Text

Supplement to License Amendment Request to Extend Reactor Trip System & Engineered Safety Features Actuation System Surveillance Time Requirements as Evaluated in WCAP-15376
	ML030640115
Person / Time
Site:	Cook
Issue date:	02/27/2003
From:	Joseph E Pollock; Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP:NRC:3311, TAC MB6324, TAC MB6325
	Download: ML030640115 (98)
	v • d • e

Indiana Michigan Power Company 500 Circle Drive Buchanan, Mi 49107 1395 INDIANA MICHIGAN POWER February 27, 2003 AEP:NRC:3311 10 CFR 50.90 Docket No.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 SUPPLEMENT TO LICENSE AMENDMENT REQUEST TO EXTEND REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM SURVEILLANCE TIME REQUIREMENTS AS EVALUATED IN WCAP-15376 (TAC Nos. MB6324 and MB6325)

Reference:

1) Letter from J. E. Pollock, I&M, to U. S. Nuclear Regulatory Commission Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Amendment Request to Extend Reactor Trip System and Engineered Safety Features Actuation System Surveillance Requirements as Evaluated in WCAP-15376," AEP:NRC:2311, dated August 30, 2002

2) Letter from W. H. Ruland, NRC, to R. H. Bryan, Westinghouse Owners Group, "Acceptance for Referencing of Topical Report WCAP-15376-P, Rev. 0, 'Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times' (TAC. No. MB0983),"

dated December 20, 2002 By Reference 1, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposed to amend Appendix A, Technical Specifications (TS), of Facility Operating Licenses DPR-58 and DPR-74. I&M proposed to revise the reactor trip system and engineered safety features actuation system surveillance requirements based on the evaluation in WCAP-15376-P, Revision 0, "Risk-Informed Assessment of the RPS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times."

U. S. Nuclear Regulatory Commission AEP:NRC:3311 Page 2 By Reference 2, the Nuclear Regulatory Commission (NRC) issued a Safety Evaluation (SE), which documents the acceptability of WCAP-15376-P for licensing applications. In Reference 1, I&M committed to review the SE when issued and supplement the license amendment request with a response to any additional stipulations made in the SE on WCAP-15376-P. This letter provides I&M's response to the conditions and limitations stipulated in the referenced SE and, with the exception of the editorial changes, withdraws the proposed changes in Reference 1 that are not bounded by WCAP-15376-P to expedite NRC review of the TS changes that are bounded by WCAP-15376-P This letter also provides additional basis for the proposed changes in relation to WCAP-15376-P and Technical Specifications Task Force (TSTF) Traveler TSTF-411, Revision 1, "Surveillance Test Interval Extension for Components of the Reactor Protection System" provides an affirmation pertaining to the statements made in this letter. Enclosure 2 provides a detailed description of the withdrawal of the proposed changes that are beyond the scope of the SE. Attachments IA and 1B to this letter provide new marked-up TS pages to replace, in their entirety, the corresponding pages submitted in Attachments 1A and 1B to Reference 1.

Attachments 2A and 2B provide new TS pages, with the changes incorporated, to replace, in their entirety, the corresponding pages submitted in Attachments 2A and 2B to Reference 1. Attachment 3 provides I&M's response to the conditions and limitations stated in Reference 2 and fulfills the commitment made in Reference 1. Attachment 4 provides a table with additional basis for the proposed changes as updated by this letter. Attachment 5 identifies the commitments documented in this letter.

The information provided in this letter consists of supporting information for the amendment request previously submitted by Reference 1. The proposed TS changes, as supplemented by this letter, remain within the scope previously proposed by Reference 1. Therefore, the No Significant Hazards Consideration evaluation and the evaluation of Environmental Considerations provided in Enclosure 2 to Reference 1 continue to bound the proposed changes.

U. S. Nuclear Regulatory Commission AEP:NRC:3311 Page 3 Should you have any questions, please contact Mr. Brian A. McIntyre, Manager of Regulatory Affairs, at (269) 697-5806.

Sincerely, J. E. Pollock Site Vice President KAS/dmb

Enclosures:

1. Affirmation

2. Detailed Description and Technical Evaluation of Proposed Change Attachments 1A and lB. Technical Specification Pages Marked to Show the Proposed Changes 2A and 2B. Technical Specification Pages with the Proposed Changes Incorporated

3. I&M Response to NRC Conditions and Limitations

4. Basis for Proposed Changes to the Technical Specifications

5. Regulatory Commitments c: K. D. Curry, Ft. Wayne AEP J. E Dyer, NRC Region III J. T. King, MPSC MDEQ - DW & RPD NRC Resident Inspector J. F. Stang, Jr, NRC Washington, DC

U. S. Nuclear Regulatory Commission AEP:NRC.3311 Page 4 bc: A C. Bakken III, w/o enclosures/attachments M. J. Finissi S. A. Greenlee R. J. Grumbir D. W. Jenkins, w/o enclosures/attachments J. A. Kobyra, w/o enclosures/attachments B. A. McIntyre, w/o enclosures/attachments J E. Newmiller J. E. Pollock, w/o enclosures/attachments D. J. Poupard M. K. Scarpello, w/o enclosures/attachments T. K. Woods, w/o enclosures/attachments

Enclosure 1 to AEP:NRC:3311 Page I AFFIRMATION I, Joseph E. Pollock, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company J. E. Pollock Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME TF2003 JULIE E. NEWMILLER Notary Public, Berrien County, MI Aug 22,2004 My Commrission Expires y Commission Expires :c*?

-° .. .

to AEP:NRC:3311 Page I DETAILED DESCRIPTION AND TECHNICAL EVALUATION OF PROPOSED CHANGE

1.0 DESCRIPTION

By Reference 1, Indiana Michigan Power Company (I&M) proposed to amend the Donald C.

Cook Nuclear Plant (CNP) Unit 1 and Unit 2 Technical Specifications (TS) to increase channel functional test (CFT) surveillance intervals for analog channels, logic cabinets, and reactor trip breakers in accordance with the evaluation and justifications presented in WCAP-15376-P, Revision 0, "Risk-Informed Assessment of the RPS and ESFAS Surveillance-Test Intervals and Reactor Trip Breaker Test and Completion Times." Additional changes were proposed to align CNP TS to NUREG-1431, Revision 2, "Standard Technical Specifications Westinghouse Plants," and to clarify existing TS. By Reference 2, the Nuclear Regulatory Commission (NRC) issued a Safety Evaluation (SE) accepting WCAP-15376-P as a generic basis for licensees to reference in license amendment requests. This enclosure withdraws the changes proposed in Reference 1, with the exception of editorial changes, that are not bounded by WCAP-15376-P or within the scope of Technical Specification Task Force (TSTF) Traveler TSTF-4 11, Revision 1, "Surveillance Test Interval Extension for Components of the Reactor Protection System."

2.0 PROPOSED CHANGE

I&M withdraws the following changes originally proposed in Reference 1:

1. Delete Mode 2 applicability for reactor trip system (RTS) Functional Unit (FU) 9, Pressurizer Pressure-Low, and FU 11, Pressurizer Water Level-High in TS Table 3.3-1, "Reactor Trip System Instrumentation."

2. Change Action 13 to the Action Statement Notations for TS Table 3.3-1. This Action, which only applied to FU 21, "Reactor Trip Breakers," in Modes 1 and 2 was proposed as:

With one Reactor Trip Breaker channel inoperable due to an inoperable diverse trip feature (Undervoltage or shunt trip attachment), restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The channel shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the channel to OPERABLE status.

3. Change the applicability of TS Table 4.3-1, "Notation (1)" to "Notation (17)" for the following RTS FUs:

1.A. Manual Reactor Trip - Shunt Trip Function 1.B. Manual Reactor Trip - Undervoltage Trip Function

23. Reactor Trip Bypass Breaker to AEP:NRC:3311 Page 2

4. Delete the TS Table 4.3-1 requirement to perform a Start-Up CFT, including deleting the applicability of Notations (1) and (11) for this S/U CFT, for the following RTS FUs:

2. Power Range, Neutron Flux

6. Source Range, Neutron Flux 21.A. Reactor Trip Breaker - Shunt Trip Function 21.B. Reactor Trip Breaker - Undervoltage Trip Function

5. Add Notation (16) to TS Table 4.3-1 to state, "Applicable to any reactor trip bypass breakers that are racked in and closed for bypassing a reactor trip breaker" and add Notation (16) to FU 23. Notation (16) was proposed only for FU 23, "Reactor Trip Bypass Breaker."

To prevent having to change Notation (17) due to the withdrawal of Notation (16), it is proposed to change Notation (16) to indicate, "Not Used." Notation (17) remains unchanged as proposed in Reference 1.

Attachments 1A and IB to this letter provide new marked-up TS pages to replace, in their entirety, the corresponding pages submitted in Attachments 1A and 1B to Reference 1.

Attachments 2A and 2B provide new TS pages, with the changes incorporated, to replace, in their entirety, the corresponding pages submitted in Attachments 2A and 2B to Reference 1.

3.0 TECHNICAL ANALYSIS

The changes withdrawn from this license amendment request were proposed to align CNP TS to NUREG-1431 and/or to clarify the existing TS, and are not bounded by WCAP-15376-P and the associated TSTF-4 11. The withdrawal is made to expedite NRC review of the proposed changes in Reference 1 that are bounded by WCAP-15376-P.

4.0 REGULATORY SAFETY ANALYSIS No Significant Hazards Consideration The proposed TS changes, as supplemented by this letter, remain within the scope previously proposed by Reference 1. Therefore, the No Significant Hazards Consideration evaluation provided in Enclosure 2 to Reference 1 bound the changes proposed in this letter and remains valid.

Applicable Regulatory Requirements/Criteria The proposed TS changes, as supplemented by this letter, remain within the scope previously proposed by Reference 1. Therefore, discussion of applicable regulatory requirements/criteria, provided in Enclosure 2 to Reference 1, remains valid.

to AEP:NRC:3311 Page 3

5.0 ENVIRONMENTAL CONSIDERATION

The proposed TS changes, as supplemented by this letter, remain within the scope previously proposed by Reference 1. Therefore, the evaluation of Environmental Considerations, provided in Enclosure 2 to Reference 1, remains valid.

6.0 REFERENCES

1. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Amendment Request to Extend Reactor Trip System and Engineered Safety Features Actuation System Surveillance Requirements as Evaluated in WCAP-15376," AEP:NRC:2311, dated August 30, 2002

2. Letter from W. H. Ruland, NRC, to R. H. Bryan, Westinghouse Owners Group, "Acceptance for Referencing of Topical Report WCAP-15376-P, Rev. 0, 'Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times' (TAC. No. MB0983)," dated December 20, 2002

ATTACHMENT 1A to AEP:NRC:3311 UNIT 1 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW THE PROPOSED CHANGES REVISED PAGES UNIT 1 1-9 3/43-5 3/4 3-8 3/4 3-12 3/4 3-13 3/43-14 3/4 3-31 3/4 3-32 3/4 3-33 3/4 3-33a 3/4 3-33b 3/4 3-34

1.0 DEFINITIONS TABLE 1.2 FREOUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

w At least once per 7 days.

M At least once per 31 days Q At least once per 92 days

,4 M~onths At least once per 214 days SA At least once per 184 days R At least once per 549 days.

S/U Prior to each reactor startup.

P Completed prior to each release.

N.A. Not Applicable.

COOK NUCLEAR PLANT-UNIT I Page 1-9 AMENDMENT 72

TABLE 3.3-1 (Continued)

z REACTOR TRIP SYSTEM INSTRUMENTATION

> MINIMUM STOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

16. Undervoltage-Reactor Coolant 4-1/bus 2 3 16 Pumps

17. Underfrequency-Reactor 4-1/bus 2 3 1 6#

"- Coolant Pumps

18. Turbine Trip A. Low Fluid Oil 3 2 2 1 7#

Pressure B. Turbine Stop Valve 4 4 4 1 7#

Closure t 19. Safety Injection Input from 2 1 2 1, 2 1 ESF

20. Reactor Coolant Pump Breaker Position Trip Above P-7 I/breaker 2 I/breaker per 1 11 operating loop

ýZ 21. Reactor Trip Breakers 2 1 2 1,2 4-, 13,['15 3", 4", 5 14

22. Automatic Trip Logic 2 1 2 1, 2 1 3", 4, 5" 14

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3-1 (Continued)

ACTION 8 (Deleted )

ACTION 9 (Deleted )

ACTION 10 (Deleted.)

I ACTION 11 With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 13 With one of the diverse trip features (Undervoltage or shunt trip attachment), inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 1.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

ACTION 14 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

ACTION 15- With the number of OPERABLE Reactor Trip Breaker channels one less than required by the Minimum Channels OPERABLE requirement for reasons other than an inoperable diverse trip feature, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . One channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Range Neutron P-6 prevents or defeats the manual Flux Channels less than 6x10" amps block of source range reactor trip.

AMENDMENT 99, 4O, 140 PLANT-UNIT II NUCLEAR PLANT-UNIT COOK NUCLEAR Page 314 3-8 Page 3/4 3-8 AMENDMENT 99,4-20, 140

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODE IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip A. Shunt Trip Function N.A. N.A. S/U(1)(10) 1, 2, 3%, 4%, 5*

B. Undervoltage Trip N.A. N.A. S/U(1)(10) 1, 2, 3%, 4%, 5*

Function

2. Power Range, Neutron S D(2,8), M(3,8), SM and S/U(1) 1,2 and' Flux and Q (6,8)

3. Power Range, Neutron N.A. R(6) 1,2 Flux, High Positive Rate

4. Power Range, Neutron N.A. R(6) QM 1,2 Flux, High Negative Rate

5. Intermediate Range, S R(6,8) 1, 2, and" Neutron Flux

6. Source Range, Neutron S R(6,14) M(14) and 2(7), 3(7), 4 and 5 Flux S/U(1)

7. Overtemperature delta T S R(9) 1, 2

8. Overpower delta T S R(9) SAM 1,2

9. Pressurizer Pressure- S R SA M 1,2 Low

10. Pressurizer Pressure- S R FA M 1,2 High

11. Pressurizer Water Level S R FA M 1,2 High

12. Loss of Flow-Single Loop S R(8) KXAM 1 COOK NUCLEAR PLANT-UNIT I Page 314 3-12 AMENDMENT 400, 4-2,0, -t-2,4, 144

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODE IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

13. Loss of Flow-Two Loops S R(8) N.A. 1

14. Steam Generator Water S R 1,2 Level -- Low-Low I

15. Steam/Feedwater Flow S R 1,2 Mismatch and Low Steam Generator Water Level

16. Undervoltage-Reactor N.A. R M 1 Coolant Pumps

17. Underfrequency-Reactor N.A. R M 1 Coolant Pumps

18. Turbine Trip A:. Low Fluid Oil Pressure N.A. N.A. S/U(1) 1,2 B. Turbine Stop Valve N.A. N.A. S/u(1) 1,2 Closure

19. Safety Injection Input from N.A. N.A. RM(4)f 1,2 ESF

20. Reactor Coolant Pump N.A. N.A. R N.A.

Breaker Position Trip

21. Reactor Trip Breaker A. Shunt Trip Function N.A. N.A. EM'6nths M 1, 2, 3*, 4*, 5*

(5)(11) and S/U(1)(1 1)

B. Undervoltage Trip N.A. N.A. ý1Months M 1, 2, 3", 4", 5*

Function (5)(11) and SIU(1)(I 1)

22. Automatic Trip Logic N.A N.A. 1, 2, 3*, 4*, 5*

23. Reactor Trip Bypass N.A. N.A. 1, 2, 3*, 4*, 5*

Breaker and(1)(13 SJU(1)(13)

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-13 AMENDMENT -t00,"1-0,144

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-1 (Continued)

NOTATION

- With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) - If not performed in previous 7 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) - Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate if absolute difference greater than or equal to 3 percent.

(4) - Manual ESF functional input check every 18 months.

(5) - Each train tested t leias every other month 92days.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

(8) - The provisions of Specification 4.0.4 are not applicable.

(9) - The provisions of Specification 4.0.4 are not applicable for f1 (delta I) and f2 (delta I) penalties, or for measurement of delta T. (See also Table 2.2-1).

(10) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12) - Local manual shunt trip prior to placing breaker in service.

(13) - Automatic Undervoltage Trip.

(14) - The provisions of Specification 4.0.4 are not applicable when leaving MODE 1. In such an event, the calibration and/or functional test shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after leaving MODE 1.

(15) IE*ach train tested at least every, other,92 days.

(16) . - Not Used.

(17) - If not performed in previous 184 days.

COOK NUCLEAR PLANT-UNIT I Page 3/4 3-14 AMENDMENT 99, 4U-2,141

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.3 INSTRUMENTATION TABLE 4 3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

1. SAFETY INJECTION, TURBINE TRIP. FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a Manual Initiation See Functional Unit 9 - ----

b Automatic Actuation Logic N.A. N.A §A M(2) NA 1,2,3,4 c Containment Pressure- High S R 9A M (3) NA 1,2,3 d Pressurizer Pressure-Low S R RAM NA. 1,2,3

e. Differential Pressure S R FA M N.A. 1,2,3 Between Steam Lines-- High

f. Steam Line Pressure-Low S R 9X M NA 1,2,3

2. CONTAINMENT SPRAY

a. Manual Initiation See Functional Unit 9.

b Automatic Actuation Logic N.A. N.A. AM (2) N.A. 1,2,3,4 C. Containment Pressure S R SM (3) N.A. 1,2,3 High-High COOK NUCLEAR PLANT-UNIT I Page 3/4 3-31 AMENDMENT 400, 4-20, 4-24 444, 4-53, 214

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REOUIRED 3 CONTAINMENT ISOLATION a- Phase "A" Isolation

1) Manual See Functional Unit 9

2) From Safety Injection N.A N.A. §X M (2) N A 1,2,3,4 Automatic Actuation Logic

b. Phase "B" Isolation

1) Manual See Functional Unit 9

2) Automatic Actuation N.A. N.A SAM (2) N.A 1,2,3,4 Logic

3) Containment Pressure S R SM (3) N.A 1,2,3 High-High c Purge and Exhaust Isolation

1) Manual See Functional Unit 9

2) Containment S R Q N A. 1,2,3,4 Radioactivity-High COOK NUCLEAR PLANT-UNIT I Page 3/4 3-32 AMENDMENT 40O0, 444, 4-5-3, 183

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

4. STEAM LINE ISOLATION

a. Manual ---- ziee 1'unctional Unit 9 b Automatic Actuation Logic N.A. N.A. M (2)

MA N.A. 1,2,3,

c. Containment Pressure S R M M(3) N.A. 1,2,3 High-High

d. Steam Flow in Two Steam S R N.A. 1,2,3 Lines--High Coincident with Tv,--Low-Low

e. Steam Line Pressure-Low S R ýxM N.A. 1,2,3

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water S R §X-M N.A. 1,2,3 Level--High-High

6. MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a Steam Generator Water Level--Low-Low S R §A M N.A. 1,2,3

b. 4 kv Bus Loss of Voltage S R M N.A. 1,2, 3 c Safety Injection N.A. N.A. SA M (2) N.A. 1,2,3

d. Loss of Main Feed Pumps N.A. N.A. R N.A. 1,2 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-33 AMENDMENT 400, 430, 4U-, 444 4._53, 214

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REOUIRED

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water S R NA 1,2,3 Level-Low-Low

b. Reactor Coolant Pump Bus N.A. R M N.A. 1,2,3 Undervoltage 8 LOSS OF POWER

a. 4 kv Bus Loss of Voltage S R M N.A 1,2,3,4 b 4 kv Bus Degraded Voltage S R M N.A 1,2,3,4 COOK NUCLEAR PLANT-UNIT I Page 3/4 3-33a AMENDMENT 400, 20, 444, 153

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED 9 Manual

a. Safety Injection (ECCS) NA N.A. N.A R 1, 2, 3, 4 Feedwater Isolation Reactor Trip (SI)

Containment Isolation Phase "A" Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System

b. Containment Spray N.A N.A. N.A R 1,2,3,4 Containment Isolation Phase "B" Containment Purge and Exhaust Isolation c Containment Isolation- NA N.A. NA R 1, 2, 3, 4 Phase "A" Containment Purge and Exhaust Isolation d Steam Line Isolation NA N.A. Q R 1, 2, 3 e Containment Air NA N.A. N A R 1, 2, 3, 4 Recirculation Fan 10 CONTAINMENT AIR RECIRCULATION FAN a Manual - See Functional Unit 9- ------ - -----

b Automatic Actuation Logic NA N.A. SA M (2) NA 1, 2, 3 c Containment Pressure - High S R RA M (3) NA 1, 2, 3 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-33b AMENDMENT 4-53, 2,04, 234

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

TABLE NOTATION (1) Deleted (2) Each train or logic channel shall be tested at least every other ý2 U1 days.

(3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-34 AMENDMENT 3.9, 204

ATTACHMENT 1B to AEP:NRC:3311 UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW THE PROPOSED CHANGES REVISED PAGES UNIT 2 1-10 3/4 3-4 3/4 3-7 3/43-11 3/4 3-12 3/4 3-13 3/4 3-30 3/4 3-31 3/4 3-32 3/4 3-33

1.0 DEFINITIONS TABLE 1 2 FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months w At least once per 7 days M At least once per 31 days Q At least once per 92 days

r-At least once per 124 days SA At least once per 184 days R At least once per 549 days S/U Prior to each reactor start-up P Completed prior to each release N.A. Not Applicable COOK NUCLEAR PLANT-UNIT 2 Page 1-10 AMENDMENT 51

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.3 INSTRUMENTATION TABLE 3.3-1 (Continued)

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

16. Undervoltage-Reactor Coolant 4-1/bus 2 3 Pumps

17. Underfrequency-Reactor 4-1/bus 2 3 Coolant Pumps

18. Turbine Trip A. Low Fluid Oil Pressure 3 2 2 1 1

B. Turbine Stop Valve 4 4 3 Closure

19. Safety Injection Input from 2 2 1,2 1 ESF

20. Reactor Coolant Pump Breaker Position Trip Above P-7 1/breaker 2 I/breaker per I 11 operating loop

21. Reactor Trip Breakers 2 2 4, 13!715 3*, 4*, 5* 14

22. Automatic Trip Logic 2 2 1,2 1 3*, 4*, 5* 14 COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-4 AMENDMENT 86, 40-7, 4-27,172

TABLE 3 3-I(Continued)

ACTION I I - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

I ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 13 - With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION

1. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

ACTION 15 - With the number of OPERABLE Reactor Trip Breaker channels one less than required by the Minimum Channels OPERABLE requirement for reasons other than an inoperable diverse trip feature, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . One channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Range P-6 prevents or defeats the Neutron Flux Channels < 6 X 10".1 amps. manual block of source range reactor trip.

AMENDMENT 86, 127 2 Page 3/4 3-7 COOK NUCLEAR PLANT-UNIT COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-7 AMENDMENT 86, 127

0 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH

'4 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip A. Shunt Trip Function N.A. N.A. S/U(1)(I0) 1, 2, 3% 4% 5*

B. Undervoltage Trip Function N.A. N.A. S/U(1)(10) 1, 2, 3, 4%, 5*

k4 1, 2 and*

2. Power Range, Neutron Flux S D(2,8), M(3,8), Q M and S/U(1) and Q (6,8)

3. Power Range, Neutron Flux, High Positive Rate N.A. R(6) 1, 2

4. Power Range, Neutron Flux, High Negative Rate N.A. R(6) Q M 1,2

5. Intermediate Range, Neutron Flux S R(6,8) S/U(4)17) 1, 2, and"

6. Source Range, Neutron Flux S R(6,14) M(14) and S/U(I) 2(7), 3(7), 4 and 5

7. Overtemperature AT S R(9) 1,2 V ~M

8. Overpower AT S R(9) 1, 2

9. Pressurizer Pressure -- Low S R stiM 1, 2

10. Pressurizer Pressure-- High S R SA M 1, 2

11. Pressurizer Water Level -- High S R SM 1, 2 z 12. Loss of Flow-Single Loop S R(8) SM 1

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED S R(8) N.A. 1 I

13. Loss of Flow-Two Loops z 14. Steam Generator Water Level -- Low-Low S R SA M 1,2

15. Steam/Feedwater Flow Mismatch and Low Steam S R 7M 1,2 W Generator Water Level

16. Undervoltage-Reactor Coolant Pumps N.A. R M 1

17. Underfrequency-Reactor Coolant Pumps N.A. R M 1

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U(I) 1,2 B. Turbine Stop Valve Closure N.A. N.A. SIU(1) 1,2

19. Safety Injection Input from EFS N.A. N.A. S M(4)t 1,2

20. Reactor Coolant Pump Breaker Position Trip N.A. N.A. R N.A.

21. Reactor Trip Breaker N.A. N.A. T 7775m (5) (11) 1,2, 3*, 4*, 5" A. Shunt Trip Function and SIU(1)(1)

B. Undervoltage Trip Function N.A. N.A. and M (5)(11) 1,2,3*,4*,5*

and S/U(1)(11)

22. Automatic Trip Logic N.A N.A. 1, 2,3", 4", 5" N.A. ,47'*hM'(ý(12) 1, 2, 3*, 4*, 5*

z 23. Reactor Trip Bypass Breaker N.A.

and S/U(1)(13)

X4 3

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-1 (Continued)

NOTATION

With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) - If not performed in previous 7 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent (3) - Compare incore to excore axial offset above 15% of RATED THERMAL POWER. Recalibrate if absolute difference greater than or equal to 3 percent.

(4) - Manual ESF functional input check every 18 months.

(5) - Each train tested Watlea every other month r2da`y's (6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

(8) - The provisions of Specification 4.0.4 are not applicable.

(9) - The provisions of Specification 4.0.4 are not applicable for f, (delta I) and f2 (delta I) penalties, or for measurement of delta T. (See also Table 2.2-1).

(10) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s).

(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers (12) - Local manual shunt trip prior to placing breaker in service.

(13) - Automatic Undervoltage Trip.

(14) - The provisions of Specification 4.0.4 are not applicable when leaving MODE 1. In such an event, the calibration and/or functional test shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after leaving MODE 1.

(15) - Each train tested at least every otr 92 days.

Si6) Not Used.

- If not performed in previous 184 days.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-13 AMENDMENT 86, 407, 128

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a Manual Initiation See Functional Unit 9 b Automatic Actuation NA NA SA M(2) NA 1,2,3,4 Logic c Containment Pressure S R (3)

MAM NA 1,2,3 High d Pressurizer Pressure S R SA M NA 1,2,3 Low e Differential Pressure S R SAM N A. 1,2,3 Between Steam Lines High f Steam Line Pressure S R SA M NA 1,2,3 Low

2. CONTAINMENT SPRAY

a. Manual Initiation - -_ See Functional Unit 9 b Automatic Actuation NA NA SXM (2) NA 1,2,3,4 Logic c Containment Pressure S R S M (3) NA 1,2,3 High-High 3 CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual - See Functional Unit 9

2) From Safety N A NA SAM (2) NA 1,2,3,4 Injection Automatic Actuation Logic b Phase "B" Isolation

1) Manual - See Functional Unit 9

2) Automatic Actuation N A NA SAM (2) NA 1,2,3,4 Logic

3) Containment S R SA M (3) NA 1,2,3 Pressure- High High COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-30 AMENDMENT 34, 4-34, 4-37, 4-58, 224

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4 3/4.3 INSTRUMENTATION I TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED c Purge and Exhaust Isolation

1) Manual See Functional Unit 9

2) Containment S R Q NA 1,2,3,4 Radioactivity - High 4 STEAM LINE ISOLATION a Manual See Functional Unit 9 b Automatic Actuation NA NA 9A M (2) NA 1,2,3 Logic c Containment Pressure S R SAM (3) NA 1,2,3 High-High d Steam Flow in Two Steam S R SA M NA 1,2,3 I

Lines -- High Coincident with T*,, - Low-Low e Steam Line Pressure S R gAM NA 1.2.3 Low 5 TURBINE TRIP AND FEEDWATER ISOLATION a- Steam Generator Water S R SixM NA 1,2,3 Level - High-High 6 MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water S R SA M NA 1,2,3 Level - Low-Low b 4 kV Bus Loss of Voltage S R M NA 1,2,3 c Safety Injection NA NA SAM (2) NA 1,2,3 d Loss of Main Feed Pumps NA NA R NA 1,2 I I

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-31 AMENDMENT 82, 97, 4-34, 4-34, 4-37, 4-59,-68, 224 I

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED 7 TURBINE DRIVEN AUXILIARY FEEDWATER PUMP a Steam Generator Water Level - Low-Low S R FA M NA 1,2,3

b. Reactor Coolant Pump NA R M N A 1,2,3 Bus Undervoltage 8 LOSS OF POWER

a. 4 kv Bus Loss of Voltage S R M NA 1,2,3,4 b 4 kv Bus Degraded S R M N.A 1,2,3,4 Voltage

9. MANUAL a Safety Injection (ECCS)

Feedwater Isolation NA NA N A R 1, 2, 3, 4 Reactor Trip (SI)

Containment Isolation Phase "A" Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System b Containment Spray NA N A N A R 1,2,3,4 Containment Isolation Phase "B" Containment Purge and Exhaust Isolation c Containment Isolation - NA NA N A R 1,2, 3,4 Phase "A" Containment Purge and Exhaust Isolation d Steam Line Isolation NA N A Q R 1,2, 3 e Containment Air NA NA N A R 1,2,3,4 Recirculation Fan 10 CONTAINMENT AIR RECIRCULATION FAN

a. Manual -- See Functional Unit 9 -

b Automatic Actuation NA N A SA M (2) N A. 1, 2, 3 Logic c Containment Pressure - S R SA M (3) NA 1, 2, 3 High COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-32 AMENDMENT 82, 9-7, 4-34, 4-37, "4-59, 4-89, 217

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

TABLE NOTATION (I) Deleted (2) Each train or logic channel shall be tested at least every other 34 days.

(3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-33 AMENDMENT 8-2, 4-34, 189

ATTACHMENT 2A to AEP:NRC:3311 UNIT I TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED REVISED PAGES UNIT 1 1-9 3/4 3-5 3/4 3-8 3/4 3-12 3/4 3-13 3/4 3-14 3/4 3-31 3/4 3-32 3/43-33 3/4 3-33a 3/4 3-33b 3/4 3-34

1.0 DEFINITIONS TABLE 1.2 FREOUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months w At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

4 Months At least once per 124 days SA At least once per 184 days.

R At least once per 549 days.

S/U Prior to each reactor startup.

P Completed prior to each release N.A. Not Applicable.

AMENDMENT 72, 1 COOK NUCLEAR PLANT-UNIT I Page 1-9

n 0 TABLE 3.3-1 (Continued) r REACTOR TRIP SYSTEM INSTRUMENTATION

>- MINIMUM STOTAL NO. CHANNELS CHANNELS APPLICABLE r" FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

16. Undervoltage-Reactor Coolant 4-l/bus 2 3 1 6#

Pumps

17. Under frequency-Reactor 4-1/bus 2 3 16

-- Coolant Pumps

18. Turbine Trip A. Low Fluid Oil 3 2 2 1 7 Pressure B. Turbine Stop Valve 4 4 4 1 7 Closure

19. Safety Injection Input from 2 1 2 1, 2 ESF

20. Reactor Coolant Pump Breaker Position Trip Above P-7 I/breaker 2 I/breaker per 1 11 operating loop z 21. Reactor Trip Breakers 2 1 2 1, 2 13, 15 3", 4, 5" 14

-- 22. Automatic Trip Logic 2 1 2 1, 2 1 3", 4", 5' 14

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3-1 (Continued)

ACTION 8 (Deleted.)

ACTION 9 (Deleted.)

ACTION 10 (Deleted.)

ACTION 11 With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 13 With one of the diverse trip features (Undervoltage or shunt trip attachment), inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 1.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

ACTION 14 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

ACTION 15 - With the number of OPERABLE Reactor Trip Breaker channels one less than required by the Minimum Channels OPERABLE requirement for reasons other than an inoperable diverse trip feature, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . One channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Range Neutron P-6 prevents or defeats the manual Flux Channels less than 6x10"' amps. block of source range reactor trip.

3/4 3-8 AMENDMENT 99, 4, 440, COOK COOK NUCLEAR PLANT-UNIT II NUCLEAR PLANT-UNIT Page Page 3/4 3-8 AMENDMENT -99,4-2-0, W4,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODE IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip A. Shunt Trip Function N.A. N.A. S/U(I)(10) 1, 2, 3%, 4%, 5" B. Undervoltage Trip NA N.A. S/U(I)(0) 1, 2, 3, 4%, 5*

Function

2. Power Range, Neutron S D(2,8), M(3,8), Q and S/U(I) 1, 2 and" Flux and Q(6, 8)

3. Power Range, Neutron N.A. R(6) Q 1,2 Flux, High Positive Rate

4. Power Range, Neutron N.A. R(6) Q 1,2 Flux, High Negative Rate

5. Intermediate Range, S R(6,8) S/U(17) 1, 2, and" Neutron Flux

6. Source Range, Neutron S R(6,14) M(14) and 2(7), 3(7), 4 and 5 Flux S/U(I)

7. Overtemperature delta T S R(9) SA 1,2

8. Overpower delta T S R(9) SA 1,2

9. Pressurizer Pressure - S R SA 1,2 Low

10. Pressurizer Pressure - S R SA 1,2 High

11. Pressurizer Water Level - S R SA 1,2 High

12. Loss of Flow-Single Loop S R(8) SA I COOK NUCLEAR PLANT-UNIT I Page 3/4 3-12 AMENDMENT 4-00, 4-2,0, 4-124, -144,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.3 INSTRUMENTATION TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODE IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

13. Loss of Flow-Two Loops S R(8) N.A. I

14. Steam Generator Water S R SA 1,2 I Level -- Low-Low 1,2

15. Steam/Feedwater Flow Mismatch and Low Steam S R SA I

Generator Water Level

16. Undervoltage-Reactor N.A. R M 1 Coolant Pumps

17. Underfrequency-Reactor N.A. R M 1 Coolant Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U(1) 1,2 B. Turbine Stop Valve N.A. N.A. S/U(1) 1,2 Closure

19. Safety Injection Input from N.A N.A. SA (4)(15) 1,2 ESF

20. Reactor Coolant Pump N.A. N.A. R N A.

Breaker Position Trip

21. Reactor Trip Breaker A. Shunt Trip Function N.A. N.A 4 Months (5)(11) 1, 2, 3*, 4%, 5*

and S/U(1)(1 1) I B. Undervoltage Trip N.A. N.A. 4 Months (5)(11) 1,2,3*, 4, 5 I

Function

22. Automatic Trip Logic

23. Reactor Trip Bypass N.A N.A.

N.A.

and S/U(1)(11)

SA(15) 4 Months (5)(12) 1, 2, 3*, 4*, 5*

1, 2, 3*, 4", 5" L

Breaker and SIU(1)(13)

AMENDMENT M0, -20,-144, 1 COOK NUCLEAR PLANT-UNIT 1 Page 314 3-13

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-1 (Continued)

NOTATION

- With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) - If not performed in previous 7 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) - Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate if absolute difference greater than or equal to 3 percent.

(4) - Manual ESF functional input check every 18 months.

(5) - Each train tested at least every other 62 days.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

(8) - The provisions of Specification 4.0.4 are not applicable.

(9) - The provisions of Specification 4.0.4 are not applicable for f1 (delta I) and f2 (delta I) penalties, or for measurement of delta T. (See also Table 2.2-1).

(10) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s)

(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12) - Local manual shunt trip prior to placing breaker in service.

(13) - Automatic Undervoltage Trip.

(14) - The provisions of Specification 4.0.4 are not applicable when leaving MODE 1. In such an event, the calibration and/or functional test shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after leaving MODE I.

(15) - Each train tested at least every other 92 days.

(16) - Not Used (17) - If not performed in previous 184 days COOK NUCLEAR PLANT-UNIT I Page 3/4 3-14 AMENDMENT 99, 4!20, M4,I

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

1. SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Manual Initiation --- - - See Functional Unit 9

b. Automatic Actuation Logic N.A. N.A SA (2) N.A. 1,2,3,4

c. Containment Pressure- High S R SA (3) N.A. 1,2,3

d. Pressurizer Pressure--Low S R SA N.A. 1,2,3

e. Differential Pressure S R SA N.A 1,2,3 Between Steam Lines- High

f. Steam Line Pressure--Low S R SA N.A. 1,2,3

2. CONTAINMENT SPRAY a Manual Initiation See Functional Unit 9-----------

b. Automatic Actuation Logic N.A. N.A. SA(2) N.A. 1,2,3,4 C. Containment Pressure S R SA (3) N.A. 1,2, 3 High-High COOK NUCLEAR PLANT-UNIT I Page 3/4 3-31 AMENDMENT 4W0, 420, 4=- 444, 4-3, 244,

314 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual See Functional Unit 9- - -

2) From Safety Injection Automatic Actuation N.A. N.A. SA (2) N.A. 1,2,3,4 I

Logic

b. Phase "B" Isolation I) Manual See Functional Unit 9------ ----- -

2) Automatic Actuation N.A. N.A. SA (2) N.A. 1,2,3,4 Logic

3) Containment Pressure S R SA (3) N.A. 1,2,3 High-High

c. Purge and Exhaust Isolation

1) Manual See Functional Unit 9 -.. . ..-------- .

2) Containment S R Q N.A. 1,2,3,4 Radioactivity--High COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-32 AMENDMENT 4W0, .144, 41.3, 483,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

4. STEAM LINE ISOLATION

a. Manual See Functional Unit 9-----

b. Automatic Actuation Logic N.A. N.A. SA (2) N.A. 1,2,3,

c. Containment Pressure S R SA (3) N.A. 1,2,3 High-High d Steam Flow in Two Steam S R SA N.A. 1,2,3 Lines--High Coincident with T3vf-Low-Low

e. Steam Line Pressure-Low S R SA N.A 1,2,3

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water S R SA N.A. 1,2,3 Level--High-High

6. MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water Level--Low-Low S R SA N.A 1,2,3

b. 4 kv Bus Loss of Voltage S R M N.A. 1,2,3

c. Safety Injection N A. N.A. SA (2) N.A. 1,2,3

d. Loss of Main Feed Pumps N.A. N.A. R N.A 1,2 COOK NUCLEAR PLANT-UNIT 1 Page 314 3-33 AMENDMENT ..00, 42, =.-, 4.44 453, 2U,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water S R SA N A. 1,2,3 Level-Low-Low

b. Reactor Coolant Pump Bus N.A. R M N A. 1,2,3 Undervoltage

8. LOSS OF POWER

a. 4 kv Bus Loss of Voltage S R M N.A. 1,2,3,4

b. 4 kv Bus Degraded Voltage S R M N.A. 1,2,3,4 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-33a AMENDMENT 41-0, 4,70, 4.44, 4,54

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED

9. Manual

a. Safety Injection (ECCS) N.A. N.A. N.A. R 1, 2, 3, 4 Feedwater Isolation Reactor Trip (SI)

Containment Isolation Phase "A" Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System

b. Containment Spray N.A. N.A N.A. R 1, 2, 3, 4 Containment Isolation Phase "B" Containment Purge and Exhaust Isolation c Containment Isolation- N.A. N.A. N.A. R 1, 2, 3, 4 Phase "A" Containment Purge and Exhaust Isolation

d. Steam Line Isolation N.A. N.A Q R 1, 2, 3

e. Containment Air N.A. N.A N.A. R 1, 2, 3, 4 Recirculation Fan

10. CONTAINMENT AIR RECIRCULATION FAN a Manual -- ----- See Functional Unit 9 ------------

b. Automatic Actuation Logic N.A. N.A SA (2) N.A. 1, 2, 3 c Containment Pressure - High S R SA (3) N.A. 1, 2, 3 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-33b AMENDMENT 4-53, 2-204, 2M,34

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

TABLE NOTATION (1) Deleted (2) Each train or logic channel shall be tested at least every other 92 days.

(3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-34 AMENDMENT 34, =0,

ATTACHMENT 2B to AEP:NRC:3311 UNIT 2 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED REVISED PAGES UNIT 2 1-10 3/4 3-4 3/43-7 3/4 3-11 3/4 3-12 3/4 3-13 3/4 3-30 3/4 3-31 3/4 3-32 3/4 3-33

1.0 DEFINITIONS I TABLE 1.2 FREOUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months w At least once per 7 days M At least once per 31 days Q At least once per 92 days 4 Months At least once per 124 days SA At least once per 184 days R At least once per 549 days S/U Prior to each reactor start-up P Completed prior to each release N.A. Not Applicable COOK NUCLEAR PLANT-UNIT 2 Page 1-10 AMENDMENTS41,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3-1 (Continued)

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

16. Undervoltage-Reactor Coolant 4-1/bus 2 3 1 Pumps

17. Underfrequency-Reactor 4-1/bus 2 3 1 Coolant Pumps

18. Turbine Trip A. Low Fluid Oil Pressure 3 2 2 1 B. Turbine Stop Valve 4 4 3 1 6#

Closure

19. Safety Injection Input from 2 2 1,2 ESF

20. Reactor Coolant Pump Breaker Position Trip Above P-7 1/breaker 2 1/breaker per 1 11 operating loop

21. Reactor Trip Breakers 2 2 1,2 13, 15 3*, 4*, 5* 14

22. Automatic Trip Logic 2 2 1,2 1 3*, 4*, 5* 14 Page 3/4 3-4 AMENDMENT 86, 107, 42.7, J2, NUCLEAR PLANT-UNIT 2 AMENDMENT 86, 4107, 4-2-7, =17, COOK NUCLEAR COOK PLANT-UNIT 2 Page 314 3-4

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3-1(Continued)

ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 13 - With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 1. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

ACTION 15 - With the number of OPERABLE Reactor Trip Breaker channels one less than required by the Minimum Channels OPERABLE requirement for reasons other than an inoperable diverse trip feature, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . One channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION P-6 With 2 of 2 Intermediate Range P-6 prevents or defeats the Neutron Flux Channels < 6 X 10" amps. manual block of source range reactor trip.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-7 AMENDMENT 86, 4=-,

) TABLE 4.3-1 0

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

z FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip A. Shunt Trip Function N.A. N.A. S/U(I)(10) 1, 2, 3, 4, 5" B. Undervoltage Trip Function N.A. N.A. S/U(1)(10) 1, 2, 3% 4, 5°

2. Power Range, Neutron Flux S D(2,8), M(3,8), Q and S/U(1) 1, 2 and' and Q(6,8)

3. Power Range, Neutron Flux, High Positive Rate N.A. R(6) Q 1,2

4. Power Range, Neutron Flux, High Negative Rate N.A. R(6) Q 1,2

5. Intermediate Range, Neutron Flux S R(6,8) S/U(17) 1, 2, and*

6. Source Range, Neutron Flux S R(6,14) M(14) and S/U(1) 2(7), 3(7), 4 and 5

7. Overtemperature AT S R(9) SA 1,2 8 Overpower AT S R(9) SA 1,2

9. Pressurizer Pressure -- Low S R SA 1,2

10. Pressurizer Pressure-- High S R SA 1,2 2 11. Pressurizer Water Level -- High S R SA 1,2

12. Loss of Flow-Single Loop S R(8) SA 1 I

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS z CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

13. Loss of Flow-Two Loops S R(8) N.A. 1 z S R SA 1,2

14. Steam Generator Water Level -- Low-Low

15. Steam/Feedwater Flow Mismatch and Low Steam S R SA "1,2 Generator Water Level

16. Undervoltage-Reactor Coolant Pumps N.A. R M 1

17. Underfrequency-Reactor Coolant Pumps N.A. R M 1

18. Turbine Trip 1,2 IZ A. Low Fluid Oil Pressure B. Turbine Stop Valve Closure N.A.

N.A.

SIU(1)

S/U(1) 1,2 N.A. N.A. SA (4)(15) 1,2

19. Safety Injection Input from EFS z.

20. Reactor Coolant Pump Breaker Position Trip N.A. N.A. R N.A.

21. Reactor Trip Breaker N.A. 4 Months (5)(11) and 1,2,33, 4 , 5*

A. Shunt Trip Function N.A.

SJU(1)(1 1) 1, 2,3",4",5 B. Undervoltage Trip Function N.A N.A. 4 Months (5)(11) and 1, 2, 3*, 4*, 5*

SIU(1)(I 1)

22. Automatic Trip Logic N.A N.A. SA (15) 1, 2, 3*, 4, 5*

23. Reactor Trip Bypass Breaker N.A. N.A. 4 Months (5)(12) and SIU(1)(13)

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-1 (Continued)

NOTATION

- With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) - If not performed in previous 7 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) - Compare incore to excore axial offset above 15% of RATED THERMAL POWER. Recalibrate if absolute difference greater than or equal to 3 percent.

(4) - Manual ESF functional input check every 18 months (5) - Each train tested at least every other 62 days.

I (6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

(8) - The provisions of Specification 4.0.4 are not applicable.

(9) - The provisions of Specification 4.0.4 are not applicable for f, (delta I) and f2 (delta I) penalties, or for measurement of delta T. (See also Table 2.2-1).

(10) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit(s)

(11) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers (12) - Local manual shunt trip prior to placing breaker in service.

(13) - Automatic Undervoltage Trip.

(14) - The provisions of Specification 4.0.4 are not applicable when leaving MODE 1. In such an event, the calibration and/or functional test shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after leaving MODE 1.

(15) - Each train tested at least every other 92 days (16) - Not Used.

(17) - If not performed in previous 184 days.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-13 AMENDMENT 86, 40-7, 28,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4 3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a Manual Initiation

- See Functional Unit 9 b Automatic Actuation N A NA SA(2) NA. 1,2,3,4 Logic c Containment Pressure S R SA (3) N A. 1,2,3 High d Pressurizer Pressure S R SA N A. 1,2,3 Low e Differential Pressure S R SA N A. 1,2,3 Between Steam Lines High f Steam Line Pressure - S R SA NA 1,2,3 Low 2 CONTAINMENT SPRAY a Manual Initiation See Functional Unit 9 b Automatic Actuation N A N A SA(2) NA 1,2,3,4 Logic c Containment Pressure S R SA (3) NA 1,2,3 High-High 3 CONTAINMENT ISOLATION a Phase "A" Isolation

1) Manual - --- See Functional Unit 9

2) From Safety N A N A SA (2) N A 1,2,3,4 Injection Automatic Actuation Logic b Phase "B" Isolation I) Manual --- - - See Functional Unit 9

2) Automatic Actuation N A NA SA (2) N A 1,2,3,4 Logic

3) Containment S R SA (3) NA 1,2,3 Pressure- High High COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-30 AMENDMENT 34, 4-34, 437, 4-58, 2-24,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REQUIRED c Purge and Exhaust Isolation I) Manual _ _- See Functional Unit 9

2) Containment S R Q NA. 1,2,3,4 Radioactivity - High 4 STEAM LINE ISOLATION a Manual - - -See Functional Unit 9 b Automatic Actuation NA N.A SA (2) N A. 1,2,3 Logic c Containment Pressure S R SA (3) N A. 1,2,3 High-High d Steam Flow in Two Steam S R SA N A. 1,2,3 Lines - High Coincident with T., - Low-Low SA N A. 1,2,3 I e Steam Line Pressure S R Low 5 TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water S R SA NA 1, 2,3 Level - High-High 6 MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a Steam Generator Water S R SA NA 1,2,3 Level - Low-Low b 4 kV Bus Loss of Voltage S R M NA 1, 2,3

c. Safety Injection NA N.A SA (2) NA 1,2,3 d Loss of Main Feed Pumps NA N.A R NA 1,2 COOK NUCLEAR PLANT-UNIT 2 Page 314 3-31 AMENDMENT 82,9-7, 4-34, 4-34, 4-3-7, 4-59, 468, 2,24,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST REOUIRED

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMP a Steam Generator Water S R SA NA 1,2,3 Level - Low-Low b Reactor Coolant Pump Bus NA R M NA 1,2,3 Undervoltage 8 LOSS OF POWER a 4 kv Bus Loss of Voltage S R M N A. 1,2,3,4 b 4 kv Bus Degraded S R M NA. 1,2,3,4 Voltage 9 MANUAL a Safety lnjection (ECCS) NA NA NA R 1,2,3,4 Feedwater Isolation Reactor Trip (SI)

Containment Isolation Phase "A" Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System b Containment Spray NA NA N A. R 1,2,3,4 Containment Isolation Phase "B" Containment Purge and Exhaust Isolation c Containment Isolation - NA N A. NA R 1,2,3,4 Phase "A" Containment Purge and Exhaust Isolation

d. Steam Line Isolation N A. NA Q R 1,2,3 e Containment Air NA NA NA R 1,2,3,4 Recirculation Fan 10 CONTAINMENT AIR RECIRCULATION FAN a Manual ------- See Functional Unit 9 b Automatic Actuation NA N A. SA (2) NA 1,2,3 Logic c Containment Pressure- S R SA (3) NA 1,2,3 High COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-32 AMENDMENT 82,97,-34,-137,

-159,-1-89, 2p-,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

TABLE NOTATION (1) Deleted (2) Each train or logic channel shall be tested at least every other 92 days (3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-33 AMENDMENT 82, 34, 4-89,

ATTACHMENT 3 TO AEP:NRC:3311 I&M RESPONSE TO NRC CONDITIONS AND LIMITATIONS to AEP:NRC:3311 Page I I&M RESPONSE TO NRC CONDITIONS AND LIMITATIONS By Reference 1, Indiana Michigan power Company (I&M) proposed to revise the Technical Specification (TS) reactor trip system (RTS) and engineered safety features actuation system (ESFAS) surveillance requirements based on the evaluation in topical report WCAP-15376-P, Revision 0, "Risk-Informed Assessment of the RPS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." The Nuclear Regulatory Commission (NRC) issued by Reference 2, a Safety Evaluation (SE) accepting WCAP-15376-P as a generic basis for licensees to reference in licensing applications. To incorporate the surveillance test interval extensions, the NRC required that an applicant for a TS amendment meet certain conditions and limitations stipulated in the SE. These conditions and limitations are addressed below for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2.

NRC Condition and Limitation 1 A licensee is expected to confirm the applicability of the topical report to their plant, and to perform a plant-specific assessment of containment failures and address any design or performance differences that may affect the proposedchanges.

I&M Response to NRC Condition and Limitation 1 Applicability of WCAP-1 5376-P to CNP In anticipation of Condition and Limitation 1 and as presented in Enclosure 2 to Reference 1, I&M has assessed the applicability of the topical report to CNP. Following guidelines and methodology previously used in the evaluation of a similar NRC-approved license amendment for the Surry Power Station, I&M completed assessments to confirm that WCAP-15376-P is applicable to the design and operation of CNP.

In Reference 1, I&M identified the licensee conditions that were anticipated for proposed requests to adopt the TS changes evaluated in WCAP-15376-P. The anticipated conditions were based on review of NRC safety evaluations for similar risk-informed topical reports, specifically WCAP-10271-P-A, including Supplements 1 and 2, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation," and WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times."

Anticipated Condition 2, on Page 18 of Enclosure 2 to Reference 1, was to confirm the applicability of the WCAP-15376-P analyses for the plant. In response to Anticipated Condition 2, I&M provided, in Attachment 3 to Reference 1, three tables listing the important parameters and assumptions made in the generic analysis that are relevant to the WCAP-1 5376-P to AEP:NRC:3311 Page 2 evaluation of changes to the ESFAS/RTS surveillance test intervals (STI), completion times (CT), and bypass times (BT). The tables also provided information for the current calculated core damage frequency (CDF) and the contribution to CDF from the anticipated transient without scram (ATWS) events.

The information provided in the tables demonstrates the applicability of the generic WCAP-15376-P analysis to the CNP system configuration and probabilistic risk assessment (PRA) model. The tables were based on implementation guidelines that were issued by the Westinghouse Owner's Group (WOG) for licensees implementing the TS allowed outage time changes justified in WCAP-14333-P-A, Revision 1. For each key parameter in the tables, the value assumed in the WCAP-15376-P analysis was compared to the CNP-specific value. The table identifies that, for each parameter, CNP design and operation is consistent with the WCAP-15376-P assumptions since the RTS and ESFAS signals are available. Based on these tables, I&M confirmed the applicability of the WCAP-15376-P topical report to the CNP PRA assessment. Another licensee received NRC approval of RPS and ESFAS surveillance frequency extension, in an earlier licensing activity, by demonstrating the plant-specific applicability of WCAP-14333-P-A using the WOG implementation guidelines.

On June 16, 2000, Virginia Electric and Power Company (VEPCO) submitted a license amendment request (Reference 3) to extend the reactor protection system (RPS) and ESFAS analog instrumentation surveillance frequency from monthly to quarterly for Surry Power Station, Units 1 and 2. Reference 3 included three implementation tables provided by the WOG to demonstrate the applicability of the WCAP-14333-P-A generic criteria to the Surry Power Station. Additional plant-specific risk assessment was only required for those changes that were outside the scope of generic WCAP-14333-P-A. In the safety evaluation report approving VEPCO's license amendment (Reference 4), the NRC concluded that the licensee provided sufficient information in its submittal to confirm the plant-specific applicability of the topical reports. The NRC further stated:

Based on the risk insights gained from the generic analyses and the information provided by the licensee, the staff finds the licensee's conclusion credible and reasonable and meets the intent of RG 1.177 and RG 1.174. The design features and operational practices listed in the submittal are generally consistent with the generic plant modeled as the topical reports. The staff finds that the topical reports are applicable for Surry.

I&M incorporated CNP-specific information into the WOG implementation tables found acceptable by the NRC in VEPCO's RPS and ESFAS surveillance frequency extension. The modified tables that demonstrate the applicability of WCAP-15376-P to CNP are provided in of Reference 1.

I&M has reviewed the Reference 1, Attachment 3 information demonstrating the applicability of WCAP-15376-P to CNP since receipt of the SE provided in Reference 2. The modified to AEP:NRC:3311 Page 3 implementation tables submitted in Attachment 3 to Reference I confirm the plant-specific applicability of WCAP-15376-P for this amendment request.

Assessment of Containment Failures I&M has prepared the following assessment of containment failures to address any design or performance differences that may affect the proposed RTS/ESFAS STI, CT and BT TS changes.

This risk assessment focuses on the postulated differences in performance between the CNP ice condenser containment design and the large, dry design that is the basis for the containment risk assessment contained in WCAP-15376-P.

The CNP-specific value for large early release frequency (LERF) is 5.59E-06 per year for both Units 1 and 2. The CNP-specific CDF values are 4.85E-05 per year for Unit 1 and 4.87E-05 per year for Unit 2. These values indicate that about 11.5 percent of CDF sequences progress to LERF for CNP. By comparison, Table 8.29 of WCAP-15376-P shows a CDF value of 5.05E-05 per year, while Table 8.32 of WCAP-15376-P presents a LERF value of 2.38E-06 per year.

These values indicate that about 4.7 percent of CDF sequences progress to LERF for the WCAP-15376-P PRA model.

An estimate of CNP-specific LERF increases attributable to the proposed TS changes may be made by applying the CNP-specific LERF-to-CDF factor to the CDF results provided in WCAP-15376-P. The cumulative effects of the STI and CT changes proposed in WCAP-15376-P result in CDF increases of 5.7E-07 per year for two-out-of-four (2/4) logic and 1.lE-06 per year for two-out-of-three (2/3) logic, as shown in Table 8.33. The cumulative effects of the STI and CT changes proposed in WCAP-15376-P on LERF for CNP may be estimated by multiplying these values by the derived CNP-specific LERF-to-CDF factor. Applying the 11.5 percent factor to Table 8.33 CDF increases yields CNP-specific estimates for LERF increases of 6.6E-08 per year for 2/4 logic and 1.3E-07 per year for 2/3 logic. These estimated increases in LERF are comparable to, or below, the Regulatory Guide (RG) 1.174 acceptance guidance of 1.OE-07 per year for very small risk increases.

An estimate of CNP-specific incremental conditional large early release probability (ICLERP) increases attributable to the proposed TS changes may be made by applying the derived CNP specific LERF-to-CDF factor to the incremental conditional core damage probability (ICCDP) results provided in Enclosure 2 to Reference 1. The reactor trip breaker (RTB) CT change proposed in WCAP-15376-P was estimated to result in an ICCDP increase of 7.4E-08. The RTB BT change proposed in WCAP-15376-P was estimated to result in an ICCDP increase of 9.9E

09. The effects of the RTB CT and BT changes proposed in WCAP-15376-P on ICLERP for CNP may be estimated by multiplying these estimated ICCDP values by the CNP-specific LERF to-CDF factor. Applying the 11.5 percent factor to the estimated ICCDP increases yields CNP specific estimates for ICLERP increases of 8.5E-09 for the proposed CT change and 1.1 E-09 for to AEP:NRC:3311 Page 4 the proposed BT change. Both of these estimated increases in ICLERP are below 5.OE-08, which is considered very small for a single TS CT change per RG 1.177.

Conclusion I&M assessments of the topical report and of postulated containment failures confirm that WCAP-15376 is applicable to the design and operation of CNP.

NRC Condition and Limitation 2 Address the Tier 2 and Tier 3 analyses including risk significant configuration insights and confirm that these insights are incorporated into the plant-specific configuration risk managementprogram.

I&M Response to NRC Condition and Limitation 2 The Tier 2 requirements of RG 1.177 state that the licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur when specific plant equipment is out of service consistent with the proposed TS change. The Tier 3 requirements of RG 1.177 require the licensee to develop a program that ensures that the risk impact of out-of service equipment is appropriately evaluated prior to performing any maintenance activity.

Tier 2 identifies and evaluates any potential risk-significant plant equipment outage configurations associated with the proposed change. RG 1.177 requirements for Tier 2 state that licensees should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur when equipment associated with the proposed TS change is out-of service.

Tier 3 provides for the establishment of an overall Configuration Risk Management Program (CRMP) and confirmation that its insights are incorporated into the decision-making process before taking equipment out-of-service prior to or during the completion time. Compared to Tier 2, Tier 3 provides additional coverage on the basis of any additional risk-significant configurations that may be encountered during maintenance scheduling over extended periods of plant operation. Tier 3 guidance is satisfied by 10 CFR 50.65(a)(4) which requires a licensee to assess and manage the increase in risk that may result from activities such as surveillance, testing, and corrective and preventive maintenance.

The Tier 2 and Tier 3 requirements of RG 1.177 have been addressed at CNP. CNP currently has in place a risk-informed on line and shutdown risk management process, to support the requirements of 10 CFR 50.65(a)(4). The Scientech Safety MonitorTM computer program is used for online risk assessment (Modes 1, 2, and 3) while the Electric Power Research Institute Outage Risk Assessment and Management (ORAMTm) computer program is used for shutdown to AEP:NRC:3311 Page 5 risk assessment (Modes 4, 5, and 6). This risk-informed process is implemented and governed by plant procedures. These procedures assure that the risk associated with the various plant configurations planned during at-power or shutdown conditions are assessed prior to entry into these configurations and appropriately managed while the plant is in these various configurations.

WCAP-15376-P provides the following Tier 2 insights which are applicable to the proposed STI extension at CNP:

- Activities that degrade the availability of auxiliary feedwater, reactor coolant system (RCS) pressure relief, ATWS mitigating system actuation circuitry (AMSAC), or turbine trip should not be scheduled when an RTB is out of service.

- Activities that could degrade the operable train of RPS including master relays, slave relays, and analog channels should not be scheduled concurrently with the out of service train.

- Activities on electrical support systems for auxiliary feedwater, RCS pressure relief, AMSAC, or turbine trip should not be scheduled during RTB maintenance.

I&M will implement administrative controls in the CNP CRMP to include the above restrictions when an RTB and/or logic cabinet is removed from service.

NRC Condition and Limitation 3 The risk impact of concurrent testing of one logic cabinet and associated reactor trip breaker needs to be evaluated on a plant-specific basis to ensure conformance with the WCAP-153 76-P, Rev. 0 evaluation,and RGs 1.174 and 1.177 guidance.

I&M Response to NRC Condition and Limitation 3 The risk associated with concurrent testing has been found to be within acceptability guidelines of WCAP-15376-P. Concurrent testing is performed on a logic cabinet and associated RTB with the use of the reactor trip bypass breaker (RTBB). The RTBB receives a trip signal from the opposite train RPS thereby maintaining trip capability. The length of time concurrent testing is performed within the out-of-service times for the RTB is consistent with the assumptions in WCAP- 15376-P.

I&M has prepared the following plant-specific risk assessment for the situation in which concurrent testing of one logic cabinet, and its associated RTB, is performed. This risk assessment includes an estimate of the ICCDP and ICLERP for this testing configuration by using methods and values obtained from the WOG response (Reference 5) to NRC request for additional information (RAI) #4 for WCAP-15376-P and Enclosure 2 to Reference 1. The to AEP:NRC:3311 Page 6 assessment uses the same CNP LERF-to-CDF factor of 11.5 percent as discussed in the response to NRC Condition and Limitation 1.

An estimate of the CNP-specific ICCDP increase attributable to the potential concurrent testing configuration may be made in a similar manner to the approach used in Enclosure 2 to Reference 1. Using the conditional CDF of 1.46E-04 per year provided in Reference 5 for the test configuration, the CNP-specific value for both units of 4.9E-05 per year and a bypass time for testing of 4-hours, an ICCDP for the year (8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months ) can be calculated:

ICCDP = (1.46E-04/yr - 4.9E-05/yr) x 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months / (8760 hrs/yr) = 4.4E-08.

Both the calculated value of 4.4E-08, assuming concurrent testing of both the RTB and logic cabinet, and the CNP-specific value of 9.9E-09 that was calculated in Enclosure 2 to Reference 1, assuming that only the RTB would be unavailable due to testing, are within the ICCDP guideline in RG 1.177 of 5.OE-07.

The corresponding ICLERP may be estimated from this ICCDP value by multiplying the CNP specific ICCDP value by the derived CNP-specific LERF-to-CDF factor. Applying the 11.5 percent factor to the ICCDP value of 4.4E-08 yields a CNP-specific estimate for the ICLERP increase of 5.1E-09. This estimated increase in ICLERP for the concurrent testing of a logic cabinet and its associated RTB is also below the acceptance guidance of 5.OE-08 provided in RG 1.177.

NRC Condition and Limitation 4 To ensure consistency with the reference plant, the model assumptionsfor human reliability in WCAP-15376-P, Rev. 0 should be confirmed to be applicable to the plant-specificconfiguration.

I&M Response to NRC Condition and Limitation 4 The general model assumptions for human reliability for the WCAP-15376-P PRA model are delineated in Section 8.1.2, "Representative PRA Model," as:

- The PSA model must allow for crediting operator actions to actuate the safety systems if the automatic signals fail. The model must also be able to account for the dependencies of subsequent operator actions on previous operator actions.

- The plant needs to have available procedures that direct the operators to trip the plant and respond to an ATWS event if the automatic action fails.

- Events also credit an operator action, as appropriate, to initiate safety injection via the SI switch in the control room.

to AEP:NRC:3311 Page 7 WCAP-15376-P, Table 8.28, "Summary of Human Error Probabilities for Operator Actions Backing Up Actuation Signals," identifies the following six specific manual actions to "back-up" the automatic systems that meet the intent of these general model assumptions. These actions are:

I. Reactor trip from the main control board trip switches

2. Reactor trip by interrupting power from the motor-generator sets

3. Manually inserting the control rods into the core

4. Safety injection from the main control board switches

5. Safety injection by manual actuations of individual components

6. Auxiliary feedwater pump start Each of these manual actions are included in CNP emergency operating procedures. In addition, I&M has determined that the CNP PRA model completely satisfies the model assumptions for four actions and partially satisfies the assumptions for the remaining two actions. The modeling of the first three operator actions is described under the manual rod insertion (MRI) operator action in CNP's Human Reliability Analysis notebook. Review of the ATWS event tree identified that operator actions following MRI have different values depending on whether MRI succeeded or failed. This observation confirms that dependencies are accounted for in subsequent operator actions.

The fourth and fifth operator actions listed in Table 8.28 are not explicitly modeled in the CNP PRA model. However, they are implicitly included in the CNP PRA model for the RPS/ESFAS.

This system model was developed at a high level from the more detailed RPS/ESFAS model developed by Westinghouse for another Westinghouse four-loop plant. The PRA model of the reactor trip and engineered safety feature signals are consistent with the evaluation of RPS unavailability performed for the TS Optimization Programs study as documented in WCAP-10271-P-A. The CNP PRA modules use signal unavailability values quantified for another Westinghouse four-loop plant. The use of these values for the CNP PRA model is conservative since the unavailablilities of the logic implementations are not significantly different and it takes fewer analog channel failures to cause signal failure for the four-loop plant that values were quantified for, than it does at CNP.

The sixth operator action is addressed in multiple procedures and human error calculations. The human error probabilities calculated for this action are less than the values used in the WCAP-15376-P PRA for corresponding actions.

In conclusion, the CNP emergency operating procedures and PRA model comply with the PRA human reliability assumptions identified in WCAP-1 5376-P.

Attachment 3 to AEP:NRC:3311 Page 8 NRC Condition and Limitation 5 For future digital upgrades with increased scope, integration and architectural differences beyond that of Eagle 21, the stafffinds the generic applicability of WCAP-15376-P, Rev. 0 to future digitalsystems not clearand should be consideredon a plant-specificbasis.

I&M Response to NRC Condition and Limitation 5 I&M incorporated Foxboro Spec 200 equipment as part of a 1994 digital upgrade of the signal conditioning and comparator circuits of the RTS and ESFAS analog channels. Reference 6 provided NRC approval of the digital upgrade project and concluded that the digital upgrade to the RTS is consistent with the current plant design and licensing basis and will provide appropriate RTS function as required. Reference 6 acknowledged that the new system processes the same inputs as the previous analog system, performs the same calculation and bistable functions, and supplies contact outputs to the reactor protection logic for initiating a reactor trip and for engineered safety feature functions. Reference 6 also noted that the system includes isolated analog outputs to indicators, recorders, plant computer and various systems.

Since CNP uses a Westinghouse design, and the Foxboro Spec 200 equipment is consistent with both the original and current plant design and licensing basis, the CNP RTS and ESFAS do not represent equipment that is of significant increased scope, integration or architectural differences than that considered generically in WCAP-15376-P. I&M has no plans for significant protection system equipment upgrades at this time; however, administrative controls will be implemented to ensure any future digital upgrades to the RTS and/or ESFAS are evaluated to ensure the generic applicability of WCAP- 15376-P is not affected.

References:

1. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Amendment Request to Extend Reactor Trip System and Engineered Safety Features Actuation System Surveillance Requirements as Evaluated in WCAP-15376," AEP:NRC:23 11, dated August 30, 2002

2. Letter from W. H. Ruland, NRC, to R. H. Bryan, WOG, "Acceptance for Referencing of Topical Report WCAP-15376-P, Rev. 0, 'Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times' (TAC. No.

MB0983)," dated December 20, 2002

3. Letter from W. R. Matthews, VEPCO, to NRC Document Control Desk, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed Technical Specifications to AEP:NRC:3311 Page 9 and Bases Change - RPS and ESFAS Analog Instrumentation Surveillance Frequency Change from Monthly to Quarterly," dated June 16, 2000

4. Letter from G. E. Edison, NRC, to D. A. Christian, VEPCO, "Surry Units 1 and 2 - Issuance of Amendments Re: Changes to Surveillance Test Intervals and Allowed Outage Times for Instrumentation Systems (TAC Nos. MA9355 and MA9356)," dated August 31, 2001

5. Letter from R. H. Bryan, WOG, to NRC Document Control Desk, "Transmittal of Response to Request for Additional Information (RAI) Numbers 4 and 11 Regarding WCAP-15376-P, Rev. 0, 'Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Inervals and Reactor Trip Breaker Test and Completion Times' (MUHP-3046)," dated January 8, 2002

6. Letter from J. B. Hickman, NRC, to E. E. Fitzpatrick, I&M, "Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 - Revised Safety Evaluation for Amendment Nos. 175 and 160 Re:

Reactor Protection System Upgrade Project (TAC Nos. M84839 and M84840)," dated May 13, 1994

ATTACHMENT 4 TO AEP:NRC:3311 BASIS FOR PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS to AEP:NRC:3311 Page I BASIS FOR PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS By Reference 1, Indiana Michigan Power Company (I&M) proposed to revise the Technical Specification (TS) reactor trip system (RTS) and engineered safety features actuation system (ESFAS) surveillance requirements based on the evaluation in WCAP-15376-P, Revision 0, "Risk-Informed Assessment of the RPS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." On December 20, 2002, the Nuclear Regulatory Commission (NRC) issued a Safety Evaluation (SE), Reference 2, accepting WCAP-15376-P as a generic basis for licensees to reference in license amendment requests. Technical Specification Task Force (TSTF) Traveler TSTF-4 11, Revision 1, "Surveillance Test Interval Extension for Components of the Reactor Protection System," provides marked-up pages of NUREG-1431, Revision 2, "Standard Technical Specifications Westinghouse Plants," (ISTS) as evaluated by WCAP- 15376-P. to Reference 1 justified each individual TS change based on whether the change was supported by the evaluation in WCAP-1 5376-P, or some other technical basis.

Subsequent to receipt of Reference 2, I&M re-evaluated each of the individual TS changes proposed by Reference 1 to distinguish between those specifically evaluated by WCAP-1 5376-P, and those that were proposed as enhancements that were not bounded by WCAP-15376-P. The purpose of this review was to expedite the NRC's review of the proposed TS changes by identifying and withdrawing those changes that were not evaluated by WCAP-15376-P. to this letter withdraws the proposed changes determined not to be evaluated by WCAP- 15376-P.

This attachment expands on the tables provided in Reference 1, Attachment 4, by providing additional justification for each individual TS change not withdrawn, and indicating the basis for determining that each change is consistent with, and bounded by, the approved changes in TSTF-41 1, Revision 1, as evaluated by WCAP-15376-P and found acceptable for use by Reference 2. Each individual TS change is assigned a sequential "Change Number" (Column 1).

The change location is identified under the "Unit/TS Page" (Column 2) heading, and a brief description of the change is presented under "Proposed Change" (Column 3). The "Basis for the Proposed Change" (Column 4) presents the technical/regulatory rationale supporting the change, including references to the corresponding ISTS functions and surveillance tests, and the determination that the change is consistent with, and bounded by, TSTF-4 11, Revision 1 and the SE. This determination is repeated in the last two columns.

Discussion of "Staggered Test Basis" Upon review of TSTF-4 11, differences between the ISTS definition of "STAGGERED TEST BASIS" and the Donald C. Cook Nuclear Plant (CNP) TS definition of "STAGGERED TEST BASIS" were noted. The following provides clarifying information about the differences and the resolution of the differences as proposed in Reference 1.

to AEP:NRC:3311 Page 2 Several changes proposed in TSTF-411 included the incorporation of the term "STAGGERED TEST BASIS" into the testing frequency. For example, Surveillance Requirement 3.3.1.4 requires the performance of a surveillance on a frequency of "62 days on a STAGGERED TEST BASIS." However, CNP's TS Tables 4.3-1 and 4.3-2, which specify the surveillance requirements for the reactor trip system (RTS) and engineered safety features actuation system (ESFAS), currently do not use the same terminology. Instead, these TS Tables typically specify the surveillance frequency, and provide an annotation that "each train is tested at least every other [frequency/2] days." Therefore, to maintain consistent terminology throughout these CNP TS tables, I&M has chosen to annotate the surveillances that are performed on a staggered test basis by providing an annotation similar to that currently provided in the TS, rather than including the term "STAGGERED TEST BASIS" into these surveillances.

To facilitate the review of this information, a list of acronyms used in this attachment is provided.

BT Bypass Time CFT Channel Functional Test CNP Donald C. Cook Nuclear Plant COT Channel Operational Test CT Completion Time ESFAS Engineered Safety Feature Actuation System FU Functional Unit I&M Indiana Michigan Power Company ISTS Improved Standard Technical Specifications M Monthly NRC Nuclear Regulatory Commission Q Quarterly RTB Reactor Trip Breaker RTBB Reactor Trip Bypass Breaker RTS Reactor Trip System S/U Start-Up SA Semi-Annual SE Safety Evaluation SG Steam Generator SI Safety Injection SR Surveillance Requirement TADOT Trip Actuating Device Operational Test TS Technical Specifications TSTF Technical Specifications Task Force UFSAR Updated Final Safety Analysis Report to AEP:NRC:3311 Page 3

References:

1. Letter from J. E. Pollock, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Amendment Request to Extend Reactor Trip System and Engineered Safety Features Actuation System Surveillance Requirements as Evaluated in WCAP-15376," AEP:NRC:231 1, dated August 30, 2002

2. Letter from W. H. Ruland, NRC, to R. H. Bryan, Westinghouse Owners Group, "Acceptance for Referencing of Topical Report WCAP-15376-P. Rev. 0, 'Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times' (TAC. No. MB0983)," dated December 20, 2002

Changes to Technical Specification Table 1.2 0 FREQUENCY NOTATION Is proposed change, consistent with'and Unit TS Page Proposed Change Basis for Proposed Change boun ed.by:

U .TSTF-411, WCAP 0 Revision 1? approval SE?

1. UI Page 1-9 Add new definition The 4 Month frequency is consistent with the TSTF-411 Surveillance Y Y for "4 Months" Frequency of"every 62 days on a STAGGERED TEST BASIS."

U2 Page 1-10 frequency as "At least once per 124 days" (D

-P-1

Changes to Technical Specification Table 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION 0

Is proposed change consistent with and

= Unit/TS Page Proposed Change', Basis for Proposed Change bounded by:

Uz TSTF-411, WCAP Revision 1? approval SE?

2. UI Page 3/4 3-5 FU 21 - Reactor Trip The proposed Action 15 will incorporate a 24-hour CT and an allowance for tII Breakers (RTBs), bypassing one channel for up to four hours for surveillance testing per Y Y U2Page3/43-4 including Reactor TS4.3.1.1.1, provided the other channel is OPERABLE. The proposed Trip Bypass Breakers 4-hour BT is consistent with the Note for ISTS 3.3.1, CONDITION 0, as (RTBBs) presented in TSTF-41 1, Revision I. The proposed 24-hour CT is consistent with ISTS 3.3.1, REQUIRED ACTION 0.1, as presented in TSTF-41 1, Add reference to Revision I, which requires the restoration of an inoperable RTB train to "Action 15 in the OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Therefore, this proposed change is "Action" column, consistent with TSTF-41 I, Revision 1.

3. UI Page 3/4 3-5 FU 21 - RTBs, As discussed above, a new Action 15 is proposed that will incorporate a including RTBBs 24-hour CT and a 4-hour BT for surveillance testing the RTBs, including Y Y U2let Pager e 3/ RTBBs. Consequently, Action 1, which does not allow a CT, and specifies a DAction I in the 2-hour BT for surveillance testing, will no longer apply to Functional Unit 21.

"Action" cointeThe. proposed Action 15 is consistent with TSTF-41 1, Revision 1, as applied "Action" column. to the RTBs, including RTBBs. Therefore, this proposed change is consistent with TSTF-41 I, Revision 1.

4. Ul Page 3/4 3-5 FU 21 - RTBs, This is an editorial change to more clearly indicate that current Action 13 and U2 Page 3/4 3-4 including RTBBs proposed Action 15 only apply to MODES I and 2, and current Action 14 n/a n/a U2ePagec3/afer only applies to MODES 3, 4 and 5. This editorial change provides Delete the comma after consistency between the Unit 1 and Unit 2 TS, and clarifies the TS without "2" in the Applicable haging the intent of the TS.

Modes column (Unit 2 chan only) Per NUREG-1431, Revision 2, Table 3.3.1-1, Function 19, "RTBs," and Function 20, "RTI3 Undervoltage and Shunt Trip Mechanisms," specify Increase the spacing different Actions for MODES I and 2 and MODES 3, 4 and 5. The current bOEte te lines f Required Actions in NUREG-1431 are similar to those in the CNP TS, and MODES I, 2 and MODES 3', 4*, 5*. are unaffected by the changes in TSTF-41 1, Revision 1. Therefore, this proposed change is consistent with NUREG-1431, Revision 2, and is not applicable to TSTF-41 I, Revision 1.

(J1

Changes to Technical Specification Table 3.3-1 >

REACTOR TRIP SYSTEM INSTRUMENTATION "Isproposed change tm " consistent with and t*l

" Unit/TSPage Proposed Change" Basis for Proposed Change bounded by:

ZTSTF-411, WCAP Revision 1? approval SE?

5. U1 Page 3/4 3-5 FU 22 - Automatic This is an editorial change to more clearly indicate that Action 1 applies only U2 Page 3/4 3-4 Trip Logic to MODES 1 and 2, and Action 14 applies only to MODES 3, 4 and 5. This n/a n/a Increase the spacing editorial change clarifies the TS without changing the intent of the TS.

between the lines for Per NUREG-1431, Revision 2, Table 3.3.1-1, Function 21, "Automatic Trip MODES 1, 2 and Logic" specifies different Actions for MODES I and 2 and MODES 3, 4 and MODES 3*, 4*, 5*. 5. The current Required Actions in NUREG-1431 are similar to those in the CNP TS, and are unaffected by the changes in TSTF-4 11, Revision 1.

Therefore, this proposed change is consistent with NUREG-1431, Revision 2, and is not applicable to TSTF-41 1, Revision 1.

6. UI Page 3/4 3-8 ACTION 15 The proposed Action 15 will incorporate a 24-hour CT and an allowance for bypassing one channel for up to four hours for surveillance testing per Y Y U2 Page 3/4 3-7 Add new Action 15 to TS 43.1.1.1, provided the other channel is OPERABLE. The proposed be applied for an RTB 4-hour BT is consistent with the Note for ISTS 3.3.1, CONDITION 0, as channel inoperable presented in TSTF-4 11, Revision 1. The proposed 24-hour CT is consistent condition for reasons with ISTS 3.3.1, REQUIRED ACTION 0.1, as presented in TSTF-411, other than an Revision 1, which requires the restoration of an inoperable RTB train to inoperable diverse trip OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Therefore, this proposed change is feature, and to specify consistent with TSTF-41 1, Revision 1.

actions required when the CT is exceeded.

Changes to Technical Specification Table 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change consistent with and,

= Unit ITS Page Proposed Change Basis forPropsed Change bounded by:

0 "TSTF"411, _ýWCAP

.. Revision 1? approval SE?

7. UI Page 3/4 3-12 FU 2 - Power TSTF-41 1, Revision I evaluates changing the specified frequency of Y Y U)

U2 Page 3/4 3-11 Range, Neutron SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined Flux in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as defined in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 2, Power Range Change the CFT Neutron Flux, which corresponds to the CNP TS requirement to perform a frequency from "M" to monthly CFT for FU 2, Power Range, Neutron Flux. The CNP TS also "Q."I require a quarterly channel calibration for the power range nuclear instrumentation. Because the channel calibration also includes the performance of a CFT, no benefit would be gained by extending the CFT for this instrumentation beyond the quarterly frequency, without a corresponding change to the channel calibration requirements. [Note: The current TS require a monthly comparison of incore and excore axial offset above 15 percent of rated thermal power. This comparison constitutes a monthly channel calibration requirement for the Power Range Neutron Flux function, and is unchanged by this license amendment.] Channel calibration surveillance interval extensions are beyond the scope of TSTF-41 1, Revision 1. Consequently, the quarterly channel calibration frequency for the Power Range Neutron Flux trip will be retained, and the proposed CFT frequency for this functional unit will be quarterly, rather than the approved semi-annual frequency.

Therefore, the proposed change to extend the CFT frequency for RTS FU 2 to quarterly is bounded by the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision I, and approved in the Reference 2 SE.

8. UI Page 3/4 3-12 FU 3 - Power Range, TSTF-411, Revision I evaluates changing the specified frequency of Neutron Flux, High SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y Positive Rate in NUREG-143 1, Revision 2, is functionally equivalent to a CFT, as defined in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 3.a, "Power Change the GET Range, Neutron Flux, High Positive Rate," which corresponds to the CNP frequency from "M" to TS requirement to perform a monthly CFT for FU 3, "Power Range, Neutron Flux, High Positive Rate." Due to the current quarterly channel calibration requirements for the power range neutron flux instrumentation, I&M proposes to change the CFT frequency for FU 2, "Power Range, Neutron Flux," to quarterly, rather than semi-annually, as evaluated in

Changes to Technical Specification Table 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 Is proposed change consistent with and Unit TS Page Proposed Change Basis for Proposecdhange bouned by:,. 0 TSTIF-411,- WCAP Revision 1? approval SE?

TSTF-411, Revision 1. Consequently, the CFT frequency for RTS FU 3, 17j which uses the same excore nuclear instrumentation as RTS FU 2, will only be extended to quarterly.

U,)

The proposed change to extend the CFT frequency for RTS FU 3 to quarterly is bounded by, the corresponding change to SR3.3.1.7, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

9. UI Page 3/4 3-12 FU 4-Power Range, TSTF-41 1, Revision I evaluates changing the specified frequency of Neutron Flux, High SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-11 Negative Rate in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as defined in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 3.b, "Power Change the CFT Range Neutron Flux Rate, High Negative Rate," which corresponds to the frequency from "M" to CNP TS requirement to perform a monthly CFT for FU 4, "Power Range, Neutron Flux, High Negative Rate." Due to the current quarterly channel calibration requirements for the power range neutron flux instrumentation, I&M proposes to change the CFT frequency for FU 2, "Power Range, Neutron Flux," to quarterly, rather than semi-annually, as evaluated in TSTF-411, Revision 1. Consequently, the CFT frequency for RTS FU 4, which uses the same excore nuclear instrumentation as RTS FU 2, will only be extended to quarterly.

The proposed change to extend the CFT frequency for RTS FU 4 from monthly to quarterly is bounded by the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision I, and approved by the Reference 2 SE.

00

Changes to Technical Specification Table 4.3-1 0 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0D

. Is proposed change.

C e... ,,consistent with and Unit/ TS Page, Proposed Change -Basis for Proposed Change . bounded by:

"_.... Revision 1? approval SE?

10. UI Page 3/4 3-12 FU 5 - Intermediate CNP's TS currently specify a Start-Up CFT for FU-5, "Intermediate Range, Range, Neutron Flux Neutron Flux," with the exception that the surveillance is only required if Y Y U2 Page 3/4 3-11 not performed in the last 7 days. NUREG-1431, Revision 2, applies the Uj Replace Notation (1) SR3.3.1.8 COT to RTS Function 4, "Intermediate Range Neutron Flux."

with Notation (17) in SR 3.3.1.8 requires the performance of a COT prior to start-up and every the CFT column.

92 days thereafter, if not performed within the previous 92 days. TSTF-41 1, Revision 1, extends the 92-day frequency and the 92-day exception to 184 days each. I&M proposes to retain the current start-up frequency for this surveillance, while changing the Notation allowing a 7-day exception to allow 184 days, as evaluated by TSTF-4 11, Revision 1. This change is acceptable because the 184-day exception has been evaluated to ensure that it provides an acceptable level of equipment reliability. If the test has been performed within the previous 184 days, the instrument channels are considered OPERABLE.

Therefore, the proposed change is consistent with, and bounded by, the corresponding change to SR 3.3.1.8, as evaluated in TSTF-41 1, Revision 1, and approved by the Reference 2 SE.

11. Ul Page 3/4 3-12 FU 7 - TSTF-41 1, Revision 1 evaluates changing the specified frequency of Overtemperature SR 3.3.1.7, "Perform COT," from 92 days to 184 days A COT, as defined U2 Page 3/4 3-11 delta T in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as defined in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 6, Change the CFT Overtemperature deltaT, which corresponds to the CNP TS requirement to "SA." perform a monthly CFT for FU 7, Overtemperature deltaT. Therefore, the proposed change to extend the CFT frequency for TS FU 7 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

0r

Changes to Technical Specification Table 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0

-Is proposed change 0 consistent with and t*l 2 Unit TS Page, Proposed Change Basis for Proposed Change bounded by:

UZ TSTF-41 1, WCAP Revision 1? 'approval SE?

o)

12. UI Page 3/4 3-12 FU 8- Overpower TSTF-41 1, Revision 1 evaluates changing the specified frequency of Y Y U2 Page 3/4 3-11 delta T SR 3.3.1.7, "Perform COT," from 92 days to 184 days A COT, as defined in NUREG-143 1, Revision 2, is functionally equivalent to a CFT, as defined Change the CFT in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 7, Overpower "frequency from W" to deltaT, which corresponds to the CNP TS requirement to perform a monthly "CFTfor FU 8, Overpower deltaT. Therefore, the proposed change to extend the CFT frequency for TS FU 8 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision 1, and approved in the Referencd 2 SE.

13. U1 Page 3/4 3-12 FU 9 - Pressurizer TSTF-41 1, Revision I evaluates changing the specified frequency of Pressure - Low SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y in NUREG-143 I, Revision 2, is functionally equivalent to a CFT, as defined Change the CFT in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 8.a, Pressurizer "frequency from W" to Pressure - Low, which corresponds to the CNP TS requirement to perform a "monthlyCFT for FU 9, Pressurizer Pressure - Low. Therefore, the proposed change to extend the CFT frequency for TS FU 9 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

14. U1 Page 3/4 3-12 FU 10- Pressurizer TSTF-41 1, Revision I evaluates changing the specified frequency of U2 Page 3/4 3-11 Pressure - High SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as defined Change the CFT in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 8 b, Pressurizer "frequency from "" to Pressure - High, which corresponds to the CNP TS requirement to perform a monthly CFT for FU 10, Pressurizer Pressure - High. Therefore, the proposed change to extend the CFT frequency for TS FU 10 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

(D C)

Changes to Technical Specification Table 4.3-1 0 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change consistent with and, F s Unit iTS Page Proposed Change Basis for Proposed Change bounded by: F TSTF-411, . WCAP Revision 1? approval SE? 0*

15. UI Page 3/4 3-12 FU 11 - Pressurizer TSTF-41 1, Revision 1 evaluates changing the specified frequency of Y Y U2 Page 3/4 3-11 Water Level - High SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined (')

Changeethe 3-11in NUREG-143 1, Revision 2, is functionally equivalent to a CFT, as defined Change the CFT in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 9, Pressurizer frequency from "M" to Water Level - High, which corresponds to the CNP TS requirement to "SA." perform a monthly CFT for FU 11, Pressurizer Water Level - High Therefore, the proposed change to extend the CFT frequency for TS FU 11 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as approved in TSTF-41 1, Revision 1.

16. U1 Page 3/4 3-12 FU 12- Loss of Flow - TSTF-41 1, Revision I evaluates changing the specified frequency of U2 Page 3/4 3-11 Single Loop SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as defined Change the CFT in the CNP TS SR 3.3.1.7 is specified for ISTS Function 10, Reactor "frequency from "M" to Coolant Flow - Low, which corresponds to the CNP TS requirement to "SA." perform a monthly CFT for FU 12, Loss of Flow - Single Loop. Therefore, the proposed change to extend the CFT frequency for TS FU 12 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

17. UI Page 3/4 3-13 FU 14- Steam TSTF-41 1, Revision I evaluates changing the specified frequency of Generator Water SR 3.3.1.7, "Perform COT," from 92 days to 184 days A COT, as defined Y Y U2 Page 3/4 3-12 Level - - Low-Low in NUREG-143 1,Revision 2, is functionally equivalent to a CFT, as defined in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 14, Steam Change the GET Generator (SG) Water Level - Low Low, which corresponds to the CNP TS "frequencyfrom "M" to requirement to perform a monthly CFT for FU 14, Steam Generator Water Level -- Low-Low. Therefore, the proposed change to extend the CFT frequency for TS FU 14 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 I, Revision 1, and approved in the Reference 2 SE.

Changes to Technical Specification Table 4.3-1 0 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change, consistent with and E Unit /TS Page Proposed Change Basis for Proposed Change bounded by:,,

TSTF-41 1, - WCAP Revision 1? approval SE?

18. UI Page 3/4 3-13 FU 15- TSTF-41 1, Revision I evaluates changing the specified frequency of Steam/Feedwater SR 3.3.1.7, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-12 Flow Mismatch and in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as defined Low Steam Generator in the CNP TS. SR 3.3.1.7 is specified for ISTS Function 15, "SG Water Water Level Level - Low, Coincident with Steam Flow/Feedwater Flow Mismatch,"

which corresponds to the CNP TS requirement to perform a monthly CFT Change the Cfm for FU 15, "Steam/Feedwater Flow Mismatch and Low Steam Generator "SA." Water Level." Therefore, the proposed change to extend the CFT frequency for TS FU 15 is consistent with, and bounded by, the corresponding change to SR 3.3.1.7, as evaluated in TSTF-41 1, Revision I, and approved in the Reference 2 SE.

19 UI Page 3/4 3-13 FU 19 - SI Input from CNP's RTS FU 19, SI Input from ESF, ensures that a reactor trip occurs when Y Y ESF the SI system is actuated. The means of actuating the SI system trips are (Ref.

U2 Page 3/4 3-12 UFSAR, Section 7.2.3):

Change the CFT 0

frequency from "M" to _ Manual Actuation (ESFAS FU l.a)

"4SA."- Containment Pressure- High (ESFAS FU I.c)

Reference to Notation - Pressurizer Pressure - Low (ESFAS FU I.d)

(15) in the CFT - Differential Pressure Between Steam Lines - High (ESFAS FU L.e) column. - Steam Line Pressure- Low (ESFAS FU 1.0 Notation (4) of the CNP TS Table 4.3-1 specifies that the manual ESF input check is performed every 18 months. This input check corresponds to the 18 month TADOT that is specified by the ISTS, and will not be affected by this change.

WCAP-15376, Revision 0 and TSTF-41 1, Revision 1, provide the technical basis for extending the SR 3.3 2 5 surveillance frequency for ISTS ESFAS Functions I.c, I.d, I.e () , and L.e (2) to semi-annual. Based on the evaluations provided in WCAP-15376 and TSTF-41 1, I&M proposes to extend the surveillance frequency of the four corresponding CNP ESFAS FUs (l.c, l.d, l.e and 1.) to semi-annual (see proposed changes to TS Table 4.3-2).

C'D The CFT for the automatic actuation of RTS FU 19 is integral to the CFTs

Changes to Technical Specification Table 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS D

,t-0 Is proposed change consistent with and "5 Unit TS Page Proposed Change, Basis for Proposed Change bounded by:

TSTF-411, WCAP Revision 1? approval SE?

performed for ESFAS FU I c, I d, I e, and l.f. To perform this surveillance, simulated signals are injected into the ESFAS channels. The logic circuitry for the function being tested generates an ESFAS Auto Actuation signal (ESFAS FU I) concurrent with a SI signal to the Reactor Trip System (RTS FU 19).

Based on this test methodology, performance of a CFT for ESFAS FU l.c, l.d, I e and L.f also constitutes a CFT performance of RTS FU 19. Consequently, the COT requirements for this RTS function were eliminated from the standard TS, with the only test remaining for this function being the 18-month TADOT.

Therefore, the proposed change to extend the CFT frequency f6r RTS FU 19, in concert with the proposed changes to extend the CFT frequency of ESFAS FU 1.c, l.d, i.e and 1.f, is consistent with, and bounded by, the surveillance interval extensions evaluated in TSTF-41 1, Rev. 1, and approved in the Reference 2 SE.

20. UI Page 3/4 3-13 FU 21.A - Reactor TSTF-41 1, Rev. 1 evaluates changing the specified frequency of SR 3.3.1.4, Trip Breaker; Shunt "Perform TADOT," from monthly to "62 days on a STAGGERED TEST Y Y U2 Page 3/4 3-12 Trip Function BASIS" SR 3.3.1.4 is specified for ISTS Function 19, RTBs, and ISTS Change the CFT Function 20, RTB Undervoltage and Shunt Trip Mechanisms, which frequncy Ccorresponds "frequency fromthe""M" to 21.A, "Reactor to the CNP TS requirement to perform a monthly CFT for FU Trip Breaker - Shunt Trip Function." A TADOT, as defined "4 Months." in NUREG-1431, Rev. 2, consists of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The intent of the CFT for TS FU 21.A is to ensure that the RTB will open within the required time upon the receipt of an auto shunt trip signal. CNP's current TS requirement for a CFT is satisfied by connecting an event recorder at the reactor trip breaker auto shunt trip panel, such that the technician monitors both the initiation of the shunt trip, when the test pushbutton is depressed and released, and the RTB status; i e., either open or closed. The position of the RTB is also verified locally and recorded. This test verifies that the RTB will open upon receipt of an auto shunt trip signal and that the response time of the RTB is within the allowed time. This test corresponds to the ISTS requirement of a TADOT. Therefore, it is appropriate to relate the ISTS TADOT frequency, as modified by TSTF-411, Rev. 1, to the CFT frequency, as specified in

Changes to Technical Specification Table 4.3-1 0 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change consistent with and, E UntI Pg rooe hneBasis for Proposed Change bounded by:

0 U ~TSTF-411ý,WCAP, Revision 1? approval SE?

CNP's current TS.

Therefore, the proposed change to extend the CFT frequency for TS M° FU 2l.A is consistent with, and bounded by, the corresponding change to SR3.3.1.4, as evaluated in TSTF-411, Rev. 1, and approved in the Reference 2 SE.

21. UI Page 3/4 3-13 FU 21.B - Reactor TSTF-41 1, Rev. 1 evaluates changing the specified frequency of SR 3 3.1.4, Y Y U2 Page 3/4 3-12 Trip Breaker; "Perform TADOT," from monthly to "62 days on a STAGGERED TEST Undervoltage Trip BASIS." SR 3.3.1.4 is specified for ISTS Function 19, RTBs, and ISTS Function Function 20, RTB Undervoltage and Shunt Trip Mechanisms, which Change the CFT corresponds to the CNP TS requirement to perform a monthly CFT for FU frequency C21.B, tho

" Reactor Trip Breaker - Undervoltage Trip Function. A TADOT, as "frequency from "M" to defined in NUREG-1431, Rev. 2, consists of operating the trip actuating "4 Months." device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The intent of the CFT for TS FU 21.B is to ensure that the RTB will open upon receipt of an RTB undervoltage trip signal. CNP's current TS requirement for a CFT is satisfied by blocking the auto shunt trip of the RTB, such that only the undervoltage trip of the breaker can occur, while depressing the test pushbuttons for the train being tested. This test verifies that the RTB will open upon receipt of an RTB undervoltage trip signal. This test corresponds to the ISTS requirement of a TADOT. Therefore, it is appropriate to relate the ISTS TADOT frequency, as modified by TSTF-41 1, Rev. 1, to the CFT frequency, as specified in CNP's current TS.

For a two-train system, such as the CNP RTBs, a "4 Month" frequency, where each train is tested at least every other 62 days (see Notation (5),

belowe), is equivalent to the ISTS frequency of "62 days on a STAGGERED TEST BASIS.

Therefore, the proposed change to extend the CFT frequency for TS FU 2 1.B is consistent with, and bounded by, the corresponding change to SR 3.3.1.4, as evaluated in TSTF-41 1, Rev. I, and approved in the Reference 2 SE.

Changes to Technical Specification Table 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change "0"d SP Bsis for Proposed Change:: :::,!: :: , consistent with and., t"l 2 Unit ITS Page Proposed Change Basis febounded by:

0 UZ TSTF-41 1, ýWCXP u'J

____ .Revision'l? approval SE?

22. UI Page 3/4 3-13 FU 22 - Automatic TSTF-41 1, Rev. 1, evaluates changing the specified frequency of RTS SR 3.3.1.5, "Perform ACTUATION LOGIC TEST," from 31 days to 92 days Y Y U2 Page 3/4 3-12 Trip Logic Change the CFT on a STAGGERED TEST BASIS. The Actuation Logic Test, as defined in frnequencF fromNUREG-1431, Rev. 2, is functionally equivalent to the CFT, that is "frequency from "MSto performed for CNP's RTS FU 22, "Automatic Trip Logic." FU 22 generates the reactor trip signals that cause the RTB of one train and RTBB of the other Replace Notation (5) train to open and shut down the reactor. Proposed Notation (15) states, with Notation (15) in "Each train tested at least every other 92 days." For a two-train system, such the CFT column, as the CNP RTS, a semi-annual surveillance frequency, where each train is tested at least every other 92 days, is equivalent to the ISTS "92 days on a STAGGERED TEST BASIS." Adding the reference to proposed Notation (15) will allow the staggered testing of FU 22 to be consistent with the staggered testing of FU 19.

Therefore, the proposed staggered semi-annual test frequency of FU 22 is consistent with, and bounded by, the corresponding change to SR 3.3.1.4, as evaluated in TSTF-41 1, Rev. 1, and approved in the Reference 2 SE.

Changes to Technical Specification Table 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 Is proposed change consistent with and SUnit /TS Page" Pr'oposed Change Basis for Proposed Change' -,boundedby:',, 0

" - TSTF-411, WA

.. Revision 1? approval SE?

23. UI Page 3/4 3-13 FU 23 - Reactor Trip TSTF-41 1, Rev. I evaluates changing the specified frequency of SR 3.3.1.4, M.,

Y Y U2 Page 3/4 3-12 Bypass Breaker "Perform TADOT," from monthly to "62 days on a STAGGERED TEST Change the CFT BASIS." SR 3.3.1.4 is specified for ISTS Function 19, RTBs, and ISTS frneq thencF frFunction 20, RTB Undervoltage and Shunt Trip Mechanisms, which "frequency from "M" to corresponds to the CNP TS requirement to perform a monthly CFT for "FU23, "Reactor Trip Bypass Breaker." A TADOT, as defined in NUREG-1431, Rev. 2, consists of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The intent of the CFT for TS FU 23 is to ensure that the RTBB will open upon receipt of a trip signal from automatic logic portion of the solid state protection system while the RTB is being tested. CNP's current TS requirement for a CFT is satisfied by verifying that the RTBB will open when the TRIP pushbutton is depressed. This test corresponds to the ISTS requirement of a TADOT. Therefore, it is appropriate to relate the ISTS TADOT frequency, as modified by TSTF-41 1, Rev. 1, to the CFT frequency, as specified in CNP's current TS. For a two-train system, such as the CNP RTBs and RTBBs, a "4 Month" frequency, where each train is tested at least every other 62 days (see Notation (5), below), is equivalent to the ISTS frequency of "62 days on a STAGGERED TEST BASIS.

Therefore, the proposed change to extend the CFT frequency for TS FU 23 is consistent with, and bounded by, the corresponding change to SR 3.3.1.4, as evaluated in TSTF-41 1, Rev. I, and approved in the Reference 2 SE.

Changes to Technical Specification Table 4.3-1 0 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change consistent with and' r= 'Unit/TSPage Proposed Change Basis for Proposed Change, bounded by: 0 U TSTF-411, WCAP Revision 1? approval SE? tTl

24. UI Page 3/4 3-13 FU 23- Reactor Trip TSTF-41 I, Revision 1, evaluates a change to the frequency of SR 3.3.1.4, Y Y U2 Page 3/4 3-12 Bypass Breaker "Perform TADOT," from "31 days on a STAGGERED TEST BASIS," to Add Notation (5)to the "62 days on a STAGGERED TEST BASIS." The TADOT frequency CFTacolumn. specified by ISTS SR 3.3.1.4 is applicable to ISTS RTS Function 19, "Reactor Trip Breakers (RTBs)," which includes any RTBBs that are racked in and closed for bypassing an RTB. By adding Notation (5) to the CFT column of TS Table 4.3-1 for RTS Functional Unit 23, "Reactor Trip Bypass Breaker," alternate trains of RTBBs will be tested every other 2 months (i.e.,

every other 62 days). This is consistent with the "62 days on a STAGGERED TEST BASIS" frequency specified in TSTF-411, Revision 14, for SR 3 3.1.4. Therefore, the proposed change to apply Notation (5) to FU 23 is consistent with, and bounded by the "62 days on a STAGGERED TEST BASIS" RTBB TADOT frequency that was evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

25 U I Page 3/4 3-14 Notation (5) TSTF-41 1, Revision I, evaluates a change to the frequency of SR 3.3.1.4, "Perform TADOT," from "31 days on a STAGGERED TEST BASIS," to Y Y tested every other "62 days on a STAGGERED TEST BASIS." The TADOT frequency month." to "Each train specified by ISTS SR 3.3.1.4 is only applicable to ISTS RTS Functions 19, mont. ato "eachteran "Reactor Trip Breakers (RTBs)," and 20, "Reactor Trip Breaker tested at least every Undervoltage and Shunt Trip Mechanisms." Function 19 includes any other 62 days." RTBB that are racked in and closed for bypassing an RTB. Notation (5) to CNP's TS Table 4.3-I will only be applicable to RTS Functional Units 21.A, "RTB - Shunt Trip Function," 21.B, "RTB - Undervoltage Trip Function,"

and 23, "Reactor Trip Bypass Breaker." For dual-train systems, such as CNP's RTBs, the revised frequency is equivalent to testing each train on a 4-month (i e., 124-day) frequency, such that alternate trains are tested every other 2 months (i.e., every other 62 days). Therefore, the proposed change to revise Notation (5) to read, "Each train tested at least every other 62 days." is consistent with, and bounded by the "62 days on a STAGGERED TEST BASIS" frequency that was evaluated in TSTF-4 11, Revision 1, and approved in the Reference 2 SE.

0D

-,4.

Changes to Technical Specification Table 4.3-1 0 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change t..

consistent wvith and

, Unit / TS Page Proposed Change -Basis for Proposed Change bounded by: 0 TSTF-411, WCAP Revision 1? approval SE?

26 UI Page 3/4 3-14 Notation (15) TSTF-41 1, Revision 1, evaluates a change to the frequency of SR 3 3.1.5, "Perform ACTUATION LOGIC TEST," from "31 days on a STAGGERED Y Y U2 Page 3/4 3-13 Add new Notation (15) TEST BASIS," to "92 days on a STAGGERED TEST BASIS." For dual to read, "Each train train systems, such as CNP's RTS, the revised frequency is equivalent to tested at least every testing each train on a semi-annual (i e., 184-day) frequency, such that other 92 days." alternate trains are tested every other 3 months (i e., every other 92 days).

Therefore, the proposed change to revise Notation (15) to read, "Each train tested at least every other 92 days." is consistent with, and bounded by the "92 days on a STAGGERED TEST BASIS" frequency that was evaluated in TSTF-41 I, Revision 1, and approved in the Reference 2 SE.

27. U1 Page 3/4 3-14 Notation (16) Reference I proposed a change to Notation (16) to clarify the CFT for the RTBBs is only required for RTBBs racked in and closed for bypassing an n/a n/a U2 Page 3/4 3-13 Add ne, Notation (16) RTB. Since this change is not within the scope ofTSTF-41 1, Rev. 1, and to to read, "Not Used" expedite NRC review of changes within the scope of TSTF-41 1, Rev. 1, this change has been withdrawn by Enclosure 2 of this letter. To prevent confusion that would result from changing Notation (17) to Notation (16), it is proposed that Notation (16) be revised to indicate "Not Used" as a placekeeper.

28. UI Page 3/4 3-14 Notation (17) TSTF-411, Revision 1, revises the note for SR 3.3.1.8, "Perform COT," to state: "Only required when not performed within previous 184 days." Y Y U2 Page 3/4 3-13 Add neiNotatton (17) SR3.3.1.8 is applicable to RTS Function 4, "Intermediate Range Neutron to read, "If not Flux." The surveillance of this function is only required during infrequent performed in previous evolutions (i.e., during start-up), so the Note provides the basis for a 184 days" minimum surveillance frequency in the event of more frequent evolutions (e g, consecutive plant start-ups within a period less than 184 days).

Therefore, the proposed new Notation (17) is consistent with, and bounded by, the corresponding Note in SR 3.3.1.8, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

CD 00

Changes to Technical Specification Table 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 0

Is proposed change consistent with and,:

"Unit/ TS Page Proposed Change Basis for Proposed Change bounded by: tri TSTF-411, ,WCAP Revision 1? approval SE? z

29. U I Page 3/4 3-31 FU l.b - SI, Turbine TSTF-41 1, Rev. I evaluates changing the specified frequency of SR 3.3.2.2, Trip, Feedwater "Perform ACTUATION LOGIC TEST," and SR 3.3.2.4, "MASTER Y Y U2 Page 3/4 3-30 Isolation, and Motor RELAY TEST," from 31 days to 92 days on a STAGGERED TEST BASIS.

Driven Auxiliary Drivendwate SR 3.3.2.2 and SR 3.3.2.4 are specified for ISTS Function l.b, "Safety Aux - Injection - Automatic Actuation Logic and Actuation Relays" (including Feedwater Pumps - Functions 5.c, "Turbine Trip and Feedwater Isolation - Safety Injection,"

Automatic Actuation and 6.d, "Auxiliary Feedwater - Safety Injection"), which correspond to the Change the CFT CNP TS requirement to perform a monthly CFT for ESFAS FU L.b, "Safety frequency from "M" to Injection, Turbine Trip, Feedwater Isolation, and Motor Driven Feedwater "SA."1 Pumps - Automatic Actuation Logic." The intent of the CFT for ESFAS FU l.b is to verify that a simulated or actual SI signal generates a turbine trip, feedwater isolation, and motor driven auxiliary feedwater pump automatic actuation. Testing the solid state protection system satisfies the current TS requirement for a CFT. The surveillance verifies that the logic output generates the SI signal and that the master relays actuate when the SI signal is applied. Slave relays are verified by checking that there is continuity through the coils of the relays. This test corresponds to the ISTS requirements of an Actuation Logic Test and a Master Relay Test.

Consequently, it is appropriate to relate the ISTS frequencies for SR 3.3.2.2 and SR 3.3.2.4, as modified by TSTF-41 1, Rev. 1, to the CFT frequency for FU I.b.

For a two-train system, such as the CNP SI System, a semi-annual surveillance frequency, where each train is tested at least every other 92 days (see change to Notation (2), below) is equivalent to the ISTS "92-days on a STAGGERED TEST BASIS" frequency.

Therefore, the proposed change to extend the CFT frequency for TS FU 1.b is consistent with, and bounded by, the corresponding changes to SR 3.3.2.2 and SR 3.3.2.4, as evaluated in TSTF-41 1, Rev. I and approved in the Reference 2 SE.

Changes to Technical Specification Table 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 Is proposed change consistent with and "nUnit / TS Page Proposed Change Basis for Proposed Change bounded by:

TSTF-411, WA Revision 1? approval SE?

30. UI Page 3/4 3-31 FU .e - SI, Turbine TSTF-411, Revision 1 evaluates changing the specified frequency of Trip, Feedwater SR 3.3.2.5, "Perform COT," from 92 days to 184 days A COT, as defined Y Y U2 Page 3/4 3-30 Isolation, and Motor in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as UJ Driven Fenedw Auxiliary xiliaerP defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function l.c,

- "Safety Injection - Containment Pressure - High 1" (including Function 5.c, Feedwater Pumps - "Turbine Trip and Feedwater Isolation - Safety Injection," and 6 d, Containment "Auxiliary Feedwater - Safety Injection"), which correspond to the CNP TS Pressure - High requirement to perform a monthly CFT for ESFAS FU l.c, "Safety Change the CFT Injection, Turbine Trip, Feedwater Isolation, and Motor Driven Feedwater frequency from "M" to Pumps - Containment Pressure - High." Therefore, the proposed change to CFT frequency for TS FU I c is consistent with, and bounded "by, the the extend corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 I, Revision I and approved in the Reference 2 SE.

31. Ul Page 3/4 3-31 FU ld - SI, Turbine TSTF-411, Revision 1 evaluates changing the specified frequency of Trip, Feedwater SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-30 Isolation, and Motor in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Driven Auxiliary defined "Safety in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 1.d, Injection - Pressurizer Pressure - Low," 5.c, "Turbine Trip and Feedwater Pumps - Feedwater Isolation - Safety Injection," and 6 d, "Auxiliary Feedwater Pressurizer Pressure Safety Injection," which correspond to the CNP TS requirement to perform

- Low a monthly CFT for ESFAS FU l.d, "Safety Injection, Turbine Trip, Change the CFT Feedwater Isolation, and Motor Driven Feedwater Pumps - Pressurizer frequency from "M" to Pressure - Low." Therefore, the proposed change to extend the CFT "11SA." frequency for TS FU l.d is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 1, Revision I and approved inthe Reference 2 SE.

cJQ tOi C:)

Changes to Technical Specification Table 4.3-2 0 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change -D consistent With'and,',

Unit /TS Page,-,-, Proposed Change Basis for Proposed Change "I, bounded by: 0-4 TSTF-411, WCAP Revision 1? approval SE?

32. Ul Page 3/4 3-31 FU L.e - SI, Turbine TSTF-411, Revision I evaluates changing the specified frequency of t->

Trip, Feedwater SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-30 Isolation, and Motor in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Driven Auxiliary defined in the CNP TS. SR 3 3.2.5 is specified for ISTS Function L.e (2),

Feedwater Pumps - "Safety Injection - Steam Line Pressure - High Differential Pressure Between Steam Lines," (including Function 5.c, "Turbine Trip and Difrerential Pressure Feedwater Isolation - Safety Injection," and 6.d, "Auxiliary Feedwater Between Steam Lines Safety Injection"), which corresponds to the CNP TS requirement to

- High perform a monthly CFT for ESFAS FU I e, "Safety Injection, Turbine Trip, Change the CFT Feedwater Isolation, and Motor Driven Feedwater Pumps - Differential Pressure Between Steam Lines -- High." Therefore, the proposed change to frequency from "M" to extend the CFT frequency for TS FU I.e is consistent with, and bounded "SA." by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 I, Revision 1 and approved in the Reference 2 SE.

33. Ul Page 3/4 3-31 FU l.f- SI, Turbine TSTF-41 1, Revision 1 evaluates changing the specified frequency of Trip, Feedwater SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-30 Isolation, and Motor in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Driven Auxiliary defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function I.e.(l),

Feedwater Pumps - "Safety "TurbineInjection - Steam Line Pressure Isolation- Low," (including Functions Trip and Feedwater - Safety Injection," and 65 c, d,

Steam Line Pressure "Auxiliary Feedwater - Safety Injection"), which corresponds to the CNP

- Low TS requirement to perform a monthly CFT for ESFAS FU 1.f, "Safety Change the CFT Injection, Turbine Trip, Feedwater Isolation, and Motor Driven Feedwater frequency from "M" to Pumps - Steam Line Pressure - Low."

"SA." Therefore, the proposed change to extend the CFT frequency for TS FU L.f is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 1, Revision 1 and approved in the Reference 2 SE.

t0J

Changes to Technical Specification Table 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 CD Is proposed change consistent with and SUnit / TS Page Proposed Change Basis for Proposed Change bounded by:,

STSTF.411, WCAP Revision 1? approval SE?

0°

34. U1 Page 3/4 3-31 FU 2.b - TSTF-41 1, Rev. 1 evaluates changing the specified frequency of SR 3.3.2.2, Containment Spray - "Perform ACTUATION LOGIC TEST," and SR 3.3.2.4, "MASTER Y Y U2 Page 3/4 3-30 Automatic Actuation RELAY TEST," from 31 days to 92 days on a STAGGERED TEST BASIS.

Logic SR 3 3.2.2 and SR 3 3.2.4 are specified for ISTS Function 2 b, "Containment Spray - Automatic Actuation Logic and Actuation Relays,"

Change the CFT which corresponds to the CNP TS requirement to perform a monthly CFT frequency from "M" to for ESFAS FU 2 b, "Containment Spray - Automatic Actuation Logic."

"SA." The intent of the CFT for ESFAS FU 2.b is to verify that a simulated or actual containment spray signal generates a containment spray automatic actuation. Testing the solid state protection system satisfies the current TS requirement for a CFT. The surveillance verifies that the logic output generates the containment spray signal and that the master relays actuate when the containment spray signal is applied. Slave relays are verified by checking that there is continuity through the coils of the relays. This test corresponds to the ISTS requirements of an Actuation Logic Test and a Master Relay Test. Consequently, it is appropriate to relate the ISTS frequencies for SR 3.3.2.2 and SR 3.3.2.4, as modified by TSTF-41 1, Rev. 1, to the CFT frequency for FU 2.b. For a two-train system, such as the CNP Containment Spray System, a semi-annual surveillance frequency, where each train is tested at least every other 92 days (see change to Notation (2), below) is equivalent to the ISTS "92-days on a STAGGERED TEST BASIS" frequency Therefore, the proposed change to extend the CFT frequency for TS FU 2.b is consistent with, and bounded by, the corresponding changes to SR 3.3.2.2 and SR 3.3.2 4, as evaluated in TSTF-41 1, Rev. I and approved in the Reference 2 SE.

tC kx)

Changes to Technical Specification Table 4.3-2 0 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Ct Is proposed change, consistent with and Unit ITS Page Proposed Change Basis for Proposed Chang bounded by:,,

-U Z TSTF-41 1' WCA Revision 1? approval SE?

O-'

35. UI Page 3/4 3-31 FU 2.c - TSTF-41 1, Revision I evaluates changing the specified frequency of u3 Containment Spray - SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-30 Containment in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Pressure--High-High defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 2 c, "Containment Spray - Containment Pressure - High - 3 (High High)," which Change the CFT corresponds to the CNP TS requirement to perform a monthly CFT for frequency from "M" to ESFAS FU 2.c, "Containment Spray - Containment Pressure -- High-High."

"SA." Therefore, the proposed change to extend the CFT frequency for TS FU 2.c is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-4 11, Revision 1 and approved in the Reference 2 SE.

36. U1 Page 3/4 3-32 FU 3.a.(2) - TSTF-41 1, Rev. I evaluates changing the specified frequency of SR 3.3.2.2, Y Y U2 Page 3/4 3-30 Containment "Perform ACTUATION LOGIC TEST," and SR 3.3.2.4, "MASTER Isolation - Phase "A" RELAY TEST," from 31 days to 92 days on a STAGGERED TEST BASIS.

Isolation - From SI SR 3.3.2.2 and SR 3.3.2.4 are specified for ISTS Functions l.b, "Safety Automatic Actuation Injection - Automatic Actuation Logic and Actuation Relays," and 3.a.(2),

"Containment Isolation - Phase A Isolation - Automatic Actuation Logic

Logic and Actuation Relays," which correspond to the CNP TS requirement to Change the CFT perform a monthly CFT for ESFAS FU 3.a (2), "Containment Isolation frequency from "M" to Phase "A" Isolation - From SI Automatic Actuation Logic." The intent of "the CFT for ESFAS FU 3.a (2) is to verify that a simulated or actual SI signal generates a Phase "A" containment isolation automatic actuation.

Testing the solid state protection system satisfies the current TS requirement for a CFT. The surveillance verifies that the logic output generates the Phase "A" containment isolation signal and that the master relays actuate when the Phase "A" containment isolation signal is applied. Slave relays are verified by checking that there is continuity through the coils of the relays. This test corresponds to the ISTS requirements of an Actuation Logic Test and a Master Relay Test. Consequently, it is appropriate to relate the ISTS frequencies for SR 3.3.2.2 and SR 3.3.2.4, as modified by TSTF-41 1, Rev. I, to the CFT frequency for FU 3.a.(2). For a two-train system, such as the CNP ESFAS, a semi-annual surveillance frequency, where each train is tested at least every other 92 days (see change to Notation (2), below) is equivalent to the ISTS "92-days on a STAGGERED

Changes to Technical Specification Table 4.3-2 0 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change 4e-

f consistent with and Unit ITS Pae Proposed Change Basis for Proposed Change bounded by: =+

0 SZ .TSTF=411, WCAP Revision 1? approval SE?

TEST BASIS" frequency.

Therefore, the proposed change to extend the CFT frequency for TS FU 3.a (2) is consistent with, and bounded by, the corresponding changes to SR 3 3 2.2 and SR 3.3 2.4, as evaluated in TSTF-41 I, Rev. I and approved in the Reference 2 SE.

37. UI Page 3/4 3-32 FU 3.b.(2) - TSTF-41 1, Rev. I evaluates changing the specified frequency of SR 3.3.2.2, Containment "Perform ACTUATION LOGIC TEST," and SR 3.3.2.4, "MASTER Y Y U2 Page 3/4 3-30 Isolation - Phase "B" RELAY TEST," from 31 days to 92 days on a STAGGERED TEST BASIS.

Isolation - Automatic SR 3.3.2.2 and SR 3.3.2.4 are specified for ISTS Function 3.b.(2),

"Containment Isolation - Phase B Isolation - Automatic Actuation Logic Actuation Logic and Actuation Relays," which corresponds to the CNP TS requirement to Change the CFT perform a monthly CFT for ESFAS FU 3.b (2), "Containment Isolation frequency from "M" to Phase "B"Isolation - Automatic Actuation Logic." The intent of the CFT "11SA." for ESFAS FU 3.b.(2) is to verify that a simulated or actual Phase "B" isolation signal generates a containment isolation automatic actuation.

Testing the solid state protection system satisfies the current TS requirement for a CFT. The surveillance verifies that the logic output k.)O generates the containment isolation signal and that the master relays actuate when the containment isolation signal is applied. Slave relays are verified by checking that there is continuity through the coils of the relays This test corresponds to the ISTS requirements of an Actuation Logic Test and a Master Relay Test. Consequently, it is appropriate to relate the ISTS frequencies for SR 3.3.2.2 and SR 3.3.2.4, as modified by TSTF-41 1, Rev 1, to the CFT frequency for FU 3 b (2) For a two-train system, such as the CNP ESFAS, a semi-annual surveillance frequency, where each train is tested at least every other 92 days (see change to Notation (2), below) is equivalent to the ISTS "92-days on a STAGGERED TEST BASIS" frequency.

Therefore, the proposed change to extend the CFT frequency for TS FU 3 b (2) is consistent with, and bounded by, the corresponding changes to SR 3.3.2.2 and SR 3.3 2 4, as evaluated in TSTF-41 1, Rev. I and approved in the Reference 2 SE. CD t*j 4ý1

Changes to Technical Specification Table 4.3-2 0 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change consistent with and Unit /TS Page Proposed Change Basis for Proposed Change , bounded by: 0 Z TSTFn411, WCAP Revision 1? approval SE?

38. UI Page 3/4 3-32 FU 3.b.(3) - TSTF-41 1, Revision I evaluates changing the specified frequency of Containment SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-30 Isolation - Phase "B" in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Isolation - defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 3.b (3),

"Containment Isolation - Phase B Isolation - Containment Pressure High -3 Conta -inment (High High)," which corresponds to the CNP TS requirement to perform a Pressure--High-High monthly CFT for ESFAS FU 3.b.(3), "Containment Isolation - Phase "B" Change the CFT Isolation - Containment Pressure -- High-High." Therefore, the proposed frequency from "M" to change to extend the CFT frequency for TS FU 3.b (3) is consistent with, "ltSA." and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 I, Revision 1, and approved in the Reference 2 SE.

39. U1 Page 3/4 3-33 FU 4.b - Steam Line TSTF-411, Rev. I evaluates changing the specified frequency of SR 3.3.2.2, Isolation - Automatic "Perform ACTUATION LOGIC TEST," and SR 3.3 2.4, "MASTER Y Y U2 Page 3/4 3-31 Actuation Logic RELAY TEST," from 31 days to 92 days on a STAGGERED TEST BASIS.

SR 3.3.2.2 and SR 3.3.2.4 are specified for ISTS Function 4.b, "Steam Line k)

Change the CFT Isolation - Automatic Actuation Logic and Actuation Relays," which frequency from "M" to corresponds to the CNP TS requirement to perform a monthly CFT for "SA." ESFAS FU 4.b, "Steam Line Isolation - Automatic Actuation Logic." The intent of the CFT for ESFAS FU 4.b is to verify that a simulated or actual SI signal generates a steam line isolation automatic actuation. Testing the solid state protection system satisfies the current TS requirement for a CFT.

The surveillance verifies that the logic output generates the steam line isolation signal and that the master relays actuate when the steam line isolation signal is applied. Slave relays are verified by checking that there is continuity through the coils of the relays. This test corresponds to the ISTS requirements of an Actuation Logic Test and a Master Relay Test.

Consequently, it is appropriate to relate the ISTS frequencies for SR 3.3.2.2 and SR 3.3.2.4, as modified by TSTF-41 1, Rev. 1, to the CFT frequency for FU 4.b. For a two-train system, such as the CNP ESFAS, a semi-annual surveillance frequency, where each train is tested at least every other 92 days (see change to Notation (2), below) is equivalent to the ISTS "92-days on a STAGGERED TEST BASIS" frequency CD Therefore, the proposed change to extend the CFT frequency for TS FU 4.b

Changes to Technical Specification Table 4.3-2 0 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Is proposed change tT*

consistent with and "Unit/TSPage Proposed Change Basis for Proposed Change bounded by: 0 U= ".TSTF-411, , WCAP i¢k Revision 1? approval SE?

is consistent with, and bounded by, the corresponding changes to SR 3.3.2.2 and SR 3.3.2.4, as evaluated in TSTF-41 1, Rev. 1 and approved in the Reference 2 SE.

40. U1 Page 3/4 3-33 FU 4.c - Steam Line TSTF-41 1, Revision 1 evaluates changing the specified frequency of Isolation - SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-31 Containment in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Pressure--High-High defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 4 c, "Steam Line Isolation - Containment Pressure - High 2," which corresponds Change the CFT to the CNP TS requirement to perform a monthly CFT for ESFAS FU 4.c, frequency from "M" to "Steam Line Isolation - Containment Pressure -- High-High." Therefore, "SA." the proposed change to extend the CFT frequency for TS FU 4 c is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

41. U I Page 3/4 3-33 FU 4.d - Steam Line TSTF-411, Revision 1 evaluates changing the specified frequency of Isolation - Steam SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-31 Flow in Two Steam in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Lines-High, defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 4.e, Coincident with "Steam Line Isolation - High Steam Flow Coincident with Tvg - Low Low,"

which corresponds to the CNP TS requirement to perform a monthly CFT Tavg-Low-Low for ESFAS FU 4.d, "Steam Line Isolation - Steam Flow in Two Steam Change the CFT Lines -- High, Coincident with T.,--Low-Low." Therefore, the proposed frequency from "M" to change to extend the CFT frequency for TS FU 4.d is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

Changes to Technical Specification Table 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0b Is proposed change consistent with and-Unit TS Page Proposed Change Basis for Proposed Change,_,-_ bounded by: 0>

Uz TSTF-411, WCAP Revision 1? approval SE?

42. U1 Page 3/4 3-33 FU 4.e - Steam Line TSTF-41 1, Revision I evaluates changing the specified frequency of -o Isolation - Steam SR 3 3 2 5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-31 Line Pressure - Low in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as U) defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 4.d.(l),

Change the CFT "Steam Line Isolation - Steam Line Pressure - Low," which corresponds to frequency from "M" to the CNP TS requirement to perform a monthly CFT for ESFAS FU 4.e, "SA." "Steam Line Isolation - Steam Line Pressure - Low." Therefore, the proposed change to extend the CFT frequency for TS FU 4.e is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

43. U1 Page 3/4 3-33 FU 5.a - Turbine TSTF-41 1, Revision I evaluates changing the specified frequency of Trip and Feedwater SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-31 Isolation -Steam in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Generator Water defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 5 b, "Turbine Trip and Feedwater Isolation - SG Water Level - High High Level-High-High (P-14)," which corresponds to the CNP TS requirement to perform a Change the CFT monthly CFT for ESFAS FU 5.a, "Turbine Trip and Feedwater Isolation frequency from "M" to Steam Generator Water Level -- High-High." Therefore, the proposed "10SA." change to extend the CFT frequency for TS FU 5.a is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 1, Revision I, and approved in the Reference 2 SE.

44. UI Page 3/4 3-33 FU 6.a - Motor TSTF-41 1, Revision I evaluates changing the specified frequency of Driven Auxiliary SR 3.3.2.5, "Perform COT," from 92 days to 184 days. A COT, as defined Y Y U2 Page 3/4 3-31 Feedwater Pumps - in NUREG-1431, Revision 2, is functionally equivalent to a CFT, as Steam Generator defined in the CNP TS. SR 3.3.2.5 is specified for ISTS Function 6 c, Water Level-Low- "Auxiliary Feedwater - SG Water Level - Low Low," which correspond to the CNP TS requirement to perform a monthly CFT for ESFAS FU 6.a, Low "Motor Driven Auxiliary Feedwater Pumps - Steam Generator Water Level Change the CFT -- Low-Low." Therefore, the proposed change to extend the CFT frequency frequency from "M" to for TS FU 6.a is consistent with, and bounded by, the corresponding change "SA." to SR 3.3 2 5, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE. CD

Changes to Technical Specification Table 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0 0-.

Is proposed change consistent with and

'0 Unit / TS Page Proposed Change Basis for Proposed Change- bouned by:

TSTF-411, I WCAP Revision 1? approval SE? U..)

45. UI Page 3/4 3-33 FU 6.c - Motor TSTF-41 1, Revision 1 evaluates changing the specified frequency of Driven Auxiliary SR 3.3.2.2, "Perform ACTUATION LOGIC TEST," and SR 3.3.2.4, Y Y U2 Page 3/4 3-31 Feedwater Pumps - "MASTER RELAY TEST," from 31 days to 92 days on a STAGGERED SI TEST BASIS. SR 3.3.2.2 and SR 3.3.2.4 are specified for ISTS Function l.b, "Safety Injection - Automatic Actuation Logic and Actuation Relays,"

Change the CFT which corresponds to the CNP TS requirement to perform a monthly CFT frequency from "M" to for ESFAS FU 6 c, "Motor Driven Auxiliary Feedwater Pumps - Safety "SA." Injection." The intent of the CFT for ESFAS FU 6 c is to verify that a simulated or actual SI signal generates a motor driven auxiliary feedwater pump automatic actuation. Testing the solid state protection system satisfies the current TS requirement for a CFT. The surveillance verifies that the logic output generates the motor driven auxiliary feedwater pump signal and that the master relays actuate when this signal is applied. Slave relays are verified by checking that there is continuity through the coils of the relays. This test corresponds to the ISTS requirements of an Actuation Logic Test and a Master Relay Test. Consequently, it is appropriate to relate the ISTS frequencies for SR 3.3.2.2 and SR 3.3.2.4, as modified by TSTF-411, Rev. I, to the CFT frequency for FU 6.c. For a two-train system, such as the CNP ESFAS, a semi-annual surveillance frequency, where each train is tested at least every other 92 days (see change to Notation (2), below) is equivalent to the ISTS "92-days on a STAGGERED TEST BASIS" frequency.

Therefore, the proposed change to extend the CFT frequency for TS FU 6 c is consistent with, and bounded by, the corresponding changes to SR 3.3.2.2 and SR 3.3.2.4, as evaluated in TSTF-41 1, Rev. 1 and approved in the Reference 2 SE.

CD t'1) 00

Changes to Technical Specification Table 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0

C'D

-Is proposed change 04 consistent with and 4-1 Unit-TS Page Proposed'Change Basis for Proposed Change bounded by: 0 TSTF-4119, WCAP Revision 1? approvalISE?

U)

46. UI Page 3/4 3- FU 7.a - Turbine The change to the COT frequency specified in TSTF-41 1, Revision 1, 33a Driven Auxiliary SR 3.3.2.5, is applicable to the ISTS ESFAS Function 6.c, "Auxiliary Y Y Feedwater Pumps - Feedwater - SG Water Level - Low Low." The ESFAS Instrumentation U2 Page 3/4 3-32 Steam Generator specified in NUREG-1431, Revision 2, "Standard Technical Specfications Water Level-Low- Westinghouse Plants," Table 3.3.2-I does not differentiate between ESFAS instrumentation functions associated with motor-driven and turbine-driven Low auxiliary feedwater pumps. Consequently, the 184-day surveillance Change the CFT frequency discussed above for TS Functional Unit 6 a, "Motor Driven frequency from "M" to Auxiliary Feedwater Pumps - Steam Generator Water Level -- Low-Low,"

"isalso applicable to TS Functional Unit 7.a, "Turbine Driven Auxiliary Feedwater Pumps - Steam Generator Water Level -- Low-Low." Therefore, the proposed change to extend the CFT frequency for TS FU 7.a, is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

47. UI Page 3/4 3- FU 10.b - TSTF-41 1, Revision I evaluates changing the specified frequency of 33b Containment Air SR 3.3.2.2, "Perform ACTUATION LOGIC TEST," and SR 3.3.24, Y Y Recirculation Fan - "MASTER RELAY TEST," from 31 days to 92 days on a STAGGERED U2 Page 3/4 3-32 Automatic Actuation TEST BASIS. The combination of Actuation Logic and Master Relay Logic Tests, as defined in NUREG-1431, Revision 2, correspond to current CFT requirements, as defined in the CNP TS. The ISTS do not include Change the CFT specifications for a containment air recirculation fan, an ESF feature which frequency from "M" to is used for the mitigation of in-containment steam release events in plants "SA." with ice condenser containment designs However, other ESFAS automatic actuation logic functions, including ISTS Function I b, "Safety Injection Automatic Actuation Logic and Actuation Relays," and 2.b, "Containment Spray - Automatic Actuation Logic and Actuation Relays," specify the performance of SR 3.3.2.2 and 3.3.2.4, which corresponds to the CNP TS requirement to perform a monthly CFT for ESFAS FU 1O.b, "Containment Air Recirculation - Automatic Actuation Logic." For a two-train system, such as the CNP ESFAS, a semi-annual surveillance frequency, where each train is tested at least every other 92 days (see change to Notation (2),

below) is equivalent to the ISTS "92-days on a STAGGERED TEST BASIS" frequency.

Changes to Technical Specification Table 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0

, *.Is proposed change consistent with and Unit ITS Page '-Proposed Change Basis for Proposed Change bounded by: 0 TSTF-411, ' WCAP Revision 1? approval SE?

Therefore, the proposed change to extend the CFT frequency for TS FU 10.b is consistent with, and bounded by, the corresponding change to SR 3.3 2.2 and SR 3.3.2.4, as evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

48. UI Page 3/4 3- FU 10.c - TSTF-41 I, Rev. 1 evaluates changing the specified frequency of SR 3.3 2.5, 33b Containment Air "Perform COT," from 92 days to 184 days. A COT, as defined in Y Y Recirculation - NUREG-1431, Rev. 2, is functionally equivalent to a CFT, as defined in the U2 Page 3/4 3-32 Containment CNP TS. The ISTS do not include specifications for a containment air Pressure - High recirculation fan, an ESF feature which is used for the mitigation of in containment steam release events in plants with ice condenser containment Change the CFT designs. However, other ESFAS function that receive an input from the frequency from "M" to Containment Pressure High signal, such as ISTS Functions 1.c, "Safety 0

"SA." Injection - Containment Pressure - High 1," and 4 c, "Steam Line Isolation Containment Pressure - High 2," specify a COT, which corresponds to the CNP TS requirement to perform a CFT. Therefore, the proposed change to extend the CFT frequency for TS ESFAS FU 10.c is consistent with, and bounded by, the corresponding change to SR 3.3.2.5, as evaluated in TSTF-4 11, Rev. 1, and approved in the Reference 2 SE. C)

49. Ul Page 3/4 3-34 Notation (2) TSTF-41 1, Revision 1, evaluates a change to the frequency of SR 3.3.2.2, "Perform ACTUATION LOGIC TEST," and SR 3.3.2.4, "Perform Y Y U2 Page 3/4 3-33 Change from "Each MASTER RELAY TEST," from " 31 days on a STAGGERED TEST train or logic channel BASIS," to "92 days on a STAGGERED TEST BASIS." For dual-train shall be tested at least systems, such as CNP ESFAS, the revised frequency is equivalent to testing every other 31 days." each train on a semi-annual (i.e., 184-day) frequency, such that alternate to "...at least every trains are tested every other 3 months (i.e., every other 92 days). For the other 92 days." CNP TS, the requirement to test ESFAS instrumentation on a staggered test basis is addressed by the application of Notation (2), which specifies that "Each train or logic channel shall be tested at least every other 31 days."

Therefore, the proposed change to revise Notation (2) to read, "Each train tested at least every other 92 days." is consistent with, and bounded by the "92 days on a STAGGERED TEST BASIS" frequency that was evaluated in TSTF-41 1, Revision 1, and approved in the Reference 2 SE.

ATTACHMENT 5 TO AEP:NRC:3311 REGULATORY COMMITMENTS The following table identifies those actions committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment Date I&M will implement administrative controls in the Donald C. The appropriate administrative Cook Nuclear Plant configuration risk management program to controls will be established include the above restrictions when a reactor trip breaker when the surveillance test and/or logic cabinet is removed from service. interval extensions are implemented following NRC approval.

I&M has no plans for significant protection system equipment The appropriate administrative upgrades at this time; however, administrative controls will be controls will be established implemented to ensure any future digital upgrades to the when the surveillance test reactor trip system and/or engineered safety feature actuation interval extensions are system are evaluated to ensure the generic applicability of implemented following NRC WCAP-15376-P is not affected. approval.

v t e Letter Sequence Supplement
TAC:MB6324, Surveillance Test Interval Extensions for Components of the Reactor Protection System - WCAP-15376-P, TSTF-41, TSTF-4 (Open) TAC:MB6325, Surveillance Test Interval Extensions for Components of the Reactor Protection System - WCAP-15376-P, TSTF-41, TSTF-4 (Open)
Initiation Acceptance... Supplement, Supplement, Supplement Administration Questions 27 March 2003 Answers 7 April 2003 Results Approval... Other:
MONTHYEAR2003-04-29 [Table View]

ML030640115

Text

Reference:

Enclosures:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 TECHNICAL ANALYSIS

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

References:

References:

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