ML050910250

WNA-LI-00039-FPL-NP, Rev 1, Licensing Input for RPS Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8, Enclosure 1D

ML050910250
Person / Time
Site:	Turkey Point
Issue date:	02/21/2005
From:	Jurczak J, Suggs C, Vaughn Thomas Westinghouse
To:	Office of Nuclear Reactor Regulation
References
WNA-LI-00039-FPL-NP, Rev 1
Download: ML050910250 (77)

Text

L-2005-006 ENCLOSURE 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 & 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-N P Revision I February 2005

Westinghouse Non-Proprietary Class 3 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-NP Revision 1 February 2005 APPROVALS Function Name and Signature Date Atos C. W. Suggs 4-Authors Principal Engineer, Des nBasiEeerinl 2/2112005 Reviewed V.M.Thomas A 2/21/2005 Pricipal Engineer. Desig t'asis Engineering ,

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

Westinghouse Electric Company, LLC P.O. Box 355 Pittsburgh, PA 15230-0355 i 2005 Westinghouse Electric Company LLC All Rights Reserved

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive 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 2/21/2005 redundant word correction.

DOCUMENT TRACEABILITY & COMPLIANCE Created to Support the Following Document(s) Document Number Revision None DETAILED RECORD OF CHANGES Revision Date

==

Description:==

Rev 0 Initial issue. 1/2005 Rev 1 :Clarifed wording/references (page 2-2), corrected redundant 2/2005 word (page 4-7)

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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 TABLE OF CONTENTS Section Title Page LIST OF CONTRIBUTORS/REVIEWERS ....................................... i TABLE OF CONTENTS ...................................... ii GLOSSARY OF TERMS ...................................... iv REFERENCES ........................................ v SECTION 1 PURPOSE AND SCOPE OF DOCUMENT ...................................... 1-1 SECTION 2 CHANGE DESCRIPTION ...................................... 2-1 SECTION 3 BEST ESTIMATE ANALYSIS ...................................... 3-1

3.1 INTRODUCTION

...................................... 3-1 3.2 INPUT PARAMETERS AND ASSUMPTIONS ...................................... 3-1

3.3 DESCRIPTION

OF ANALYSIS AND EVALUATION ...................................... 3-4 3.4 ACCEPTANCE CRITERIA AND RESULTS ...................................... 3-4

3.5 CONCLUSION

S ...................................... 3-5 SECTION 4 DETAILED EVALUATION OF NON-LOCA EVENTS ...................................... 4-1 SECTION 5

SUMMARY

OF EVALUATIONS ...................................... 5-1 SECTION 6 CONCLUSION ...................................... 6-1 INDEX OF TABLES Table Title Page Table 1 Turbine Trip without Reactor Trip from P-8 Setpoint of 40% Power 3-4 WNA-LI-00039-FPL-NP, Rev. I ii

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 ACRONYMS AND TRADEMARKS The following abbreviations and acronyms are defined to allow an understanding of their use within this document.

Acronyms Definition BOL Beginning-of-Life DNB Departure from Nucleate Boiling FPL Florida Power & Light FWCS Feedwater Control System l&C Instrumentation and Control LAR License Amendment Request NSSS Nuclear Steam Supply System PTN Plant Turkey Point PORV Power-Operated Relief Valve RCS Reactor Coolant System RPS Reactor Protection System RTP Rated Thermal Power SG Steam Generator UFSAR Updated Final Safety Analysis Report All other product and corporate names used in this document may be trademarks or registered trademarks 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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 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, Proposed Anticipatory Trip Modification, October 1980.

4. PCWG-2779, 'Turkey Point Units 3 & 4 (FPUFLA): Approval of Category IV PCWG Parameters to Support Reduced Feedwater Temperature in Conjunction with Uprate Program," 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 Power Uprate 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 to Condenser Controls."

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

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

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

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

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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8

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

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

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

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

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

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

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

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

20. FPL, Document 5614-M-313, Revision 38, 'Turkey Point Nuclear Unit 4 Instrument Setpoint List," 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, Revision 14, "Logic Diagram Units 3 & 4, Nuclear Instrumentation Permissives and Blocks."

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

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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8

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 and 221.

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 Protection Systems, Turkey Point Units 3 and 4, March 1990 (Last Page of Front Matter)

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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 SECTION 1 PURPOSE 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 to permissive 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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 SECTION 2 CHANGE DESCRIPTION The current design at Turkey Point Units 3 and 4 for the reactor trip on turbine trip automatically blocks the function when power levels are below the P-7 setpoint (10% power). The Turkey Point plants are designed with 50% load rejection capability. With the load rejection capacity at the Turkey Point Units, load rejections of up to 50% should not require a reactor trip if all control systems function as designed. Therefore, it is possible to increase the setpoint for the reactor trip on turbine trip interlock up to 50% power. This would result in blocking the direct reactor trip on turbine trip for load rejection events up to 50% power levels. By implementing the block of reactor trip on turbine trip at a permissive with a higher setpoint, there is a decrease in potentially unnecessary challenges to the reactor protection system and an increase in plant availability. For the Turkey Point Units 3 and 4, the P-8 permissive has been selected, but with the setpoint reduced from 45% to 40% power, for conservatism.

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

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

To satisfy the NRC requirements in Item II.K.3.10, a best estimate plant specific analysis was performed to show that the implementation of the block of reactor trip on turbine trip at the P-8 WNA-LI-00039-FPL-NP, Rev. I 2-1

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 setpoint will not result in challenges to the pressurizer PORVs. The best estimate analysis was performed using the proposed P-8 setpoint of 40% power. The results show that for this setpoint value, the pressurizer PORVs will not be challenged.

The proposed P-8 setpoint change affects the Reactor Coolant Flow - Low Reactor Trip logic (Technical Specifications Table 3.3-1, Function 10) and the Reactor Coolant Pump Breaker Position Reactor Trip logic (Technical Specifications Table 3.3-1, Function 18). For these functions the single loop and single breaker trips are enabled above P-8 and the two loop and two breaker trips are enabled between P-7 and P-8. Changing the P-8 setpoint from 45% to 40% is in the conservative direction and is allowable according to the current Trip Setpoint Allowable Value (Technical Specifications Table 2.2-1, Function 17.c). Currently, between 40% and 45% power, low flow in two loops or two RCP breakers open is required for a reactor trip. Following the P-8 setpoint change, between 40% and 45% power, low flow in only one loop or only one RCP breaker open will initiate a reactor trip. For events analyzed in UFSAR Chapter 14 which credit the low flow or RCP breaker trips, the analyses and results are bounding since the P-8 setpoint change is in the conservative direction. The current 3% difference between the Trip Setpoint and the Allowable Value specified in Technical Specification Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints is based on the allowable value of the associated protection function as documented in the current setpoint analysis (Reference 31). The basis for the 3% allowance is unaffected by this change.

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

(Last Page of Section 2)

WNA-LI-00039-FPL-NP, Rev. I 2-2

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 SECTION 3 BEST ESTIMATE ANALYSIS

3.1 INTRODUCTION

Westinghouse has performed an evaluation for the Reactor Protection System (RPS) modifications to change the reactor trip on turbine trip interlock from permissive P-7 to permissive P-8 at Turkey Point Units 3 and 4 (Reference 30). This evaluation has been performed to determine the impact of a 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 of increasing the interlock setpoint for the turbine trip without a reactor trip because of the potential to increase the probability of a stuck open pressurizer PORV. The NRC position was addressed in NUREG- 0737, Item lI.K.3.10. In order to reduce the likelihood of opening the pressurizer PORV following 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 to determine if the implementation of the P-8 permissive (40%) will result in challenges to the pressurizer PORVs.

3.2 INPUT PARAMETERS AND ASSUMPTIONS A best estimate analytical study was performed to determine the transient plant response to a turbine trip without reactor trip transient. The analysis was performed using the LOFTRAN computer code (Reference 5) model of Turkey Point Units 3 and 4. The LOFTRAN computer code was 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 well as the various control and protection systems. Since the object of this study was primarily to determine the peak in pressurizer pressure following the initiation of the transient, assumptions were made that would contribute to a conservatively high prediction of pressurizer pressure.

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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 These assumptions follow: a,b,c a,b,c

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

• Minimum SGTP level was chosen since it results in the highest pressurizer insurge for a fixed full power Tavg.

• Maximum feedwater temperature was chosen since it results in the highest steam/feedwater flow and therefore, the smallest steam dump capacity (in fraction of rated steam flow).

3. Best estimate Beginning-of-Life (BOL) reactivity parameters were used. BOL reactivity parameters have lower differential rod worth and less negative moderator temperature coefficient 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., best estimate analysis) and are at their [ ]b.c power value.

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

a,b,c WNA-LI-00039-FPL-NP, Rev. I 3-2

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P8 a,u,c

7. The Pressurizer Pressure control system, Steam Dump control system (to condenser, load rejection mode) and Steam Generator (SG) Level control system were assumed operational and in the automatic mode of control. Steam dump to the atmospheric relief valves is not credited in the analysis.

a,b,c

10. Since the analyses are best estimate analyses, the parameter values including the control systems setpoints and PORV setpoints were assumed at nominal conditions without uncertainties and/or tolerances.

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

were taken from the following references:

Steam Dump Control Settings: References 7 & 8 Pressurizer Pressure Control Settings: 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 WNA-LI-00039-FPL-NP, Rev. I 3-3

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8

3.3 DESCRIPTION

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

]bc with all normal control systems assumed operational. This best estimate 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 and cubic 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 challenging the PORVs or the SG safety valves. The results of this best estimate analysis are shown in Table 1.

a,b,c Table I Turbine Trip without Reactor Trip from P-8 Setpoint of 40% Power WNA-LI-00039-FPL-NP, Rev. I 3-4

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8

3.5 CONCLUSION

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

(Last Page of Section 3)

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

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 SECTION 4 DETAILED EVALUATION OF NON-LOCA EVENTS Each of the non-LOCA accident analyses described in Chapter 14 of the Turkey Point UFSAR was reviewed 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 accident analyses. This review is summarized below for each event.

1. UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL FROM A SUBCRITICAL CONDITION Event Definition: A rod cluster control assembly (RCCA) bank withdrawal accident is defined as an uncontrolled addition of reactivity to the reactor core caused by withdrawal of one or more RCCA banks, resulting in a power excursion. This could occur with the reactor either subcritical, at hot zero power, or at power.

Plant Operating Conditions: The plant is assumed to be operating at the no-load reactor coolant average 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 generator is not on-line. Therefore, the proposed changes have no effect on this accident scenario and the 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 the core 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 Rated Thermal Power are analyzed. For all cases analyzed, the results show that integrity of the core is maintained by the reactor protection system (RPS) as the departure from nucleate boiling (DNBR) remains above the safety analysis limit value.

WNA-LI-00039-FPL-NP, Rev. I 4-1

Florida Power &Light Licensing Input for RPS Turkey Point Units 3 &4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 Effect of Proposed Change: In this scenario, the reactor trip is provided by the automatic actuation of the first primary side reactor protection signal reached: either overtemperature delta temperature (OTAT) or power range high neutron flux. Neither of these reactor protection functions is affected by the turbine trip signal, 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 UFSAR remain valid.

3. ROD CLUSTER CONTROL ASSEMBLY (RCCA) DROP Event Definition: The dropped RCCA accident is initiated by a single electrical or mechanical failure which causes any number and combination of rods from the same group of a given bank to 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 is not 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 or backup trip. Therefore, the proposed changes have no effect on this accident scenario and the conclusions 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 inadvertent opening of the primary water makeup control valve and failure of the blend system, either by controller 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 for power 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 to reach the OTAT trip setpoint resulting in a reactor trip. In Mode 2, the power range high neutron flux (low setpoint) function provides the trip. Neither of these reactor trip functions is affected by the reactor trip on turbine trip setpoint or by the change in the P-8 setpoint. Therefore, the WNA-LI-00039-FPL-NP, Rev. I 4-2

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 proposed changes have no effect on this accident scenario and the conclusions of the UFSAR remain 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 condition results in the sudden introduction of colder water into the core from the idle loop which could cause 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 of the plant with one or more loops out of service. Therefore, this event no longer applies and has been removed from the plant's licensing basis. The proposed changes, therefore, have no effect 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 the steam generators or a decrease in feedwater temperature. This event will result in an increase in the heat transfer rate from primary to secondary in the steam generators and a consequential reduction in primary system temperature and pressure. The transient responses for an excessive feedwater flow event to one steam generator were analyzed for four cases: two cases at hot full power (one case with automatic rod control and one without) and two cases at hot zero power (one case with automatic rod control and one without).

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

Effect of Proposed Change: For full power conditions, feedwater isolation and turbine trip (with subsequent reactor trip signal on turbine trip) occur on the high-high steam generator water level 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-7 setpoint. The proposed change is to move the reactor trip on turbine trip from P-7 to P-8, which results in the trip function being disabled below 40% power (the new P-8 setpoint). As noted WNA-LI-00039-FPL-NP, Rev. 1 4-3

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 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 trip function is still available at 100% power. For the zero power cases, the reactor may be tripped by the power range high neutron flux trip (low setting), although this trip is not credited and the minimum DNBR would not change significantly without it.

For power levels below 40%, where the turbine trip / reactor trip function is disabled, the event would be no more severe than the 100% power case. Although credited in the analysis for the full power cases, the reactor trip on turbine trip is not a critical function that is required in order to get acceptable results. By the time that the high-high SG level trip is reached, the plant reaches 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 the high-high steam generator level trip. The reactor trip on turbine trip is modeled since it is expected to occur. However, it is not considered primary protection for the event and the minimum DNBR calculated would not change significantly without it.

Based on the above, the proposed changes do not invalidate the results of the analysis and the conclusions 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 steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. The reactor control system is designed to accommodate a 10% step-load increase or a 5% per minute ramp load increase in the range of 15 to 100% power. Any loading rate 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 does not occur in this analysis. Therefore, the proposed changes have no effect on this accident scenario and the conclusions of the UFSAR remain valid.

8. LOSS OF REACTOR COOLANT FLOW A) Flow Coastdown Accidents WNA-LI-00039-FPL-NP, Rev. 1 4-4

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 &4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 Event Definition: The loss of flow incident can result from a mechanical or electrical failure in a reactor 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. Bounding analyses are performed at full power since this is the most conservative in terms of potential consequences, specifically a more limiting minimum DNBR.

Effect of Proposed Change: The low primary coolant loop flow, RCP undervoltage, RCP underfrequency, and RCP breaker position reactor trip functions provide the necessary protection 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 the analysis. This permissive defines the highest steady state power level at which the reactor can operate with one RCS loop inactive without violating the N-1 core thermal limits. A reduction in the P-8 setpoint is conservative since the core thermal limits are less likely to be violated at lower power levels. Therefore, the proposed changes have no effect on this accident scenario and 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 an instantaneous seizure of a single RCP rotor which results in a rapid reduction in reactor coolant loop flow.

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

Effect of Proposed Change: The reactor trip is initiated by low primary coolant loop flow. This trip is not affected by the reactor trip on turbine trip setpoint. Therefore, the proposed changes have 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 a complete loss of steam load from full power without a direct reactor trip, or a turbine trip without a direct reactor trip.

WNA-LI-00039-FPL-NP, Rev. I 4-5

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 &4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 Plant Operating Condition: The analysis assumes a complete loss of steam load from full power with 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 pressurizer pressure, or low-low steam generator water level signals. The loss of external electrical load/turbine trip event from a full power condition bounds a turbine trip with no subsequent reactor 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 conclusions of 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 in the 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 from 102% of Rated Thermal Power. Maximum initial RCS temperature and pressure conditions are assumed.

Effect of Proposed Change: The reactor trip is initiated by low - low steam generator water level. This trip is not affected by the changes. Therefore, the proposed changes have no effect on 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 all power to the plant auxiliaries: i.e., the RCPs, condensate pumps, etc. The loss of power may be caused by a complete loss of the offsite grid accompanied by a turbine generator trip at the station, 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 water level 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 the UFSAR remain valid.

WNA-LI-00039-FPL-NP, Rev. I 4-6

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 &4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8

12. RUPTURE OF A STEAM PIPE Event Definition: A rupture of a steam pipe is assumed to include any accident which results in an uncontrolled steam release from a steam generator. Such a release may result from either the 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 shutdown conditions.

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 Safety Injection Signal. The limiting zero power analysis does not specifically credit the reactor trip system. Only the Engineered Safety Features Actuation System (ESFAS) is needed to limit the consequent of the analyzed events. Therefore, the proposed changes have no effect on this accident 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 housing such 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 low setpoint, or the high setpoint. These trip mechanisms are not affected by the proposed changes. Therefore, the proposed changes have no effect on this accident scenario and the conclusions of the UFSAR remain valid.

(Last Page of Section 4)

WNA-LI-00039-FPL-NP, Rev. I 4-7

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 SECTION 5

SUMMARY

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 on turbine 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 analyzed at 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 further boron precipitation The changes do not affect the normal plant operating parameters, the safeguards systems actuation or accident mitigation capabilities important to LOCA, or the assumptions used in the LOCA related accidents. Nor do the changes create conditions more limiting than those assumed in these analyses.

2. NON-LOCA RELATED EVALUATION Each of the non-LOCA accident analyses described in Chapter 14 of the Turkey Point UFSAR was reviewed with respect to changing the reactor trip on turbine trip permissive from P-7 to P-8 and the P-8 setpoint from 45% to 40%. Based on this review, it is concluded that the proposed changes have no effect on the accident analyses. This review is summarized above in Section 4, Evaluation of Non-LOCA Events.

It is concluded that the non-LOCA safety analyses presented in Chapter 14 of the UFSAR are not adversely affected by changing the reactor trip on turbine trip permissive from P-7 to P-8 or by changing the P-8 setpoint from 45% to 40%. Additionally, normal plant operating parameters, accident mitigation capabilities, and assumptions used in the non-LOCA transients WNA-LI-00039-FPL-NP, Rev. I 5-1

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 are not adversely affected. The changes will not create conditions more limiting than those considered in the current non-LOCA analyses. Therefore, changing the reactor trip on turbine trip permissive from P-7 to P-8 and changing the P-8 setpoint from 45% to 40% do not alter the conclusions 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-8 setpoint from 45% to 40% do not affect either the inside or outside containment MSLB mass and energy release, or the calculations for the steam mass release used as input to the radiological dose evaluation. For this event, High Containment Pressure initiates safety injection and the safety injection signal produces a reactor trip signal. The reactor trip on turbine trip function is not credited for this event. The changes do not affect the normal plant operating 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 mass and 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-8 setpoint from 45% to 40% do not affect the SGTR analysis methodology or assumptions. For this event, Low Pressurizer Pressure initiates safety injection and the safety injection signal produces a reactor trip signal. The reactor trip on turbine trip function is not credited for this event. These changes do not alter the current SGTR event analysis results. Thus, the conclusions presented in the UFSAR remain valid with respect to the SGTR event.

(Last Page of Section 5)

WNA-LI-00039-FPL-NP, Rev. 1 5-2

Florida Power & Light Licensing Input for RPS Turkey Point Units 3 & 4 Modifications Changing Reactor Trip on Turbine Trip Permissive from P-7 to P-8 SECTION 6 CONCLUSION 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 10CFR50.92, 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.

(Last Page of Section 6)

WNA-LI-00039-FPL-NP, Rev. I 6-1

L-2005-006 ENCLOSURE 2 Page 1 of 4 ENCLOSURE 2 NO SIGNIFICANT HAZARDS CONSIDERATIONS

L12005-006 ENCLOSURE 2 Page 2 or 4 Introduction The proposed amendments revise the Turkey Point Technical SpecNfications for several Reactor Trip System 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 water level - 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 17c, Reactor Trip System Interlock, Power Range Neutron Flux, P-8, the allowable value is being changed from < 48.0% rated thermal power (RTP) to < 43% RTP, and the 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 water level - 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 water level - 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 no significant hazards consideration are included in the Commission's regulation, 10 CFR 50.92, which states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Each standard is discussed as follows:

(1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes revise the operability requirements, surveillance requirements and the interlock setpoint for two Reactor Trip System functional units. The affected trip functional units are not initiators of any accident previously evaluated. The proposed changes to the affected trip functional units do not adversely affect the initiators of any accident previously evaluated. A best estimate analysis has shown that a turbine trip without a reactor trip below 40% power does not challenge the pressurizer PORVs or the steam generator safety valves;

L-2005-006 ENCLOSURE 2 Page 3 of 4 thereby, not adversely affecting the probability of a small break LOCA due to a stuck open PORV, 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 increased by the proposed changes.

The steam/feedwater flow mismatch coincident with steam generator water level - low reactor trip is not credited as a primary trip in any previously evaluated accidents. The reactor trip on turbine trip below the P-8 interlock is not credited as a primary trip in any previously evaluated accidents. Therefore, the mitigation functions that have been assumed in the accident analyses will continue to be performed by the systems and components currently credited in the analyses; and the accident analysis results are not affected by the changes to the affected trip functional units. The P-8 setpoint is not an initial condition of any accident previously evaluated. Therefore, the accident analysis results are not affected by changes to the P-8 setpoint. No safety analyses previously performed in the Turkey Point Units 3 and 4 UFSAR required reanalysis for these proposed changes. All accident analyses acceptance criteria continue to be met. The proposed changes do not create any new credible limiting single failure. As a result, the consequences of any accident previously evaluated are not significantly increased by the proposed changes.

In conclusion, operation of the facility in accordance with the proposed amendments does not involve a significant increase in the probability or consequences of any accident previously evaluated.

(2) Operation of the facility in accordance with the proposed amendments would not create the 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 causal mechanisms. The proposed changes do not adversely affect previously identified accident initiators and do not create any new accident initiators. No new limiting single failures or accident scenarios are created by the proposed changes. No new challenges to any installed safety system are created by these proposed changes. The proposed changes do not result in any event previously deemed incredible being made credible.

The steam/feedwater flow mismatch coincident with steam generator water level - low reactor 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 different kind of accident from any previously evaluated.

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

L-2005-006 ENCLOSURE 2 Page 4 or 4 the associated trip functional units operate or function, the changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Therefore, operation of the facility in accordance with the proposed amendments does not create 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 not involve 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 criteria are unaffected. The current UFSAR safety analyses remain bounding. No UFSAR Chapter 14 events explicitly credit the steam / feedwater flow mismatch reactor trip function and the reactor trip on turbine trip function below the P-8 setpoint value. The safety systems credited in 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 Reactor Coolant Flow - Low Reactor Trip and the Reactor Coolant Pump Breaker Position Reactor Trip. Therefore, the proposed changes do not result in a significant reduction in a margin of safety; and operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

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

L-2005-006 ENCLOSURE 3 Page 1 of 2 ENCLOSURE 3 ENVIRONMENTAL CONSIDERATION

L-2005-006 ENCLOSURE 3 Page 2 of 2 Environmental Consideration The proposed license amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The proposed amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released off-site, and no significant increase in individual or cumulative occupational radiation exposure. FPL has concluded that the proposed amendments involve no significant hazards consideration, and therefore, meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Hence, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment need not be prepared in connection with issuance of the amendments.

L-2005-006 ENCLOSURE 4 Page I of 17 ENCLOSURE 4 PROPOSED MARK-UP OF AFFECTED TECHNICAL SPECIFICATIONS AND (FOR INFORMATION ONLY) BASES PAGES

L-2005-006 ENCLOSURE4 Page 2 of17 Provided For Information Only - No Change To This Page TABLE 2.2-1 ig fRELACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ALLOWABLE fflj FUNCTIONAL UNIT VALUE TRIP SETPOINT Z 1. Manual Reactor Trip NA. NA.

2. Power Range. Neutron Flux

a. High Setpolnt S 112.0% of RTP** s 109.0% of RTP'*

b. Low Setpolnt s28.0% of RTP' S 25% of RTP*

3. Intermediate Range, Neutron Flux S 31.0% of RTP S 25% of RTP**

4. Source Range. Neutron Flux 5 5 10o cps 51.4 X 10 cps

5. Overtemperature AT See Note 2 See Note 1

6. Overpower AT See Note 4 See Note 3

7. Pressurizer Pressure-Low 1817 psIg a 1835 pslg

8. Pressurizer Pressure-High 5 2403 psrg s 2385 pslg

9. Pressurizer Water Level-Hlgh 5 92.2% of Instrument span 5 92% of Instrument span t10. Reactor Coolant Flow-Low 2 88.8% of loop design flow 290% of loop design fow' K II. Steam Generator Water Level Low-Low Ž 8.15% of narrow range Ž10%of narrow range m Instrument span Instrument span

-a CD Loop design flow = 85,000 gpm RTP = Rated Thermal Power

,-2005-006 ENCLOSURE 4 Page 3 or17 I

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ALLOWABLE FUNCTIONAL UNIT VALUE TRIP SETPOINT C

12. SteamlFeedwater Flow Misrrlatch Feed FlowS 23.9% below Feed Flow S20% below rated Coincident with rated Steam Flow Steam Flow co go Steam GeneratorWate (DI-Low+ 8.15% of narrow range a O1% of narrow range instrument span instrument span

13. Undervoltage-4.18 kV BussSes A and B 2Ž69% bus voltage 2 70% bus voltage

14. Underfreruencv - TrID of Re actor Coolant k 55.9 Hz Ž56.1 Hz Pump Breaker(s) Open

15. Turbine Trip

a. Auto Stop Oil Pressure 2Ž42 psig Ž45 psig

b. Turbine Stop Valve Closure Fully Closed** Fully Closed-*

16. Safety Injection Input from ESF N.A. NA.

> 17. Reactor Trip System Interlocks m

a. Intermediate Range Neutron Flux. P-6 26.0 X IO 'I amps Nominal 1 X 10-'° amps I

0 oi mSwiuh e Ives are fully clsed.

f - / A SE IT TE ce co

L,2005-006 ENCLOSURE 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.

L2005-006 ENCLOSURE 4 Page 5 of 17 TABLE 2.2-1 (Continuecdl REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

• 0 0 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 NA Trip

19. Reactor Trip Breakers N.A.l N.A.

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

m z

z0 0

z CD.

,{SR ¢se o tt RER

,-4

L-2005-006 ENCLOSURE 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,2005-006 ENCLOSURE 4 Page 7 of 17 ProviIded For Information Only - No Change To This Page TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION z MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

-I FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE _MODES_ ACTION C

2

1. Manual Reactor Trip 2 1 2 1,2 Cft Ca) 2 1 2 3-.4',5- 9

2. Power Range. Neutron Flux

a. High Setpolnt 4 2 3 1.2 2

b. Low Setpoint 4 2 3 1##, 2 2

3. Intermediate Range, Neutron Flux 2. 1 2 3

4. Source Range, Neutron Flux I 4

a. Startup 2 2 2#

b. Shutdown*' 2 0 2 3,4,5 5

c. Shutdown 2 1 2 3', 4*, 5' 9

5. Overtemperature AT 3 2 2 1.2 13

6. Overpower AT 3 2 2 1,2 13

7. Pressurizer Pressure-Low 3 2 2 1 6 2 (Above P-7) z 2 2 1,2

8. Pressurizer Pressure-High 3 6 z .3 2 2 I 13 0 9. Pressurizer Water Level-High fn (Abova P-7) I CD

10. Reactor Coolant Flow-Low 0

a. Single Loop (Above P4) 3/loop 21loop ZIioop 1 6

b. Two Loops (Above P-7 3/10p 2Jnoop 211oop I 6 and below P-8)

L-2005-006 ENCLOSURE 4 Page 8 of 17 TABLE 3.3-1 (Continued)l

2 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM A TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION (n

z 11. Steam GeneratorWater 3/stm. gen. 2/stm. gen. 2/stm. gen. 1,2 6 Level- Low-Low 90

12. Steam Generator Water Level- 2 stm. gen. 1 stm. gen. 1stm. gen. 1.2 6 Low Coincident V~frSti level and level coin- level and Feedwater F i smatch J 2 stm.lfeed- cident with 2 stm./feed-water flow I stni.feed- water Row AA mismatch in water flow mismatch in each stm. gen. mismatch In same stm. gen.

SUPERSC" same stm. or 2 stm. gen.

• _,.r gen. level and 1 C.,

stmifeedwater flow mismatch Insame stm.

gen.

13. Undervottage-4.16 KV Busses 2/bus 1bus on 21bus 1 12 z A and B (Above P-7) both busses

14. Underfrequency-Trip of Reactor 2Jbus 1 to trip 2/hus 1 11 z

Coolant Pump Breaker(s) Open RCPs---

z4 (Above P-7) 0

15. Turbine Trip (Above

a. Autostop Oil Pressure 3 2 2 1i 12

b. Turbine Stop Valve Closure 2 2 2 1 12 t-L IA/KST rzTr l E )

fl_(SC*- ne7t ?41e} _

L-2005-006 ENCLOSURE 4 Page 9 of 17 INSERT NEW FOOTNOTE FOR TS PAGE 3/4 3-3:

+ 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-2005-006 ENCLOSURE 4 Page 10 of 17 Provided For Information Only - No Change To This Page .

TABLE 4.3-'

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK- CALIBRATION TEST TEST LOGIC TEST IS REQUIRED C4

1. Manual Reactor Trip NA N.A. N.A. R(11) N.A. . 23*.4* 5*

C.)

2. Power Range. Neutron Flux

a. HIgh Setpoint S D(2. 4), Q N.A. NA 1.2 M(3. 4), I 0(4.B).

R(4)

b. Low Setpodnt S R(4j SAJ(1) NA NA 1 '. 2 (a

3. IntermedIlate Range, S R(4) SAU(1) NA. N.A. 111*, 2 Neutron Flux

4. Source Range, Neutron Flux S R(4) SUM1), 0(9) NA NA 2'.3,4,5

5. Overtemperature AT S R a NA N.A. 1,2

6. Overpower AT S R a NA N.A. 1.2

7. Pressurizer Pressure-Low S R a N.A. N.A. 1

8. Pressurizer Pressure-HIgh S R 0 NA N.A. 1,2 I

9. Pressurizer Water Level-Hfgh S R a NA. NA 1 z

n0 10. Reactor Coolant Flow-Low S R a NA. NA 1 m 11. Steam Generator Water Level- S R a N.A. NA 1,2 Low-Low z

0

!a1 C4 co 0

L-2005-006 ENCLOSURE 4 Page 11 of 17 TABLE 4.3.1 m REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP aI ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH Z CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE C

FUNCTIONAL UNIT CHECK CLIREATION TEST TEST LOGIC TEST IS REQUIRED

-P1 12. Steam GeneratorWater S R a N.A. N.A. 1 2 I Level-Low Coincident with

_,bPjer Flow t15R A

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

14. Underfrequency-Trip of NA. R NA NA. NA. I Reactor Coolant Pump Breakers(s) Open Se 15. Turbine Trip 9 a. Autostop Oil Pressure N.A. R NA. SNU(1, 10) N.A. I

b. Turbine Stop Valve Closure NA R NA. SIU(1, 10) NA I

16. Safety Injection Input from ESF N.A. NA. NA. R NA. 1,2

17. Reactor Trip System Interlocks

a. Intermediate Range z Neutron Flux. P-8 NA. R(4) R NA. NA 26 a

r-

b. Low Power Reactor Trips Block. P-7 NA. R(4) R NA. NA. I z (Indudes P-10 Input 0 and Turbine First Stage Pressure)

CD c. Power Range Neutron Flux P4 NA. R(4) R NA. NA 1 NsreRTmvo74 4sTF

L-2005-006 ENCLOSURE 4 Page 12 of 17 INSERTNEWFOOTNOTE FORTS PAGE 3/4 3-9:

+ Not applicable to Unit 3.

Only applicable to Unit 4 through Cycle 22 operation.

Not applicable to Unit 4 starting with Cycle 23 operation.

1-2005-006 ENCLOSURE 4 Page 13 of 17 Peiced= No.: hocedwe Thk: Page:

17 Approval Date:

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

(/IV Mtd;Teo ATTACHMENT I O0 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 the consequences 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 R or a turbine first stage pressure at approximately 10% of full power equivalent), an automatic cr trip 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 RTI:D _III:N Ab FPWERS an automatic Reactor ti occur if the flow in any single loop drop eo 0M of nominal full loop flow. Conversely, an decreasing power between P-8 and the P7aiutmi Reactor trip will occur on low reactor coolant flow 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 event of a sustained steam/feedwater flow mismatch resulting from loss of normal feedwater. The specified sctpoint 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 is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Trip System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Stcanm/Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or equal to 0.665 x 10 lbs/hour. The Steam Generator Water Level-Low portion of the trip is activated when the water level drops below 10%, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam gencrators is 'Minimized.

C%s T-iFoo7NoTE c set~nee ?c-sq)

Vwo7 a need - ha_

.urorrlrormru-

L-2005-006 ENCLOSURE 4 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.

L-2005-006 ENCLOSURE 4 Page 15of17 Prcedure No.: Procedure oak: pagee 18 Arano" Datc; 0*ADM1-536 Technical Specification Bases Control Program 9/16104 l (F R 0 ) ATTACHMENT 1 N_ (Page 8 of 103)

TECHNICAL SPECIFICATION BASES 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 of complete loss of forced coolant flow. The specified setpoint assures a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. The delay is set so that the time required for a signal to reach the Reactor trip breakers following the trip of at least one undervoltage relay in both of the associated Units 4.16 kV busses shall not exceed 1.3 seconds. On decreasing power the Undervoltage Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine first stage pressure at approximately 100 of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Tripo A Turbine trip initiates a eactor trip. On decreasing power, the Reactor Trip from the Turbine trip is autornatically blce pyn ower le-we! ef nppo.vir.-tl -F9RAfETN D lllERlMAIL romrRsn wt a _tiiefr!saer~faey19-CAl QY and on increasing powrr, reinstated automatically by.

Safetv Iniection Input from ESF ~

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

Reactor Coolant Pump Breaker Position Tri/

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

NOSCR / -0077Vo-rr (seeot pag)

WV97:PSimJrakav

L-2005-006 ENCLOSURE 4 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

L-2005-006 ENCLOSURE 4 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)

TECHNICAL SPECIFICATION BASES 2.2 LIMITING SAFETY SYSTEM SETTINGS (Continued)

Reactor Coolant Pump Breaker Position Trip (Continued)

Undcrfrequency sensors are also installed on the 4.16 kV busses to detect underfrequency and initiate breaker trip on underfrequency. The undcrfrequency trip setpoints preserve the coast down energy of the 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., prevents premature block of Source Range trip) and deenergizes the high voltage to the detectors.

On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

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

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

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

WO70PSqmtfmmav

L-2005-006 ENCLOSURE5 Page 1 or 8 ENCLOSURE 5 RE-TYPED TECHNICAL SPECIFICATIONS PAGES AND (FOR INFORMATION ONLY) BASES PAGES

T-2005-006 ENCLOSURES5 Page 2 of 8

-I C TABLE 2.2-1 (Continued) m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

-D 0 ALLOWABLE z VALUE TRIP SETPOINT

-i FUNCTIONAL UNIT C Feed Flow < 23.9% below z 12. Steam/Feedwater Flow AAismatch Feed Flow < 20% below rated

=1 Coincident with rated Steam Flow Steam Flow Co Steam Generator Water Level-Low+ 2 8.15% of narrow range 2 10% of narrow range I instrument span instrument span

13. Undervoltage - 4.16 kV IBusses A and B 2 69% bus voltage 2 70% bus voltage

14. Underfrequencv-Tripol f Reactor Coolant 2 55.9 Hz Ž56.1 Hz Pump Breaker(s) Open 90 15. Turbine Trip

a. Auto Stop Oil Pressure 2 42 psig 2 45 psig

b. Turbine Stop Valve Closure Fully Closed*** Fully Closed***

16. Safety Injection Input from ESF N.A. N.A.

> 17. Reactor Trip System Interlocks m

Z a. Intermediate Range Neutron Flux, P-6 26.0X 10-11 amps Nominal 1 X 10 -10 amps m

z 0

CD

^*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-2005.006 ENCLOSURE-S Page3of 8

--I C TABLE 2.2-1 (Continued) m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

-U 0 ALLOWABLE z TRIP SETPOINT FUNCTIONAL UNIT VALUE

-I C

tt7 z b. Low Power Reactor Trips Block, P-7 co Ca 1) P-10 input S 13.0% RTP** Nominal 10% of RTP**

CA,

2) Turbine First Stage Pressure

• 13.0% Turbine Power Nominal 10% Turbine Power

c. Power Range Neutron Flux, P-8 S 48.0%+ RTP** Nominal 45%+ of RTP**

Nominal 40%++ of RTP**

• 43.0%++ RTP

d. Power Range Neutron Flux, P-1 0 Nominal 10% of RTP**

10 2 7.0% RTP**

18. Reactor Coolant Pump Breaker Position N.A. N.A.

Trip

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

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

z 0

m z

--4 z

0 C,,

z 0

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.

L-2005006 ENCLOSURES5 Page 4 of 8

-4 TABLE 3.3-1 (Continued)

M REACTOR TRIP SYSTEM INSTRUMENTATION m

MINIMUM 0 TOTAL NO. CHANNELS CHANNELS APPLICABLE z FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

-4 C 11. Steam Generator Water 3/stm. gen. 2/stm. gen. 2/stm. gen. 1,2 6 z Level--Low-Low co U)

CIO

12. Steam Generator Water Level-- 2 stm. gen. 1 stm. gen. 1 stm. gen. 1,2 6 Low Coincident With Steam/ level and level coin- level and Feedwater Flow Mismatch+ 2 stm./feed- cident with 2 stm./feed- I water flow 1 stm./feed- water flow mismatch in water flow mismatch in each stm. gen. mismatch in same stm.

same stm. gen.

co gen. or 2 stm. gen.

level and 1 stm./feedwater flow mismatch in same stm.

gen.

m 13. Undervoltage--4.16 KV Busses 2/bus 1/bus on 2/bus 1 12 z A and B (Above P-7) both busses 0

m 14. Underfrequency-Trip of Reactor 2/bus 1 to trip 2/bus 1 11 z

-4 Coolant Pump Breaker(s) Open RCPs***

z (Above P-7) 0 C,,

15. Turbine Trip (Above P-8) I

a. Autostop Oil Pressure 3 2 2 1 12 z b. Turbine Stop Valve Closure 2 2 2 1 12 0

+ 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-2005.006 ENCLOSURE S Page 5 of 8 C

z TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 TRIP Z ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH C

z CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE 1 FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED cn co 12. Steam Generator Water S R a N.A. N.A. 1,2 00 Level--Low Coincident with Steam/Feedwater Flow Mismatch+

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

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

' 15. Turbine Trip

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

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

16. Safety Injection Input from ESF N.A. N.A. N.A. R N.A. 1,2

17. Reactor Trip System

> Interlocks K

m a. Intermediate Range Z Neutron Flux, P-6 N.A. R(4) R N.A. N.A. 2**

0 m b. Low Power Reactor H Trips Block, P-7 N.A. R(4) R N.A. N.A. 1 Z (includes P-10 input En and Turbine First Stage Pressure) z 0Z c. Power Range Neutron Flux, P-8 N.A. R(4) R N.A. N.A. 1

+ 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-2005-006 ENCLOSURE 5 Page 6 of 8 Priccdurc No.: Pocedurc Titc: PaCe:

17 Approv-Al DatI.

O-ADM-536 Technical Specification Bases Control Program D  : xxx AITACHIMIENT I (Page 7 of 103)

TECHNICAL SPECIFICATION BASES 2.2 LIMITING SAFETY SYSTEM SETTINGS (Continued)

Reactor Coolant Flow The Reactor Coolant Flow-Low trip provides core protection to prevent DNB by mitigating the consequences 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 POWER or a turbine first stage pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of loop design flow. Above P-8, an automatic Reactor trip will occur if the flow in any single loop drops below 90%'of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Lcvel The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam/fecdwater flow mismatch resulting from loss of normal fccdwatcr. The specified sctpoint 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 is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Trip System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam/Fecdwatcr Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or cqual to 0.665 x 1O6 Ibs/hour. The Steam Generator Water Lcvel-Low portion of the trip is activated when the water level drops below 10%, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor trip before the steam generators arc dry. Thercf6re, the required capacity and starting time requirements of the auxiliary feedwater pumps arc reduced and the resulting 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

L-2005-006 ENCLOSURE 5 Page 7 of 8 Procedure No.: Procedure

Title:

'age:

18 n Approval Date:

0-A1)11-536 Technical Specification Bases Control Program XXXXX ATTACHMENT I (Page 8 of 103)

TECHNICAL SPECIFICATION BASES 2.2 LIMITING SAFETY SYSTEM SETTINGS (Continued)

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 of complete loss of forced coolant flow. Thc spccificd setpoint assures a Reactor trip signal'is generated before the Low Flow Trip Setpoint is reached. Time delays arc incorporated in the Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. The delay is set so that the time required for a signal to reach the Reactor trip breakers following the trip of at least one undervoltage relay in both of the associated Units 4.16 kV busses shall not cxcced 1.3 seconds. On decreasing power the Undervoltage Bus trips are automatically blocked by P-7 (a'power level of approximately 10% of RATED THERMAL POWER with a turbine first stage pressure at approximately 10% 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 is automatically 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 ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

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

L-2005-006 ENCLOSURE 5 Page 8 of 8 Procedure No.: Proeedare

Title:

Pae:

1 19

. . Approval Date:

O-ADIN1-536 Technical Specification Bases Control Program XXXYXp ATTACHMENT I (Page 9 of 103)

TECIINICAL SPECIFICATION BASES 2.2 LIMITING SAFETY SYSTEM SETTINGS (Continued)

Reactor Coolant Pump Breaker Position Trip (Continucd)

Underfrcquency sensors are also installed on the 4.16 kV busses to detect undcrfrequency and initiate breaker trip on underfrequency. The underfrequency trip sctpoints preserve the coast down energy of the 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., prevents premature block of Source Range trip) and deenergizes the high voltage to the dctcctors.

On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

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

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops, and one or more reactor coolant pump breakers open, and turbine trip. On decreasing power, the P-8 interlock automatically blocks the trip on low flow in one 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 and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Rangc high voltage power. On decreasing power, the Intermediate Range trip and the-Low Selpoint Power Range trip are automatically reactivated. P-10 also provides input to P-7. The trip sctpoint on increasing power shall be Ž 10% and the reset point shall be less than or equal to 10%.

wQY-nPVM.e L-2005-006 ENCLOSURE 6 Page 1 of 9 ENCLOSURE 6 CAW-05-1958, Application For Withholding Proprietary Information From Public Disclosure

Westinghouse Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 3744643 Document Control Desk Directfax: (412) 3744011 Washington, DC 20555-0001 e-mail: greshajaewestinghouse.com Our ref: CAW-05-1958 February 22, 2005 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

"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)

The proprietary information for which withholding is being requested in the above-referenced report is further 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 basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

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 the Westinghouse affidavit should reference this letter, CAW-05-1958, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

Very truly yIrs, 6 JA. Gresham, Manager Regulatory Compliance and Plant Licensing Enclosures cc: B. Benney L. Feizollahi A BNFL Group company e

-11,I

CAW-05-1958 bcc: J. A. Gresham (ECE 4-7A) IL R. 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)

A BNFL Group company

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 Seal Sharon L Rod, Notary Publc Monroeville Boro, Allegheny County My 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) 1am 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 competitive advantage.

(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 NRC in 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 the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so 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 being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types 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 to make the number of copies of the information contained in these reports which are necessary for its internal 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 public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its initernal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

L-2005-006 ENCLOSURE 7 Page I of 9 ENCLOSURE 7 CAW-05-1959, Application For Withholding Proprietary Information From Public Disclosure

Westinghouse . .

Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 3744643 Document Control Desk Directfax: (412) 374-4011 Washington, DC 20555-0001 e-mail: greshajaewestinghouse.com Our ref: CAW-05-1959 February 22, 2005 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

"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)

The proprietary information for which withholding is being requested in the above-referenced report is further 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 basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

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 the Westinghouse affidavit should reference this letter, CAW-05-1959, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

Very truly yours,

41. A. Gresham, Manager A/ Regulatory Compliance and Plant Licensing Enclosures cc: B. Benney L. Feizollahi A BNFL Group company

CAW-05-1959 bcc: J. A. Gresham (ECE 4-7A) IL R. 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)

A BNFL Group company

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 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. 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 County My 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) 1am 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) 1have 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 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 competitive advantage.

(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 NRC in 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 the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so 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 being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types 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 to make the number of copies of the information contained in these reports which are necessary for its internal 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 public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.