ML041840116

Proposed Technical Specifications Change TS-431, Extended Power Uprate UFSAR Review Matrix, Enclosure 6

ML041840116
Person / Time
Site:	Browns Ferry
Issue date:	06/30/2004
From:	Schaefer K Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
References
DRF 0000-0010-9439, TVA-BFN-TS-431 NEDO-33101
Download: ML041840116 (183)

Text

ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS - 431 -

REQUEST FOR LICENSE AMENDMENT FOR EXTENDED POWER UPRATE OPERATION BROWNS FERRY EXTENDED POWER UPRATE UFSAR REVIEW MATRIX This enclosure provides a matrix identifying sections in the UFSAR that are currently under evaluation for change for EPU implementation. TVA will complete the final UFSAR changes following approval of this change.

Browns Ferry Extended Power Uprate UFSAR Review Matrix

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ENCLOSURE 14 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT I PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS - 431 -

REQUEST FOR LICENSE AMENDMENT FOR EXTENDED POWER UPRATE OPERATION NON-PROPRIETARY VERSION OF NEDC-33101 P BROWNS FERRY UNIT 1 SAFETY ANALYSIS REPORT FOR EXTENDED POWER UPRATE (See Attached)

GE Nuclear Energy NEDO-33 101 Class I DRF 0000-0010-9439 June 2004 BROWNS FERRY UNIT 1 SAFETY ANALYSIS REPORT FOR EXTENDED POWER UPRATE K. T. Schaefer, et. al.

GE Nuclear Energy 175 CurtnerAve., San Jose, CA 95125 NEDO-33101 Class I DRF 0000-0010-9439 June 2004 Browns Ferry Unit 1 Safety Analysis Report For Extended Power Uprate Prepared by: K. T. Schaefer, et. al.

A-~'7I/a Approved by I:

L. W. King, Project Manager GE Nuclear Energy Approved by 1 alentc, Unit I Engineering Manager

[ennessee Valley Authority

NEDO-33101 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Tennessee Valley Authority (TVA) and GE, Contract Order No. 00001704 Release 00248, effective February 5, 2003, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than TVA, or for any purpose other than that for which it is intended, is not authorized; and, with respect to any unauthorized use, GE 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-33101 Table Of Contents Page EXECUTIVE

SUMMARY

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1. INTRODUCTION .......... . 1-1 1.1 Report Approach . . .1-1 1.2 Purpose and Approach . . .1-1 1.2.1 Uprate Analysis Basis .1-2 1.2.2 Computer Codes .1-2 1.2.3 Approach .1-2 1.2.4 Unit 1 Recovery Configuration Basis .1-3 1.3 Uprated Plant Operating Conditions ....................................... . 1-4 1.3.1 Reactor Heat Balance. 1-4 1.3.2 Reactor Performance Improvement Features. 14 1.4 Summary and Conclusions . . . 1-5 1.5 References . . .1-5

2. REACTOR CORE AND FUEL PERFORMANCE ............... ................... 2-1 2.1 Fuel Design And Operation . . .2-1 2.1.1 Fuel Thermal Margin Monitoring Threshold .2-2 2.2 Thermal Limits Assessment . . .2-2 2.2.1 Safety Limit Minimum Critical Power Ratio .2-2 2.2.2 Minimum Critical Power Ratio Operating Limit .2-3 2.2.3 MAPLHGR and Maximum LHGR Operating Limits .2-3 2.3 Reactivity Characteristics . . .2-3 2.3.1 Power/Flow Operating Map .2-3 2.4 Stability. . . 2-4 2.5 Reactivity Control. . . 2-4 2.5.1 Control Rod Drive System. 2-4 2.5.2 Control Rod Drive Positioning and Cooling .2-5 2.5.3 Control Rod Drive Integrity Assessment .2-5 2.6 References .............  ; 2-6

3. REACTOR COOLANT AND CONNECTED SYSTEMS ....................................... 3-1 3.1 Nuclear System Pressure Relief . . .3-1 3.1.1 MSRV Setpoint Tolerance .3-1 3.2 Reactor Overpressure Protection Analysis . . .3-1 3.3 Reactor Vessel and Internals . . .3-2 3.3.1 Reactor Vessel Fracture Toughness ........................... 3-2 3.3.2 Reactor Vessel Structural Evaluation ........................... 3-3 3.3.3 Reactor Internal Pressure Differences ........................... 3-3 3.3.4 Reactor Internals Structural Evaluation ........................... 3-4 3.3.5 Flow Induced Vibration ........................... 3-8 3.3.6 Steam Separator and Dryer Performance ........................... 3-9 3.4 Reactor Recirculation System .. . 3-10 3.5 Reactor Coolant Pressure Boundary Piping .. . 3-10 iii

NEDO-33101 3.5.1 Pipe Stresses ............................. 3-11 3.5.2 Pipe Supports ............................. 3-11 3.5.3 Piping Flow Induced Vibration ............................. 3-12 3.6 Main Steam Line Flow Restrictors ............................. 3-12 3.7 Main Steam Isolation Valves ..............................  ;.3-12 3.8 Reactor Core Isolation Cooling ............................. 3-13 3.9 Residual Heat Removal System ............................. 3-14 3.9.1 Shutdown Cooling Mode .3-15 3.9.2 Suppression Pool Cooling Mode .3-15 3.9.3 Containment Spray Cooling Mode .3-15 3.9.4 Supplemental Spent Fuel Pool Cooling .3-16 3.9.5 Steam Condensing Mode .3-16 3.9.6 Standby Cooling/Crossties . ................ 3-16 3.10 Reactor Water Cleanup System ............................. 3-16 3.11 Balance-Of-Plant Piping Evaluation ............................. 3-17 3.11.1 Pipe Stresses ............................. 3-18 3.11.2 Pipe Supports ............................. 3-19 3.11.3 Erosion/Corrosion ............................. 3-19 3.12 References .............................. 3-20

4. ENGINEERED SAFETY FEATURES .. 4-1 4.1 Containment System Performance ........................... 4-1 4.1.1 Containment Pressure and Temperature Response .4-1 4.1.2 Containment Dynamic Loads .4-3 4.1.3 Containment Isolation .4-5 4.1.4 Generic Letter 89-10 Program .4-5 4.1.5 Generic Letter 89-16 .4-5 4.1.6 Generic Letter 95-07 .4-6 4.1.7 Generic Letter 96-06 .4-6 4.2 Emergency Core Cooling Systems ............................ 4-6 4.2.1 High Pressure Coolant Injection System .4-6 4.2.2 Low Pressure Coolant Injection .4-7 4.2.3 Core Spray System......... ... 4-7 4.2.4 Automatic Depressurization System .4-7 4.2.5 ECCS Net Positive Suction Head .4-8 4.3 Emergency Core Cooling System Performance ....................................... 4-10 4.4 Main Control Room Atmosphere Control System . .4-11 4.5 Standby Gas Treatment System .. 4-11 4.6 Main Steam Isolation Valve Leakage Control System . .4-11 4.7 Post-LOCA Combustible Gas Control .. 4-11 4.8 References .. 4-12

5. INSTRUMENTATION AND CONTROL . .5-1 5.1 NSSS Monitoring and Control Systems .. 5-1 5.1.1 Control Systems Evaluation .5-1 5.1.2 Neutron Monitoring System .5-1 5.1.3 Rod Worth Minimizer .5-2 iv

NEDO-33101 5.2 BOP Monitoring and Control Systems ............ .. ................ 5-2 5.2.1 Pressure Control System .............................. 5-2 5.2.2 Feedwater Control System .............................. 5-3 5.2.3 Leak Detection System .............................. 5-3 5.3 Instrument Setpoints .................. 5-4 5.3.1 High-Pressure Scram. 5-5 5.3.2 High-Pressure ATWS Recirculation Pump Trip .5-5 5.3.3 Main Steam Relief Valve. 5-5 5.3.4 Main Steam High Flow Isolation. 5-5 5.3.5 Neutron Monitoring System .5-6 5.3.6 Main Steam Line High Radiation Scram .5-6 5.3.7 Low Steam Line Pressure MSIV Closure (RUN Mode) .5-6 5.3.8 Reactor Water Level Instruments .5-6 5.3.9 Main Steam Tunnel High Temperature Isolation .5-7 5.3.10 Low Condenser Vacuum .5-7 5.3.11 TSV Closure and TCV Fast Closure Scram Bypass. 5-7 5.3.12 Rod Worth Minimizer .5-7 5.3.13 Pressure Regulator .5-8 5.3.14 Feedwater Flow Setpoint for Recirculation Cavitation Protection ............. ............ 5-8 5.3.15 RCIC Steam Line High Flow Isolation ..................................................... 5-8 5.3.16 HPCI Steam Line High Flow Isolation ..................................................... 5-8 5.4 References ...................................................... 5-8

6. ELECTRICAL POWER AND AUXILIARY SYSTEMS . .6-1 6.1 AC Power .................................. 6-1 6.1.1 Off-Site Power System .................................. 6-1 6.1.2 On-site Power Distribution System .................................. 6-2 6.2 DC Power .................................. 6-2 6.3 Fuel Pool .................................. 6-3 6.3.1 Fuel Pool Cooling .................................. 6-3 6.3.2 Crud Activity and Corrosion Products .................................. 6A 6.3.3 Radiation Levels .................................. 6-4 6.3.4 Fuel Racks .................................. 6-4 6.4 Water Systems ................................... 6-4 6.4.1 Service Water Systems .6-4 6.4.2 Main Condenser/Circulating Water/Normal Heat Sink Performance .6-6 6.4.3 Reactor Building Closed Cooling Water System .6-6 6.4.4 Raw Cooling Water System .6-7 6.4.5 Ultimate Heat Sink .6-7 6.5 Standby Liquid Control System ......................................... 6-7 6.6 Power Dependent HVAC ....................................... .. 6-8 6.7 Fire Protection ....................................... 6-9 6.7.1 10 CFR 50 Appendix R Fire Event....................................... 6-9 6.8 Systems Not Impacted By Extended Power Uprate ....................................... 6-11 6.8.1 Systems with No Impact....................................................................................... 6-11 6.8.2 Systems With Insignificant Impact ....................................... 6-11 6.9 References . . .6-11 v

NEDO-33101

7. POWER CONVERSION SYSTEMS .......................................... 7-1 7.1 Turbine-Generator .......................................... . 7-1 7.2 Condenser And Steam Jet Air Ejectors .......................................... 7-1 7.3 Turbine Steam Bypass .......................................... 7-2 7.4 Feedwater And Condensate Systems ................ .. ....................... 7-2 7.4.1 Normal Operation ......................................... 7-3 7.4.2 Transient Operation .......................................... 7-3 7.4.3 Condensate Demineralizers ......................................... 7-3

8. RADWASTE AND RADIATION SOURCES .................. ....................... 8-1 8.1 Liquid And Solid Waste Management . . .8-1 8.2 Gaseous Waste Management . . .8-1 8.2.1 Offgas System .8-2 8.3 Radiation Sources In The Reactor Core . . .8-2 8.3.1 Normal Operation .8-2 8.3.2 Normal Post-Operation .8-2 8.4 Radiation Sources In Reactor Coolant . . .8-3 8.4.1 Coolant Activation Products .8-3 8.4.2 Activated Corrosion and Fission Products .8-3 8.5 Radiation Levels. . . 8-4 8.5.1 Normal Operation .8-4 8.5.2 Normal Post-Operation .8-4 8.5.3 Post Accident .8-5 8.6 Normal Operation Off-Site Doses . . .8-5

9. REACTOR SAFETY PERFORMANCE EVALUATIONS ........................................ 9-1 9.1 Reactor Transients .. 9-1 9.1.1 Fuel Thermal Margin Events .9-2 9.1.2 Power and Flow Dependent Limits .9-2 9.1.3 Loss of Feedwater Flow Event .9-2 9.2 Design Basis Accidents .. 9-3 9.3 Special Events .. 9-3 9.3.1 Anticipated Transient Without Scram .9-3 9.3.2 Station Blackout ............................................... 9-5 9.4 References .. 9-6

10. OTHER EVALUATIONS .. 10-1 10.1 High Energy Line Break .. 10-1 10.1.1 Temperature, Pressure and Humidity Profiles .10-1 10.1.2 Pipe Whip and Jet Impingement .10-1 10.1.3 Internal Flooding from High Energy Line Breaks .10-2 10.2 Moderate Energy Line Break .. 10-2 10.3 Environmental Qualification .. 10-2 10.3.1 Electrical Equipment .10-2 10.3.2 Mechanical Equipment With Non-Metallic Components .10-3 10.3.3 Mechanical Component Design Qualification .10-3 10.4 Testing .. 10-3 10.4.1 Recirculation Pump Testing .................. 10-5 vi

NEDO-33101 10.4.2 10 CFR 50 Appendix J Testing .............................. 10-5 10.4.3 Main Steam Line, Feedwater and Reactor Recirculation Piping Flow Induced Vibration Testing ................................ 10-5 10.5 Individual Plant Evaluation ................................. .. 10-6 10.5.1 Initiating Event Frequency .10-7 10.5.2 Component and System Reliability .10-8 10.5.3 Operator Response .10-9 10.5.4 Success Criteria .10-10 10.5.5 External Events .10-10 10.5.6 Shutdown Risk .10-11 10.5.7 Probabilistic Risk Assessment Quality .10-11 10.6 Operator Training And Human Factors . . ............................... 10-12 10.7 Plant Life ................................. 10-14 10.7.1 RPV Internal Components ................................. 10-14 10.7.2 Flow Accelerated Corrosion ................................ 10-15 10.8 References ................................. 10-16

11. LICENSING EVALUATIONS ..................... ........... 11-1 11.1 Other Applicable Requirements .................................. 11-1 11.1.1 NRC and Industry Communications ................................. 11-1 11.1.2 Plant-Unique Items ................................. 11-1 11.2 References ................................. 11-3.

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NEDO-33101 Tables Table 1-1 Glossary of Terms ............................................................... 1-6 Table 1-2 Browns Ferry Current and EPU Plant Operating Conditions .................................. 1-13 Table 1-3 Computer Codes Used For EPU Analyses .............................................................. 1-14 Table 3-1 Browns Ferry Unit I Upper Shelf Energy Equivalent Margin Analysis for 32 EFPY .............................................................. 3-21 Table 3-2 Browns Ferry Unit I Upper Shelf Energy Equivalent Margin Analysis for 54 EFPY ............................................................... 3-23 Table 3-3 Browns Ferry Unit 1 Adjusted Reference Temperatures ......................................... 3-25 Table 3-4 Browns Ferry Unit 1 CUFs of Limiting Components .............................................. 3-26 Table 3-4 Browns Ferry Unit 1 CUFs of Limiting Components .............................................. 3-27 Table 3-5 Browns Ferry Unit 1 RIPDs for Normal Conditions (psid) ..................................... 3-28 Table 3-6 Browns Ferry Unit 1RIPDs for Upset Conditions (psid) ........................................ 3-29 Table 3-7 Browns Ferry Unit 1 RIPDs for Faulted Conditions (psid) ...................................... 3-30 Table 3-8 Browns Ferry Unit 1 Reactor Internal Components - Summary of Stresses ............ 3-31 Table 4-1 Browns Ferry Containment Performance.Analysis Results ..................................... 4-14 Table 4-2 Browns Ferry Short-Term Containment Input to NPSH Analysis ........................... 4-15 Table 4-3 Browns Ferry Long-Term Containment Input to NPSH Analysis ........................... 4-15 Table 4-4 Browns Ferry EPU DBA LOCA NPSH Margins and Containment Overpressure Credit ............................................................... 4-16 Table 4-5 Browns Ferry ECCS Performance Analysis Results ................................................ 4-17 Table 5-1 Browns Ferry Unit I Analytical Limits For Setpoints ............................................... 5-9 Table 6-1 Brown Ferry EPU Plant Electrical Characteristics ................................................... 6-12 Table 6-2 Brown Ferry Offsite Electric Power System ............................................................ 6-12 Table 6-3 Browns Ferry Unit 1 Spent Fuel Pool Parameters .................................................... 6-13 Table 6-4 Browns Ferry Effluent Discharge Comparison ......................................................... 6-14 Table 6-5 Browns Ferry Appendix R Fire Event Evaluation Results ....................................... 6-15 Table 6-6 Browns Ferry Systems With No Impact ............................................................... 6-16 Table 6-7 Browns Ferry Systems With Insignificant Impact ................................................... 6-17 Table 9-1 Browns Ferry Unit 1 Parameters Used for Transient Analysis .................................. 9-7 Table 9-2 Browns Ferry Unit 1 Transient Analysis Results ....................................................... 9-8 Table 9-3 Browns Ferry Key Inputs for ATWS Analysis .......................................................... 9-9 Table 9-4 Browns Ferry ATWS Analysis Results ............................................................... 9-9 viii

NEDO-33101 Table 10-1 Browns Ferry Unit 1 High Energy Line Breaks ................................................... 10-17 Table 10-2 Browns Ferry Unit 1 Equipment Qualification for EPU ...................................... 10-18 Table 10-3 Browns Ferry Summary of CDF and LERE ......................................................... 10-19 Table 10-4 Summary of the Initiator Contributions to CDF and LERF for Browns Ferry Unit I

................................................................... 10-20 1-...........................................

Table 10-5 Frequency Weighted Fractional Importance to Core Damage of Operator Actions Used in Browns Ferry Unit 1 PRA ............................................................... 10-24 Table 10-6 Browns Ferry PRA Peer Review Results ............................................................. 10-26 Table 10-7 Browns Ferry Unit I FAC Parameter Comparison for EPU .............. .................. 10-27 ix

NEDO-33101 Figures Figure 1-1. Browns Ferry EPU Heat Balance - Nominal ....................................................... 1-19 Figure 1-2. Browns Ferry EPU Heat Balance - Overpressure Protection Analysis ................... 1-20 Figure 2-1. Browns Ferry Power/Flow Operating Map ........................................................ 2-7 Figure 3-1. Browns Ferry Response to MSIV Closure with Flux Scram ................. ................ 3-33 Figure 3-2. Browns Ferry Response to Turbine Trip with Bypass Failure and Flux Scram .... 3-34 Figure 4-1. Browns Ferry Time-integrated Containment Hydrogen Generation ..................... 4-18 Figure 4-2. Browns Ferry Uncontrolled H2 and 02 Concentrations in Drywell and Wetwell..4-19 Figure 4-3. Browns Ferry Drywell Pressure Response to CAD Operation Without Venting.. 4-20 Figure 4-4. Browns Ferry CAD System Nitrogen Volume Requirement ................................. 4-21 Figure 9-1. Browns Ferry Turbine Trip with Bypass Failure ................................................... 9-10 Figure 9-2. Browns Ferry Generator Load Rejection with Bypass Failure ................. ............. 9-11 Figure 9-3. Browns Ferry Feedwater Controller Failure - Maximum Demand ........................ 9-12 Figure 9-4. Browns Ferry Feedwater Controller Failure - Maximum Demand with Bypass OOS

........................................................................ 9-13 91.....................................

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NEDO-33101 EXECUTIVE

SUMMARY

This report summarizes the results of all significant safety evaluations performed that justify uprating the licensed thermal power at Browns Ferry Nuclear Plant Unit 1 (hereafter, Browns Ferry unless explicitly noted). The requested license power level is an increase to 3952 MWt from the current licensed reactor thermal power of 3293 MWt.

This report follows the NRC approved generic format and content for BWR EPU licensing reports, documented in NEDC-32424P-A, "Generic Guidelines For General Electric Boiling Water Reactor Extended Power Uprate," commonly called "ELTRI." Per ELTR1, every safety issue that should be addressed in a plant-specific EPU licensing report is addressed in this report.

For issues that have been evaluated generically, this report usually only references the (NRC approved) generic evaluations in either ELTRI or NEDC-32523P-A, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," which is commonly called "ELTR2."

It is not the intent of this report to address all the details of the analyses and evaluations reported herein. For example, only previously NRC-approved or industry-accepted methods were used for the analyses of accidents and transients, as documented in ELTRI. Therefore, the safety analysis methods have been previously addressed, and thus, are not addressed in this report.

Also, event and analysis descriptions that are already provided in other licensing reports or the UFSAR are not repeated within this report. This report summarizes the results of the significant operational and safety evaluations needed to justify a licensing amendment to allow for EPU operation.

Uprating the power level of nuclear power plants can be done safely within plant-specific limits and is a cost-effective way to increase installed electrical generating capacity. Many light water reactors have already been uprated worldwide.

An increase in electrical output of a BWR plant is accomplished primarily by generation and supply of higher steam flow for the turbine generator. Browns Ferry, as originally licensed, has an as-designed equipment and system capability to accommodate steam flow rates at least 5%

above the current rating. Also, Browns Ferry has sufficient design margins to allow Browns Ferry to be safely uprated up to 120% of its OLTP.

A higher steam flow is achieved by increasing the reactor power along slightly modified 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.

Detailed 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 Browns Ferry can safely operate at the requested license power level of 3952 MWt. However, non-safety power generation modifications must be implemented in order to obtain the electrical power output associated with 100% of the EPU RTP level. These modifications have been evaluated and they do not constitute a material alteration to Browns Ferry, as discussed in 10 CFR 50.92.

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NEDO-33101 The evaluations and reviews were conducted in accordance with the criteria in Appendix B of ELTRI. The results of the following evaluations and reviews, presented in this report, were found to be acceptable:

• All safety aspects that are affected by the increase in thermal power and operating pressure were evaluated;

• Evaluations were performed using NRC-approved or industry-accepted analysis methods;

• No change, requiring compliance with a more recent industry code and/or standard, is being requested;

• The UFSAR will be updated for the EPU related changes, after EPU is implemented, per the requirements in 10 CFR 50.7 1(e);

• Modifications will be implemented under 10 CFR 50.59;

• Systems and components affected by EPU were reviewed to ensure there is no significant challenge to any safety system;

• Compliance with current plant environmental regulations were reviewed;

• Potentially affected commitments to the NRC have been reviewed;

• Planned changes not yet implemented have also been reviewed for the effects of EPU;

• All EPU related Technical Specification changes are identified and justified.

The Browns Ferry licensing requirements have been reviewed, and it is concluded that this EPU 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 Browns Ferry, which might cause a significant reduction in a margin of safety.

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NEDO-33101

1. INTRODUCTION 1.1 REPORT APPROACH Uprating the power level of nuclear power plants can be done safely within certain plant-specific limits. Most GE BWR plants have the capability and margins for an uprating of 5 to 20%

without major NSSS hardware modifications. Many light water reactors have already been uprated world wide. Over a thousand MWe have already been added by uprate in the United States. Several BWR plants are among those that have already been uprated. The following evaluation supports an EPU to 3952 MWt, which corresponds to 120% of the OLTP. The OLTP level is 3293 MWt.

This report follows the NRC approved generic format and content for EPU licensing reports, as described in Section 3.0 and Appendices A & B of ELTRI (Reference 1), and the NRC staff position letter reprinted in ELTRI. The analytical methodologies used for ECCS-LOCA evaluations, containment evaluations, transient evaluations, and piping evaluations are documented in ELTRI,Section I and Appendices D, E, G and K. The limitations on use of these methods as defined in the NRC staff position letter reprinted in ELTRI were followed for this EPU analysis.

Many of the component, system and performance evaluations contained within this report have been generically evaluated in ELTR2 (Reference 2), and found to be acceptable. The ELTR2 generic evaluations are based on (1) a 20% thermal power increase, (2) an increased operating dome pressure to 1095 psia, (3) a reactor coolant temperature increase to 5560 F, and (4) steam and FW increases of about 24%. The plant conditions assumed in the ELTR2 evaluations bound the conditions for this EPU.

A glossary of terms is provided in Table 1-1.

1.2 PURPOSE AND APPROACH An increase in electrical output of a BWR is accomplished primarily by generation and supply of higher steam flow to the turbine generator. Most BWRs, as originally licensed, have an as-designed equipment and system capability to accommodate steam flow rates at least 5% above the original rating. In addition, continuing improvements in the analytical techniques (computer codes) based on several decades of BWR safety technology, plant performance feedback, operating experience, and improved fuel and core designs have resulted in a significant increase in the design and operating margin between the calculated safety analyses results and the current plant licensing limits. The available margins in calculated results, combined with the as-designed excess equipment, system, and component capabilities (1) have allowed many BWRs to increase their thermal power ratings by 5% without any NSSS hardware modification, and (2) provide for power increases up to 20% with some non-safety hardware modifications. These power increases involve no significant increase in the hazards presented by the plants as approved by the NRC at the original license stage.

The method for achieving higher power is to use the MELLLA power/flow map, and increase core flow along the MELLLA flow control lines. However, there is no increase in the maximum allowable recirculation flow value. EPU operation does slightly (< 3 %1o)increase reactor vessel dome pressure, to help provide sufficient pressure control and turbine flow capabilities to control the inlet pressure conditions at the turbine.

1-1

NEDO-33101 1.2.1 Uprate Analysis Basis Brown Ferry Unit 1 is currently licensed to 3293 MWt and most of the current safety analyses are based on this value. The EPU RTP level included in this evaluation is 120% of the OLTP.

Plant specific EPU parameters are listed in Table 1-2. The EPU safety analyses are based on a power level of 1.02 times the EPU power level, unless the Regulatory Guide 1.49 two percent power factor is already accounted for in the analysis methods, consistent with the methodology described in Reference 3.

1.2.2 Computer Codes NRC approved or industry-accepted computer codes and calculational techniques are used to demonstrate compliance with the applicable regulatory acceptance criteria. The application of these codes to the EPU analyses complies with the limitations, restrictions and conditions specified in the approving NRC SER where applicable for each code. Any exceptions to the use of the code or conditions of the applicable SER are noted in Table 1-3.

1.2.3 Approach The planned approach to achieving the EPU RTP level consists of: (1) an increase in the core thermal power with a more uniform (flattened) power distribution to create increased steam flow, (2) a corresponding increase in the FW system flow, (3) no increase in maximum allowable core flow, (4) a small (<3%/o) increase in reactor vessel dome pressure, and (5) reactor operation primarily along an extension of the MELLLA rod/flow control lines. This approach is based on, and is consistent with, the NRC-approved BWR generic EPU guidelines that are presented in ELTRI. The plant-unique evaluations are based on a review of plant design and operating data, as applicable, to confirm excess design capabilities, and, if necessary, identify any items which may require modifications associated with the EPU. For some items, bounding analyses and evaluations in ELTR2 demonstrate plant operability and safety. The generic analyses and evaluations in ELTR2 are based on a 20% of original licensed power increase. For the Browns Ferry EPU, the conclusions of system/component acceptability stated in ELTR2 are bounding. The scope and depth of the evaluation results provided herein are established based on the generic BWR EPU guidelines and unique features of the plant. The results of the following evaluations, presented in this report, were found to be acceptable:

(a) Reactor Core and Fuel Performance: Specific analyses required for EPU have been performed for a representative fuel cycle with the reactor core operating at EPU conditions.

Specific core and fuel performance is evaluated for each operating cycle, and will continue to be evaluated and documented for the operating cycles that implement EPU.

(b) RCS and Connected Systems: Evaluations of the NSSS components and systems have been performed at EPU conditions. These evaluations confirm the acceptability of the effects of the higher power and the associated change in process variables (i.e., increased pressure, temperature, and steam and FW flows). Safety-related equipment performance is the primary focus in this report, but key aspects of reactor operational capability are also included.

(c) Engineered Safety Feature Systems: The effects of EPU power operation on the Containment, ECCS, Standby Gas Treatment system and other ESFs have been evaluated for key events. The evaluations include the containment responses during limiting A0Os, special events, ECCS-LOCA, and MSRV containment dynamic loads.

1-2

NEDO-33101 (d) Control and Instrumentation: The control and instrumentation signal ranges and analytical limits for setpoints have been evaluated to establish the effects of the changes in various process parameters such as power, pressure, neutron flux, steam flow and FW flow. As required, setpoint evaluations have been performed to determine the need for any TS allowable value changes for various functions (e.g., MSL high flow isolation setpoints).

(e) Electrical Power and Auxiliary Systems: Evaluations have been performed to establish the operational capability of the plant electrical power and distribution systems and auxiliary systems to ensure that they are capable of supporting safe plant operation at the EPU RTP power level.

(f) Power Conversion Systems: Evaluations have been performed to establish the operational capability of various non-safety BOP systems and components to ensure that they are capable of delivering the increased power output, and/or the modifications necessary to obtain full EPU power.

(g) Radwaste Systems and Radiation Sources: The liquid and gaseous waste management systems have been evaluated at limiting conditions for EPU to show that applicable release limits continue to be met during operation at higher power. The radiological consequences have been evaluated for EPU to show that applicable regulations have been met for the EPU power conditions. This evaluation includes the effect of higher power level on source terms, on-site doses and off-site doses, during normal operation.

(h) Reactor Safety Performance Evaluations: The limiting UFSAR analyses for design basis events are performed as part of the EPU evaluation. ELTRI identifies the limiting analyses that require reanalysis for EPU. The EPU results in no new limiting event beyond those identified in ELTRI. All limiting accidents and transients are analyzed based upon limiting conditions for the EPU and show continued compliance with regulatory requirements.

(i) Additional Aspects of EPU: HELB and EQ evaluations are performed at bounding conditions for the EPU to show the continued operability of plant equipment under the EPU conditions. The effects of the EPU on the plant IPE are analyzed to demonstrate that there are no new vulnerabilities to severe accidents.

(j) Licensing Evaluations: The applicable plant licensing commitments, IE Bulletins, Circulars, Notices, etc. are evaluated for the effects of the EPU. [

)) Items unique to Browns Ferry Unit I are shown to be acceptable for EPU operation.

1.2A Unit 1 Recovery Configuration Basis Part of Browns Ferry Unit 1 recovery process is to update the Unit 1 configuration to be operationally the same as Units 2 & 3, and thus, some non-EPU related Units 2/3 changes are being addressed concurrently with EPU. EPU and other licensing evaluations are based on the Unit 1 configuration as it is planned to be as of the recovery restart date. All EPU changes that potentially could affect safety and/or require NRC approval (i.e., license amendment) per ELTRI are evaluated as documented within this report. Other Unit 1 recovery changes that require license amendments are or will be addressed separately.

1-3

NEDO-33101 1.3 UPRATED PLANT OPERATING CONDITIONS The following evaluations justify increasing the licensed thermal power to 120% of the OLTP value. This new RTP value provides an increase of steam flow to approximately 123% of the original value, and a corresponding increase in electrical power output. To accomplish this performance increase, the rated thermal power is increased to 3952 MWt. The operating reactor vessel dome pressure also is increased, to help retain adequate control of the reactor/turbine. The following descriptions provide information on the original and the EPU plant operating conditions.

1.3.1 Reactor Heat Balance The operating pressure, the total core flow, and the coolant thermodynamic state characterize the thermal hydraulic performance of a BWR reactor core. The EPU values of these parameters are used to establish the steady state operating conditions and as initial and boundary conditions for the required safety analyses. The EPU values for these parameters are determined by performing heat (energy) balance calculations for the reactor system at EPU conditions.

The reactor heat balance relates the thermal-hydraulic parameters to the plant steam and FW flow conditions for the selected core thermal power level and operating pressure. Operational parameters from actual plant operation are considered (e.g., steam line pressure drop) when determining the expected EPU conditions. The thermal-hydraulic parameters define the conditions for evaluating the operation of Browns Ferry at EPU conditions. The thermal-hydraulic parameters obtained for the EPU conditions also define the steady state operating conditions for equipment evaluations. Heat balances at appropriately selected conditions define the initial and boundary conditions for plant safety analyses.

Figure 1-1 shows the EPU heat balance at 100% of EPU and 100% rated core flow. Figure 1-2 shows the EPU heat balance at 102% of EPU and 100% core flow.

Table 1-2 provides a summary the reactor thermal-hydraulic parameters for the current rated and EPU conditions. At EPU conditions, the nominal operating reactor vessel dome pressure is slightly (< 3%/o) increased, to help provide sufficient pressure control range and turbine flow capabilities to control the inlet pressure conditions at the turbine. Section 7.1 demonstrates that the operating pressure increase with some non-safety BOP modifications provide sufficient operating turbine inlet pressure to ensure good pressure control characteristics.

1.3.2 Reactor Performance Improvement Features Because Browns Ferry Unit 1 is being upgraded to be operationally the same as Units 2/3, all of the Units 2/3 reactor performance improvement features will be implemented at Browns Ferry Unit 1. The reactor performance improvement features and the equipment allowed to be OOS are listed in Table 1-2. When limiting, the input parameters related to the performance improvement features or the equipment OOS have been included in the safety analyses for EPU. The use of these performance improvement features and allowing for equipment OOS are to be allowed during EPU operation. The evaluations that are dependent upon cycle length are performed for EPU assuming a 24-month cycle.

1-4

NEDO-33101 Some of the Units 2/3 reactor performance improvement features have also been licensed for Unit 1. The performance improvement features that have not yet been NRC approved for Unit 1, are documented in separate licensing submittals.

1.4

SUMMARY

AND CONCLUSIONS This evaluation encompasses an EPU to 120% of OLTP. The strategy for achieving higher power is to extend the MELLLA power/flow map region along the upper boundary extension.

The Browns Ferry licensing requirements have been reviewed to demonstrate that this uprate can be accommodated without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any existing regulatory limit or design allowable limit applicable to Browns Ferry which might cause a reduction in a margin of safety.

1.5 REFERENCES

1. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor EPU," (ELTRI), Licensing Topical Reports NEDC-32424P-A, Class III (Proprietary),

February 1999; and NEDO-32424, Class I (Non-proprietary), April 1995.

2. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor EPU," (ELTR2), Licensing Topical Reports NEDC-32523P-A, Class III (Proprietary),

February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999.

3. GE Nuclear Energy, "General Electric Standard Application for Reactor Fuel, GESTAR-II," NEDE-2401 1-P-A-14, April 2000.

1-5

NEDO-33101 Table 1-1 Glossary of Terms Term Definition AC Alternating current ADS Automatic Depressurization System ADHR Auxiliary Decay Heat Removal AL Analytical Limit ALARA As Low as Reasonably Achievable ANS American Nuclear Society ANSI American National Standards Institute AOO Anticipated operational occurrences (moderate frequency transient events)

AOP Abnormal Operating Procedure AOV Air Operated Valve APLHGR Average Planar Linear Heat Generation Rate APRM Average Power Range Monitor ARI Alternate Rod Insertion ARTS APRM/RBM/Technical Specifications ASME American Society of Mechanical Engineers AST Alternate Source Term ATWS Anticipated Transient Without Scram AV Allowable Value BHP Brake horse power BIIT Boron injection initiation temperature BOP Balance-of-plant B&PV Boiler and Pressure Vessel BPWS Banked Position Withdrawal Sequence BTU British Thermal Unit BWR Boiling Water Reactor BWROG BWR Owners Group BWRVIP BWR Vessel and Internals Project CAD Containment Atmosphere Dilution CBDT Cause-based decision tree CDF Core damage frequency CFD Condensate filter demineralizer CFR Code of Federal Regulations CLTP Current Licensed Thermal Power 1-6

NEDO-33101 Term Definition CO Condensation oscillation COLR Core Operating Limits Report CPR Critical Power Ratio CRD Control Rod Drive CRDA Control Rod Drop Accident CRDH Control Rod Drive Hydraulic CREVS Control Room Emergency Ventilation System CRHZ Control Room Habitability Zone CSC Containment Spray Cooling CS Core Spray CSS Core support structure CST Condensate Storage Tank CUF Cumulative usage factor DBA Design basis accident DC Direct current DEGB Double-ended guillotine break DHR Decay heat removal DLO Dual (recirculation) loop operation ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EECW Emergency Equipment Cooling Water EFPY Effective full power years EHC Electro-hydraulic control ELTR Extended (Power Uprate) Licensing Topical Report ELTRI Generic Guidelines for General Electric Boiling Water Reactor EPU ELTR2 Generic Evaluations of General'Electric Boiling Water Reactor EPU EOC End of cycle EOP Emergency Operating Procedure EPRI Electric Power Research Institute EPU Extended Power Uprate EQ Environmental qualification ESF Engineered Safety Feature FAC Flow Accelerated Corrosion FFWTR Final Feedwater Temperature Reduction FHA Fuel Handling Accident FIV Flow induced vibration FLIM Failure likelihood index methodology 1-7

NEDO-33101 Term Definition FPCC Fuel Pool Cooling and Cleanup FW Feedwater FWHOOS Feedwater heater out of service GC Generic Communication GDC General Design Criteria GE General Electric Company GENE General Electric Nuclear Energy GL Generic Letter HCR Human cognitive reliability HELB High Energy Line Break HEPA High Efficiency Particulate Air HEP Human error probability Hga Inches of mercury absolute HPCI High Pressure Coolant Injection HPT High-pressure turbine HRA Human Reliability Assessment HVAC Heating Ventilating and Air Conditioning HWC Hydrogen Water Chemistry HWWV Hardened Wetwell Vent HX Heat exchanger IASCC Irradiation-assisted stress corrosion cracking ICF Increased Core Flow ICS Integrated computer system IE Inspection and Enforcement IEB Inspection and Enforcement Bulletin IEEE Institute of Electrical and Electronics Engineers IGSCC Intergranular stress corrosion cracking ILBA Instrument Line Break Accident IORV Inadvertent Opening of a Relief Valve IPE Individual Plant Evaluation IRM Intermediate Range Monitor ISP Integrated Surveillance Program LCO Limiting Conditions for Operation LCS Leakage Control System LDS Leak Detection System LERF Large early release frequency LHGR Linear Heat Generation Rate 1-8

NEDO-33101 Term Definition LOCA Loss-Of-Coolant Accident LOFW Loss-of-feedwater LOOP Loss of Offsite Power LPCI Low Pressure Coolant Injection LPRM Local Power Range Monitor LPSP Low Power Setpoint LTTIP Long Term Torus Integrity Program MAAP Modular Accident Analysis Program MAPLHGR Maximum Average Planar Linear Heat Generation Rate MBTU Millions of BTUs MCPR Minimum Critical Power Ratio MELB Moderate Energy Line Break MEOD Maximum Extended Operating Domain (MELLLA with ICF)

MELLLA Maximum Extended Load Line Limit Analysis MeV Million Electron Volts Mlb Millions of pounds MOV Motor operated valve MS Main steam MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSIV-LCS Main Steam Isolation Valve Leakage Control System MSL Main steam line MSLB Main Steamline Break MSLBA Main Stearnline Break Accident MSRV Main steam relief valve MSRVDL Main steam relief valve discharge line MSVV Main steam valve vault MVA Mega Volt Amps Mvar Megavar MWe Megawatts-electric MWt Megawatt-thermal NA Not Applicable NDE Non-destructive examination NPSH Net positive suction head NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system NTSP Nominal Trip Setpoint 1-9

NEDO-33101 Term Definition NUREG Nuclear Regulations OLMCPR Operating Limit Minimum Critical Power Ratio OLTP Original Licensed Thermal Power 00S Out-of-service OPRM Oscillation Power Range Monitor ORAM Outage Risk Assessment Management AP Differential pressure - psi P2 5 25% of EPU Rated Thermal Power PCS Pressure Control System PCT Peak cladding temperature PRA Probabilistic Risk Assessment PRFO Pressure Regulator Failure - Open PRNMS Power Range Neutron Monitoring System PSA Probabilistic Safety Analysis PSF Performance-shaping factor psi Pounds per square inch psia Pounds per square inch - absolute psid Pounds per square inch - differential psig Pounds per square inch - gauge P-T Pressure-temperature RBCCW Reactor Building Closed Cooling Water RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RCW Raw Cooling Water RHR Residual Heat Removal RHRSW Residual Heat Removal Service Water RIPD Reactor internal pressure difference(s)

RPS Reactor Protection System RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RSLB Recirculation system line break kRS Reactor Recirculation System RTP Rated Thermal Power RTNDT Reference temperature of nil-ductility transition RWCU Reactor Water Cleanup 1-10

NEDO-33101 Term Definition RWE Rod Withdrawal Error RWM Rod Worth Minimizer Salt EPU alternating stress intensity Sm Code allowable stress limit SAR Safety Analysis Report SBO Station blackout SCW Stator Cooling Water SDC Shutdown Cooling SER Safety Evaluation Report SFP Spent fuel pool SGTS Standby Gas Treatment System SHB Shroud head bolts SIL Service Information Letter SJAE Steam Jet Air Ejector SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single-loop operation SOV Solenoid Operated Valve SP Setpoint SPC Suppression Pool Cooling SPDS Safety Parameter Display System SR Surveillance Requirement SRM Source Range Monitor SRP Standard Review Plan SRSS Square Root of the Sum of the Squares SRV Safety relief valve SRVDL Safety relief valve discharge line SSC Structure, system, component SSDS Safe Shutdown System SSP Supplemental surveillance capsule program TAF Top of active fuel TB Turbine Bypass TBOOS Turbine Bypass Out-of-Service TBV Turbine Bypass Valve TFSP Turbine first stage pressure TCV Turbine Control Valve T/G Turbine/generator 1-11

NEDO-33101 Term Definition TLD Thermoluminescent dosimeter TRM Technical Requirements Manual TS Technical Specification(s)

TSV Turbine Stop Valve TVA Tennessee Valley Authority T.. Time available UFSAR Updated Final Safety Analysis Report UHS Ultimate heat sink USE Upper Shelf Energy VFD Variable Frequency Drive 1-12

NEDO-33101 Table 1-2 Browns Ferry Current and EPU Plant Operating Conditions Current*

Licensed EPU Parameter Value Value Thermal Power (MWt) 3293 3952 Vessel Steam Flow (Mlb/hr)** 13.37 16.44 Full Power Core Flow Range (Mlb/hr) 76.9 to 107.6 101.5 to 107.6

% Rated 75 to 105 99 to 105 Maximum Nominal Dome Pressure (psia) 1020 1050 Maximum Nominal Dome Temperature (0 F) 547 551 Pressure at upstream side of turbine stop valve (TSV) (psia) 988 988 Full Power Feedwater Flow (Mlb/hr) 13.32 16.39 Temperature (0 F) 376.6 394.5 Core Inlet Enthalpy (Btu/lb) *** 521.6 523.2

• Based on current reactor heat balance.

• • At normal FW heating.

• At 100% core flow condition.

Performance improvement features:

(1) ARTS-MELLLA with PRNMS (2) EOC Coastdown (GESTAR Generic Analysis)

(3) SLO (4) FFWTR (5) FWHOOS (6) One MSRV OOS (7) 3% MSRV Setpoint tolerance (8) ICF (9) EOCRPT OOS (10) TBOOS (11) 24-month fuel cycle (12) Reactor Level 3 Reduction 1-13

NEDO-33101 Table 1-3 Computer Codes Used For EPU Analyses Task Computer Version NRC Comments Code or Approved Revision Reactor Heat Balance ISCOR 09 Y(l) NEDE-2401 IP Rev. 0 SER Reactor Core and Fuel TGBLA 04 Y NEDE-30130-P-A Performance PANACEA 10 Y NEDE-30130-P-A ISCOR 09 Y(l) NEDE-2401 IP Rev. 0 SER RPV Fluence DORTGOIV 01 N (2)

TGBLA 06 Y (3)

Reactor Vessel Integrity ANSYS 6.1 N (4)

- Stress Evaluation Reactor Internal ISCOR 09 Y(l) NEDE-2401 IP Rev. 0 SER Pressure Differences LAMB 07 (5) NEDE-20566-P-A Transient Analysis PANACEA 10 Y NEDE-30130-P-A (6)

ISCOR 09 Y(M) NEDE-2401 IP Rev. 0 SER ODYN 10 Y NEDO-24154-A SAFER 04 (7) NEDC-32424P-A NEDC-32523P-A, (8) (9) (10)

TASC 03A Y NEDC-32084P-A Rev. 2 ATWS ODYN 10 Y NEDE-24154P-A Supp. 1, Vol. 4 STEMP 04 (1 1)

PANACEA 10 Y NEDE-30130-P-A ISCOR 09 Y(l) NEDE-2401IP Rev. 0 SER TASC 03A Y NEDC-32084P-A Containment System SHEX 05 Y (12)

Response M3CPT 05 Y NEDO-10320, April 1971 LAMB 08 (5) NEDE-20566-P-A September 1986 Appendix R Fire GESTR 08 (7) NEDE-23785-1-PA, Rev. I Protection SAFER 04 (7) (8) (9) (10)

SHEX 04 Y (12)

Reactor Recirculation BILBO 04V NA NEDE-23504, February 1977 (13)

System 1-14

NEDO-33101 Task Computer Version NRC Comments Code or Approved Revision ECCS-LOCA LAMB 08 Y NEDE-20566-P-A GESTR 08 Y NEDE-23785-1-PA, Rev. 1, (8),

(9), NEDC-23785P-A, Vol III, Supp 1, Rev. I SAFER 04 Y NEDE-23785-1 -PA, Rev. 1, (8),

(9), (10), NEDC-23785P-A, Vol 111, Supp 1, Rev. I ISCOR 09 Y(l) NEDE-24011PRev.0SER TASC 03A Y NEDC-32084P-A, Rev. 2 Fission Product ORIGEN2 2.1 N Isotope Generation and Depletion Inventory Code MS Piping Analysis TPIPE 16 Y Structural Analysis Program ME 150 17 (14) Structural Analysis Program HYTRAN 1.6 (14) Hydraulic Transient Analysis RELAP 5 3.2 Y Used for hydraulic modeling of two-phase flow.

HELB-OPC Mass and RELAP 5 MOD 3.2 Y HELB-OPC mass flow rate and Energy Releases enthalpy data.

HELB-OPC GOTHIC 6.1a Y HELB-OPC temperature, pressure Subcompartment and relative humidity profiles for Analysis reactor building areas.

Individual Plant RISKMAN Windows NA RISKMAN is used as the Code for Evaluation 5.02 many IPE submittals to NRC.

MAAP4.0.4, Rev 3 NA MAAP is used for the thermal-hydraulic analysis for many IPE submittals to NRC BOP Performance PEPSE 64A (15) Used to develop the turbine cycle heat balance Multi-flow 1.10 (15) Used for hydraulic modeling of the condensate and FW systems RELAP5 MOD 3.2 Y Used for hydraulic modeling of the heater drain system (two phase flow)

Condenser Evaluation PEPSE 64A (15) Used to develop the turbine cycle heat balance.

Raw Water Cooling Multi-flow 1.10 (15) Used for hydraulic modeling of Evaluation the raw water cooling system.

1-15

NEDO-33101 Task Computer Version NRC Comments Code or Approved Revision Reactor Recirculation HYTRAN 1.6 (15) Used to develop force time Vibration Monitoring history files for pressure pulsation occurring during steady state operation of the reactor recirculation pumps and piping.

TPIPE 16 Y Used to evaluate steady state vibration of the recirculation and attached piping.

Condensate Pump, TPIPE 16 Y Used to evaluate piping Condensate Booster stresses and to determine Pump, and Feedwater support loads for modifications Pump modifications, made to the condensate piping and Steam Packing Multi-flow 1.10 (15) Used for hydraulic modeling of Exhauster Bypass the condensate and FW Valve modifications. systems FW Heater SAP 2000 (15) Used to evaluate stresses in the Evaluations pass partition plates inside the FW heaters Turbine Building WTRCOIL 1.1 (15) Used to evaluate the HVAC Evaluation performance of the cooling coils for increased temperatures due to EPU

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

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

(2) CCC-543, "TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport Version 2.8.14," Radiation Shielding Information Center (RSIC), January 1994.

(3) Letter, S.A. Richards (USNRC) to G. A. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II - Implementing Improved GE Steady-State Methods (TAC No. MA6481)," November 10, 1999.

1-16

NEDO-33101 (4) The code has been reviewed and approved by GENE for "Level-2" application, and is part of GENE's standard design process.

(5) The LAMB code is approved for use in ECCS-LOCA applications (NEDE-20566P-A and NEDO-20566A), but no approving SER exists for the use of LAMB in the evaluation of reactor internal pressure differences or containment system response. The use of LAMB for these applications is consistent with the model description of NEDE-20566P-A.

(6) The physics code PANACEA provides inputs to the transient code ODYN. The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-24011-P-A). The use of PANAC Version 10 in this application was initiated following approval of Amendment 13 of GESTAR II by letter from G.C.

Lainas (NRC) to J.S. Charnley (GE), MFN 028-086,

Subject:

"Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-A Amendment 13, Rev. 6 General Electric Standard Application for Reactor Fuel," March 26, 1998.

(7) The ECCS-LOCA codes are not explicitly approved for Transient or Appendix R usage. The staff concluded that SAFER is qualified as a code for best estimate modeling of loss-of-coolant accidents and loss of inventory events via the approval letter and evaluation for NEDE-23785P, Revision 1, Volume II. (Letter, C.O. Thomas (NRC) to J.F. Quirk (GE), "Review of NEDE-23785-1 (P),

"GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volumes I and II", August 29, 1983.) In addition, the use of SAFER in the analysis of long term Loss-of-Feedwater events is specified in the approved LTRs for power uprate: "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32424P-A, February 1999 and "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," NEDC-32523P-A, February 2000. The Appendix R events are similar to the loss of FW and small break LOCA events.

(8) Letter, J.F. Klapproth (GE) to USNRC, Transmittal of GE Proprietary Report NEDC-32950P "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model," dated January 2000 by letter dated January 27, 2000.

(9) Letter, S.A. Richards (NRC) to J.F. Klapproth, "General Electric Nuclear Energy (GENE) Topical Reports GENE (NEDC)-32950P and GENE (NEDC)-32084P Acceptability Review," May 24, 2000.

(10) "SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-Jet Pump Plants," NEDE-30996P-A, General Electric Company, October 1987.

(11) The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool heatup. The use of STEMP was noted in NEDE-24222, "Assessment of BWR Mitigation of ATWS, Volume I & H (NUREG-0460 Alternate No. 3) December 1, 1979." The code has been used in ATWS applications since that time. There is no formal NRC review and approval of STEMP or the ATWS topical report.

(12) The application of the methodology in the SHEX code to the containment response is approved by NRC in the letter to G. L. Sozzi (GE) from A. Thadani (NRC), "Use of the SHEX Computer Program and ANSVANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.

1-17

NEDO-33101 (13) Not a safety analysis code that requires NRC approval. The code application is reviewed and approved by GENE for "Level-2" application and is part of GENE's standard design process. Also, the application of this code has been used in previous power uprate submittals.

(14) Code provides input to TPIPE. TPIPE has been used by TVA to support submittals to NRC.

(15) Code used to determine nonsafety-related parameters and information.

1-18

NEDO-33101

• Conditions at upstreamside of TSV Core Therxml Power 3951.6 Pump Heating 10.6 Cleanup Losses -4.4 OtherSystemLosses -1.1 Turbine Cycle Use 3956.7 MWt Figure 1-1. Browns Ferry EPU Heat Balance - Nominal

(@ 100% Power and 100% Core Flow) 1-19

NEDO-33101 Legend

1. - Flow, lbm'hr 1070 H = Enthalpy, Btullbm F = Temperature, IF M - Moisture, %

P = Pressure, psia I iilI t ! I Main SteamF1nw 16.831E+06 #

• 1189.6 H

• OA9 M

• Carryunder= 035% 1005 P

• 4031 Main Feed Flow MWt .

Wd = 100 % 16.914E+06 # 16.781E+06 #

A 526.4 H 372.7 H 372.4 H 531.9 °F Total 396.9 °F 396.6 OF Core Flow 102.5E+06 /

Ah= 1.1 H 1.333E+05 #

Jt I52.4 413.1 H 434.2 °F H

Cleanup Demineralizer System I

5.OOOE 404 # Control IRod Drive 1.333E+05 #

48.0 H Feed Flow 525.3 H 77.0 °F 53 1.0 °F

• Conditions at upstream side of TSV Core Thermal Power 4030.6 Pump Heating 10.6 Cleanup Losses -4.4 Other System Losscs -1.1 Turbine Cycle Use 4035.7 MWt Figure 1-2. Browns Ferry EPU Heat Balance - Overpressure Protection Analysis

(@ 102% Power and 100% Core Flow) 1-20

NEDO-33101

2. REACTOR CORE AND FUEL PERFORMANCE 2.1 FUEL DESIGN AND OPERATION EPU increases the average power density proportional to the power increase. Browns Ferry Unit 1 is currently licensed with an average bundle power of 4.31 MW/bundle. The average bundle power for EPU is 5.17 MW/bundle. The Browns Ferry Unit 1 EPU average bundle power is within the range of other operating BWRs.

The average power density has some effects on operating flexibility, reactivity characteristics and energy requirements. The additional energy requirements for EPU are met by an increase in bundle enrichment, an increase in the reload fuel batch size, and/or changes in fuel loading pattern to maintain the desired plant operating cycle length. The power distribution in the core is changed to achieve increased core power, while limiting the MCPR, LHGR and MAPLHGR in any individual fuel bundle to be within its allowable value as defined in the COLR.

At the OLTP or the EPU conditions, all fuel and core design limits continue to be met by planned deployment of fuel enrichment and burnable poison. This is supplemented by core management control rod pattern and/or core flow adjustments. ((

)) However, revised loading patterns, larger batch sizes and potentially new fuel designs may be used to provide additional operating flexibility and maintain fuel cycle length.

The EPU evaluations assume a reference equilibrium core of GE14 fuel. No new fuel product line designs are introduced for EPU, and EPU does not require a change to any fuel design limit. The fuel design limits are established for all new fuel product line designs as a part of the fuel introduction and reload analyses. ((

The reactor core design power distribution usually represents the most limiting thermal operating state at design conditions. It includes allowances for the combined effects on the fuel heat flux and temperature of the gross and local power density distributions, control rod pattern, and reactor power level adjustments during plant operation. NRC approved core design methods were used to analyze core performance at the EPU RTP level. Detailed fuel cycle calculations of a representative core design for this plant demonstrate the feasibility of EPU RTP operation while maintaining fuel design limits. Thermal-hydraulic design and operating limits ensure an acceptably low probability of boiling transition-induced fuel cladding failure occurring in the core, even for the most severe postulated operational transients. As needed, limits are also placed on fuel APLHGR and/or fuel rod LHGR in order to meet both peak cladding temperature limits for the limiting LOCA and fuel mechanical design bases.

The subsequent reload core designs for operation at the EPU RTP level will take into account the above limits, to ensure acceptable differences between the licensing limits and their corresponding operating values.

EPU may result in a small change in fuel burnup, the amount of fuel to be used, and isotopic concentrations of the radionuclides in the irradiated fuel relative to the original level of burnup.

NRC-approved limits for burnup on the fuel designs are not exceeded. Also, due to the higher steady-state operating power associated with the EPU, the short-term curie content of the reactor fuel increases. The effects of higher power operation on radiation sources and design basis 2-1

NEDO-33101 accident doses are discussed in Sections 8 and 9.2, respectively. EPU has some effects on operating flexibility, reactivity characteristics, and energy requirements. These issues are discussed in the following sections based on GE experience and fuel characteristics.

2.1.1 Fuel Thermal Margin Monitoring Threshold The power level above which fuel thermal margin monitoring is required changes with EPU.

The original plant operating licenses set this monitoring threshold at a typical value of 25% of RTP. ((

1))

The fuel thermal margin monitoring threshold is scaled down, if necessary, ((

)) Therefore, the Browns Ferry fuel thermal monitoring threshold is lowered to 23%

(( ))-.

A change in the fuel thermal monitoring threshold also requires a corresponding change to the TS reactor core safety limit for reduced pressure or low core flow. The above discussion is consistent with the TS related discussion in Section 9.1.

2.2 THERMAL LIMITS ASSESSMENT Operating limits ensure that regulatory and/or safety limits are not exceeded for a range of postulated events (e.g., transients, LOCA). This section addresses the effects of EPU on thermal limits. A reference equilibrium core of GE14 fuel is used for the EPU evaluation. Cycle-specific core configurations, evaluated for each reload, confirm EPU capability, and establish or confirm cycle-specific limits, as is currently the practice.

))

2.2.1 Safety Limit Minimum Critical Power Ratio The SLMCPR can be affected slightly by EPU due to the flatter power distribution inherent in the increased power level. ((

] The SLMCPR analysis reflects the actual plant core loading pattern and is performed for each plant reload core. [

))

2-2

NEDO-33101 2.2.2 Minimum Critical Power Ratio Operating Limit The OLMCPR is determined on a cycle-specific basis from the results of the reload transient analysis, as described in Sections 5.3.2 and 5.7.2.1 of ELTRI and Section 3.4 of ELTR2 (Reference 2). This approach does not change for EPU. The required OLMCPR is not expected to significantly change (< 0.03) as shown in Table 3-1 of ELTRI and Figure 5-3 of ELTR2 and from experience with other uprated BWRs. For the reference equilibrium core of GE14 fuel, the OLMCPR for EPU RTP operation is shown in Table 9-2.

(( )) The OLMCPR is calculated by adding the change in MCPR due to the limiting AOO event to the SLMCPR, and is determined on a cycle specific basis. EPU does not change the method used to determine this limit. The effect of EPU on AJO events is addressed in Section 9.1. ((

))

2.2.3 MAPLHGR and Maximum LHGR Operating Limits The MAPLHGR and maximum LHGR limits are maintained as described in subsection 5.7.2.2 of ELTRl. No significant change in operation is anticipated due to the EPU, based on experience from other BWR uprates. The ECCS performance is addressed in Section 4.3, and uses a reference equilibrium core of GE14 fuel for EPU. ((

1))

2.3 REACTIVITY CHARACTERISTICS All minimum shutdown margin requirements apply to cold conditions and are maintained without change.

Operation at higher power could reduce the hot excess reactivity during the cycle. This loss of reactivity does not affect safety, and is not expected to significantly 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.

Increasing hot reactivity may result in less hot-to-cold reactivity differences, and therefore, smaller cold shutdown margins. However, this potential loss in margin can be accommodated through core design, and current design and TS cold shutdown margin requirements are not affected. If needed, a bundle design with improved shutdown margin characteristics can be used to preserve the flexibility between hot and cold reactivity requirements for future cycles.

2.3.1 Power/Flow Operating Map The EPU analyses and evaluations conservatively assume the MELLLA and ICF operating domains. The EPU power/flow operating map (Figure 2-1) includes the operating domain changes for the EPU, and also shows the applicable Browns Ferry performance improvement features (e.g., MELLLA and ICF) addressed in Section 1.3.2. The changes to the power/flow operating map are ((

)) The maximum thermal operating power and 2-3

NEDO-33101 maximum core flow shown on Figure 2-1 correspond to the EPU RTP and the previously analyzed core flow range when rescaled so that EPU RTP is equal to 100% rated. The power/flow operating map changes, incorporated into Figure 2-1, are consistent with the changes shown in Figure 5-1 of ELTRI.

The details of the reactor operating domain for EPU conditions are provided in Figure 2-1. The operating domain for EPU is defined by the following boundaries:

• the MELLLA upper load line, extended up to the EPU RTP level;

• the maximum EPU RTP corresponding to 120% of the OLTP; and

• the ICF line up to EPU RTP at 105% of rated core flow.

Consistent with ELTR1, these boundaries define an increase in the extent of the operating domain above the OLTP between the extended (relative to OLTP) MELLLA upper load line and ICF line.

Thermal hydraulic instability exclusion regions are not shown on Figure 2-1.

Analyses and evaluations have been performed to demonstrate that Browns Ferry may increase core flow to operate within the region of the operating map bounded by the constant speed line between 100P/105F and 52.5P/112.6F for EOC coastdown at constant maximum pump speed line.

EPU does not affect SLO because the maximum attainable thermal power during SLO is less than OLTP, and is limited by the available recirculation flow. SLO is bounded by the MELLLA domain in terms of absolute thermal power versus core flow. Therefore, a separate SLO power/flow operating map is not needed for EPU.

2.4 STABILITY For Browns Ferry, protection against thermal-hydraulic instabilities will be addressed in a separate submittal.

2.5 REACTIVITY CONTROL 2.5.1 Control Rod Drive System The CRD system is used to control core reactivity by positioning neutron absorbing control rods within the reactor and to scram the reactor by rapidly inserting withdrawn control rods into the core. No change is made to the control rods due to the EPU. The effect on the nuclear characteristics of the fuel is discussed in Section 2.3.

The increase in dome pressure due to uprate produces a corresponding increase in the bottom head pressure. Initially, rod insertion is slowed down due to the increased pressure. As the scram insertion continues beyond about 5% insertion, the reactor pressure becomes the primary source of pressure to complete the scram.

2-4

NEDO-33101

))

2.5.2 Control Rod Drive Positioning and Cooling For normal CRD insertion and withdrawal, the required minimum differential pressure between the drive water header and the vessel bottom head is 250 psi. The CRD pumps were evaluated against this requirement and were found to have sufficient capacity. The flows required for CRD cooling and driving are ensured by the automatic opening of the system flow control valve, thus compensating for any increase in reactor pressure.

)) and the automatic operation of the system flow control valve maintains the required drive water pressure and cooling water flow rate. Therefore, the CRD positioning and cooling functions are not affected. The CRD cooling and normal CRD positioning functions are operational considerations, not safety-related functions, and are not affected by EPU operating conditions.

Plant operating data has confirmed that the CRD system flow control valve operating position has sufficient operating margin.

2.53 Control Rod Drive Integrity Assessment

[I )) The CRD system introduces changes in core reactivity by positioning neutron absorbing control rods within the reactor. It is also required to scram the reactor by rapidly inserting withdrawn rods into the core. ((

))the CRD systems for BWR/2-6 are acceptable for EPUs as high as 20% above OLTP. A confirmatory evaluation was performed for EPU. The results of this evaluation are summarized below.

For EPU, the operating dome pressure is increased 30 psi to 1035 psig, which increases the reactor bottom head operating pressure to approximately 1070 psig. The CRD mechanisms are designed for 1250 psig, and thus, the CRD mechanism structural and functional integrity remains acceptable. The CRD mechanism design pressure bounds both the EPU operating pressure and the EPU vessel dome pressure scram analytical limit (see Table 5-1) plus 35 psid at the reactor bottom head.

The CRD components that form part of the primary pressure boundary are designed in accordance with the applicable ASME B&PV Code,Section III. The applicable code effective date for the initially supplied CRDs is the 1968 Edition up to and including the Summer 1970 Addenda. BWR/6 drives, modified for BWR/2-5 use, have replaced the original drives and the applicable code effective date is the 1974 Edition, including the Winter 1975 Addenda.

The CRD mechanism indicator tube has a calculated primary membrane plus bending stress of 20,790 psi. The allowable stress is conservatively specified as 26,060 psi (i.e., 1.5 Sm). The maximum stress on this component results from the maximum CRD internal hydraulic pressures

[I )) caused by a postulated abnormal operating condition. This postulated abnormal operating condition assumes a failure of the system pressure regulating valve that 2-5

NEDO-33101 subsequently applies the maximum pump discharge pressure to the CRD mechanism internal components. Because the reactor operating condition does not affect the CRD pump discharge pressure, the abnormal pressure condition assumed in the analysis remains the same regardless of reactor pressure. Therefore, the stresses for the limiting CRD component are maintained within the allowable stress criteria for the maximum EPU operating pressure. The postulated abnormal pressure operating condition bounds the ASME reactor overpressure limit.

The CRD mechanism has also been subjected to extensive testing at 1250 psig, which is higher than the maximum EPU pressure. Based on the demonstrated performance of the CRD mechanism at high pressures, it is concluded that any additional deformation resulting from the EPU pressure increase is not significant and of no consequence.

EPU increases the reactor bottom head temperature by approximately 40 F to 532 0F. This effect is considered to be insignificant, (( 11 The analysis for cyclic operation of the CRD was conservatively evaluated in accordance with the applicable requirements specified in the ASME B&PV Code,Section III. For example, when considering the loadings resulting from scram with a leaking scram discharge valve, scram with a failed buffer, and scram without CRD cooling water flow, the CRD main flange yields a fatigue usage factor of (( )), which is much less than the allowable limit of 1.0. All requirements are satisfied when considering the increased EPU vessel bottom head pressure, thereby satisfying the peak stress intensity limits governed by fatigue.

))

2.6 REFERENCES

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

2. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2), Licensing Topical Reports NEDC-32523P-A, Class III (Proprietary), February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999.

2-6

Core Flow (MIbhr) 0 10 20 30 40 50 60 70 80 90 100 110 120 120 4500 110 100 4000 90 3500 Pi so 3000 0 70 W

4 Li 60 0 0

I-50 2000 EW~

2500 eœ 2L1 40 30 20 500 10 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core J3ow (°/0)

Figure 2-1. Browns Ferry Power/Flow Operating Map

NEDO-33101

3. REACTOR COOLANT AND CONNECTED SYSTEMS 3.1 NUCLEAR SYSTEM PRESSURE RELIEF The nuclear system pressure relief system prevents overpressurization of the nuclear system during AOOs, the plant ASME Upset overpressure protection event, and postulated ATWS events. The MSRVs along with other functions provide this protection. An evaluation was performed in order to confirm the adequacy of the pressure relief system for EPU conditions.

The adequacy of the pressure relief system is also demonstrated by the overpressure protection evaluation performed for each reload core and by the ATWS evaluation performed for EPU (Section 9.3.1).

The MSRV setpoints are increased consistent with the operating dome pressure increase. The analytical limits for the EPU analyses, which use the upper tolerance limits of the MSRV setpoints, are as shown in Table 5-1.

The updated nominal MSRV setpoints ensure that adequate differences (simmer margin) between operating pressure and the MSRV setpoints are maintained. Also, the setpoints are high enough to prevent an increase in the number of unnecessary MSRV actuations due to normal plant maneuvers at the EPU conditions. Based on the analysis described in Section 3.2, it is concluded that the nuclear boiler pressure relief system has the capability to provide sufficient overpressure protection to accommodate the EPU. These changes are consistent ((

))

3.1.1 MSRV Setpoint Tolerance MSRV setpoint tolerance is independent of EPU. EPU evaluations are performed using the existing MSRV setpoint tolerance analytical limit of 3% as a basis. Actual historical in-service surveillance of MSRV setpoint performance test results are monitored separately for compliance to the TS requirements.

3.2 REACTOR OVERPRESSURE PROTECTION ANALYSIS The design pressure of the reactor vessel remains at 1250 psig. The acceptance limit for pressurization events is the ASME code allowable peak pressure of 1375 psig (110% of design value). The overpressure protection analysis description and analysis method are provided in Section 5.5.1.4 and Appendix E of ELTRI (Reference 1). As shown in Table E-1 of ELTRI, the limiting pressurization events are the MSIV closure and turbine trip with turbine bypass failure.

Both events are (conservatively) analyzed assuming a failure of the valve position scram. The analyses also assume that the events initiate at a reactor dome pressure of 1055 psig (which is higher than the nominal EPU dome pressure), the MSRV analytical limits in Table 5-1, and one MSRV (with the lowest setpoint) OOS. Starting from 102% of EPU RTP, the calculated peak RPV pressure, located at the bottom of the vessel, is 1342 psig. The corresponding calculated maximum reactor dome pressure is 1314 psig. The peak calculated RPV pressure remains below the 1375 psig ASME limit, and the maximum calculated dome pressure remains below the TS 1325 psig Safety Limit. Therefore, there is no decrease in margin of safety. The results of the EPU overpressure protection analysis are given in Figures 3-1 and 3-2 and are consistent ((

1]

3-1

NEDO-33101 3.3 REACTOR VESSEL AND INTERNALS The RPV structure and support components form a pressure boundary to contain the reactor coolant and moderator, and form a boundary against leakage of radioactive materials into the drywell. The RPV also provides structural support for the reactor core and internals.

Comprehensive reviews have assessed the effects of increased power conditions on the reactor vessel and its internals. These reviews and associated analyses show continued compliance with the original design and licensing criteria for the reactor vessel and internals.

33.1 Reactor Vessel Fracture Toughness The Browns Ferry Unit 1 neutron fluence is reanalyzed for EPU, using neutron transport methodology (Reference 3) consistent with Regulatory Guide 1.190. The Regulatory Guide 1.190 fluxes are conservatively applied for the entire 60 year plant life. The revised fluence is used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G. The results of these evaluations indicate that:

(a) The USE remains bounded by the BWROG equivalent margin analysis, thereby demonstrating compliance with Appendix G. The USE evaluation results are provided in Tables 3-1 and 3-2.

(b) The beltline material RTNDT remains below 200'F.

(c) The TS P-T curves will be revised considering the increases in shifts affecting the beltline portion of the curves. These curves will be provided in a separate licensing submittal. The hydrotest pressure for EPU is the minimum nominal operating pressure.

(d) The 40-year life (32 EFPY) shift is increased, and consequently, requires a change in the adjusted reference temperature, which is the initial RTNDT plus the shift. This shift is used to revise the P-T curves. These values and the 60-year life (54 EFPY) shift are provided in Table 3-3.

(e) The surveillance program consists of three capsules. All three capsules for Browns Ferry Unit 1 have been in the reactor vessel since plant startup. EPU has no effect on the existing surveillance schedule. The current schedule calls for the removal of the first set of specimens from the RPV at the end of the first cycle after restart (U 1C7 Refueling Outage),

which most closely represents 8 EFPY of operation. Per the intent of ASTM El 85-82 and compliance with 10 CFR 50 Appendix H, the remaining specimens would be withdrawn every 6 EFPY after withdrawal of the first set of samples. This would mean that the second and third sets of specimens are presently scheduled for removal at 14 EFPY and 20 EFPY, respectively.

The maximum nominal operating dome pressure for EPU is changed to be consistent with Units 2/3, where the hydrostatic and leakage test pressures are acceptable for a power uprate with a 30 psi operating pressure increase. Because the vessel is in compliance with the regulatory requirements, operation with EPU does not have an adverse effect on the reactor vessel fracture toughness.

3-2

NEDO-33101 3.3.2 Reactor Vessel Structural Evaluation The effect of EPU 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 summer 1965, 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. New stresses were determined by scaling the "original" stresses based on the EPU conditions of pressure, temperature and flow. The analyses were performed for the design at Normal and Upset, and Emergency and Faulted conditions. Any increase in annulus pressurization, jet reaction, pipe restraint, or fuel lift loads, was considered in the analysis of the components affected for Normal, Upset, Emergency and Faulted conditions.

3.3.2.1 Design Conditions Because there are no changes in the design conditions (vessel pressure and temperature) due to EPU, the design stresses are unchanged and the Code requirements are met.

3.3.2.2 Normal and Upset Conditions The reactor coolant temperature and flows at EPU 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 EPU conditions. A primary plus secondary stress analysis was performed showing EPU stresses still meet the requirements of the ASME Code,Section III, subsection NB. Lastly, the fatigue usage was evaluated for the limiting location of components with a usage factor greater than 0.5. The Browns Ferry fatigue analysis results for the limiting components are provided in Table 3-4. The Browns Ferry analysis results for EPU show that all components meet their ASME Code requirements.

Browns Ferry Unit 1 FW nozzles with the triple-sleeve, double-seal, thermal sleeve design are qualified by UT inspection methods based on ASME Section XI Code, which are approved by NRC.

FFWTR is considered in the fatigue usage evaluation and included with the system cycling effects shown in Table 3-4.

3.3.2.3 Emergency and Faulted Conditions Emergency and faulted conditions did not change for any component. Therefore, Code requirements are met for all RPV components.

3.3.3 Reactor Internal Pressure Differences The increase in core average power, pressure, and core flow region resulted in higher core loads an'd RIPDs due to the higher core exit steam quality. The maximum acoustic and flow-induced loads, following a postulated recirculation line break, are shown to be insignificantly affected by EPU, including the ICF conditions.

3-3

NEDO-33101 The RIPDs are calculated for Normal (steady-state operation), Upset, and Faulted conditions for all major reactor internal components. [

Tables 3-5 through 3-7 compare results for the various loading conditions between original analysis results and operation with EPU for the vessel internals that are affected by the changed RIPDs.

3.3.4 Reactor Internals Structural Evaluation The reactor internals consist of the CSS components and non-CSS components. The reactor internals are not certified to the ASME code, except the CRD as noted, however; the requirements of the code are used as guidelines in their design basis analysis. The evaluations and stress reconciliation in support of the thermal power increase are performed consistent with the design basis analysis of the components. The reactor internal components evaluated are:

Core Support Structure Components

• Shroud

• Shroud Support

• Core Plate

• Top Guide

• Control Rod Drive Housing

• Control Rod Guide Tube

• Orificed Fuel Support

• Fuel Channel Non-Core Support Structure Components

• Steam Dryer

• Feedwater Sparger

• Jet Pumps

• Core Spray Line and Sparger

• Access Hole Cover

• Shroud Head and Steam Separator Assembly

• In-core Housing and Guide Tube

• Vessel Head Cooling Spray Nozzle

• Jet Pump Instrument Penetration Seal

• Differential Pressure and Standby Liquid Control Line

• Control Rod Drive 3-4

NEDO-33101 The original configurations of the internal components are considered in the EPU evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation.

The effects on the loads as a result of the thermal-hydraulic changes due to EPU are evaluated for the reactor internals. All applicable loads and load combinations are considered consistent with the existing design basis analysis. These loads include the RIPDs, seismic loads, flow induced and acoustic loads due to RSLB-LOCA, and thermal loads. The RIPDs increase for some components/loading conditions as a result of EPU. The flow conditions and thermal effects were considered in the evaluation, as applicable. The seismic response is unaffected by EPU. The acoustic and flow induced loads in the annulus as a result of the RSLB-LOCA are included in the evaluation, and are bounded by pre-EPU values.

The EPU loads are compared to those in the existing design basis analysis. If the loads do not increase due to EPU, then the existing analysis results bound the EPU conditions, and no further evaluation is required or performed. If the loads increase due to the EPU, then the effect of the load increase is evaluated further. ((

))

Table 3-8 presents the governing stresses for the various reactor internal components. All stresses are within allowable limits and the reactor internal components are demonstrated to be structurally adequate for EPU.

The following reactor vessel internals are evaluated for the effects of changes in loads due to EPU.

a) Shroud: ((

3] Therefore, the structural integrity of the Shroud is acceptable for EPU.

b) Shroud Support: [

)) Therefore, the structural integrity of the Shroud Support is acceptable for EPU.

c) Core Plate: ((

3-5

NEDO-33101

)) Therefore, the core plate remains structurally qualified for EPU.

d) Top Guide: ((

)) Therefore, the structural integrity of the top guide is acceptable for EPU.

e) Control Rod Drive Housing: ((

Therefore, the structural integrity of the CRD housing is acceptable for EPU.

f) Control Rod Guide Tube: ((

)) Therefore, the structural integrity of the control rod guide tube is acceptable for EPU.

g) Orificed Fuel Support: f[

)) Therefore, the structural integrity of the orificed fuel support is acceptable for EPU.

h) Fuel Channel: ((

)) Therefore, the structural integrity of the fuel channels is acceptable for EPU.

i) Steam Dryer: ((

)) In response to the recent dryer failures observed at another BWR site during EPU operation, a detailed evaluation will be performed to examine dryer components susceptible to failure at EPU conditions. The results of the quantitative evaluation will be used to identify any additional modifications needed to maintain steam dryer structural integrity at EPU conditions. If any steam dryer components requiring modification are identified, these modifications will be implemented prior to operation at the EPU conditions. Refer to subsection 3.3.5.

j) Feedwater Sparger: ((

3-6

NEDO-33101

)) Therefore, the structural integrity of the FW sparger is acceptable for EPU.

k) Jet Pumps: ((

)) Therefore, the structural integrity of the jet pump assembly is acceptable for EPU.

1) Core Spray Lines and Spargers: ((

)) Therefore, the structural integrity of the core spray line and the spargers is acceptable for EPU.

m) Access Hole Cover (Bolted Design): ((

)) Therefore, the structural integrity of the Access Hole Cover is acceptable for EPU.

n) Shroud Head and Steam Separator Assembly (including Shroud Head Bolts): ((

)) Therefore, the structural integrity of the shroud head and steam separator assembly is acceptable for EPU.

o) In-Core Housing and Guide Tube: [(

)) Therefore, the structural integrity of the In-core Housing and Guide Tube is acceptable for EPU.

p) Vessel Head Cooling Spray Nozzle: ((

] Therefore, the structural integrity of the vessel head cooling spray nozzle is acceptable for EPU.

q) Jet Pump Instrument Penetration Seal: ((

)) Therefore, the structural integrity of the jet pump instrument penetration seal is acceptable for EPU.

3-7

NEDO-33101 r) Core Differential Pressure and Liquid Control Line: ((

)) Therefore, the structural integrity of the differential pressure and standby liquid control line is acceptable for EPU.

s) Control Rod Drive: ((

)) Therefore, the structural integrity of the CRD is acceptable for EPU.

3.3.5 Flow Induced Vibration The core flow dependent RPV internals (in-core guide tube and control rod guide tube components) are acceptable for EPU operation, because the maximum core flow does not change.

EPU operation increases the steam production in the core, resulting in an increase in the core pressure drop. There is only a small increase (2.3%) in maximum drive flow at EPU conditions as compared to OLTP. The increase in power may increase the level of reactor internals vibration. Analyses were performed to evaluate the effects of FIV on the reactor internals at EPU conditions. This evaluation used a bounding reactor power of 3952 MWt and 105% of rated core flow. This assessment was based on vibration data obtained during startup testing of Browns Ferry Unit 1. For components requiring an evaluation but not instrumented in Browns Ferry Unit 1, vibration data acquired during the startup testing from similar plants or acquired outside the RPV is used. The expected vibration levels for EPU were estimated by extrapolating the vibration data recorded in Browns Ferry Unit 1 or similar plants and on GE BWR operating experience. These expected vibration levels were then compared with the established vibration acceptance limits. The following components were evaluated:

a) Shroud b) Shroud head and moisture separator c) Jet pumps d) FW sparger e) In-core guide tubes f) Control rod guide g) Steam dryer h) Jet pump sensing lines The results of the vibration evaluation show that continuous operation at a reactor power of 3952 MWt and 105% of rated core flow does not result in any detrimental effects on the safety-related reactor internal components.

3-8

NEDO-33101

))

During EPU operation, the components in the upper zone of the reactor, such as the moisture separators and dryer, are mostly affected by the increased steam flow. Components in the core region and components such as the core spray line are primarily affected by the core flow.

Components in the annulus region such as the jet pump are primarily affected by the recirculation pump drive flow and core flow. Because the increase in drive flow is small with core flow remaining the same at EPU conditions, the resulting change in vibration is negligible for components in the annulus and core regions. Only the moisture separator and dryer are significantly affected by EPU conditions. The steam dryer and moisture separators are not safety-related components. However, the moisture separator loads act on the shroud through the shroud head. Because the shroud is a safety-related component, the separator/shroud structure was tested at various power conditions up to rated power during startup. The separator/shroud structure was evaluated from these test data.

The calculations for EPU 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 less than the ASME Code criteria of 13,600 psi;

• The modes are absolute summed; and

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

Based on the above, it is concluded that FIV effects are expected to remain within acceptable limits at EPU conditions.

In addition to the above components, which were evaluated for flow induced vibration per the requirement of NRC Regulatory Guide 1.20, a supplemental evaluation was performed for additional components. The following components were additionally evaluated for EPU conditions and determined to be structurally adequate to withstand the effects of flow induced vibrations: Guide Rods, Top Head Instrument Nozzle, Head Spray Nozzle, Top Head Vent Nozzle, Core Spray Sparger, Core Spray Piping, Fuel Assembly, Shroud Head Bolts, Steam line Nozzle, Water Level Instrument Nozzle, and Top Guide.

3.3.6 Steam Separator and Dryer Performance The performance of the steam separators and dryer has been evaluated to ensure that the quality of the steam leaving the reactor pressure vessel remains acceptable at EPU conditions. EPU 3-9

NEDO-33101 increases the saturated steam generated in the reactor core. At constant core flow, this results in an increase in the separator inlet quality and dryer face velocity and a decrease in the water level inside the dryer skirt. These factors, in addition to the core radial power distribution, affect the steam separator-dryer performance. Steam separator-dryer performance was evaluated at EPU equilibrium cycle limiting conditions of high radial power peaking and the applicable core flow range shown on the power-flow map (Figure 2-1). The predicted steam moisture content is acceptable based on a revised moisture content performance specification of < 0.3 weight %.

The operation and performance of the NSSS components with moisture content up to the revised performance specification was determined to be acceptable. The effect of increasing steam moisture content on the radiation source terms is addressed in Section 8.0.

3.4 REACTOR RECIRCULATION SYSTEM The EPU power condition is accomplished by operating along extensions of current rod lines on the power/flow map (Figure 2-1) with no increase in the maximum core flow at EPU RTP. The core reload analyses are performed with the most conservative allowable core flow. The evaluation of the reactor recirculation system performance at the EPU RTP will ensure that adequate core flow can be maintained.

((I SLO is unchanged by EPU.

The system piping has been reviewed for operation at the uprated conditions and found to meet its design requirements (see Section 3.5). System components (e.g., pumps and valves) will be evaluated at EPU conditions to ensure that safety and design objectives are met.

3.5 REACTOR COOLANT PRESSURE BOUNDARY PIPING An alternative piping evaluation process, ((

)) was used for the Browns Ferry EPU RCPB piping inside the primary containment. A description of this piping evaluation process is included within this section.

Browns Ferry incorporated the pressure, temperature, and flow changes due to EPU into the TPIPE analysis computer model for all of the RCPB piping systems listed below. New support loads, penetration loads, equipment nozzle loads, piping movements, pipe break/crack locations, and maximum stress values for each code equation were generated based on the TPIPE analysis results. The analysis results were evaluated and documented in the associated calculations. The EPU pressures and temperatures were incorporated into thermal operating modes calculations for each RCPB piping system. These thermal operating modes calculations were used as design input documents for the piping stress analysis calculations. Also included in the above analysis 3-10

NEDO-33101 effort were changes due to NRC IEB 79-14 walkdown data, seismic design criteria and spectra changes, and piping and piping component replacement design changes. All of these changes combined affected the analysis results requiring support modifications and additions for each piping system.

The RCPB piping systems evaluated for EPU include RRS, MS (including MSIV drains), RCIC, HPCI, FW (including Reactor Vessel Level Indicating System), RWCU, CS, SLCS, RHR, RPV Head Vent, RPV Bottom Drain, MSRVDLs, CRDH and Primary Chemistry Sampling.

3.5.1 Pipe Stresses Operation at the EPU conditions increases stresses in the piping systems and loads on piping system components and supports due to higher operating temperatures, pressures and flow rates internal to the pipes. The changes due to NRC IEB 79-14 walkdown data, seismic design criteria and spectra changes, and piping and piping component replacement design changes included in the analysis with EPU also affected the analysis results. The BFN Unit 1 piping analyses effort stress results, including EPU as well as the other changes discussed above, for each of the RCPB systems were checked against the USAS-B31.1.0, 1967 Code stress criteria and found acceptable. All fatigue usage factors satisfy the code requirements. For high energy lines, the postulated break/crack locations were identified based on the analyses results and were evaluated. The higher EPU pressure effects on the pipe whip restraints have been evaluated and found acceptable. Equipment nozzle loads were checked against the applicable allowable for the equipment and found acceptable. Potential clearance problems identified by the NRC IEB 79-14 walkdown between the piping and other plant commodities (including pipe whip restraints) were evaluated and found acceptable. Clearances will also be checked during plant heat up for restart testing. Other items including penetration anchors, flanges, etc. as applicable to each piping problem were also evaluated and found acceptable. The analysis results and associated evaluations discussed herein are documented in the analysis calculation for each of the RCPB piping problems. The TSV closure transient affects on the MS piping increased due to the EPU pressure and flow changes. The MS analysis included the TSV transient and the stresses from this event were found acceptable.

Flow-accelerated corrosion for all potentially affected piping systems is addressed in 3.11.3.

3.5.2 Pipe Supports Operation at the EPU conditions slightly increases the pipe support loadings due to increases in the temperature, pressure and flow of the affected piping systems. A much larger affect on the pipe supports for the RCPB systems for Browns Ferry Unit I were the changes due to NRC IEB 79-14 walkdown data, seismic design criteria and spectra changes, and piping and piping component replacement design changes incorporated into the piping analysis. The Browns Ferry Unit 1 support calculations incorporated the analysis loads for EPU conditions and the loads from the other analysis changes. The TSV transient loads were included in the calculations for the MS supports. All existing RCPB piping supports were qualified as is or were modified as required to meet design criteria. New supports added were qualified based on the new analyses loads. These support changes will be installed in the field prior to EPU implementation.

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NEDO-33101 3.5.3 Piping Flow Induced Vibration Key applicable structures include the MS system piping and suspension, the FW system piping and suspension, and the RRS piping and suspension. In addition, branch lines attached to the MS system piping or FW system piping are considered.

RRS drive flow is not significantly increased (< 5%) during EPU operation. [

Even though the RRS flow rates for EPU are essentially the same as previously experienced and tested, testing of the FIV levels of the RRS piping will be performed.

The MS and FW piping have increased flow rates and flow velocities in order to accommodate EPU. As a result, the MS and FV piping experience increased vibration levels, approximately proportional to the square of the flow velocities. The ASME Code and nuclear regulatory guidelines require some vibration test data be taken and evaluated for these high-energy piping systems during initial operation at EPU conditions. Vibration data for the MS and FW piping inside and outside containment will be acquired using remote sensors, such as displacement probes, velocity sensors, and accelerometers. A piping vibration startup test program will be performed and the results will be reviewed for acceptability. The FIV testing will be performed during EPU power ascension.

The safety-related thermowells and sample probes in the MS, FW and RRS piping systems were evaluated, and found to be adequate for the increased MS, FW and RRS flows as a result of EPU.

3.6 MAIN STEAM LINE FLOW RESTRICTORS The increase in steam flow rate has no significant effect on flow restrictor erosion. There is no effect on the structural integrity of the MS flow element (restrictor) due to the increased differential pressure because the restrictors were designed and analyzed for the choke flow condition.

After a postulated steamline break outside containment, the coolant flow in the broken steamline increases until it is limited by the MSL flow restrictor. However, the resulting doses, considering the 30 psi increase in dome pressure, remain below the 10 CFR 50.67 guidelines, and are bounded by the analysis discussed in Section 9.2.

Therefore, the MSL flow restrictors are not significantly affected by EPU.

3.7 MAIN STEAM ISOLATION VALVES The MSIVs are part of the RCPB, and perform the safety function of steam line isolation during certain abnormal events. The MSIVs must be able to close within a specified time range at all design and operating conditions. They are designed to satisfy leakage limits set forth in TS.

The MSIVs have been (( )) evaluated, as discussed in Section 4.7 of ELTR2. The evaluation covers both the effects of the changes to the structural capability of the MSIV to meet pressure boundary requirements, and the potential effects of EPU-related changes to the safety functions of the MSIVs. ((

3-12

NEDO-33101

)) The MSIVs are modified to reduce their pressure drop. The modified valves will continue to meet the closure stroke time requirement. The MSIVs are also modified due to higher valve stem forces experienced by the increased steam flow rate. Therefore, ((

)) and the MSIVs are acceptable for EPU operation.

3.8 REACTOR CORE ISOLATION COOLING The RCIC system evaluation scope is provided in Section 5.6.7 of ELTRI.

The RCIC system is required to maintain sufficient water inventory in the reactor to permit adequate core cooling following a reactor vessel isolation event accompanied by loss of flow from the FW system. The system design injection rate must be sufficient for compliance with the system limiting criteria to maintain the reactor water level above TAF at the EPU conditions.

The RCIC system is designed to pump water into the reactor vessel over a wide range of operating pressures. As described in Section 9.1.3, this event is addressed on a plant specific basis. The results of the Browns Ferry plant-specific evaluation indicate adequate water level margin above TAF at the EPU conditions. Thus, the RCIC injection rate is adequate to meet this design basis event.

An operational requirement is that the RCIC system can restore the reactor water level while avoiding ADS timer initiation and MSIV closure activation functions associated with the low-low-low reactor water level setpoint (Level 1). This requirement is intended to avoid unnecessary initiations of these safety systems. The results of the Browns Ferry plant-specific evaluation indicates that the RCIC system is capable of maintaining the water level outside the shroud above nominal Level I setpoint through a limiting LOFW event at the EPU conditions.

Thus, the RCIC injection rate is adequate to meet the requirements for inventory makeup. (See Section 9.1)

((

))

Operation of the RCIC system at EPU conditions does not have any effect on the availability or the reliability of the system, and does not invalidate any of the original design pressures or temperatures for the system components.

The RCIC surveillance test range pressure is (in part) based on the maximum normal reactor dome pressure. Because the maximum normal reactor dome pressure increases by 30 psi, the RCIC surveillance test range also should be increased by 30 psi. The surveillance test range is 3-13

NEDO-33101 increased from

• 1010 psig and 2 920 psig to < 1040 psig and 2 950 psig, consistent with the 30 psi increase to the nominal reactor operating pressure.

II )) because there are no physical changes to the pump suction configuration, and no changes to the system flow rate or minimum atmospheric pressure in the suppression chamber or CST. EPU does not affect the capability to transfer the RCIC pump suction on high suppression pool level or low CST level from its normal alignment, the CST, to the suppression pool, and does not change the existing requirements for the transfer. For ATWS (Section 9.3.1) and Appendix R (Section 6.7.1), operation of the RCIC system at suppression pool temperatures greater than the operational limit may be accomplished by using the dedicated CST volume as the source of water. Therefore, the specified operational temperature limit for the process water does not change with the EPU. ((

)) Operation of the RCIC system during Station Blackout events is discussed in Section 9.3.2.

The reactor system response to a Loss of Feedwater transient with RCIC is discussed in Section 9.1.

For Browns Ferry, a portion of the CST volume (135,000 gallons) is reserved for RCIC operation by the use of a standpipe in the tank. The increase in reactor decay heat due to EPU reduces the amount of time that RCIC can maintain reactor vessel level in hot shutdown conditions utilizing this reserve volume from greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to a little less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

This is not a safety-related function and procedures are in place to direct the establishment of additional sources of water if the CST level approaches the top of the standpipe. Additionally, the UFSAR provides a discussion of suppression pool temperature following reactor vessel isolation with RCIC operation. Operation of RCIC during this time would not be affected by EPU conditions; however, the energy added to the suppression pool from the MSRV discharge would increase due to the increased decay heat associated with EPU.

)) Therefore, the RCIC system is acceptable for EPU.

3.9 RESIDUAL HEAT REMOVAL SYSTEM The RHR system evaluation process is described in Section 5.6.4 of ELTRI. The following results for the RHR system evaluation ((

1]

The RHR system is designed to restore and maintain the reactor coolant inventory following a LOCA and remove reactor decay heat following reactor shutdown for normal, transient, and accident conditions. The EPU effect on the RHR system is a result of the higher decay heat in the core corresponding to the uprated power and the increased amount of reactor heat discharged into the containment during a LOCA. The RHR system is designed to operate in the LPCI mode, SDC mode, SPC mode, CSC mode, Supplemental Spent Fuel Pool Cooling and Standby Cooling/Crossties mode.

The LPCI mode, as it relates to the LOCA response, is discussed in Section 4.2.2.

3-14

NEDO-33101 The SPC mode is manually initiated following isolation transients or a postulated LOCA to maintain the containment pressure and suppression pool temperature within design limits. The CSC mode reduces drywell pressure, drywell temperature and suppression chamber pressure following an accident. The adequacy of these operating modes is demonstrated by the containment analysis (Section 4.1).

The higher suppression pool temperature and containment pressure during a postulated LOCA (Section 4.1) do not affect hardware capabilities of RHR equipment to perform the LPCI, SPC, and CSC functions.

The Supplemental Spent Fuel Pool Cooling mode, using existing RHR heat removal capacity, provides supplemental fuel pool cooling capability in the event that the fuel pool heat load exceeds the heat removal capability of the FPCC system. The adequacy of fuel pool cooling, including use of the Supplemental Spent Fuel Pool Cooling mode, is addressed in Section 6.3.

3.9.1 Shutdown Cooling Mode

((

))

3.9.2 Suppression Pool Cooling Mode The functional design basis as stated in the UFSAR for the Suppression Pool Cooling Mode during normal plant operation is to control the initial pool temperature below the TS limit to so that the pool temperature immediately after a blowdown does not exceed the condensation limit in the event of a design basis LOCA, and to ensure the long-term pool temperature does not exceed the torus attached piping analysis limit. The EPU maximum suppression pool temperature (Section 4.1.1) is utilized as the torus attached piping analysis temperature limit in the torus attached piping analysis, therefore, this objective is met for EPU.

The increase in decay heat due to EPU increases the heat input to the suppression pool resulting in slightly higher containment temperature and pressure during the initial stages of a LOCA. The EPU effect on the containment (drywell and torus) temperature, pressure and condensation limit after a design basis LOCA is described in the containment analysis (Section 4.1.1).

As shown in Section 4.2.5, there is adequate NPSH margin during the RHR pump operation under the post-LOCA operating conditions.

3.9.3 Containment Spray Cooling Mode The CSC mode provides water from the suppression pool to spray headers in the drywell and suppression chamber to reduce containment pressure and temperature during post-accident 3-15

NEDO-33101 conditions. Following EPU, increases in the post-LOCA containment spray temperature correspond to the increase in suppression pool temperature. The rate of increase has a negligible effect on the calculated values of drywell pressure, drywell temperature, and suppression chamber pressure since these parameters reach their highest values prior to actuation of the containment spray, as shown in Sections 4.1.1.2 and 4.1.1.3.

The CSC mode is used to reduce containment pressure following a LOCA, which can affect the available NPSH. The adequacy of NPSH margin during the RHR pump operation under the post-LOCA operating conditions is discussed in Section 4.2.5.

3.9A Supplemental Spent Fuel Pool Cooling The RHR Supplemental Spent Fuel Pool Cooling Mode, using the existing RHR heat removal capacity, provides supplemental fuel pool cooling in the event that the fuel pool heat load exceeds the heat removal capability of the FPCC system due to off loading of the entire core.

This mode operates along with the FPCC system to maintain the Fuel Pool temperature within acceptable limits during a reactor cold shutdown. The increased short-term fuel pool heat load due to EPU does not exceed the combined heat removal capacities of this mode and FPCC system. (See Section 6.3.)

3.9.5 Steam Condensing Mode Steam Condensing mode of RHR is not installed at Browns Ferry.

3.9.6 Standby Cooling/Crossties Standby Cooling/Crossties utilizes the standby coolant supply connection and the RHR crossties to provide additional long-term redundancy to the emergency core and containment cooling systems. This function is not affected by EPU because the performance requirements for the emergency core and containment cooling systems are not changed.

3.10 REACTOR WATER CLEANUP SYSTEM RWCU system operation at the EPU RTP level slightly increases the pressure and temperature within the RWCU system. This system is designed to remove solid and dissolved impurities from recirculated reactor coolant, thereby reducing the concentration of radioactive and corrosive species in the reactor coolant. The system is capable of performing this function at the EPU RTP level.

Based on operating experience, the FW iron input to the reactor increases as a result of the increased FW flow. This input increases the calculated reactor water iron concentration from 19.3 ppb to 23.7 ppb. However, this change is considered insignificant, and does not affect RWCU.

The effects of EPU on the RWCU system functional capability have been reviewed, and the system can perform adequately during EPU with the original RWCU system flow. This RWCU system flow results in a slight increase in the calculated reactor water conductivity (from 0.10IOS/cm to 0.11 pIS/cm) because of the increase in FW flow. The present reactor water conductivity limits are unchanged for EPU, and the actual conductivity remains within these limits.

3-16

NEDO-33101 The system piping and components have been reviewed for operation at the uprated conditions (pressure and temperature) and found to meet its safety and design objectives, including maintaining structural integrity during normal, upset, emergency and faulted conditions. In the event of a HELB in the system piping, appropriate isolation shall be achieved (see Section 4.1.3).

Refer to Sections 3.5 and 3.11 for evaluation of pipe and support adequacy, and Section 10.1 for the HELB evaluation.

3.11 BALANCE-OF-PLANT PIPING EVALUATION The BOP piping systems evaluation consists of a number of piping subsystems that move fluid through systems outside the RCPB piping.

For some BOP piping systems, the flow, pressure, temperature, and mechanical loads do not increase. [

))

The evaluation of the Reactor Building BOP piping and supports was performed in a manner similar to the evaluation to the evaluation of RCPB piping systems and supports (Section 3.5),

where the affected analyses and calculations incorporated the EPU conditions and compared the results to the applicable code equation limits (ASME Section III, Division I through the September 1977 Addenda, Subsections NC (torus attached piping only) or USAS B31.1.0 1967 Power Piping Code equations. The original Codes of record (as referenced in the appropriate calculations) and Code allowables were used and no new assumptions were introduced. As with the RCPB piping, the Reactor Building BOP piping analyses effort included changes due to NRC IEB 79-14 walkdown data, seismic design criteria and spectra changes (not applicable for torus attached piping), and piping and piping component replacement design changes. Other items including penetration anchors, RPV nozzles, flanges, commodity clearances, etc. were also evaluated for effects by the analyses results. The analyses results and associated evaluations discussed herein are documented in the analysis calculation for each of the Reactor Building BOP piping problems. Some of the torus piping calculations had small changes which did not require the TPIPE analysis to be redone, however, the previous analyses results were increased by hand calculations for the change in temperatures, and the new loads and stresses were verified to be acceptable. The torus attached piping stress calculations were revised to document EPU changes and evaluation results.

For Turbine Building BOP piping, if a pipe stress calculation was available, the thermal stresses were factored by the percent increase in the EPU temperature over the original analyzed temperature, and compared to the B31.1.0 Power Piping Code allowable values. For Turbine Building piping where pipe stress calculations do not exist, the increase in EPU Temperature is addressed by qualitatively evaluating the available flexibility in the piping system or by performing a new flexibility analysis. The TSV closure transient effects on the MS piping increased due to the EPU pressure and flow changes. The MS analysis included the TSV transient, and the stresses from this event were found to be acceptable. For those EPU pressure changes greater than original design pressure, the increased pressure stress was calculated for code acceptance.

3-17

NEDO-33101 The DBA-LOCA dynamic loads, including the pool swell loads vent thrust loads, CO loads and chugging loads were re-evaluated and found acceptable, and there are no resulting effects on the torus shell attached structures. The torus post-LOCA temperature increases were included in the thermal operating mode calculations used as input for the torus attached piping stress evaluations performed for the Unit 1 restart. The larger thermal displacements of the torus due to the increased post-LOCA temperatures were also included in the evaluation of the torus attached piping and documented in the associated calculation.

The effects of the EPU conditions have been evaluated for the following piping systems:

• MS - Outside Containment

• Extraction Steam, Heater Vents and Drains

• FW and Condensate

• RWCU - Outside Containment

• RHR - Outside Containment

• RHR Service Water - Outside Containment

• CS - Outside Containment - Pump Suction / Pump Discharge

• HPCI - Outside Containment

• RCIC - Outside Containment

• SLCS - Outside Containment

• CRD

• EECW

• RBCCW

• SFC/ADHR

• RCW/SCW

• SGTS

• Off Gas

• Torus Attached Piping including ECCS Suction Strainers 3.11.1 Pipe Stresses Operation at the EPU conditions increases stresses on piping and piping system components due to higher operating temperatures, pressures and flow rates internal to the pipes. For those systems with existing analysis documentation, the maximum stress levels and fatigue analysis results were reviewed based on specific increases in temperature, pressure and flow rate. For those systems that do not require a detailed analysis, pipe routing and flexibility were evaluated and determined to be acceptable. Based on the design margins between actual stresses and code limits, all piping is below the code allowables of the applicable code of record: USAS B3 1.1.0 -

1967 Power Piping Code or ASME Boiler and Pressure Vessel Code - Section III, Division I 3-18

NEDO-33101 through the summer 1977 Addenda (torus attached piping only). For high energy lines in the Reactor Building, the postulated break/crack locations are identified based on the analyses results, and will be evaluated prior to EPU implementation. The higher EPU pressure effects on the pipe whip restraints have been evaluated and found acceptable. Reactor Building BOP piping equipment nozzle loads were checked against the applicable allowables for the equipment and found acceptable. Potential clearance problems, identified by the NRC IEB 79-14 walkdown, between the piping and other plant commodities (including pipe whip restraints) were evaluated and found acceptable.

3.11.2 Pipe Supports Operation at the EPU conditions slightly increases the pipe support loadings due to increases in the temperature of the affected piping systems.

The pipe supports of the Reactor Building systems affected by EPU loading increases were also affected by changes due to the NRC IEB 79-14 walkdown data, seismic design criteria and spectra changes (except torus attached pipe), and piping and piping component replacement design changes. All of these changes combined resulted in various support modifications to meet design requirements. The pipe supports have been re-evaluated based on the pipe stress analysis results, which include the effects of EPU conditions. All modifications required to the supports will be complete prior to EPU implementation.

Turbine Building BOP piping supports were reviewed based on the percent increase in thermal loads due to the EPU temperature increase over the original design temperature. Any modifications required to these supports will be complete prior to EPU implementation.

3.11.3 Erosion/Corrosion The integrity of high energy piping systems is assured by proper design in accordance with the applicable codes and standards. A consideration in assuring proper design and maintaining system operation within the design is the allowable piping thickness values. Piping thickness values of carbon steel piping and components can be affected by FAC. Browns Ferry has an established program for monitoring pipe wall thinning in single phase and two-phase carbon steel piping. Process variables that influence FAC at Browns Ferry are moisture content, water chemistry, temperature, oxygen, flow path geometry and velocity, and material composition.

EPU operation results in some changes to parameters affecting FAC in those systems associated with the turbine cycle (e.g., condensate, FW, MS, extraction steam). The evaluation of and inspection for FAC in BOP systems is addressed by compliance with NRC Generic Letter 89-08, "Erosion/Corrosion in Piping." The Browns Ferry FAC program currently monitors the affected systems (see Section 10.7). Continued monitoring of the systems provides a high level of confidence in the integrity of potentially susceptible piping systems. Appropriate changes to piping inspection frequency will be implemented to ensure adequate margin is maintained for those systems where process conditions change. This includes adjustments to predict material loss rates to project the need for maintenance/replacement prior to reaching minimum wall thickness requirements. The program provides assurance that the EPU does not adversely affect piping systems potentially susceptible to pipe wall thinning due to FAC.

3-19

NEDO-33101 3.12 REFERENCES

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

2. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2), Licensing Topical Reports NEDC-32523P-A, Class III (Proprietary), February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999.

3. GE Nuclear Energy "GE Methodology to RPV Fast Neutron Flux Evaluations," Licensing Topical Report NEDC-32983P, Class III (Proprietary), August 2000, and NEDO-32983-A, Class I (Non-proprietary), December 2001.

4. TVA letter, Browns Ferry Nuclear Plant (BEN) - Units 2 and 3 - NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," Revised Commitments for Feedwater Nozzle Nondestructive Examination (TAC Nos. M08436 and M08437), R08001023713, October 23, 2000.

3-20

NEDO-33101 Table 3-1 Browvns Ferry Unit 1 Upper Shelf Energy Equivalent Margin Analysis for 32 EFPY a) Plate EMA 32 EFPY-Plant Applicability Verification Form Surveillance Plate USE--Not Available:

%Cu = N/A 15' Capsule Fluence= N/A 2 nd Capsule Fluence = N/A 1I5 Capsule Measured % Decrease = N/A (Charpy Curves) 2nd Capsule Measured % Decrease = N/A (Charpy Curves) lit Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2) 2nd Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2)

Limiting Beltline Plate USE (Heat B5864-1):

%Cu= 0.15 32 EFPY I/4T Fluence= 7.91 E+17 n/cm2 R.G. 1.99 Predicted % Decrease = 13.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 13.5% < 21%, so vessel plates are bounded by equivalent margin analysis 3-21

NEDO-33101 Table 3-1 (continued) b) Weld EMA 32 EFPY-Plant Applicability Verification Form Surveillance Weld USE-Not Available:

%Cu= N/A 1"S Capsule Fluence = N/A 2nd Capsule Fluence = N/A ISt Capsule Measured % Decrease = N/A (Charpy Curves) 2 nd Capsule Measured % Decrease = N/A (Charpy Curves)

I" Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2) 2 nd Capsule R.G. 1.99 Predicted % Decrease = N/A (RG. 1.99, Figure 2)

Limiting Beltline Weld USE (Heat 406L44):

%Cu = 0.27 32 EFPY 1/4T Fluence = 7.91 E+17 n/cm 2 R.G. 1.99 Predicted % Decrease = 23.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 23.5% < 34%, so vessel welds are bounded by equivalent margin analysis 3-22

NEDO-33101 Table 3-2 Browns Ferry Unit I Upper Shelf Energy Equivalent Margin Analysis for 54 EFPY a) Plate EMA 54 EFPY--Plant Applicability Verification Form Surveillance Plate USE-Not Available:

%Cu = N/A 15' Capsule Fluence = N/A 2nd Capsule Fluence = N/A 1St Capsule Measured % Decrease = N/A (Charpy Curves) 2nd Capsule Measured % Decrease = N/A (Charpy Curves) 1I Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2) 2 nd Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2)

Limiting Beltline Plate USE (Heat B5864-1):

%Cu= 0.15 32 EFPY 1/4T Fluence = 1.33 E+18 n/cm 2 R.G.1.99 Predicted % Decrease = .15.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 15.5% < 23.5%, so vessel plates are bounded by equivalent margin analysis 3-23

NEDO-33101 Table 3-2 (continued) b) Weld EMA 54 EFPY-Plant Applicability Verification Form Surveillance Weld USE-Not Available:

%Cu= N/A Ist Capsule Fluence = N/A 2 nd Capsule Fluence = N/A 1't Capsule Measured % Decrease = N/A (Charpy Curves) 2 nd Capsule Measured % Decrease = N/A (Charpy Curves) 1St Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2) 2 nd Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2)

Limiting Beltline Weld USE (Heat 406L44):

%Cu= 0.27 32 EFPY I/4T Fluence = 1.33 E+18 n/cm 2 R.G. 1.99 Predicted % Decrease = 26.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 26.5% < 39%, so vessel welds are bounded by equivalent margin analysis 3-24

NEDO-33101 Table 3-3 Browns Ferry Unit I Adjusted Reference Temperatures a) 32 EFPY BROWNS FERRY I Loowerltrrmndleat Plte ad Axla Welds Th es ihn Iches

• 6.13 Ratio Peak/Locatlon -1.00 32EFPYPeakLD. Ruence- 1.410E18 iVcm2 32 EFPY Peakl14 T uece 9.76E+17 n/aTr2 32 EFPY Peak 14 Y fluee 9.76E+17 rVrI2 Lower Plates and Axial Welds & Lower to Lowor4ritermdlade G0r1hWeld Thckness InsInch- 6.13 Ratio Peak/ Location

• 0.81 D eneS* 1.14E-18 32 EFPY Peak lO. rntma2 32EFPYPeAk1I4Tfluknae- 7.91E+17 Vcm'2 32 EFPYPeak14T Sten- 791E+17 aVaM2 FEATOR -"al 114T 32 EFPY 32 EFFY 32 EFPY COMPONENT HEATILOT %Cu %Nl CF RTndt Flurnc a RTndt a, a Margin Shift ART F riem-2 F F .F *F PLATES:

Lwe Shell 6-127-1 1.099.1 0.14 0.60 100 -20 7.91E+17 37 0 17 34 71 51 6-127-2 B5864 1 0.15 0.44 101 *20 7.91E+17 38 0 17 34 72 52 6-127-4 A1009.1 0.14 0.50 96 -10 7.91E-17 35 0 17 34 69 59 L erjnr4termedle Shell 6-139-19 C2684-2 0.12 0.53 82 14 9.760E17 34 0 17 34 67 81 6-139-20 C2868-2 0.09 0.48 58 30 9.7?0+17 24 0 12 24 48 78 6t139-21 C2753-1 0.08 0.50 51 2 9.76E+17 21 0 11 21 42 44 WELDS:

AxW elds ESW 0.24 0.37 141 23.1 9.76E.17 58 13 28 62 120 143 Lower to Lower4ntrnnodlat 0GirthWeld WF154 406L 0.27 060 184 20 7 E910+17 C8 10 28 59 12$ 148

• SW dq.u7 . b1..pW- aAW fli9. Sp.4. -id l - sm .bk.

3-25

NEDO-33101 Table 3-3 (continued)

BROWNS FERRY I Lorr4ntarmodlate Plates and Axial Welds Thickness hIInches

• 6.13 RatloPeak/Location- 1.00 54EFPYPeakL.Dblence- 238E*18 VCm^'2 54 EFPY Peak 1/4 T lence l .BSE18 rdam^2 54 EFPY Peak 1/4T tkbse

• 1.65E 1t n/csm2 Loer Plates and Axial Welds S Lower to Lower4npternedite Girth Weld Thickness h Indis- 6.13 Raio Peak/Loaion - 0.1 54 EFPY Peak tI.Ikonce- 1.93E.18 rnm^2 54 EFPYPeak 1/4T b*ence.1.33E.-16 ncm^2 54 EFPY Peak 1/4Ttfluence. 1.33E.l5 nr=%i2 HEAT OR Initia 1/4 T 54 EFPY 54 EFPY 54 EFPY COMPONENT HEATLOT %Cu %NI CF RTndt Fklunce RTndt . Uarghin M. Shitl ART

_ *F n/cm^2 *F *F .F *F PLATES:

L er Shell 6-127.1 A0999-1 0.14 0.60 100 .20 1.33=18 48 0 17 34 82 62 6-127.2 r35864-1 0.15 0.44 101 -20 .33&E-1 48 0 17 34 82 62 6-127.4 A1009-1 0.14 0.50 96 .10 1.33E+18 4S 0 17 34 so 70 Lrr4ntennetate Shell 6-139-19 C28U-2 0.12 0.53 62 14 1.651+18 43 0 17 34 77 91 6-13920 C28S8-2 0.09 0.48 58 30 1.65E+18 30 0 15 30 61 91 6-139-21 C2753 1 0.08 0.50 51 2 1.65E+18 27 0 13 27 53 55 WELDS:

Axial Welcd ESW _ 0.24 0.37 141 23.1 1.65E+18 74 13 28 62 136 159 Lowr to Lower4nt dsateGirth Weld WFt54 I -40644 0.27 0.60 184 20 133E+1B 68 10 28 59 147 167 t SW OdVry ted W.. IIA* f9. Spft.id hldbhW - -w 3-26

NEDO-33101 Table 34 Browns Ferry Unit 1 CUFs of Limiting Components P + Q Stress (ksi) CUF Component(l) Current EPU Allowable Current EPU Allowable (ASME (40-year Code Limit) life)

Feedwater Nozzle 37.3 57.7 69.9 (3Sm) 0.984 0.654() 1.0 Main Closure Studs 103.3 103.3 110.1 (3Sm) 0.762 0.575 1.0 Support Skirt 73.5°) 51.3°) 80.1 (3Sm) 0.904 0.066(4) 1.0 Recirculation Outlet 75.5 78.2 80.1 (3Sm, 0.779 0.494 1.0 Nozzle Notes:

1. Only components with Usage Factor greater than 0.5 are included in this table.

2. The values shown are excluding thermal bending, as permitted by the ASME Code, to show Code compliance.

3. Feedwater Nozzle CUF based on a 12-year seal refurbishment cycle; however, Brown Ferry Unit 1 intends to revise its commitment regarding FW nozzle refurbishment and inspection to be consistent with the Browns Ferry Units 2 and 3 submittal to the NRC (Reference 4).

4. Support skirt was reanalyzed using finite element analysis. Original analysis used conservative computational methods. This results in a large peak stress, which in turn produces large Ke factor and results in very large alternating stress, Sa. The large Sa gives low allowable cycles resulting in high CUF. The finite element analysis done for EPU considerably reduces the peak stress and Ke. The reduced Ke gives larger allowable cycles and with plant specific cycles (reduced number of actual cycles) results in very low CUF for the EPU analysis.

3-27

NEDO-33101 Table 3-5 Browns Ferry Unit I RIPDs for Normal Conditions (psid)

Parameter OLTP* EPU*

Core Plate and Guide Tube 22.18 24.40 Shroud Support Ring and Lower Shroud 30.29 32.89 Upper Shroud 8.11 8.55 Shroud Head 8.24 9.43 Shroud Head to Water Level (Irreversible**) 12.24 Shroud Head to Water Level (Elevation**) 0.94 Top Guide 0.68 0.61 Steam Dryer 0.27 0.42 Fuel Channel Wall 13.31

• 105% core flow

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

• Not available at OLTP conditions 3-28

NEDO-33101 Table 3-6 Brovwns Ferry Unit 1 RIPDs for Upset Conditions (psid)

Parameter OLTP* EPU*

Core Plate and Guide Tube 24.58 26.80 Shroud Support Ring and Lower Shroud 32.69 35.29 Upper Shroud 12.17 12.82 Shroud Head 12.36 14.14 Shroud Head to Water Level (Irreversible**) 18.36 Shroud Head to Water Level (Elevation**) 1.41 Top Guide <1.00 0.92 Steam Dryer 0.41 0.62 Fuel Channel Wall 16.21

• 105% core flow

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

• • • Not available at OLTP conditions 3-29

NEDO-33101 Table 3-7 Browns Ferry Unit 1 RIPDs for Faulted Conditions (psid)

Parameter OLTP* EPU*

Core Plate and Guide Tube 32.0 28.5 Shroud Support Ring and Lower Shroud 53.0 51.0 Upper Shroud 29.0 29.0 Shroud Head 29.5 29.5 Shroud Head to Water Level (Irreversible**) 32.0 Shroud Head to Water Level (Elevation**) 1.4 Top Guide 1.3 1.1 Steam Dryer 5.8 7.7****

Fuel Channel Wall 15.5

• 105% core flow

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

• • • Not available at OLTP conditions

• • • • These pressure drops are for a MSLB outside primary containment. The steam dryer pressure drop is greatest for the high flow, low power condition (interlock point). The interlock condition has not changed with the EPU.

3-30

NEDO-33101 Table 3-8 Browns Ferry Unit 1 Reactor Internal Components - Summary of Stresses Item Component Location Category/ Stress/Load Pre-EPU EPU Allowable Service Category Basis Value Condition Value*

1 Shroud Normal! Bounded by Pre-EPU design basis Loads/Stresses Upset Emergency/

Fault.

2 Shroud Support Design Stress (psi) 24,500 30,062 34,950 Operating I I I 3 Shroud Support Faulted Bounded by Pre-EPU design basis Loads/Stresses 4 Core Plate Normal! Buckling/Sliding 25.2 l 26.8 28.0 Upset AP (psid) l l l 5 Core Plate Emergency/ Bounded by Pre-EPU design basis Loads/Stresses Faulted 6 Top Guide Normal/ Bounded by Pre-EPU design basis Loads/Stresses Upset Emergency/

Faulted 7 CRD Housing Qualitative Assessment (See Section 3.3.4(e))

8 Control Rod Guide Normal! AP Buckling 0.24 0.26 0.40 Tube Upset (p/pC) I I I 9 Control Rod Guide Emergency/ Bounded by Pre-EPU design basis Loads/Stresses Tube Faulted 10 Orificed Fuel Support Normal! Stress (psi) l 12,413 1 12,527 l 15,580 Upset __ _ I _ _ I _ I 11 Orificed Fuel Support Emergency! Bounded by Pre-EPU design basis Loads/Stresses Faulted 12 Fuel Channel Qualified per Proprietary Fuel Design Basis 13 Steam Dryer (Hood) Normal! Stress (psi) 4,054 1 5,027 16,950 Upset _

14 Steam Dryer Faulted Bounded by Pre-EPU design basis Loads/Stresses 15 Feedwater Sparger Normal! Pm + Pb + Q - 70,800 70,910 76,500 Slotted Ring Upset Therm. Bending

___ _(psi) .

16 Feedwater Sparger Normal! Pm + Pb (psi) 5,190 6,990 21,450 Header Pipe/Tee Upset 17 Feedwater Sparger Emergency Pm + Pb (psi) 6,020 7,820 28,600 Header Pipe/Tee I I 3-31

NEDO-33101 Item Component Location Category/ Stress/Load Pre-EPU EPU Allowable Service Category Basis Value Condition Value*

18 Feedwater Sparger Faulted Pm + Pb (psi) 33,690 35,490 42,900 Header Pipe/Tee 19 Jet Pump Normal/ Qualitative Assessment (See Section 3.3.4(k))

Upset Emergency/

Faulted 20 Core Spray Line and Qualitative Assessment (See Section 3.3.4(1))

Sparger 21 Access Hole Cover Normal! Pm + Pb (psi) 6,756 l 7,093 l 34,950

_ (Bolted Design) Upset l1 22 Access Hole Cover Emergency/ Bounded by Pre-EPU design basis Loads/Stresses (Bolted Design) Faulted 23 Shroud Head and Normal/ Pm + Pb (psi) 33,993 34,489 34,950 Steam Separator Upset Assembly (SHB) 24 SHB Emergency Pm + Pb (psi) 31,348 34,671 52,425 25 SHB Faulted Pm + Pb (psi) 41,432 41,758 69,900 26 In-Core Housing and Qualitative Assessment (See Section 3.3.4(o))

Guide Tube 27 Vessel Head Cooling Qualitative Assessment (See Section 3.3.4(p))

Spray Nozzle 28 Jet Pump Instrument Qualitative Assessment (See Section 3.3.4(q))

Penetration Seal 29 Differential Pressure Qualitative Assessment (See Section 3.3.4(r))

and Standby Liquid Control Line 30 Control Rod Drive Qualitative Assessment (See Section 3.3.4(s))

• Pre-EPU Basis Values are based on 105% OLTP with pressure increase, which bounds the 100% OLTP condition.

3-32

NEDO-33101

-Vessel Press Risa (Psi) 41-SafetyValve Fkyw

-Refief Valve Flow

-Bypass Valve Flow 2710' 250 2 1755 Is 127Ao iSa

.0o 0 a0 1.0 2.0 30 41 to 10 7.0 10 *0 Thm (sec) mA to so 4* 6o 6.0 ?A t0o to O00 1.0 t5 10 4.0 sO .

.0 O aa t0 Time (sec) Time (sec)

Figure 3-1. Browns Ferry Response to MSIV Closure with Flux Scram (102% EPU power, 105% core flow, and 1055 psig initial dome pressure) 3-33

NEDO-33101

'nor

-- Noutron Fhtx -_-Vessel Press Rise (psi)

-- Ave Surface Heat Flux -*-SafetyVate Flaw

-*-Core Irdet Flow --* Relet Valve Flow

-- Bypass VaNe Flow

.C GA --*.-Core Infet Subcoolng is

'a-a 12t.

I 0

'-0 0 2.0 25

• O 5o to ?Af 0a Is LO sa 40 (0 5 .

Tkm. (sec) * (sc)

-0

.9 12 at 0a0 1i0 za 3A AB to £ 3.0 a0 so ea 70 Tkm (see) Time, (sec)

Figure 3-2. Browns Ferry Response to Turbine Trip with Bypass Failure and Flux Scram (102% EPU power, 105% core flow, and 1055 psig initial dome pressure) 3-34

NEDO-33101

4. ENGINEERED SAFETY FEATURES NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 6.1.1, subsection I states, "Engineered safety features (ESF) are provided in nuclear plants to mitigate the consequences of design basis or loss-of-coolant accidents." The Browns Ferry plant features evaluated within this section are designed to (directly) mitigate the consequences of postulated accidents, and thus, are classified in the Browns Ferry UFSAR as engineered safety features.

4.1 CONTAINMENT SYSTEM PERFORMANCE This section addresses the effect of the EPU on various aspects of the Browns Ferry containment system performance.

The UFSAR provides the containment responses to various postulated accidents that validate the design basis for the containment. Operation at the EPU RTP causes changes to some of the conditions for the containment analyses. For example, the short-term DBA LOCA containment response during the reactor blowdown is governed by the blowdown flow rate. This blowdown flow rate is dependent on the reactor initial thermal-hydraulic conditions, such as vessel dome pressure and the mass and energy of the vessel fluid inventory, which change slightly at the EPU RTP. Also, the long-term heatup of the suppression pool following a LOCA or a transient is governed by the ability of the RHR system to remove decay heat. Because the decay heat depends on the initial reactor power level, the long-term containment response is affected by EPU. The containment pressure and temperature responses have been reanalyzed, as described in Section 4.1.1, to demonstrate Browns Ferry's acceptability for operation at EPU RTP.

The analyses were performed in accordance with Regulatory Guide 1.49 and ELTRI (Reference 1) using GE codes and models (References 2 through 5). The GE methods have been reviewed and approved by the NRC (References 6 and 7). Confirmatory calculations with the SHEX code and the NRC-accepted HXSIZ code show a difference of less than 10 F in peak suppression pool temperature between the two codes. Therefore, the use of the SHEX code for Browns Ferry complies with the NRC requirements for use in EPU analyses presented in Reference 8.

The effect of EPU on the containment dynamic loads due to a LOCA or MSRV discharge has also been evaluated as described in Section 4.1.2. These loads were previously defined generically during the Mark I Containment LTTIP as described in Reference 9, and accepted by the NRC per References 6 and 7. Plant-specific dynamic loads were also defined (Reference 10), which were accepted by the NRC in Reference 11. The evaluation of the LOCA containment dynamic loads is based primarily on the results of the short-term analysis described in Section 4.1.1.3. The MSRV discharge load evaluation is based on 30 psi increases in the MSRV opening setpoints at EPU conditions.

4.1.1 Containment Pressure and Temperature Response Short-term and long-term containment analyses results are reported in the UFSAR. The short-term analysis is directed primarily at determining the drywell pressure response during the initial blowdown of the reactor vessel inventory to the containment following a large break inside the drywell. The long-term analysis is directed primarily at the suppression pool temperature 4-1

NEDO-33101 response, considering the decay heat addition to the suppression pool. The effects of EPU on the events yielding the limiting containment pressure and temperature responses are provided below.

4.1.1.1 Long-Term Suppression Pool Temperature Response Short-term and long-term containment analysis results are reported in the UFSAR. The long-

.term analysis is directed primarily at the pool temperature response, considering the decay heat addition to the pool.

(a) Bulk Pool Temperature The long-term bulk pool temperature response with EPU was evaluated for the DBA LOCA.

The analysis was performed at 102% of EPU RTP. Table 4-1 compares the calculated peak values for LOCA bulk pool temperature. The current analyses have been performed using the same RHR containment cooling capability based on K =223 BTU/sec-0 F/HX and a service water temperature of 95 0F. The EPU analysis was performed using a realistic decay heat model (ANS/ANSI 5.1 with 2cs uncertainty). Benchmark calculations were made as requested by the NRC in Reference 8. The Browns Ferry calculated peak bulk suppression pool temperatures are provided in Table 4-1 for both 102% of OLTP and 102% of EPU RTP.

Based on the analysis and limit values shown in Table 4-1, the peak bulk pool temperature with EPU is acceptable from a structural design standpoint.

The containment response used for NPSH evaluations is calculated using Browns Ferry inputs to maximize suppression pool temperature and minimize containment pressure, similar to the DBA-LOCA analysis using the same methodology. The suppression pool temperature and corresponding wetwell pressure for the short-term and long-term NPSH containment analyses are used in the evaluation of the available NPSH for the CS and the RHR pumps. The results of that evaluation are provided in Section 4.2.5.

(b) Local Pool Temperature with MSRV Discharge The local pool temperature limit for MSRV discharge is specified in NUREG-0783, because of concerns resulting from unstable condensation observed at high pool temperatures in plants without quenchers. The MSRV discharge quenchers at Browns Ferry are slightly below the elevation of the ECCS suction line penetration. The peak local suppression pool temperature at Browns Ferry has been evaluated for EPU and meets the NUREG-0783 criteria. Therefore, the peak local suppression pool temperature at Browns Ferry is acceptable for EPU conditions.

However, it is necessary to ensure that steam ingestion in the ECCS suction line is not of concern during steam MSRV discharge at high suppression pool temperature, because the top of the ECCS suction strainers at Browns Ferry are located above the T-quenchers. Per Reference 12, TVA addressed ECCS suction separation. TVA evaluated the physical configuration of the suppression pool, MSRV T-Quenchers, and ECCS suction strainers utilizing the information contained in NEDO-30832 (Reference 13), the NRC SER and the associated Brookhaven report.

Based on this evaluation, the ECCS suction piping would not ingest steam bubbles that could later collapse and induce water hammer loads. These conclusions remain valid for the EPU conditions.

4-2

NEDO-33101 4.1.1.2 Short-Term Gas Temperature Response The drywell airspace temperature limit is specified in Table 4-1. This limit is based on a bounding analysis of the superheated gas temperature reached during the steam blowdown to the drywell during a LOCA. The changes in the reactor vessel conditions at EPU increase the expected peak drywell gas temperature following a LOCA by -l IF. Therefore, the drywell gas temperature response with EPU does not exceed the limit.

Short-term containment response analyses for DBA-LOCA demonstrate that operation at EPU RTP does not result in exceeding the containment design limits. These analyses cover the blowdown period when the maximum drywell airspace temperature occurs. The analyses were performed at 102% of EPU RTP, using the methods reviewed and accepted by the NRC during the Mark I Containment LTTIP. The calculated peak drywell airspace temperatures are provided in Table 4-1. The total time that the drywell airspace temperature exceeds the containment structural design basis temperature of 281'F is less than one minute. This short duration is not sufficient for the average shell temperature to exceed the containment structural designi temperature.

The wetwell gas space peak temperature response is calculated assuming thermal equilibrium between the pool and wetwell gas space for the short-term containment response. The wetwell gas space increases by the same amount as the bulk pool temperature. The short-term wetwell gas space temperatures, due the EPU conditions, remain below the suppression chamber design temperatures. Therefore, the short-term wetwell gas temperature responses for EPU are acceptable.

4.1.1.3 Short-Term Containment Pressure Response Short-term containment response analyses were performed for the limiting DBA LOCA, which assumes a double-ended guillotine break of a recirculation suction line, to demonstrate that EPU does not result in exceeding the containment design limits. The short-term analysis covers the blowdown period during which the maximum drywell pressures and differential pressures between the drywell and wetwell occur. These analyses were performed at 102% of EPU RTP, using methods reviewed and accepted by the NRC during the Mark I Containment LTTIP with the break flow calculated using a more detailed RPV model (Reference 5) previously approved by the NRC. The results of these short-term analyses are summarized in Table 4-1 for comparison to the drywell design pressure. As shown by these results, the maximum drywell pressure values at the EPU conditions are bounded by the UFSAR analysis value and by the design pressure.

4.1.2 Containment Dynamic Loads 4.1.2.1 Loss-of-Coolant Accident Loads The LOCA containment dynamic loads analysis for EPU is based primarily on the short-term LQCA analyses. These analyses were performed as described in Section 4.1.1.3, using the Mark I Containment LTTIP method, except that the break flow was calculated using a more detailed RPV model (Reference 5). The application of this model to EPU containment evaluations is identified in ELTRI. These analyses provide calculated values for the controlling parameters for the dynamic loads throughout the blowdown. The key parameters are drywell 4-3

NEDO-33101 and wetwell pressure, vent flow rates and suppression pool temperature. The LOCA dynamic loads for EPU include pool swell, CO, and chugging loads. For Mark I plants like the Browns Ferry units, the vent thrust loads are also evaluated.

The short-term containment response conditions with EPU are within the range of test conditions used to define the pool swell and CO loads for Browns Ferry. The peak drywell pressure from these analyses is given in Table 4-1. The long-term response conditions at EPU conditions when chugging would occur are within the conditions used to define the chugging loads. The vent thrust loads at EPU conditions are calculated to be less than the plant-specific values calculated during the Mark I Containment LTTIP. Therefore, the LOCA dynamic loads are not affected by EPU.

4.1.2.2 Main Steam Relief Valve Loads The MSRV air-clearing loads include MSRVDL loads, suppression pool boundary pressure loads and drag loads on submerged structures. These loads are influenced by MSRV opening setpoint pressure, the initial water leg in the MSRVDL, MSRVDL geometry, and suppression pool geometry. For first MSRV actuations, following an event involving RPV pressurization, the controlling parametric change introduced by power uprate, which can affect the MSRV loads, is the increase in MSRV opening setpoint pressure. An increase in the MSRV opening setpoint pressure results in higher MSRV flow rates, and therefore, higher MSRV loads. An additional parametric change with EPU, which is specific to MSRV loads for subsequent actuations (second pops), is the time between MSRV actuations. EPU may result in changes in the time between MSRV actuations, which can affect the MSRVDL water level at the time of subsequent actuations. A higher water level at the time of a second pop results in higher MSRV loads.

The MSRV analytical limits for setpoints with EPU are shown in Table 5-1, which shows a 30 psi increase in the MSRV analytical limits. The increased MSRV loads resulting from this increase in the setpoint pressures were compared with plant unique design limits calculated during the Mark I Containment LTTIP (Reference 10). The comparison shows there is sufficient conservatism in the pre-EPU containment MSRV load definition to accommodate the increased MSRV loads due to EPU. Therefore, EPU does not affect the first actuation MSRV load definitions.

Subsequent actuations loads may be affected by changes in the MSRV discharge line water level in addition to the increase in the loads due to the pressure setpoint change. The water level in the MSRV discharge line is a function of the MSRVDL reflood response and the time between subsequent actuations. The MSRVDL reflood height response following a subsequent actuation is a function of the vacuum breaker size and MSRVDL geometry, which are unchanged with EPU. The increased MSRV loads resulting from subsequent actuations were compared with plant unique design limits calculated during the Mark I Containment LTTIP (Reference 10). The comparison shows there is sufficient conservatism in the pre-EPU containment MSRV load definition to accommodate the increased MSRV loads due to subsequent actuations. Therefore, EPU does not affect the subsequent actuations MSRV load definitions.

4-4

NEDO-33101 4.1.2.3 Subcompartment Pressurization The annulus pressure load on the biological shield wall due to a postulated break in a 4-inch jet pump instrument line nozzle is evaluated at EPU conditions. The annulus pressure load (2.4 psid) evaluated in UFSAR Section 12.2.2.6 remains bounding compared to the 102% of EPU annulus pressure load of 2.3 psid for normal FW temperatures. For FFWTR at 102% of EPU conditions, the annulus pressure load is 2.6 psid. The biological shield wall and component designs remain adequate, because there is substantial margin to the structural design value of 19 psid.

4.1.3 Containment Isolation The system designs for containment isolation are not affected by EPU. The capabilities of isolation actuation devices to perform during normal operations (including the 30 psi operating pressure increase) and under post-accident conditions have been determined to be acceptable.

Therefore, the Browns Ferry containment isolation capabilities are not adversely affected by the EPU.

The AOV and SOV parameters (temperature, pressure, flow) were reviewed, and no changes to the functional requirements of any AOV/SOV were identified as a result of operating at EPU conditions.

Operation at EPU conditions is within the pressure and temperature capabilities of the AOVs and SOVs. Therefore, the AOVs and SOVs remain capable of performing their design basis function.

4.1.4 Generic Letter 89-10 Program The MOV process parameters (temperature, pressure, flow) were reviewed and no significant changes to the functional requirements of the GL 89-10 MOVs were identified as a result of operating at EPU conditions.

Operation at the EPU conditions increases post-accident room temperatures (< 100 F) where the MOVs are located. Operation at the increased EPU conditions is within the pressure and ambient temperature capability of the GL 89-10 MOVs. The GL 89-10 MOVs will be capable of performing their design basis functions.

4.1.5 Generic Letter 89-16 In response to Generic Letter 89-16, Browns Ferry installed a HWWV system. The current design of the HWWV was based on 1.05% of 3293 MWt (OLTP). Therefore, at the EPU RTP conditions, the existing HWWV exhausts a smaller percentage of RTP. Based on the design, the HWWV would exhaust approximately 0.88% RTP at 3952 MWt (EPU RTP) and is designed to be operational during a SBO.

The primary objective of the HWWV is to preclude primary containment failure due to overpressurization, given a loss of decay heat removal (TW sequence) event. Using the ANSI/ANS-5.1-1979 decay heat (nominal) curve, 0.88% RTP is reached at approximately 5.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. From EPU condition, the containment pressure at 5.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is 46.4 psig, which is below the containment design pressure and primary containment pressure limit of 56 psig. At the EPU conditions, decay heat will be below the relieving capacity of the HWWV before containment 4-5

NEDO-33101 pressure reaches the design pressure limit, therefore, the existing HWNWV meets the intent of Generic Letter 89-16 for EPU conditions.

4.1.6 Generic Letter 95-07 MOVs used as containment or HELB isolation valves have been reviewed for the effects of operations at EPU conditions, including thermal binding and pressure locking (Generic Letter 95-07). The operability of MOVs is documented as part of the Browns Ferry GL 89-10 program.

4.1.7 Generic Letter 96-06 The Browns Ferry evaluations for Generic Letter 96-06, "Assurance of Equipment Operability and ContainmentIntegrity DuringDesign-BasisAccident Conditions," were accomplished using the peak drywell temperature (3360 F) for a MSLB inside containment. The equipment and containment remain within their design allowables for EPU conditions.

4.2 EMERGENCY CORE COOLING SYSTEMS Each ECCS is discussed in the following subsections. The effect on the functional capability of each system due to EPU is addressed. The ECCS performance evaluation is contained in Section 4.3.

4.2.1 High Pressure Coolant Injection System The HPCI system is designed to pump water into the reactor vessel over a wide range of operating pressures. The primary purpose of the HPCI is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immediately depressurize the reactor vessel. In this event, the HPCI system maintains reactor water level and helps depressurize the reactor vessel. The adequacy of the HPCI system is demonstrated in Section 4.3.

)) The conclusions are that for EPU conditions the pump and turbine remain within their allowable operating envelopes, the HPCI system is capable of delivering its design injection flow rate, and the turbine has the capacity to develop the required horsepower and speed.

4-6

NEDO-33101

))

The HPCI surveillance test range pressure is (in part) based on the maximum normal reactor dome pressure. Because the maximum normal reactor dome pressure increases by 30 psi, the HPCI surveillance test range also should be increased by 30 psi. Therefore, the surveillance test range is increased from

• 1010 psig and 2 920 psig to < 1040 psig and > 950 psig, consistent with the 30 psi increase to the nominal reactor operating pressure.

4.2.2 Low Pressure Coolant Injection The LPCI mode of the RHR system is automatically initiated in the event of a LOCA. When operating in conjunction with other ECCS, the LPCI mode provides adequate core cooling for LOCA events.

The increase in decay heat due to EPU could increase the calculated PCT following a postulated LOCA by a small amount. The ECCS performance evaluation presented in Section 4.3 demonstrates that the existing LPCI mode performance capability, in conjunction with the other ECCS, is adequate to meet the post-LOCA core cooling requirement for EPU RTP conditions.

((1 4.2.3 Core Spray System The CS system is automatically initiated in the event of a LOCA. When operating in conjunction with other ECCS, the CS system provides adequate core cooling for LOCA events.

There is no change in the reactor pressures at which the CS system is required to operate.

The increase in decay heat due to EPU could increase the calculated PCT following a postulated LOCA by a small amount. The ECCS performance evaluation presented in Section 4.3 demonstrates that the existing CS system performance capability, in conjunction with the other ECCS as required, is adequate to meet the post-LOCA core cooling requirement for the EPU conditions. ((

4.2.4 Automatic Depressurization System The ADS evaluation scope is provided in Section 5.6.8 of ELTRI.

The ADS uses MSRVs to reduce reactor pressure following a small break LOCA, when it is assumed that the high pressure ECCS has failed. This function allows LPCI and CS to inject coolant into the vessel. Plant design requires a minimum flow capacity for the MSRVs, and that ADS initiates following confirmatory signals and associated time delay(s). The required flow capacity and ability to initiate ADS on appropriate signals are not affected by EPU. The ADS 4-7

NEDO-33101 initiation logic and ADS valve control (( are adequate for EPU conditions.

4.2.5 ECCS Net Positive Suction Head Following a LOCA, the RHR and CS pumps operate to provide the required core and containment cooling. Adequate margin (NPSH available minus NPSH required) is required during this period to ensure the essential pump operation. The limiting NPSH conditions occur during either short-term or long-term post-LOCA pump operation and depend on the total pump flow rates, debris loading on the suction strainers, and suppression pool temperature.

TVA, previously requested containment overpressure credit for Browns Ferry Units 2 and 3 (Reference 20). In that submittal, TVA indicated that the need for containment overpressure in the short term was based on RHR requirements, and in the long term was based on CS requirements. The pre-EPU analysis indicates that up to 3 psi of overpressure credit (considering whole number value) is required for the short-term case for RHR pump operation to maintain adequate NPSH. One (1) psi of overpressure credit is required pre-EPU for the long-term case for CS pump adequate NPSH.

For the pre-EPU and the EPU analyses, the assumptions used maximized suppression pool temperature and minimized containment pressure. EPU RTP operation increases the reactor decay heat, which increases the heat addition to the suppression pool following a LOCA.

Therefore, changes in vapor pressure corresponding to the increase in suppression pool temperatures affect the NPSH margin. After 10 minutes, operation of the RHR pumps for containment cooling in the containment spray mode with continued operation of a CS loop for ECCS injection is also assumed.

The NPSH margins were calculated based on conservatively assuming RHR maximum flow rates and CS design flow rates during the short-term, and RHR and CS design flow rates during the long-term. The system flow rates for the short-term case are 42,000 gpm total RHR flow and 12,500 gpm total CS flow. The system flow rates for the long-term case are 13,000 gpm total RHR flow and 6,250 gpm total CS flow. The methodology used to determine the amount of debris generated and transported to the ECCS strainers is generally based on NEDO-32686, the BWROG Utility Resolution Guidance for ECCS Suction Strainer Blockage. The minimum quantity of paint chips recommended by this guidance is 85 lbs. BFN has identified a maximum surface area allowable of 157ft2 for unqualified coatings within the primary containment, which represents an additional 18 lbs of debris; therefore, 103 lbs of paint debris were assumed in sizing the strainers. This quantity did not change with EPU. Because the ECCS pump flow rates were unchanged for EPU, strainer approach velocities were not affected. Therefore, the debris loading on the suction strainers for EPU is the same as the pre-EPU condition. The assumptions used in the Browns Ferry Unit 1 ECCS NPSH calculations for friction loss, static head, strainer loss, flow, and NPSH required have not changed since the responses to NRC GL 97-04 (Reference 15) and NRC Bulletin 96-03 (Reference 21).

The short-term EPU NPSH analysis (0 to 600 seconds) indicates that with a containment overpressure (suppression chamber air space pressure) credit of 3 psi the RHR pumps have adequate NPSH margin. The short-term analysis also indicates that greater than 3 psi of overpressure is available from the beginning of the event until approximately 350 seconds. From 350 seconds to 600 seconds, the short-term analysis (using inputs that conservatively maximized 4-8

NEDO-33101 suppression pool temperature and minimized containment pressure) indicates an available overpressure of less than 3 psig. For the brief time that the short-term analysis indicates that less than 3 psi is available, the RHR pumps only require 2.5 psi. In addition, historical plant testing has demonstrated that the RHR pumps are capable of operating for short periods of time at NPSH values less than (approximately 9 feet) the manufacturer's required NPSH without degradation or substantial loss of flow. Therefore, RHR pump operation is not adversely affected by containment pressure less than 3 psi. This was previously presented for pre-EPU conditions and approved by the NRC in Reference 16. In the SER accompanying Reference 16, the NRC stated that "the use of 3 psi of containment overpressure above the initial airspace pressure is acceptable for the first 10 minutes after a LOCA." Reference 16 also concludes that CS pump operation is not affected by this lower containment overpressure during the short term.

The long-term EPU NPSH analysis (0 until the end of the event) indicates that up to 2 psi (considering whole number value) containment overpressure credit is required when the suppression pool temperature exceeds 18 1F to obtain adequate NPSH margin for the long-term operation of the CS pumps. This is an increase from the 1 psi of overpressure credit currently approved for pre-EPU conditions. The long-term analysis demonstrates that greater than 4 psi of containment overpressure is available during this period.

Tables 4-2 and 4-3 provide the results of the short-term and long-term containment response.

Table 4-4 provides the suppression pool temperature and required containment overpressure required to maintain NPSH margins during the DBA LOCA for EPU conditions.

Based on the above, Browns Ferry is requesting approval of 3 psi of overpressure credit to meet both the short-term and long-term NPSH requirements. A single containment overpressure credit value is requested both to account for potential future contingencies and to provide consistency between the inputs to the short- and long-term analyses. Other means to increase the NPSH margin were found to not be feasible.

One RHR pump is required to operate during either the SBO or an Appendix R fire event. EPU RTP operation increases the reactor decay heat, which increases the heat addition to the suppression pool following these events (see Sections 6.7.1 and 9.3.2). As a result, the long-term peak suppression pool water temperature and peak containment pressure increase. The NPSH evaluation at these peak pool temperatures shows adequate NPSH margins during the SBO and the Appendix R events with containment overpressures of 1 psi and 10 psi, respectively.

The HPCI system primary function is to provide reactor inventory makeup water and assist in depressurizing the reactor during an intermediate or small break LOCA. The HPCI system can operate with suction from the suppression pool at a temperature below 140'F during the first 10 minutes after initiation of the event. EPU has an insignificant effect on the time for the suppression pool temperature to reach 140'F. If the HPCI pump operates beyond the first 10 minutes following the event, the reactor operator may terminate HPCI pump operation when the suppression pool temperature reaches 140'F. The HPCI pump NPSH margin remains adequate as long as the suppression pool temperature does not exceed 140'F during HPCI operation.

HPCI system operation is credited during ATWS, Appendix R, and SBO events. The suppression pool temperature does not affect the NPSH margin, because the HPCI pump takes suction from the CST during these events.

4-9

NEDO-33101 4.3 EMERGENCY CORE COOLING SYSTEM PERFORMANCE The Browns Ferry ECCS for each unit is designed to provide protection against postulated LOCAs caused by ruptures in the primary system piping, and the ECCS performance characteristics do not change for EPU. The ECCS-LOCA performance analysis demonstrates that 10 CFR 50.46 requirements continue to be met following EPU operating conditions.

((i The EPU effect on PCT for small recirculation line breaks is larger than the EPU effect on PCT for large line breaks. The increased decay heat associated with EPU results in a longer ADS blowdown time leading to a later ECCS system injection and a higher PCT for the small break LOCA. As a result, the limiting LOCA case that defines the Browns Ferry Licensing Basis PCT at EPU for GE14 fuel is a small recirculation discharge line break with Battery failure.

The effects on compliance with the other acceptance criteria of 10 CFR 50.46 for the limiting large and small breaks are evaluated for the Browns Ferry EPU. For power uprates, there is a negligible effect on compliance with the other acceptance criteria of 10 CFR 50.46 (local cladding oxidation, core-wide metal-water reaction, coolable geometry and long-term cooling).

The local cladding oxidation and core-wide metal-water reaction were calculated and determined to be within the 10 CFR 50.46 acceptance criteria. Coolable geometry and long-term cooling have been dispositioned generically for BWRs. These generic dispositions are not affected by EPU.

The Licensing Basis PCT is determined based on the calculated Appendix K PCT at rated core flow with an adder to account for uncertainties. For the EPU, the GE13 Licensing Basis PCT is 17800 F at rated core flow. The comparable GE13 Licensing Basis PCT for the 105% OLTP conditions is 18101F at rated core flow. For the EPU, the GE14 Licensing Basis PCT is 1830'F at rated core flow. The comparable GE14 Licensing Basis PCT for the 105% OLTP conditions is 1760'F at rated core flow. At EPU conditions, the limiting break size is the large break for GE13 and the 0.06 R2 small break for GE14. The results of these analyses are provided in Table 4-5. The changes in PCT are small when compared to the PCT margin to the 10 CFR 50.46 licensing limit of 22000 F.

Reference 17 provides justification for the elimination of the 1600 0 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 in Reference 18 accepted this position by noting that since plant-specific Upper Bound PCT calculations have been performed for all plants, other means may be used to demonstrate compliance with the original SER limitations. 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. The changes in magnitude of the PCT due to EPU demonstrate that this plant configuration does not invalidate the existing Upper Bound PCT calculation. After the implementation of EPU, the Licensing Basis PCT will 4-10

NEDO-33101 continue to bound the Upper Bound PCT. Therefore, the Licensing Basis PCT is sufficiently conservative.

For SLO, a multiplier is applied to the Two-Loop Operation PLHGR and MAPLHGR limits.

This multiplier ensures that the Two-loop Upper Bound PCT is also bounding for the SLO case.

The SLO PCT values are well below the 2200'F limit.

4.4 MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM The CREVS processes outside air needed to provide ventilation and pressurization of the CRHZ during accident conditions. The CREVS units are started and the CRHZ is isolated on receipt of a primary containment isolation signal or high radiation signal in the Control Building intake duct. When the CRHZ is isolated, a fixed amount of outside air is filtered.

TVA has submitted a request for an amendment to the plant-operating license that supports the full scope implementation of an AST for Units 1, 2 and 3 (Reference 19). The TVA request includes the radiological dose consequences for the design bases accidents and includes the CREVS operational parameters following EPU conditions.

4.5 STANDBY GAS TREATMENT SYSTEM The SGTS is designed to maintain secondary containment at a negative pressure and to filter the exhaust air for removal of fission products potentially present during abnormal conditions. By limiting the release of airborne particulates and halogens, the SGTS limits Control Room and off-site dose following a postulated design basis accident.

TVA has submitted a request for an amendment to the plant-operating license that supports the full scope implementation of an AST for Units 1, 2 and 3 (Reference 19). The TVA request includes the radiological dose consequences for the design bases accidents and includes the SGTS operational parameters following EPU conditions.

4.6 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM Browns Ferry does not use a Main Steam Isolation Valve Leakage Control System (MSIV-LCS).

4.7 POST-LOCA COMBUSTIBLE GAS CONTROL The Combustible Gas Control System is designed to maintain the post-LOCA concentration of oxygen or hydrogen in the containment atmosphere below the lower flammability limit.

As a result of EPU, the post-LOCA production of hydrogen and oxygen by radiolysis increases proportionally with power level. This increase in radiolysis has an effect on the time available to start the system before reaching procedurally controlled limits, but does not affect the ability of the system to maintain oxygen below the lower flammability limit of 5% by volume as specified in Safety Guide 7. The required start time for the CAD system decreases from 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> to 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />, as a result of EPU. This reduction in required CAD initiation time does not affect the ability of the operators to respond to the postulated LOCA. The integrated hydrogen production rates from radiolysis and metal-water reaction are shown in Figure 4-1. Uncontrolled hydrogen and oxygen concentrations in the drywell and wetwell are shown in Figure 4-2, and the Drywell Pressure Response to CAD operation without venting is shown in Figure 4-3.

4-11

NEDO-33101 The TS requires sufficient on-site storage of nitrogen in each of the two 4000-gallon storage tanks to maintain containment oxygen below 5% during the 7-day period following the postulated LOCA. As a result of increased production rate of radiolytic gas following EPU operation, the required 7-day volume of nitrogen increases to 197,000 scf, which exceeds the available 191,000 scf supply required by the Unit I TS. An evaluation was performed to determine the amount needed to maintain a 4-day supply following the postulated LOCA. This resulted in a nitrogen volume of 104,834 scf that is less than the available 191,000 scf supply required by the TS. The TS Bases liquid nitrogen 7-day requirement is conservative, because additional liquid nitrogen can be delivered within one day or less. Two liquid nitrogen distribution facilities are located within 1-day travel distance from Browns Ferry. Each facility is capable of delivering 5000 gallons or more of liquid nitrogen to Browns Ferry with less than 4 days notice. The historical average delivery time is 1 day. The TS are not changed, however, the TS Bases will be revised to a 4-day nitrogen storage requirement to accommodate EPU operations. The CAD System Nitrogen Volume Requirements are shown in Figure 4-4.

4.8 REFERENCES

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

2. GE Nuclear Energy, "The GE Pressure Suppression Containment System Analytical Model," NEDM-10320, March 1971.

3. GE Nuclear Energy, "The General Electric Mark III Pressure Suppression Containment System Analytical Model," NEDO-20533, June 1974.

4. GE Nuclear Energy, "Maximum Discharge of Liquid-Vapor Mixtures from Vessels,"

NEDO-21052, September 1975.

5. GE Nuclear Energy, "General Electric Model for LOCA Analysis In Accordance With 10 CFR 50 Appendix K," NEDE-20566-P-A, September 1986.

6. NUREG-0800, U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 6.2.1.1.C, "Pressure - Suppression Type BWR Containments," Revision 6, August 1984.

7. NUREG-0661, "Mark I Containment Long-Term Program Safety Evaluation Report," July 1980.

8. Letter to Gary L. Sozzi (GE) from Ashok Thadani (NRC) on the Use of the SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis, July 13, 1993.

9. GE Nuclear Energy, "Mark I Containment Program Load Definition Report," NEDO-21888, Revision 2, November 1981.

10. BFN Report CEB-83-34, "Browns Ferry Nuclear Plant Torus Integrity Long-Term Program Plant Unique Analysis Report," Rev. 2, December 10, 1984.

4-12

NEDO-33101

11. Letter from USNRC to H. G. Paris, TVA, entitled "MARK I CONTAINMENT PROGRAM - Browns Ferry Nuclear Plant Units 1, 2 and 3," May 6, 1985 (A02 850513 002).

12. TVA Letter, from T. E. Abney to USNRC, "Browns Ferry Nuclear Plant (BFN) - Unit 2 and 3, Corrected Information for Technical Specification Change Request TS-384, Power Uprate - (TAC NOS M99711 and M99712)," R08 991201 679, December 1, 1999.

13. GE Nuclear Energy, "Elimination of Limit on Local Suppression Pool Temperature for SRV Discharge with Quenchers," NEDO-30832, Class I, December 1984.

14. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," Licensing Topical Reports NEDC-32523P-A, Class III, February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement I Volume II, April 1999.

15. TVA Letter, "Browns Ferry Nuclear Plant (BFN) Unit 1 - Response to NRC Generic Letter (GL) 97-04, Assurance of Sufficient Net Positive Suction Head (NPSH) For Emergency Core Cooling And Containment Heat Removal Pumps," May 6, 2004.

16. NRC Letter, "Browns Ferry Nuclear Plants, Units 2 and 3 - Issuance of Amendments Regarding Crediting of Containment Overpressure for Net Positive Suction Head Calculations for Emergency Core Cooling Pumps (TAC Nos. MA3492 and MA3493),"

September 3, 1999.

17. GE Nuclear Energy, "GESTR-LOCA and SAFER Models For Evaluation of Loss-of-Coolant Accident, Additional Information For Upper Bound PCT Calculation," NEDE-23785P-A, Volume III, Supplement 1, Revision 1, March 2002.

18. NRC Letter, Stuart A. Richards (NRC) to James F. Klapproth (GENE), Review of NEDE-23785P, Vol. III, Supplement 1, Revision 1, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplement 1, Additional Information for Upper Bound PCT Calculation," (TAC No. MB2774), February 1, 2002.

19. TVA Letter, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - License Amendment

- Alternative Source Term," R08 020731 649, July 31, 2002, including Tech. Spec. No. 405 (TVA-BFN-TS-405).

20. TVA Letter, "Browns Ferry Nuclear Plant (BFN) - Units 2 And 3 - License Amendment Regarding Use of Containment Overpressure for Emergency Core Cooling System (ECCS)

Pump Net Positive Suction Head (NPSH) Analyses," September 4, 1998.

21. TVA Letter, "Browns Ferry Nuclear Plant (BFN) - NRC Bulletin No. 96-03, Potential Plugging of Emergency Core Cooling Suction (ECCS) Strainers by Debris in Boiling Water Reactors (TAC Nos. M96135, M96136, M96137)," July 25, 1997.

4-13

NEDO-33101 Table 4-1 Browns Ferry Containment Performance Analysis Results OLTP(I) EPU Limit Parameter (Historical) (Current Methods)

Peak Drywell Pressure (psig) 49.6 48.5(2) 56 Peak Drywell Temperature (oF)(3 ) 294 295.2(2) 340/281 Peak Bulk Pool Temperature (IF) 170 187.3(4, 5 281 Peak Wetwell Pressure (psig) *27 30.5 56 (1) Unit 1 UFSAR Section 14.11.3 values.

(2) LAMB mass and energy release data used as input to M3CPT.

(3) The acceptance limit for drywell airspace temperature is 340'F, while the shell design value is 281 0 F. The listed peak values are for airspace temperature.

(4) Uses ANS/ANSI 5.1 (+2a uncertainty) decay heat model.

(5) Service water temperature of 950F.

4-14

NEDO-33101 Table 4-2 Browns Ferry Short-Term Containment Input to NPSH Analysis Time After LOCA Suppression Pool Pressure Suppression Pool Temperature (sec) (psia) (OF) 0 14.4 95 54.34 36.73 125.9 101.31 37.98 136.4 151.47 29.67 139.0 201.84 22.73 143.1 304.94 17.95 148.3 351.75 17.23 149.9 399.94 16.92 151.2 500.87 16.81 153.4 600.12 16.81 155.4 Table 4-3 Browns Ferry Long-Term Containment Input to NPSH Analysis Time After LOCA Suppression Pool Pressure Suppression Pool Temperature (sec) (psia) OF) 0 14.4 95 99.63 38.45 141.0 197.82 36.34 142.6 297.76 34.35 143.7 408 31.00 146.2 607 24.43 152.8 4,134 19.90 175.8 7,105 20.69 181.9 14,682 20.99 186.6 37,426 20.05 181.9 50,180 19.27 176.7 4-15

NEDO-33101 Table 4-4 Browns Ferry EPU DBA LOCA NPSH Margins and Containment Overpressure Credit Time Suppression RHR CS After Pool Containment pump pump LOCA Temperature Overpressure NPSH NPSH Description/Basis (sec) (O) Required (psi) margin margin

_ _ _ __ __ _ _ __ _ _ _ _ _ _ (fi) (fi)

Short-term analysis.

600 155.4 2.46 0 6.75 Overpressure required to meet RHR NPSH I3 requirements 601 152.4 0 12.95 6.55 Long-term analysis Greater than 0 psi of 4,150 175.83 0 6.32 0 overpressure required 4__150__ 175__83_ _ - _ for long-term for CS pumps Greater than I psi of 7,090 181.85 1 6.32 0 overpressure required for long-term for CS pumps 14,700 186.6 1.90 6.32 0 Peak Suppression Pool temperature Less than 1psi of 37,500 181.85 1 6.32 0 overpressure required for long-term for CS pumps 4-16

NEDO-33101 Table 4-5 Browns Ferry ECCS Performance Analysis Results Parameter 105% OLTP EPU 10 CFR 50.46 Limit Method SAFER/GESTR SAFERIGESTR Power 105% OLTP 120% OLTP

1. Licensing Basis < 1810 (GE13) < 1780 (GE13) < 2200 PCT (TF) < 1760 (GE14)(') < 1830 (GE14)

2. Cladding <2.0 <3.0 < 17 Oxidation (%

Original Clad Thickness)

3. Hydrogen < 0.1 < 0.1 < 1.0 Generation

(% core wide metal-water reaction)

4. Coolable OK OK Meet 1 and 2, above Geometry

5. Core Long-Term OK OK Core flooded to TAF or Cooling Core flooded to jet pump suction elevation and at least one core spray system is operating at rated flow.

(1) An update of the Licensing Basis PCT at 105% OLTP was calculated for the EPU analysis.

This allows for comparison with the EPU Licensing Basis PCT results.

4-17

NEDO-33101 en as E 15J Z

.0 c

E In To a,

cm 0 a* 0 0

a) 10 a) o _-

QO1 Qi 1 10 100 Tirm ftOC(da)

Figure 4-1. Browns Ferry Time-integrated Containment Hydrogen Generation 4-18

NEDO-33101 30 c)

E 0

0 cz C

C)

C 0

C.

0.01 0.1 1 10 100 Time After LOCA (days)

Figure 4-2. Browns Ferry Uncontrolled Hz and O2 Concentrations in Dryivell and Wetwell 4-19

NEDO-33101 60 50 In 0.

40 E

(n en 30 E

cC)

'U 20 10 O _~

0.01 0.1 1 10 100 Time After LOCA (days)

Figure 4-3. Browns Ferry Drywell Pressure Response to CAD Operation Without Venting 4-20

NEDO-33101 1,

In c

Y.1.

0

-I-E C.

a 0

• 0 0

• 0 lat 0

10 c

O L.

z 0 20 40 60 80 100 Time After LOCA (days)

Figure 4-4. Browns Ferry CAD System Nitrogen Volume Requirement 4-21

NEDO-33101

5. INSTRUMENTATION AND CONTROL The safety-related and major (non-safety) process monitoring instruments, controls and trips (analytical limits for setpoints) that could be affected by the EPU are addressed below.

The following evaluations are based on the NRC approved guidelines in Appendix F of ELTRI (Reference 1).

5.1 NSSS MONITORING AND CONTROL SYSTEMS The instruments and controls that directly interact with or control the reactor are usually considered within the NSSS. The NSSS process variables, instrument setpoints and Regulatory Guide 1.97 instrumentation that could be affected by the EPU were evaluated. As part of the EPU implementation, the NRC approved TVA setpoint methodology (Reference 2) is used to generate the allowable values and (nominal trip) setpoints related to the analytical limit changes shown in Table 5-1.

The following summarizes the results of the NSSS evaluations.

5.1.1 Control Systems Evaluation Changes in process variables and their effects on instrument setpoints were evaluated for the EPU operation to determine any related changes. Process variable changes are implemented through changes in plant procedures.

TS instrument AVs and/or setpoints are those sensed variables, which initiate protective actions.

The determination of instrument AVs and setpoints is based on plant operating experience and the conservative ALs used in specific licensing safety analyses. The settings are selected with sufficient margin to preclude inadvertent initiation of the protective action, while assuring that adequate operating margin is maintained between the system settings and the actual limits.

Increases in the core thermal power, pressure and steam flow affect some instrument setpoints, as described in Section 5.3. These setpoints were adjusted to maintain comparable differences between system settings and actual limits, and were reviewed to assure that adequate operational flexibility and necessary safety functions are maintained at the EPU RTP level.

5.1.2 Neutron Monitoring System The APRM power signals are rescaled to the EPU RTP level, such that the indications read 100% at the new licensed power level.

EPU implementation has little effect on the IRM overlap with the SRMs and the APRMs. Using normal plant surveillance procedures, the IRMs may be adjusted, as required, so that overlap with the SRMs and APRMs remains adequate. No change is needed in the APRM downscale setting.

At EPU RTP, the average flux experienced by the detectors increases due to the average power increase in the core. The maximum flux experienced by an LPRM remains approximately the same because the peak bundle powers do not appreciably increase. Due to the increase in neutron flux experienced by the LPRMs and TIPs, the neutronic life of the LPRM detectors may be reduced and radiation levels of the TIPs may be increased. LPRMs are designed as replaceable components. The LPRM accuracy at the increased flux is within specified limits, 5-1

NEDO-33101 and LPRM lifetime is an operational consideration that is handled by routine replacement. TIPs are stored in shielded rooms. A small increase in radiation levels is accommodated by the radiation protection program for normal plant operation.

The increase in power level at the same APRM reference level results in increased flux at the LPRMs that are used as inputs to the RBM. The RBM instrumentation is 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 the higher average local flux. The change in performance does not have a significant effect on the overall RBM performance.

The Neutron Monitoring Systems installed at Browns Ferry are in accordance with the requirements established by the GE design specifications.

5.1.3 Rod Worth Minimizer The RWM is a normal operating system that does not perform a safety-related function. The function of the RWM is to support the operator by enforcing rod patterns, until reactor power has reached appropriate levels. ((

']The power-dependent instrument setpoints for the RWM are included in the TS (see Section 5.3.12).

))

5.2 BOP MONITORING AND CONTROL SYSTEMS Operation of Browns Ferry at the EPU RTP level has minimal effect on the BOP system instrumentation and control devices. Based on the EPU operating conditions for the power conversion and auxiliary systems, most process control values and instrumentation have sufficient range/adjustment capability for use at the expected EPU conditions. However, some (non-safety) modifications may be needed to the power conversion systems to obtain full EPU RTP.

5.2.1 Pressure Control System The PCS is a normal operating system that provides fast and stable responses to system disturbances related to steam pressure and flow changes so that reactor pressure is controlled within its normal operating range. This system does not perform a safety function. Pressure control operational testing is included in the EPU implementation plan as described in Section 10.4 to ensure that adequate turbine control valve pressure control and flow margin is available.

(( ))

5.2.1.1 EHC Turbine Control System The turbine EHC system was reviewed for the increase in core thermal power and the associated increase in rated steam flow. For EPU conditions, a second steam line resonance compensator card is installed to attenuate third harmonic resonance. In addition, new diode function generator calibration curves and control valve hydraulic accumulators are installed. The control systems are expected to perform normally for EPU RTP operation.

5-2

NEDO-33101 No modification to the turbine control valves or the turbine bypass valves is required for operation at the EPU throttle conditions. However, if needed, the load and load set meters will be replaced and their starting point redlines will be revised. Normal manual operator controls are used in conjunction with the associated operating procedures. Confirmation testing will be performed during power ascension (see Section 10.4).

5.2.1.2 Turbine Steam Bypass System The Turbine Steam Bypass System is a normal operating system that is used to bypass excessive steam flow. The bypass flow capacity is included in some AOO evaluations (Section 9.1).

These evaluations demonstrate the adequacy of the bypass system. Some of the limiting events in the reload analyses take credit for the availability of the bypass system. The reload analyses are used to establish the core operating limits.

5.2.2 Feedwater Control System The FW control system controls reactor water level during normal operations. (The capacity of the FW pumps to adequately support EPU RTP operation is discussed in Section 7.4.) The minimum excess flow capacity requirement for adequate reactor water level control is approximately 5% of the operating point flow rate. The control signal range is capable of accessing as much of the flow as needed. Therefore, the capacity is sufficient for acceptable control.

The control system itself is adjusted to provide acceptable operating response on the basis of unit behavior. It will be set up to cover the current power range using startup and periodic testing.

An expansion of the steam flow signal range (part of the three-element control mode) is planned to ensure full control near the EPU RTP event with one MSIV closed. No changes in the operating water level or water level trip setpoints are required for the EPU; therefore, margin for trip avoidance is maintained. For EPU, the FW flow control system device settings have the sufficient adjustment ranges to ensure satisfactory operation. However, this will be confirmed by performing unit tests during the power ascension to the EPU conditions (Section 10.4). ((

1]

Failure of this system is evaluated in the reload analysis for each reload core with the FW controller failure-maximum demand event. A LOFW transient event can be caused by downscale failure of the controls. The LOFW event is discussed in Section 9.1.3.

5.2.3 Leak Detection System The instrument setpoints associated with primary system leak detection have been evaluated with respect to the slightly higher operating steam flow, steam pressure and FW temperature for EPU.

Each of the systems, where leak detection potentially could be affected by the EPU, is addressed below.

MS Valve Vault and Tunnel Temperature Based Leak Detection:

EPU increases the RWCU, FW and MS pressures and temperatures, which potentially affects the MSL tunnel and MSVV LDSs. The pipe break mass and energy calculations have been performed to establish ALs that support the current TS values. Thus the temperature based leak detection is not adversely affected and the ALs in the TS do not require a change.

5-3

NEDO-33101 RCIC System Temperature Based Leak Detection:

EPU increases the RCIC steam pressure and temperature. The pipe break mass and energy calculations have been performed to establish ALs that support the current TS values. Thus the RCIC temperature based leak detection 'is not adversely affected and the ALs in the TS do not require a change.

HPCI System Temperature Based Leak Detection:

EPU increases the HPCI steam pressure and temperature. The pipe break mass and energy calculations have been performed to establish ALs that support the current TS values. Thus the HPCI temperature based leak detection is not adversely affected and the ALs in the TS do not require a change.

Non-Temperature Based Leak Detection:

See Section 5.3.4, 5.3.15 and 5.3.16 for a discussion of the effects of EPU on the non-temperature based leak detection.

5.3 INSTRUMENT SETPOINTS TS instrument AVs and their associated NTSPs are provided for those sensed variables that initiate protective actions and are generally associated with the safety analysis. TS AVs are highly dependent on the results of the safety analysis. The safety analyses generally establishes the ALs. The determination of the TS AVs and the instrument NTSPs includes consideration of measurement uncertainties and is derived from the ALs. The settings are selected with sufficient margin to minimize inadvertent initiation of the protective action, while assuring that adequate operating margin is maintained between the system settings and the actual limits. There is margin in the safety analysis process that is considered in establishing the setpoint process used to establish the TS AVs and setpoints.

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

All TS instruments were evaluated for effects from EPU using the existing TVA setpoint methodology (Reference 2). This methodology is consistent with NRC Regulatory Guide 1.105, and has been previously reviewed by the NRC. This evaluation included a review of environmental (i.e., radiation and temperature) effects, process (i.e., measured parameter) effects and analytical (i.e., AL and margins) effects on the subject instruments.

The instrument function AL is the value used in the safety analyses to demonstrate acceptable nuclear safety system performance is maintained. The AV and NTSP are then chosen/calculated such that the instrument functions before reaching the AL under the worst-case environmental/event conditions. Instrument NTSPs account for measurable instrument characteristics (e.g., drift, accuracy, repeatability).

Table 5-1 summarizes the current and EPU ALs.

5-4

NEDO-33101

))

5.3.1 High-Pressure Scram During a pressure increase transient, that is not terminated by a direct scram or high neutron flux scram, the high-pressure scram terminates the event. The reactor vessel high-pressure scram signal settings are maintained slightly above the reactor vessel maximum normal operating pressure and below the specified AL. The setting permits normal operation without spurious scrams, yet provides adequate margin to the maximum allowable reactor vessel pressure.

((I 5.3.2 High-Pressure ATWS Recirculation Pump Trip The ATWS-RPT trips the pumps during plant transients associated with increases in reactor vessel dome pressure and/or low reactor water level. The ATWS-RPT is designed to provide negative reactivity by reducing core flow during the initial part of an ATWS. The ATWS-RPT high pressure setpoint is a significant factor in the analysis of the peak reactor vessel pressure from an ATWS event. The ATWS-RPT low reactor water level setpoint is not a significant factor for the limiting ATWS events, and thus, the low reactor water level setpoint is not affected by EPU.

The major consideration for the ATWS-RPT high pressure setpoint is an increase in the calculated peak vessel pressure during a hypothetical ATWS event, because of the higher initial power, dome pressure and MSRV setpoint values. For EPU, the high pressure setpoint was included in the ATWS evaluation discussed in Section 9.3.1. This evaluation concludes that the calculated peak vessel pressure remains below its allowable limit for an ATWS event.

Therefore, the high pressure ATWS-RPT setpoint is increased to account for the increased MSRV setpoints and operating dome pressure.

5.3.3 Main Steam Relief Valve Because there is an increase in reactor operating dome pressure, the MSRV ALs for setpoints need to be increased (see Table 5-1), as discussed in Section 3.1. This is consistent ((

)) The updated values were used in the overpressure protection and transient analyses discussed in Sections 3.2 and 9.1.

5.3.4 Main Steam High Flow Isolation The MSL high flow isolation is used to initiate the isolation of the Group 1 primary containment isolation valves. The only safety analysis event that credits this trip is the MSLBA. For this accident, there is a diverse trip from high area temperature. There is sufficient margin to choke fl6w, so the AL for EPU is maintained at the current 144 percent of rated steam flow in each MSL.

No new instrumentation is required because the existing instrumentation has the required upper range limit to re-span the instrument loops for the higher steam flow condition. A new setpoint 5-5

NEDO-33101 is calculated using the methodology noted in Section 5.3, and no TS change is required. This ensures that sufficient margin to the trip setpoint exists to allow for normal plant testing of the MSIVs and turbine stop and control valves. This approach is consistent with Section F.4.2.5 of ELTRI.

5.3.5 Neutron Monitoring System The AL for the APRM Neutron Flux Scram remains the same in terms of percent power, and thus, the percent power values for the TS AV and the NTSP do not change.

For DLO, the clamped AL, AV and NTSP retain the MELLLA domain values in percent RTP.

The SLO AL for the fixed (clamped) APRM scram is evaluated to be the same as for DLO.

A new nominal trip setpoint and AV are calculated for the APRM setdown using Browns Ferry current design basis methodology. This methodology is based on GE NEDC-31336, which has been evaluated and accepted by the NRC (Reference 3).

The severity of rod withdrawal error during power operation event is dependent upon the RBM rod block setpoint. This setpoint is only applicable to a control rod withdrawal error. ((

)) The flow biased RBM is clamped based on its power value at 100% core flow and 100% power. The RBM setpoints are based on the cycle-specific RWE transient analysis, and thus, are confirmed or revised (as needed) via the reload core design, review and approval process.

The ALs for the above trips are provided in Table 5-1 5.3.6 Main Steam Line High Radiation Scram Browns Ferry does not have a MSL radiation level scram.

53.7 Low Steam Line Pressure MSIV Closure (RUN Mode)

The purpose of this setpoint is to initiate MSIV closure on low steam line pressure when the reactor is in the RUN mode. This setpoint is not changed for EPU, as discussed in Section F.4.2.7 of ELTRI.

5.3.8 Reactor Water Level Instruments The reactor water level trip values used in the safety analyses do not require changing as a result of EPU.

5-6

NEDO-33101 5.3.9 Main Steam Tunnel High Temperature Isolation At EPU conditions, the increase in ambient temperature is not significant (< 1IF), and no change to the MSL Tunnel High Temperature Isolation setpoint is required.

5.3.10 Low Condenser Vacuum Browns Ferry does not have a low condenser vacuum MSIV isolation or scram trip.

5.3.11 TSV Closure and TCV Fast Closure Scram Bypass The TSV closure and TCV fast closure scram bypass allows these scrams to be bypassed, when reactor power is sufficiently low, such that the scram function is not needed to mitigate a T/G trip. This power level is the AL for determining the actual trip setpoint, which comes from the TFSP. The TFSP setpoint is chosen to allow operational margin so that scrams and recirculation pump trips can be avoided, by transferring steam to the turbine bypass system during T/G trips at low power.

Based on the guidelines in Section F.4.2.3 of ELTRI, the TSV Closure and TCV Fast Closure Scram Bypass AL is reduced (see Table 5-1). ((

)) The new AL is based on a reactor steam flow within approximately 1% of the original steam flow. Due to changes in the turbine, a new first stage pressure setpoint will be determined.

EPU results in an increased power level and the HPT modifications result in a change to the relationship of turbine first-stage pressure to reactor power level. The TFSP setpoint is used to reduce scrams and recirculation pump trips at low power levels where the turbine steam bypass system is effective for turbine trips and generator load rejections. In the safety analysis, this trip bypass only applies to events at low power levels that result in a turbine trip or load rejection.

Maintaining the AL at the same absolute power as for the current setpoint, maintains the same transient analysis basis and scram avoidance range of the bypass valves.

Because the HPT is modified to support achieving the uprated level, a new AL (in psig) corresponding to the same absolute power as the current AL is established. Therefore, a new setpoint is calculated using the methodology as noted in Section 5.3, and the TS applicable condition in percent RTP has been changed. The AV (in psig) for Browns Ferry will be revised prior to EPU implementation.

To ensure that the new value is appropriate, EPU plant ascension startup test or normal plant surveillance will be used to validate that the actual plant interlock is cleared consistent with the safety analysis 5.3.12 Rod Worth Minimizer The Rod Worth Minimizer LPSP is used to bypass the rod pattern constraints established for the control rod drop accident at greater than a pre-established low power level. The measurement parameters are FW and steam flow. ((

))

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NEDO-33101 5.3.13 Pressure Regulator The PCS is discussed within Section 5.2.1. The PCS provides the means by which the operating pressure setpoint of the reactor is established, provides for loading of the main turbine generator relative to reactor power, and provides for control of the main turbine bypass valves. The PCS controlling pressure signal is reactor pressure.

The reactor dome pressure is changed for EPU, and the increased steam flow results in a somewhat greater steam line pressure loss. Therefore, the pressure regulator operational setpoint must be adjusted to achieve the desired reactor pressure.

The small differences in tuning parameter values will be reconfirmed during the power ascension testing. Specific EHC and steam bypass control system tests will be performed during the initial EPU ascension phase, as summarized in Section 10.4.

53.14 Feedwater Flow Setpoint for Recirculation Cavitation Protection The current value of the FW flow setpoint remains unchanged in terms of actual FW flow rate, because the cavitation interlock requirement is not based on the percentage of rated flow.

However, the relative setpoint, as it appears on the power/flow map, is reduced slightly to account for EPU RTP. This is consistent with Section F.4.2.6 of ELTRI.

53.15 RCIC Steam Line High Flow Isolation For EPU, the AL (in % steam line flow) remains based on the maximum rated steam flow to the RCIC turbine. Due to the 30 psi increase in MSRV setpoints, the RCIC turbine speed and hence RCIC steam flow rate are increased. By maintaining the same AV (in differential pressure at the flow measuring device), the system will isolate at the same absolute steam flow as pre-EPU.

Therefore, there is no change required for the RCIC steam line high flow isolation AL and AV.

5.3.16 HPCI Steam Line High Flow Isolation For EPU, the AL (in % steam line flow) remains based on the maximum rated steam flow to the HPCI turbine. Due to the 30 psi increase in MSRV setpoints, the HPCI turbine speed and hence HPCI steam flow rate are increased. By maintaining the same AV (in differential pressure at the flow measuring device), the system will isolate at the same absolute steam flow as pre-EPU.

Therefore, there is no change required for the HPCI steam line high flow isolation AL and AV.

5.4 REFERENCES

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

2. TVA Branch Technical Instruction, EEB-TI-28, Setpoint Calculations, Revision 5, February 25, 2000.

3: Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report Instrumentation Setpoint Methodology General Electric Company NEDC-31336, Revision 1, November 6, 1995.

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NEDO-33101 Table 5-1 Browns Ferry Unit I Analytical Limits For Setpoints Analytical Limits Parameter Current J EPU APRM Calibration Basis (MWt) 3293 l 3952 APRM Simulated Thermal Power Scram DLO Fixed No change SLO Fixed No change DLO Flow Biased (%RTP) ( 0.66WD + 73% 0.55WD + 67.5%

6 (WD - AW)+ 73% .5 -AW)+

SLO Flow Biased (%RTP) ( X2)0.6 67.5%

APRM Neutron Flux Scram No change APRM Setdown Scram (%RTP) (l) 25 l 23 Rod Block Monitor 6 6 WD + 69% 0.55WD + 63.5%

DLO Flow Biased (%RTP) (0X4) ,.

SLO Flow Biased (%RTP) (IX2X4) 0.66(WD - AW)+ 69% 0.55(WD- anV) +

__ __ __ _ __ _ __ _ __ __ __ _ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _6 3 .5 %

Rod Block Monitor Upscale Function Ranges

-Low Power Range No change Intermediate Power Range No change 1I' High Power Range No change Typical Low Trip Setpoint (4) No change Typical Intermediate Trip Setpoint (4) No change Typical High Trip Setpoint (4) No change Vessel High Pressure Scram (psig) 1071 1101 High Pressure ATWS RPT (psig) 1146.5 1177 1105+3% 1135+3%

Main Steam Relief Valve Setpoints (psig) 1115+3% 1145+3%

1125+3% 1155+3%

TSV & TCV Scram Bypass (%RTP) (IX3) 30 26 Main Steam Line High Flow Isolation No change Main Steam Line Tunnel High Temperature Isolation No change Feedwater Flow Cavitation Interlock No change Low Steam Line Pressure MSIV Closure (Run Mode) No change RCIC Steam Line High Flow Isolation No change HPCI Steam Line High Flow Isolation No change 5-9

NEDO-33101

1. No credit is taken in any safety analysis.

2. WD is % recirculation drive flow where 100% drive flow is that required to achieve 100% core flow at 100% power, and AW is the difference between the DLO and SLO drive flow at the same core flow.

3. TS AV provided.

4. Changed on a cycle-specific basis and documented in the COLR.

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NEDO-33101

6. ELECTRICAL POWER AND AUXILIARY SYSTEMS 6.1 AC POWER The Browns Ferry AC power supplies each include both off-site and on-site power. The on-site power distribution system consists of transformers, buses, and switchgear. AC power to the distribution system is provided from the offsite transmission system or from onsite Diesel Generators. The Browns Ferry EPU plant electrical characteristics are shown in Table 6-1.

6.1.1 Off-Site Power System The existing off-site electrical equipment was determined to be adequate for normal operation with the EPU electrical output as shown in Table 6-2. The only significant change in electrical load demand is due to the replacement with larger motors for the Condensate Booster and Condensate Pumps due to increased flow demand at EPU conditions. The review concluded the following:

• The Main Isolated Phase Bus Duct is to be modified/up-rated to have a continuous current rating of 36,740 Amperes and an asymmetrical current rating of 346,989 amps to support the Generator output at EPU conditions.

• The Tap Isolated Phase Bus Duct is to be modified/up-rated to have an asymmetrical current rating of 602,143 amps to support the Generator output at EPU conditions.

• The Generator breaker is to be modified/up-rated to have a continuous current rating of 36,740 Amperes and an asymmetrical current rating of 204,529 amps to support the Generator output at EPU conditions.

• The existing main power transformers are being upgraded as a material condition improvement due to obsolescence. The replacement transformers are adequate for operation with the EPU-related electrical output of the generator.

• Changes will be required to plant operating procedures to prevent automatic transfer to the 161-kV system when any unit is in backfeed or when any USST B is out of service to avoid overloading any of the 161 -kV power supply circuits.

• The existing 500-kV switchyard buses, breakers, and switches are adequate for EPU operations. However, additional breakers are being added to increase the operating flexibility of the 500-kV switchyard.

• The protective relaying for the main generator, transformer, and switchyard is adequate for the EPU generator output.

A Transmission System Study has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criteria 17 (10 CFR 50, Appendix A) and to analyze for unit/grid stability. The study documented that no additional changes are required for Browns Ferry's offsite power system to continue to meet GDC-17 requirements. Analyses in the study also determined that operation at EPU electrical outputs will not have a significant adverse effect on reliability of the offsite electrical system or on the stability of the Browns Ferry units.

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NEDO-33101 6.1.2 On-site Power Distribution System The on-site power distribution system loads were reviewed under normal and emergency operating scenarios for EPU conditions. Loads were computed based on equipment nameplate data or brake horsepower (BHP) as applicable. These loads were used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the EPU conditions is achieved for normal and emergency conditions by operating equipment within the nameplate rating running kW or applicable BHP.

The only significant change in electrical load demand is associated with power generation system motors for the condensate and condensate booster pumps. These system pumps experience increased flow demand at EPU conditions and will be replaced with higher capacity pumps and motors. To support these load increases, modifications to the onsite electrical system will be performed prior to EPU operation. Load flow and short circuit calculations were performed to verify the adequacy of the on-site AC system for the proposed changes. The existing protective relay settings are adequate to accommodate the increased load on the 4kV power system.

Selective coordination is maintained between the pump motor breakers and the 4kV Unit Board main feeder breakers.

Significant changes to the on-site power analysis include:

• The BHP of the Reactor Recirculation Pump motors increases 17% for EPU, but remains within its motor uprate analysis capability.

• The Recirculation MG sets have been replaced with Variable Frequency Drives. The capability of .the VFD is 9000 HP, which is adequate for the expected Reactor Recirculation Pump motor load of 8550 BHP.

• The electrical load demand associated with power generation system motors for the condensate pumps and condensate booster pumps increase for EPU. These system pumps experience increased flow demand at EPU conditions and will be replaced with higher capacity pumps and motors.

EPU conditions are achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the required pump motors for both normal and emergency operating conditions.

Units 1 and 2 share four independent safety-related diesel generator units coupled, as an alternate source of power, to four independent 4160-V boards. ((

)) The systems have sufficient capacity to support all required loads to achieve and maintain safe shutdown conditions and to operate the ECCS equipment following postulated accidents and transients.

6.2 DC POWER The DC power distribution system provides control and motive power for various systems/components within the plant. In normal and emergency operating conditions, loads are computed based on equipment nameplate ratings. These loads are used as inputs for the 6-2

NEDO-33101 computation of load, voltage drop, and short circuit current values. The load addition for control logic relays associated with on-site power system changes are within existing margins.

Operation at the EPU conditions does not increase any load beyond nameplate rating or revise any component operating duty cycle; therefore, the DC power distribution system remains adequate.

6.3 FUEL POOL The fuel pool systems consist of storage pools, fuel racks, the FPCC system, and the ADHR.

The objective of the fuel pool system is to provide specially-designed underwater storage space for the spent fuel assemblies. The objective of the fuel pool systems is to remove the decay heat from the fuel assemblies and maintain the fuel pool water within specified temperature limits.

6.3.1 Fuel Pool Cooling The Browns Ferry SFP bulk water temperature must be maintained below the licensing limit of 150'F. The limiting condition is a full core discharge with all remaining storage locations filled with used fuel from prior discharges. A normal batch offload (approximately 332 fuel bundles) is assumed for outage planning with the additional assumptions in either case (batch or full core) of only one of two trains of the FPCC system and only one of two trains of the non-safety ADHR system available, 24-month fuel cycle, ANSI/ANS 5.1-1979 + 2a, and GE-14 fuel. The RHR system supplemental fuel pool cooling mode may be used to augment the capacity of the FPCC system when the ADHR system is unavailable. The batch and full core offload scenarios were also analyzed with only one of two trains of FPCC system and one train of RHR in the supplemental fuel pool cooling mode. The key results of these analyses are presented in Table 6-

3. The temperature requirement ensures operator comfort (an operational requirement), and provides ample margin against an inventory loss in the fuel pool due to evaporation or boiling.

The EPU SFP heat load is higher than the pre-EPU heat load. The EPU heat loads at the limiting full core offload condition and the normal batch offload are calculated and then the bulk pool temperature is determined to evaluate the FPCC system adequacy. EPU does not affect the heat removal capability of the FPCC system, the ADHR, or the supplemental fuel pool cooling mode of the RHR system. EPU results in slightly higher core decay heat loads during refueling. Each reload affects the decay heat generation in the SFP after a batch discharge of fuel from the reactor. The full core offload heat load in the SFP reaches a maximum immediately after the full core discharge. Based on the heat load evaluations, the SFP bulk temperature remains less than 150'F for either core offload case, and thus, is acceptable for EPU conditions.

The SFP normal makeup source is from the Seismic Category II Condensate Storage system, with a capacity of 100 gpm and is not affected by EPU and remains adequate for EPU conditions.

In the unlikely event of a complete loss of SFP cooling capability, Table 6-3 shows that the SFP could reach the boiling temperature and produce a maximum boil-off rate of 104 gpm. Two Seismic Category I emergency makeup sources, the RHR/RHR Service Water crosstie and the EECW system, each have a makeup capability of at least 150 gpm.

Prior to each refueling outage, calculations are performed to determine the actual pool heat load and determine which equipment must be placed in service to maintain pool temperature.

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NEDO-33101 Administrative controls are used to ensure that the fuel pool cooling capacity is not exceeded during core offload. Existing plant instrumentation and procedures provide adequate indications and direction for monitoring and controlling SFP temperature and level during normal batch offloads and the unexpected case of the limiting full core offload. Symptom based operating procedures exist to provide mitigation strategies including placing additional cooling trains or systems in service, stopping fuel movement, and initiating make-up, if necessary. The symptom based entry conditions and mitigation strategies for these procedures do not require changes for EPU.

6.3.2 Crud Activity and Corrosion Products The crud in the SFP would increase by approximately 2%, assuming that all residual crud in the RCS is transported to the SFP. This is based on an RWCU system removal efficiency of 90%

and approximately 23% increase in FW flow for EPU. However, the increase is insignificant, and SFP water quality is maintained by the FPCC system.

6.3.3 Radiation Levels The normal radiation levels around the SFP may increase slightly primarily during fuel handling operation. Current Browns Ferry radiation procedures and radiation monitoring program would detect any changes in radiation levels and initiate appropriate actions.

6.3.4 Fuel Racks The increased decay heat from the EPU results in a higher heat load in the fuel pool during long-term storage. The fuel racks are designed for higher temperatures (212 0F) than the licensing limit of 1500 F. There is no effect on the design of the fuel racks because the original fuel pool design temperature is not exceeded.

6.4 WATER SYSTEMS The Browns Ferry water systems are designed to provide a reliable supply of cooling water for normal operation and design basis accident conditions.

6.4.1 Service Water Systems The Browns Ferry water systems consist of safety-related and nonsafety-related service water systems, the circulating water system and main condenser, and the ultimate heat sink. The safety-related service water systems include the EECW system, the RHRSW system, and the UHS. The nonsafety-related service water systems include the RBCCW system and the RCW system.

6.4.1.1 Safety-Related Loads The safety-related service water systems are designed to provide a reliable supply of cooling water during and following a design basis accident for the following essential equipment and systems:

• RHR HXs;

• SFP HXs, as needed for supplemental cooling; 6-4

NEDO-33101

• SFP emergency make-up, if necessary;

• Standby core and containment cooling emergency backup, if necessary; and

• EECW system.

The evaluation of the systems performance is given in the following subsections.

6.4.1.1.1 Emergency Equipment Cooling Water System [KTS227]

The safety-related performance of the EECW system during and following the most demanding design basis event, the LOCA, for the following equipment and systems is not dependent on RTP:

• EDG Engine Coolers;

• RHR Pump Seal Coolers;

• Control Air Compressors;

• Diesel Generator Building Chillers;

• Electric Board Room ACU Condensers and Chillers;

• Control Bay Chillers;

• H2 -02 Analyzers;

• Control Room Compressors; and

• RBCCW Heat Exchangers.

The diesel generator loads, RBCCW HXs, control air compressor loads, RHR pump seal loads, H2-02 Analyzer loads remain unchanged for LOCA conditions following uprated operation.

The building cooling loads (Area Cooling Units) also remain the same as that for rated operation because the equipment performance in these areas has remained unchanged for post-LOCA conditions. The RHR and CS Room Cooler post-LOCA heat loads increase slightly because of room temperature increases at EPU conditions (< 2F for RHR and < 3F for CS), but remain within the current design limits.

The EECW system is a shared system with the capacity to supply cooling water to all three units.

EPU does not significantly increase equipment cooling water loads, and thus, the capacity of the EECW system remains adequate.

6.4.1.1.2 Residual Heat Removal Service Water System The containment cooling analysis in Section 4.1.1 shows that the post-LOCA RHR heat load increases due to an increase in the maximum suppression pool temperature that occurs following a LOCA. The post-LOCA containment and suppression pool responses have been calculated based on an energy balance between the post-LOCA heat loads and the existing heat removal capacity of the RHR and RHRSW systems. As discussed in Sections 3.11 and 4.1.1, the existing suppression pool structure and associated equipment have been reviewed for acceptability based on this increased post-LOCA suppression pool temperature. Therefore, the containment cooling analysis and equipment review demonstrate that the suppression pool temperature can be 6-5

NEDO-33101 maintained within acceptable limits in the post-accident condition based on the existing capability of the RHRSW system.

With EPU, the RHRSW system has sufficient capacity to supply adequate cooling and makeup to the spent fuel pool heat exchangers and spent fuel pool, respectively. In addition, the RHRSW system has sufficient capacity to serve as a standby coolant supply for long term core and containment cooling as required for EPU conditions. The RHRSW system flow rate is not changed.

6.4.2 Main Condenser/Circulating Water/Normal Heat Sink Performance The main condenser, circulating water, and 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. Maintaining adequately low condenser pressure ensures the efficient operation of the turbine-generator and minimizes wear on the turbine last stage buckets.

EPU operation increases the heat rejected to the condenser and, therefore, reduces the difference between the operating pressure and the recommended maximum condenser pressure. If condenser pressures approach the main turbine backpressure limitation, then reactor thermal power reduction would be required to reduce the heat rejected to the condenser and maintain condenser pressure within the main turbine requirements.

The performance of the main condenser was evaluated for EPU. This evaluation is based on a design duty over the actual range of circulating water inlet temperatures, and confirms that the condenser, circulating water system, and heat sink are adequate for EPU operation. Current main turbine backpressure limitations may require load reductions at the upper range of the anticipated circulating water inlet temperatures.

6.4.2.1 Discharge Limits The state discharge limits were compared to the current discharges and bounding analysis discharges, as shown in Table 6-4. This comparison indicates that Browns Ferry will comply with the state discharge limit during operation at EPU. Based on recorded historical data, the administrative control procedures presently in place remain valid to ensure EPU operation remains within state discharge limits.

6.4.3 Reactor Building Closed Cooling Water System The heat loads on the RBCCW system increase < 0.3%. The RBCCW heat loads are mainly dependent on the reactor vessel temperature and/or flow rates in the systems cooled by the RBCCW. The change in vessel temperature is minimal and does not result in any significant increase in drywell cooling loads. The flow rates in the systems cooled by the RBCCW (e.g.,

Recirculation and RWCU pumps cooling) do not change due to EPU and, therefore, are not affected by EPU. The operation of the remaining equipment cooled by the RBCCW (e.g.,

sample coolers and drain sump coolers) is not power-dependent, and is not affected by EPU.

The RBCCW system contains sufficient redundancy in pumps and heat exchangers to ensure that adequate heat removal capability is available during normal operation. Sufficient heat removal capacity is available to accommodate the small increase in heat load due to EPU.

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NEDO-33101 6.4.4 Raw Cooling Water System The temperature of RCW system discharge results from the heat rejected to the RCW system via components cooled by the system. The power dependent heat loads on the RCW system that are increased by EPU are those related to the operation of the RBCCW system, the condensate pumps, condensate booster pumps, and the isolated phase bus duct air HX. The increase in RCW system discharge temperature from these sources due to EPU is < 1PF, which is minimal and within equipment tolerances.

6.4.5 Ultimate Heat Sink The UHS is the Wheeler Reservoir/Tennessee River. The upstream temperature of the river is unaffected by operations at EPU conditions. The existing UHS system provides a sufficient quantity of water at a temperature less than 95'F (design temperature) to perform its safety-related functions for EPU. In the TS, the UHS maximum allowable temperature is 950F, and thus, the TS is not changed for EPU.

The UFSAR includes a discussion relative to heatup of the downstream portion of the pool that would exist following the loss of the downstream dam on the Tennessee River. The river thermal rise post-shutdown would increase due to the increase in decay heat associated with EPU conditions but would not significantly affect this event.

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 some or all of the control rods cannot be inserted. This manually operated system pumps a highly enriched sodium pentaborate solution into the vessel to provide neutron absorption and achieve a subcritical reactor condition. SLCS is designed to inject over a wide range of reactor operating pressures.

)) The TS minimum available volume of sodium pentaborate solution associated with this increase is bounded by the volume requested in Reference 1.

The boron injection rate requirement for the limiting ATWS event with SLCS injection is not increased for EPU.

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NEDO-33101

)) The SLCS pumps are positive displacement pumps, where small pressure changes in the MSRV setpoint would have no effect on the rated injection flow to the reactor. As discussed below, the resulting increase in system operating pressure has not reduced the SLCS pump relief valve pressure margin below recommended levels. Therefore, the capability of the SLCS to provide its backup shutdown function is not affected by EPU.

The surveillance test pressure is based on the maximum SLCS injection pressure and allowances for system test inaccuracies. This test pressure is increased from the pre-EPU value of 1275 to 1325 psig to account for the increase in system injection pressure resulting from EPU conditions.

Based on the results of the plant-specific ATWS analysis, the maximum reactor lower plenum pressure following the limiting ATWS event reaches 1204 psig during the time the SLCS is analyzed to be in operation. Consequently, there is a corresponding increase in the maximum pump discharge pressure and a decrease in the operating pressure margin for the pump discharge relief valves. The operation of the pump discharge system was analyzed to confirm that the pump discharge relief valves re-close in the event that the system is initiated before the time that the reactor pressure recovers from the first transient peak. The evaluation compared the calculated maximum reactor pressure needed for the pump discharge relief valves to re-close with the lower reactor pressure expected during the time the MSRVs are cycling open and closed prior to the time when rated SLCS injection is assumed in the ATWS analysis. Consideration was also given to system flow, head losses for full injection, and cyclic pressure pulsations due to the positive displacement pump operation in determining 'the pressure margin to the opening set point for the pump discharge relief valves. The pump discharge relief valves are periodically tested to maintain this tolerance. Therefore, the current SLCS process parameters associated with the minimum boron injection rate are not changed.

((

)) The evaluation shows that EPU has no adverse effect on the ability of the SLCS to mitigate an ATWS event.

6.6 POWER DEPENDENT HVAC The HVAC systems consist mainly of heating, cooling supply, exhaust, and recirculation units in the turbine building, reactor building, and the drywell. EPU results in slightly higher process temperatures and small increases in the heat load due to higher electrical currents in some motors and cables.

The affected areas are the drywell; the steam tunnel and the ECCS rooms in the reactor building; and the FW heater bay, condenser, and the condensate/FW pump areas in the turbine building.

Other areas in the reactor building and the turbine building are unaffected by the EPU because the process temperatures remain relatively constant.

The increased heat loads during normal plant operation result in < 3F increase in the drywell and the MSL tunnel, due to the 30 psi reactor pressure increase. In the turbine building, the 6-8

NEDO-33101 maximum temperature increase in the FW heater bay, condensate/FW pump, and condenser areas is < 2F.

Based on a review of design basis documents, the design of the HVAC is adequate for the EPU with the exception of the condensate and condensate booster pump motor coolers. Replacement of or modification to these pump motors, described in Section 7.4, may require modifications to their coolers.

6.7 FIRE PROTECTION This section addresses the effect of EPU on the fire protection program, fire suppression and detection systems, and reactor and containment system responses to postulated 10 CFR 50 Appendix R fire events.

(( )] Any changes in physical plant configuration or combustible loading as a result of modifications to implement the EPU, will be evaluated in accordance with the Browns Ferry modification and fire protection programs. The safe shutdown systems and equipment used to achieve and maintain cold shutdown conditions do not change, and are adequate for the EPU conditions. The scope of operator actions required to mitigate the consequences of a fire are not affected. Therefore, the fire protection systems and analyses are not affected by EPU.

The reactor and containment responses to the postulated 10 CFR 50 Appendix R fire event at EPU conditions are evaluated in Section 6.7.1. The results show that the peak fuel cladding temperature, reactor pressure, and containment pressures and temperatures are below the acceptance limits and demonstrate that there is sufficient time available for the operators to perform the necessary actions to achieve and maintain cold shutdown conditions. Therefore, the fire protection systems and analyses are not adversely affected by EPU.

6.7.1 10 CFR 50 Appendix R Fire Event A plant-specific evaluation was performed to demonstrate safe shutdown capability in compliance with the requirements of 10 CFR 50 Appendix R assuming EPU conditions. The limiting Appendix R fire event from the current Browns Ferry Unit 2/3 analysis was reanalyzed for Browns Ferry Unit 1 assuming EPU. The fuel heatup analysis was performed using the SAFER/GESTR-LOCA analysis model. The containment analysis was performed using the SHEX model. Justification for using SAFERIGESTR-LOCA and SHEX models for EPU calculations is presented in Section 4. These are the same analysis methodologies that were used for the Unit 2/3 Appendix R Fire event analysis. This evaluation determined the effect of EPU on fuel cladding integrity, reactor vessel integrity, and containment integrity as a result of the fire event.

The postulated Appendix R fire event using the minimum SSDS was analyzed for the three cases described below:

Case 1: No spurious operation of plant equipment occurs and the operator initiates three MSRVs 25 minutes into the event.

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NEDO-33101 Case 2: One MSRV opens immediately due to a spurious opening signal generated as a result of the fire. The MSRV is reclosed 10 minutes into the event by operator action. The operator initiates three MSRVs 20 minutes into the event.

Case 3: One MSRV opens immediately as in Case 2, but remains open throughout the event.

The operator initiates three MSRVs 20 minutes into the event.

The above are the same cases as those described in the Browns Ferry Fire Protection Report (Reference 2), except as described below.

These cases were evaluated for EPU with some reduction in conservatism in the analytical assessment; as compared to the methods used currently for Units 2 and 3.

For pre-EPU analysis for Browns Ferry Units 2 and 3, for all cases it was conservatively assumed that the LPCI injection does not occur until reactor pressure is

• 200 psig, instead of the standard injection point of 319.5 psig, which delays LPCI injection into the vessel. For the EPU assessment, the analysis is based on the reactor vessel pressure reaching 385 psig, and then the LPCI injection valve is opened by operator action. LPCI flow to the vessel begins at 319.5 psig.

This adjustment to the analysis does not affect any operator action because the current procedures direct the operations staff open the LPCI injection valve when RPV pressure is 5 450 psig.

The bounding PCT case is Case 1. For this case, the time available to the operator to open three MSRVs is 25 minutes at the EPU conditions. The Browns Ferry Units 2 and 3 pre-EPU analysis determined the 3 MSRVs were required to be opened within 30 minutes. This reduction in the time available does not have any effect because the procedures will require this action to be completed within 20 minutes. For OLTP and EPU, the PCTs are calculated using conservative LPCI performance characteristics (e.g., minimum flow rate as functions of vessel pressure).

In addition, spurious operation of HPCI was reviewed in accordance with Reference 2. The HPCI system was assumed to initiate at the onset of the Appendix R event, and flow at its normal flow rate. The time at which the reactor vessel water level would reach the MSLs is greater than 6 minutes. Therefore, the procedures will require HPCI isolation prior to 6 minutes during an Appendix R event.

The results of the Appendix R evaluation for EPU provided in Table 6-5 demonstrate that the fuel cladding integrity, reactor vessel integrity, and containment integrity are maintained and that sufficient time is available for the operator to perform the necessary actions. The current exemption for the momentary core uncovery during depressurization remains necessary for EPU.

EPU does not affect any other exemptions described in Reference 2. No changes are necessary to the equipment required for safe shutdown for the Appendix R event. One train of systems remains available to achieve and maintain safe shutdown conditions from either the main control room or the remote shutdown panel. Therefore, EPU has no adverse effect on the ability of the systems and personnel to mitigate the effects of an Appendix R fire event, and satisfies the requirements of Appendix R with respect to achieving and maintaining safe shutdown in the event of a fire.

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NEDO-33101 6.8 SYSTEMS NOT IMPACTED BY EXTENDED POWER UPRATE 6.8.1 Systems with No Impact Similar to the systems listed in Table J-1 of ELTRI (Reference 3), the systems in Table 6-6 are not affected by operation of the plant at the EPU power level.

6.8.2 Systems With Insignificant Impact The systems affected in a very minor way by operation of Browns Ferry at the uprated power level are listed in Table 6-7. This listing is similar to the systems listed in Table J-2 of ELTR1.

For these systems, the effects of EPU are insignificant with respect to their design and operation.

6.9 REFERENCES

1. TVA Letter, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - License Amendment

- Alternative Source Term," R08 020731 649, July 31, 2002, including Tech. Spec. No. 405 (TVA-BFN-TS-405).

2. Tennessee Valley Authority, "Fire Protection Report," Vol. 1, Revision 16, January 2001.

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

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NEDO-33101 Table 6-1 Brown Ferry EPU Plant Electrical Characteristics Parameter Value Generator Output (MWe) 1280 Rated Voltage (kV) 22.0 Power Factor 0.962 Generator Output (MVA) 1330 Current Output (kA) 36.740 Isolated Phase Bus Duct Rating (kA) 36.740 Main Transformers Rating (MVA) 1500 Transformer Output (MVA) 1300 Table 6-2 Brown Ferry Offsite Electric Power System Component Rating EPU Output Generator (MVA) 1330 1330 Isolated Phase Bus Duct (kA) 36.740 36.740 Main Transformers (MVA) 1500 1330(')

Auxiliary Transformer (MVA) 72(2) 4 0 (3)

Switchyard (limiting) (MVA) 1750(4) 1300(5)

1. Auxiliary transformers loading of 30 MVA used for conservative evaluation of main transformer.

2. Two auxiliary transformers rated 40 MVA and 32 MVA.

3. Determined using actual plant data with estimated additional loading due to EPU.

4. Seven (7) 500 kV lines each rated at 1750 MVA.

5. 1330 MVA Unit I Generator rating minus 30 MVA station load.

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NEDO-33101 Table 6-3 Browns Ferry Unit 1 Spent Fuel Pool Parameters Limiting Full Conditions / Parameter Batch Core Ofiload Limit Configuration 1:

One train each of FPCC and ADHR in service (I)

Peak SFP Temperature (0F) 99.1 121.5 125 (Batch) 150 (Full Core)

Time to Peak SFP Temperature (hr) 80 109 NA Time to boil from loss of all cooling at 14 5 NA peak temperature (hr)

Boil off rate (gpm) 48 104 150 Configuration 2:

One train each of FPCC and RHR supplemental fuel pool cooling mode in service Peak SFP Temperature (0F) 124.9 149.8 125 (Batch) 150 (Full Core)

Time to Peak SFP Temperature (hr) 130(2) 229(3) NA Time to boil from loss of all cooling at 13 4 NA peak temperature (hr)

Boil off rate (gpm) 42 80 150

1. Assumes core offload begins 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after reactor shutdown to allow for cooldown, vessel head removal, refueling cavity filling, and other refueling preparations.

2. Assumes core offload begins 95 hours0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br /> after reactor shutdown and includes 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of in-vessel stay time because the RHR supplemental fuel pool cooling mode has less heat removal capacity than the ADHR system and 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> to allow for cooldown, vessel head removal, refueling cavity filling, and other refueling preparations.

3. Assumes core offload begins 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> after reactor shutdown and includes 115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br /> of in-vessel stay time because the RHR supplemental fuel pool cooling mode has less heat removal capacity than the ADHR system and 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> to allow for cooldown, vessel head removal, refueling cavity filling, and other refueling preparations.

6-13

NEDO-33101 Table 6-4 Browans Ferry Effluent Discharge Comparison Parameter State Unit 1 @

Limit EPU Flow (million gallons/day) None 1008 Downstream Temperature 24-hour avg. (IF) 90.0 90.0 Downstream Temperature 1-hour avg. (IF) 93.0 93.0 In-stream Delta T. 24-hour avg. (OF) 10.0 <10.0 Chlorine (average/day) mg/L/per day 0.064 < 0.064 Net heat addition (MBTU/hr) None 9,284 6-14

NEDO-33101 Table 6-5 Browns Ferry Appendix R Fire Event Evaluation Results Parameter 105% EPU App. R Criteria OLTPO')

Cladding Heatup (PCT), 0F 1485 1428 < 1500 Primary System Pressure, psig 1150 1150 < 1375 Primary Containment Pressure, psig 18.6 13.6 < 56 Suppression Pool Bulk Temperature, IF 212 227 < 281°

• 2274)

NPSH (2) Yes Yes Adequate for system using suppression pool water source

1. Values based on Browns Ferry Units 2 and 3 at 105% OLTP.

2. NPSH demonstrated adequate, see Section 4.2.5.

3. Containment structure design limit.

4. Torus attached piping limit.

6-15

NEDO-33101 Table 6-6 Browns Ferry Systems With No Impact Svstem Number Svstem Title 012 Auxiliary Boiler 018 Fuel Oil 020 Lubricating Oil 027 Condenser Circulating Water Tube Cleaning 028 Water Treatment 029 Potable Water 032 Control Air 033 Service Air 034 Vacuum Priming 036 Auxiliary Boiler Feedwater Secondary Treatment 037 Gland Seal Water 038 Insulating Oil 040 Drainage 043 Sampling and Water Quality 044 Building Heating 049 Breathing Air 050 Raw Water Chemical Treatment 051 Chemical Treatment 052 Seismic Monitoring 056 Temperature Monitoring 058 Biothermal Facility 064C Secondary Containment 079 Fuel Handling and Storage 086 Diesel Generator Starting Air 099 Reactor Protection 111 Cranes 112 Shop Equipment 200 200 Series display boards 244 Communications 247 240 VAC Lighting System 258 Operations Recorder 260 Security 301 Sewage Disposal 302 Elevators 315 Microwave 327 Flood Protection 417 Meteorological Tower 6-16

NEDO-33101 Table 6-7 Browns Ferry Systems With Insignificant Impact Svstem Number System Title 004 Hydrogen Water Chemistry 009 Control Bay Panels 030 Normal Ventilation 053 Demineralizer Backwash Air 055 Annunciator 080 Drywell Temperature Monitoring 925 Control Panels 6-17

NEDO-33101

7. POWER CONVERSION SYSTEMS 7.1 TURBINE-GENERATOR The turbine and generator was originally designed with a maximum flow-passing capability and generator output in excess of rated conditions to ensure that the original rated steam-passing capability and generator output is achieved. This excess design capacity ensures that the turbine and generator meet rated conditions for continuous operating capability with allowances for variations in flow coefficients from expected values, manufacturing tolerances, and other variables that may adversely affect the flow-passing capability of the units. The difference in the steam-passing capability between the design condition and the rated condition is called the flow margin.

At EPU RTP and reactor dome pressure of 1050 psia, the turbine operates at an increased rated throttle steam flow of 16.44 Mlb/hr and at a throttle pressure of 988 psia. To maintain control capability GE uses a minimum target value of approximately 3% throttle flow ratio, with controllability confirmed by unit testing as described in Section 10.4. For operation at EPU, the high pressure turbine has been redesigned with replacement diaphragms and buckets and a new rotor for at least the minimum target throttle flow margin, to increase its flow passing capability.

The generator is rated at 1330 MVA, which results in a rated electrical output (gross) of 1280 MWe at a power factor of 0.962.

The expected environmental changes, such as diurnal heating and cooling effects changing cycle efficiency, periodically require management of reactor power to remain within the generator rating. The required variations in reactor power do not approach the magnitude of changes periodically required for surveillance testing and rod pattern alignments and other occasional events requiring de-rating, such as equipment out of service for maintenance.

As part of the EPU on Unit 1, the original shrunk on LP rotors will be replaced with monoblock design rotors. A turbine missile probability assessment is therefore not applicable since the turbines will have integral (i.e. monoblock) wheels in lieu of shrunk-on wheels (Ref. NUREG-1048).

The overspeed calculation compares the entrapped steam energy contained within the turbine and the associated piping, after the stop valves trip, and the sensitivity of the rotor train for the capability of overspeeding. The entrapped energy increases slightly for the EPU conditions. The hardware modification design and implementation process establishes the overspeed trip settings to provide protection for a turbine trip.

7.2 CONDENSER AND STEAM JET AIR EJECTORS The condenser converts the steam discharged from the turbine to water to provide a source for the condensate and FW systems. The SJAE remove noncondensable gases from the condenser to improve thermal performance.

The condenser and SJAE functions are required for normal plant operation and are not safety-related.

The condensers were evaluated for performance at EPU conditions based on a maximum cold water temperature of 90'F and current circulating water system flow. Additional analysis at 7-1

NEDO-33101 EPU conditions also determined the condenser back pressure would be below the 5" Hga design limit, assuming cleanliness levels as low as 85%.

Due to the increase in condensate flow rate associated with EPU conditions, the retention time of condensate in the condenser hotwell is slightly reduced to 1.7 minutes. Condenser hotwell capacities and level instrumentation are adequate for EPU conditions. Periodic eddy current testing and water chemistry monitoring are performed to monitor the effects of EPU RTP operation on the condenser tubes.

The design of the condenser air removal system is not adversely affected by EPU and no modification to the system is required. The physical size of the primary condenser and evacuation time are the main factors in establishing the capabilities of the vacuum pumps. These parameters do not change. Because flow rates do not change, there is no change to the two-minute holdup time in the pump discharge line routed to the reactor building vent stack. The design capacity of the SJAEs is not affected by EPU, because they were originally designed for operation at greater than warranted flows.

7.3 TURBINE STEAM BYPASS The Turbine Steam Bypass system provides a means of accommodating excess steam generated during normal plant maneuvers and transients.

The turbine bypass valves were initially rated for a total steam flow capacity of not less than 25% of the original rated reactor steam flow, or -3.56 Mlb/hr. Each of nine bypass valves is designed to pass a steam flow of -400,000 lbm/hr, and does not change at EPU RTP. The small pressure increase (<3%) due to the EPU is within the original design capability of the turbine bypass system. At EPU conditions, rated reactor steam flow is 16.44 Mlb/hr, resulting in a bypass capacity of 21.7% of EPU rated steam flow. The bypass capacity at Browns Ferry remains adequate for normal operational flexibility at EPU RTP.

The bypass capacity is used as an input to the reload analysis process for the evaluation of transient events that credit the Turbine Steam Bypass System (see Section 9.1).

7.4 FEEDNVATER AND CONDENSATE SYSTEMS The FW and condensate systems do not perform a system level safety-related function, and are designed to provide a reliable supply of FW at the temperature, pressure, quality and flow rate as required by the reactor. Therefore, these systems are not safety-related. However, their performance has a major effect on plant availability and capability to operate at the EPU condition. Modifications to some nonsafety-related equipment in the FW and condensate systems are necessary to attain full EPU core thermal power. Implementation of these modifications is reviewed per the site 10 CFR 50.59 process.

For EPU, the FW and condensate systems meet the following performance criteria with modifications to some nonsafety-related equipment:

1.- The systems provide a reliable supply of FW at the EPU dome pressure with sufficient capacity to supply the steady-state FW flow demanded at the EPU condition.

7-2

NEDO-33101

2. The systems have the capacity to provide at least 105% of the EPU FW flow. This ensures that Browns Ferry remains available during water level transients, avoids scrams, and minimizes challenges to plant safety systems.

3. The FW system is capable of providing adequate FW flow at the expected operating pressure, and to provide unit trip avoidance when one FW pump is tripped.

4. The runout capacity of the FW system in the limiting pump alignment does not exceed the performance capacity assumed in the transient analyses.

7A.1 Normal Operation System operating flows at EPU increase approximately 20% of rated flows at the OLTP. The condensate and FW systems will be modified to assure acceptable performance with the new system operating conditions.

The FW heaters will be analyzed and verified to be acceptable for the higher FW heater flows, temperatures, and pressures for the EPU, and re-rated prior to implementation of EPU. The performance of the FW heaters will be monitored during the EPU power ascension program.

7.4.2 Transient Operation To account for FW demand transients, the FW system was evaluated to ensure that a minimum of 5% margin above the EPU FW flow was available. For system operation with all system pumps available, the predicted operating parameters were acceptable and within the component capabilities.

The FW system post, feed pump trip capacity was evaluated to confirm that with the modifications to the FW and Condensate system configurations, the capability to supply the transient flow requirements is maintained or increased.

A transient analysis was performed (Section 9.1.3) to determine the reactor level response following a single FW pump trip. The results of the analysis show that the system response is adequate during the EPU conditions.

7.4.3 Condensate Demineralizers The effect of EPU on the CFDs was reviewed. The system requires modification to support CFD full flow operation during backwashing and pre-coating without requiring a plant power reduction. The system experiences slightly higher loadings resulting in slightly reduced CFD run times. However, the reduced run times are acceptable (refer to Section 8 for the effects on the radwaste systems).

7-3

NEDO-33101

8. RADWASTE AND RADIATION SOURCES 8.1 LIQUID AND SOLID WASTE MANAGEMENT The Liquid and Solid Radwaste systems collects, monitors, processes, stores, and returns processed radioactive waste to the plant for reuse or for discharge.

The single largest source of liquid and wet solid waste is from the backwash of condensate demineralizers. EPU results in an increased flow rate through the condensate demineralizers, resulting in a reduction in the average time between backwashes. This reduction does not affect plant safety. Similarly, the RWCU filter-demineralizer requires more frequent backwashes due to higher levels of impurities as a result of the increased FW flow.

The floor drain collector subsystem and the waste collector subsystem both receive periodic inputs from a variety of sources. EPU does not affect system operation or equipment performance. Therefore, neither subsystem is expected to experience a large increase in the total volume of liquid and solid waste due to operation at EPU conditions.

The increased loading of soluble and insoluble species increases the volume of the liquid processed wastes by 5.6% and the volume of the solid processed wastes by 21.7%. The total volume of liquid and solid processed waste does not significantly increase (as compared to the Radwaste System capacity) because the only increase in processed waste is due to more frequent backwashes of the condensate demineralizers and RWCU filter demineralizers. The total liquid and solid increases are within the Radwaste System capacity. Therefore, EPU does not have an adverse effect on the processing of liquid and solid radwaste, and there is no significant environmental effect.

The increases in the liquid and the solid processed waste are based on the increase due to the FW flow increase. The percentage bounding value for the increase in liquid and solid processed waste is equal to or less than that of the FW flow percentage increase.

8.2 GASEOUS WASTE MANAGEMENT The gaseous waste management systems collect, control, process, store, and dispose of gaseous radioactive waste generated during normal operations. The gaseous waste management systems include the Offgas System and various building ventilation systems. The systems are designed to meet the requirements of 10 CFR 20 and 10 CFR 50, Appendix I.

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

Building ventilation systems control airborne radioactive gases by using combinations of devices such as HEPA and charcoal filters, and radiation monitors that signal automatic isolation dampers or trip supply and/or exhaust fans, or by maintaining negative air pressure, where required, to limit migration of gases. ((

)) This is because the amount of fission products released into the coolant depends on the number and nature of the fuel rod defects, and is approximately linear with respect to core thermal power. The 8-1

NEDO-33101 concentration of coolant activation products in the steam remains nearly constant. The release limit is an administratively controlled variable, and is not a function of core power. The gaseous effluents are well within limits at original power operation and remain well within limits following implementation of EPU. There are no significant environmental effects due to EPU.

8.2.1 Offgas System The primary function of the Offgas System is to process and control the release of gaseous radioactive effluents to the site environs so that the total radiation exposure of persons in offsite areas is within the guideline values of 10 CFR 50, Appendix I.

The radiological release rate is administratively controlled to remain within existing site release rate limits and is a function of fuel cladding performance, main condenser air inleakage, charcoal adsorber inlet dew point, and charcoal adsorber temperature.

Because EPU affects the flow rate of radiolytic hydrogen and oxygen to the Offgas System, the catalytic recombiner temperature and offgas condenser heat load are affected. The Browns Ferry radiolytic decomposition rate is based upon Browns Ferry design specifications adjusted for EPU power level. The EPU analysis for the Offgas System utilized a higher decomposition rate that is more conservative than the Browns Ferry plant specific decomposition rate. The EPU hydrogen flow rates and concentrations are still within the design limits of the Offgas System. The catalytic recombiner and offgas condenser, as well as downstream components, have sufficient design margin to handle the increase in thermal power for EPU without exceeding the system design limits of temperature, flow rates, or heat loads.

In addition, HWC operation when used will cause a reduction in core radiolysis. The combination of the HWC injected hydrogen plus the reduced radiolysis is expected to produce a lower net hydrogen flow to the Offgas System.

8.3 RADIATION SOURCES IN THE REACTOR CORE 8.3.1 Normal Operation 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 or activity released per unit of reactor power. Therefore, for EPU, the percent increase in the operating source terms is no greater than the percent increase in power.

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

The second set of post-operation source data consists primarily of nuclide activity inventories for fission products in the fuel. These data are needed for post-accident and spent fuel pool 8-2

NEDO-33101 evaluations, which are performed in compliance with regulatory guidance that applies different release and transport assumptions to different fission products. The core fission product inventories for these evaluations are based on an assumed fuel irradiation time, which develops "equilibrium" activities in the fuel (typically 3 years). Most radiologically significant fission products reach equilibrium within a 60-day period. ((

))The radionuclide inventories are provided in terms of Curies per megawatt of reactor thermal power at various times after shutdown.

The results of the plant-specific radiation sources are included in the LOCA, FHA, and CRDA radiological analyses presented in Section 9.2. Plant specific analyses for NUREG-0737, Item ll.B.2, post-accident mission doses have been performed. The results of this assessment are accounted for in the Browns Ferry radiation protection program.

8.4 RADIATION SOURCES IN REACTOR COOLANT Radiation sources in the reactor coolant include activation products, activated corrosion products and fission products.

8.4.1 Coolant Activation Products During reactor operation, the coolant passing through the core region becomes radioactive as a result of nuclear reactions. The coolant activation, especially N-16 activity, is the dominant source in the turbine building and in the lower regions of the drywell. The activation of the water in the core region is in approximate proportion to the increase in thermal power.

-] The activation products in the steam from EPU are bounded by the existing design basis concentration. The margin in the design basis for reactor coolant activation concentrations significantly exceeds potential increases due to EPU. Therefore, no change is required in the activation design basis reactor coolant concentrations for EPU.

8.4.2 Activated Corrosion and Fission Products The reactor coolant contains activated corrosion products, which are the result of metallic materials entering the water and being activated in the reactor region. Under EPU conditions, the FW flow increases with power, and the activation rate in the reactor region increases with power. The net result is an increase in the activated corrosion product production.

The total activated corrosion product activity is approximately 3% higher than the original design basis activity as a consequence of EPU. However, the sum of the activated corrosion product activity and the fission product activity remains a small fraction (< 3%) of the total design basis activity.

8-3

NEDO-33101 Fission products in the reactor coolant are separable into the products in the steam and the products in the reactor water. The activity in the steam consists of noble gases released from the core plus carryover activity from the reactor water. This activity is the noble gas offgas that is included in the plant design. The calculated offgas rates for EPU after thirty minutes decay are well below the original design basis of 0.35 curies/sec. Therefore, no change is required in the design basis for offgas activity for EPU.

The fission product activity in the reactor water, like the activity in the steam, is the result of minute releases from the fuel rods. Fission product activity levels in the reactor water were calculated to be higher than previous calculated data, increasing from current values due to EPU.

These activity levels remain a fraction (<2%) of the design basis fission product activity.

Therefore, the activated corrosion product and fission product activities design bases are unchanged for EPU.

For the EPU, normal radiation sources increase slightly. Shielding aspects of the plant were conservatively designed for total normal radiation sources. Thus, the increase in radiation sources does not affect radiation zoning or shielding and plant radiation area procedural controls will compensate for increased normal radiation sources.

8.5 RADIATION LEVELS 8.5.1 Normal Operation For EPU, normal operation radiation levels increase slightly. For conservatism, many aspects of Browns Ferry were originally designed for higher-than-expected radiation sources. Thus, the increase in radiation levels does not affect radiation zoning or shielding in the various areas of Browns Ferry because it is offset by conservatism in the original design, source terms used and analytical techniques.

The normal operating doses are generally based on dose rate measurements at various locations during plant operation at OLTP conditions. The normal doses specified for OLTP conditions are increased by 20% with the exception of four zones where additional data was available to demonstrate that the normal doses specified for these areas contained sufficient margin to account for a 20% increase in the observed dose rate. The increased normal radiation doses are evaluated and determined to have no adverse effect on safety-related plant equipment, as indicated in Sections 10.3.1 and 10.3.2. Individual worker exposures can be maintained within acceptable limits by controlling access to radiation areas using the site ALARA program. Procedural controls compensate for increased radiation levels.

8.5.2 Normal Post-Operation Post-operation radiation levels in most areas of Browns Ferry are expected to increase by no more than the percentage increase in power level. In a few areas near the reactor water piping and liquid radwaste equipment, the increase could be slightly higher. Regardless, individual worker exposures can be maintained within acceptable limits by controlling access to radiation areas using the site ALARA program. Procedural controls compensate for increased radiation levels. Radiation measurements will be made at selected power levels to ensure the protection of personnel.

8-4

NEDO-33101 8.5.3 Post Accident The 100-day post-accident radiation doses are expected to increase by 12% or less at EPU RTP compared to the post-accident doses for Browns Ferry Units 2 and 3 at 105% OLTP conditions.

For some areas, the post-accident doses specified for Browns Ferry Units 2 and 3 at 105% OLTP conditions are bounding for the EPU conditions. The increased post-accident radiation doses have no adverse effect on safety-related plant equipment as indicated in Sections 10.3.1 and 10.3.2. Plant specific analyses for NUREG-0737, Item ll.B.2, post-accident mission doses have been performed.

Section 9.2 addresses the accident doses for the Main Control Room.

8.6 NORMAL OPERATION OFF-SITE DOSES The primary sources of normal operation offsite doses are (1) airborne releases from the offgas system and (2) gamma shine from the plant turbines.

The increase in activity levels is proportional to the percentage increase in core thermal power.

The TS limits implement the guidelines of 10 CFR 50, Appendix I. EPU does not involve significant increases in the offsite dose from noble gases, airborne particulates, iodine, tritium, or liquid effluents. Present offsite radiation levels form a negligible portion of background radiation. Therefore, the normal offsite doses are not significantly affected by operation at EPU and remain below the limits of 10 CFR 20, 10 CFR 50, Appendix I, and 40 CFR 190.

Browns Ferry Unit 1 is implementing zinc injection to limit the increase in normal radiation doses from the implementation of hydrogen water chemistry. The EPU increase in steam flow results in higher levels of N-16 and other activation products in the turbines. The increased flow rate and velocity, which result in shorter travel times to the turbine and less radioactive decay in transit, lead to higher radiation levels in and around the turbines and offsite skyshine dose. Any discernible increase in radiation as a result of increased N-16 would be measured on the site environmental TLD stations. Past history from these TLD stations for the implementation of HWC and the 5% power increase at Units 2 and 3 has not shown any discernible increase in radiation at offsite locations. Therefore, it is unlikely that the increase in N-16 source term due to EPU results in any measurable dose to the public.

8-5

NEDO-33101

9. REACTOR SAFETY PERFORMANCE EVALUATIONS 9.1 REACTOR TRANSIENTS The UFSAR evaluates the effects of a wide range of potential plant A0Os, commonly referred to as transients. Disturbances to Browns Ferry caused by a malfunction, a single equipment failure or an operator error are investigated according to the type of initiating event per Chapter 14 of the Browns Ferry UFSAR. Appendix E of ELTRI (Reference 1) identifies the limiting events to be considered in each category of events. The generic guidelines also identify the analytical methods, the operating conditions that are to be assumed, and the criteria that are to be applied.

The following paragraphs address each of the limiting events and provide a summary of the resulting transient safety analysis. The results given here are for a representative core, and show the overall capability of the design to meet all transient safety criteria for EPU operation.

Table E-1 of ELTRI provides the specific events to be analyzed for EPU, the power level to be assumed, and the computer models to be used. The transients that are not listed in Table E-1, are generally milder versions of the analyzed events. ((

))

The reactor operating conditions that apply most directly to the transient analysis are summarized in Table 9-1. They are compared to the conditions used for the UFSAR analyses.

An equilibrium core of GE14 fuel was used as the representative fuel cycle for the EPU transient analyses. Most of the transient events are analyzed at the full power and maximum allowed core flow operating point on the power/flow map, shown in Figure 2-1. Direct or statistical allowance for 2% power uncertainty is included in the analysis. ((

))The SLMCPR in Table 9-1 was used to calculate the OLMCPR value(s) required for the analyzed events. For all pertinent events, one MSRV is considered to be OOS, and the MSRV setpoint tolerance is considered to be + 3%. A discussion of other equipment OOS options is provided in Section 1.3.2.

The transient events of each category from Table E-1 of ELTRI were analyzed. Their inputs and results revise the licensing basis for the transient analysis to the EPU RTP. The overpressurization analysis is provided in Section 3.2. Other transient analysis results for the full EPU RTP condition are provided in Table 9-2. The most limiting transient event results are shown in Figures 9-1 through 9-4. As shown in the table and figures, no change to the basic characteristics of any of the limiting events is caused by EPU.

The severity of transients at less than rated power are not significantly affected by the EPU, because of the protection provided by the ARTS power and flow dependent limits.

The historical 25% of RTP value for the TS Safety Limit, some thermal limits monitoring LCOs thresholds, and some SRs threshold's is based on (( )) analyses (evaluated up to -50% of original RTP) applicable to the plant design with highest average bundle power (the BWR6) for all of the BWR product lines. As originally licensed, the highest average bundle power (at 100%

RTP) for any BWR6 is 4.8 MWt/bundle. The 25% RTP value is a conservative basis, as described in the TS, ((

9-1

NEDO-33101 1]

9.1.1 Fuel Thermal Margin Events

((

9.1.2 Power and Flow Dependent Limits The OLMCPR, LHGR, and/or MAPLHGR thermal limits are modified by a flow factor when Browns Ferry is operating at less than 100% core flow. This flow factor is primarily based upon an evaluation of the slow recirculation increase event. ((

3]

Similarly, the thermal limits are modified by a power factor when Browns Ferry is operating at less than 100% power. ((

))

((

9.13 Loss of Feedwater Flow Event For the LOFW event, adequate transient core cooling is provided by maintaining the water level inside the core shroud above the TAF. A plant specific analysis was performed at EPU 9-2

NEDO-33101 conditions. This analysis assumed failure of the HPCI system and used only the RCIC system to restore the reactor water level. Because of the extra decay heat from EPU, slightly more time is required for the automatic systems to restore water level. Operator action is only needed for long-term plant shutdown. The results of the LOFW analysis show that the minimum water level inside the shroud is 58 inches above the TAF at EPU conditions. After the water level is restored, the operator manually controls the water level, reduces reactor pressure, and initiates RHR shutdown cooling. This sequence of events does not require any new operator actions or shorter operator response times. Therefore, the operator actions for a LOFW transient do not significantly change for EPU.

As discussed in Section 3.8, an operational requirement is that the RCIC system restores the reactor water level while avoiding ADS timer initiation and MSIV closure activation functions associated with the low-low-low reactor water level setpoint (Level 1). This requirement is intended to avoid unnecessary initiations of these safety systems, and is not a safety-related function. The results of the LOFW analysis for Brown Ferry show that the nominal Level 1 setpoint trip is avoided.

The. loss of one FW pump event only addresses operational considerations to avoid reactor scram on low reactor water level (Level 3). This requirement is intended to avoid unnecessary reactor shutdowns. Because the MELLLA region is extended along the existing upper boundary to the EPU RTP, there is no increase in highest flow control line for EPU. Therefore, the results of the loss of one FW pump event are insignificant. A plant-specific evaluation confirms that the level is maintained above Level 3.

9.2 DESIGN BASIS ACCIDENTS This section addresses the radiological consequences of DBAs.

Plant specific radiological dose consequence analyses have been performed for the DBAs at EPU conditions utilizing AST in accordance with 10 CFR 50.67. The results of these analyses for the LOCA, the CRDA, the FHA, and the MSLB are provided in the AST license amendment submittal (Reference 3). The calculated doses remain within applicable regulatory acceptance criteria.

The ILBA analysis was also performed at EPU conditions utilizing AST. The radiological consequences of this event remain bounded by the other postulated line breaks.

9.3 SPECIAL EVENTS 9.3.1 Anticipated Transient Without Scram The overpressure evaluation includes consideration of the most limiting RPV overpressure case.

((

A LOOP does not result in a reduction in the RHR SPC capability relative to the MSIVC and PRFO cases. With the same RHR SPC capability, the containment response for the MSIVC and PRFO cases bound the LOOP case.

9-3.

NEDO-33101 Browns Ferry meets the ATWS mitigation requirements defined in 10 CFR 50.62:

a. Installation of an ARI system;

b. Boron injection equivalent to 86 gpm; and

c. Installation of automatic RPT logic (i.e., ATWS-RPT).

In addition, plant-specific ATWS analysis is performed to ensure that the following ATWS acceptance criteria are met:

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

b. Peak suppression pool temperature less than 281 0 F (Wetwell shell design temperature); and

c. Peak containment pressure less than 56 psig (Drywell design pressure).

The limiting events for the acceptance criteria discussed above are the PRFO event and the MSIVC event.

The ATWS analyses have been performed for 105% OLTP and for EPU RTP to demonstrate the effect of the EPU on the ATWS acceptance criteria. There is no change to the assumed operator actions for the EPU ATWS analysis, and there is no change to the required hot shutdown boron weight. The key inputs to the Browns Ferry ATWS analysis are provided in Table 9-3.

The analysis was performed using the ODYN code. The results of the analysis are provided in Table 9-4.

The results of the ATWS analysis meet the above ATWS acceptance criteria. Therefore, the response to an ATWS event at EPU is acceptable.

Coolable core geometry is ensured by meeting the 22000 F peak cladding temperature and the 17% local cladding oxidation acceptance criteria of 10 CFR 50.46. ((

9.3.1.1 ATWS with Core Instability The effects of an ATWS with core instability event occurs at natural circulation following a recirculation pump trip. It is initiated at approximately the same power level as before EPU, because the MELLLA upper boundary is not increased. The core design necessary to achieve EPU operations may affect the susceptibility to coupled thermal-hydraulic/neutronic core oscillations at the natural circulation condition, but would not significantly affect the event progression.

94

NEDO-33101 Several factors affect the response of an ATWS instability event, including operating power and flow conditions and core design. The limiting ATWS core instability evaluation presented in References 4 and 5 was performed for an assumed plant initially operating at OLTP and the MELLLA minimum flow point. ((

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

JI The conclusion of Reference 5 and the associated NRC SER that the analyzed operator actions effectively mitigate an ATWS instability event are applicable to the operating conditions expected for EPU.

Initial operating conditions of FWHOOS and FFWTR do not significantly affect the ATWS instability response reported in References 4 and 5. The limiting ATWS evaluation assumes that all FW heating is lost during the event and the injected FW temperature approaches the lowest achievable main condenser hot well temperature. The minimum condenser hot well temperature is not affected by FWHOOS or FFWTR. Thus, as compared to the event initiated from a normal FW temperature condition, the event initiated from either the FWHOOS or FFWTR condition would have less moderator reactivity insertion based on a smaller temperature difference between the initial and final FW temperatures. Therefore, the power oscillation for FWHOOS or FFWTR is expected to be no worse than for the normal temperature condition.

9.3.2 Station Blackout SBO was reevaluated using the guidelines of NUMARC 87-00. The plant response to and coping capabilities for an SBO event are affected slightly by operation at EPU, due to the increase in the initial power level and decay heat. Decay heat was conservatively evaluated assuming end-of-cycle (24-month) for GE-14 fuel. There are no changes to the systems and equipment used to respond to an SBO, nor is the required coping time changed.

Areas containing equipment necessary to cope with a SBO event were evaluated for the effect of loss-of-ventilation due to an SBO. The evaluation shows that equipment operability is bounded due to conservatism in the existing design and qualification bases. The battery capacity remains adequate to support HPCI/RCIC operation after EPU. Adequate compressed gas capacity exists to support the MSRV actuations.

The current CST inventory reserve (135,000 gal.) for HPCI/RCIC use ensures that adequate water volume is available to remove decay heat, depressurize the reactor and maintain reactor vessel level above the top of active fuel (approximately 122,000 gal. required). Peak containment pressure and temperature remain within design bases. Consistent with the DBA-LOCA condition, the required NPSH margin for the RHR pumps has been evaluated (see Section 4.2.5), and a component acceptability review has been completed (see Section 3.9).

Based on the above evaluations, Browns Ferry continues to meet the requirements of 10 CFR 50.63 after the EPU.

9-5

NEDO-33101

9.4 REFERENCES

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

2. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2), Licensing Topical Reports NEDC-32523P-A, Class III, February 2000; NEDC-32523P-A, Supplement 1 Volume I, February 1999; and Supplement 1 Volume II, April 1999.

3. TVA Letter, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - License Amendment

- Alternative Source Term," R08 020731 649, July 31, 2002, including Tech. Spec. No. 405 (TVA-BFN-TS-405).

4. GE Nuclear Energy, "ATWS Rule Issues Relative to BWR Core Thennal-Hydraulic Stability," NEDO-32047-A.

5. GE Nuclear Energy, "Mitigation of BWR Core Thermal-Hydraulic Instabilities in ATWS,"

NEDO-32164.

9-6

NEDO-33101 Table 9-1 Browns Ferry Unit 1 Parameters Used for Transient Analysis Parameter Base UFSAR EPU Rated Thermal Power (MWt) 3293 3952 Analysis Power (% Rated) 100.0 / 104.5 100 / 102 Analysis Dome Pressure (psia) 1020 1050 Analysis Turbine Pressure (psig) 960 973 2.3 Rated Vessel Steam Flow (Mlb/hr) 13.37 16.44 Analysis Steam Flow (% Rated) 105.0 100.0 Rated Core Flow (Mlb/hr) 102.5 102.5 Rated Power Core Flow Range (% Rated) 100 - 105 99 - 105 Maximum Core Flow 4 (Mlb/lr) 107.6 107.6 Normal Feedwater Temperature (F) 376.0 394.5 Feedwater Temperature Reduction (AT IF) 0.0 54.7 Steam Bypass Capacity (% Rated Stearnflow) 26.2 21.7 Reactor Low Water Level 3 Scram 538 518 (inches above vessel zero)

No. of MSRVs assumed in the analysis 12 12 5 MCPR Safety Limit 1.07 1.08

1. GEMINI analysis at 100% and at 102% when required to be consistent with the Browns Ferry UFSAR

2. EPU analysis based on calculated steam line pressure drop.

3. EPU input value represents the conservative value (lowest steamline pressure) for all three BFN units.

4. All Analysis at maximum core flow unless explicitly noted otherwise.

5. A low pressure bank setpoint MSRV is assumed OOS for transient analysis.

9-7

NEDO-33101 Table 9-2 Browns Ferry Unit 1 Transient Analysis Results Peak OLMCPR Neutron Peak Heat Event Flux (% of Flux (% of ACPR Option A Option B Rated Rated EPU) EPU)

Generator Load Rejection with 680 133 0.26 1.42 1.39 Bypass Failure Turbine Trip With Bypass Failure 673 132 0.26 1.42 1.39 Feedwater Controller Failure Max 629 136 0.26 1.41 1.38 Demand Feedwater Controller Failure with 742 141 0.31 1.47 1.44 TBOOS Pressure Regulator Downscale (l) (l) (l) (l) (l)

Failure Loss of Feedwater Heating (2) (2) 0.13 1.21 Inadvertent HPCI Actuation 112 109 0.06 1 (

Rod Withdrawal Error (2) 0.19 (4) 1.27 Slow Recirculation Increase (5) (5) (5) MCPRf Fast Recirculation Increase 181 94 0.14 (6) (6)

Generator Load Rejection with 590 129 0.22 1.37 1.34 Bypass _7 MSIV Closure - All Valves 123 100 0.03 (7)

MSIV Closure - One Valve 130 106 0.06 (7)

Loss of Feedwater Flow 100 100 (5) (5) (5)

Loss of One Feedwater Pump 100 100 (5) (5) (5)

1. Not required, based on UFSAR 14.5.2.8.

2. Peak neutron flux and peak heat flux are not reported for the slow transients.

3. HPCI is bounded by Loss of Feedwater Heating.

4. With rod block monitor setpoint of 111%.

5. . Not applicable.

6. Fast recirculation increase is bounded by off-rated limits.

7. Bounded by the Generator Load Rejection with Bypass Failure.

9-8

NEDO-33101 Table 9-3 Browns Ferry Key Inputs for ATWS Analysis Input Variable 105% OLTP EPU Reactor power (MWt) 3458 3952 Reactor dome pressure (psia) 1050 1050 MSRV capacity of 13 valves (Mlbm/hr) 11.31 11.31 High pressure ATWS-RPT (psig) 1177 1177 Number of MSRVs OOS 1 1 Table 9-4 Browns Ferry ATWVS Analysis Results Parameter 105% OLTP EPU Peak vessel bottom pressure (psig) 1368 1484 Peak suppression pool temperature (0 F) 214.6 214.1 Peak containment pressure (psig) 21.7 21.4 Peak cladding temperature (0 F) 1476 1453 Local cladding oxidation (%) <17 <17 9-9

NEDO-33101

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-. s SoffaceHeat Flux 32s-e 0tusiW

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Figure 9-1. Browns Ferry Turbine Trip with Bypass Failure (100% EPU RTP & 105% Core Flow) 9-10

NEDO-33101 Vessel PressRise(psi)

Sufet Vs"e Fow

_ Roo VaeF Flow Bypass Valt. F0.

ruo WA V

.Ir

s As 2 la 0 go as IIftt Tkne (e.43 10 to to 4C to to is 15 lot 15 44 04 to Thke (*e) Thow (oc Figure 9-2. Browns Ferry Generator Load Rejection with Bypass Failure (100% EPU RTP & 105% Core Flow)

I I

9-It

NEDO-33101

-Net~n P

-Caow

-CAon Surec Fit Hall

-*Cons khlt F W

-- cam~ kWeE I

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at I.0 non agce of to Ad0 I8. Doe ma Time("c} Time (sec)

Figure 9-3. Browns Ferry Feedwater Controller Failure - Maximum Demand (100% EPU RTP, 105% Core Flow & 394.5 0F Feedwater Temp.)

9-12

NEDO-33101 F iu INeutrAwl

-Core nlet I 0:

5\l1N WI.

to U "A "I5 Tim. (se4)

VI As "af MD 355 45 A l4e 41.. is. 35 Tn.(soc) ax (Se)

Figure 9-4. Browns Ferry Feedwater Controller Failure - Maximum Demand with Bypass QoS (100% EPU RTP, 105% Core Flow & 394.5 0F Feedwater Temp.)

9-13

NEDO-33101

10. OTHER EVALUATIONS 10.1 HIGH ENERGY LINE BREAK HELBs are evaluated for their effects on structures and on equipment qualification. The Browns Ferry Unit I EPU requires a small (< 3%) increase in the RPV dome operating pressure to supply more steam to the turbine. The slight increase in the vessel pressure and temperature will result in a small increase in the mass and energy release rates following HELBs.

10.1.1 Temperature, Pressure and Humidity Profiles The HELB analysis evaluation was made for all systems evaluated in the UFSAR. The evaluation shows that the affected building and cubicles that support the safety-related function are designed to withstand the resulting pressure and thermal loading following a HELB. The equipment and systems that support a safety-related function are also qualified for the environmental conditions imposed upon them.

At EPU, HELBs outside the primary containment can cause the sub-compartment pressure and temperature profiles to increase compared to lower power levels. The relative humidity change is negligible. The effects of HELBs were not calculated for Unit I at OLTP conditions, but have been calculated at EPU conditions, including the effect of the 30 psi in reactor dome pressure increase. However, a representative comparison of the effects from the increase in reactor power for EPU can be demonstrated based on the results of the Browns Ferry Units 2 and 3 calculations for 105% OLTP and 120% OLTP, which were both performed considering the increased reactor dome pressure. Table 10-1 shows a comparison of results for 105% OLTP and 120% OLTP. As can be seen, the increase in blowdown rate at 120% OLTP is small and thus the resulting temperature and pressure profiles are bounded by the profiles for 105% OLTP. The methodology used to obtain the Unit I EPU results is consistent with the methodology used to obtain Unit 2 and Unit 3 EPU results. The Unit 1 EPU results, which are used to establish safety and design limits, are in excellent agreement with the Unit 2 and Unit 3 EPU results. The effects of HELBs in the reactor building have been calculated for breaks in MS, FW and RWCU as well as the RCIC and HPCI steam lines.

The effects of the Unit 1 reactor building subcompartment pressure, temperature and relative humidity on equipment qualification is discussed in Section 10.3.

10.1.2 Pipe Whip and Jet Impingement Pipe whip and jet impingement loads resulting from high energy pipe breaks are directly proportional to system pressure. Because the operating pressure increase is small (< 4%), pipe whip and jet impingement loads do not significantly change. Existing calculations for existing whip restraints and jet impingement shields have been reviewed and determined to be adequate for EPU conditions. Acceptance criteria used to evaluate existing jet impingement targets did not require quantitative evaluations contingent on the internal pipe pressure for all targets. To the extent that pipe pressure was used in quantitative evaluations and based on the methodology used for qualitative assessments, the small change in operating pressure would not affect the conclusions of the jet impingement evaluations. Walkdowns and evaluations for whip restraints and impingement targets for current modifications are in progress with completion of 10-1

NEDO-33101 confirmatory walkdowns scheduled after modifications are completed. All pipe rupture evaluations will be done prior to EPU implementation.

10.1.3 Internal Flooding from High Energy Line Breaks The only high energy liquid filled lines in the reactor building are RWCU and FW. The mass release for the RWCU breaks were calculated using the 30 psi reactor pressure increase. The resulting flood levels were determined for the affected flood areas. The safety-related equipment was evaluated for effects as described in Section 10.3.

The flooding from a FW line break is dependent upon the maximum water levels in the hotwells and not reactor vessel conditions. Feedwater system hardware changes have been evaluated and the flooding rate from a FW line break is acceptable. Because the water level in the hotwells, the existing draining systems and existing flood barriers are not changing, the existing FW break flooding analysis is valid for the EPU condition.

10.2 MODERATE ENERGY LINE BREAK MELBs are evaluated for their effects on EQ.

System design limits (design pressure) used as input to the MELB flooding analyses are not changed by EPU. Therefore, the MELB internal flooding evaluations are not affected by the EPU and the design change process ensures continued evaluation of all changes for effect on MELB flooding.

10.3 ENVIRONMENTAL QUALIFICATION Safety-related components are required to be qualified for the environment in which they are required to operate. Table 10-2 provides a listing of the parameter used in EQ.

103.1 Electrical Equipment The safety-related electrical equipment was reviewed to ensure the existing qualification for the normal and accident conditions, expected in the area where the devices are located, remain adequate.

10.3.1.1 Inside Containment EQ for safety-related electrical equipment located inside the containment is based on MSLB and/or DBAILOCA conditions and their resultant temperature, pressure, humidity and radiation consequences, and includes the environments expected to exist during normal plant operation.

Normal temperatures are.expected to increase slightly, but remain bounded by the normal temperatures used in the EQ analyses. The post-accident peak temperature and pressure do not significantly increase for EPU. The long-term post-accident temperatUres inside containment increase. However, the increase in long-term post-accident temperatures was determined not to adversely affect the qualification of safety-related electrical equipment.

The current radiation levels under normal plant conditions were conservatively evaluated to increase in proportion to the increase in reactor thermal power. The accident radiation levels increase by < 14% above the OLTP levels. The total integrated doses (normal plus accident) for EPU conditions were determined not to adversely affect qualification of the equipment located 10-2

NEDO-33101 inside containment. The increased radiation doses resulted in a reduction of the radiation life of some solenoids located inside containment. However, the qualified life based on thermal aging is shorter than the radiation life for these solenoids. Therefore, the equipment qualified life was not reduced due to the increased radiation doses.

10.3.1.2 Outside Containment Accident temperature, pressure and humidity environments as well as flood levels used for qualification of equipment located in harsh environments outside containment result from an MSLB, or other HELBs, whichever is limiting for each plant area. The HELB temperature and pressure profiles were determined for EPU conditions. The accident temperature in the reactor building resulting from a LOCA/MSLB inside containment has been determined. The long-term post-accident temperatures were evaluated and used to establish the qualification of safety-related electrical equipment. The normal humidity and pressure conditions do not change as a result of EPU. The normal temperature changes are discussed in Section 6.6.

The current radiation levels under normal plant conditions were conservatively evaluated to increase in proportion to the increase in reactor thermal power. The accident radiation levels increase by < 14% above the OLTP levels. The total integrated doses (normal plus accident) for EPU conditions were evaluated and used to establish the qualification of the equipment located outside of containment.

10.3.2 Mechanical Equipment With Non-Metallic Components The changes to normal and post accident ambient conditions for safety-related equipment, as a result of EPU conditions, are discussed in Section 10.3.1. Evaluation of the safety-related mechanical equipment with non-metallic components identified some equipment potentially affected by the EPU conditions. These effects were evaluated and determined not to have an adverse effect on the functional capability of non-metallic components in the mechanical equipment both inside and outside containment.

10.3.3 Mechanical Component Design Qualification The process fluid operating conditions of equipment/components (pumps, heat exchangers, etc.)

in certain systems are affected by operation at EPU due to slightly increased temperatures, pressure, and in some cases, flow. The effects of increased fluid induced loads on safety-related components are described within Sections 3 and 4.1. Increased nozzle loads and component support loads, due to the revised operating conditions, were evaluated within the piping assessments within Section 3. These increased loads are insignificant, and become negligible (i.e., remain bounded) when combined with the governing dynamic loads. Therefore, the mechanical components and component supports are adequately designed for EPU conditions.

10.4 TESTING Compared to the initial startup program, ([

)) EPU requires only a limited subset of the original startup test program. As applicable to this plant's design, testing for EPU is consistent with the descriptions in Section 5.11.9 and Appendix L, Section L.2 of ELTRI. Specifically, the following testing will be performed during the initial power ascension steps for EPU:

10-3

NEDO-33101

1. Testing will be performed in accordance with the TS Surveillance Requirements on instrumentation that is re-calibrated for EPU conditions. Overlap between the IRM and APRM will be assured.

2. Steady-state data will be taken during power ascension beginning at 90% OLTP and continuing at each EPU power increase increment. These data will allow system performance parameters to be projected through the EPU power ascension.

3. EPU power increases above the 100% OLTP will be made along an established flow control/rod line in increments of equal to or less than 5% power. Steady-state operating data, including fuel thermal margin, will be taken and evaluated at each step. Routine measurements of reactor and system pressures, flows, and vibration will be evaluated from each measurement point, prior to the next power increment. Radiation measurements will be made at selected power levels to ensure the protection of personnel.

4. Control system tests will be performed for the reactor FW/reactor water level controls, pressure controls, and recirculation flow controls, if applicable. These operational tests will be made at the appropriate plant conditions for that test at each of the power increments, to show acceptable adjustments and operational capability.

5. Testing will be done to confirm the power level near the turbine first-stage scram bypass setpoint.

6. A test specification is being prepared which identifies the EPU tests, the associated acceptance criteria and the appropriate test conditions. All testing will be done in accordance to written procedures as required by 10 CFR 50, Appendix B, Criterion XI.

The same performance criteria will be used as in the original power ascension tests, unless they have been replaced by updated criteria since the initial test program. ((

))

For EPU, Browns Ferry does not intend to perform large transient testing for involving an automatic scram from a high power. Transient experience at high powers at operating BWR plants, including Browns Ferry Units 2 and 3, has shown-a close correlation of the plant transient data to the evaluated events. The operating history of Browns Ferry demonstrates that previous transient events from full power are within expected peak limiting values. The transient analyses demonstrate that safety criteria are met, and that this uprate does not cause any previous non-limiting events to become limiting. Based on the similarity of plants, past transient testing, past analyses, and the evaluation of test results, the effects of the EPU RTP level can be analytically determined on a plant specific basis. No new systems or features were installed for mitigation of rapid pressurization anticipated operational occurrences for this EPU. A scram from high power level results in an unnecessary and undesirable transient cycle on the primary system. Therefore, additional transient testing involving a scram from high power levels is not justifiable. Should any future large transients occur, Browns Ferry procedures require verification that the actual plant response is in accordance with the predicted response. Plant event data recorders are capable of acquiring the necessary data to confirm the actual versus expected response.

Further, the important nuclear characteristics required for transient analysis are confirmed by the steady state physics testing. Transient mitigation capability is demonstrated by other tests 10-4

NEDO-33101 required by the TS. In addition, the limiting transient analyses are included as part of the reload licensing analysis.

10.4.1 Recirculation Pump Testing Vibration testing of the recirculation pumps is not required, because there is no change in the maximum core flow. To maintain the same core flow with the increased core pressure drop (due to an increase in steam production), recirculation flow (drive flow) increases slightly (< 3%).

The "containment noise" observed in a BWR/4 -251 in 1994 is not expected at Browns Ferry.

At that plant, an increase in containment noise and vibration levels during plant operation was observed at increased recirculation pump speeds. Based on test results, the utility concluded that the increased noise was a direct result of the RHR check valve not being properly seated. The testing demonstrated that the containment noise levels were greatly attenuated when the RHR check valve was properly seated. Thus, this phenomenon is unrelated to EPU and no containment noise is expected due to EPU.

10.4.2 10 CFR 50 Appendix J Testing The Browns Ferry 10 CFR 50 Appendix J test program is required by the Technical Specifications. This test program periodically pressurizes the containment (Type A test), the containment penetrations (Type B test), and the containment isolation valves and test boundary (Type C tests) to the calculated peak containment pressure (P.), and measures leakage. Resulting from the EPU, P. changes to 48.5 psig, as shown in Table 4-1. Therefore, the 10 CFR 50 Appendix J test program is revised to reflect this calculated peak containment pressure value.

10.4.3 Main Steam Line, Feedwater and Reactor Recirculation Piping Flow Induced Vibration Testing Piping for the MS, FW and RRS will be monitored for vibrations during initial plant operation for the new EPU operating conditions. This test program will show that these piping systems are vibrating at acceptable levels during initial plant operation at the EPU conditions.

The MS and FW piping systems are normally affected by an EPU, because their mass flow rates and operating pressures usually increase at EPU. The mass flow rates in these systems typically increase in proportion to the EPU power level increase. The flow induced vibration levels simultaneously increase in proportion to the increase in the fluid density and the square of the fluid velocity at these higher mass flow rates.

The RRS piping is affected by several modifications for the Unit 1 restart, and there is a small recirculation drive flow increase for EPU, and thus, vibration monitoring will also be performed on this system.

The MS, FW. and RRS piping inside containment will be monitored with remote vibration sensors. Also, the MS and FW piping outside containment will be monitored with remote sensors or hand-held instruments. The vibration monitoring devices will be located on portions of the piping system determined to be most susceptible to vibration.

Acceptable vibration criteria will be established for these locations prior to testing. Vibration monitoring of these piping systems will be taken at power levels below the final, maximum extended power level. Vibration data is typically collected at 50%, 75%, 100%, 105%, l 10%,

10-5

NEDO-33101 115% and 120% of OLTP. The measured vibration levels at each power level will be compared to the acceptance criteria to verify the piping is below the acceptance criteria prior to moving to the next power level. In this manner, the vibration monitoring testing can proceed as Browns Ferry Unit 1 operates for the first time at each new power level, and at the same time avoid the remote possibility of incurring high vibrations and damaging the plant equipment (piping),

before appropriate corrective actions can take place.

10.5 INDIVIDUAL PLANT EVALUATION PRAs are performed to evaluate the risk of plant operation.

The Browns Ferry Unit I PRA was performed incorporating analyses inputs, results, and modifications associated with the restart of Unit 1 at EPU power level. The Unit 1 PRA assumes that Units 2 and 3 are also operational at EPU power levels.

To ensure that risk-significant effects of Unit 1 operation at EPU conditions are represented in the Unit 1 PRA, all associated plant modifications were systematically reviewed to identify their effect on the elements of a risk assessment. Specifically, modifications were reviewed with respect to their potential effect on the PRA model.

Regulatory Guide 1.174 provides the guidance framework for using PRA in risk-informed decisions for plant-specific changes to the licensing basis. The acceptance guidelines consider both the magnitude and size of the changes to CDF and LERF. The direct applicability of Regulatory Guide 1.174 to Browns Ferry Unit 1 is uncertain as the Unit 1 PRA represents the base case rather than a change from a preexisting base case. Nevertheless, the guidance offered by Regulatory Guide 1.174 does offer a framework in assisting in the interpretation of the numerical results of the PRA.

As stated in Regulatory Guide 1.174,

• "When the calculated increase in CDF is in the range of 10.6 per reactor year to 1O's per reactor year, applications will be considered only if it can be reasonably shown that the total CDF is less than 104 per reactor year (Region II)."

• "When the calculated increase in LERF is in the range of 107 per reactor year to 106 per reactor year, applications would be considered only if it can be reasonably shown that the total CDF is less than 10'5 per reactor year (Region II)."

As shown in Table 10-3, the total CDF for Browns Ferry Unit 1 is below the guideline value of 104 per reactor year. Also, as shown in Table 10-3, the total LERF for Browns Ferry Unit 1 is below the guideline value of 10'5 per reactor year.

An uncertainty analysis was performed to more fully describe the core damage frequency. The mean core damage frequency is 1.66E-6. The fifth percentile is 5.13E-7 while the ninety-fifth percentile is 4.65E-6. The ratio of the ninety-fifth percentile to the median for this distribution is 3.9.

The plant-specific MAAP model was used to support the system success criteria determination and sequence timing. The RISKMAN"h integrated PRA computer code was used to perform the necessary data and system analyses and to represent the response of the operators and plant 10-6

NEDO-33101 systems to the initiators considered. The EPRI HRA Calculator was used to quantify all operator actions considered in the Unit I PRA.

10.5.1 Initiating Event Frequency Forty initiating event categories are considered in the Unit 1 PRA: twenty-one transient initiator categories; thirteen LOCA initiator categories; and six internal flooding initiator categories. The initiator categories for Unit 1 are similar to the ones used in the Units 2 and 3 PRAs. The Unit 1 initiator list is an expansion of the list used for Units 2 and 3. The Unit 1 PRA was conducted following the guidance of the ASME PRA Standard and DG 1122. The initiator categories for the Unit 1 PRA are defined at a more detailed level as compared to the Unit 2 and Unit 3 PRAs, thus facilitating scrutiny as encouraged by the ASME Standard. Specifically, the changes facility the tracing of success criteria in the event model.

The initiating event category occurrence frequencies have been updated for the Unit 1 PRA reflecting plant operational experience gained since the conduct of the Unit 2 and Unit 3 PRAs.

For this reason, the initiating event frequencies used in the Unit 1 PRA differ slightly from those used in the Unit 2 and Unit 3 PRAs.

One initiator category that appears in the Unit 2 and Unit 3 PRAs that was eliminated in the Unit 1 PRA is "inadvertent opening of two SRVs." A review of the SRV design found no credible mechanism for the inadvertent opening of two SRVs. A review of industry operational experience supported the elimination of this initiator category from further consideration.

The EPU does not result in plant equipment operation beyond the design ratings and conditions.

The transient categories with the most potential to be affected by the EPU are those associated with trip set points, such as reactor scram, system isolations, and operating equipment trips. A review of these conditions concluded that the operational margins remain within values consistent with the PRA models regarding initiating event category frequencies and that changes are not required to reflect EPU conditions.

The frequency of loss of offsite power events (either "loss of 500 kV to the plant" or "loss of 500 kV to one unit") due to grid instabilities is not affected by EPU. In addition to the loss of offsite power represented by these two initiators, there is the potential that the grid is lost following the trip of a unit. Rapid separation of a large generating unit from the grid has the potential to cause Browns Ferry grid instability and loss of offsite power. This possibility is represented by a unique top event. A grid stability analysis has been performed, considering the increase in electrical output, to demonstrate conformance to General Design Criteria 17. In addition to the normal configuration, the analysis considers various transmission system contingencies. As a result of this analysis, TVA will continue compensating electrical generation operation limits to ensure that the probability of induced grid loss due to the trip of a Browns Ferry unit is not increased following implementation of EPU.

The Unit 1 PRA identifies four initiators as a result of the loss or degradation of support systems:

loss of I&C bus A; loss of I&C bus B; loss of plant air; and loss of RCW. The duty on these systems is essentially unchanged as a result of the EPU. Therefore the frequency of these initiators does not change under EPU conditions. The Unit 2 and Unit 3 PRAs identify a fifth support system initiator: loss of RBCCW. At this time of the development of the Unit 2 and 10-7

NEDO-33101 Unit 3 PRAs, loss of RBCCW would result in loss of drywell control air. This dependency will be removed prior to the restart of Unit 1.

The nominal RPV pressure is increased 30 psi. The LOCA frequencies used in the PRA are based on NUREG/CR-5750, Rates of Initiating Events at US Nuclear Power Plants:. 1987-1995.

The small break LOCA frequency is estimated from available US operational experience combined in a Bayesian update of the frequency presented in WASH-1400. Larger break frequencies rely on through-wall crack analyses that are then combined with a conditional probability of a through-wall crack transitioning into a rupture. The initiator frequency estimation process does not explicitly involve consideration of pressure. Hence the 30 psi increase in operating pressure did not factor into the estimation of LOCA frequencies. It is noted that any impact on LOCA frequencies due to such a small increase in pressure would be much smaller than the uncertainty expressed in these frequencies.

ATWS is modeled as five separate initiating events in the Browns Ferry Unit 1 PRA. In the Unit 2 and Unit 3 PRAs, ATWS scenarios are addressed within the event sequence models for individual transient initiator categories. This modeling approach difference was made in order to provide models that more clearly communicate their underlying logic as encouraged by the ASME Standard.

Reactor Protection System reliability is taken from NUREG/CR-5500. The total frequency of ATWS is determined by multiplying the frequency of the transient events by the likelihood of failure to scram. As discussed above, the frequency of occurrence of the transient initiators is conservatively assumed not to change after the modifications are implemented. No restart or EPU modifications are anticipated to affect the likelihood of scram.

The initiators identified for the Unit 1 PRA are presented in Table 10-4. Table 10-4 also summarizes the mean CDF and LERF contributions for the Unit 1 initiator groups.

10.5.2 Component and System Reliability No increase in component failure rates is anticipated as a result of operation at EPU conditions.

Under EPU conditions, equipment operating limits, and/or ratings are not exceeded. Existing plant component monitoring programs detect degradation if it occurs and corrective action is taken in a timely manner. It is possible that EPU conditions may result in selected components requiring refurbishment or replacement more frequently; however, the functionality and reliability of components and systems is maintained at the current standard.

The component unavailability database has been updated for the Unit 1 PRA reflecting plant operational experience gained since the conduct of the Unit 2 and Unit 3 PRAs. For this reason, the component database used in the Unit 1 PRA differs slightly from that used in the Unit 2 and Unit 3 PRAs.

One significant difference between the Unit 1 PRA and the Unit 2 and Unit 3 PRAs is found in the modeling of common cause failures between the HPCI and RCIC systems. All three PRAs include the consideration of the possibility of common cause failures between these two systems.

In the Unit I PRA, however, a detailed evaluation of the common cause coupling resulted in a lower likelihood of failure of both systems.

10-8

NEDO-33101 A second significant difference is found in the data developed to represent the diesel generators.

The additional operational experience reflected in the Unit 1 PRA as compared to the Unit 2 and Unit 3 PRAs resulted in lower assessed unavailabilities for the diesel generators.

10.5.3 Operator Response Operator response capabilities were analyzed for the Unit I PRA. The assessment of operator response conforms to the guidelines for Human Reliability Assessment (HRA) contained in Table 4.5.5 of ASME RA-S-2002. The EPRI HRA CalculatorTM was used to quantify both the pre-initiator and post-initiator operator actions associated with activities that pass screens as potentially significant to risk. The evaluation of operator response for the Unit I PRA represents a complete reevaluation relative to the assessment conducted for the Unit 2 and Unit 3 PRAs.

The operator response analysis addresses human activities that have a potential for causing the failure that could contribute to the prevention of core damage or release of radioactive materials should core damage occur. These include the following types of actions.

1. Specific routine pre-initiator activities that, if not completed correctly, may impact the availability of equipment necessary for the successful performance of system safety functions following the initiating events in the PRA model. These include:

• Undetected misalignment of a SSC

• Calibration error that has an adverse impact on the automatic initiation of standby safety equipment.

• Work practices that could introduce a mechanism that simultaneously affects equipment in different trains of either a redundant system or diverse systems.

2. Post-initiating event or dynamic actions identify the set of operator responses required within the accident sequences to initiate (for those systems not automatically initiated), operate, control, isolate, or terminate those systems and components used in preventing or mitigating core damage as defined by the success criteria for the plant conditions at the time the action is needed.

3. Recovery actions take credit for those actibns performed by the on-shift personnel either in response to procedural direction or as skill-of-the-craft to recover a failed function, system or component that is used in the performance of a response action in dominant sequences.

Human activities could also lead to an initiating event, but they are not quantified explicitly.

Rather, any actions that have initiated a plant transient are included in the data set that contributes to the frequency of the applicable initiating event. This directly accounts for the extensive operating history of the plant or group of plant upon which the frequency is based.

The most significant difference in the evaluation of operator actions for the Unit 1 PRA as compared to the Unit 2 and Unit 3 PRAs is associated with the depressurization of the vessel following failure of high pressure level control. The assessment to support the Unit 1 PRA resulted in the decrease in the failure likelihood of this action by approximately a factor of 2.

Table 10-5 lists the operator actions with the largest contributions to core damage frequency as measured by fractional importance values. Actions with fractional importance values greater 10-9

NEDO-33101 than 1E-6 are listed. The action with the largest fractional importance is the failure of the operators to depressurize the vessel given failure of high pressure injection systems to maintain vessel level.

10.5.4 Success Criteria The response to an initiator is represented in the PRA models by a set of discrete requirements for the operation of individual systems and the performance of specific operator actions. These scenario-specific requirements define the success criteria for system operation and operator actions to fulfill the critical safety functions necessary to maintain the reactor fuel in a safe condition. The critical safety functions are reactivity control, RPV pressure control, containment pressure control, and RPV inventory makeup. These individual criteria were reassessed at the EPU RTP conditions and increased decay heat.

The scram function is not affected by EPU because operation of the CRD system and the RPS is not affected by the EPU.

The thermal hydraulic analyses supporting the success criteria were performed using MAAP 4.0.

The impact of EPU on the system success criteria for Unit I was identical to the impact for Units 2 and 3. Specifically, the only change in success criteria for EPU conditions as compared to power conditions reflecting operation at 105% of the OLTP is associated with level control using enhanced CRD flow. Analyses of EPU conditions conclude that enhanced CRD system operation (i.e., operation of both pumps for vessel injection) alone is not sufficient to prevent fuel damage if the RPV remains at high pressure. If the vessel is successfully depressurized within six hours following a successful scram, enhanced CRD system operation is sufficient to provide successful RPV level control.

Analyses performed for Unit 1 did indicate that, following a transient initiator and loss of all high pressure injection, operation of a single SRV would depressurize the vessel in a timely manner allowing successful level control by low pressure injection systems. This criterion is implemented in the Unit 1 PRA for transient response, but not in the Unit 2 or Unit 3 PRAs.

The Unit 1 PRA also incorporates a different model for the battery boards as compared to the Unit 2 and Unit 3 PRAs. In the latter, it was assumed that the battery charger alone would not be sufficient to provide power to the boards under load. In the Unit 1 PRA, additional plant information was used to allow credit for the chargers acting alone to provide power to these boards.

10.5.5 External Events The effect of the restart and EPU was reviewed to determine whether any new plant vulnerabilities exist from the occurrence of internal fires, seismic events, and other external events. Equipment changes associated with the EPU are minor and do not affect reliability. The EPU does not affect any existing structures or fire zones and therefore no new vulnerabilities are introduced.

The Browns Ferry IPEEE (Reference 2) and Seismic IPEEE Report (Reference 3) were reviewed to determine whether there were any existing conditions where the EPU could introduce new vulnerabilities.

10-10

NEDO-33101 The IPEEE review concluded that there are no new fire-induced vulnerabilities associated with the EPU. The fire zones, fire loading, and safe shutdown paths for Browns Ferry do not change for EPU; therefore there is no increase in the vulnerability to internal fires.

Because the EPU modifications do not affect the structures or component anchoring, no new vulnerabilities are introduced as a result of a seismic event.

The IPEEE states that the Browns Ferry Plant/Facilities design is robust in relation to the 1975 SRP Criteria and a walk-down did not reveal any potential significant vulnerability that were not included in the original design basis analysis. Because there are no external or other structural changes associated with the EPU, there are no new vulnerabilities introduced from wind or flood events.

There are no changes in the EPU that could be affected by transportation or nearby facility accidents, thus there are no new vulnerabilities introduced from transportation and nearby facility accidents.

10.5.6 Shutdown Risk A PRA model to evaluate shutdown risk, specifically CDF or LERF, has not been developed for Browns Ferry; however simplified risk evaluation tools are utilized. Browns Ferry utilizes a defense-in-depth approach to managing risk during plant shutdowns. To assist in the management of risk during shutdowns, TVA uses EPRI's computer code, ORAM. This process specifically monitors the safety functions: shutdown cooling; fuel pool cooling; inventory control; offsite AC power; onsite AC power; primary and secondary containment; and reactivity control.

EPU increases the amount of decay heat following shutdown, which has the greatest effect on RHR capability. Prior to each outage, a decay heat prediction based on best-estimate values (i.e.

no statistical uncertainty or added conservatism applied) is determined. This decay heat prediction is used to create a fuel pool "time to boil off curve" which is then used by the outage planning team to ensure that heat removal systems are available and that contingency plans are made for maintenance and testing of systems. The incremental decay heat due to the EPU will slightly extend the time that the existing RHR systems will need to remain in service during plant shutdown and remain available right after shutdown.

The Browns Ferry TS and TRM address the above requirements regarding shutdown risk management concepts. Browns Ferry procedures provide complete and consistent implementation of the steps required to ensure that shutdown risks attributes are controlled.

Therefore, EPU has no effect on the process controls for shutdown risk management and a negligible effect on the overall ability of Browns Ferry to adequately manage shutdown risk.

10.5.7 Probabilistic Risk Assessment Quality The Unit I PRA was built on more than 10 years of analysis effort associated with the Unit 2 and 3 PRAs. The Unit 2 and 3 PRAs are maintained and were updated as recently as early 2003.

TVA procedures provide the details describing the use of the PRA at Browns Ferry to support the Maintenance Rule. The PRA assists in establishing performance criteria, balancing unavailability and reliability for risk significant SSCs and goal setting and provides input to the 10-11

NEDO-33101 onsite Expert Panel for the risk significance determination process when revisions to the PRA take place. Functions are potentially considered risk significant if any of the following conditions are satisfied:

• Functions modeled in the level 1 PRA are found to have a risk achievement worth greater than or equal to 2.0;

• Functions modeled in the level 1 PRA are found to have a risk reduction worth of less than or equal to 0.995; or

• Functions modeled in the level 1 PRA are found to have a cumulative contribution of 90% of the CDF.

Because the PRAs are actively used at Browns Ferry, a formal process is in place to evaluate and resolve PRA model-related issues as they are identified. The PRA Update Report is evaluated for updating every other refueling outage. The administrative guidance for this activity is contained in TVA Procedures.

During November 1997, TVA participated in a PRA Peer Review Certification of the Browns Ferry Unit 2 and 3 PRAs administered under the auspices of the BWROG Peer Certification Committee. The purpose of the peer review process is to establish a method of assessing the technical quality of the PRA for its potential applications.

The Peer Review evaluation process utilized a tiered approach using standardized checklists allowing for a detailed review of the elements and the sub-elements of the Browns Ferry PRA to identify strengths and areas that needed improvement. The review system used allowed the Peer Review team to focus on technical issues and to issue their assessment results in the form of a "grade" of 1 through 4 on a PRA sub-element level. To reasonably span the spectrum of potential PRA applications, the four grades of certification as defined by the BWROG document "Report to the Industry on PSA Peer Review Certification Process - Pilot Plant Results" were employed.

The Browns Ferry Unit 2 and 3 Peer Review presented in Table 10-6 resulted in a consistent evaluation across all elements and sub-elements. Also, during the Unit 2 and 3 PRA updates in 2003, all significant findings (i.e., designated as Level A or B) from the Peer Certification were resolved, resulting in all PRA elements now having a minimum certification grade of 3. The Unit 1 PRA has incorporated the findings of the Units 2 and 3 PRAs Peer Review. Revision of the Unit 1 PRA to resolve the findings of the Units 2 and 3 Peer Review represents the equivalent of an independent Peer Review for the Unit 1 PRA. The previous conducted Peer Review was effectively an administrative and technical Peer Review of the Unit 1 PRA. Similar models, processes, policies, approaches, reviews and management oversight were utilized to develop the Unit 1 PRA.

Additionally, the Unit 1 PRA was developed with the explicit goal of meeting the category 2 requirements of the ASME PRA Standard as well as DG-1 122. The Unit 1 PRA has had internal independent review and was conducted under TVA's quality assurance program.

10.6 OPERATOR TRAINING AND HUMAN FACTORS Some additional training is required to enable plant operation at the EPU RTP level. However, the additional training for EPU conditions is minimal.

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NEDO-33101 For EPU conditions, operator responses to transient, accident and special events are not affected.

Most abnormal events result in automatic plant shutdown (scram). Some abnormal events result in automatic RCPB pressure relief, ADS actuation and/or automatic ECCS actuation (for low water level events). EPU does not change any of the automatic safety functions. After the applicable automatic responses have initiated, the subsequent operator actions (e.g., maintaining safe shutdown, core cooling, and containment cooling) for plant safety do not change for EPU.

The analog and digital inputs for the ICS and SPDS will be reviewed to determine the effects from EPU. This includes required changes to monitored points, calculations, alert and trip setpoints. Various changes in EOP curves and limits, if required, will also require an update of the SPDS. Any changes required to the ICS and SPDS computer will be completed prior to startup at EPU conditions.

Following a review of the EPU modifications and identified key procedure changes, recommendations for operator training and simulator changes and a final determination of the operator training needs will be made, consistent with the Browns Ferry training program for selection of modifications for operator training. Any modifications required for EPU will be evaluated for its effect on the ICS and SPDS and any required changes (including any new monitoring points) will be addressed as a part of the modification. Any changes made will be discussed as a part of the operator-training program for EPU.

Training required to operate the plant following EPU will be conducted prior to restart of the unit at the EPU conditions. Data obtained during startup testing will be incorporated into the training as needed. The classroom training will cover various aspects of EPU including changes to parameters, setpoints, scales, procedures, systems and startup test procedures. The classroom training will be combined with simulator training. The simulator training will include, as a minimum, a demonstration of transients that show the greatest change in plant response at EPU RTP compared to OLTP.

Installation of the EPU changes to the Browns Ferry Simulator is planned for approximately three to four months prior to EPU implementation. The two training cycles prior to EPU implementation will complete the recommended operator classroom and simulator training for EPU implementation. The simulator changes will include hardware changes for new or modified control room instrumentation and controls, software updates for modeling changes due to EPU setpoint changes and re-tuning of the core physics model for cycle specific data. The Simulator ICS will also be updated for EPU modifications.

Operating data will be collected during EPU implementation and start-up testing. This data will be compared to simulator data as required by ANSI/ANS 3.5-1985, Section 5.4.1. Simulator acceptance testing will be conducted to benchmark the simulator performance based on design and engineering analysis data as required in ANSIIANS 3.5-1985. ANSI/ANS 3.5 is endorsed by Regulatory Guide 1.149, Revision 3 and 10 CFR 55. The simulator acceptance testing for EPU is planned to be complete within 30 days after steady state operation at 120% OLTP.

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NEDO-33101 10.7 PLANT LIFE 10.7.1 RPV Internal Components The plant life evaluation identifies degradation mechanisms influenced by increases in fluence and flow.

Browns Ferry has a procedurally controlled program for the inspection of selected RPV internal components in order to ensure their continued structural integrity. The inspection techniques utilized are primarily for the detection and characterization of service-induced, surface-connected planar discontinuities, such as IGSCC and IASCC, in welds and in the adjacent base material. Browns Ferry belongs to the BWRVIP organization and implementation of the procedurally controlled program is consistent with the BWRVIP issued documents. The inspection strategies recommended by the BWRVIP consider the effects of fluence on applicable components and are based on component configuration and field experience.

Components selected for inspection include those that are identified as susceptible to in-service degradation and augmented examination is conducted for verification of structural integrity.

These components have been identified through the review of NRC IEBs, BWRVIP documents, and recommendations provided by GE SILs. The inspection program provides performance frequency for NDE and associated acceptance criteria. Components inspected include the following:

• CS piping

• CS spargers

• Core shroud and core shroud support (includes access hole covers)

• Jet pumps and associated components

• Top guide

• Lower plenum

• Vessel ID attachment welds

• Instrumentation penetrations

• Steam dryer drain channel welds

• FW spargers

• Core plate

• SLC piping (internal to RPV)

Continued implementation of the current procedure program assures the prompt identification of any degradation of reactor vessel internal components experienced during EPU operating conditions. Browns Ferry utilizes mitigation techniques, such as HWC and NMC, to mitigate the potential for IGSCC and IASCC. Reactor vessel water chemistry conditions are also maintained consistent with the EPRI (Reference 4) and established industry guidelines. EPRI periodically updates the water chemistry guidelines, as new information becomes available. As EPRI updates occur, they are reviewed for possible incorporation into the BFN Chemistry Program.

The peak fluence increase experienced by the reactor internals does not exceed the threshold value that reflects a characteristic rise in intergranular cracking (Reference 5). The current 10-14

NEDO-33101 inspection strategy for the reactor internal components is expected to be adequate to manage any potential effects of EPU.

10.7.2 Flow Accelerated Corrosion The Browns Ferry procedurally controlled FAC program activities predict, detect, and monitor wall thinning in piping and components due to FAC. The FAC program is based on the EPRI guidelines in NSAC-202L, R2, "Recommendations for an Effective Flow-Accelerated Corrosion Program". The FAC program specifications and requirements ensure that the FAC program is implemented as required by NRC Generic Letter (GL) 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning."

The FAC program uses selective component inspections to provide a measure of confidence in the condition of systems susceptible to FAC for each unit. These selective inspections are the basis for qualifying un-inspected components for further service. In addition to this aggressive monitoring program, selected piping replacements have been performed to maintain suitable design margins and FAC-resistant replacement materials are used to mitigate occurrences of FAC.

A CHECWORKSTnI FAC model (in accordance with the CHECWORKSTh FAC users guide and EPRI modeling guidelines) has been developed for Browns Ferry to predict the FAC wear rate (single and two-phase fluids) and the remaining service life for each piping component. The controlled CHECWORKSTm FAC model is updated after each refueling outage. The FAC models are also used to identify FAC examination locations for the outage examination list and uses empirical data input to the model.

Process variables that influence FAC at Browns Ferry:

• Moisture content

• Water chemistry

• Temperature

• Oxygen

• Flow path geometry and velocity

• Material composition Browns Ferry has predicted EPU system operating conditions that are used as inputs to the CHECWORKSTm FAC model. EPU affects moisture content, temperature, oxygen and flow velocity but these remain within the CHECWORKS'm FAC model parameter bounds. Table 10-7 compares key parameter values affecting FAC. EPU parameter values result in changes consequential enough to heat balance models that certain systems or portion of systems (extraction steam) see a disproportional increase in wear compared to the percent power increase, i.e., FAC wear rate will increase in some systems and decrease in others.

The CHECWORKSTh ' FAC program targets FAC susceptible piping and components and includes the installation of FAC resistant material. Based on experience at pre EPU operating conditions and previous FAC modeling results, it is anticipated that the EPU operating conditions will result in changes for the CHECWORKS9M model. The changes may then result 10-15

NEDO-33101 in additional inspection scope, unless carbon steel piping and components have been replaced with FAC resistant material.

The BFN FAC program will continue to adequately manage loss of material due to flow accelerated corrosion, such that the piping and components will continue to perform their intended functions at EPU conditions.

10.8 REFERENCES

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

2. Tennessee Valley Authority, "Browns Ferry Nuclear Plant, Individual Plant Examination for External Events (IPEEE)," July 1995.

3. Tennessee Valley Authority, "Seismic IPEEE Report, Browns Ferry Nuclear Plant;"

Revision 0, June 1996.

4. EPRI Report TR-103515, R2, February, 2000, "BWR Water Chemistry Guidelines - 2000 Revision."

5. Engineering Materials Handbook on Stress Corrosion Cracking, 1992, Chapter 6, "Irradiation-Assisted Stress-Corrosion Cracking," Peter L. Anderson.

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NEDO-33101 Table 10-1 Browns Ferry Unit I High Energy Line Breaks Increase (Change from 105% OLTP) Due to EPU Break Location Mass Release Pressure Temperature MSLB in Steam Tunnel or MSVV No change No change No change FW Line Breaks in Steam Tunnel or MSVV + 12.3% No change < 0.5'F RCIC Steam Line Breaks in Reactor Building or MSVV No change No change No change HPCI Steam Line Breaks in Reactor Building No change No change No change RWCU Breaks in Reactor Building + 4.4% < 0.07 psi <9.30 F 10-17

NEDO-33101 Table 10-2 Browns Ferry Unit 1 Equipment Qualification for EPU Parameter EPU Effect Normal EPU plant operation radiation increase inside containment for EQ < 20%

Accident EPU radiation increase inside containment for EQ <14%

Accident EPU peak temperature inside containment for EQ 3360F(I)

Accident EPU peak pressure inside containment for EQ 50.64 psigd2)

Accident EPU temperature assumption outside containment for EQ 311°F Accident EPU pressure outside containment for EQ 7.24 psig Normal EPU radiation level increase outside containment for EQ < 20%

Accident EPU radiation increase outside containment for EQ <14%

(1) Based on MSLB inside containment.

(2) Conservative value used in EQ.

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NEDO-33101 Table 10-3 Browns Ferry Summary of CDF and LERF Parameter J Value Total CDF (ye', mean value) 1.66 E-6 LERF (yr', mean value) 2.93 E-7 10-19

NEDO-33101 Table 10-4 Summary of the Initiator Contributions to CDF and LERF for Browns Ferry Unit 1 Initiator Category Contribution to CDF and LERF Initiator Category Mean frequency CDF l LERF (events per year) l l Transient initiator categories Inadvertent Opening 4.36E-2 4.92E-9 4.63E-1 1 of One SRV Spurious Scram at 8.70E-2 1.04E-8 1.22E-10 Power Loss of 500kV 9.73E-3 1.68E-8 3.26E-10 Switchyard to Plant Loss of 500kV 2.30E-2 3.81E-8 6.31E-10 Switchyard to Unit Loss of 4.10E-3 3.43E-10 < IE-12 Instrunentation and Control Bus 1A Loss of 4.IOE-3 9.67E-9 1.38E-10 Instrumentation and Control Bus lB Total Loss of 9.09E-3 1.49E-8 2.68E-10 Condensate Flow Partial Loss of 1.80E-2 1.97E-9 1.39E-11 Condensate Flow MSIV Closure 5.70E-2 8.60E-8 1.58E-9 Turbine Bypass 1.96E-3 2.82E-9 4.21E-1 1 Unavailable Loss of Condenser 9.72E-2 1.48E-7 2.78E-9 Vacuum Total Loss of 2.58E-2 3.90E-8 6.64E-10 Feedwater Partial Loss of 2.58E-2 2.79E-9 2.31E-11 Feedwater 10-20

NEDO-33101 Initiator Category Contribution to CDF and LERF Initiator Category Mean frequency CDF LERF (events per year)

Loss of Plant Control 1.20E-2 2.07E-8 3.64E-10 Air Loss of Offsite Power 6.43E-3 2.21E-7 1.58E-9 Loss of Raw Cooling 7.95E-3 1.96E-8 2.93E-10 Water Momentary Loss of 7.17E-3 6.64E-10 2.29E-12 Offsite Power Turbine Trip 5.09E-1 6.03E-8 9.88E-10 High Pressure Trip 4.30E-2 4.85E-9 4.57E-1 1 Excessive Feedwater 2.60E-2 2.8 1E-9 2.33E-11 Flow Other Transients 3.70E-1 4.74E-8 7.06E-10 ATWS Categories Turbine Trip ATWS N/A 1.57E-7 8.49E-8 LOSP ATWS N/A 2.02E-9 1.07E-9 Loss of Condenser N/A 4.87E-8 2.63E-8 Heat Sink ATWS Inadvertent Opening N/A 1.26E-8 6.76E-9 of SRV ATWS Loss of Feedwater N/A 1.01E-7 5.44E-8 ATWS LOCA initiator categories Breaks Outside 6.67E-4 1.90E-8 3.22E-10 Containment Excessive LOCA 9.39E-9 9.19E-9 4.55E-1 1 (reactor vessel failure)

Ifterfacing Systems 1.42E-4 4.20E-7 9.48E-8 LOCA 10-21

NEDO-33101 Initiator Category Contribution to CDF and LERF Initiator Category Mean frequency CDF l LERF (events per year) l l Large LOCA - Core Spray Line Break Loop I 1.57E-6 3.41E-9 1.48E-10 Loop II 1.57E-6 3.41E-9 1.48E-10 Large LOCA -

Recirculation Discharge Line Break Loop A 1.10E-5 1.86E-8 1.16E-9 Loop B 1.1OE-5 1.86E-8 1.16E-9 Large LOCA -

Recirculation Suction Line Break Loop A 7.85E-7 4.33E-9 7.66E-1 1 Loop B 7.85E-7 4.33E-9 7.66E-l 1 Other Large LOCA 1.57E-6 2.49E-9 1.48E-10 Medium LOCA Inside 4.OOE-5 2.64E-8 4.27E-9 Containment Small LOCA Inside 5.OOE-4 1.35E-9 1.73E-10 Containment Very Small LOCA 3.38E-3 2.72E-10 less than IE-12 Inside Containment Internal flooding initiator categories EECW Flood in 1.20E-3 6.87E-ll less than IE-12 Reactor Building -

shutdown units EECW Flood in 1.85E-6 2.99E-9 3.OOE-10 Reactor Building -

operating unit 10-22

NEDO-33101 Initiator Category Contribution to CDF and LERF Initiator Category Mean frequency CDF LERF (events per year)

Flood from the 1.22E-4 1.OOE-9 7.51E-12 Condensate Storage Tank Flood from the Torus 1.22E-4 4.08E-8 5.8 1E-9 Large Turbine 3.65E-3 9.49E-9 1.18E-10 Building Flood Small Turbine 1.65E-2 5.1 lE-9 5.05E-1 1 Building Flood 10-23

NEDO-33101 Table 10-5 Frequency Weighted Fractional Importance to Core Damage of Operator Actions Used in Browns Ferry Unit 1 PRA Frequency-Weighted Database Fractional Importance Variable Operator Action Description to Core Damage HPRVD1 OPERATOR FAILS TO INITIATE 2.8105E-001 DEPRESSURIZATION HOREEI OPERATOR FAILS TO ALIGN SWING RHRSW 2.0980E-002 PUMPS FOR EECW (SCENARIO REQUIRES 2 PUMPS TO BE ALIGNED)

HOTAF1 OPERATOR FAILS TO CONTOL LEVEL AT TAF 1.7612E-002 DURING ATWS - UNISOLATED VESSEL HOU11 OPERATOR FAILS TO ALIGN THE RHR UNIT 9.2941 E-003 1/UNIT 2 CROSSTIE HODWS I OPERATOR FAILS TO ALIGN FOR DRYWELL 5.3865E-003 SPRAY. THIS IS A NON ATWS SCENARIO.

HOTAF2 OPERATOR FAILS TO CONTOL LEVEL AT TAF 3.8830E-003 DURING ATWS- ISOLATED VESSEL H0HPC1 OPERATOR FAILS TO CONTROL LEVEL WITH 3.6075E-003 HPCIIRCIC - THIS IS A NON ATWS SCENARIO HOHPR1 OPERATOR FAILS TO CONTROL LEVEL WITH 3.6075E-003 HPCI/RCIC FOLLOWING LEVEL 8 TRIP HRSPC1 OPERATOR FAILS TO LOCALLY RECOVER SP 2.1276E-003 COOLING FAILURE HRSPC2 OPERATORS FAIL TO ALIGN SUPPRESSION 1.7454E-003 POOL COOLING DURING ATWS HOREE2 OPERATOR FAILS TO ALIGN SWING RHRSW 1.6831 E-003 PUMPS FOR EECW (SCENARIO REQUIRES I PUMP TO BE ALIGNED)

HOSLI OPERATOR FAILS TO INITIATE STANDBY 1.6377E-003 LIQUID CONTROL - VESSEL IS NOT ISOLATED FROM CONDENSER 10-24

NEDO-33101 Frequency-Weighted Database Fractional Importance Variable Operator Action Description to Core Damage HOHPC2 OPERATOR FAILS TO CONTROL LEVEL WITH 1.3671E-003 HPCIIRCIC DURING AN ATWS HOX2 OPERATOR FAILS TO CROSSTIE 4 KV 8.0552E-004 SHUTDOWN BOARD HOSL2 OPERATOR FAILS TO INITIATE STANDBY 3.6644E-004 LIQUID CONTROL - VESSEL IS ISOLATED FROM CONDENSER HRSPCI OPERATOR FAILS TO ALIGN SUPPRESSION 2.7164E-004 POOL COOLING - THIS IS A NON ATWS SCENARIO HOADI OPERATOR FAILS TO INHIBIT ADS 2.4140E-004 HODSBI OPERATOR FAILS TO ALIGN DIESEL BOARD 6.6053E-005 FOR DIESEL C HOXI OPERATOR FAILS TO ALIGN BATTERY 4.6077E-005 CHARGER 2B TO 250V DC BATTERY BOARD HOR480 OPERATOR FAILS TO RECOVER 480 SHUTDOWN 4.2578E-005 BOARD HPCRD1 OPERATOR FAILS TO ESTABLISH ADDITIONAL 1.8067E-005 CRD FLOW PER APP 5B HOLPCI OPERATOR FAILS TO CONTROL LPCI/CS 1.7648E-005 INJECTION HORTB I OPERATOR FAILS TO PROVIDE BACKUP TRIP 5.72166E-006 SIGNAL 10-25

NEDO-33101 Table 10-6 Browns Ferry PRA Peer Review Results CERTIFICATION PRA ELEMENT GRADE INITIATING EVENTS 3 ACCIDENT SEQUENCE EVALUATION 3 THERMAL HYDRAULIC ANALYSIS 2 SYSTEMS ANALYSIS 3 DATA ANALYSIS 2 HUMAN RELIABILITY ANALYSIS 3 DEPENDENCY ANALYSIS 3 STRUCTURAL RESPONSE 3 QUANTIFICATION 3 CONTAINMENT PERFORMANCE ANALYSIS 2 MAINTENANCE AND UPDATE PROCESS 3 10-26

NEDO-33101 Table 10-7 Browns Ferry Unit 1 FAC Parameter Comparison for EPU CHECKNVORKS ^ EPU Parameter Allowable Input Typical Range of Comments Values Steam Flow 1 - 100,000,000 930,000 to Values Within (Ibm/hr) 13,840,000 Allowable Velocity (ft/sec) Calculated in program 132 to 171 Values Within Allowable Steam Quality 0 to 100 92.8 to 98.2 Values Within

(%) Allowable Operating 0 to 750 314 to 406 Values Within temperature (OF) Allowable 10-27

NEDO-33101

11. LICENSING EVALUATIONS 11.1 OTHER APPLICABLE REQUIREMENTS The analysis, design, and implementation of EPU were reviewed for compliance with the current plant licensing basis acceptance criteria and for compliance with new regulatory requirements and operating experience in the nuclear industry. ((

)) The associated tables identify the issues that are generically evaluated, and issues to be evaluated on a plant-unique basis. The applicable plant-unique evaluations have been performed for the subjects addressed below.

11.1.1 NRC and Industry Communications All of the issues from the following NRC and industry communications are either generically evaluated in ELTR2 (as supplemented), or are evaluated on a plant-specific basis as part of the EPU program. These evaluations conclude that every issue is either (1) not affected by the EPU, (2) already incorporated into the generic EPU program, or (3) bounded by the plant-specific EPU evaluations. The NRC and industry communications evaluated cover the subjects listed below.

• CFRs

• NRC TMI Action Items

• NRC Action Items (Formerly Unresolved Safety Issues) and New Generic Issues

• NRC Regulatory Guides

• NRC Generic Letters

• NRC Bulletins

• NRC Information Notices

• NRC Circulars

• INPO Significant Operating Event Reports (applicable to the EPU)

• GE Services Information Letters

• GE Rapid Information Communication Service Information Letters 11.1.2 Plant-Unique Items Plant-unique items whose previous evaluations could be affected by operation at the EPU RTP level have been identified. These are (1) the NRC and Industry communications discussed above, (2) the safety evaluations for work in progress and not yet integrated into the plant design, (3) the temporary modifications that could have been reviewed prior to the EPU and still exist after EPU implementation, and (4) the plant EOPs. These items will be reviewed for possible effect by the EPU, and items affected by the EPU will be revised prior to EPU implementation.

11-1

NEDO-33101 11.1.2.1 Commitments to the NRC Prior to EPU implementation, the potentially power dependent NRC commitments are reviewed for required changes due to EPU conditions prior to EPU implementation. The commitments that are affected by EPU will be updated to account for the effects of EPU.

11.1.2.2 10 CFR 50.59 Evaluations 10CFR50.59 evaluations performed for work in progress and 10CFR50.59 evaluations completed but not yet included in the UFSAR are reviewed prior to EPU implementation for required changes due to EPU conditions. No 10 CFR 50.59 evaluation process change is required for EPU.

11.1.2.3 Temporary Modifications Pre-existing Temporary Modifications, Technical Operability Evaluations, Open Work Orders that will be in effect after EPU implementation will be reviewed and revised, if necessary, to include EPU conditions.

11.1.2.4 Emergency and Abnormal Operating Procedures EOPs and AOPs can be affected by EPU. Some of the EOPs variables and limit curves depend upon the value of rated reactor power. Some AOPs may be affected by plant modifications to support the higher power level.

EOPs include variables and limit curves, defining conditions where operator actions are indicated. Some of these variables and limit curves depend upon the RTP value. Changing some of the variables and limit curves requires modifying the values in the EOPs and updating the support documentation. EOP curves and limits may also be included in the safety parameter display system and will be updated accordingly.

The charts and tables used by the operators to perform the EOPs are reviewed for any required changes prior to each core reload. The EOPs will be reviewed for any changes required prior to EPU implementation. The operators will receive training on these procedures as described in Section 10.6.

AOPs include event based operator actions. Some of these operator actions may be influenced by plant modifications required to support the increase in rated reactor power. Changing some of the operator actions may require modifications to the AOPs and updating the support documentation. The Browns Ferry AOPs will be reviewed for any effects of power uprate prior to EPU implementation. Some of the setpoints used in the AOPs change due to EPU. The operators will receive training on these procedures as described in Section 10.6.

The Browns Ferry EOPs are reviewed for any effects of the EPU, and the EOPs will be updated, as necessary. This review is based on Section 2.3 of ELTR2, which includes a list of operator action levels, which are sensitive to the EPU.

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NEDO-33101

11.2 REFERENCES

1. GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," (ELTR2), Licensing Topical Reports NEDC-32523P-A, Class M (Proprietary), February 2000; NEDC-32523P-A, Supplement I Volume I, February 1999; and Supplement 1 Volume II, April 1999.

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ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 1 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS - 431 -

REQUEST FOR LICENSE AMENDMENT FOR EXTENDED POWER UPRATE OPERATION GENERAL ELECTRIC REPORT BROWNS FERRY UNIT I SAFETY ANALYSIS FOR EXTENDED POWER UPRATE See Attached:

• Assessment of General Electric Evaluations for Extended Power Uprate

• Browns Ferry Units 1, 2, and 3 Comparison of ELTR Generic Evaluations to the PUSAR

• NEDC-331 01 P, "Browns Ferry Unit 1 Safety Analysis Report for Extended Power Uprate," (PUSAR) Revision 0.

i

Assessment of General Electric Evaluations For Extended Power Uprate In order for TVA to ensure that the EPU products provided by GE were based on approved methodologies, consistent with engineering requirements, and compatible with the design, operating, and licensing basis of BFN, the following were performed:

1) Each GE task product was reviewed and comments resolved through the TVA EPU Project staff prior to issue. The TVA EPU Project staff was composed of highly qualified engineers with diverse BWR experience, including system engineering and qualified station SRO experience. The task product review included participation review by site organizations (operations, engineering, maintenance, training, etc.) to ensure that the products were compatible with the design, operation, and licensing bases of BFN and included the consideration of the latest industry operating experience.

2) Each final GE task product was design verified in accordance with the GE Quality Assurance Program, by a GE technical verifier and a GE Regulatory Services verifier, with oversight by the responsible GE technical manager and GE Project Manager. Prior to acceptance by TVA, proper resolution and incorporation of all comments was verified.

3) Within the TVA EPU Project, a separate TVA project task performed additional reviews of GE specified power uprate process parameter change effects on the plant design basis. Plant design basis calculations and analyses were reviewed to confirm the need for and accomplishment of their revision. This task accomplished the consideration of plant process parameter changes and the effect on plant engineering analysis and operating setpoints due to EPU.

4) In selected cases, TVA contracted with engineering firms to independently perform specified engineering evaluations to verify the GE task evaluation work and results.

5) A review of the generic evaluations credited in the BFN PUSAR was performed to ensure that the generic evaluations provided in ELTR 1 and ELTR 2 were applicable to BFN. The summary of the results of this review are provided in the attached Table.

6) In order to validate the quality of the GE design products, a formal, independent Nuclear Assurance assessment was conducted in April 2002. A group of BFN personnel including representatives of Site Engineering, EPU Project, Nuclear Fuels, and Nuclear Assurance conducted a review of GE task output documentation (products) and task supporting documentation (design basis) developed for the EPU project. Task products, such as task scoping documents, draft task reports, and final task reports were verified to E4-1

provide consistent and current information accurately reflecting design input information and adequate task to task communication in the associated evaluations and calculations. The assessment included a review of task product basis documents and supporting evaluations contained in GE design record files at GE offices in San Jose, California and Wilmington, North Carolina.

7) The GE work was subject to TVA procedures that control the handling of issues identified as Conditions Adverse To Quality. During the progress of the EPU Project, GE conditions adverse to quality that emerged and their corrective actions were monitored. Parallel TVA CAQs were issued to track issues to resolution through any necessary revision of BFN design basis documents.

These in-process and independent reviews and assessments performed and conducted during the EPU project provided reasonable assurance that the products and activities met the TVA expectations and requirements and were based on approved methodologies, consistent with all engineering requirements, and compatible with the design, operating, and licensing basis of BFN.

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Browns Ferry Units 1, 2 and 3 Comparison of ELTR Generic Evaluations to the PUSAR PUSAR ELTR 1 / ELTR Parameter(s) /

Section Topic ELTR 2 Requirement(s) I BFN PUSAR Comparison Section Assumption(s) 1.1, 1.3, Reactor ELTR 1,

• 20% Thermal Power Increase.

• 20% Thermal Power Increase Table 1-2 Thermal - Section from OLTP.

Hydraulic 1.0, Table (15% Increase from CLTP for Parameters 5.1, Units 2 & 3)

Appendix C.2

• 24% Steam Flow Increase.

• Approximately 23% Steam Flow Increase from OLTP.

(16% Increase from CLTP for Units 2 & 3)

• 1095 psia Operating Dome

• 1050 psia Operating Dome Pressure. Pressure.*

• 556 0F Dome Temperature.

• 550.5 OF Dome Temperature.*

• 99% to 110% Full Power Core Flow Range.

• 99% to 105% Full Power Core Flow Range.

• No change from CLTP conditions for Units 2 & 3.

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Browns Ferry Units 1, 2 and 3 Comparison of ELTR Generic Evaluations to the PUSAR PUSAR ELTR 1 / ELTR Parameter(s) /

Section Topic ELTR 2 Requirement(s) I BFN PUSAR Comparison Section Assumption(s) 2.3.1, Power ELTR 1, . The upper boundary shall be

• The maximum EPU RTP (Points Figure 2-1 Flow Appendix limited to the uprated power D, E, & F of PUSAR Figure 2-1)

Operating C.2.3, level. corresponds to 120% of the Map Figure 5-1, OLTP.

ELTR 2, Section 3.2

• The right side of the operating

• The maximum core flow shown on range shall be the same core PUSAR Figure 2-1 corresponds to flow limit as currently licensed. the previously analyzed core flow range when resealed so that EPU RTP is equal to 100% rated.

• The left (lower core flow) side of

• Point D of PUSAR Figure 2-1 the operating map will be corresponds to 99% core flow at bounded by the new lower limits 100% EPU RTP.

provided in Table C-1. (99% for BWR-3 and 4.)

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Browns Ferry Units 1, 2 and 3 Comparison of ELTR Generic Evaluations to the PUSAR PUSAR ELTR 1 / ELTR Parameter(s) /

Section Topic ELTR 2 Requirement(s) I BFN PUSAR Comparison Section Assumption(s) 3.7 Main Steam ELTR 2,

• 20% Thermal Power Increase.

• 20% Thermal Power Increase Isolation Section 4.7 from OLTP. (15% Increase from Valves CLTP for Units 2 & 3)

• 24% Steam Flow Increase.

• Approximately 23% Steam Flow Increase from OLTP. (16%

Increase from CLTP for Units 2 &

3)

. 1095 psia Operating Dome Pressure.

• 1050 psia Operating Dome Pressure.*

• 556 OF Dome Temperature.

• 550.5 OF Dome Temperature.*

• No change from CLTP conditions for Units 2 & 3.

4.2.1 High ELTR 2, * < 75 psi increase in Reactor

• 30 psi increase for Unit 1. No Pressure Section 4.2 operating pressure. change from CLTP conditions for Coolant Units 2 & 3.

Injection

• The HPCI hydraulic control

• Modification to be installed on Unit modification described in GE SIL 1. Modification installed on Units No. 480 should be installed: 2 & 3.

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