycle that implement the TPO uprate.

9.1.1 Alternate Shutdown Cooling Evaluation Alternate Shutdown Cooling Mode is not part of the plant licensing basis.

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

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

With the exception of the Main Steam Line Break Accident, radiological consequences due to postulated DBA events have been evaluated and analyzed to show that NRC regulations are met for 2% above the CLTP. The MSLBA was evaluated using the Technical Specification limit on reactor coolant activity. The limit on reactor coolant activity is unchanged for the TPO uprate 9-1

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION condition. Therefore, the radiological consequences associated with a postulated DBA firom TPO uprate conditions are bounded by these analyses. The evaluation/analysis was based on the methodology, assumptions, and analytical techniques described in the RGs, the Standard Review Plan (SRP) (where applicable), and in previous Safety Evaluations (SEs).

9.3 SPECIAL EVENTS 9.3.1 Anticipated Transient Without Scram CNS meets the following ATWS mitigation equipment requirements defined in 10 CFR 50.62:

1. Installation of an Alternate Rod Insertion (ARD) system;

2. Boron injection equivalent to 86 gpm; and

3. InstallatiOn of automatic RPT logic (i.e., ATWS-RPT).

There are no changes in the equipment for the TPO uprate. The performance characteristics of the equipment do not change because operating conditions (operating pressure, SRV setpoints, and maximum rod line) do not change.

The CNS-specific analysis at the CLTP demonstrates that the following ATWS acceptance criteria are met:

I. Peak vessel bottom pressure less than ASME Service Level C limit of 1500 psig;

2. Peak clad temperature within the 10 CFR 50.46 limit of 2200'F;

3. Peak clad oxidation within the requirements of 10 CFR 50.46;

4. Peak suppression pool temperature less than 197'F; and

5. Peak containment pressure less than 56 psig.

TLTR Section 5.3.5, TLTR Appendix L, present a generic evaluation of the sensitivity of an ATWS to a change in power typical of the TPO uprate. The evaluation is based on previous analyses for power uprate projects. For a TPO uprate, if a plant has sufficient margin for the projected changes in peak parameters given in TLTR Section L.3.5, ((

)) The previous ATWS analysis, performed at 100% of CLTP, demonstrated a margin of 193 psi to the peak vessel bottom head pressure limit and a margin of 160F to the pool temperature limit. ((

9-2

NEDO-338385 REVISION 0 NON-PROPRIETARY INFORMATION 9.3.2 Station Blackout TLTR Appendix L provides a generic evaluation of a potential loss of all alternating current power supplies based on previous plant response and coping capability analyses for typical power uprate projects. The previous power uprate evaluations have been performed according to the applicable bases for the plant (e.g., the bases, methods, and assumptions of RG 1 155 and/or NUMARC 87-00). This evaluation is for confirmation of continued compliance to 10 CFR 50.63. It is recognized that this evaluation is dependent upon many plant-specific design and equipment parameters.

Specifically, the following main considerations were evaluated:

" The adequacy of the condensate/reactor coolant inventory.

" The capacity of the Class IE batteries.

" The SBO compressed Nitrogen requirements.

" The ability to maintain containment integrity.

• The effect of loss of ventilation on rooms that contain equipment essential for plant response to a SBO event Applicable operator actions have previously been assumed consistent with the plant Emergency Procedure Guidelines. These are the currently accepted procedures for each plant and SBO analysis. For the TPO uprate, there is no significant change in the time available for the operator to perform these assumed actions.

['

)) CNS cunrently has margins of 31,658 gallons to the available condensate storage inventory volume and 139 1Fto the containment peak temperature limit. ((

)) Therefore, no CNS-specific SBO analysis is performed for the TPO uprate.

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NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 10.0 OTHER EVALUATIONS 10.1 HIGH ENERGY LINE BREAK Because the TPO uprate system operating temperatures and pressures change only slightly, there is no significant change in High Energy Line Break (HELB) mass and energy releases. The FW lines, near the pump discharge, increase < 2°F and < 5 psi. The recirculation lines increase < I° and < 1 psi due to the slightly higher core pressure drop. Vessel dome pressure and other portions of the RCPB remain at current operating pressure or lower. Therefore, the consequences of any postulated HELB would not significantly change. The postulated break locations remain the same because the piping configuration does not change due to the TPO uprate.

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

10.1.1 Steam Line Breaks The critical parameter affecting the high-energy steam line break analysis is the reactor vessel dome pressure. Because there is no pressure increase for the TPO, the MSL pressure decreases and there is a slight decrease in the main steam line break (MSLB) blowdown rate. The MSLB is used with a concurrent FW line break to: establish the peak pressure and the temperature environment in the MS tunnel, The TPO uprate increases the FW temperature <2'F and pressure < 5 psi, which results in an insignificant increase in the FW mass and energy release.

Design margins within the HELB analysis for a MSLB with a-concurrent FW line break provide adequate margin to the limits in the steam tunnel.

10.1.2 Liquid Line Breaks 10.1.2.1 Feedwater Line Breaks FW line breaks are assumed concurrently with an MSLB, as discussed in Section 10.1.1.

10.1.2.2 ECCS Line Breaks HPCI: Steam line breaks in the HPCI pump/turbine room and the MS tunnel are the limiting breaks for structural design and equipment qualification. Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase.

Therefore, the previous HELB analysis is bounding for the TPO uprate condition.

10-1

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION The HPCI and other ECCS lines are normally isolated from the reactor vessel, and a failure of one of these lines would result in a non-limiting break inside drywell, which would be bounded by other line breaks. Because these lines are normally isolated, the TPO uprate does not affect their line break analyses, for breaks outside drywell, 10.1.2.3 RCIC System Line Breaks Steam line breaks in the RCIC pump/turbine room are the limiting breaks for structural design and equipment qualification. Because there is no increase in the reactor dome pressure relative to the original analysis, the mass flow rate does not increase. Therefore, the previous HELB analysis is bounding for the TPO uprate conditions.

10.1.2.4 RWCU System Line Breaks The RWCU system line breaks are the limiting breaks for structural design and equipment qualification in several areas of the plant. As a result of the small increase in recirculation temperature with no pressure increase, the blowdown rate decreases slightly and the energy increases slightly. The original analysis was generally performed with conservative model assumptions. These conservatisms more than offset the effects of the temperature change, so the original HELB analysis is bounding.

10.1.2.5 CRD System Line Breaks The CRD pipe rupture analysis is not affected by the TPO uprate.

1.0.1.2.6 Building Heating Line Breaks Building heating lines are not connected to the reactor-turbine primary loop. Therefore, building heating lines are not affected.

10.1.2.7 Pipe Whip and Jet Impingement Because there is no change in the nominal vessel dome pressure, pipe whip and jet impingement loads do not significantly change. Existing calculations supporting the dispositions of potential targets of pipe whip and jet impingement from postulated HELBs have been reviewed and determined to be adequate for the safe shutdown effects in the TPO RTP conditions. Existing pipe whip restraints, jet impingement shields, and their supporting structures are also adequate for the TPO uprate conditions.

10.1.2.8 Internal Flooding from HELB None of the plant flooding zones contains a potential HELB location affected by the reactor operating conditions changed for the TPO uprate. The systems containing potential HELBs in the plant flooding areas are the CS, RHR, SLC, HPCI, RCIC, CRD, MS, FW, Sampling, RWCU, SW, Bleed Steam and Heating Steam. The systems' operational modes evaluated for HELB are 10-2

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION not affected by the TPO uprate, nor are the plant internal flooding analysis or safe shutdown analysis.

10.2 MODERATE ENERGY LINE BREAK None of the plant flooding zones contains a potential Moderate Energy Line Break (MELB) location affected by the reactor operating conditions changed for the TPO uprate. The following systems contain potential MELB locations in plant flooding zones: Standby Liquid Control, Reactor Core Isolation Cooling, Residual Heat Removal, High Pressure Coolant Injection, Core Spray, Liquid Radwaste, Fuel Pool Cooling & Demineralizer, Reactor Equipment Cooling, Turbine Equipment Cooling, Service Water and Service Water Booster, Fire Protection, Makeup Water Treatment, Circulating Water, Condensate DI Units, Condensate - Feedwater, and Drainage.

No system operational modes evaluated for MELB are affected by the TPO uprate. Therefore, the plant internal flooding analysis and safe shutdown analysis are not affected.

10.3 ENVIRONMENTAL QUALIFICATION Safety-related components must be qualified for the environment in which they operate. The TPO increase in power level increases the radiation levels experienced by equipment during normal operation and accident conditions. Because the TPO uprate does not increase the nominal vessel dome pressure, there is a very small effect on pressure and temperature conditions experienced by equipment during normal operation and accident conditions. The resulting environmental conditions are bounded by the existing environmental parameters specified for use in the environmental qualification program.

.10.3.1 Electrical Equipment The safety-related electrical equipment was reviewed to ensure that the existing qualification for the normal and accident conditions expected in the area where the devices are located remain adequate: Conservatisms in, the equipment qualifications were originally applied to the environmental parameters, and no change is needed for the TPO uprate.

10.3.1.1 Inside Containment Environmental qualification (EQ) for safety-related electrical equipment located inside the containment is based on MSLBA 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. The current accident conditions for temperature and pressure are based on analyses initiated from > 102% of CLTP. Normal temperatures may increase slightly near the FW and reactor recirculation lines and will be evaluated through the EQ'temperature monitoring'program, which tracks such information for equipment aging considerations. The

-current radiation levels under normal plant conditions also increase slightly. The current plant 10-3

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION environmental envelope for radiation is not exceeded by the changes resulting from the TPO uprate.

10.3.1.2 Outside Containment Accident temperature, pressure, and humidity environments used for qualification of equipment outside containment result from an MSLB in the pipe tunnel, or other HELBs, whichever is limiting for each area. The 14ELB pressure and temperature profiles bound the TPO uprate conditions. There is adequate margin in the qualification envelopes to accommodate the small changes due to TPO conditions. Maximum accident radiation levels used for qualification of equipment outside containment are from a DBA-LOCA.

10.3.2 Mechanical Equipment With Non-Metallic Components Operation at the TPO RTP level increases the normal process temperature very slightly in the FW and reactor recirculation piping. The slight increase in normal and accident radiation was evaluated in Section 10.3. Evaluation of the safety-related mechanical equipment with non-metallic components for temperature and radiation is not part of CNS environmental qualification program licensing basis.

10.4 TESTING The TPO uprate power ascension is based on the guidelines in TLTR Section L.2. Pre-operational tests are not needed because there are no significant changes to any plant systems or components that require such testing.

In preparation for operation at TPO uprate conditions, routine measurements of reactor and system pressures, flows, and selected major rotating equipment vibration are taken near 95% and 100% of CLTP, and at 100% of TPO RTP. The measurements will be taken along the same rod patternine used for the increase to TPO RTP. Core power from the APRMs is re-scaled to the TPO RTP before exceeding the CLTP and any necessary adjustments will be made to the APRM alarm and trip settings.

The turbine pressure controller setpoint will be readjusted at < 95% of CLTP and held constant.

The setpoint is reduced so the reactor dome pressure is the same at TPO RTP as for the CLTP.

Adjustment of the pressure setpoint before taking the baseline power ascension data establishes a consistent basis for measuring the performance of the reactor and the turbine control valves.

Demonstration of acceptable fuel thermal margin will be performed prior to and during power ascension to the TPO RTP at each steady-state heat balance point defined above. Fuel thermal margin will be projected to the TPO RTP point after the measurements taken at 95% and 100%

of CLTP to show the estimated margin. The thermal margin will be confirmed by the measurements taken at full TPO RTP conditions. The demonstration of core and fuel conditions will be performed with the methods currently used at CNS.

10-4

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION Performance of the pressure and FW/level control systems will be recorded at each steady-state point defined above, The checks will utilize the methods and criteria described in the original startup testing of these systems to demonstrate acceptable operational capability, Water level changes of -3 inches and pressure setpoint step changes of +/-3 psi will be used, If necessary, adjustments will be made to the controllers and actuator elements.

The increase in power for the TPO uprate is sufficiently small that large transient tests are not necessary. High power testing performed during initial startup demonstrated the adequacy of the safety and protection systems for such large transients. Operational occurrences have shown the unit response is clearly bounded by the safety analyses for these events. ((

10.5 OPERATOR TRAINING AND HUMAN FACTORS No additional training (apart from normal training for plant changes) is required to operate the plant in the TPO uprate condition. For TPO uprate conditions, operator response to transient, accident, and special events is not affected. Operator actions for maintaining safe shutdown, core cooling, containment cooling, etc., do not change for the TPO uprate. Minor changes to the power/flow map, flow-referenced setpoint, and the like, will be communicated through normal operator training. Simulator changes and validation for the TPO uprate will be performed in accordance with established CNS plant certification testing procedures.

10.6 PLANT LIFE Two degradation mechanisms may be influenced by the TPO uprate: (1) Irradiation Assisted Stress Corrosion Cracking (IASCC) and (2) FAC. The increase in irradiation of the core internal components influences IASCC. The increases in steam and FW flow rate influence FAC.

However, the sensitivity to the TPO uprate is small -and various programs are -currently implemented to -monitor the aging of plant components, including .EQ, FAC, and In-service Inspection. EQ is addressed in Section 10.3, and FAC is addressed in Section 3.5. These programs address the degradation mechanisms and do not change for the TPO uprate. Plant Specific: The core internals see a slight increase in fluence, but the inspection strategy used at CNS, based on the Boiling Water Reactor Vessel and Internals Project (BWRVIP), is sufficient to address the increase. The Maintenance Rule also provides oversight for the other mechanical and electrical components, important to plant safety, to guard against age-relaied degradation.

The longevity of most equipment is not affected by the TPO uprate because there is no significant change in the operating conditions. No additional maintenance, inspection, testing, or surveillance procedures are required.

10-5

NEDO-33385 REVISION 0 NON-PROPRIETARY INFORMATION 10.7 NRC AND INDUSTRY COMMUNICATIONS NRC and Industry communications are generically addressed in the TLTR, Section 10.8, Per the TLTR, it is not necessary to review prior dispositions of NRC and industry communications and no additional information is required in this area.

10.8 PLANT PROCEDURES AND PROGRAMS Plant procedures and programs are in place to:

1. Monitor and maintain instrument calibration during normal plant operation to assure that instrument uncertainty is not greater than the uncertainty used to justify the TPO uprate;

2. Control the software and hardware configuration of the associated instrumentation;

3. Peiform corrective actions, where required, to maintain instrument uncertainty within litnits;

4. Report deficiencies of the associated instruments to the manufacturer; and

5. Receive and resolve the manufacturer's deficiency reports.

10.9 EMERGENCY OPERATING PROCEDURES The Emergency Operating Procedures (EOP) action thresholds are plant unique and will be addressed using standard procedure updating processes. It is expected that the TPO uprate will have a negligible or no effect on the operator action thresholds and to the EOPs in general.

10.10 INDIVIDUAL PLANT EXAMINATION (IPE)

The IPE (PRA) will not be updated, because the change in plant risk from the subject power uprate is insignificant. This conclusion is supported by NRC Regulatory Issue Summary (RIS) 2002-03. In response to feedback received during the public workshop held on August 23, 2001, the Staff wrote, "The NRC has generically determined that measurement uncertainty recapture power uprates have an insignificant impact on plant risk. Therefore,- no risk information is requested to support such applications."

10-6

NEDO-33 385 REVISION 0 NON-PROPRIETARY INFORMATION

11.0 REFERENCES

1. GE Nuclear Energy, "Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization," Licensing Topical Report, NEDC-32938P-A, Revision 2, Class I1 (Proprietary), May 2003.

2. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," (ELTR 1), Licensing Topical Reports NEDC-32424P-A, Class 11I (Proprietary), February 1999, and NEDO-32424, Class I (Non-proprietary), April 1995.

3. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2), Licensing Topical Reports NEDC-32523P-A, Class HI (Proprietary), February 2000, NEDC-32523P-A, Supplement I Volume I, February 1999; and Supplement I Volume 11, April 1999.

4. NRC Regulatory Issue Summary 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," dated January 31, 2002.

5. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis

,Methodology for Reload Applications," August 1996.

6. "Radiation Embrittlement of Reactor Vessel Materials," Regulatory Guide 1.99, Revision 2, May 1988.

7. NEDE-23785P-A, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement I Additional Information for Upper Bound PCT Calculation", Supplement 1, Revision 1, March 2002.

8. GE-NE-0000-0037-8293-R0, "NPPD Cooper Nuclear Station SAFER/GESTR Loss-of-Coolant Accident ECCS Analysis with Elimination of 1600 F Upper Bound PCT Limit",

September 2005.

9. GE Nuclear Energy, "Constant Pressure Power Uprate," Licensing Topical Report NEDC-33004P-A, Revision 4, Class .111 (Proprietary), June 2003.

10. GE Nuclear Energy, "General Electric Instrument Setpoint Methodology," NEDC-31336P-A, Class 3 (Proprietary), September 1996.

11-1

NLS2007069 ENCLOSURE 6 AFFIDAVITS OF WITHHOLDING PURSUANT TO 10 CFR 2.390 CAMERON INTERNATIONAL CORPORATION APPENDIX K MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE COOPER NUCLEAR STATION DOCKET NO. 50-298, DPR-46

Measurement Systems Caldon Ultrasonics Technology Center 1000 McClaren Woods Drive Coraopolis, PA 15108 Tel 724-273-9300 Fax 724-273-9301 CAMERON www.c-a-m.com September 19, 2007 CAW 07-18 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Caldon Ultrasonics Engineering Report: ER-592 Rev.2 "Bounding Uncertainty Analysis for Thermal Power Determination at Cooper NPPD Using the LEFM/ + System" Gentlemen:

This application for withholding is submitted by Cameron International Corporation, a Delaware Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, pursuant to the provisions of paragraph (b)(1) of Section 2.390 of the Commission's regulations. It contains trade secrets and/or comnercial information proprietary to Cameron and customarily held in confidence.

The proprietary information for which withholding is being requested is identified in the subject submittal. In conformance with 10CFR Section 2.390, Affidavit CAW 07-18 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure.

Accordingly, it is respectfully requested that the subject information, which is proprietary to Cameron, be withheld from public disclosure in accordance with I OCFR Section 2.390 of the Commission's regulations.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference CAW 07-18 and should be addressed to the undersigned.

Very truly yours, Calvin R. Hastings General Manager Enclosures (Only upon separation of the enclosed confidential material should this letter and affidavit be released.)

September 18, 2007 CAW 07-19 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared Calvin R. Hastings, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Cameron International Corporation, a Delaware Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

Calvin R. Hastings il General Manager Sworn to and subscribed before me this / -___day of

___________ ,2007 tary PubliAý COMMONWEALTH OF PENNSYLVANIA Notarial Seal Joann B. Thomas, Notary Public Findlay Twp., Alegheny County My Commission Expires July 28,2011 Member, Pennsylvania Association of Notaries

1. I am the General Manager of Caldon Ultrasonics Technology Center, and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of Cameron.

2. I am making this Affidavit in conformance with the provisions of 10CFR Section 2.390 of the Commission's regulations and in conjunction with the Cameron application for withholding accompanying this Affidavit.

3. I have personal knowledge of the criteria and procedures utilized by Cameron in designating information as a trade secret, privileged or as confidential commercial or financial information. The material and information provided herewith is so designated by Cameron, in accordance with those criteria and procedures, for the reasons set forth below.

4. Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Cameron.

(ii) The information is of a type customarily held in confidence by Cameron and not customarily disclosed to the public. Cameron has a rational basis for determining the types of information customarily held in confidence by it and, in that connection utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Cameron policy and provides the rational basis required. Furthermore, the information is submitted voluntarily and need not rely on the evaluation of any rational basis.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Cameron's competitors without license from Cameron constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, and assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Cameron, its customer or suppliers.

(e) It reveals aspects of past, present or future Cameron or customer funded development plans and programs of potential customer value to Cameron.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Cameron system, which include the following:

(a) The use of such information by Cameron gives Cameron a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Cameron competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Cameron ability to sell products or services involving the use of the information.

(c) Use by our competitor would put Cameron at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Cameron of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Cameron in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Cameron capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence, and, under the provisions of IOCFR Section 2. 390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld is the submittal titled Caldon Ultrasonics Engineering Report: ER-592 Rev. 2 "Bounding Uncertainty Analysis for Thermal Power Determination at Cooper NPPD Using the LEFM V + System" and is designated therein in accordance with 10CFR §§ 2.390(b)(1)(i)(A,B), with the reason(s) for confidential treatment noted in the submittal and further described in this affidavit. This information is voluntarily submitted for use by the NRC Staff in their review of the accuracy assessment of the proposed methodology for LEFM CheckPlus Systems used by Cooper NPPD for an MUR UPRATE.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Cameron because it would enhance the ability of competitors to provide similar flow and temperature measurement systems and licensing defense services for commercial power reactors without coimnensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Cameron effort and the expenditure of a considerable sum of money.

In order for competitors of Cameron to duplicate this information, similar products would have to be developed, similar technical programs would have to be performed, and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods.

Further the deponent sayeth not.

Measurement Systems Caldon Ultrasonics Technology Center 1000 McClaren Woods Drive Coraopolis, PA 15108 Tel 724-273-9300 Fax 724-273-9301 CAMERON .c-m.com September 19, 2007 CAW 07-19 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Caldon Ultrasonics Engineering Report: ER-614 Rev. 1 "LEFM,/+ Meter Factor Calculation and Accuracy Assessment for Cooper NPPD" Gentlemen:

This application for withholding is submitted by Cameron International Corporation, a Delaware Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, pursuant to the provisions of paragraph (b)(1) of Section 2.390 of the Commission's regulations. It contains trade secrets and/or commercial information proprietary to Cameron and customarily held in confidence.

The proprietary information for which withholding is being requested is identified in the subject submittal. In conformance with I OCFR Section 2.390, Affidavit CAW 07-19 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure.

Accordingly, it is respectfully requested that the subject information, which is proprietary to Cameron, be withheld from public disclosure in accordance with 10CFR Section 2.390 of the Cormnission's regulations.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference CAW 07-19 and should be addressed to the undersigned.

Very truly yours, Calvin R. Hastings General Manager Enclosures (Only upon separation of the enclosed confidential material should this letter and affidavit be released.)

September 19, 2007 CAW 07-19 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared Calvin R. Hastings, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Cameron International Corporation, a Delaware Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

Calvin R. Hastinge General Manager Sworn to and subscribed before me this ______day of Z 2007 NotWSeal Joann B.Th"ma, Notaty Putbic FW y Twp.. AJ "eny CowY My Commissýo Expres Juiy 28,2011 Member, Pennsylvania Association of Notaries

1. I am the General Manager of Caldon Ultrasonics Technology Center, and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of Cameron.

2. I am making this Affidavit in conformance with the provisions of 10CFR Section 2.390 of the Commission's regulations and in conjunction with the Cameron application for withholding accompanying this Affidavit.

3. I have personal knowledge of the criteria and procedures utilized by Cameron in designating information as a trade secret, privileged or as confidential commercial or financial information. The material and information provided herewith is so designated by Cameron, in accordance with those criteria and procedures, for the reasons set forth below.

4. Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Cameron.

(ii) The information is of a type customarily held in confidence by Cameron and not customarily disclosed to the public. Cameron has a rational basis for determining the types of information customarily held in confidence by it and, in that connection utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Cameron policy and provides the rational basis required. Furthermore, the information is submitted voluntarily and need not rely on the evaluation of any rational basis.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Cameron's competitors without license from Cameron constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, and assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Cameron, its customer or suppliers.

(e) It reveals aspects of past, present or future Cameron or customer funded development plans and programs of potential customer value to Cameron.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Cameron system, which include the following:

(a) The use of such information by Cameron gives Cameron a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Cameron competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Cameron ability to sell products or services involving the use of the information.

(c) Use by our competitor would put Cameron at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Cameron of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Cameron in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Cameron capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence, and, under the provisions of I OCFR Section 2. 390, it is to be received in confidence by the Commission.

(iv) The inform-ation sought to be protected is not available in public sources or available information has not been previously employed in the same manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld is the submittal titled Caldon Ultrasonics Engineering Report: ER-614 Rev. 1 "LEFM ,/+ Meter Factor Calculation and Accuracy Assessment for Cooper NPPD" and is designated therein in accordance with 10CFR §§ 2.390(b)(1)(i)(A,B), with the reason(s) for confidential treatment noted in the submittal and further described in this affidavit. This information is voluntarily submitted for use by the NRC Staff in their review of the accuracy assessment of the proposed methodology for LEFM CheckPlus Systems used by Cooper NPPD for an MUR UPRATE.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Cameron because it would enhance the ability of competitors to provide similar flow and temperature measurement systems and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Cameron effort and the expenditure of a considerable sum of money.

In order for competitors of Cameron to duplicate this information, similar products would have to be developed, similar technical programs would have to be performed, and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods.

Further the deponent sayeth not.