Turkey Point Units 3 & 4, WNA-LI-00039-FPL-NP, Rev 1 ...

L-2005-006ENCLOSURE I

Enclosure ID (Non-Proprietary)

ENCLOSURE ID

LICENSING INPUT FOR RPS MODIFICATIONS

CHANGING REACTOR TRIP ON TURBINE TRIP PERMISSIVE

FROM P-7 TO P-8

(contains non-proprietary information)

Westinghouse Non-Proprietary Class 3

Florida Power & LightTurkey PointUnits 3 & 4

Licensing Input for RPSModifications Changing ReactorTrip on Turbine Trip Permissive

from P-7 to P-8

WNA-LI-00039-FPL-N P

Revision I

February 2005

Westinghouse Non-Proprietary Class 3

Florida Power & LightTurkey PointUnits 3& 4

Licensing Input for RPS ModificationsChanging Reactor Trip on TurbineTrip Permissive from P-7 to P-8

WNA-LI-00039-FPL-NP

Revision 1February 2005

APPROVALS

Function Name and Signature Date

Atos C. W. Suggs 4-Authors Principal Engineer, Des nBasiEeerin| 2/2112005

Reviewed V. M. Thomas A 2/21/2005Pricipal Engineer. Desig t'asis Engineering ,

Approved J. A. Jurczak ,, 2/21/2005Manager, Design Bas is /

Westinghouse Electric Company, LLCP.O. Box 355

Pittsburgh, PA 15230-0355

i 2005 Westinghouse Electric Company LLCAll Rights Reserved

Florida Power & LightTurkey Point Units 3 & 4

Licensing Input for RPSModifications ChangingReactor Trip on Turbine TripPermissive from P-7 to P-8

LIST OF CONTRIBUTORSIREVIEWERS

Name Date

Natalie R. Jurcevich 9/01/04

J. Seenu Srinivasan 9/01/04

Edward M. Monahan 11/19/04

REVISION HISTORY

RECORD OF CHANGES

Revision Revision Made By Description Date

0 C. W. Suggs Initial issue. 1/11/2005

I C. W. Suggs Clarified wording/references and 2/21/2005redundant word correction.

DOCUMENT TRACEABILITY & COMPLIANCE

Created to Support the Following Document(s) Document Number Revision

None

DETAILED RECORD OF CHANGES

RevisionDescription:

Rev 0 Initial issue.Rev 1 :Clarifed wording/references (page 2-2), corrected redundant

word (page 4-7)

Date

1/20052/2005

WNA-LI-00039-FPL-NP, Rev. I i



Section

SECTION 1

SECTION 2

SECTION 33.13.23.33.43.5

SECTION 4

SECTION 5

SECTION 6

TABLE OF CONTENTS

Title Page

LIST OF CONTRIBUTORS/REVIEWERS ....................................... iTABLE OF CONTENTS ...................................... ii

GLOSSARY OF TERMS ...................................... iv

REFERENCES ........................................ v

PURPOSE AND SCOPE OF DOCUMENT ...................................... 1-1

CHANGE DESCRIPTION ...................................... 2-1

BEST ESTIMATE ANALYSIS ...................................... 3-1

INTRODUCTION ...................................... 3-1

INPUT PARAMETERS AND ASSUMPTIONS ...................................... 3-1

DESCRIPTION OF ANALYSIS AND EVALUATION ...................................... 3-4ACCEPTANCE CRITERIA AND RESULTS ...................................... 3-4CONCLUSIONS ...................................... 3-5

DETAILED EVALUATION OF NON-LOCA EVENTS ...................................... 4-1

SUMMARY OF EVALUATIONS ...................................... 5-1

CONCLUSION ...................................... 6-1

INDEX OF TABLES

Table Title

Table 1 Turbine Trip without Reactor Trip from P-8 Setpoint of 40% Power

Page

3-4

WNA-LI-00039-FPL-NP, Rev. I ii

Florida Power & Light Licensing Input for RPSTurkey Point Units 3 & 4 Modifications Changing

Reactor Trip on Turbine TripPermissive from P-7 to P-8

ACRONYMS AND TRADEMARKS

The following abbreviations and acronyms are defined to allow an understanding of their use withinthis document.

Acronyms Definition

BOL Beginning-of-LifeDNB Departure from Nucleate BoilingFPL Florida Power & LightFWCS Feedwater Control Systeml&C Instrumentation and ControlLAR License Amendment RequestNSSS Nuclear Steam Supply SystemPTN Plant Turkey PointPORV Power-Operated Relief ValveRCS Reactor Coolant SystemRPS Reactor Protection SystemRTP Rated Thermal PowerSG Steam GeneratorUFSAR Updated Final Safety Analysis Report

All other product and corporate names used in this document may be trademarks or registeredtrademarks of other companies, and are used only for explanation and to the owners' benefit,without intent to infringe.

WNA-LI-00039-FPL-NP, Rev. I i~I



GLOSSARY OF TERMS

The following definitions are provided for the special terms used in this document.

Term Definitions

None.

WNA-LI-00039-FPL-NP, Rev. I iv



REFERENCES

Following is a list of references used throughout this document.

1. Not used.

2. Florida Power & Light Turkey Point Units 3 and 4, Westinghouse Functional Logic Diagrams,883D988 Sheet 2, Rev. 3; Sheet 11, Rev. 5; and Sheet 16, Rev. 5.

3. NUREG-0737, -Clarification of TMI Action Plan Requirements," Item Il.K.3.10, ProposedAnticipatory Trip Modification, October 1980.

4. PCWG-2779, 'Turkey Point Units 3 & 4 (FPUFLA): Approval of Category IV PCWGParameters to Support Reduced Feedwater Temperature in Conjunction with UprateProgram," June 14, 2002.

5. WCAP-7907-P-A, "LOFTRAN Code Description," April 1984.

6. Westinghouse Letter, 95-JB-UP-5478, "Turkey Point Plant-Units 3 & 4, Thermal PowerUprate Project, Final Margin to Trip Evaluation," December 1995.

7. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 7A, Revision 1, "Steam Dump toCondenser Controls."

8. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 7B, Revision 0, "Steam Dump toCondenser Controls."

9. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 2A, Revision 2, "PressurizerPressure Control."

10. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 2B, Revision 0, "PressurizerPressure Control."

11. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 3D, Revision 1, "PressurizerLevel Control and Protections."

WNA-LI-00039-FPL-NP, Rev. I v



12. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 8A4, Revision 4, "S/G A LevelNarrow Range."

13. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 8B4, Revision 3, 'S/G B LevelNarrow Range."

14. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 8C4, Revision 3, "S/G C LevelNarrow Range."

15. FPL, Turkey Point Units 3 and 4, Drawing 5610-J-844, Sheet 5H, Revision 2, "ReactorTemperature Controls (Tavg - Tref)."

16. FPL, Turkey Point Units 3 and 4, Drawing 561 0-J-844, Sheet 5J, Revision 1, "ReactorTemperature Control (Power Mismatch)."

17. FPL, Turkey Point Units 3 and 4, Drawing 5610-T-D-12A, Sheet 1, Revision 12, 'Rod ControlSystem."

18. FPL, Turkey Point Units 3 and 4, Drawing 5610-T-D-12B, Sheet 1, Revision 10, "Tavg Controland Insertion Limit Alarms."

19. FPL, Document 5613-M-313, Revision 40, "Turkey Point Nuclear Unit 3 Instrument SetpointList," 2-23-04.

20. FPL, Document 5614-M-313, Revision 38, 'Turkey Point Nuclear Unit 4 Instrument SetpointList," 11-26-02.

21. FPL, Turkey Point Units 3 and 4, Reactor Trip Signals Functional Logic Drawing, 5610-T-1,Sheet 2, Revision 20.

22. FPL, Turkey Point Units 3 and 4, Functional Logic Drawing, 5610-T-L1, Sheet 17, Revision14, "Logic Diagram Units 3 & 4, Nuclear Instrumentation Permissives and Blocks."

23. Turkey Point Units 3 and 4, Design Basis Document, Reactor Protection System andEngineered Safety Features Actuation System, 5610-049-DB-001, Rev. 11.

WNA-LI-00039-FPL-NP, Rev. I vi



24. Turkey Point Units 3 and 4 UFSAR (Updated Final Safety Analysis Report), Chapter 7, Rev.U4C21.

25. Turkey Point Units 3 and 4 UFSAR, Chapter 14, Rev. U4C21.

26. Turkey Point Units 3 and 4 Plant Technical Specifications, Through Amendment Nos. 224 and221.

27. FPL Purchase Order 00071279, Revision 004.

28. Westinghouse Offer NA-MKTG-04-39, dated February 24, 2004.

29. Westinghouse Offer NA-MKTG-04-48, dated March 16, 2004.

30. Westinghouse Letter, FPL-04-292, MP7-P8 RPS Modification Best Estimate Analysis, Rev. 4,"December 17, 2004.

31. WCAP 12201, 'Basis Document for Westinghouse Setpoint Methodology for ProtectionSystems, Turkey Point Units 3 and 4, March 1990

(Last Page of Front Matter)

WNA-LI-00039-FPL-NP, Rev. I vii



SECTION 1PURPOSE AND SCOPE OF DOCUMENT

The purpose of this document is to provide licensing support for the Reactor Protection System(RPS) modifications to change the reactor trip on turbine trip interlock from permissive P-7 topermissive P-8 and to change the P-8 setpoint from 45% to 40% at Turkey Point Units 3 and 4.

This report provides the following information:

* A description of the changes* Licensing justification for changing the reactor trip on turbine trip interlock from P-7 to P-8

and changing the P-8 setpoint from 45% to 40%

(Last Page of Section 1)

WNA-LI-00039-FPL-NP, Rev. I 1-1



SECTION 2CHANGE DESCRIPTION

The current design at Turkey Point Units 3 and 4 for the reactor trip on turbine trip automaticallyblocks the function when power levels are below the P-7 setpoint (10% power). The Turkey Pointplants are designed with 50% load rejection capability. With the load rejection capacity at theTurkey Point Units, load rejections of up to 50% should not require a reactor trip if all controlsystems function as designed. Therefore, it is possible to increase the setpoint for the reactor tripon turbine trip interlock up to 50% power. This would result in blocking the direct reactor trip onturbine trip for load rejection events up to 50% power levels. By implementing the block of reactortrip on turbine trip at a permissive with a higher setpoint, there is a decrease in potentiallyunnecessary challenges to the reactor protection system and an increase in plant availability. Forthe Turkey Point Units 3 and 4, the P-8 permissive has been selected, but with the setpoint reducedfrom 45% to 40% power, for conservatism.

A review of the safety analyses in Chapter 14 of the Turkey Point UFSAR has been performed inorder to confirm that the safety analysis results are not adversely affected by this proposedmodification.

In addition to evaluating the UFSAR Chapter 14 licensing basis, an evaluation has been performedto determine the impact of a turbine trip without reactor trip on the pressurizer power-operated reliefvalves (PORVs). Following the Three Mile Island event, the NRC expressed concern about theimplementation of blocking the reactor trip on turbine trip function on a permissive with an increasedsetpoint because of the potential to increase the probability of a stuck open pressurizer PORV. TheNRC position is addressed in NUREG-0737, Item II.K.3.10. The NRC has stated that theanticipatory trip modification proposed by some licensees to confine the range of use to high-powerlevels should not be made until it has been shown on a plant-by-plant basis that the probability of asmall-break loss-of-coolant accident (LOCA) resulting from a stuck-open power-operated reliefvalve (PORV) is substantially unaffected by the modification. Therefore, the proposed P-8permissive setpoint (40% power) must be evaluated to determine if it meets the acceptancecriterion for the turbine trip without reactor trip.

To satisfy the NRC requirements in Item II.K.3.10, a best estimate plant specific analysis wasperformed to show that the implementation of the block of reactor trip on turbine trip at the P-8




setpoint will not result in challenges to the pressurizer PORVs. The best estimate analysis wasperformed using the proposed P-8 setpoint of 40% power. The results show that for this setpointvalue, the pressurizer PORVs will not be challenged.

The proposed P-8 setpoint change affects the Reactor Coolant Flow - Low Reactor Trip logic (TechnicalSpecifications Table 3.3-1, Function 10) and the Reactor Coolant Pump Breaker Position Reactor Triplogic (Technical Specifications Table 3.3-1, Function 18). For these functions the single loop and singlebreaker trips are enabled above P-8 and the two loop and two breaker trips are enabled between P-7and P-8. Changing the P-8 setpoint from 45% to 40% is in the conservative direction and is allowableaccording to the current Trip Setpoint Allowable Value (Technical Specifications Table 2.2-1, Function17.c). Currently, between 40% and 45% power, low flow in two loops or two RCP breakers open isrequired for a reactor trip. Following the P-8 setpoint change, between 40% and 45% power, low flow inonly one loop or only one RCP breaker open will initiate a reactor trip. For events analyzed in UFSARChapter 14 which credit the low flow or RCP breaker trips, the analyses and results are bounding sincethe P-8 setpoint change is in the conservative direction. The current 3% difference between the TripSetpoint and the Allowable Value specified in Technical Specification Table 2.2-1, Reactor Trip SystemInstrumentation Trip Setpoints is based on the allowable value of the associated protection function asdocumented in the current setpoint analysis (Reference 31). The basis for the 3% allowance isunaffected by this change.

The P-7 interlock receives input from the Power Range Neutron Flux instrumentation and the Turbinefirst stage pressure. The P-8 interlock only receives input from the Power Range Neutron Fluxinstrumentation. This represents a logic change for the reactor trip on turbine trip function. Thischange is acceptable because the P-8 interlock will continue to receive reliable input from thePower Range Neutron Flux instrumentation, and the accident analyses do not credit the Turbinefirst stage pressure input to the permissive as a trip initiator or as an accident mitigation function.





SECTION 3BEST ESTIMATE ANALYSIS

3.1 INTRODUCTION

Westinghouse has performed an evaluation for the Reactor Protection System (RPS) modificationsto change the reactor trip on turbine trip interlock from permissive P-7 to permissive P-8 at TurkeyPoint Units 3 and 4 (Reference 30). This evaluation has been performed to determine the impact ofa turbine trip without reactor trip on the pressurizer Power Operated Relief Valves (PORVs).

Following the Three Mile Island event, the NRC has expressed a concern on the implementation ofincreasing the interlock setpoint for the turbine trip without a reactor trip because of the potential toincrease the probability of a stuck open pressurizer PORV. The NRC position was addressed inNUREG- 0737, Item lI.K.3.10. In order to reduce the likelihood of opening the pressurizer PORVfollowing a turbine trip without a reactor trip, the P-8 setpoint will be reduced from 45% to 40%.

To satisfy the NRC requirements, a best estimate plant specific analysis was performed todetermine if the implementation of the P-8 permissive (40%) will result in challenges to thepressurizer PORVs.

3.2 INPUT PARAMETERS AND ASSUMPTIONS

A best estimate analytical study was performed to determine the transient plant response to aturbine trip without reactor trip transient. The analysis was performed using the LOFTRANcomputer code (Reference 5) model of Turkey Point Units 3 and 4. The LOFTRAN computer codewas previously used with best estimate methodology for the Turkey Point Units 3 and 4 uprate.This computer model simulates overall thermal-hydraulic and nuclear response of the NSSS as wellas the various control and protection systems. Since the object of this study was primarily todetermine the peak in pressurizer pressure following the initiation of the transient, assumptionswere made that would contribute to a conservatively high prediction of pressurizer pressure.




These assumptions follow: a,b,c

a,b,c

* Minimum Tavg was chosen since it results in the smallest temperature error for thesteam dump system and therefore the largest plant heatup.

* Minimum SGTP level was chosen since it results in the highest pressurizer insurgefor a fixed full power Tavg.

* Maximum feedwater temperature was chosen since it results in the higheststeam/feedwater flow and therefore, the smallest steam dump capacity (in fraction ofrated steam flow).

3. Best estimate Beginning-of-Life (BOL) reactivity parameters were used. BOL reactivityparameters have lower differential rod worth and less negative moderator temperaturecoefficient and thus, using BOL parameters in the analysis yield more conservative results,which bound the full cycle of operation.

; 4. Initial RCS conditions such as, Tavg and pressure, are without any uncertainties (i.e., bestestimate analysis) and are at their [ ]b.c power value.

5. Minimum overall heat transfer coefficient (UA) for fuel to coolant, consistent with BOLconditions.

a,b,c



Licensing Input for RPSModifications ChangingReactor Trip on Turbine TripPermissive from P-7 to P8

a,u,c

7. The Pressurizer Pressure control system, Steam Dump control system (to condenser, loadrejection mode) and Steam Generator (SG) Level control system were assumed operationaland in the automatic mode of control. Steam dump to the atmospheric relief valves is notcredited in the analysis.

a,b,c

10. Since the analyses are best estimate analyses, the parameter values including the controlsystems setpoints and PORV setpoints were assumed at nominal conditions withoutuncertainties and/or tolerances.

A majority of the LOFTRAN fluid systems thermal-hydraulic data was taken from the uprate analysisperformed in 1995 (Reference 6). The control system settings (the gains and time constants, etc.)were taken from the following references:

Steam Dump Control Settings:

Pressurizer Pressure Control Settings:

References 7 & 8

References 9 & 10

Pressurizer Level Control Settings: Reference 11

Steam Generator Level Control Settings: References 12,13 & 14

Rod Control Settings: Reference 15, 16, 17 & 18




3.3 DESCRIPTION OF ANALYSIS AND EVALUATION

A best estimate analysis for a turbine trip without reactor trip transient from the proposed P-8setpoint was performed to determine if the pressurizer PORVs are challenged. The turbine tripwithout a reactor trip transient was initialized from an initial power level of [

]bc with all normal control systems assumed operational. This bestestimate analysis addresses the NRC position in NUREG-0737, Item II.K.3.10 (Reference 3).

3.4 ACCEPTANCE CRITERIA AND RESULTS

The acceptance crterion for the turbine trip without a reactor trip transient from the proposed P-8 abc

nservatism, the results were further evaluated to determine if the SG safety valves were abc

would be challenged. Also, the peak pressurizer levels were reported (as % of the tap-to-tap span andcubic feet) to illustrate that the pressurizer does not go water solid during the transient.

The proposed P-8 setpoint of 40% accommodates a turbine trip without a reactor trip without challengingthe PORVs or the SG safety valves. The results of this best estimate analysis are shown in Table 1.

a,b,cTable I

Turbine Trip without Reactor Trip from P-8 Setpoint of 40% Power




3.5 CONCLUSIONS

The turbine trip without reactor trip transient from the proposed P-8 setpoint of 40% power will not resultin challenging the pressurizer PORVs with best estimate simulation.


WNA-LI-00039-FPL-NP, Rev. 1 3-5



SECTION 4DETAILED EVALUATION OF NON-LOCA EVENTS

Each of the non-LOCA accident analyses described in Chapter 14 of the Turkey Point UFSAR wasreviewed to evaluate the effect of moving the reactor trip on turbine trip function from P-7 (10%power) to P-8 (540% power). Additionally, the change in the P-8 setpoint itself from 45% to 40%was assessed for the Loss of Flow events (all other events are unaffected by the change in P-8).Based on this review, it is concluded that the proposed changes have no effect on the accidentanalyses. This review is summarized below for each event.

1. UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL FROM ASUBCRITICAL CONDITION

Event Definition: A rod cluster control assembly (RCCA) bank withdrawal accident is defined asan uncontrolled addition of reactivity to the reactor core caused by withdrawal of one or moreRCCA banks, resulting in a power excursion. This could occur with the reactor eithersubcritical, at hot zero power, or at power.

Plant Operating Conditions: The plant is assumed to be operating at the no-load reactor coolantaverage temperature with a power level of 1x10 9 of nominal.

Effect of Proposed Change: In this scenario, the reactor is not critical and the turbine generatoris not on-line. Therefore, the proposed changes have no effect on this accident scenario andthe conclusions of the UFSAR remain valid.

2. UNCONTROLLED CONTROL ROD ASSEMBLY WITHDRAWAL AT POWER

Event Definition: This event is defined as the inadvertent addition of positive reactivity to thecore caused by the uncontrolled withdrawal of an RCCA bank(s) while at power.

Plant Operating Conditions: Initial power levels of 100, 80, 60 and 10 percent of nominal RatedThermal Power are analyzed. For all cases analyzed, the results show that integrity of the coreis maintained by the reactor protection system (RPS) as the departure from nucleate boiling(DNBR) remains above the safety analysis limit value.




Effect of Proposed Change: In this scenario, the reactor trip is provided by the automaticactuation of the first primary side reactor protection signal reached: either overtemperature deltatemperature (OTAT) or power range high neutron flux. Neither of these reactor protectionfunctions is affected by the turbine trip signal, or by the change in the P-8 setpoint. Thereforethe proposed changes have no effect on this accident scenario and the conclusions of theUFSAR remain valid.

3. ROD CLUSTER CONTROL ASSEMBLY (RCCA) DROP

Event Definition: The dropped RCCA accident is initiated by a single electrical or mechanicalfailure which causes any number and combination of rods from the same group of a given bankto drop to the bottom of the core.

Plant Operating Condition: The analysis is performed with the plant at full power.

Effect of Proposed Change: There is no reactor trip credited in the analysis. This analysis isnot affected by the reactor trip on turbine trip setpoint or by the change in the P-8 setpoint value.The reactor trip on turbine trip function is not credited for this event as either a primary orbackup trip. Therefore, the proposed changes have no effect on this accident scenario and theconclusions of the UFSAR remain valid.

4. CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION

Event Definition: This event is the inadvertent dilution of the reactor coolant system (RCS)boron concentration. This event is caused by a chemical and volume control system (CVCS)malfunction or faulty operator action. The limiting scenario considered is the inadvertentopening of the primary water makeup control valve and failure of the blend system, either bycontroller or mechanical failure, resulting in the addition of unborated water into the RCS.

Plant Operating Condition: The analysis is performed for an inadvertent dilution of the RCS forpower operation (mode 1), startup (hot zero power) and refueling modes of plant operation.

Effect of Proposed Change: In Mode 1, the power and temperature rise will cause the reactor toreach the OTAT trip setpoint resulting in a reactor trip. In Mode 2, the power range high neutronflux (low setpoint) function provides the trip. Neither of these reactor trip functions is affected bythe reactor trip on turbine trip setpoint or by the change in the P-8 setpoint. Therefore, the




proposed changes have no effect on this accident scenario and the conclusions of the UFSARremain valid.

5. STARTUP OF AN INACTIVE REACTOR COOLANT LOOP

Event Definition: The inadvertent startup of an idle loop while operating in an N-1 loop conditionresults in the sudden introduction of colder water into the core from the idle loop which couldcause an unplanned reactivity insertion and power increase.

Plant Operating Condition: N/A. See below.

Effect of Proposed Change: The Turkey Point Technical Specifications preclude operation ofthe plant with one or more loops out of service. Therefore, this event no longer applies and hasbeen removed from the plant's licensing basis. The proposed changes, therefore, have noeffect on this accident scenario.

6. EXCESSIVE FEEDWATER FLOW AND REDUCTION IN FEEDWATER ENTHALPY INCIDENT

Event Definition: This event is defined as an increase in feedwater flow to one or more of thesteam generators or a decrease in feedwater temperature. This event will result in an increasein the heat transfer rate from primary to secondary in the steam generators and a consequentialreduction in primary system temperature and pressure. The transient responses for anexcessive feedwater flow event to one steam generator were analyzed for four cases: two casesat hot full power (one case with automatic rod control and one without) and two cases at hotzero power (one case with automatic rod control and one without).

Plant Operating Condition: This event is analyzed at power levels corresponding to zero andfull load.

Effect of Proposed Change: For full power conditions, feedwater isolation and turbine trip (withsubsequent reactor trip signal on turbine trip) occur on the high-high steam generator waterlevel signal. For the zero power cases, although there is no reactor trip credited in the analysis,the reactor may be tripped by the power range high neutron flux trip (low setting).

Currently, the reactor trip on turbine trip function is disabled below 10% power via the P-7setpoint. The proposed change is to move the reactor trip on turbine trip from P-7 to P-8, whichresults in the trip function being disabled below 40% power (the new P-8 setpoint). As noted




above, cases are performed for the Feedwater Malfunction analysis at 100% and 0% power.These cases are unaffected by the proposed changes, since the reactor trip on turbine tripfunction is still available at 100% power. For the zero power cases, the reactor may be trippedby the power range high neutron flux trip (low setting), although this trip is not credited and theminimum DNBR would not change significantly without it.

For power levels below 40%, where the turbine trip / reactor trip function is disabled, the eventwould be no more severe than the 100% power case. Although credited in the analysis for thefull power cases, the reactor trip on turbine trip is not a critical function that is required in orderto get acceptable results. By the time that the high-high SG level trip is reached, the plantreaches a semiequilibrium state. The RCS temperature, power, and the DNBR are leveling off.The event is effectively terminated when the turbine is tripped and feedwater isolated via thehigh-high steam generator level trip. The reactor trip on turbine trip is modeled since it isexpected to occur. However, it is not considered primary protection for the event and theminimum DNBR calculated would not change significantly without it.

Based on the above, the proposed changes do not invalidate the results of the analysis and theconclusions of the UFSAR remain valid.

7. EXCESSIVE LOAD INCREASE INCIDENT

Event Definition: An excessive load increase event is defined as a rapid increase in the steamflow that causes a power mismatch between the reactor core power and the steam generatorload demand. The reactor control system is designed to accommodate a 10% step-loadincrease or a 5% per minute ramp load increase in the range of 15 to 100% power. Any loadingrate in excess of these values may cause a reactor trip by the reactor protection system.

Plant Operating Condition: The event is analyzed at full power conditions and assumes a 10%step load increase.

Effect of Proposed Change: Although the RPS is assumed to be operable, a reactor trip doesnot occur in this analysis. Therefore, the proposed changes have no effect on this accidentscenario and the conclusions of the UFSAR remain valid.

8. LOSS OF REACTOR COOLANT FLOW

A) Flow Coastdown Accidents




Event Definition: The loss of flow incident can result from a mechanical or electrical failure in areactor coolant pump (RCP), or from a fault in the power supply of these pumps.

Plant Operating Condition: The plant is assumed to be operating at full power. Boundinganalyses are performed at full power since this is the most conservative in terms of potentialconsequences, specifically a more limiting minimum DNBR.

Effect of Proposed Change: The low primary coolant loop flow, RCP undervoltage, RCPunderfrequency, and RCP breaker position reactor trip functions provide the necessaryprotection for this event. These trips are not affected by the reactor trip on turbine trip setpoint.

The change in P-8 setpoint from 45% to 40% also does not have any adverse effect on theanalysis. This permissive defines the highest steady state power level at which the reactor canoperate with one RCS loop inactive without violating the N-1 core thermal limits. A reduction inthe P-8 setpoint is conservative since the core thermal limits are less likely to be violated atlower power levels. Therefore, the proposed changes have no effect on this accident scenarioand the conclusions of the UFSAR remain valid.

B) Locked Rotor Accident

Event Description: The design basis reactor coolant pump shaft seizure event is defined as aninstantaneous seizure of a single RCP rotor which results in a rapid reduction in reactor coolantloop flow.

Plant Operating Condition: This event is analyzed assuming that the plant is operating atmaximum reactor coolant pressure and temperature, and maximum power when the eventoccurs.

Effect of Proposed Change: The reactor trip is initiated by low primary coolant loop flow. Thistrip is not affected by the reactor trip on turbine trip setpoint. Therefore, the proposed changeshave no effect on this accident scenario and the conclusions of the UFSAR remain valid.

9. LOSS OF EXTERNAL ELECTRICAL LOAD

Event Definition: The loss of external electrical load and/or turbine trip event is defined as acomplete loss of steam load from full power without a direct reactor trip, or a turbine trip withouta direct reactor trip.




Plant Operating Condition: The analysis assumes a complete loss of steam load from full powerwith no credit taken for the direct reactor trip on turbine trip.

Effect of Proposed Change: Protection for this event is provided by the OTAT, high pressurizerpressure, or low-low steam generator water level signals. The loss of external electricalload/turbine trip event from a full power condition bounds a turbine trip with no subsequentreactor trip from 10% power (P-7) as well as from 40% power (proposed value for P-8).Therefore, the proposed changes have no effect on this accident scenario and the conclusionsof the UFSAR remain valid.

10. LOSS OF NORMAL FEEDWATER FLOW

Event Definition: The design basis loss of normal feedwater event is defined as a reduction inthe capability of the secondary system to remove heat generated in the reactor core.

Plant Operating Condition: A complete loss of main feedwater flow is assumed to occur from102% of Rated Thermal Power. Maximum initial RCS temperature and pressure conditions areassumed.

Effect of Proposed Change: The reactor trip is initiated by low - low steam generator waterlevel. This trip is not affected by the changes. Therefore, the proposed changes have no effecton this accident scenario and the conclusions of the UFSAR remain valid.

11. LOSS OF NON-EMERGENCY AC POWER TO THE PLANT AUXILIARIES

Event Definition: A complete loss of non-emergency AC power may result in the loss of allpower to the plant auxiliaries: i.e., the RCPs, condensate pumps, etc. The loss of power maybe caused by a complete loss of the offsite grid accompanied by a turbine generator trip at thestation, or by a loss of onsite non-emergency AC distribution system.

Plant Operating Condition: The plant is initially operating at 102% of rated thermal power.Maximum initial RCS temperature and pressure conditions are assumed.

Effect of Proposed Change: The reactor trip is initiated by the low - low steam generator waterlevel trip function. This trip mechanism is not affected by the proposed changes. Therefore,the proposed changes have no effect on this accident scenario and the conclusions of theUFSAR remain valid.




12. RUPTURE OF A STEAM PIPE

Event Definition: A rupture of a steam pipe is assumed to include any accident which results inan uncontrolled steam release from a steam generator. Such a release may result from eitherthe opening of a steam generator relief or safety valve, or from a steam system pipe break.

Plant Operating Condition: The analysis assumes that the reactor is initially at hot shutdownconditions.

Effect of Proposed Change: Protection for this event is provided by the overpower reactor trips(neutron flux and AT), and the reactor trip occurring in conjunction with receipt of the SafetyInjection Signal. The limiting zero power analysis does not specifically credit the reactor tripsystem. Only the Engineered Safety Features Actuation System (ESFAS) is needed to limit theconsequent of the analyzed events. Therefore, the proposed changes have no effect on thisaccident scenario and the conclusions of the UFSAR remain valid.

13. RUPTURE OF A CONTROL ROD MECHANISM HOUSING - RCCA EJECTION

Event Definition: This event is an assumed failure of a control rod mechanism pressure housingsuch that the RCS pressure would eject the control rod and drive shaft.

Plant Operating Condition: Both full and zero power cases are analyzed.

Effect of Proposed Change: The reactor will trip on either the power range high neutron flux lowsetpoint, or the high setpoint. These trip mechanisms are not affected by the proposedchanges. Therefore, the proposed changes have no effect on this accident scenario and theconclusions of the UFSAR remain valid.





SECTION 5SUMMARY OF EVALUATIONS

The following evaluations were completed in support of the changes:

1. LOSS OF COOLANT ACCIDENT (LOCA) AND LOCA-RELATED EVALUATIONS

The following LOCA related analyses are not adversely affected by changing the reactor trip onturbine trip permissive from P-7 to P-8 or by changing the P-8 setpoint from 45% to 40%:

1. Large and small break LOCA

2. Reactor vessel and loop LOCA blowdown forces (the LOCA blowdown forces are analyzedat 100% power, thus reactor trip is not a relevant input to this analysis)

3 Post-LOCA long term core cooling subcriticality

4. Post-LOCA long term core cooling minimum flow and hot leg switchover to prevent furtherboron precipitation

The changes do not affect the normal plant operating parameters, the safeguards systemsactuation or accident mitigation capabilities important to LOCA, or the assumptions used in theLOCA related accidents. Nor do the changes create conditions more limiting than thoseassumed in these analyses.

2. NON-LOCA RELATED EVALUATION

Each of the non-LOCA accident analyses described in Chapter 14 of the Turkey Point UFSARwas reviewed with respect to changing the reactor trip on turbine trip permissive from P-7 to P-8and the P-8 setpoint from 45% to 40%. Based on this review, it is concluded that the proposedchanges have no effect on the accident analyses. This review is summarized above in Section4, Evaluation of Non-LOCA Events.

It is concluded that the non-LOCA safety analyses presented in Chapter 14 of the UFSAR arenot adversely affected by changing the reactor trip on turbine trip permissive from P-7 to P-8 orby changing the P-8 setpoint from 45% to 40%. Additionally, normal plant operatingparameters, accident mitigation capabilities, and assumptions used in the non-LOCA transients




are not adversely affected. The changes will not create conditions more limiting than thoseconsidered in the current non-LOCA analyses. Therefore, changing the reactor trip on turbinetrip permissive from P-7 to P-8 and changing the P-8 setpoint from 45% to 40% do not alter theconclusions presented in the UFSAR.

3. MAIN STEAMLINE BREAK (MSLB) MASS AND ENERGY RELEASE

Changing the reactor trip on turbine trip permissive from P-7 to P-8 and changing the P-8setpoint from 45% to 40% do not affect either the inside or outside containment MSLB massand energy release, or the calculations for the steam mass release used as input to theradiological dose evaluation. For this event, High Containment Pressure initiates safetyinjection and the safety injection signal produces a reactor trip signal. The reactor trip onturbine trip function is not credited for this event. The changes do not affect the normal plantoperating parameters, input assumptions including accident mitigation capabilities, results,parameters or conclusions of the MSLB mass and energy release analyses and calculations.Therefore, the conclusions presented in the UFSAR remain valid with respect to MSLB massand energy release rates and steam mass release calculations.

9. STEAM GENERATOR TUBE RUPTURE (SGTR) EVALUATION

Changing the reactor trip on turbine trip permissive from P-7 to P-8 and changing the P-8setpoint from 45% to 40% do not affect the SGTR analysis methodology or assumptions. Forthis event, Low Pressurizer Pressure initiates safety injection and the safety injection signalproduces a reactor trip signal. The reactor trip on turbine trip function is not credited for thisevent. These changes do not alter the current SGTR event analysis results. Thus, theconclusions presented in the UFSAR remain valid with respect to the SGTR event.





SECTION 6CONCLUSION

The analyses and evaluations have shown that, with all control systems assumed operational,setting the P-8 setpoint to 40% power, and applying this permissive to block reactor trip on turbine

trip below this power level is acceptable.

Based on this evaluation, blocking the automatic reactor trip on turbine trip at the proposed P-8

permissive setpoint (40% power) can be supported, since the current licensing basis safety

analyses have been shown to remain valid, and the requirements of NUREG 0737, Item II.K.3.10

have been met.

A review was performed in accordance with 1OCFR50.92, to determine if the proposed changes to

the Turkey Point Units 3 and 4 Technical Specifications involve a significant hazards consideration.Based on the review it has been determined that the proposed change of the reactor trip on turbine

trip permissive from P-7 to P-8 and of the P-8 setpoint from 45% to 40% power does not (1)

significantly increase the probability or consequences of an accident previously evaluated, (2) does

not create the possibility of a new or different kind of accident than any accident already evaluated

and (3) does not involve a significant reduction in a margin of safety, and therefore does not involve

a significant hazards consideration.



L-2005-006ENCLOSURE 2

Page 1 of 4

ENCLOSURE 2

NO SIGNIFICANT HAZARDS CONSIDERATIONS

L12005-006ENCLOSURE 2

Page 2 or 4

Introduction

The proposed amendments revise the Turkey Point Technical SpecNfications for several Reactor TripSystem functional units as described below:

* Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints

Trip functional unit 12, steam/feedwater flow mismatch coincident with steam generator waterlevel - low, together with the associated allowable values, and trip setpoints are being deleted.

* Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints

Trip functional unit 1 7c, Reactor Trip System Interlock, Power Range Neutron Flux, P-8, theallowable value is being changed from < 48.0% rated thermal power (RTP) to < 43% RTP, andthe trip setpoint is being changed from nominal 45% of RTP to nominal 40% of RTP.

* Table 3.3-1, Reactor Trip System Instrumentation

Trip functional unit 12, steam/feedwater flow mismatch coincident with steam generator waterlevel - low, the operability requirements are being deleted.

* Table 4.3-1, Reactor Trip System Instrumentation Surveillance Requirements

Trip functional unit 12, steam/feedwater flow mismatch coincident with steam generator waterlevel - low, the surveillance requirements are being deleted.

Determination of No Significant Hazards Consideration

The standards used to arrive at a determination that a request for amendment involves a nosignificant hazards consideration are included in the Commission's regulation, 10 CFR 50.92, whichstates that no significant hazards considerations are involved if the operation of the facility inaccordance with the proposed amendment would not (1) involve a significant increase in theprobability or consequences of an accident previously evaluated; or (2) create the possibility of anew or different kind of accident from any accident previously evaluated; or (3) involve a significantreduction in a margin of safety. Each standard is discussed as follows:

(1) Operation of the facility in accordance with the proposed amendment would notinvolve a significant increase in the probability or consequences of an accidentpreviously evaluated.

The proposed changes revise the operability requirements, surveillance requirements and theinterlock setpoint for two Reactor Trip System functional units. The affected trip functionalunits are not initiators of any accident previously evaluated. The proposed changes to theaffected trip functional units do not adversely affect the initiators of any accident previouslyevaluated. A best estimate analysis has shown that a turbine trip without a reactor trip below40% power does not challenge the pressurizer PORVs or the steam generator safety valves;


Page 3 of 4

thereby, not adversely affecting the probability of a small break LOCA due to a stuck openPORV, or an excessive cooldown event due to a stuck open steam generator safety valve.As a result, the probability of any accident previously evaluated is not significantly increasedby the proposed changes.

The steam/feedwater flow mismatch coincident with steam generator water level - lowreactor trip is not credited as a primary trip in any previously evaluated accidents. Thereactor trip on turbine trip below the P-8 interlock is not credited as a primary trip in anypreviously evaluated accidents. Therefore, the mitigation functions that have been assumedin the accident analyses will continue to be performed by the systems and componentscurrently credited in the analyses; and the accident analysis results are not affected by thechanges to the affected trip functional units. The P-8 setpoint is not an initial condition ofany accident previously evaluated. Therefore, the accident analysis results are not affectedby changes to the P-8 setpoint. No safety analyses previously performed in the Turkey PointUnits 3 and 4 UFSAR required reanalysis for these proposed changes. All accident analysesacceptance criteria continue to be met. The proposed changes do not create any new crediblelimiting single failure. As a result, the consequences of any accident previously evaluated arenot significantly increased by the proposed changes.

In conclusion, operation of the facility in accordance with the proposed amendments does notinvolve a significant increase in the probability or consequences of any accident previouslyevaluated.

(2) Operation of the facility in accordance with the proposed amendments would not createthe possibility of a new or different kind of accident from any previously evaluated.

No changes are being made to the plant that would introduce any new accident causalmechanisms. The proposed changes do not adversely affect previously identified accidentinitiators and do not create any new accident initiators. No new limiting single failures oraccident scenarios are created by the proposed changes. No new challenges to any installedsafety system are created by these proposed changes. The proposed changes do not result inany event previously deemed incredible being made credible.

The steam/feedwater flow mismatch coincident with steam generator water level - lowreactor trip is not credited as an inhibitor of any potential or actual accident initiators. So,deletion of this reactor trip functional unit will not create the possibility of a new or differentkind of accident from any previously evaluated.

Changing the interlock for the reactor trip on turbine trip from P-7 to P-8 changes the powerlevel associated with enabling and disabling the reactor trip on turbine trip function. Theturbine pressure input to the reactor protection system permissives is not an accident initiatorand is not credited in the accident analyses. Changing the P-8 allowable and trip setpointvalues changes the power level associated with enabling and disabling the reactor tripfunctions currently associated with P-8. The change does not affect how the associated tripfunctional units operate or function. Since these interlock changes do not affect the way that


Page 4 or 4

the associated trip functional units operate or function, the changes do not create thepossibility of a new or different kind of accident from any previously evaluated.

Therefore, operation of the facility in accordance with the proposed amendments does notcreate the possibility of a new or different kind of accident from any previously evaluated.

(3) Operation of the facility in accordance with the proposed amendments would notinvolve a significant reduction in a margin of safety.

No UFSAR safety analyses were changed or modified as a result of these proposed changes.Therefore, all margins associated with the current UFSAR safety analyses acceptance criteriaare unaffected. The current UFSAR safety analyses remain bounding. No UFSAR Chapter14 events explicitly credit the steam / feedwater flow mismatch reactor trip function and thereactor trip on turbine trip function below the P-8 setpoint value. The safety systems creditedin the safety analyses will continue to be available to perform their mitigation functions.Changing the P-8 setpoint from 45% to 40% is in the conservative direction for the ReactorCoolant Flow - Low Reactor Trip and the Reactor Coolant Pump Breaker Position ReactorTrip. Therefore, the proposed changes do not result in a significant reduction in a margin ofsafety; and operation of the facility in accordance with the proposed amendments would notinvolve a significant reduction in a margin of safety.

Based on the above, it has been determined that the proposed amendment does not (1) involve asignificant increase in the probability or consequences of an accident previously evaluated, (2) createthe possibility of a new or different kind of accident from any previously evaluated, or (3) involve asignificant reduction in a margin of safety; and therefore does not involve a significant hazardsconsideration.


Page 1 of 2

ENCLOSURE 3

ENVIRONMENTAL CONSIDERATION


Page 2 of 2

Environmental Consideration

The proposed license amendments change requirements with respect to installation or use of afacility component located within the restricted area as defined in 10 CFR Part 20. The proposedamendments involve no significant increase in the amounts and no significant change in the types ofany effluents that may be released off-site, and no significant increase in individual or cumulativeoccupational radiation exposure. FPL has concluded that the proposed amendments involve nosignificant hazards consideration, and therefore, meet the eligibility criteria for categorical exclusionset forth in 10 CFR 51.22(c)(9). Hence, pursuant to 10 CFR 51.22(b), an environmental impactstatement or environmental assessment need not be prepared in connection with issuance of theamendments.


Page I of 17

ENCLOSURE 4

PROPOSED MARK-UP OF

AFFECTED TECHNICAL SPECIFICATIONS

AND

(FOR INFORMATION ONLY) BASES PAGES

L-2005-006ENCLOSURE4

Page 2 of17

Provided For Information Only - No Change To This Page

ig fRELACTOR

fflj FUNCTIONAL UNIT

Z 1. Manual Reactor Trip

2. Power Range. Neutron Fluxa. High Setpolntb. Low Setpolnt

3. Intermediate Range, Neutron Flux

4. Source Range. Neutron Flux

5. Overtemperature AT

6. Overpower AT

7. Pressurizer Pressure-Low8. Pressurizer Pressure-High

9. Pressurizer Water Level-Hlgh

t 10. Reactor Coolant Flow-Low

-a

K II. Steam Generator Water Level Low-Lowm

CD

Loop design flow = 85,000 gpmRTP = Rated Thermal Power

TABLE 2.2-1

TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

ALLOWABLEVALUE

NA.

S 112.0% of RTP**s28.0% of RTP'

S 3 1.0% of RTP

51.4 X 105 cps

See Note 2

See Note 4

1817 psIg5 2403 psrg

5 92.2% of Instrument span

2 88.8% of loop design flow

Ž 8.15% of narrow rangeInstrument span

TRIP SETPOINT

NA.

s 109.0% of RTP'*S 25% of RTP*

S 25% of RTP**

5 10o cps

See Note 1

See Note 3

a 1835 pslgs 2385 pslg

5 92% of Instrument span

290% of loop design fow'

Ž10% of narrow rangeInstrument span

,-2005-006ENCLOSURE 4

Page 3 or17

IC

cogo

FUNCTIONAL UNIT

12. SteamlFeedwater Flow MisrrCoincident with

Steam GeneratorWate (

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

ALLOWABLEVALUE

latch Feed FlowS 23.9% belowrated Steam Flow

DI-Low+ 8.15% of narrow rangeinstrument span

Ses A and B 2Ž69% bus voltage

actor Coolant k 55.9 Hz

13. Undervoltage-4.18 kV Buss

14. Underfreruencv - TrID of RePump Breaker(s) Open

15. Turbine Trip

a. Auto Stop Oil Pressure

b. Turbine Stop Valve Closure

16. Safety Injection Input from ESF

> 17. Reactor Trip System Interlocksm

a. Intermediate Range Neutron Flux. P-6

I0

oi mSwiuh e Ives are fully clsed.

-f / A SE IT T E ceco

TRIP SETPOINT

Feed Flow S20% below ratedSteam Flow

a O1% of narrow rangeinstrument span

2 70% bus voltage

Ž 56.1 Hz

Ž 45 psig

Fully Closed-*

NA.

Nominal 1 X 10 -'° amps

2Ž42 psig

Fully Closed**

N.A.

26.0 X IO 'I amps

L,2005-006ENCLOSURE 4

Page 4 of 17

INSERTADDITONAL FOOTNOTES FOR TS page 2-5:

+ Not applicable to Unit 3.Only applicable to Unit 4 through Cycle 22 operation.Not applicable to Unit 4 starting with Cycle 23 operation.

L2005-006ENCLOSURE 4

Page 5 of 17

TABLE 2.2-1 (Continuecdl

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS*00 AALLOWABLE

FUNCTIONAL UNIT VALUE TRP SETPOINT

Z b. Low Power Reactor Trips Block, P-7 *

1) P-10 input s 13.0%A RTP'* Nominal 10% of RTP*'

A 2) Turbine First Stage Pressure S 13.0% Turbine Power Nominal 10% Turbine Power

c. Power Range Neutron Flux, P-8 m

d. Power Range Neutron FIux, P-10 7.0% RTP'* Nominal 10% of RTP'*

18. Reactor Coolant Pump Breaker PositIon NA NATrip

19. Reactor Trip Breakers N.A.| N.A.

20. Automatic Trip and Interlock Logic N.A. NA.

mz

z0

z

o RER

0

CD.,{SR se ¢ tt

,-4


Page 6 of 17

INSERTADDITONAL FOOTNOTES FOR TS page 2-6:

+ Only applicable to Unit 4 through Cycle 22 operation.

++ Applicable to Unit 3.Applicable to Unit 4 starting with Cycle 23 operation.

L,2005-006ENCLOSURE 4

Page 7 of 17

z-I

C2

CftCa)

Provi

FUNCTIONAL UNIT

1. Manual Reactor Trip

2. Power Range. Neutron Fluxa. High Setpolntb. Low Setpoint

3. Intermediate Range, Neutron Flux

4. Source Range, Neutron Fluxa. Startupb. Shutdown*'c. Shutdown


6. Overpower AT

7. Pressurizer Pressure-Low(Above P-7)

8. Pressurizer Pressure-High

9. Pressurizer Water Level-High(Abova P-7)

10. Reactor Coolant Flow-Lowa. Single Loop (Above P4)b. Two Loops (Above P-7

and below P-8)

Ided For Information Only - No Change To This Page

TABLE 3.3-1

REACTOR TRIP SYSTEM INSTRUMENTATION

MINIMUMTOTAL NO. CHANNELS CHANNELS

OF CHANNELS TO TRIP OPERABLE

2 1 22 1 2

44

2.

222

3

3

3

3

.3

3/loop3/10p

22

1

I01

2

2

2

2

2

21loop2Jnoop

33

2

222

2

2

2

2

2

ZIioop211oop

APPLICABLE_MODES_

1,23-.4',5-

1.21##, 2

2#3,4,53', 4*, 5'

1.2

1,2

1

1,2

I

1I

ACTION

9

223

4

59

13

13

62z

z0fn

CD0

6

13

I

66


Page 8 of 17

;2

A

z

(n

90

TABLE 3.3-1 (Continued)l


FUNCTIONAL UNIT

11. Steam GeneratorWaterLevel- Low-Low

12. Steam Generator Water Level-Low Coincident V~frStiFeedwater F i smatch J

AA

SUPERSC"*_,.r

TOTAL NO.OF CHANNELS

3/stm. gen.

2 stm. gen.level and2 stm.lfeed-water flowmismatch ineach stm. gen.

MINIMUMCHANNELS CHANNELS

TO TRIP OPERABLE

2/stm. gen. 2/stm. gen.

APPLICABLEMODES

1,2

1.2

6

ACTION

1 stm. gen.level coin-cident withI stni.feed-water flowmismatch Insame stm.gen.

1 stm. gen.level and2 stm./feed-water Rowmismatch insame stm. gen.or 2 stm. gen.level and 1stmifeedwaterflow mismatchIn same stm.gen.

6

C.,

z

z

z40

t-L

13. Undervottage-4.16 KV BussesA and B (Above P-7)

14. Underfrequency-Trip of ReactorCoolant Pump Breaker(s) Open(Above P-7)

15. Turbine Trip (Abovea. Autostop Oil Pressureb. Turbine Stop Valve Closure

IA/KST rzTr l E )fl_(SC*- ne7t ?41e} _

2/bus

2Jbus

32

1bus onboth busses

1 to tripRCPs---

22

21bus

2/hus

1 12

1 11

22

1i1

1212


Page 9 of 17

INSERT NEW FOOTNOTE FOR TS PAGE 3/4 3-3:



Page 10 of 17

Provided For Information Only - No Change To This Page .

C4

C.)

TABLE 4.3-'

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

FUNCTIONAL UNIT

1. Manual Reactor Trip2. Power Range. Neutron Flux

a. HIgh Setpoint

CHANNELCHECK-

NA

S

(a

b. Low Setpodnt3. IntermedIlate Range,

Neutron Flux4. Source Range, Neutron Flux


6. Overpower AT

7. Pressurizer Pressure-Low8. Pressurizer Pressure-HIgh9. Pressurizer Water Level-Hfgh

10. Reactor Coolant Flow-Low11. Steam Generator Water Level-

Low-Low

SS

SS

S

SSSSS

CHANNELCALIBRATION

N.A.

D(2. 4),M(3. 4),0(4.B).R(4)

R(4j

R(4)

R(4)

R

R

R

R

R

R

R

ANALOGCHANNELOPERATIONALTEST

N.A.

Q N.A.

TRIPACTUATINGDEVICEOPERATIONALTEST

R(11)

ACTUATIONLOGIC TEST

N.A.

NA

MODES FORWHICHSURVEILLANCEIS REQUIRED

.23*. 4* 5*

SAJ(1)

SAU(1)

SUM1), 0(9)a

a

a

0

aa

a

1.2

NA

NA.

NA

NA

NA

N.A.

NA

NA.

NA.

N.A.

NAN.A.

NAN.A.N.A.

N.A.N.A.NANANA

1 '. 2

111*, 2

I

2'.3,4,5

1,2

1.2

1

1,2

1

1

1,2

Izn0m

z0!a1C4

co0


Page 11 of 17

m

aI

ZC

-P1

TABLE 4.3.1


CHANNEL CHANNELFUNCTIONAL UNIT CHECK CLIREATION

12. Steam GeneratorWater S RLevel-Low Coincident with

_,b Pjer Flow t15R A

13. Undervoltage-4.16kV NA. RBusses A and B

ANALOGCHANNELOPERATIONALTEST

a

TRIPACTUATINGDEVICEOPERATIONALTEST

N.A.

ACTUATIONLOGIC TEST

N.A.

MODES FORWHICHSURVEILLANCEIS REQUIRED

1 2 I

NA NA. N.A. I

9Se

14. Underfrequency-Trip ofReactor Coolant PumpBreakers(s) Open

15. Turbine Tripa. Autostop Oil Pressureb. Turbine Stop Valve

Closure

16. Safety Injection Inputfrom ESF

17. Reactor Trip SystemInterlocksa. Intermediate Range

Neutron Flux. P-8

b. Low Power ReactorTrips Block. P-7(Indudes P-10 Inputand Turbine FirstStage Pressure)

NA.

N.A.

NA

N.A.

NA.

NA.

R

R

NA

NA.

R NA.

za

r-

z0

CD

NA.

R(4)

R(4)

R(4)

NA.

R

R

R

NA.

SNU(1, 10)

SIU(1, 10)

R

NA.

NA.

NA.

NA.

N.A.

NA

NA.

NA

NA.

NA

I

I

I

1,2

26

I

1c. Power Range Neutron

Flux P4 NA.

NsreRT mvo74 4sTF


Page 12 of 17

INSERTNEWFOOTNOTE FORTS PAGE 3/4 3-9:


1-2005-006ENCLOSURE 4

Page 13 of 17

Peiced= No.: hocedwe Thk: Page:

17Approval Date:

O-ADM-536 Technical Specification Bases Control Program 9116/04

(/IV Mtd;Teo ATTACHMENT IO0 eAL (Page7of 103)

TECHNICAL SPECIFICATION BASES

2.2 LTMiTING SAFETY SYSTEM SETTINGS (Continued)

Reactor Coolant Flow

The Reactor Coolant Flow-Low trip provides core protection to prevent DNB by miti ating theconsequences of a loss of flow resulting from the loss of onc or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately I 0/o of RATED THERMAL P0 Ror a turbine first stage pressure at approximately 10% of full power equivalent), an automatic crtrip will occur if the flow in more than one loop drops below 90% of loop design flow. Abov P-8_ ar.. C¢. Of appimintielyll 4556 of R TI:D _III:N Ab FPWERS an automatic Reactor tioccur if the flow in any single loop drop eo 0M of nominal full loop flow. Conversely, andecreasing power between P-8 and the P7aiutmi Reactor trip will occur on low reactor coolantflow in more than one loop and below P- teri on is automatically blocked.

Steam Generator Water LevelPAEMT

The Steam Gencrator Water Level Low-Low trip protects the reactor from loss of beat sink in the eventof a sustained steam/feedwater flow mismatch resulting from loss of normal feedwater. The specifiedsctpoint provides allowances for starting delays of the Au~xiliary Feedwater System.

Steam/Feedwater Flow Mismatch and Low Steam Generator Watee

The StearnfFeedwater Flow Mismatch in coincidence with a Steam Generator Water Level-Low trip isnot used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functionalcapability of the specified trip settings and thereby enhance the overall reliability of the Reactor TripSystem. This trip is redundant to the Steam Generator Water Level Low-Low trip. TheStcanm/Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds thefeedwater flow by greater than or equal to 0.665 x 10 lbs/hour. The Steam Generator Water Level-Lowportion of the trip is activated when the water level drops below 10%, as indicated by the narrow rangeinstrument. These trip values include sufficient allowance in excess of normal operating values topreclude spurious trips but will initiate a Reactor trip before the steam generators are dry. Therefore, therequired capacity and starting time requirements of the auxiliary feedwater pumps are reduced and theresulting thermal transient on the Reactor Coolant System and steam gencrators is 'Minimized.

C %s T-i Foo7NoTEc set~ nee ?c-sq)

Vwo7 need - ha_-1 -a .urorrlrormru-


Page 14 of 17

INSERT NEW FOOTNOTE FOR TS BASES PAGE 17:

+ A power level of approximately 40% of RATED THERMAL POWER- applicable to Unit 3-- applicable to Unit 4 starting with Cycle 23 operation

A power level of approximately 45% of RATED THERMAL POWER- applicable to Unit 4 through Cycle 22 operation

++ Not applicable to Unit 3.Only applicable to Unit 4 through Cycle 22 operation.Not applicable to Unit 4 starting with Cycle 23 operation.


Page 15of17

Prcedure No.: Procedure oak: pagee18

Arano" Datc;0*ADM1-536 Technical Specification Bases Control Program 9/16104 |

(F R 0 ) ATTACHMENT 1N_ (Page 8 of 103)


2.2 LIMITING SAFETY SYSTEM SETTINGS (Continued)

Undervoltaze - 4.16 kV Bus A and B Trips

The 4.16 kV Bus A and B Undervoltage trips provide core protection against DNB as a result ofcomplete loss of forced coolant flow. The specified setpoint assures a Reactor trip signal is generatedbefore the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Undervoltage trips toprevent spurious Reactor trips from momentary electrical power transients. The delay is set so that thetime required for a signal to reach the Reactor trip breakers following the trip of at least oneundervoltage relay in both of the associated Units 4.16 kV busses shall not exceed 1.3 seconds. Ondecreasing power the Undervoltage Bus trips are automatically blocked by P-7 (a power level ofapproximately 10% of RATED THERMAL POWER with a turbine first stage pressure at approximately100 of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trip o

A Turbine trip initiates a eactor trip. On decreasing power, the Reactor Trip from the Turbine trip isautornatically blce pyn ower le-we! ef nppo .vir.-tl -F9 RAfETN D lllERlMAIL romrRsnwt a _tiiefr!saer~faey19-CAl QY and on increasingpowrr, reinstated automatically by.

Safetv Iniection Input from ESF ~

If a Reactor trip has not already been generated 'by the Reactor Trip System instrumentation, the SFautomatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a fetyInjection. The ESF instrumentation channels which initiate a Safety Injection signal are s wn inTable 3.3-3.

Reactor Coolant Pump Breaker Position Tri/

The Reactor Coolant Pump Breaker Position Trips a anticipatory trips which provide reacar coreprotection against DNB. The open/close position s assure a reactor trip signal is generated fore thelow flow trip sctpoint is reached. Their functio capability at the open/close position ttings isrequired to enhance the overall reliability of the Rea r Protection System. Above P-7 (a wcr levelof approximately 10% of RATED THERMAl OWER or a turbine first stage ressure atapproximately 10% of full power equivalent) a to atic reactor trip will occur if re than onereactor cooglant pump breaker is opened. Cbri P-8V pa"c. l of rqlkys x..tcI G fITDUtERMAL rPE) ) an automatic reactor trill occur if one reactor coolant pump breaker isopened. On decreasing power between P-8 and P-7, an automatic reactor trip will occur if more thanone reactor coolant pump breaker is opened and below P-7 the trip function is automatically blocked.

NOSCR / -0077Vo-rr (seeot pag)

WV97:PSimJrakav


Page 16 or 17

INSERT NEW FOOTNOTE FOR TS BASES PAGE 18:

+ A power level of approximately 40% of RATED THERMAL POWER- applicable to Unit 3- applicable to Unit 4 starting with Cycle 23 operation

A power level of approximately 45% of RATED THERMAL POWER- applicable to Unit 4 through Cycle 22 operation


Page 17 or 17

Pxeedue No-: Pr0=d ~tk: Page:19

Appronl Date

O-ADM-536 Technical Specification Bases Control Program 9116104

ATTACHMENT I(Page 9 of 103)



Reactor Coolant Pump Breaker Position Trip (Continued)

Undcrfrequency sensors are also installed on the 4.16 kV busses to detect underfrequency and initiatebreaker trip on underfrequency. The undcrfrequency trip setpoints preserve the coast down energy ofthe reactor coolant pumps, in case of a grid frequency decrease so DNB does not occur.

Reactor Trip S3stem Tnterlocks

The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manual block of the Source Range trip (i.e., preventspremature block of Source Range trip) and deenergizes the high voltage to the detectors.On decreasing power, Source Range Level trips are automatically reactivated and highvoltage restored.

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more thanone reactor coolant loop, more than one reactor coolant pump breaker open, reactorcoolant pump bus undervoltage and underfrequency, Turbine-tri pressurizer lowpressure and pressurizer high level. On decreasing power, the above listed trips areautomatically blocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or morereactor coolant loops, and one or more reactor coolant pump breakers ope Ondecreasing power, the P-8 interlock automatically blocks the trip on low flow f onecoolant loo ,one coolant pump breaker open. <

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip andthe Low Setpoint Power Range trip; and automatically blocks the Source Range trip anddeenergizes the Source Range high voltage power. On decreasing power, theIntermediate Range trip and the Low Setpoint Power Range trip are automaticallyreactivated. P-10 also provides input to P-7. The trip setpoint on increasing power shallbe > 10% and the reset point shall be less than or equal to 10Mo.

WO70PSqmtfmmav

L-2005-006ENCLOSURE5

Page 1 or 8

ENCLOSURE 5

RE-TYPED TECHNICAL SPECIFICATIONS PAGES

AND

(FOR INFORMATION ONLY) BASES PAGES

T-2005-006ENCLOSURES5

Page 2 of 8

-IC

m-D0z-i

Cz=1Co

90

FUNCTIONAL UNIT

12. Steam/Feedwater Flow ACoincident with

Steam Generator Water



ALLOWABLEVALUE

Aismatch Feed Flow < 23.9% belowrated Steam Flow

Level-Low+ 2 8.15% of narrow rangeinstrument span

Busses A and B 2 69% bus voltage

f Reactor Coolant 2 55.9 Hz

13. Undervoltage - 4.16 kV I

14. Underfrequencv-TripolPump Breaker(s) Open

15. Turbine Trip

a. Auto Stop Oil Pressure

b. Turbine Stop Valve Closure

16. Safety Injection Input from ESF

TRIP SETPOINT

Feed Flow < 20% below ratedSteam Flow

2 10% of narrow range Iinstrument span

2 70% bus voltage

Ž 56.1 Hz

2 45 psig

Fully Closed***

N.A.

Nominal 1 X 10 -10 amps

2 42 psig

Fully Closed***

N.A.

26.0X 10-11 amps

> 17. Reactor Trip System InterlocksmZ a. Intermediate Range Neutron Flux, P-6

mz

0CD

^*Limit switch is set when Turbine Stop Valves are fully closed.o + Not applicable to Unit 3.

Only applicable to Unit 4 through Cycle 22 operation.Not applicable to Unit 4 starting with Cycle 23 operation.

L-2005.006ENCLOSURE-S

Page3of 8

--IC

m-U0z

tt7

-ICz

coCaCA,



ALLOWABLEFUNCTIONAL UNIT

b. Low Power Reactor Trips Block, P-7

1) P-10 input

2) Turbine First Stage Pressure

c. Power Range Neutron Flux, P-8

d. Power Range Neutron Flux, P-1 0

VALUE

S 13.0% RTP**

• 13.0% Turbine Power

S 48.0%+ RTP**

• 43.0%++ RTP''

2 7.0% RTP**

TRIP SETPOINT

Nominal 10% of RTP**

Nominal 10% Turbine Power

Nominal 45%+ of RTP**

Nominal 40%++ of RTP**

Nominal 10% of RTP**10

mz0

mz--4z0C,,

z0

18. Reactor Coolant Pump Breaker PositionTrip

19. Reactor Trip Breakers

20. Automatic Trip and Interlock Logic

'' RTP = RATED THERMAL POWER

+ Only applicable to Unit 4 through Cycle 22 operation.

++ Applicable to Unit 3.Applicable to Unit 4 starting with Cycle 23 operation.

N.A.

N.A. N.A.

N.A.

N.A.

N.A.

L-2005006ENCLOSURES5

Page 4 of 8

-4M

m

0z

-4CzcoU)CIO



co

FUNCTIONAL UNIT

11. Steam Generator WaterLevel--Low-Low

12. Steam Generator Water Level--Low Coincident With Steam/Feedwater Flow Mismatch+

13. Undervoltage--4.16 KV BussesA and B (Above P-7)

14. Underfrequency-Trip of ReactorCoolant Pump Breaker(s) Open(Above P-7)

15. Turbine Trip (Above P-8)a. Autostop Oil Pressureb. Turbine Stop Valve Closure

TOTAL NO.OF CHANNELS

3/stm. gen.

2 stm. gen.level and2 stm./feed-water flowmismatch ineach stm. gen.

2/bus

2/bus

32

CHANNELSTO TRIP

2/stm. gen.

1 stm. gen.level coin-cident with1 stm./feed-water flowmismatch insame stm.gen.

1/bus onboth busses

1 to tripRCPs***

22

MINIMUMCHANNELSOPERABLE

2/stm. gen.

1 stm. gen.level and2 stm./feed-water flowmismatch insame stm.gen.or 2 stm. gen.level and 1stm./feedwaterflow mismatchin same stm.gen.

2/bus

2/bus

APPLICABLEMODES

1,2

1,2

I

ACTION

6

6

mz0

mz-4z0C,,

z0

1

1

12

11

I22

11

1212


L-2005.006ENCLOSURE S

Page 5 of 8Cz TABLE 4.3-1


0 TRIPZ ANALOG ACTUATING MODES FOR

CHANNEL DEVICE WHICHC CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCEz1 FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIREDcn

co 12. Steam Generator Water S R a N.A. N.A. 1,200 Level--Low Coincident with

Steam/Feedwater FlowMismatch+

13. Undervoltage - 4.16 kV N.A. R N.A. N.A. N.A. 1Busses A and B

14. Underfrequency- Trip of N.A. R N.A. N.A. N.A. 1Reactor Coolant PumpBreakers(s) Open

' 15. Turbine Trip° a. Autostop Oil Pressure N.A. R N.A. S/U(1, 10) N.A. 1

b. Turbine Stop ValveClosure N.A. R N.A. S/U(1, 10) N.A. 1

16. Safety Injection Inputfrom ESF N.A. N.A. N.A. R N.A. 1,2

17. Reactor Trip System:> InterlocksK a. Intermediate RangemZ Neutron Flux, P-6 N.A. R(4) R N.A. N.A. 2**0

m b. Low Power ReactorH Trips Block, P-7 N.A. R(4) R N.A. N.A. 1Z (includes P-10 inputEn and Turbine First

Stage Pressure)

z0Z c. Power Range Neutron

Flux, P-8 N.A. R(4) R N.A. N.A. 1



Page 6 of 8

Priccdurc No.: Pocedurc Titc: PaCe:

17Approv-Al DatI.

O-ADM-536 Technical Specification Bases Control Program D : xxx

AITACHIMIENT I(Page 7 of 103)



Reactor Coolant Flow

The Reactor Coolant Flow-Low trip provides core protection to prevent DNB by mitigating theconsequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWERor a turbine first stage pressure at approximately 10% of full power equivalent), an automatic Reactortrip will occur if the flow in more than one loop drops below 90% of loop design flow. Above P-8, anautomatic Reactor trip will occur if the flow in any single loop drops below 90%'of nominal full loopflow. Conversely, on decreasing power between P-8 and the P-7 an automatic Reactor trip will occuron low reactor coolant flow in more than one loop and below P-7 the trip function is automaticallyblocked.

Steam Generator Water Lcvel

The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the eventof a sustained steam/fecdwater flow mismatch resulting from loss of normal fccdwatcr. The specifiedsctpoint provides allowances for starting delays of the Auxiliary Feedwater System.

Steam/Feedwater Flow Mismatch and Low Steam Generator Water lTcvefl4

The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator XVatcr Levcl-Low trip isnot used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functionalcapability of the specified trip settings and thereby enhance the overall reliability of the Reactor TripSystem. This trip is redundant to the Steam Generator Water Level Low-Low trip. TheSteam/Fecdwatcr Flow Mismatch portion of this trip is activated when the steam flow exceeds thefeedwater flow by greater than or cqual to 0.665 x 1O6 Ibs/hour. The Steam Generator Water Lcvel-Lowportion of the trip is activated when the water level drops below 10%, as indicated by the narrow rangeinstrument. These trip values include sufficient allowance in excess of normal operating values topreclude spurious trips but will initiate a Reactor trip before the steam generators arc dry. Thercf6re, therequired capacity and starting time requirements of the auxiliary feedwater pumps arc reduced and theresulting thermal transient on the Reactor Coolant System and steam generators is minimized.

+ A power level of approximately 40% of RATED THERMAL POWER- applicable to Unit 3- applicable to Unit 4 starting with Cycle 23 operation

A power level ofapproxinately 45% of RATED THERMAL POWER- applicable to Unit 4 through Cyclc 22 opcration

+F Not applicable to Unit 3.Only applicable to Unit 4 through Cycle 22 operation.Not applicable to Unit 4 starting with Cycle 23 operation.

woY nPQY.s?-Yat


Page 7 of 8

Procedure No.: Procedure Title: 'age:

18n Approval Date:

0-A1)11-536 Technical Specification Bases Control Program XXXXX




Undervoltagc - 4.16 kV Bus A and B Trips

The 4.16 kV Bus A and B Undervoltage trips provide core protection against DNB as a result ofcomplete loss of forced coolant flow. Thc spccificd setpoint assures a Reactor trip signal'is generatedbefore the Low Flow Trip Setpoint is reached. Time delays arc incorporated in the Undervoltage trips toprevent spurious Reactor trips from momentary electrical power transients. The delay is set so that thetime required for a signal to reach the Reactor trip breakers following the trip of at least oneundervoltage relay in both of the associated Units 4.16 kV busses shall not cxcced 1.3 seconds. Ondecreasing power the Undervoltage Bus trips are automatically blocked by P-7 (a'power level ofapproximately 10% of RATED THERMAL POWER with a turbine first stage pressure at approximately10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trip

A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor Trip from the Turbine trip isautomatically blocked by P-8+; and on increasing power, reinstated automatically by P-8.

Safety Iniection Input from ESF

If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESFautomatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a SafetyInjection. The ESF instrumentation channels which initiate a Safety Injection signal are shown inTable 3.3-3.

Reactor Coolant Pump Breaker Position Trip

The Reactor Coolant Pump Breaker Position Trips are anticipatory trips'which provide reactor coreprotection against DNB. The opcn/close position trips assure a reactor trip signal is generated before thelow flow trip setpoint is reached. Thcir functional capability at the openiclose position settings isrequired to enhance the overall reliability of the Reactor Protection System. Above P-7 (a power levelof approximately 10% of RATED THERMAL POWER or a turbine first stage pressure atapproximately 10% of full power equivalent) an automatic reactor trip will occur if more than onereactor coolant pump breaker is opened. Above P-8+, an automatic reactor trip will occur if one reactorcoolant pump breaker is opened. On decreasing power between P-8 and P-7, an automatic reactor tripwill occur if more than one reactor coolant pump breaker is opened and below P-7 the trip function isautomatically blocked.

+ A powver level of approximately 40%M1 of RATED THERMAL POWER-- applicable to Unit 3- applicable to Unit 4 starting with Cycle 23 operation

A power lcvel of approximately 45% of RATED THER1MSAL POWER- applicable to Unit 4 through Cycle 22 operation

vv�, uro,,ru-,',,vuy


Page 8 of 8

Procedure No.: Proeedare Title: Pae:

1 19. . Approval Date:

O-ADIN1-536 Technical Specification Bases Control Program XXXYXp


TECIINICAL SPECIFICATION BASES


Reactor Coolant Pump Breaker Position Trip (Continucd)

Underfrcquency sensors are also installed on the 4.16 kV busses to detect undcrfrequency and initiatebreaker trip on underfrequency. The underfrequency trip sctpoints preserve the coast down energy ofthe reactor coolant pumps, in case of a grid frequency decrease so DND does not occur.

Reactor Trip System Intcrlocks

The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manual block of the Source Range trip (i.e., preventspremature block of Source Range trip) and deenergizes the high voltage to the dctcctors.On decreasing power, Source Range Level trips are automatically reactivated and highvoltage restored.

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more thanone reactor coolant loop, more than one reactor coolant pump breaker open, reactorcoolant pump bus undervoltage and underfrequency, pressurizer low pressure andpressurizer high Icvel. On decreasing power, the above listed trips are automaticallyblocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or morereactor coolant loops, and one or more reactor coolant pump breakers open, and turbinetrip. On decreasing power, the P-8 interlock automatically blocks the trip on low flow inone coolant loop, one coolant pump breaker open, and turbine trip.

P-10 On increasing power, P-10 allows the manual block of the Intcrmediate Range trip andthe Low Setpoint Power Range trip; and automatically blocks the Source Range trip anddeenergizes the Source Rangc high voltage power. On decreasing power, theIntermediate Range trip and the-Low Selpoint Power Range trip are automaticallyreactivated. P-10 also provides input to P-7. The trip sctpoint on increasing power shallbe Ž 10% and the reset point shall be less than or equal to 10%.

wQY-nPVM.e-1-


Page 1 of 9

ENCLOSURE 6

CAW-05-1958, Application For Withholding ProprietaryInformation From Public Disclosure

Westinghouse Westinghouse Electric CompanyNuclear ServicesP.O. Box 355Pittsburgh, Pennsylvania 15230-0355USA

U.S. Nuclear Regulatory CommissionDocument Control DeskWashington, DC 20555-0001

Direct tel:Directfax:

e-mail:

(412) 3744643(412) 3744011greshajaewestinghouse.com

Our ref: CAW-05-1958

February 22, 2005

APPLICATION FOR WITHHOLDING PROPRIETARYINFORMATION FROM PUBLIC DISCLOSURE

Subject: "Florida Power & Light Turkey Point Units 3 & 4, Licensing Input forDeletion of Steam / Feedwater Flow Mismatch Reactor Trip, WNA-LI-00038-FPL-P,Revision 1. February 2005" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report isfurther identified in Affidavit CAW-05-1958 signed by the owner of the proprietary information,Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basison which the information may be withheld from public disclosure by the Commission and addresses withspecificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission'sregulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Florida Power & Light.

Correspondence with respect to the proprietary aspects of the application for withholding or theWestinghouse affidavit should reference this letter, CAW-05-1958, and should be addressed toJ. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric CompanyLLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

Very truly yIrs,

6 JA. Gresham, ManagerRegulatory Compliance and Plant Licensing

Enclosures

cc: B. BenneyL. Feizollahi

A BNFL Group company

e-11, I

CAW-05-1958

bcc: J. A. Gresham (ECE 4-7A) ILR. Bastien, IL (Nivelles, Belgium)C. Brinkman, I L (Westinghouse Electric Co., 12300 Twinbrook Parkway, Suite 330, Rockville, MD 20852)RCPL Administrative Aide (ECE 4-7A) IL (letter and affidavit only)


CAW-05-1958

AFFIDAVIT

COMMONWEALTH OF PENNSYLVANIA:

ss

COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly

sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of

Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this

Affidavit are true and correct to the best of his knowledge, information, and belief:

J. A. Gresham, Manager

Regulatory Compliance and Plant Licensing

Sworn to and subscribed

before me this c~jday

of c 2005

,w< ,,Notary Public

Notarlai SealSharon L Rod, Notary Publc

Monroeville Boro, Allegheny CountyMy Commnissbon Expires January 29.2007

Member, Pennsylvania Association Of Notares

2 CAW-05-1 958

(1) I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse

Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the

function of reviewing the proprietary information sought to be withheld from public disclosure in

connection with nuclear power plant licensing and rule making proceedings, and am authorized to

apply for its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the

Commission's regulations and in conjunction with the Westinghouse "Application for

Withholding" accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating

information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations,

the following is furnished for consideration by the Commission in determining whether the

information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held

in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not

customarily disclosed to the public. Westinghouse has a rational basis for determining

the types of information customarily held in confidence by it and, in that connection,

utilizes a system to determine when and whether to hold certain types of information in

confidence. The application of that system and the substance of that system constitutes

Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several

types, the release of which might result in the loss of an existing or potential competitive

advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component,

structure, tool, method, etc.) where prevention of its use by any of

3 CAW-05-1958

Westinghouse's competitors without license from Westinghouse constitutes a

competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or

component, structure, tool, method, etc.), the application of which data secures a

competitive economic advantage, e.g., by optimization or improved

marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his

competitive position in the design, manufacture, shipment, installation, assurance

of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or

commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded

development plans and programs of potential commercial value to Westinghouse.

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

There are sound policy reasons behind the Westinghouse system which include the

following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive

advantage over its competitors. It is, therefore, withheld from disclosure to

protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such

information is available to competitors diminishes the Westinghouse ability to

sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by

reducing his expenditure of resources at our expense.

4 CAW-05-1958

(d) Each component of proprietary information pertinent to a particular competitive

advantage is potentially as valuable as the total competitive advantage. If

competitors acquire components of proprietary information, any one component

may be the key to the entire puzzle, thereby depriving Westinghouse of a

competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of

Westinghouse in the world market, and thereby give a market advantage to the

competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and

development depends upon the success in obtaining and maintaining a


(iii) The information is being transmitted to the Commission in confidence and, under the

provisions of 10 CFR Section 2.390, it is to be received in confidence by the

Commission.

(iv) The information sought to be protected is not available in public sources or available

information has not been previously employed in the same original manner or method to

the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is

appropriately marked in "Florida Power & Light Turkey Point Units 3 & 4, Licensing

Input for Deletion of Steam / Feedwater Flow Mismatch Reactor Trip, WNA-LI-00038-

FPL-P, Revision 1, February 2005," (Proprietary) being transmitted by the Florida Power

& Light Company letter and Application for Withholding Proprietary Information from

Public Disclosure, to the Document Control Desk. The proprietary information for

Turkey Point Units 3 & 4 is expected to be applicable for other licensee submittals in

response to certain NRC requirements for justification of steam/feedwater flow mismatch

reactor trip elimination.

This information is part of that which will enable Westinghouse to:

5 CAW-05-1958

(a) Provide an approved, fault tolerant design.

(b) Provide a design configuration that has been certified by an approved process.

(c) Provide basis information for related accident analyses.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for

purposes of eliminating the steam/feed flow mismatch reactor trip.

(b) Westinghouse can sell support and defense of steam/feed flow mismatch reactor

trip elimination.

(c) The information requested to be withheld reveals the distinguishing aspects of a

methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the

competitive position of Westinghouse because it would enhance the ability of

competitors to provide similar design modifications, bases, and licensing defense services

for commercial power reactors without commensurate expenses. Also, public disclosure

of the information would enable others to use the information to meet NRC requirements

for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of

applying the results of many years of experience in an intensive Westinghouse effort and

the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical

programs would have to be performed and a significant manpower effort, having the

requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

* PROPRIETARY INFORMATION NOTICE

Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRCin connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning theprotection of proprietary information so submitted to the NRC, the information which is proprietary in theproprietary versions is contained within brackets, and where the proprietary information has been deletedin the non-proprietary versions, only the brackets remain (the information that was contained within thebrackets in the proprietary versions having been deleted). The justification for claiming the informationso designated as proprietary is indicated in both versions by means of lower case letters (a) through (f)located as a superscript immediately following the brackets enclosing each item of information beingidentified as proprietary or in the margin opposite such information. These lower case letters refer to thetypes of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a)through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE

The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted tomake the number of copies of the information contained in these reports which are necessary for itsinternal use in connection with generic and plant-specific reviews and approvals as well as the issuance,denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license,permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on publicdisclosure to the extent such information has been identified as proprietary by Westinghouse, copyrightprotection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC ispermitted to make the number of copies beyond those necessary for its initernal use which are necessary inorder to have one copy available for public viewing in the appropriate docket files in the public documentroom in Washington, DC and in local public document rooms as may be required by NRC regulations ifthe number of copies submitted is insufficient for this purpose. Copies made by the NRC must includethe copyright notice in all instances and the proprietary notice if the original was identified as proprietary.


Page I of 9

ENCLOSURE 7

CAW-05-1959, Application For Withholding ProprietaryInformation From Public Disclosure

W estinghouse Westinghouse Electric Company. . Nuclear Services

P.O. Box 355Pittsburgh, Pennsylvania 15230-0355USA

U.S. Nuclear Regulatory Commission Direct tel: (412) 3744643Document Control Desk Directfax: (412) 374-4011Washington, DC 20555-0001 e-mail: greshajaewestinghouse.com

Our ref: CAW-05-1959

February 22, 2005

APPLICATION FOR WITHHOLDING PROPRIETARYINFORMATION FROM PUBLIC DISCLOSURE

Subject: "Florida Power & Light Turkey Point Units 3 & 4, Licensing Input forRPS Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8,WNA-LI-00039-FPL-P, Revision 1, February 2005" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report isfurther identified in Affidavit CAW-05-1959 signed by the owner of the proprietary information,Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basison which the information may be withheld from public disclosure by the Commission and addresses withspecificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission'sregulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Florida Power & Light.

Correspondence with respect to the proprietary aspects of the application for withholding or theWestinghouse affidavit should reference this letter, CAW-05-1959, and should be addressed toJ. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric CompanyLLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

Very truly yours,

41. A. Gresham, ManagerA/ Regulatory Compliance and Plant Licensing

Enclosures

cc: B. BenneyL. Feizollahi


CAW-05-1959

bcc: J. A. Gresham (ECE 4-7A) ILR. Bastien, IL (Nivelles, Belgium)C. Brinkman, IL (Westinghouse Electric Co., 12300 Twinbrook Parkway, Suite 330, Rockville, MD 20852)RCPL Administrative Aide (ECE 4-7A) IL (letter and affidavit only)


CAW-05-1 959

AFFIDAVIT

COMMONWEALTH OF PENNSYLVANIA:

ss

COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me dulysworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf ofWestinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in thisAffidavit are true and correct to the best of his knowledge, information, and belief:

J. A. resham, Manager

Regulatory Compliance and Plant Licensing

Sworn to and subscribed

before me this K ay

of 2005

Notary Public

Notarial Sea]Sharon L Rod, Notary Pubric

Monroeville Boro, Adlegheny CountyMy Commission Expires January 29,2007

Member. Pennsyrvaria Association Of Notares

2 CAW-05-1959

(1) I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse

Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the

function of reviewing the proprietary information sought to be withheld from public disclosure in

connection with nuclear power plant licensing and rule making proceedings, and am authorized to

apply for its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the

Commission's regulations and in conjunction with the Westinghouse "Application for

Withholding" accompanying this Affidavit.

(3) 1 have personal knowledge of the criteria and procedures utilized by Westinghouse in designating

information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations,

the following is furnished for consideration by the Commission in determining whether the

information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held

in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not

customarily disclosed to the public. Westinghouse has a rational basis for determining

the types of information customarily held in confidence by it and, in that connection,

utilizes a system to determine when and whether to hold certain types of information in

confidence. The application of that system and the substance of that system constitutes

Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several

types, the release of which might result in the loss of an existing or potential competitive

advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component,

structure, tool, method, etc.) where prevention of its use by any of

3 CAW-05-1959

Westinghouse's competitors without license from Westinghouse constitutes a

competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or

component, structure, tool, method, etc.), the application of which data secures a

competitive economic advantage, e.g., by optimization or improved

marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his

competitive position in the design, manufacture, shipment, installation, assurance

of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or

commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded

development plans and programs of potential commercial value to Westinghouse.

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

There are sound policy reasons behind the Westinghouse system which include the

following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive

advantage over its competitors. It is, therefore, withheld from disclosure to

protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such

information is available to competitors diminishes the Westinghouse ability to

sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by

reducing his expenditure of resources at our expense.

4 CAW-05-1959

(d) Each component of proprietary information pertinent to a particular competitive

advantage is potentially as valuable as the total competitive advantage. If

competitors acquire components of proprietary information, any one component

may be the key to the entire puzzle, thereby depriving Westinghouse of a


(e) Unrestricted disclosure would jeopardize the position of prominence of

Westinghouse in the world market, and thereby give a market advantage to the

competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and

development depends upon the success in obtaining and maintaining a


(iii) The information is being transmitted to the Commission in confidence and, under the

provisions of 10 CFR Section 2.390, it is to be received in confidence by the

Commission.

(iv) The information sought to be protected is not available in public sources or available

information has not been previously employed in the same original manner or method to

the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is

appropriately marked in "Florida Power & Light Turkey Point Units 3 & 4, Licensing

Input for RPS Modifications Changing Reactor Trip on Turbine Trip Permissive from

P-7 to P-8, WNA-LI-00039-FPL-P, Revision 1, February 2005," (Proprietary) being

transmitted by the Florida Power & Light Company letter and Application for

Withholding Proprietary Information from Public Disclosure, to the Document Control

Desk. The proprietary information for Turkey Point Units 3 & 4 is expected to be

applicable for other licensee submittals in response to certain NRC requirements for

justification of changing reactor trip on turbine trip permissive from P-7 to P-8.

This information is part of that which will enable Westinghouse to:

5 CAW-OS-1959

(a) Perform analysis in support of the noted modification.

(b) Provide basis information for related accident analyses and reactor protection system

functions.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for

purposes of changing reactor trip on turbine trip permissive from P-7 to P-8.

(b) Westinghouse can sell support and defense of changing reactor trip on turbine

trip permissive from P-7 to P-8.

(c) The information requested to be withheld reveals the distinguishing aspects of a

methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the

competitive position of Westinghouse because it would enhance the ability of

competitors to provide similar design modifications, analyses, bases, and licensing

defense services for commercial power reactors without commensurate expenses. Also,

public disclosure of the information would enable others to use the information to meet

NRC requirements for licensing documentation without purchasing the right to use the

information.

The development of the technology described in part by the information is the result of

applying the results of many years of experience in an intensive Westinghouse effort and

the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical

programs would have to be performed and a significant manpower effort, having the

requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE

Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRCin connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning theprotection of proprietary information so submitted to the NRC, the information which is proprietary in theproprietary versions is contained within brackets, and where the proprietary information has been deletedin the non-proprietary versions, only the brackets remain (the information that was contained within thebrackets in the proprietary versions having been deleted). The justification for claiming the informationso designated as proprietary is indicated in both versions by means of lower case letters (a) through (f)located as a superscript immediately following the brackets enclosing each item of information beingidentified as proprietary or in the margin opposite such information. These lower case letters refer to thetypes of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a)through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE

The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted tomake the number of copies of the information contained in these reports which are necessary for itsinternal use in connection with generic and plant-specific reviews and approvals as well as the issuance,denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license,permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on publicdisclosure to the extent such information has been identified as proprietary by Westinghouse, copyrightprotection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC ispermitted to make the number of copies beyond those necessary for its internal use which are necessary inorder to have one copy available for public viewing in the appropriate docket files in the public documentroom in Washington, DC and in local public document rooms as may be required by NRC regulations ifthe number of copies submitted is insufficient for this purpose. Copies made by the NRC must includethe copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Turkey Point Units 3 & 4, WNA-LI-00039-FPL-NP, Rev 1 ...

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