AEP-NRC-2019-39, Application to Revise Technical Specifications to Adopt TSTF-569, Revise Response Time Testing Definition.

Application to Revise Technical Specifications to Adopt TSTF-569, Revise Response Time Testing Definition.

ML19310D766
Person / Time
Site:	Cook
Issue date:	10/31/2019
From:	Lies Q Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2019-39
Download: ML19310D766 (25)

Text

m INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER One Cook Place Bridgman, Ml 49106 A unit ofAmerican Electric Power lndianaMichiganPower.com October 31, 2019 AEP-NRC-2019-39 10 CFR 50.90 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant, Unit 1 and Unit 2 Application to Revise Technical Specifications to Adopt TSTF-569, "Revise Response Time Testing Definition."

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP), hereby requests an amendment to the Technical Specifications (TS) for CNP Unit 1 and Unit 2.

l&M requests adoption of TSTF-569, "Revise Response Time Testing Definition," which is an.

approved change to the Improved Standard Technical Specifications, into the CNP Unit 1 and Unit 2 TS. The proposed amendment revises the TS Definitions for Engineered Safety Feature Response Time and Reactor Trip System Response Time. provides an affirmation statement pertaining to the information contained herein. provides a description and assessment of the proposed changes. Enclosures 3 and 4 provide existing Unit 1 and Unit 2 TS pages, respectively, marked up to show the proposed changes.

New clean Unit 1 and Unit 2 TS pages with proposed changes incorporated will be provided to the Nuclear Regulatory Commission (NRC) Licensing Project Manager when requested.

Enclosures 5 and 6 to this letter provide existing Unit 1 and Unit 2 TS Bases pages, respectively, marked up to reflect the proposed changes. TS Bases markups are included for information only.

Changes to the existing TS Bases, consistent with the technical and regulatory analyses, will be implemented under TS 5.5.12, "Technical Specifications Bases Control Program."

l&M requests NRC review and approval of the proposed changes by September 2020 to support the CNP Unit 1 2020 fall outage. Once approved, the amendment shall be implemented within 60 days.

In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Michigan state officials.

U. S. Nuclear Regulatory Commission AEP-NRC-2019-39 Page 2 There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely, 2~A~

~nelies Site Vice President Indiana Michigan Power Company JMT/kmh

Enclosures:

1. Affirmation

2. Description and Assessment of the Technical Specification Changes

3. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked To Show Proposed Changes

4. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes

5. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Bases Pages Marked To Show Proposed Changes (For Information Only)

6. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Bases Pages Marked To Show Proposed Changes (For Information Only) c: R. J. Ancona - MPSC R. F. Kuntz - NRC, Washington, D.C.

EGLE - RMD/RPS NRG Resident Inspector D. J. Roberts - NRC Region Ill A. J. Williamson - AEP Ft. Wayne, w/o enclosures

Enclosure 1 to AEP-NRC-2019-39 AFFIRMATION I, Q. Shane Lies, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of l&M, and that the statements made and the matters set forth herein pertaining to l&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company a2t:~1G Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS 31 DAY OF Qc..,mber 2019

~~-p~~

My Commission Expires I 0\ f 2, .2025 c-.

Enclosure 2 to AEP-NRC-2019-39 Description and Assessment of Technical Specification Changes

1.0 DESCRIPTION

Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP)

Unit 1 and Unit 2, requests adoption of TSTF-569, "Revise Response Time Testing Definition,"

which is an approved change to the Improved Standard Technical Specifications (ISTS), into the TechnicaLSpecifications (TS) for CNP Units 1 and 2. The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System {RTS)

Response Time.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation l&M has reviewed the safety evaluation for TSTF-569 provided to the Technical Specifications Task Force in a letter dated August 14, 2019. This review included a review of the Nuclear Regulatory Commission (NRC) staff's evaluation, as well as the information provided in TSTF-569. l&M has concluded that the justifications presented in TSTF-569 and the safety evaluation prepared by the NRC staff are applicable to CNP Unit 1 and Unit 2 and justify this amendment for the incorporation of the changes to the CNP Unit 1 and Unit 2 TS.

2.2 Variations l&M is proposing the following variations from the TS or TS Bases changes described in TSTF-569. These variations do not affect the applicability of TSTF-569, or the _NRC staffs safety evaluation, to the proposed license amendment.

The changes to TS Bases in TSTF-569 refer to SR 3.3.1.16, "RTS Instrumentation," stating "Verify RTS RESPONSE TIME is within limits." The corresponding Surveillance Requirement in the CNP Unit 1 and Unit 2 TS is SR 3.3.1.19.

The changes to TS Bases in TSTF-569 refer to SR 3.3.2.10, "ES FAS Instrumentation," stating "Verify ESFAS

• RESPONSE TIMES are within . limit." The corresponding Surveillance Requirement in the CNP Unit 1 and Unit 2 TS is SR 3.3.2.12.

The changes to TS Bases B 3.3.1, "RTS Instrumentation" in TSTF-569 refer to WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements,"

January 1996 and WCAP-14036-P, Revision 1; "Elimination of Periodic Protection Channel Response Time Tests,,; December 1995 as References 10 and 15, respectively. Due to the numbering sequence of the References in the CNP Unit 1 and Unit 2 TS Bases, the corresponding

. numbers are References 13 and 14, respectively.

Enclosure 2 to AEP-NRC-2019-39 Page 2 The changes to TS Bases B 3.3.2, "ESFAS Instrumentation" in TSTF-569 refer to WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996 and WCAP-14036-P,. Revision 1, "Elimination of Periodic Prdtection Channel Response Time Tests," December 1995 as References 14 and 15, respectively. Due to the numbering sequence of the References in the CNP Unit 1 and Unit 2 TS Bases, the corresponding numbers are References 12 and 13, respectively.

These variations only effect changes to the CNP Unit 1 and Unit 2 TS Bases pages. Enclosures 5 and 6 to this letter provide existing Unit 1 and Unit 2 TS Bases pages, respectively, marked up to reflectthe proposed changes. TS Bases markups are included for information only. Changes to the existing TS Bases, consistent with the technical and regulatory analyses, will be implemented under TS 5.5.12, "Technical Specifications Bases Control Program."

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) requests adoption of TSTF-569, "Revise Response Time Testing Definition," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the CNP Unit 1 and Unit 2 TS. The proposed amendment revises the TS Definitions for Engineered Safety Feature (ESF) Response Time and Reactor Trip System (RTS) Response Time.

l&M has evaluated whether 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:No The proposed change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC approved methodology and apply a bounding response time for some components in lieu of measurement. The requirem~nt for the instrumentation to actuate within the response time assumed in the accident analysis is unaffected.

ThE:l response time associated with the RTS and ESF instrumentation is not an initiator of any accident. Therefore, the proposed change has no significant effect on the probability of any accident previously evaluated.

• The affected RTS and ESF instrumentation are assumed to actuate their respective components within the required response time to mitigate accidents previously evaluated.

Revising the TS definition for RTS and ESF instrumentation response times, to allow an NRC approved methodology for verifying response times for some components, does not alter the surveillance requirements that verify the RTS and ESF instrumentation response times are within the required limits. As such, the TS will continue to assure that the RTS

l to AEP-NRC-2019-39 Page 3 and ESF instrumentation actuate their associated components within the specified

. response time to accomplish the required safety functions assumed in the accident analyses. Therefore, the assumptions used in any accidents previously evaluated are unchanged and there is no significant increase in the consequences.

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

Response:No The proposed change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC approved methodology and apply a bounding response time for some components in lieu of measurement. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The proposed change does not alter any assumptions made in the safety analyses. The proposed change does not alter the limiting conditions for operation for the RTS or ESF instrumentation, nor does it change the Surveillarice Requirement to verify the RTS and ESF instrumentation response times are within the required limits. As such, the proposed change does not alter the operability requirements for the RTS and ESF instrumentation, and therefore, does not introduce any new failure modes.

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

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

Response:No The proposed. change revises the TS Definition of RTS and ESF instrumentation response time to permit the licensee to evaluate using an NRC approved methodology and apply a bounding response time for some components in lieu of measurement. The proposed change has no effect on the required RTS and ESF instrumentation response times or setpoints assumed in the safety analyses and the TS requirements to verify those response times and setpoints. The proposed change does not alter any Safety Limits or analytical limits in the safety analysis. The proposed change does not alter the TS operability requirements for the RTS and ESF instrumentation. The RTS and ESF instrumentation actuation of the required systems and components at the required setpoints and within the specified response times will continue to accomplish the design basis safety functions of the associated systems and components in the same manner as before. As such, the RTS and ESF instrumentation will continue to perform the required safety functions as assumed in the safety analyses for all previously evaluated accidents.

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

Enclosure 2 to AEP-NRC-2019-39 Page 4 Based on the above, l&M concludes that the proposed change 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.

3.2 Conclusions 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.

4.0 *. ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility.

• component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. .Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b ), no environmental impact statement or environmental assessment need to be prepared in connection with the proposed change.

Enclosure 3 to AEP-NRC-2019-39 Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes 1.1-3 1.1-5

Definitions 1.1 1.1 Definitions dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE(ESF)RESPONS E when the monitored parameter exceeds its actuation setpoint TIME at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been reviousl reviewed and a roved by the NRG, or the com onents have been evaluated in accordance with an NRG a roved methodolo .

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM. program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank,

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and Cook Nuclear Plant Unit 1 1.1-3 Amendment No. 287, 29:8, ~. 345

Definitions 1.1 1.1 Definitions RATED THERMAL POWER ATP shall be a total reactor core heat transfer rate to the (ATP) reactor coolant of 3304 MWt.

REACTOR TRIP SYSTEM The ATS RESPONSE TIME shall be that time interval from (ATS) RESPONSE TIME . when the monitored parameter exceeds its ATS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been reviousl reviewed and a roved by the NRG, or the com onents have been evaluated in ccordance with an NRG a roved methodolo .

SHUTDOWN MARGIN (SOM) SOM shall be the instantaneous amount of reactivity by which the reactor is subcritical o{ would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SOM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay.

The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other

• designated components in the associated function.

Cook Nuclear Plant Unit 1 1.1-5 Amendment No. 287

l Enclosure 4 to AEP-NRC-2019-39 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show

, Proposed Changes 1.1-3 1.1-5

Definitions 1.1 1.1 Definitions dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE(ESF)RESPONS E when the monitored parameter exceeds its actuation setpoint TIME at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been reviousl reviewed and a roved by the NRG, or the com onents have been evaluated in accordance with an NRG a roved methodolc:i .

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).

LEAKAGE LEAKAGE shall be: *

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank,

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and

c. Pressure Boundary LEAKAGE Cook Nuclear Plant Unit 2 1.1-3 Amendment No. 269

Definitions 1.1 1.1 Definitions RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3468 MWt.

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been reviousl reviewed and a roved by the NRG, or the com onents have been evaluated in accordance with an NRG a roved methodolo .

SHUTDOWN MARGIN (SD~) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay.

The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testirig of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

Cook Nuclear Plant Unit 2 1.1-5 Amendment No. 269

Enclosure 5 to AEP-NRC-2019-39 Donald C. Cook Nuclear Plant Unit 1 Technical Specification Bases Pages Marked to Show Proposed Changes (For Information Only)

B 3.3.1-47 B 3.3.1-48 B 3.3.2-41 B 3.3.2-42

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from:

( 1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g.,

vendor) test measurements, or (3) utilizing vendor engineering

. specifications. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," (Ref. 13) provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified ih the WCAP. Response time verification for other sensor types

. must be demonstrated by test.

WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 14) provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time.

The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

!INSERT PJ The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.19 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.

The response time testing of the neutron flux signal portion of the channel shall be measured from either the detector output or the input of the first electronic component in the channel.

Cook Nuclear Plant Unit 1 B 3.3.1-47 Revision No. 63

RTS Instrumentation B 3.3.1 BASES REFERENCES 1. Regulatory Guide 1.105, Revision 3, "Setpoints for Safety Related Instrumentation."

2. UFSAR, Chapter 7.

3. Technical Requirements Manual.

4. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.

5. UFSAR, Table 7.2-1.

6. UFSAR,Table14.1-2.

7. 10 CFR 50.49.

8. WCAP-12741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.

9. UFSAR, Chapter 14.

10. WCAP-10271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System,"

including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.

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

12. UFSAR, Table 7.2-6.

13. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Ti.me Testing Requirements," January 1996.

14. WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.

15. Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing."

Cook Nuclear Plant Unit 1 B 3.3.1-48 Revision No. 5

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)

Table 7.2-7 (Ref. 11 ). Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).

For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.

Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from:

(1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place; onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," dated January 1996 (Ref. 12), provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.

WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 13) provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time.

The allocations for sensor, signal conditioning, and actuation logic

• response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response tim~ provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

!INSERT Bl Cook Nuclear Plant Unit 1 B 3.3.2-41 Revision No. 63

ESFAS Instrumentation B 3.3.2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

BASES SURVEILLANCE REQUIREMENTS (continued)

This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 850 psig in the.SGs.

REFERENCES 1. Technical Requirements Manual.

2. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.

3. UFSAR, Table 7.2-1.

4. UFSAR, Table 14.1-2.

5. 10 CFR 50.49.

6. WCAP-12741, "Westinghouse Menu Driven Setpoint Calculation Program (STEP IT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.

7. UFSAR, Chapter 14.

8. WCAP-14333-P-A, Revision 1, October 1998.

9. WCAP-10271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection lnstrumentatio.n System,"

including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.

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

11. UFSAR, Table 7.2-7.

12. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response TimeTesting Requirements," January 1996.

13. WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.

14. Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing."

Cook Nuclear Plant Unit 1 B 3.3.2-42 Revision No. 63

Insert A The response time may be verified for components that repl 9ce the components that were previously evaluated in Ref. 13 and Ref. 14, provided that the components. have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing," (Ref. 15).

Insert B The response time may be verified for components that replace the components that were previously evaluated in Ref. 12 and Ref. 13, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing," (Ref. 14).

Enclosure 6 to AEP-NRC-2019-39 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Bases Pages Marked to Show Proposed Changes (For Information Only)

B 3.3.1-47 B 3.3.1-48 B 3.3.2-42 B 3.3.2-43

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

Response time may be verified by actual response time tests in any series of sequential,. overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from:

(1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g.,

vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," (Ref. 13) provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP .. Response time verification for other sensor types must be demonstrated by test.

WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 14) provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the prolection system channel response time.

The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

!INSERTAj The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.19 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.

The response time testing of the neutron flux signal portion of the channel shall be measured from either the detector output or the input of the first electronic component in the channel.

Cook Nuclear Plant Unit 2 B 3.3.1-47 Revision No. 60

RTS Instrumentation B 3.3.1 BASES REFERENCES 1. Regulatory Guide 1.105, Revision 3, "Setpoints for Safety Related Instrumentation." *

2. UFSAR, Chapter 7.

3. Technical Requirements Manual.

4. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.

5. UFSAR, Table 7.2-1.

6. UFSAR, Table 14.1.0-4.

7. 10 CFR 50.49.

8. WCAP-12741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License

• Amendments 175 and 160, dated May 13, 1994.

9. UFSAR, Chapter 14.

10. WCAP-10271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System,"

including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.

11. WCAP-15376, "Risk-Informed Assessment of the RTS and ES FAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," October 2000.

12. UFSAR, Table 7.2-6.

13. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.

14. WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.

15. Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing."

Cook Nuclear Plant Unit 2 B 3.3.1-48 Revision No. 5

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMEN!S (continued) replaced without verification testing.

• One example where response time could be affected is replacing the sensing assembly of a transmitter.

!INSERT Bl The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 850 psig in the SGs.

REFERENCES 1. Technical Requirements Manual.

2. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.

3. UFSAR, Table 7.2-1.

4. UFSAR, Table 14.1.0-4.

5. 10 CFR 50.49.

6. - WCAP-12741, "Westinghouse Menu Driven Setpoint Calculation Program {STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.

7. UFSAR, Chapter 14.

8. WCAP-14333-P-A, Revision 1, October 1998.

9. WCAP-10271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System,"

including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.

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

11. UFSAR, Table 7.2-7.

Cook Nuclear Plan.t Unit 2 B 3.3.2-42 Revision No. 60

ESFAS Instrumentation B 3.3.2 BASES REFERENCES (continued)

12. WCAP-13632-P-A, Revision 2, "Elimination of Pressure ~ensor Response Time Testing Requirements," January 1996.

13. WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.

14. Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing."

Cook Nuclear Plant Unit 2 B 3.3.2-43 Revision No. 58

Insert A The response time may be verified for components that replace the components that were previously evaluated in Ref. 13 and Ref. 14, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing," (Ref. 15).

Insert B The response time may be verified for components that replace the components that were previously evaluated in Ref. 12 and Ref, 13, provided that the components have beein evaluated in accordance with the NRC approved methodology as.discussed in Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel Response Time Testing," (Ref. 14).