Text

License Amendment Request to Extend Reactor Trip System & Engineered Safety Features Actuation System Surveillance Time Requirements as Evaluated in WCAP-15376
	ML022490127
Person / Time
Site:	Cook
Issue date:	08/30/2002
From:	Joseph E Pollock; Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP:NRC:2311, RG-1.174, RG-1.177 WCAP-15376
	Download: ML022490127 (108)
	v • d • e

Indiana Michigan Power Company 500 Circle Drive

,, Buchanan, MI 49107 1373 INDIANA MICHIGAN POWER August 30, 2002 AEP:NRC:2311 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

SUBJECT:

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 Time Requirements as Evaluated in WCAP-15376

REFERENCE:

WCAP-15376, Revision 0, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," dated October 2000

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend Appendix A, Technical Specifications (TS), of Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to revise the Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS) Surveillance Requirements in TS 3/4.3.1 and TS 3/4.3.2, respectively, by increasing the channel operational test surveillance intervals for analog channels, logic cabinets, and reactor trip breakers, and increasing the completion time and bypass time for the reactor trip breakers in accordance with the evaluation and justifications presented in the referenced document, WCAP-15376.

Additionally, I&M proposes to remove Mode 2 applicability for the RTS low pressurizer pressure and high pressurizer water level trips and to add a note to TS Table 4.3-1 clarifying that channel functional testing requirements for the reactor trip bypass breakers are only applicable when they are racked in and closed for bypassing a reactor trip breaker. I&M also proposes format and capitalization changes to the affected TS pages that improve appearance but do not affect any requirements.

U. S. Nuclear Regulatory Commission AEP:NRC:2311 Page 2 These changes are generically evaluated in WCAP-15376, which is currently under review by the Nuclear Regulatory Commission (NRC). Any additional conditions stipulated in the Safety Evaluation Report that documents NRC approval of WCAP-15376 will be addressed by a supplement to this license amendment request.

The approach used in WCAP-15376 is consistent with the NRC's approach for evaluating risk-informed changes to a plant's current licensing basis as presented in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," and RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications." No new plant specific risk-informed analyses were required to support this proposed amendment request.

Enclosure 1 to this letter provides an oath and affirmation affidavit pertaining to the requested amendment. Enclosure 2 provides a detailed description and technical analysis to support the proposed amendment, including an evaluation of significant hazards considerations pursuant to 10 CFR 50.92(c) and an environmental assessment. Attachments IA and lB provide TS pages that are marked to show the proposed changes for Unit 1 and Unit 2, respectively.

Attachments 2A and 2B provide TS" pages with the proposed changes incorporated for Unit 1 and Unit 2, respectively. Attachment 3 provides an analysis to determine the applicability of WCAP-15376 to CNP. Attachment 4 lists each proposed TS change and the justification for each proposed change.

Attachment 5 contains a list of new regulatory commitments made in this letter.

I&M requests approval of the proposed amendment within six months from the receipt of this request. Once approved, the amendment will be implemented within 30 days.

No pending amendment requests affect the TS pages that are submitted in this request. If future submittals affect these TS pages, I&M will coordinate the changes to the pages with the NRC Project Manager to ensure proper TS page control when the associated license amendment requests are approved.

U. S. Nuclear Regulatory Commission AEP:NRC:2311 Page 3 Should you have any questions, or require additional information, please contact Mr. Brian A. McIntyre, Manager of Regulatory Affairs, at (269) 697-5806.

Sincerely, J. E. Pollock Site Vice President NH/jen

Enclosures:

1 Notarized Affidavit 2 Description and Technical Analysis for the Proposed Changes Attachments:

IA and 1B Technical Specification Pages Marked to Show Proposed Changes 2A and 2B Technical Specification Pages with the Proposed Changes Incorporated 3 Applicability of WCAP-15376 Analyses 4 Detailed Summary and Justification of Proposed Technical Specification Changes 5 Regulatory Commitments c: K. D. Curry J. E. Dyer MDEQ - DW & RPD NRC Resident Inspector R. Whale

Enclosure 1 to AEP:NRC:2311 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 THI SO DAY OF Ut , 2002 JUUE E.NEWMILLER J. Notary Public, Berrien County, Mi My Commission Expires Aug 22,2004 Notary i ublic My Commission Expires 8e'*

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to AEP:NRC:2311 Page I License Amendment Request Extend Reactor Trip System and Engineered Safety Features Actuation System Surveillance Time Requirements as Evaluated in WCAP-15376

1.0 DESCRIPTION

Indiana Michigan Power Company (I&M) proposes to amend Attachment A, Technical Specifications (TS) of Facility Operating Licenses DPR-58 and DPR-74 for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, respectively. The proposed changes will revise the TS to increase the channel operational test (COT) surveillance test intervals (STIs) for Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS) analog channels, logic cabinets, and reactor trip breakers (RTBs), and the completion time (CT) and bypass time (BT) for the RTBs.

These changes are generically evaluated in WCAP-15376, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times" (Reference 1), which is currently under review by the Nuclear Regulatory Commission (NRC). Any additional conditions stipulated in the Safety Evaluation Report (SER) that documents NRC approval of WCAP-15376 will be addressed by a supplement to this License Amendment Request.

2.0 PROPOSED CHANGE

S The proposed changes are presented in the following table:

Table 2.1 Proposed Changes for RTS and ESFAS TS Based on WCAP-15376 Component Surveillance Test Interval Completion Time and Bypass Changes Time Changes (3)

Logic Cabinet 1 month to 6 months None Proposed Analog Channels 1 month to 6 months None Proposed Power Range Neutron Flux 1 month to quarterly None Proposed

& related instrumentation(4)

Reactor Trip Breakers 1 month to 4 months CT: 0(2) hours to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months BT: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to AEP:NRC:2311 Page 2 Notes:

(1) Includes the Reactor Trip Bypass Breakers (RTBB) when in use.

(2) The current TS does not provide for a restoration time. It specifies that the reactor be placed in Hot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(3) CT is also referred to as allowed outage time (AOT) in WCAP-15376.

(4) Includes RTS Functional Units (FU) 2, Power Range, Neutron Flux; FU 3, Power Range, Neutron Flux, High Positive Rate; and FU 4, Power Range, Neutron Flux, High Negative Rate.

The following changes are proposed specifically for the implementation of the revised surveillance intervals/times presented in Table 2.1:

1. Add the definition of a four-month frequency to TS Table 1.2, "Frequency Notation."

2. Delete the Modes 1 and 2 applicability of Action 1 and add Action 15 for RTS FU 21, "Reactor Trip Breakers," in TS Table 3.3-1.

3. Incorporate various editorial changes to RTS FU 21, "Reactor Trip Breakers," and FU 22, "Automatic Trip Logic," in TS Table 3.3-1 to clarify which Actions apply in each Mode.

Specifically, in Modes 1 and 2, for FU 21, Action 13 and proposed Action 15 will apply, and for FU 22, Action 1 applies. In Modes 3, 4, and 5, for FU 21 and FU 22, Action 14 applies.

This editorial change involves increasing the vertical spacing between the lines for these two sets of Actions, and for Unit 2 only deleting the comma after the "2" in the Applicable Modes column.

4. Change Action 13 to the Action Statement Notations for TS Table 3.3-1. This Action, which only applies to FU 21, "Reactor Trip Breakers," in Modes 1 and 2 will state:

"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."

5. Add Action 15 to the Action Statement Notations for TS Table 3.3-1. This Action, which will apply only to FU 21, "Reactor Trip Breakers," in Modes 1 and 2 will state:

"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."

to AEP:NRC:2311 Page 3

6. Change the Channel Functional Test (CFT) frequency specified in TS Table 4.3-1, "Reactor Trip System Instrumentation Surveillance Requirements," from monthly (M) to semi annually (SA) for the following RTS FUs:

7. Overtemperature delta T (OTAT)

8. Overpower delta T (OPAT)

9. Pressurizer Pressure - Low

10. Pressurizer Pressure - High

11. Pressurizer Water Level - High

12. Loss of Flow- Single Loop

14. Steam Generator Water Level - Low-Low

15. Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level

19. Safety Injection Input from ESF

22. Automatic Trip Logic

7. Change the CFT frequency specified in TS Table 4.3-1 from monthly (M) to quarterly (Q) for the following RTS FUs:

2. Power Range, Neutron Flux

3. Power Range, Neutron Flux, High Positive Rate

4. Power Range, Neutron Flux, High Negative Rate

8. Change the CFT frequency specified in TS Table 4.3-1 from monthly (M) to "4 Months," for the following RTS FUs:

21.A. Reactor Trip Breaker- Shunt Trip Function 21.B. Reactor Trip Breaker -Undervoltage Trip Function

23. Reactor Trip Bypass Breaker

9. 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

5. Intermediate Range, Neutron Flux

23. Reactor Trip Bypass Breaker

10. Delete the TS Table 4.3-1 requirement to perform a Start-Up (S/U) 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

11. Change the TS Table 4.3-1, "Notation (5)" surveillance frequency from "Each train tested every other month" to "Each train tested at least every other 62 days."

to AEP:NRC:2311 Page 4

12. Revise the CFT for FU 23, "Reactor Trip Bypass Breaker," by applying Notation (5) to the Channel Functional Test Column in TS Table 4.3-1.

13. Add Notation (15) to TS Table 4.3-1 to state, "Each train tested at least every other 92 days."

Notation (15) will be applied to the CFT requirements of FU 19, "Safety Injection Input from ESF," and FU 22, "Automatic Trip Logic."

14. 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." Notation (16) will be applicable only to FU 23, "Reactor Trip Bypass Breaker."

15. Add Notation (17) to TS Table 4.3-1 to state, "If not performed in previous 184 days."

16. Change the CFT frequency specified in TS Table 4.3-2, "Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements," from monthly (M) to semi annually (SA) for the following ESFAS FUs:

1. Safety Injection, Turbine Trip, Feedwater Isolation, and Motor Driven Auxiliary Feedwater Pumps

b. Automatic Actuation Logic

c. Containment Pressure - High

d. Pressurizer Pressure - Low

e. Differential Pressure Between Steam Lines - High

f. Steam Line Pressure - Low

2. Containment Spray

b. Automatic Actuation Logic

c. Containment Pressure - High-High

3. Containment Isolation

a. Phase "A" Isolation

2) From Safety Injection Automatic Actuation Logic

b. Phase "B" Isolation

2) Automatic Actuation Logic

3) Containment Pressure - High-High

4. Steam Line Isolation

b. Automatic Actuation Logic

c. Containment Pressure - High-High

d. Steam Flow in Two Steam Lines - High Coincident with Tag - Low-Low

e. Steam Line Pressure - Low

5. Turbine Trip and Feedwater Isolation

a. Steam Generator Water Level - High-High

6. Motor Driven Auxiliary Feedwater Pumps

a. Steam Generator Water Level - Low-Low

c. Safety Injection to AEP:NRC:2311 Page 5

7. Turbine Driven Auxiliary Feedwater Pumps

a. Steam Generator Water Level - Low-Low

10. Containment Air Recirculation Fan

b. Automatic Actuation Logic

c. Containment Pressure - High

17. Change the TS Table 4.3-2, "Table Notation (2)" frequency from "at least every other 31 days," to "at least every other 92 days."

The following associated changes to the RTS/ESFAS TS Tables are also proposed:

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

2. Make format and capitalization changes to the affected TS pages that improve appearance but are not intended to introduce other changes.

The proposed changes are consistent with the evaluation and justifications presented in WCAP-15376, and the Improved Standard Technical Specifications in NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," Revisions 0, 1, and 2 (References 2, 3, and 4).

Each proposed TS change and the justification for each proposed change is listed in . TS Tables 1.2, 3.3-1, 4.3-1, and 4.3-2 marked-up with the proposed changes are presented in Attachments IA and 1B for Units I and 2, respectively. The TS pages with the proposed changes incorporated are presented in Attachments 2A and 2B.

I&M is not proposing to adopt the following TS changes, which were included in the generic evaluations in WCAP-15376:

1. RTS FU 5: Intermediate Range, Neutron Flux, is evaluated for a 6-month STI in WCAP-15376. For CNP, the CFT for the intermediate range, neutron flux FU is required to be current prior to each start-up. (Note: Channel functional tests, or CFTs, as per the CNP TS, are referred to as channel operability tests, or COTs, in WCAP-15376.) The STI for the intermediate range, neutron flux FU will remain in accordance with CNP's current licensing basis.

2. RTS FU 6: Source Range, Neutron Flux, is evaluated for a 6-month STI in Mode 2 per WCAP-15376. The CNP TS require the source range, neutron flux CFT to be performed monthly in Modes 2, 3, 4, and 5, and prior to start-up. Except for core physics testing, CNP does not routinely operate the plant in Mode 2 during startup or shutdown. The routine startup procedure from a non-refueling outage is to steadily increase power into Mode 1 following criticality. During a routine shutdown from Mode 1, the plant is procedurally tripped from low in the power range, thereby quickly passing through Mode 2 into Mode 3.

to AEP:NRC:23 11 Page 6 Since the STI is not applicable in Mode 1, the potential for a reactor trip at power due to human error while performing a source range CFT does not exist. The STI for the source range, neutron flux will remain in accordance with CNP's current licensing basis.

3. WCAP-15376 generically evaluates a 6-month STI for master relays, while retaining the 3-month STI for the slave relays. Surveillance testing of the master and slave relays at-power is not specified in the CNP TS and CNP has not committed to testing these relays at-power. In accordance with TS Table 4.3-2, "Engineered Safety Features Actuation System Instrumentation Surveillance Requirements," FU 9, "Manual," a trip actuating device operational test is performed at each refueling. This includes the master and slave relays, including their associated output contacts. The master relay actuation, including the associated output contacts, and slave relay coil continuity are procedurally checked at-power.

As a part of implementing the proposed TS changes, the master relays and the slave relay coil continuity will continue to be procedurally tested at the frequency specified in TS for the automatic actuation logic. This is not a commitment to test the master or slave relays at-power.

These proposed TS changes do not involve changes to actuation setpoints/setpoint tolerance, testing acceptance criteria, or channel response times. No hardware changes are proposed nor required to implement these changes at CNP.

In summary, I&M proposes to amend the Unit 1 and Unit 2 TS Tables 1.2, 3.3-1, 4.3-1, and 4.3-2. For the RTS and ESFAS, I&M proposes to increase the CFT STI for the RTBs, logic cabinets, and analog channels. I&M also proposes to increase the allowed CT and the BT for the RTBs in accordance with WCAP-15376. Additional TS changes that are necessary to support the generic changes evaluated in WCAP-15376 are also proposed.

3.0 BACKGROUND

In the early 1980s, in response to growing concerns of the impact of current testing and maintenance requirements on plant operation, particularly as related to instrumentation systems, the Westinghouse Owner's Group (WOG) initiated a program to develop a justification to be used to revise generic and plant-specific instrumentation TS. Operating plants experienced many inadvertent reactor trips and safeguards actuations during performance of instrumentation surveillance, causing unnecessary transients and challenges to safety systems. Significant time and effort on the part of the operating staff was devoted to performing, reviewing, and documenting and tracking the various surveillance activities, which in many instances seemed unwarranted based on the high reliability of the equipment. Significant benefits for operating plants appeared to be achievable through revision of instrumentation test and maintenance requirements. The results of these studies and the recommended changes to the testing of reactor protection and engineered safeguards instrumentation are documented in WCAP-10271-P-A, and WCAP-10271-P-A, Supplement 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System" (References 5 and 6), and to AEP:NRC:2311 Page 7 WCAP-10271-P-A, Supplement 2, Revision 1 "Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System" (Reference 7).

The NRC completed an evaluation of surveillance testing at power, which indicated that testing in many areas could be reduced without any significant reduction in safety. These findings and recommendations are documented in NUREG-1366, "Improvement to the Technical Specifications Surveillance Requirements," and Generic Letter 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation." Reduced surveillance testing of the Reactor Protection System (RPS) and ESFAS analog instrumentation was recommended. In 1998, the NRC completed a review of the WOG Topical Report WCAP-14333-P-A, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," (Reference 11). Additional time for testing and extended allowed outage time for the RPS and ESFAS were approved by the NRC on July 15, 1998 (Reference 12).

The WOG submitted WCAP-15376, Revision 0, to the NRC for review on November 8, 2000 (Reference 13). In WCAP-15376, the WOG proposed generic extensions of the STIs for the RTS and ESFAS analog channels, logic cabinets, and the master relays. In addition, it proposed the generic extension of the STI, CT, and BT for the RTBs. The WCAP-15376 evaluation used the CT and BT values in WCAP-14333-P-A as the basis for determining the increased risk of further relaxation of the STIs, and the RTB CT and BT.

Surveillance Extensions Evaluated in WCAP-10271-P-A In WCAP-10271-P-A, and its supplements (References 5, 6, and 7), the WOG evaluated the impact of increased STIs, CTs, and BTs and their effect on core damage frequency (CDF) and public risk. The time extensions addressed in this WCAP were for the following RPS and ESFAS components:

"* STI

- analog channels

" CT and BT

- analog channels

- logic cabinet,

- master relays, and

- slave relays In February 1985, the NRC issued the SER (Reference 8) for WCAP-10271 and Supplement 1.

This SER approved quarterly testing, if conducted on a staggered basis, six hours to place a failed channel in a tripped mode, increased AOTs for testing and maintenance, and testing in bypass for analog channels of the RPS. In February 1989, the NRC issued the SER (Reference

9) for WCAP-10271, Supplement 2, and Supplement 2, Revision 1. The SER approved quarterly to AEP:NRC:2311 Page 8 testing, six hours to place a failed channel in a tripped mode, increased AOTs for testing and maintenance, and testing in bypass for analog channels of the ESFAS. By a letter dated April 30, 1990 (Reference 10), the NRC issued the Supplemental SER (SSER) for WCAP-10271, Supplement 2 and Supplement 2, Revision 1. The SSER approved the STI and AOT extensions for the ESFAS functions that were not included in the Westinghouse Standard Technical Specifications (i.e., functions associated with the Safety Injection, Steam Line Isolation, Main Feedwater Isolation, and Auxiliary Feedwater Pump Start signals).

To incorporate these extended times, the NRC required that an applicant for a TS amendment must meet certain plant-specific conditions stipulated in the SERs and SSER. The generically-approved changes to CTs, BTs, and the analog channel STI specified in WCAP-10271-P-A, including its supplements, have not been amended into the CNP TS.

However, since these stipulated plant-specific conditions constitute part of the basis for the changes evaluated in WCAP-15376, I&M has addressed each of the conditions stipulated in the WCAP-10271-P-A SERs and SSER. The stipulated plant-specific conditions and the I&M evaluations are presented below:

1. SER Stipulated Condition - The RPS SER (Reference 8) required the use of a staggered test plan for the RTS channels changed to the extended test frequency.

Response - In the subsequent SER for WCAP-10271-P-A, Supplement 2 (Reference 9), the NRC did not repeat this requirement for ESFAS channels and withdrew the requirement for the RPS channels. Therefore, this stipulated condition is no longer applicable.

2. SER Stipulated Condition - The RPS SER (Reference 8) required that plant procedures require a common cause evaluation for RPS channels changed to the extended test frequency and additional testing for plausible common cause failures.

Response - In accordance with CNP's Corrective Action Program and existing plant procedures, installed equipment (including RTS/ESFAS components) failures and setpoint deficiencies require immediate Shift Manager notification and entry into the corrective action process, via the initiation of an action request/condition report. The "extent of condition" portion of the condition report evaluation, and root cause evaluation (if required), evaluates similar components on the other train, unit, etc. to determine if the failure mechanism could be common to other channels. Additionally, if the failed component is required in the current operating mode, and there is reasonable assurance to believe that the component is not capable of performing its safety function, the operability determination will require a check of the operability of redundant equipment.

3. SER Stipulated Condition - The RPS SER (Reference 8) required installed hardware capability for testing in the bypass mode. Approval of routine channel testing in a bypassed condition is contingent on the capability of the RPS design to allow such testing without lifting leads or installing temporary jumpers.

to AEP:NRC:2311 Page 9 Response - The protection system installed at CNP does not have the hardware capability for full bypassing of each RTS and ESFAS instrument channel. Only those instrument channels that have hardware installed to permit testing in bypass without lifting leads or installing jumpers are routinely tested in bypass. The system design at CNP provides a test switch that interrupts the channel output to the logic circuitry during individual channel testing. When activated, the test switch de-energizes the associated logic input causing that portion of the logic to be actuated (partial trip) accompanied by a partial trip alarm and channel status light actuation in the control room. Analog channel testing of an inoperable channel typically will be completed with the channel being tested in the tripped condition.

However, due to the potential of having to perform periodic surveillance testing with one channel in the trip system already inoperable, the TS currently allow a bypass time for surveillance testing of the remaining operable channels of the same functional unit. This situation is non-routine and applies to a failed channel that must be bypassed to facilitate surveillance testing of the other channels associated with a given trip system function. For this infrequent situation, the channel cannot be assumed operable and is therefore considered to be in a state of bypass during the surveillance testing of the other associated channels.

Other than those channels that have installed bypass capability, I&M does not plan to implement routine testing in bypass of the RTS/ESFAS functions.

4. SER Stipulated Condition - The RPS SER (Reference 8) indicated that, for channels that provide input to both the RPS and the ESFAS, the more stringent ESFAS requirements still apply.

Response - The extensions generically approved in the SER and SSER for the ESFAS analog channels (References 9 and 10) were the same as those approved for the RPS analog channels; therefore, this condition is no longer applicable.

5. SER Stipulated Condition - The RPS SER (Reference 8) requires confirmation that the instrument setpoint methodology includes sufficient margin to offset the drift anticipated as a result of less frequent surveillance.

Response - The application of this stipulated condition to the WCAP-15376 surveillance intervals is addressed in Section 4.0 of this enclosure.

6. SER Stipulated Condition - The ESFAS SER (Reference 9) requires confirmation of the applicability of the generic analyses to the plant.

Response - The information provided in Attachment 3 demonstrates the applicability of the generic WCAP-15376 analysis to CNP's ESFAS and RTS. These tables are based on implementation guidelines that were issued by the WOG for licensees implementing the TS AOT changes that were justified in WCAP-14333-P-A. The tables were modified, as appropriate, to reflect the information specified in WCAP-15376.

to AEP:NRC:2311 Page 10

7. SSER Stipulated Condition - Create a separate new action statement for Modes 1 and 2 for RPS Automatic Trip and Interlock Logic Functional Units.

Response - The generically-approved CT and BT changes specified in WCAP-10271-P-A, including its supplements, have not been amended into the CNP TS; therefore, this stipulated condition does not apply.

Surveillance Extensions Evaluated in WCAP-14333-P-A In WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times" (Reference 11), the WOG proposed and justified further extensions of the RPS and ESFAS CTs and BTs for the analog channels, logic cabinets, master relays, and slave relays. This WCAP did not propose additional extensions of any STI. Time extensions were addressed in WCAP-14333-P-A for the following RPS and ESFAS components:

"* CT and BT

- analog channels

" CT Only

- logic cabinets

- master relays

- slave relays The NRC approved these generic CT and BT extensions in an SER dated July 15, 1998 (Reference 12). The SER that approved the generic changes in WCAP-14333-P-A indicated that the increase in CDF and Large Early Release Frequency (LERF) for those plants that have not implemented the changes evaluated in WCAP-10271-P-A, and its supplements, is small (specifically, for 2-out-of-3 and 2-out-of-4 logic, the increase in CDF is small, approximately 3.1% and 2.3% respectively). The LERF would increase by only four percent for both 2-out-of-3 and 2-out-of-4 logic schemes. In approving the proposed TS changes, the NRC concluded that implementation of the changes specified in WCAP-14333-P-A would result in a very small quantitative impact on plant risk.

The staff required that an applicant for a proposed amendment to incorporate these extended times into their TS must meet certain plant-specific conditions stipulated in the SER. The generically-approved changes to CTs and BTs specified in WCAP-14333-P-A have not been amended into the CNP TS. However, since these plant-specific conditions share a commonality with the changes evaluated in WCAP-15376, I&M has addressed each of the conditions stipulated in the SER for WCAP-14333-P-A (Reference 12). The stipulated plant-specific conditions and the I&M responses are presented below:

to AEP:NRC:2311 Page 11

1. SER Stipulated Condition - Confirm the applicability of the WCAP-14333-P-A analysis for their plant.

Response - The information provided in Attachment 3 demonstrates the applicability of the generic WCAP-15376 analysis to CNP's ESFAS and RTS. These tables are based on implementation guidelines that were issued by the WOG for licensees implementing the TS AOT changes that were justified in WCAP-14333-P-A. The tables were modified, as appropriate, to reflect the information specified in WCAP-15376.

2. SER Stipulated Condition - Address the Tier 2 and Tier 3 analyses, including the configuration risk management program (CRMP) insights, by confirming that these insights are incorporated into the referencing licensee's decision-making process before taking equipment out of service.

Response - The response to this stipulated condition is addressed in Section 4.0 of this enclosure.

4.0 TECHNICAL ANALYSIS

Current Design Basis The CNP RTS and ESFAS utilize a solid-state protection system to initiate reactor trip signals from 1-out-of-2, 2-out-of-3, and 2-out-of-4 protective logic. The system is designed to permit any one analog channel to be maintained, tested, or calibrated during power operation without a system trip, unless a trip condition actually exists. However, this does not include sensor calibration or such backup trips as reactor coolant pump breakers or manual trips. With two exceptions, the active parts of the system continue to satisfy the single failure criterion during surveillance testing activities, since the channel under test is either tripped or makes use of superimposed test signals that do not negate the process signal. The first exception is that 1-out-of-2 systems will not meet the single failure criterion during channel bypass provided that acceptable reliability of operation can be otherwise demonstrated and bypass time interval is short. The second exception is that the containment spray actuation channels are tested by bypassing or negating the channel under test. This is acceptable because the protection system contains four channels, and the 2-out-of-4 trip logic reduces to 2-out-of-3 during the test. CNP does not have the capability to test all of the analog channels in bypass.

The proposed changes extend the STIs, and the RTB CT and BT requirements that are specified in TS Tables 3.3-1, 4.3-1, and 4.3-2. The STI changes for the RTS and ESFAS components reduce the probability of reactor trip during component testing activities. Testing 2-out-of-3 or 2-out-of-4 logic with one channel in the tripped state makes the reactor more vulnerable to a spurious trip. The proposed STI changes will reduce the required testing on the RTS and ESFAS components without significantly impacting their reliability, and reduce the potential for reactor to AEP:NRC:2311 Page 12 trips and actuation of engineered safety features associated with more frequent testing of these components.

The CT and BT extensions for the RTB provide additional time to complete test and maintenance activities while at power, potentially reducing the number of forced outages related to compliance with the RTB CT. WCAP-15376 generically documented that averting the risk associated with a plant shutdown and restart offsets the small increase in risk from the proposed CT change due to increased signal unavailability while at power. The incurred and offsetting risks were compared. The results of the comparison are presented below:

" The generic probability of core damage is 4.7E-07 per year, based on the generic conditional core damage frequency (CCDF) of 3.OE-06 per year, and the probability of inducing a transient event during the shutdown and startup.

" The change in the generic CDF from the WCAP-14333-P-A base case, due to increasing the RTB CT to 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months plus 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to complete the mode change) and the RTB BT to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , is 1.OE-08 per year.

WCAP-15376 provides the technical basis and methodology to justify extending the STI for the analog channels, logic cabinets, RTBs and master relays, and the CT and BT for the RTBs. A risk-informed approach was used to justify these changes. The approach is consistent with that recommended by the NRC in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," dated July 1, 1998 (Reference 14), and RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1, 1998 (Reference 15).

This approach provides an evaluation of the impact on plant risk using the three-tiered approach as presented in RG 1.177. Tier 1, "PRA Capability and Insights," assesses the impact of the proposed STI changes on CDF and LERF, along with the impact of the proposed RTB CT and BT time changes on CDF. LERF, incremental conditional core damage probability, and incremental conditional large early release probability (ICLERP). Tier 2, "Avoidance of Risk Significant Plant Configurations," considers potential risk-significant plant operating configurations. Tier 3, "Risk-Informed Plant Configuration Control and Management,"

considers risk evaluations of configurations when entered on a plant-specific basis.

Since the extensions generically approved in WCAP-10271-P-A, and its supplements, and WCAP-14333-P-A have not been incorporated into the CNP TS, the CNP TS times are closely aligned with pre-technical specification optimization program (Pre-TOP) values. The tables below compare the current CNP TS times, the Pre-TOP times, and the proposed TS changes.

to AEP:NRC:2311 Page 13 Table 4.1 STI Comparison Component Current Pre-TOP STI Proposed STI Values(b)

STI Values(a)

Analog Channels 1 month 1 month 6 months Power Range Neutron Flux 1 month I month 3 months

& related instrumentation(c)

Logic Cabinets 1 month 2 months 6 months Reactor Trip Breakers 1 month 2 months 4 months Notes:

(a) Values given in WCAP-15376, Table 1.1.

(b) Values given in WCAP- 5376, Table 4.1.

(c) Includes Functional Units 2, 3 and 4 Table 4.2 CT and BT Comparison Component Current TS Pre-TOP Values(b) Proposed Values(c)

CT BT CT BT CT BT Reactor Trip Breakers owa hours 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Notes:

(a) The current TS does not provide for a restoration time. It specifies that the reactor be placed in Hot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

(b) Values provided in WCAP-15376, Table 1.1.

(c) Values provided in WCAP-15376, Table 4.1. The proposed CT change is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months plus an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to make the mode change. The current TS also specifies 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to place the reactor in Hot Standby, which will be unchanged. A proposed new action in TS Table 3.3-1 incorporates these proposed changes.

WCAP-15376 compares the cumulative impact of the proposed STI changes, and RTB CT and BT changes on CDF using the Pre-TOP values as the basis. This comparison credits the expected reduction in reactor trips due to the reduced analog channel testing resulting from extending the analog channel STIs from monthly to quarterly, as evaluated in WCAP-10271-P-A, and its supplements. The comparison indicates that the cumulative impact on CDF when using the Pre-TOP values as the base case is 5.7E-07 per year for 2-out-of-4 logic and 1.LE-06 per year 2-out-of-3 logic. These are small increases to the CDF per the acceptance criteria of 1.OE-06 per year given in RG 1.174.

WCAP-15376 does not make a similar comparison to Pre-TOP values for incremental conditional core damage probability (ICCDP), LERF or ICLERP. Since the incremental probability assessments are a direct function of the duration of the single AOT under to AEP:NRC:2311 Page 14 consideration, the only AOTs to be considered are the proposed increase in the RTB CT to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months plus 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the mode change, and the RTB BT to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The plant-specific ICCDP is calculated by substituting the CNP Unit 1 and Unit 2 CDF of 4.9E-05 per year for the WCAP-15376 base case CDF of 5.05E-05 per year in the ICCDP equation. The CNP ICCDP calculations yield an increase of:

For RTB CTs:

ICCDP = (7.07E-05/yr - 4.9E-05/yr) x 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months / (8760 hrs/yr) = 7.4E-08 For RTB BTs:

ICCDP = (7.07E-05/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) = 9.9E-09 These plant-specific increases are comparable to the WCAP-15376 calculated ICCDP increase of 6.92E-08 and 9.22E-09, respectively. Both the plant-specific and WCAP-15376 calculated increases are below 5.OE-07, which is considered very small for a single TS CT per RG 1.174.

The plant-specific LERF for both Unit 1 and Unit 2 is 5.59E-06 per year. This value is the same order of magnitude as the WCAP-15376 base case LERF values of 2.38E-06 per year for 2-out-of-4 logic and 2.44E-06 per year for 2-out-of-3 logic. WCAP-15376 determined that the combined impact on LERF due to the proposed changes in the STIs, and the RTB CT and BT, was an increase of 3.09E-08 per year for 2-out-of-4 logic and 5.68E-08 per year for 2-out-of-3 logic. These increases in LERF are small based on the RG 1.174 guidance of 1.OE-07 per year.

Therefore, the increase in the plant-specific LERF would also be small since the WCAP-15376 base case and the plant-specific LERF values are the same order of magnitude.

The impact of the RTB AOT increase on CDF and LERF is small and the ICCDP is acceptable; therefore, the ICLERP will also be acceptable. Consequently, WCAP-15376 indicated that detailed calculations to determine the impact on ICLERP are not necessary.

For WCAP-10271-P-A and WCAP-14333-P-A, the NRC stipulated that for a proposed amendment to incorporate these extended times into a facility's TS, the applicant must address certain plant-specific conditions contained in the SERs. As similar stipulations are anticipated for the approval of WCAP-15376, the plant-specific conditions stipulated for these two earlier WCAPs have been applied to the WCAP-15376 changes. I&M's responses to these conditions are presented below:

1. Anticipated Condition - Confirm that any increase in instrument drift due to the extended STIs is properly accounted for in the setpoint calculation methodology.

Response: I&M has evaluated the RTS and ESFAS instrumentation in accordance with the NRC's guidance provided in Reference 16. Reference 16 specifies the licensee activities to be performed to support a conclusion that "the setpoint drift which could be expected under the extended STIs has been studied and either (1) has been shown to remain within the to AEP:NRC:2311 Page 15 existing allowance in the RPS and ESFAS instrument setpoint calculation methodology, or (2) the allowance and setpoint have been adjusted to account for the expected drift." I&M has completed a two-part study that concluded that an adjustment of the setpoint calculation methodology is not required.

The first part of the study reviewed the Foxboro Spec 200 components in the RTS and ESFAS analog channels. This study encompassed the analog channels listed in TS Tables 4.3-1 and 4.3-2 that are subject to this request, except for the nuclear instrumentation channels. The study determined that the manufacturer's drift specifications are bounded by the one percent rack drift assumption used in the setpoint calculation methodology. An evaluation of representative field data determined that the field data is consistent with the manufacturer's drift specifications.

The second part of the study reviewed the power range nuclear instrumentation. This part of the study was based, in part, on approximately 21 months of monthly CFT data and quarterly channel calibration data. The review of the monthly and quarterly surveillance data found no instances where either the neutron flux high or low bistable setpoints, neutron flux high positive rate bistable setpoint, or neutron flux high negative rate bistable setpoint had drifted beyond the allowed tolerance value. All of these neutron flux trips use the same bistable model.

The power range nuclear instrumentation calibration method uses front panel-mounted analog test meters and indication lights as the testing instrumentation. Because of the inherent uncertainties induced by this calibration method that are not related to instrumentation drift, such as analog meter resolution, parallax, and specified meter accuracy, a rigorous statistical rack drift analysis could not be performed. Therefore, the proposed quarterly STI for the analog channels associated with the excore power range nuclear instrumentation was qualitatively evaluated for each individual functional unit.

The TS for the Power Range, Neutron Flux high trip functional units include two monitoring requirements above 15% rated thermal power (RTP) and a quarterly channel calibration requirement. The TS require that the excore nuclear instrumentation indications be compared daily to a calorimetric heat balance, and monthly to the incore axial imbalance.

If either comparison is found to be outside the TS-specified limits, a channel re-calibration is required. These two frequent monitoring requirements provide assurance that the neutron flux high trip bistable input remains within the required 'limits throughout the STI.

Therefore, based on the current TS requirements for daily and monthly channel comparisons, it is expected that excessive instrumentation drift would be identified prior to exceeding the values assumed in the setpoint calculations. Consequently, the rack drift assumptions in the setpoint calculation methodology bound the expected performance of the power range neutron flux channels over the proposed quarterly STI. The high neutron flux trip is credited in the safety analysis to provide a reactor trip for several fault events.

However, additional automatic reactor trip functions, such as OPAT, OTAT, neutron flux to AEP:NRC:2311 Page 16 high positive rate trips, safety injection actuation and/or manual operator action provide defense-in-depth protection. Based on the evaluated and expected drift, the frequent monitoring of the excore nuclear instrumentation indication/bistable input signal and the established defense-in-depth features, an extended CFT STI for the Power Range Neutron Flux functional unit (high and low setpoints) will not necessitate an adjustment of the setpoint calculation methodology for this instrumentation.

Because the TS 1.9 Channel Calibration definition requires that channel calibrations also include the performance of a CFT, no benefit would be gained by extending the Power Range, Neutron Flux CFT to semi-annual, as proposed in WCAP-15376, without a corresponding change to the channel calibration requirements. However, channel calibration surveillance interval extensions are beyond the scope of WCAP-15376 and this license amendment request; therefore, the Power Range Neutron Flux (high and low setpoints) channel calibrations will be retained at the current quarterly interval, and the CFT for this functional unit will only be extended from monthly to quarterly.

Similarly, it is proposed that the CFT STIs for the Power Range, Neutron Flux, High Positive Rate (FU 3) the Power Range, Neutron Flux, High Negative Rate (FU 4) functional units be extended from monthly to quarterly. The neutron flux rate trips use the same excore nuclear instrumentation output as the Power Range Neutron Flux trip. To provide assurance that the neutron flux rate trips remain within the values specified by the TS Limiting Safety System Setpoints, surveillance procedures specify conservative time constants with respect to these Limiting Safety System Setpoints. A comparison of the neutron flux high positive rate and high negative rate trip instrumentation surveillance procedures to the limiting safety system setpoints confirmed that the time constants are set conservatively for both rate functions.

The neutron flux high positive rate and high negative rate trips are not explicitly credited in the safety analyses, but do provide defense-in-depth for other actions credited for the mitigation of certain uncontrolled reactivity insertion events. For the rod cluster control assembly (RCCA) ejection fault condition, the neutron flux high positive rate provides defense-in-depth for the power range neutron flux reactor trip. The safety analysis for the RCCA ejection credits the power range neutron flux reactor trip, but does not take credit for the neutron flux high positive rate trip. Only the power range neutron flux (high and low trip setpoints) function is modeled in the RCCA ejection safety analysis. For the misaligned or dropped RCCA fault, the neutron flux high negative rate trip provides defense-in-depth for the operator actions. The neutron flux high negative rate is not credited in the safety analysis for providing a reactor trip. Following the identification of an RCCA group misalignment condition by the reactor operator, the operator is required to take action as required by TS and operating instructions. By procedure, the reactor operator controls the recovery from the dropped RCCA fault.

to AEP:NRC:2311 Page 17 In summary, a quarterly STI is proposed for the neutron flux high positive rate and the neutron flux high negative rate functional units. This STI extension is justified based upon:

"* Neither of the two neutron flux rate trips are credited for a reactor trip in the safety analyses,

"* Defense-in-depth is provided for neutron flux rates high and low functional units,

"* The excore nuclear instrumentation signal output will continue to be monitored on a daily and monthly frequency in accordance with TS, and

"* A review of 21 months of surveillance test data indicates that there is sufficient reason to expect the bistables to remain within their required setpoint.

Therefore, a quarterly STI for the neutron flux high positive rate and the neutron flux high negative rate functional units is justified, and will not necessitate an adjustment of the setpoint calculation methodology for this instrumentation.

I&M proposes a semi-annual CFT STI for the OTAT and OPAT functional units, FU 7 and FU 8, respectively. The OTAT and OPAT trips also use the power range excore nuclear instrumentation upper and lower flux signals as inputs by way of the f(AI) function. The power range instrumentation is calibrated to represent the core flux pattern. The method of programming the power range input portion [f(AI)] module in the overall OTAT and OPAT protection logic has been implemented to be compatible with the refueling interval calibration requirement of TS Tables 4.3-1 and 2.2-1. As such the f1(AI) input portion of the OTAT function is set conservatively with respect to the assumptions in the setpoint methodology and the f2 (AI) input portion of the OPAT function is set to zero, as specified in TS Table 2.2-1. Consequently, implementation of semi-annual CFTs for the OTAT or OPAT functions will not cause the instrument drift for the f(AI) module to exceed the calibration limits for the refueling interval calibration. The remaining portion of the OTAT and OPAT functions is bounded by a review of approximately 64 months of surveillance data that was performed in the first part of I&M's instrument drift/calculation methodology study. The OTAT and OPAT functions provide a back-up trip function to the power range neutron flux high trip function. For some fault conditions, the OTAT/OPAT functions are credited in the safety analysis, but always in conjunction with other diverse trip functions, such as power range neutron flux high and pressurizer pressure-low, and/or operator action.

In summary, a semi-annual STI is proposed for the OTAT and OPAT functional units. The OTAT and OPAT STI extension is justified based upon:

"* A review of approximately 64 months of channel calibration data for the OPAT function found no instances where the as-found data exceeded the allowed tolerances,

"* The evaluated drift data and the resultant expectation that the drift assumptions in the setpoint calculation methodology will remain bounding, to AEP:NRC:2311 Page 18

"* The frequent TS-required monitoring of the power range excore nuclear instrumentation output signal in conjunction with the fact that the f2(AI) portion of the function will exhibit zero drift, and

"* Defense-in-depth is provided by diverse reactor trip functions.

Therefore, a semi-annual STI for the OPAT and OTAT functional units is justified, and will not necessitate an adjustment of the setpoint calculation methodology for this instrumentation.

2. Anticipated Condition - Confirm the applicability of the WCAP-15376 analyses for their plant.

Response: The information provided in Attachment 3 demonstrates the applicability of the generic WCAP-15376 analysis to CNP's ESFAS and RTS. These tables are based on implementation guidelines that were issued by the WOG for licensees implementing the TS AOT changes that were justified in WCAP-14333-P-A. The tables were modified, as appropriate, to reflect the information specified in WCAP-15376.

3. Anticipated Condition - Confirm that plant procedures require a common cause evaluation for RTS channels changed to the extended test frequency and additional testing for plausible common cause failures.

Response - In accordance with CNP's Corrective Action Program and existing plant procedures, installed equipment (including RTS/ESFAS components) failures and setpoint deficiencies require immediate Shift Manager notification and entry into the corrective action process, via the initiation of an action request/condition report. The "extent of condition" portion of the condition report evaluation, and root cause evaluation (if required),

evaluates similar components on the other train, unit, etc. to determine if the failure mechanism could be common to other channels. Additionally, if the failed component is required in the current operating mode, and there is reasonable assurance to believe that the component is not capable of performing its safety function, the operability determination will require a check of the operability of redundant equipment.

4. Anticipated Condition - Address the Tier 2 and Tier 3 analyses including the configuration risk management program (CRMP) insights, by confirming that these insights are incorporated into the referenced licensee's decision-making process before taking equipment out of service.

Response: The Tier 2 and Tier 3 requirements 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). 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 to AEP:NRC:2311 Page 19 prior to entry into these configurations and appropriately managed while the plant is in these various configurations.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Determination I&M has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No.

The proposed changes to the STIs and RTB CT and BT reduce the potential for inadvertent reactor trips and spurious actuations, and therefore do not increase the probability of any accident previously evaluated. The proposed changes do not change the response of the plant to any accidents and have an insignificant impact on the reliability of the RTS and ESFAS signals. These changes satisfy the acceptance criteria specified in the NRC's regulatory guidance for evaluating risk-informed changes in RG 1.174 and RG 1.177. The RTS and ESFAS will continue to perform their functions with high reliability as originally assumed in the safety analysis, and the increase in risk is within the acceptance criteria of existing regulatory guidance; therefore, there will not be a significant increase in the consequences of any accidents.

The RTS and ESFAS are not accident initiators or precursors in the safety analysis. No new initiators are created by this activity. The proposed changes do not change any RTS or ESFAS setpoints, nor do they alter the accident mitigation function of any system, structure or component, design assumptions, conditions or configuration of the facility, or the manner in which the plant is operated and maintained. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types or amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

to AEP:NRC:2311 Page 20

2. Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated?

Response: No.

No new system interfaces or interactions are created. The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed changes do not result in a change in the manner in which the RTS and ESFAS provide plant protection. The RTS and ESFAS will continue to have the same setpoints after the proposed changes are implemented. The proposed changes to STI, CT, and BT do not change any existing accident scenarios, do not alter assumptions made in the safety analysis, nor create any new or different accident scenarios.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by these changes. Redundant RTS and ESFAS trains are maintained, and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also maintained. All signals credited as primary or secondary, and all operator actions credited in the accident analyses will remain the same.

The proposed changes will not result in plant operation in a configuration outside the design basis. The calculated impact on risk is insignificant and meets the acceptance criteria contained in RG 1.174 and RG 1.177.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, I&M concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

to AEP:NRC:2311 Page 21 5.2 Applicable Regulatory Requirement/Criteria 5.2.1 Regulations TS 3/4.3.1 and 3/4.3.2 specify the Limiting Conditions for Operation and Surveillance Requirements for instrumentation in the RTS and ESFAS, respectively. I&M is proposing a license amendment that would modify these requirements by extending STIs for many of the RTS and ESFAS analog channels and logic cabinets, extending the CTs and BTs for the RTBs and RTBBs, and deleting the Mode 2 applicability for the Pressurizer Pressure-Low and Pressurizer Water Level-High FUs. The proposed changes are based on the generic TS changes that were evaluated and justified by the WOG in WCAP-15376 and NRC-approved WCAP-14333-P-A. The Technical Analysis for the proposed license amendment demonstrates that the generic analyses of WCAP-14333-P-A and WCAP-15376 are applicable to the proposed changes to the CNP TS and that these proposed changes are, therefore, appropriate for implementation at CNP. The Technical Analysis provides the basis for I&M's determination that the proposed amendment does not involve significant hazards considerations as described in 10 CFR 50.92.

No other regulations or TS will be affected by the proposed amendment.

5.2.2. Updated Final Safety Analysis Report (UFSAR)

The CNP UFSAR does not specify STIs for the RTS/ESFAS instrumentation that is affected by this proposed license amendment, nor does it specify RTB or RTBB surveillance CT or BT. A study performed by CNP concluded that the proposed STI extensions will not necessitate an adjustment of the established setpoint calculation methodology. The proposed changes do not impact the instrumentation setpoints or method of operation of this instrumentation; therefore, the functions and performance requirements of this instrumentation are unaffected by the proposed amendment.

Consequently, the proposed amendment does not affect the ability of the RTS and ESFAS to fulfill their intended safety functions.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Precedent Regulatory Activity The NRC required that an applicant for a TS amendment to incorporate the generic changes evaluated in WCAP-10271-P-A and its supplements, and WCAP-14333-P-A, to AEP:NRC:2311 Page 22 must meet certain plant-specific conditions stipulated in the SERs and SSER. The generically-approved changes to CTs, BTs, and the analog channel STI specified in these WCAPs have not been amended into the CNP TS. Furthermore, the generic changes evaluated in WCAP-15376 are still being reviewed by the NRC, and have not yet been submitted by any licensees for plant-specific application at other nuclear plants.

However, the plant-specific conditions stipulated in the SERs and SSER for WCAP-10271-P-A and its supplements, and WCAP-14333-P-A, constitute part of the basis for the changes evaluated in WCAP-15376. I&M has addressed each of the conditions stipulated in the WCAP-10271-P-A and WCAP-14333-P-A SERs and SSER.

Additionally, licensee submittals to adopt the two earlier WCAPs and WOG correspondence pertaining to the NRC's review of WCAP-15376 were reviewed to ascertain the nature of the plant-specific issues that may be specified in an SER for WCAP-15376.

On June 16, 1999, Virginia Electric and Power Company (VEPCO) submitted a license amendment request (Reference 17) to extend the RPS and ESFAS analog instrumentation surveillance frequency from monthly to quarterly for the Surry Power Station, Units 1 and 2. This request, which was supplemented by licensee correspondence dated September 27, 2000, and June 6, 2001 (References 18 and 19), was approved by the NRC in an SER dated August 31, 2001 (Reference 20). Reference 17 included three tables, which were originally provided by the WOG to aid in the implementation of WCAP-14333-P-A. The tables in Reference 17 demonstrate the applicability of the WCAP-14333-P-A generic criteria to the Surry Power Station. I&M has modified these tables to demonstrate the applicability of the WCAP-15376 generic criteria to the CNP proposed changes. These implementation guideline tables are provided in Attachment 3 of this submittal.

6.0 ENVIRONMENTAL CONSIDERATION

I&M has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. I&M has determined that this license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes surveillance requirements, and the amendment meets the following specific criteria.

(i) The amendment involves no significant hazards consideration.

As demonstrated in Section 5.1, No Significant Hazards Determination, this proposed amendment does not involve a significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

to AEP:NRC:2311 Page 23 The proposed amendment involves a revision to the TS affecting the testing requirements for the RTS and ESFAS. The reactor protection circuitry, redundancy, and diversity is not changed. This change does not impact any systems or activities associated with processing or discharge of effluents. Therefore, this change does not change the types or increase the amounts of effluent that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

No new effluents or effluent release paths are created by the proposed TS changes to the RTS and ESFAS surveillance testing requirements. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure resulting from this change.

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. WCAP-15376, Revision 0, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," dated October 2000

2. NUREG-1431, Revision 0, "Standard Technical Specifications, Westinghouse Plants, Specifications, Volume 1," dated September 1, 1992

3. NUREG-1431, Revision 1, "Standard Technical Specifications, Westinghouse Plants, Specifications, Volume 1," dated April 1, 1995

4. NUREG-1431, Revision 2, "Standard Technical Specifications, Westinghouse Plants, Specifications, Volume 1," dated April 1, 2001

5. WCAP-10271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System;" dated May 1986

6. WCAP-10271-P-A, Supplement 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System;" dated July 1983

7. WCAP-10271-P-A, Supplement 2, Revision 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System;" dated June 1990 to AEP:NRC:2311 Page 24

8. Letter from C. C. Thomas (NRC) to J. J. Sheppard (WOG), "Acceptance for Referencing of Licensing Topical Report WCAP-10271, Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation Systems," dated February 21, 1985 (NRC Safety Evaluation for WCAP-10271-P-A and WCAP-10271-P-A, Supplement 1)

9. Letter from C. E. Rossi (NRC) to R. A. Newton (WOG-WEPC), "Westinghouse Topical Reports WCAP-10271, Supplement 2, and WCAP-10271, Supplement 2, Revision 1, Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," dated February 22, 1989 (NRC Safety Evaluation for WCAP-10271-P-A Supplement 2 and Supplement 2, Revision 1)

10. Letter from C. E. Rossi (NRC) to G. T. Goering (WOG-NSP), "Westinghouse Topical Reports WCAP-10271, Supplement 2, Revision 1, Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," dated April 30, 1990 (NRC Supplemental Safety Evaluation for WCAP-1027 1-P-A, Supplement 2, Revision 1)

11. WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," dated October 1998

12. Letter from T. H. Essig (NRC) to L. F. Liberatori Jr. (WOG), "Review of Westinghouse Owners Group Topical Reports WCAP-14333P and WCAP-14334NP, Dated May, 1995, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,"

dated July 15, 1998 (NRC Safety Evaluation for WCAP-14333-P-A)

13. Letter from R. H. Bryan (WOG) to the NRC, "Transmittal of Reports: WCAP-15376-P, Rev. 0, (Proprietary) and WCAP-15377-NP, Rev. 0 (Non-Proprietary), Entitled "Risk Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times" (MUHP-3 045)," dated November 8, 2000

14. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," dated July 1, 1998

15. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1, 1998

16. Letter from C. E. Rossi (NRC) to R. F. Janecek (BWR Owner's Group), "Staff Guidance for Licensee Determination that the Drift Characteristics for Instrumentation Used in RPS Channels are Bounded by NEDC-30851P Assumptions when the Functional Test Interval is Extended from Monthly to Quarterly," dated April 27, 1988 to AEP:NRC:2311 Page 25

17. Letter from W. R. Matthews (VEPCO) to the NRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed Technical Specifications and Bases Change - RPS and ESFAS Analog Instrumentation Surveillance Frequency Change from Monthly to Quarterly," dated June 16, 2000

18. Letter from L. N. Hartz (VEPCO) to the NRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed Technical Specifications and Bases Change

- RPS and ESFAS Analog Instrumentation Surveillance Frequency Change from Monthly to Quarterly," dated September 27, 2000

19. Letter from L. N. Hartz (VEPCO) to the NRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed Technical Specifications and Bases Change

- RPS and ESFAS Analog Instrumentation Surveillance Frequency Change from Monthly to Quarterly, Response to Request for Additional Information," dated June 6, 2001

20. 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

ATTACHMENT 1A TO AEP:NRC:2311 UNIT 1 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW PROPOSED CHANGES REVISED PAGES UNIT 1 1-9 3/43-4 3/43-5 3/4 3-8 3/4 3-12 3/43-13 3/43-14 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 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.

Atleast 124 dTys 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.

I N.A. Not Applicable.

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

TABLE 3.3-I (Continue) 0 REACTOR TRIP SYSTEM INSTRUMENTATION 0

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION FUNCTIONAL UNIT 4 2 3 6'

9. Pressurizer Pressure - Low 4 2 3 1,2 6'

10. Pressurizer Pressure -- High 1-,4 71

11. Pressurizer Water Level - 3 2 2 I

I- 12.

High Loss of Flow - Single Loop (Above P-8) 3/loop 2/loop in any operating loop 2/loop in each operating loop 1 7' 7#

13. Loss of Flow - Two Loops 3/loop 2/loop in two 2/loop in each 1 (Above P-7 and below P-8) operating loops operating loop

14. Steam Generator Water 3/loop 2/loop in any 2/loop in each 1,2 7' Level -- Low-Low operating loop operating loop 2/loop-level and 1/loop-level 1/loop-level and 1,2 7V

15. Steam/Feedwater Flow Mismatch and Low Steam 2/loop-flow coincident with 2/loop-flow Generator Water Level mismatch in same 1/loop-flow mis-match or loop mismatch in 2/loop-level and same loop 1/loop-flow mismatch N

0

n 0 TABLE 3.3-1 (Continued)

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

16. Undervoltage-Reactor Coolant 4-1/bus 2 3 1 6' Pumps

17. Underfrequency-Reactor 4-1/bus 2 3 1 6' "CoolantPumps

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 1 ESF

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

> operating loop

21. Reactor Trip Breakers 2 1 2 1, 2 4-r 13,M 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.

Witho fbin inoperable diverse trip features ACTION 13 (Undervoltage or shunt trip attachmen OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or -.

4n1IoT S l declare the breaker inoperable and apply A^ 4-ON 1. The breaker IjTabi 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 =hun 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.

WffI N5-7 N-thte nuirb~r-of'OPEPA-BLE Rýea~joi Tr~ip -lr~B-r-ýik i~ onlne le6s than iequixedbly the, Minimum Channels OPERABLE requirement for ~reasons other than an inoperable diverse trip1 feature, restore the inoperIable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT1 STANDBY'Nwithin the fiollowing 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 surveillanqce testngpe.§pecificattioti~Jj,3L'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" I amps. block of source range reactor trip.

AMEINJJMEIN I 9, +W, iqu NUCLEAR PLANT-UNIT 1I COOK NUCLEAR Page 3/4 3-8 Page 314 3-8 AMVEN*DM ENT 99,

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 CHECK CALIBRATION TEST REQUIRED FUNCTIONAL UNIT

1. Manual Reactor Trip A. Shunt Trip Function N.A. N.A. 1, 2, 3*, 4*, 5:

B. Undervoltage Trip N.A. N.A. S/U(4-t(10Yd() 1, 2, 3% 4% 5" Function V M and ( I 0)*

2. Power Range, Neutron S D(2,8), M(3,8), 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) 1,2 Flux, High Negative Rate Intermediate Range, S R(6,8) SIUf1-)U1) 1, 2, and*

5.

Neutron Flux

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

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

8. Overpower delta T S R(9) 1,2 Pressurizer Pressure S R 1,-2

9. -

Low S R 1,2

10. Pressurizer Pressure -

High

11. Pressurizer Water Level S R High

12. Loss of Flow-Single Loop S R(8) 1 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-12 AMENDMENT IN0, =1,, 4t24, 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 R 1,2 I

15. Steam/Feedwater Flow S 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(i) 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. n M(4) 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. =.on sM 1, 2, 3, 4, 5" (5)(1 1)-and B. Undervoltage Trip N.A. N.A. 1, 2, 3*, 4*, 5*

Function (5)(11)-an ont n M/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 S/U--( 13)0 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-13 AMENDMENT M4},1,20, 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 ea every other meoth=-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 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.

[TTT~pp ýwai37 ýeýacrtipbýtypas bereer i~q o2 Ermt reactor p[) i pea ipe yiT~ekr-ffa r a di COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-44 AMENDMENT 99, 4-20, 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. SM(2) N.A. 1,2,3,4

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

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

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

2. CONTAINMENT SPRAY

a. Manual Initiation See Functional Unit 9 b Automatic Actuation Logic N.A. N.A. SM(2) N.A. 1,2,3,4 C. Containment Pressure-- High S R F. M (3) N.A. 1,2,3 High COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-31 AMENDMENT 4-00, 420J, -144, M, 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 DEVICE WHICH CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE CHANNEL CHANNEL TEST TEST REQUIRED FUNCTIONAL UNIT CHECK CALIBRATION

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual ________ _______--See Functional Unit 9 - -

NA N A. SM (2) NA 1,2,3,4

2) From Safety Injection Automatic Actuation Logic

b. Phase "B" Isolation

1) Manual ------- - ,ie runctioni il Unit9 ---------

2) Automatic Actuation N A. N.A. SM (2) NA 1,2,3,4 Logic R SM (3) N.A. 1,2,3

3) Containment Pressure- S 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 I

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-32 AMENDMENT 4-00,4:44, M15, 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 See Functional Unit 9---

b Automatic Actuation Logic NA NA SM (2) N A. 1,2,3, c Containment Pressure- High S R SM (3) NA 1,2,3 High d Steam Flow in Two Steam S R N.A. 1,2,3 Lines-High Coincident with Tag--Low-Low e Steam Line Pressure-Low S R N.A. 1,2,3

5. TURBINE TRIP AND FEEDWATER ISOLATION a Steam Generator Water S R N.A. 1,2,3 Level--High-High 6 MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water Level--Low-Low S R N.A. 1,2,3 b 4 kv Bus Loss of Voltage S R M NA 1,2,3

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

d. Loss of Main Feed Pumps N A. N A. R NA 1,2 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-33 AMENDMENT 00, 4-20, 4-24,144 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 REQUIRED

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS

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

b. Reactor Coolant Pump Bus NA 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 4-00,4t-20, 144, 153

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

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 NA 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. F M (2) N A. 1,2,3

c. Containment Pressure - High S R S M (3) N A. 1,2,3 COOK NUCLEAR PLANT-UNIT I Page 3/4 3-33b AMENDMENT 4-53,Z*04, 234

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.3 INSTRUMENTATION TABLE 4 3-2 (Continued)

TABLE NOTATION (1) Deleted (2) Each train or logic channel shall be tested at least every other F 2: 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 -39,204

ATTACHMENT lB TO AEP:NRC:2311 UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW PROPOSED CHANGES REVISED PAGES UNIT 2 1-10 3/43-3 3/43-4 3/43-7 3/43-11 3/43-12 3/43-13 3/4 3-30 3/4 3-31 3/43-32 3/43-33

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 At least once per 184 days SA At least once per 549 days R

S/r Prior to each reactor start-up P Completed prior to each release I

N.A. Not Applicable COOK NUCLEAR PLANT-UNIT 2 Page 1-10 AMENDMENT 51 I

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

9. Pressurizer Pressure-Low 4 2 3 1,.2 6#

10. Pressurizer Pressure--High 4 2 3 1,2 6#

3 2 2 1,2 7#

11. Pressurizer Water Level-High 3/loop 2/loop in any 2/loop in each 1 7#

12. Loss of Flow - Single Loop (Above P-8) operating loop operating loop

13. Loss of Flow - Two Loops 3/loop 2/loop in two 2/loop in each (Above P-7 and below P-8) operating loops operating loop 1,2 I

14. Steam Generator Water Level-- 3/loop 2/loop in any 2/loop in each Low-Low operating loop operating loop

15. Steam/Feedwater Flow 2/loop-level and I/loop-level I/loop-level 1,2 7#

Mismatch and Low Steam 2/loop-flow coincident with and 2/loop Generator Water Level mismatch in same I/loop-flow flow mismatch loop mismatch in or 2/loop-level same loop and I/loop flow mismatch AMENDMENT 45,107 2 Page 3/4 3-3 COOK NUCLEAR PLANT-UNIT 2 COOK Page 3/4 3-3 AMENDMENT 45, 107

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 I Pumps

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

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

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

20. Reactor Coolant Pump Breaker Position Trip Above P-7 l/breaker 2 I/breaker per 1 11 operating loop 13

21. Reactor Trip Breakers 2 1 2 1,27 3*, 4*, 5* 14 2 1 2 1,2 1

22. Automatic Trip Logic 3*, 4*, 5* 14 172 Page 314 3-4 AMENDMENT 86, 40, 42, PLANT-UNIT 22 NUCLEAR PLANT-UNIT COOK NUCLEAR Page 3/4 3-4 AMENDMENT 86, :1W-, IV-, 172

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 .

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 ba6TdTr'p rne 7pera1- iv "o

features (Undervoltage or shunt trip attachIment) inoperabe, restore it e___ r a e to OPERABLEstatus within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or keiq -I a 6 hot declare the breaker inoperable and apply ACTION 1. The breaker anne shall not be bypassed while one of the diverse trip features is inoerable except for the time required for performing maintenance to restore the breaker j 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 Wihtenme fOEAL eco rpBreaker 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 4.3.1I.1I.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 10.11 amps. manual block of source range reactor trip.

AMENDMENT 86,127 Page 314 3-7 COOK NUCLEAR PLANT-UNIT 22 NUCLEAR PLANT-UNIT Page 3/4 3-7 AMENDMENT 86, 127

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES 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(4-I)X(1O 1, 2, 3%4% 5*

B. Undervoltage Trip Function N.A. N.A. S/U(-14.)'(t(7.j 1, 2, 3%, 4% 5*

2. Power Range, Neutron Flux S D(2,8), M(3,8), qLM and SýU(4!) 1, 2 and*

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) 1,2

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

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

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

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

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

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

11. Pressurizer Water Level -- High S R 1,4

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

0 TABLE 4.3-1 (Continued) 0 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST REQUIRED FUNCTIONAL UNIT S R(8) N.A. 1

13. Loss of Flow-Two Loops

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

15. Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level N.A. R M 1

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

17. Underfrequency-Reactor Coolant Pumps U 18. Turbine Trip N.A. S/U(1) 1,2 A. Low Fluid Oil Pressure N.A.

N.A. N.A. SIU(1) 1,2 B. Turbine Stop Valve Closure N.A.

M(4)*I3 1,2

19. Safety Injection Input from EFS N.A.

N.A. N.A. R N.A.

20. Reactor Coolant Pump Breaker Position Trip

21. Reactor Trip Breaker 1, 2, 3%, 4', 5*

N.A. N.A. 7 M(5)(11 A. Shunt Trip Function N.A. N.A. S(5)(11) 1, 2, 3*, 4*, 5*

B. Undervoltage Trip Function and S/U(i)(1 1)

N.A N.A. 1, 2, 3, 4, 5"

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

23. Reactor Trip Bypass Breaker s/U(-14(13) L

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

NOTATI ON

- 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 tiies every other menth 37s.

(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.

M E i1e that are ackF*candtc or i jifhe tot g ay COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-13 AMENDMENT 86,1t07, 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 CHECK CALIBRATION TEST TEST REQUIRED FUNCTIONAL UNIT 1 SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a Manual Initiation See Functional Unit 9 b Automatic Actuation NA NA f M(2) N A. 1,2,3,4 Logic

c. Containment Pressure S R SM (3) N A. 1,2,3 High S R NA 1,2,3

d. Pressurizer Pressure I Low S R NA 1,2,3 e Differential Pressure Between Steam Lines High R NA 1.2.3 f Steam Line Pressure S Low 2 CONTAINMENT SPRAY a Manual Initiation - See Functional Unit 9

b. Automatic Actuation NA NA M(2) NA 1,2,3,4 Logic c Containment Pressure S R SM (3) NA 1,2,3 High-High

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual See Functional Unit 9

2) From Safety NA NA. fM(2) NA 1,2,3,4 Injection Automatic Actuation Logic b Phase "B" Isolation I) Manual See Functional Unit 9

2) Automatic Actuation NA NA M(2) NA 1,2,3,4 Logic

3) Containment S R SM (3) NA 1,2,3 Pressure- High High I

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-30 AMENDMENT 34,434,43-7, 4158, 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 c Purge and Exhaust Isolation

1) Manual See Functional Unit 9

2) Containment S R Q N.A 1,2,3,4 Radioactivity - High 4 STEAM LINE ISOLATION

a. Manual See Functional Unit 9 b Automatic Actuation NA NA. *M(2) NA 1,2,3 Logic

c. Containment Pressure S R SM (3) NA 1,2,3 High-High d Steam Flow in Two Steam S R NA 1,2,3 Lines - High Coincident with Tavg - Low-Low e Steam Line Pressure S R N A. 1,2,3 Low 5 TURBINE TRIP AND FEEDWATER ISOLATION a Steam Generator Water S R N.A 1,2,3 Level - High-High 6 MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a Steam Generator Water S R N A. 1,2,3 Level - Low-Low b 4 kV Bus Loss of Voltage S R M N A. 1,2,3

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

d. Loss of Main Feed Pumps NA N.A R N A. 1,2 I COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-31 AMENDMENT 82, 97, M-t, 4-34, 4-3-7,4-59, 468, 224 I

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.3 INSTRUMENTATION TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUTREMENTS 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 FM NA 1,2,3

b. Reactor Coolant Pump Bus N.A 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 N.A 1,2,3,4 Voltage 9 MANUAL a Safety Injection (ECCS)

Feedwater Isolation N A. NA NA 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 N.A N.A NA 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. 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 b Automatic Actuation NA NA. (2) N A. 1,2,3 Logic

c. Containment Pressure S R (3) N A. 1,2,3 High COOK NUCLEAR PLANT-UNIT 2 Page 314 3-32 AMENDMENT 82, 9-7, 4-34, M-3, 4-59, 489, 217

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 234 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,-t-34, 189

ATTACHMENT 2A TO AEP:NRC:2311 UNIT 1 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED REVISED PAGES UNIT 1 1-9 3/43-4 3/43-5 3/43-8 3/4 3-12 3/43-13 3/4 3-14 3/4 3-31 3/43-32 3/4 3-33 3/4 3-33a 3/4 3-33b 3/4 3-34

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.

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 -7.2, 1 COOK NUCLEAR PLANT-UNIT 1 Page 1-9

¢n TABLE 3.3-1 (Continue) 0 0 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE z FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 4 2 1 6#

9. Pressurizer Pressure - Low 3

10. Pressurizer Pressure -- High 4 2 3 1,2

11. Pressurizer Water Level - 3 2 2 1 High

12. Loss of Flow - Single Loop 3/loop 2/loop in any 2/loop in each 1 (Above P-8) operating loop operating loop

13. Loss of Flow - Two Loops 3/loop 2/loop in two 2/loop in each 1 (Above P-7 and below P-8) operating loops operating loop

14. Steam Generator Water 3/loop 2/loop in any 2/loop in each 1,2 Level -- Low-Low operating loop operating loop

15. Steam/Feedwater Flow 2/loop-level and 1/loop-level 1/loop-level and 1,2 Mismatch and Low Steam 2/loop-flow coincident with 2/loop-flow mis Generator Water Level mismatch in same I/loop-flow match or 2/loop loop mismatch in level and I/loop same loop flow mismatch

-0

z

n 0

0 TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 3 1

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

17. Underfrequency-Reactor Coolant Pumps

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

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

20. Reactor Coolant Pump Breaker Position Trip Above P-7 l/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 I 3", 4", 5" 14

314 3/4.3 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS INSTRUMENTATION I 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 Mimmum 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 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. I 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"1 amps. block of source range reactor trip.

AMI*INUIVIEIN I J9, +U, 441J, I PLANT-UNIT I1 COOK NUCLEAR PLANT-UNIT Page 314 3-8 Page 3/4 3-8 AMIENDMENTI -99, 42-0, 40,

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 FUNCTIONAL SURVEILLANCE CHANNEL CHANNEL CHECK CALIBRATION TEST REQUIRED FUNCTIONAL UNIT

1. Manual Reactor Trip N.A. N.A. S/U(1O)(17) 1, 2, 3%, 4%, 5*

A. Shunt Trip Function 1, 2, 3%, 4%, 5*

N.A. N.A. S/U(1O)(17) I B. Undervoltage Trip Function

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

Flux and Q(6,8)

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

4. Power Range, Neutron Flux, High Negative Rate R(6,8) S/U(17) 1, 2, and *

5. Intermediate Range, S Neutron Flux

6. Source Range, Neutron S R(6,14) M(14) 2(7), 3(7), 4 and 5 I Flux R(9) SA 1,2

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

8. Overpower delta T S R SA 1

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

10. Pressurizer Pressure - S High S R SA 1

11. Pressurizer Water Level -

High S R(8) SA 1

12. Loss of Flow-Single Loop AMIENDMENT 1-00, t=O, V-4, 144, 1 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-12

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

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS CHANNEL MODE IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST REOUIRED FUNCTIONAL UNIT S R(8) N.A. 1

13. Loss of Flow-Two Loops

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

15. Steam/Feedwater Flow S R SA 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(l) 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 I 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 B. Undervoltage Trip N.A. N.A. 4 Months (5)(1 1) 1, 2, 3% 4*, 5*

Function

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

N.A. N.A. 4 Months 1, 2, 3*, 4*, 5*

23. Reactor Trip Bypass Breaker (5)(12)(16) and S/U(13)(17)

AMENDMENT t400, =-2,444,1 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-13

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

NOTAT ION 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.

- Compare incore to excore axial imbalance above 15% of RATED THERMAL POWER. Recalibrate (3) 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 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) - Applicable to any reactor trip bypass breakers that are racked in and closed for bypassing a reactor trip breaker.

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

(N)OT *NUICLAEAR PLANT-UNIT 1 . . .

Page 3/4 3-14 AMENDMENT 99, I-., 141, co o

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 FUNCTIONAL OPERATIONAL SURVEILLANCE CHANNEL CHANNEL TEST TEST REQUIRED FUNCTIONAL UNIT CHECK CALIBRATION

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

a. Manual Initiation See Functional Unit 9 N A. N.A. SA (2) N.A. 1,2,3,4

b. Automatic Actuation Logic

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

f. Steam Line Pressure-Low

2. CONTAINMENT SPRAY

a. Manual Initiation See Functional Unit 9 -

N.A N.A. SA (2) N.A. 1,2,3,4

b. Automatic Actuation Logic S R SA (3) N A. 1,2,3 C. Containment Pressure- High High COOK NUCLEAR PLANT-UNIT 1 Page 314 3-31 AMENDMENT 4-00, 1-20, 4 4-44, 4-53,24A4,

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 CHECK CALIBRATION TEST TEST REQUIRED FUNCTIONAL UNIT 3 CONTAINMENT ISOLATION

a. Phase "A" Isolation

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

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

1) Manual ------ --- - ee 1rufction2 dUnit 9 -----------

2) Automatic Actuation NA NA SA (2) NA 1,2,3,4 Logic S9 R SA (3) N.A. 1,2,3

3) Containment Pressure High-High c Purge and Exhaust Isolation

1) Manual Functional Unit 9 ---

2) Containment S R Q NA 1,2,3,4 Radioactivity--High COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-32 AMENDMENT 4-00,4-44,4t53,4"8-3,

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

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

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 Level-Low-Low S R SA N.A. 1,2,3 I b 4 kv Bus Loss of Voltage S R M N A. 1,2,3 I

c Safety Injection N.A. N.A SA (2) NA 1,2,3 d Loss of Main Feed Pumps N A. N A. R NA 1,2 COOK NUCLEAR PLANT-UNIT 1 Page 314 3-33 AMENDMENT 4-00, 4-20, 4-21,444 4-53, 2U4,

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 CHECK CALIBRATION TEST TEST REQUIRED FUNCTIONAL UNIT

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS

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

b. Reactor Coolant Pump Bus N A.

Undervoltage 8 LOSS OF POWER a 4 kv Bus Loss of Voltage S R M N.A 1,2,3,4 S R M N A. 1,2,3,4 b 4 kv Bus Degraded Voltage AMENDMENT 4-00,4t-2-0,444,4t-53, 1 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-33a

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. NA 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 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 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--5-3,2"4, 2M3,

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 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 I Page 314 3-34 AMENDMENT-39,M*,

ATTACHMENT 2B TO AEP:NRC:2311 UNIT 2 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED REVISED PAGES UNIT 2 1-10 3/43-3 3/43-4 3/43-7 3/43-11 3/43-12 3/4 3-13 3/4 3-30 3/4 3-31 3/43-32 3/4 3-33

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 I 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-5-1, I

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 CHANNELS TO TRIP OPERABLE MODES ACTION FUNCTIONAL UNIT 4 2 3 I 6#1

9. Pressurizer Pressure-Low

10. Pressurizer Pressure-High 4 2 3 1,2

11. Pressurizer Water Level-High 3 2 2 7 # 1

12. Loss of Flow - Single Loop 3/loop 2/loop in any 2/loop in each 1 (Above P-8) operating loop operating loop

13. Loss of Flow - Two Loops 3/loop 2/loop in two 2/loop in each 7#

(Above P-7 and below P-8) operating loops operating loop

14. Steam Generator Water Level 3/loop 2/loop in any 2/loop in each 1,2 Low-Low operating loop operating loop 2/loop-level and I/loop-level I/loop-level 1,2

15. Steam/Feedwater Flow Mismatch and Low Steam 2/loop-flow coincident with and 2/loop Generator Water Level mismatch in same 1/loop-flow flow mismatch loop mismatch in or 2/loop-level same loop and I/loop flow mismatch AMENDMENT 45, 107, Page 3/4 3-3 PLANT-UNIT 22 NUCLEAR PLANT-UNIT COOK NUCLEAR Page 3/4 3-3 AMENDMENT 45,4:-07, I

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 CHANNELS TO TRIP OPERABLE MODES ACTIO FUNCTIONAL UNIT 4-1/bus 2 3 1 6#

16. Undervoltage-Reactor Coolant Pumps 4-1/bus 31 6#

17. Underfrequency-Reactor 2 Coolant Pumps

18. Turbine Trip 3 2 1 7#

A. Low Fluid Oil Pressure 2 4 4 3 1 6#

B. Turbine Stop Valve Closure 1 2 1,2

19. Safety Injection Input from 2 ESF

20. Reactor Coolant Pump Breaker Position Trip 1/breaker per 11 Above P-7 1/breaker 2 operating loop 1,2 13, 15

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

22. Automatic Trip Logic 3*, 4*, 5* 14 AMENDMENT 86, 4()7, 42, m,

Page 314 3-4 COOK NUCLEAR PLANT-UNITPLANT-UNIT 22 Page 3/4 3-4 AMENDMENT 86, 1W0, I---, 1-72, 1

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 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. I 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, 47, PLANT-UNIT 22 Page 3/4 3-7 AMENDMENT 86, 7, COOK NUCLEAR PLANT-UNIT Page 3/4 3-7

TABLE 4.3-1 0

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

1. Manual Reactor Trip z N.A. N.A. S/U(10)(17) 1, 2,3%4%, 5 A. Shunt Trip Function 1, 2,3, 4, 5" N.A. N.A. S/U(10)(17)

B. Undervoltage Trip Function S D(2,8), M(3,8), Q 1, 2 and"

2. Power Range, Neutron Flux and Q(6,8)

N.A. R(6) Q 1,2

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

4. Power Range, Neutron Flux, High Negative Rate S R(6,8) S/U(17) 1, 2, and

5. Intermediate Range, Neutron Flux S R(6,14) M(14) 2(7), 3(7), 4 and 5

6. Source Range, Neutron Flux S R(9) SA 1,2

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

8. Overpower AT S R SA 1

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

10. Pressurizer Pressure -- High z S R SA 1

11. Pressurizer Water Level -- High S R(8) SA 1

12. Loss of Flow-Single Loop

-i I

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

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

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

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

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

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

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

1, 2, 3*, 4*, 5*

a. B. Undervoltage Trip Function

22. Automatic Trip Logic N.A.

N.A N.A.

N.A.

4 Months (5)(11)

SA (15) 1, 2, 3%, 4*, 5*

23. Reactor Trip Bypass Breaker N.A. N.A. 4 Months (5)(12)(16) 1, 2, 3%4*, 5*

and S/U(13)(17)

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).

- The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the (11) 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.

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

(16)

- Applicable to any reactor trip bypass breakers that are racked in and closed for bypassing a reactor trip

[

breaker.

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

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-13 AMENDMENT 86,:0q,4t-18,

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 I. 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) NA 1,2,3 High d Pressurizer Pressure S R SA NA 1,2,3 Low e Differential Pressure S R SA NA 1,2,3 Between Steam Lines High f Steam Line Pressure S R SA N-A 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) N A. 1,2,3 High-High 3 CONTAINMENT ISOLATION a Phase "A" Isolation

1) Manual See Functional Unit 9

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

1) Manual See Functional Unit 9

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

3) Containment S R SA (3) N A. 1,2,3 Pressure- High High COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-30 AMENDMENT -34,-134, t3-7,-1-58, am4,

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

c. Purge and Exhaust Isolation

1) Manual See Functional Unit 9

2) Containment S R Q N.A 1,2,3,4 Radioactivity - High 4 STEAM LINE ISOLATION 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) 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.,g - Low-Low e Steam Line Pressure S R SA NA 1,2,3 Low 5 TURBINE TRIP AND FEEDWATER ISOLATION a Steam Generator Water S R SA N.A. 1,2,3 I 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 N A. 1,2,3 c Safety Injection NA NA SA (2) N A. 1,2,3 I

d. Loss of Main Feed Pumps N A. NA R N A. 1,2 COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-31 AMENDMENT 82, 97, 4-34,434, 4-37,459, 468, 224,

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.3 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 SA NA 1,2,3 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 S R M NA 1,2,3,4 Voltage 9 MANUAL a Safety Injection (ECCS)

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

c. Containment Isolation - NA N.A 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 N A. N A. NA R 1,2,3,4 Recirculation Fan 10 CONTAINMENT AIR RECIRCULATION FAN a Manual -See Functional Unit 9 b Automatic Actuation N.A N A. SA (2) N.A 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, 134, 4-3-7, 459,4-89, 1-,

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 IAMENDMENT 8-Z, 4-34, M8,

ATTACHMENT 3 TO AEP:NRC:2311 APPLICABILITY OF WCAP-15376 ANALYSES The tables in this attachment list the important parameters and assumptions made in the generic analysis that are relevant to the surveillance test intervals (STI), completion time (CT), and bypass time (BT) evaluation. Information is also provided for the current calculated core damage frequency (CDF), and the contribution to CDF from the anticipated transients without scram (ATWS) events. The values used for determining CDF are based on those reported to the Nuclear Regulatory Commission (NRC) in response to the Individual Plant Examination (IPE) per Generic Letter (GL) 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54(f)" (Reference 1), as updated through mid-2001.

The information provided in the attached tables demonstrates the applicability of the generic WCAP-15376, Revision 0, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times" (Reference 2), analysis to Donald C. Cook Nuclear Plant (CNP). These tables are based on implementation guidelines that were issued by the Westinghouse Owner's Group (WOG) for licensees implementing the Technical Specification allowed outage time (AOT) changes that were justified in WCAP-14333-P-A, Revision 1 (Reference 3). The tables were modified, as appropriate, to reflect the information specified in WCAP-15376.

REFERENCES

1. Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities 10 CFR 50.54(f)," dated November 23, 1988

2. WCAP-15376, Revision 0, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," dated October 2000

3. WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," dated October 1998 to AEP:NRC:2311 Page 2 Table 3.1 WCAP-15376 Implementation Guidelines: Applicability of the Analysis GeneralParameters Parameter WCAP 15376 Analysis CNP Plant-specific Logic Cabinet Type (1) Relay & SSPS SSPS Component Test Intervals (2)

Analog Channels 6 months 1 month (currently) c")

6 months (proposed) (15)

Logic Cabinets (SSPS) 6 months 1 month (currently) 6 months (proposed) 6 months Infrequent (12)

Master Relays (SSPS) 3 months Infrequent (12)

Slave Relays Reactor Trip Breakers 4 months 1 month (currently) 4 months (proposed)

Analog Channel Calibrations (3) yes Infrequent (13)

Done At-Power (per Periodic Tests) 18 months 18 months (16)

Interval Typical At-Power Maintenance Intervals (4)

Analog Channels 24 months Infrequent Logic Cabinets (SSPS) 18 months Infrequent Master Relays (SSPS) Infrequent (5) Infrequent Slave Relays Infrequent (5) Infrequent Reactor Trip Breakers 12 months Infrequent AMSAC (6) Credited for AFW pump start Provides AFW pump start Risk Comparison Total Transient Event Frequency (7) 3.6 events/calendar year (14) 1.34 events/calendar year -Unit 1 1.34 events/calendar year -Unit 2 ATWS Contribution to CDF (current 8.4E-06 events/reactor year (14) 6.8E-07 events/reactor year-Unit 1 PRA model) (8) 5.8E-07 events/reactor year- Unit 2 Total CDF from Internal Events (current 5.8E-05 events/reactor year (14) 4.9E-05 events/reactor year-Unit 1 PRA model) (9) 4.9E-05 events/reactor year-Unit 2 Total CDF from Internal Events (IPE) (1) Not Applicable 7.1E-05 events/reactor year-Unit 1 7.1E-05 events/reactor year-Unit 2 Notes:

1. Indicates the type of logic cabinet; SSPS or Relay (both are included in WCAP-15376).

2. Applicable test intervals are provided for CNP, Units 1 and 2. With the exception of the Power Range Neutron Flux and associated rate trip instrumentation, the proposed test intervals will be equal to those used in WCAP-15376; therefore, the WCAP-15376 analysis is considered applicable to CNP.

to AEP:NRC:2311 Page 3

3. Indicates if channel calibration is done at-power. If channel calibrations are done at-power, the calibration interval is provided. If channel calibrations are not done at-power, or if the calibration interval is equal to or greater than that used in WCAP-15376, the analysis is applicable. For CNP, the Power Range, Neutron Flux channel calibrations are performed quarterly at power. The frequency for these channel calibrations is not changed by the proposed license amendment. Other channels are only calibrated at-power infrequently. Therefore, the WCAP-15376 analysis is applicable.

4. Indicates applicable typical maintenance intervals for CNP. At CNP, at-power maintenance on this equipment is only performed infrequently (i.e., intervals are equal to or greater than those used in WCAP-15376); therefore, the analysis is applicable to CNP.

5. While at power, only corrective maintenance is done on the master and slave relays. The maintenance interval on typical relays is relatively long (i.e., experience has shown they typically do not fail completely). Failure of slave relays usually involves failure of individual contacts. Because "infrequent" slave relay failures are the norm at CNP, the WCAP-15376 analysis is applicable to CNP.

6. Because AMSAC will initiate AFW pump start at CNP, the WCAP-15376 analysis is applicable to CNP.

7. Includes total frequency for initiators requiring a reactor trip signal to be generated for event mitigation. This is required to assess the importance of ATWS events to CDF. Does not include events initiated by a reactor trip.

8. Indicates the ATWS contribution to CDF (from at-power, internal events). This is required to determine if the ATWS event is a large contributor to CDF.

9. Indicates the total CDF from internal events (including internal flooding) for the most recent PRA model update. This is required for comparison to the NRC's risk-informed CDF acceptance guidelines.

10. Indicates the total CDF from internal events from the IPE model referenced in Generic Letter 88-20 response. This value differs from the most recent PRA model update CDF, so a concise list of reasons has been provided, in bulletized form, describing the differences between the models that account for the change in CDF.

See Appendix A, "A Description of the Major Differences Between the Individual Plant Examination (IPE) and Current Probabilistic Risk Analysis (PRA) Model," to this Attachment for a list and discussion of PRA model updates.

11. The current analog channel test interval is I month, as the STI increase evaluated in WCAP-10271 has not been implemented at CNP. However, the analyses of WCAP-14333-P-A and WCAP-15376 still remain applicable.

12. The master and slave relays at CNP do not have TS requirements.

to AEP:NRC:2311 Page 4

13. During normal operation, only the Power Range, Neutron Flux nuclear instrumentation has an at power channel calibration. The frequency for this calibration is quarterly, and will not be changed by this licensing activity.

14. The risk analysis of WCAP-14333-P-A, formed the basis for the risk analysis of WCAP-15376; therefore, the specified risk values are the same as the WCAP-14333-P-A values.

15. The channel functional test surveillance intervals for applicable RTS and ESFAS functional units are proposed to be extended to semi-annual with the exception of Power Range, Neutron Flux (FU 2) and the High Positive Rate and High Negative Rate Power Range, Neutron Flux trips (FU 3 and 4), which are proposed to be extended to quarterly.

16. The channel calibration for the Power Range, Neutron Flux (FU 2) trip is performed quarterly, and will remain at the current quarterly interval.

to AEP:NRC:2311 Page 5 Table 3.2 WCAP-15376 Implementation Guidelines: Applicability of Analysis (Continued)

Reactor Trip Actuation Signals Event WCAP 15376 Analysis CNP Plant-specific Assumption Parameter (1)

Signal Actuation Source Large LOCA Not Required N/A Medium LOCA Not Required N/A Small LOCA Nondiverse (2) w/OA (3) Agree Steam Generator Tube Rupture Nondiverse w/OA Agree Interfacing System LOCA Not Required N/A Reactor Vessel Rupture Not Required N/A Secondary Side Breaks Inside Nondiverse w/OA Agree Containment Secondary Side Breaks Outside Nondiverse w/OA Agree Containment Transient Events, such as:

Positive Reactivity Insertion Diverse (4)w/OA Agree Loss of Reactor Coolant Flow Diverse (4)w/OA Agree Total or Partial Loss of Main Diverse (4)w/OA Agree Feedwater Loss of Condenser Diverse (4)w/OA Agree**

Turbine Trip Diverse (4)w/OA Agree Loss of 250 VDC Bus Diverse (4)w/OA Agree Loss of two 120VAC Vital Diverse (4)w/OA Agree Instrument Panels Loss of Instrument Air Diverse (4)w/OA Agree**

Inadvertent Opening of a Steam Diverse (4)w/OA Agree**

Valve Spurious Safety Injection Diverse (4)w/OA Agree Reactor Trip Generated by RPS Agree Loss of Offsite Power Not Required by RPS N/A Station Blackout Not Required by RPS N/A Loss of Service Water or Component Nondiverse w/OA Agree**

Cooling Water

- Events without automatic protection are addressed by procedure, which directs a reactor trip when required to maintain plant control or safety margins.

to AEP:NRC:2311 Page 6 Notes:

1. "Agree" indicates that CNP's plant design and operation is consistent with the WCAP-15376 analysis, that is, the noted reactor trip signals are available at a minimum. As "agree" is listed for each applicable event, the WCAP-15376 analysis is applicable to CNP.

2. Nondiverse means that (at least) one signal will be generated to initiate reactor trip for the event.

3. "OA" indicates that an operator could take action to initiate reactor trip for the event, that is, there is sufficient time for action and procedures are in place that will instruct the operator to take action.

4. Diverse means that (at least) two signals will be generated to initiate reactor trip for the event.

to AEP:NRC:2311 Page 7 Table 3.3 WCAP-15376 Implementation Guidelines: Applicability of Analysis (Continued)

EngineeredSafety FeaturesActuation Signals Safety Event WCAP 15376 Analysis Assumption Plant-specific Function Signal Actuation Source Parameter (1)

Safety Large LOCA Nondiverse Signal (2) Agree Injection Medium LOCA Nondiverse Signal, Agree OA(3) by SI Switch on Main Control Board Small LOCA Nondiverse Signal, Agree OA by SI Switch on Main Control Board, OA of Individual Components Interfacing Systems Nondiverse Signal, Agree LOCA OA by SI Switch on Main Control Board, OA of Individual Components SG Tube Rupture Nondiverse Signal, Agree OA by SI Switch on Main Control Board, OA of Individual Components Secondary Side Breaks Nondiverse Signal, Agree OA by SI Switch on Main Control Board, OA of Individual Components Auxiliary Events Generating SI Pump Actuation on SI Signal Agree Feedwater Signal Pump Start Nondiverse Signal Agree Transient Events AMSAC Transient__Events Operator Action Main Secondary Side Breaks Nondiverse Signal Agree Feedwater Isolation Steamlme Secondary Side Breaks Nondiverse Signal Agree Isolation Containment All Events Nondiverse Signal Agree Spray Actuation Containment All Events From SI Signal Agree Isolation Containment All Events From SI Signal Agree Cooling Notes:

1. "Agree" indicates that CNP's plant design and operation is consistent with the WCAP-15376 analysis, that is, the noted engineered safety features actuation signals are available at a minimum.

As each event "agrees" with the corresponding event assumed in WCAP-15376, the WCAP-15376 analysis is applicable to CNP.

to AEP:NRC:2311 Page 8

2. Nondiverse means that (at least) one signal will be generated to initiate the engineered safety feature noted for the event.

3. "OA" indicates that an operator could take action to initiate the engineered safety feature for the event, that is, there is sufficient time for action and procedures are in place that will instruct the operator to take action.

to AEP:NRC:2311 Page 9 Appendix A A Description of the Major Differences Between the Individual Plant Examination (IPE) and Current Probabilistic Risk Analysis (PRA) Model Indiana Michigan Power Company (I&M) submitted the initial CNP IPE to the NRC staff for review on May 1, 1992 (Reference 1). The IPE data analysis was revised in a model completed in June 1994. In response to the NRC's requests for additional information, the Human Reliability Analysis (HRA) was revised in June 1994 and October 1995, and submitted to the NRC on October 26, 1995 (Reference 2). The NRC staff evaluation report was sent to I&M on September 6, 1996 (Reference 3), and concluded that the IPE satisfied the requirements of Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities 10 CFR 50.54(f)," dated November 23, 1988, and the guidance given in NUREG-1335, "Individual Plant Examination: Submittal Guidance, Final Report," dated August 1, 1989.

Revisions to the CNP IPE not considered in the staff evaluation report included updating plant specific data in May 1996, changing from a linked fault tree model to a top logic model in August 1997, and a major revision in 2001. The table below provides a summary of these changes.

Rev. H/H1 PA-96-03 Rev. HH 2001 PRA Version Rev. 0 Rev. I Update Converted Major Revised Updated tontop Rvson Reason for Submittal of Initial Updated Data HRA Data to Top Revision revision Ini Data ho g Data Logic (described PE Analysis Methodology Analysis Model below)

Date of May-92 Jun-94 Oct-95 May-96 Aug-97 Jun-01 revision Summary of the 2001 PRA Update The following general changes were made to the PRA model:

The existing Computer Aided Fault Tree Analysis (CAFTA) model was converted to a WinNUPRATM model to better support implementation of Safety MonitorTM for on-line and shutdown risk evaluation.

The PRA was updated to include new plant specific data, procedure and/or design changes, revision of the treatment of common cause failures to comply with the latest methodology, and removal of conservative assumptions and simplifications.

to AEP:NRC:2311 Page 10 The IPE was a single unit model and applied only to an operating unit. The 2001 update created a dual unit model including inter-unit dependencies and spanned all modes of operation (operating and shutdown). This effort included the development of Safety MonitorTM full power models based on the updated PRA and development of and inclusion of a shutdown risk model, which can be used to support assessment and management of shutdown risk.

The following specific changes were made:

A. Initiating Events Large-break and medium-break loss of coolant accidents (LOCAs), steam generator tube ruptures, and steam line breaks were subdivided into the individual contributions from each loop and four separate initiating events were evaluated for each of these categories.

Initiators for loss of a single direct current train were added for each train separately.

The loss of offsite power initiator was divided into loss of offsite power to a single unit and loss of offsite power to both units (dual unit loss of offsite power) to improve modeling of the unit crossties.

Similarly, loss of ESW was split to consider the loss of a single unit's ESW separately from a total (dual unit) loss of ESW to improve modeling of the unit crossties.

Initiating event frequencies were reassessed based on updated plant-specific data and new generic data. In addition, a number of the frequencies were obtained from models built into the overall PRA as transfers from other initiators. The initiators included:

1. Consequential medium-break and small-break LOCAs resulting from a reactor coolant system power operated relief valve or safety relief valve failing to reclose.

2. Station blackouts.

3. Anticipated Transients Without Scram (ATWS) events.

Also, several initiating event frequencies were obtained from detailed system models:

1. Loss of ESW to a single unit.

2. Loss of ESW to both units.

3. Loss of component cooling water (CCW).

4. Loss of 250 volt direct current busses.

to AEP:NRC:2311 Page I11 B. Fault Trees The fault tree models were revised to incorporate design changes and operational changes.

Individual component common cause groups were identified for Multiple Greek Letter method common cause analysis.

The models were revised to support the implementation of Safety MonitorTM.

The heat removal function was removed from the recirculation model, and this function was included in a separate long term cooling model.

Extensive changes were made in the ESW system model to properly account for interactions between units for this shared system.

The 4160 volts alternating current system model was changed to address a reconfiguration of the reserve auxiliary transformers.

C. Reliability and Unavailability Data Revision of component failure data analysis included collecting and analyzing more recent CNP failure data for the time period since the previous update and the enhancement of common cause failure data for all components.

D. Human Reliability Analysis Evaluation of human error probabilities were limited to those affected by changes in procedures or were new to the updated model. The principal re-evaluation involved the revised Emergency Operating Procedure for switching to cold leg recirculation.

The revised procedure for a loss of CCW was also used to update the associated human error probabilities.

The net result was to add or revise 30 human error probabilities (20% of the total human interaction events).

Results The CDF is less than that from the 1995 update of 7.14E-05 per year. This can be attributed to a number of factors including a reduction in LOCA-initiating event frequencies, the removal of conservative assumptions, and the more detailed and complete modeling of ESW crossties between units.

to AEP:NRC:2311 Page 12 The Unit 2 results are almost identical to those for Unit 1 with the differences being due to minor differences in power supply arrangements to support systems and ATWS unfavorable exposure times.

The distribution of the contributions to the results has changed from the 1995 update. The station blackout contribution is now 36% of the total CDF and is higher than the 1995 result.

Sequences related to a loss of all ESW contribute approximately 24% of the total CDF. The most significant contributors are loss of ESW either as the initiator or following a normal transient initiator with subsequent loss of ESW combined with failure to recover ESW.

Small-break LOCA is still an important contributor (17%) to CDF. The importance of small break LOCA has decreased from the 1995 evaluation due to the reduced initiator frequency. The contribution to the total of steam generator tube ruptures has been reduced due to more detailed modeling while the contribution from steamline breaks has gone up because of an increase in assessed secondary side pipe break frequency.

The dominant contributors to LERF are loss of offsite power initiated sequences that make up approximately 50% of the total. Steam generator tube ruptures, loss of ESW, and small-break LOCAs each contribute about 10% to the total LERF.

Peer Review The 2001 update was the model provided to the Westinghouse Owners Group PRA Peer Review Team for review. The PRA peer review was performed in September 2001.

The summary of strengths and areas for improvement has been extracted from the draft report and is provided below.

Strengths

"* PRA Notebooks are mostly well constructed and useful.

"* Good interaction with plant personnel/functions, good input into Human Reliability Analysis.

"* Use of a Risk-Informed Steering Committee.

"* Broad scope PRA and Information Tools (e.g., On-Line Safety MonitorTM, shutdown model, external events models).

"* Strong attention to detail in the modeling and quantification process and documentation.

"* Highly sophisticated single fault tree model able to be used for PRA or Safety MonitorTM quantification.

to AEP:NRC:2311 Page 13 Areas for Improvement

"* Better estimate success criteria analyses to remove conservatisms.

"* Should re-create some of the analytical bases for IPE success criteria.

"* Internal flooding analysis should be updated.

"* Common cause process could be improved, and plant-specific common cause screening should be considered.

"* Highly sophisticated single fault tree model able to be used for PRA or Safety MonitorTM quantification requires high degree of attention to quantification process.

All elements of the PRA Peer Review received a grade of contingent 3. An aspect of a grade 3 is that the PRA can be used in licensing submittals to the NRC to support positions concerning absolute levels of safety significance if supported by deterministic evaluations. The contingent designation indicates that the peer review team identified Facts and Observations (F&O) that either require resolution immediately (Level A) or at the next scheduled update (Level B) for an element. Once the necessary F&O(s) for an element are resolved, the grade is no longer considered to be contingent. I&M has undertaken an update to the PRA model to address all of the F&Os leading to the contingent designation.

Facts and Observations The peer review team identified four Level A F&Os, the most significant, and 24 Level B F&Os.

Many of the F&Os identified by the peer review team were resolved shortly after the review team completed their evaluation by providing additional information or explanation to support the analysis (i.e., no changes to the model were required). For example, to resolve many of the common cause failure (CCF) issues, CNP obtained clarification from the author of NUREG/CR 5485. The clarification provided by the author resolved the issue with no model changes required. None of the Level A or Level B F&Os are relevant to the RTS or ESFAS.

References:

1. Letter from E. E. Fitzpatrick, I&M, to U. S. Nuclear Regulatory Commission Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Individual Plant Examination Submittal Response to Generic Letter 88-20," AEP:NRC:1082E, dated May 1, 1992

2. Letter from E. E. Fitzpatrick, I&M, to U. S. Nuclear Regulatory Commission Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Individual Plant Examination Response to NRC Audit Concerns and Request for Additional Information,"

AEP:NRC:10820, dated October 26, 1995

3. Letter from J. B. Hickman, NRC, to E. E. Fitzpatrick, I&M, "Review of D. C. Cook Individual Plant Examination Submittal - Internal Events (TAC Nos. M74398 and M74399)," dated September 6, 1996

ATTACHMENT 4 TO AEP:NRC:2311 DETAILED

SUMMARY

AND JUSTIFICATION OF PROPOSED TECHNICAL SPECIFICATION CHANGES The proposed surveillance test interval (STI), completion time (CT), bypass time (BT), and associated changes to the Unit 1 and Unit 2 Technical Specifications (TS) are presented by the TS Table Number. The affected TS tables are:

"* Table 1.2, "Frequency Notation"

"* Table 3.3-1, "Reactor Trip System Instrumentation"

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

"* Table 4.3-2, "Engineered Safety Features Actuation System Instrumentation Surveillance Requirements" The changes are individually addressed in the tables presented in this attachment.

to AEP:NRC:2311 Page 2 Table 4.1 Proposed TS Definition Changes Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 1-9 Add to Notation column: This change is supported by WCAP- 15376 (Reference 1) 2 1-10 "4 Months" (Table 1.2) 1 1-9 Add to Frequency This change is supported by WCAP-15376.

2 1-10 column:

(Table 1.2) "At least once per 124 1 days" Table 4.2 Proposed Applicable Mode, and Reactor Trip Breaker (RTB) CT and BT Changes Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/43-4 Functional Unit (FU) 9- NUREG-1431, Revision 1, Table 3.3.1-1, referenced in 2 3/4 3-3 Pressurizer Pressure- WCAP-15376 contains FU 8.a., Pressurizer Pressure-Low.

(Table 3.3-1) Low: Only Mode 1 is listed in the "Applicable Modes" column.

Delete reference to Donald C. Cook Nuclear Plant's (CNP) FU 9 has a P-7 Mode 2 in the "Applicable Modes" interlock (Updated Final Safety Analysis Report (UFSAR)

"Aclumn. MTable 7.2-1). Per CNP TS Table 3.3-1, the P-7 interlock column, blocks the FU 9 reactor trip logic signal below 11% of rated thermal power, which encompasses Mode 2 (defined as less than or equal to 5% of rated thermal power). Therefore, CNP FU 9 does not perform a reactor trip function in Mode 2, and Mode 2 should be removed from the "Applicable Modes" column.

This is consistent with NUREG-1431, Revisions 1 and 2 (Reference 2, 3) 1 3/43-4 FU 11 -Pressurizer NUREG-1431, Rev. 1, Table 3.3.1-1, referenced in 2 3/4 3-3 Water Level-High: WCAP-15376, contains FU 9, Pressurizer Water Level (Table 3.3-1) Delete reference to High. Only Mode I is listed in the "Applicable Modes" Mode 2 in the column.

"Applicable Modes" CNP's FU II has a P-7 interlock (UFSAR Table 7.2-1).

column. Per CNP TS Table 3.3-1, the P-7 interlock blocks the FU 11 reactor trip logic signal below 11% of rated thermal power, which encompasses Mode 2 (defined as less than or equal to 5% of rated thermal power). Therefore, FU 11 does not perform a reactor trip function in Mode 2, and Mode 2 should be removed from the "Applicable Modes" column.

This is consistent with NUREG-1431, Revisions 1and 2.

to AEP:NRC:2311 Page 3 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/4 3-5 FU 21 - RTB, including Proposed Action 15 is specific to the RTBs and RTBBs 2 3/4 3-4 the Reactor Trip Bypass when racked in and closed for bypassing an RTB in (Table 3.3-1) Breakers (RTBB): Modes 1 and 2.

Add reference to Action This change is supported by WCAP-15376.

15 in the "Action" column.

1 3/4 3-5 FU 21 - RTBs, including Due to the proposed changes, reference to Action 1 will no 2 3/4 3-4 the RTBBs: longer be applicable for FU 21 and should be deleted. A (Table 3.3-1) Delete reference to new Action 15 has been proposed to replace Action 1.

Action 1 in the "Action" Proposed Action 15 is specific to the increased CT and BT column. evaluated in WCAP-15376 for the RTBs, and RTBBs, when racked in and closed for bypassing an RTB in Modes 1 and 2.

This change is supported by WCAP-15376.

1 3/4 3-5 FU 21 - RTB, including This is an editorial change to clarify that current Action 13 2 3/4 3-4 the RTBBs: and proposed Action 15 only apply to Modes 1 and 2, and (Table 3.3-1) Delete the comma after current Action 14 only applies to Modes 3, 4 and 5.

"2" in the Unit 2 "Appicable Modes2This change is consistent with the NUREG-1431 STS and Applicable Modes WCAP-15376.

column, and increase the vertical space between the lines for Modes 1, 2 and Modes 3, 4, 5.

1 3/4 3-5 FU 22 - Automatic Trip Thus is an editorial change to clarify that Action 1 applies to 2 3/4 3-4 Logic: Increase the Modes land 2, and Action 14 only applies to Modes 3, 4, (Table 3.3-1) vertical space between and 5.

the lines for Modes 1,2 This change is consistent with the NUREG-1431 STS and and Modes 3, 4, WCAP-15376.

1 3/4 3-8 Action Statements: Action 13 currently specifies the required actions to be 2 3/4 3-7 Change Action 13 to taken when an RTB channel is determined to be inoperable (Table 3.3-1) reflect RTB channel as a result of an inoperable diverse trip feature. The TS inoperability condition currently specifies that this Action will be taken when the when a diverse trip diverse trip feature is determined to be inoperable and that feature is inoperable, and the breaker will be declared inoperable after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , at to specify actions which time Action I will be applied. The change is required when AOT is proposed to clarify that the action is applied to the exceeded. Functional Unit (i.e., the RTB channel), rather than the diverse trip feature. The proposed change specifies that the RTB channel is declared inoperable as a result of an inoperable diverse trip feature. Additionally, as a result of deleting Action 1 from the RTB requirements in Table 3.3 1, the reference to Action 1 is deleted from Action 13, and replaced with the equivalent text.

This proposed change is consistent with the NUREG-1431 STS methodology, and is supported by WCAP- 15376.

to AEP:NRC:2311 Page 4 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/4 3-8 Action Statements: The proposed Action 15 will incorporate a 24-hour AOT 2 3/4 3-7 Add Action 15 specific and an allowance for bypassing one channel for up to 4 (Table 3.3-1) for the RTBs, including hours for surveillance testing per TS 4.3.1.1.1, provided the the RTBBs. other channel is OPERABLE. The proposed 24-hour AOT is consistent with the WCAP-15376 evaluation. The 4-hour BT is consistent with the WCAP-15376 evaluation, as well as the TS mark-ups provided with Technical Specification Traveler Form (TSTF) 411, Revision 1 (Reference 5),

which proposes to revise the STS to reflect the changes evaluated in WCAP-15376. This change is consistent with NRC guidance provided in an NRC letter to NEI dated July 26, 2002 (Reference 6).

This change is supported by WCAP-15376.

Table 4.3 Proposed Reactor Trip System (RTS) TS STI and Associated Changes Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/4 3-12 FU l.A. -Manual Proposed Action (17) is applied to various specific S/U 2 3/4 3-11 Reactor Trip - Shunt surveillances that will only be performed if they have not (Table 4.3-1) Trip Function: Replace been performed in the previous 184 days. Consistent with the applicability of the semi-annual STIs of WCAP-15376, it is acceptable to Note (1) with Note (17) apply this extended exception criterion to this functional unit.

This change is supported by WCAP-15376.

1 3/4 3-12 FU 1.B. -Manual Proposed Action (17) is applied to various specific S/U 2 3/4 3-11 Reactor Trip - surveillances that will only be performed if they have not (Table 4.3-1) Undervoltage Trip been performed in the previous 184 days. Consistent with Function: Replace the the semi-annual STIs of WCAP-15376, it is acceptable to applicability of Note (1) apply this extended exception criterion to this functional with Note (17) unit.

This change is supported by WCAP-15376.

to AEP:NRC:2311 Page 5 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/43-12 FU 2- Channel A change to Note (1) is proposed to specify that 2 3/4 3-11 Functional Test: requirements annotated with this notation are only required (Table 4.3-1) Delete the Start-Up CFT, if not performed within the previous 184 days. The including Notation (1) proposed 184-day interval is greater than the proposed applicability quarterly (92 days) interval for this functional unit's CFT.

Consequently, a CFT will be required at start-up for this instrumentation to satisfy the Quarterly CFT requirement, if one has not been performed within the previous 92 days.

Therefore, the SlU CFT requirement is redundant to the S/U requirement, and should be deleted to minimize potential misinterpretations of this requirement.

This change is consistent with the surveillance requirements of NUREG-1431, Revision 2 and is supported by WCAP-15376.

1 3/4 3-12 FU 2 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test:

(Table 4.3-1) Change Monthly to Quarterly 1 3/4 3-12 FU 3 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test:

(Table 4.3-1) Change Monthly to Quarterly 1 3/4 3-12 FU 4 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Tezt:

(Table 4.3-1) Change Monthly to Quarterly 1 3/4 3-12 FU 5 - Intermediate Proposed Action (17) is applied to various specific S/U 2 3/4 3-11 Range, Neutron Flux: surveillances that will only be performed if they have not (Table 4.3-1) Replace the applicability been performed in the previous 184 days. Consistent with of Note (1) with Note the semi-annual STIs of WCAP-15376, it is acceptable to (17) apply this extended exception criterion to this functional unit.

This change is supported by WCAP-1 5376.

1 3/43-12 FU 6 - Channel A change to Note (1) is proposed to specify that 2 3/4 3-11 Functional Test: requirements annotated with this notation are only required (Table 4.3-1) Delete the Start-Up if not performed within the previous 184 days. The Channel Functional Test, proposed 184-day interval is greater than the current including Notation (1) monthly (31 days) interval for this functional unit's CFT.

applicability Consequently, a CFT will be required at start-up for this instrumentation to satisfy the monthly CFT requirement, if one has not been performed within the previous 31 days.

Therefore, the S/U CFT requirement is redundant to the S/U requirement, and should be deleted to minimize potential misinterpretations of this requirement.

This change is consistent with the surveillance requirements of NUREG-1431, Revision 2 and is supported by WCAP-15376.

to AEP:NRC:2311 Page 6 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/4 3-12 FU 7 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test:

(Table 4.3-1) Change Monthly to Semi-Annually 1 3/4 3-12 FU 8 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test:

(Table 4.3-1) Change Monthly to Semi-Annually 1 3/4 3-12 FU 9 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test:

(Table 4.3-1) Change Monthly to Semi-Annually 1 3/4 3-12 FU 9 - Mode in which Per UFSAR Table 7.2-1, the CNP FU 9 has a P-7 interlock.

2 3/4 3-11 Surveillance Required: Per CNP TS Table 3.3-1 the P-7 interlock blocks the FU 9 (Table 4.3-1) Delete Mode 2. reactor trip logic signal below 11% of rated thermal power, which encompasses Mode 2 (defined as less than or equal to 5% of rated thermal power). Therefore, CNP FU 9 does not perform a reactor trip function in Mode 2, and Mode 2 should be removed from the "Mode in Which Surveillance Required" column.

This is consistent with NUREG-1431, Revisions 1 and 2.

1 3/4 3-12 FU 10 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test:

(Table 4.3-1) Change Monthly to Semi-Annually 1 3/4 3-12 FU 11 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test:

(Table 4.3-1) Change Monthly to Senm-Annually 1 3/4 3-12 FU 11 - Modes in which Per UFSAR Table 7.2-1, the CNP FU 11 has a P-7 2 3/4 3-11 Surveillance Required: interlock. Per CNP TS Table 3.3-1 the P-7 interlock blocks (Table 4.3-1) Delete Mode 2. the FU 11 reactor trip logic signal below 11% of rated thermal power, which encompasses Mode 2 (defined as less than or equal to 5% of rated thermal power).

Therefore, CNP FU 11 does not perform a reactor trip function in Mode 2, and Mode 2 should be removed from the "Mode in Which Surveillance Required" column.

This is consistent with NUREG-1431, Revisions 1 and 2.

1 3/4 3-12 FU 12 - Channel This change is supported by WCAP-15376.

2 3/4 3-11 Functional Test.

(Table 4.3-1) Change Monthly to Semi-Annually 1 3/4 3-13 FU 14 - Channel This change is supported by WCAP-15376.

2 3/4 3-12 Functional Test:

(Table 4.3-1) Change Monthly to I Semi-Annually to AEP:NRC:2311 Page 7 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/4 3-13 FU 15 - Channel This change is supported by WCAP-1 5376.

2 3/4 3-12 Functional Test:

(Table 4.3-1) Change Monthly to Semi-Annually 1 3/4 3-13 FU 19 - Channel NUREG-1431, Rev. 1, only specifies a trip actuating device 2 3/4 3-12 Functional Test: operational test (TADOT) frequency of 18 months for the (Table 4.3-1) Change Monthly to Safety Injection Actuation Input from the Engineered Semi-Annually Safety Features Actuation System (ESFAS) to the RTS (STS RTS FU 17, and Surveillance Requirement (SR) 3.3.1.14). Therefore, it was not necessary to specifically evaluate an STI extension for this functional unit in WCAP-15376. However, an STI extension to semi-annual was justified in WCAP-15376 for the ESFAS actuation logic (STS ESFAS FU L.b, and SR 3.3.2.2). This actuation logic output is the same signal that is the input to the CNP RTS-FU 19. I&M proposes to increase the STI for CNP RTS-FU 19 to semi-annual. This is consistent with the STI extensions approved in WCAP-15376 for STS ESFAS FU 1, Safety Injection, inputs and actuation logic.

This change is supported by WCAP-15376.

1 3/4 3-13 FU 19 - Channel A new Notation (15) has been proposed which is specific to 2 3/4 3-12 Functional Test: FU 22, "Automatic Trip Logic." Proposed Notation (15)

(Table 4.3-1) Add reference to continues the staggered testing of FU 22. FU 19, "Safety notation "(15)" Injection Input from ESF," is the ESFAS actuation signal that is directly input to the RTS actuation logic. 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.

The proposed staggered semi-annual testing of FU 19 is more conservative than that specified in NUREG-1431, Revisions 1 and 2, which has an STI of 18 months for the ESFAS signal input to the RTS (STS RTS FU 17).

1 3/43-13 FU 21.A - Channel A change to Note (1) is proposed to specify that 2 3/4 3-12 Functional Test: requirements annotated with this notation are only required (Table 4.3-1) Delete the Start-Up if not performed within the previous 184 days. The Channel Functional Test, proposed 184-day interval is greater than the proposed 4 including Notations (1) Month (124 days) interval for this functional unit's CFT.

and (11) applicability Consequently, a CFT will be required at start-up for this instrumentation to satisfy the 4 month CFT requirement, if one has not been performed within the previous 124 days.

Therefore, the S/U CFT requirement is redundant to the S/U requirement, and should be deleted to minimize potential misinterpretations of this requirement. Note (11) still applies to the "4-Month" CFT requirement.

This change is consistent with the surveillance requirements of NUREG-1431, Revision 2 and is supported by WCAP-15376.

to AEP:NRC:2311 Page 8 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/4 3-13 FU 21.A - Channel This change is supported by WCAP-15376.

2 3/4 3-12 Functional Test:

(Table 4.3-1) Change Monthly to 4 Months 1 3/43-13 FU 21.B - Channel A change to Note (1) is proposed to specify that 2 3/4 3-12 Functional Test: requirements annotated with this notation are only required (Table 4.3-1) Delete the Start-Up if not performed within the previous 184 days. The Channel Functional Test, proposed 184-day interval is greater than the proposed 4 including Notations (1) Month (124 days) interval for this functional unit's CFT.

and (11) applicability Consequently, a CFT will be required at start-up for this instrumentation to satisfy the 4 month CFT requirement, if one has not been performed within the previous 124 days.

Therefore, the S/U CFT requirement is redundant to the S/U requirement, and should be deleted to minimize potential misinterpretations of this requirement. Note (11) still applies to the "4-Month" CFT requirement.

This change is consistent with the surveillance requirements of NUREG-1431, Revision 2 and is supported by WCAP-15376 1 3/4 3-13 FU 21.B - Channel This change is supported by WCAP-15376.

2 3/4 3-12 Functional Test:

(Table 4.3-1) Change Monthly to 4 Months 1 3/4 3-13 FU 22 - Channel This change is supported by WCAP-15376.

2 3/4 3-12 Functional Test:

(Table 4.3-1) Change Monthly to Seml-Annually 1 3/4 3-13 FU 22 - Channel A new Notation (15) has been added to be specific to 2 3/4 3-12 Functional Test: FU 22.

(Table 4.3-1) Change notation reference "(5)" to "(15)" This change is supported by WCAP-15376.

1 3/4 3-13 FU 23 - Channel This change is supported by WCAP-15376.

2 3/4 3-12 Functional Test:

(Table 4.3-1) Change Monthly to 4 Months 1 3/4 3-13 FU 23 - Channel Notation (5) specifies that each train shall be tested at least 2 3/4 3-12 Functional Test: every other 62 days For the two trains of Reactor Trip (Table 4.3-1) Add reference to Bypass Breakers, notation (5) results in testing each train Notation "(5)" every 4 months. In accordance with STS terminology, this change is consistent with testing each train every 4 months "on a staggered test basis.'"

Therefore, this change is supported with the NUREG-1431 STS and WCAP-15376.

to AEP:NRC:231 1 Page 9 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) 1 3/4 3-13 FU 23 - Channel A new Notation (16) is proposed. It is a clarification of the 2 3/4 3-12 Functional Test: surveillance testing requirements for the RTBBs. The (Table 4.3-1) Add reference to RTBBs do not perform a safety function when they are not proposed Notation racked in and closed. Therefore, operability testing is not

"(16)" required. Once an RTBB is racked in and closed, thereby bypassing the associated reactor trip breaker, functional testing is required. An STI of 4 months is proposed for the RTBBs.

This clarification is consistent with NUREG-1431, Revision 2, and the STI is supported by WCAP-15376.

1 3/4 3-13 FU 23 - Reactor Trip Proposed Action (17) is applied to various specific S/J 2 3/4 3-12 Bypass Breaker: surveillances that will only be performed if they have not (Table 4.3-1) Replace the start-up been performed in the previous 184 days. Consistent with applicability of Note (1) the semi-annual STIs of WCAP-15376, it is acceptable to with Note (17) apply this extended exception criterion to this functional unit.

This change is supported by WCAP-15376.

1 3/4 3-14 Notation (5): Change This change is supported by WCAP-15376.

2 3/4 3-13 "Each train tested every (Table 4.3-1) other month" to "Each train tested at least every other 62 days" 1 3/4 3-14 Add Notation (15) The proposed change to original Notation (5) will make it 2 3/4 3-13 "(15) - Each train tested specific to the proposed STI for FUs 21 and 23. As a result (Table 4.3-1) at least every other 92 of the proposed change to Notation (5), this note will no days" longer be applicable to FU 22. Proposed Notation (15) is specific to FUs 19 and 22.

This change is supported by WCAP-15376.

1 3/4 3-14 Add Notation (16): This is a clarification of the surveillance testing 2 3/4 3-13 "Applicable to any requirements for the RTBBs. The RTBBs do not perform a (Table 4.3-1) reactor trip bypass safety function when they are not racked in and closed.

breakers that are racked Therefore, operability testing is not required. Prior to in and closed for racking in and closing an RTBB (thereby bypassing the bypassing a reactor trip associated reactor trip breaker), functional testing is breaker." required. A 4-month STI is proposed for the RTBBs.

This clarification is consistent with NUREG-1431, Revision 2, and the STI is supported by WCAP-15376.

to AEP:NRC:2311 Page 10 Unit TS Page Proposed Change Explanation of Change/Justification (TS Section) I 1 3/43-14 Add Notation (17): This note is applicable to several FUs in the CNP TS. The 2 3/4 3-13 "If not performed in ability of these FUs to perform their intended function(s) is (Table 4.3-1) previous 184 days." not impacted by a reactor startup. This note in the STS has been revised as the STI extensions for the applicable FUs have been approved in WCAP-10271-P-A and WCAP-14333-P-A, and is evaluated and justified in WCAP-15376. This change is consistent with NUREG-1431, Revision 2 and WCAP-15376 TS changes.

For those FUs that have a more frequent STI than semi-annual, the TS Surveillance requirement 4.3.1.1.1 requires that the surveillance testing be current prior to entering an applicable mode.

This change is supported by WCAP-15376.

______ .1_____________ .1.

to AEP:NRC:2311 Page 11 Table 4.4 Proposed ESFAS TS STI Changes Unit TS Page Proposed Explanation of Change/Justification (TS Section) ESFAS TS STI Changes 1 3/4 3-31 FU L.b - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-31 FU L.c - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-31 FU L.d - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-31 FU L.e - Channel This change is supported by WCAP-1 5376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-31 FU L.f- Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-31 FU 2.b - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-31 FU 2.c - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-32 FU 3.a.2) - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-32 FU 3.b.2) - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4 3-2) Change Monthly to Semi-Annually 1 3/4 3-32 FU 3.b.3) - Channel This change is supported by WCAP-15376.

2 3/4 3-30 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-33 FU 4.b - Channel This change is supported by WCAP-15376.

2 3/4 3-31 Functional Test:

(Table 4.3-2) Change Monthly to I Semi-Annually to AEP:NRC:2311 Page 12 Unit TS Page Proposed Explanation of Change/Justification (TS Section) ESFAS TS STI Changes 1 3/43-33 FU 4.c - Channel NUREG-1431, Rev. 1, Table 3.3.2-1, referenced in 2 3/4 3-31 Functional Test: WCAP-15376, contains three containment pressure trips; a (Table 4.3-2) Change Monthly to Containment Pressure - High 1 with the lowest setpoint, Semi-Annually Containment Pressure - High 2 with an intermediate pressure setpoint, and a Containment Pressure - High 3 with the highest setpoint. The STS table specifies the Containment Pressure - High 2 as the input to ESFAS FU 4.c.

The corresponding CNP functional unit, FU 4.c., uses the Containment Pressure - High-High, (highest pressure setpoint) as the input.

CNP, an ice condenser design, only has two corresponding signals, a Containment Pressure - High and a Containment Pressure - High-High. Both of these trips are derived from the same analog channel. WCAP-15376 evaluates a semi annual STI for the "Containment Pressure - High-High" analog channel that provides input for FU 2.c. and FU 3(b)3. The STI relaxations in WCAP-15376 are applicable to the actuation logic for all analog signals processed through the solid state protection system.

Therefore, I&M proposes to increase the STI for CNP FU 4.c to semi-annual This change is supported by WCAP-15376.

1 3/4 3-33 FU 4.d - Channel This change is supported by WCAP-15376.

2 3/4 3-31 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-33 FU 4.e - Channel This change is supported by WCAP-15376.

2 3/4 3-31 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-33 FU 5.a - Channel This change is supported by WCAP-15376.

2 3/4 3-31 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-33 FU 6.a - Channel This change is supported by WCAP-15376.

2 3/4 3-31 Functional Test:

(Table 4.3-2) Change Monthly to Sermi-Annually 1 3/4 3-33 FU 6.c - Channel This change is supported by WCAP-15376.

2 3/4 3-31 Functional Test.

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-33a FU 7.a - Channel This change is supported by WCAP-1 5376.

2 3/4 3-32 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually to AEP:NRC:2311 Page 13 Unit TS Page Proposed Explanation of Change/Justification (TS Section) ESFAS TS STI Changes 1 3/4 3-33b FU 1O.b - Channel This change is supported by WCAP-15376.

2 3/4 3-32 Functional Test:

(Table 4.3-2) Change Monthly to Semi-Annually 1 3/4 3-33b FU 10.c - Channel NUREG-1431, Rev. 1, Table 3.3.2-1 (referenced in 2 3/43-32 Functional Test: WCAP-15376) does not contain a Containment Air (Table 4.3-2) Change Monthly to Recirculation Fan functional unit. CNP's containment is an Semi-Annually ice condenser design, which uses containment air recirculation fans for mitigation of in-containment steam release events. CNP's TS FU 10, "Containment Air Recirculation Fan," uses Containment Pressure-High as the input. This is the same analog channel that is used as an input to CNP's FU 1, which has a proposed semi-annual STI. Per the justifications in WCAP-15376, the STI relaxations are applicable to the actuation logic for all analog signals processed through the solid state protection system.

Therefore, I&M proposes to increase the STI for CNP FU 1O.c to semi-annual.

This change is supported by WCAP-15376.

1 3/4 3-34 Table Notation (2): This change is supported by WCAP-15376.

2 3/4 3-33 Change "31" to "92" (Table 4.3-2) days

References:

1. WCAP-15376, Revision 0, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," dated October 2000

2. NUREG-1431, Revision 1, "Standard Technical Specifications, Westinghouse Plants, Specifications, Volume 1," dated April 1, 1995

3. NUREG-1431, Revision 2, "Standard Technical Specifications, Westinghouse Plants, Specifications, Volume 1," dated April 1, 2001

4. NUREG-1431, Revision 0, "Standard Technical Specifications, Westinghouse Plants, Specifications, Volume 1," dated September 1, 1992

5. TSTF-41 1, Revision 1, "Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376-P)," dated August 7, 2002

6. Letter from W. D. Beckner, NRC, to A. Pietrangelo, NEI, addressing inconsistencies between TSTF-41 1 and TSTF-418, dated July 26, 2002

ATTACHMENT 5 TO AEP:NRC:2311 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 Any additional conditions stipulated in the Safety Evaluation Within 60 days of receipt Report (SER) that documents NRC approval of WCAP-15376 will of the WCAP-15376 SER be addressed by a supplement to this License Amendment Request.

ML022490127

Contents

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SUBJECT:

REFERENCE:

Dear Sir or Madam:

Enclosures:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

References:

SUMMARY

References:

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ML022490127

Text

SUBJECT:

REFERENCE:

Dear Sir or Madam:

Enclosures:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

References:

SUMMARY

References:

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