Text

Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3 and to Request an Alternative to the ASME Code
	ML16208A076
Person / Time
Site:	Cook
Issue date:	07/21/2016
From:	Lies Q; AEP Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP-NRC-2016-29
	Download: ML16208A076 (70)
	v • d • e

m INOIANA MICHIGAN POWER A unit of American Electric Power July 21, 2016 Docket Nos.: 50-315 50-316 U.S. NuclearRegulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 lndianaMichiganPower.com AEP-NRC-2016:-29 10 CFR 50.90 10 CFR 50.55a Donald C. Cook Nuclear Plant, Units 1 and 2 APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-545, REVISION 3, "TS INSERVICE TESTING PROGRAM REMOVAL & CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING," AND TO REQUEST AN ALTERNATIVE TO THE ASME CODE Pursuant to 10 CFR 50.90, Indiana Michigan Power (l&M), the *licensee for Donald C. Cook Nuclear Plant (CNP), is submitting a request for an amendment to the Technical Specifications (TS) for CNP Units 1 and 2. The proposed change revises the TSs to eliminate the Section 5.5.6, "lnservice Test Program." A new defined term, "lnservice Testing Program," is added to the TS Definitions section.

This request is consistent with Technical Specification Task Force - 545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing."

Pursuant to 10 CFR 50.55a(z), the application also proposes an alternative to the testing

frequencies in the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code, by adoption of approved Code Case OMN-20, "lnservice Test Frequency," for the current 10 year lnservice Testing interval. provides an affirmation statement pertaining to the information contained herein. provides a description and assessment of the proposed TS changes. Enclosures 3 and 4 provide Unit 1 and Unit 2 TS pages marked to show the proposed changes. Enclosures 5 and 6 provide Unit 1 and Unit 2 TS Bases pages marked to show the proposed changes. Bases markups are included for information only. New clean Unit 1 and Unit 2 TS pages with proposed changes incorporated will be provided to the U. S. Nuclear Regulatory Commission (NRC)

Licensing Project Manager when requested. Enclosure 7 provides the request for an alternative to the ASME Code.

l&M would like to request NRG review and approval of the proposed change and relief request commensurate with the NRC's normal review schedule. Once approved, the amendment shall be implemented within 9"0 days.

U.S. Nuclear Regulatory Commission Page2 AEP-NRC-2016-29 Copies of this letter and its enclosures are being transmitted to the. Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, 1~1~

Quinton S. Lies Site Vice President Indiana Michigan Power Company DB/mll

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 Ba~es 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)

7.

Description and Assessment of the Proposed Alternative to the ASME Code c:

R. J. Ancona, MPSC A. W. Dietrich, NRC, Washington, D.C.

MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC, Region Ill A. J. Williamson, AEP Ft. Wayne, w/o enclosures to AEP-NRC-2016-29 AFFIRMATION I, Quinton S. 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

?_~J.

Q'~n S. Lies Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS/) \\st DAY OF ;JLL}\\J 2016

I My Commission Expires 0 \\ I~\\ l d...0\\~

to AEP-NRC-2016-29 Description and Assessment of Technical Specification Changes This letter is a request to amend Operating License Numbers DPR-58 and DPR-74 for Donald C.

Cook Nuclear Plant (CNP) Unit 1 and Unit 2, respectively. In this request, Indiana Michigan Power Company (l&M), the licensee for CNP, proposes to implement Technical Specification Task Force (TSTF) change traveler TSTF-545, Revision 3, "TS lnservice Testing Program Removal &

Clarify SR Usage Rule Application to Section 5.5 Testing," dated October 21, 2015, at CNP.

1.0 DESCRIPTION

The proposed change eliminates the Technical Specifications (TS), Section 5.5.6, "lnservice Test (IST) Program," to remm/e requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, "lnservice Test Frequency." A new defined term, "lnservice Testing Program," is added to TS Section 1.1, "Definitions."

The proposed change to the TS is consistent with TSTF-545, Revision 3, "TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing."

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation l&M has reviewed the Safety Evaluation (SE) published in the Federal Register Notice of Availability on March 28, 2016, (81 FR 17208) to the TSTF in a letter dated December 11, 2015, Nuclear Regulatory Commission (NRC) Agencywide Documents Access and Management System, Accession No. ML15314A305. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-545. l&M concluded that the justifications presented in TSTF-545, and the model safety evaluation prepared by the NRC staff are applicable to CNP Units 1 and 2 and justify this amendment for the incorporation of the changes to the CNP TS.

CNP Unit 1 was issued a construction permit on March 25, 1969, and the provisions of 10 CFR 50.55a(f)(1) are applicable.

CNP Unit 2 was issued a construction permit on March 25, 1969, and the provisions of 10 CFR 50.55a(f)(1) are applicable.

2.2 Optional Changes and Variations r

No technical variances are proposed in this amendment request. The following bulleted items identify administrative type variations. These differences do not result in any technical conflict with the intent of TSTF-545 or the associated model SE. Therefore, l&M has concluded the following variations are both acceptable and permit NRC review of the proposed adoption of TSTF-545 under the Consolidated Line Item *Improvement Process.

to AEP-NRC-2016-29 Page2 Some minor formatting clean-up is performed, where needed, to adjust available space on a page or in a table cell.

The CNP TSs include plant-specific surveillances that are not contained in NUREG-1431 and/or have numbering that does not match NUREG-1431.

These differences are administrative and do not affect the applicability of TSTF-545 to the CNP TSs.

With reference to Surveillance Requirement (SR) 3.4.10:1 and SR 3.7.2.1, the IST reference also appears in the SR column of the table and, therefore, is capitalized consistent with the intent of TSTF-545.

l&M performed a search of the entire CNP Unit 1 and Unit 2 TSs for the key phrase "inservice testing program" and "IST". In addition to that already described above, the TS Bases may h~ve included other examples of this phrase, all of which are not capitalized as a defined term, consistent with TSTF-545.

(Note that the TS Bases markups included in and 6 are submitted for information only).

TSTF-545 deletes the IST program, TS 5.5.6, and renumbers all subsequent TS programs.

This also impacts several TS Bases references.

l&M proposes to retain the TS 5.5.6 reference, now shown as "DELETED", and not change the subsequent TS program numbers. These program numbers are referenced in a number of station procedures. By maintaining the current program numbering, excessive administrative burden is avoided.

Based on this approach, several TSTF-545 TS Bases markup pages associated with the TSTF-545 program numbering are not included in Enclosure 5 and 6 of this application.

3.0 REGULATORY ANALYSIS

3. 1 No significant Hazards Consideration Determination l&M requests adoption of the TS changes described in TSTF-545, "TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing," which is an approved change to the Improved Standard Technical Specifications, into the CNP Unit 1 and 2 TSs. The proposed change revises the TS Chapter 5, "Administrative Controls," Section 5.5, "Programs and Manuals,"

to delete the "lnservice Testing Program" specification.

Requirements in the IST Program are removed, as they are duplicative of requirements in the ASME Operations and Maintenance (OM)

Code, as clarified by Code Case OMN-20, "lnservice Test Frequency." Other requirements in Section 5.5 are eliminated because the NRG has determined their appearance in the TS is contrary to regulations. A new defined term, "lnservice Testing Program," is added, which references the requirements of Title 10 of the code of Federal Regulations (10 CFR), Part 50, paragraph 50.55a(f).

l&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 or consequences of an accident previously evaluated?

Response: No

/ to AEP-NRC-2016-29 Page 3 The proposed change revises TS Chapter 5, "Administrative Controls," Section 5. 5, "Programs and Manuals," by eliminating the "lnservice Testing Program" specification. Most requirements in the IST Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, "lnservice Test Frequency." The remaining requirements in the Section 5.5.6 IST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, "lnservice Testing Program," is added to the TS, which references the requirements of 10 CFR 50.55a(f).

Performance of IST is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change.

lnservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components.

As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.

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

2.

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

Response: No The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed.

The proposed change does not alter the types of insevice testing performed.

In most cases, the frequency of IST is unchanged.

However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

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 change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling to AEP-NRC-2016-29 Page4 and consideration of plant operating conditions that may not be suitable for performance of the required testing.

The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS SR 3.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability.

This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the TS provide actions to ensure that the margin of safety is protected.

The proposed change also eliminates a statement. that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety.

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

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.

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 1 O CFR 20, or would change an inspection or SR.

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 inqividual 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 1 O CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

to AEP-NRC-2016-29 Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes

UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS.

TABLE OF CONTENTS Chapter/Specification 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility.......................................................... ;............................................... :********* 5.1-1

. 5.2 Organization....................................................................................................................... 5.2-1 5.2.1 Onsite and Offsite Organizations.................................................................................... 5.2.:.1 5.2.2 Unit Staff........................................................................................................................ 5.2-1 5.3 Unit Staff Qualifications..................................................................................................... 5.3-1

.5.4 Procedures....................................................................................................................... 5.4-1

5.5 Programs and Manuals..................................................................................................... 5.5-1 5.5.1 Offsite Dose Calculation Manual (ODCM)............................................................. :........ 5.5-1 5.5.2 Leakage Monito ring Program.................................. *................................ *................... :........................................ 5.5-2 5.5.3 Radioactive Effluent Controls Program........................................................................... 5.5-2 5.5.4 Component Cyclic or Transient Limits........,..................................... :.............................. 5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program..................................................... 5.5-4 5.5.6 lnservice Testing ProgramlDeletedl................................................................................. 5.5-4 5.5.7 Steam Generator (SG) Program..................................................................................... 5.5-5 5.5.8 Secondary Water Chemistry Program............................................................................ 5.5-8 5.5.9 Ventilation Filter Testing Program (VFTP)....................................................................... 5.5-8 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program.............................. 5.5-11 5.5.11 Diesel Fuel Oil Testing Program....................................... ;............................................. 5.5-11 5.5.12 Technical Specifications (TS} Bases Control Program................................................... 5.5-12 5.5.13 Safety Function Determination Program (SFDP)............................................................ 5.5-13 5.5.14 Containment Leakage Rate Testing Program................................................................. 5.5-14.

5.5.15 Battery Monitoring and Maintenance Program******-~*-************'**************************************** 5.5-15 5.5.16 Control Room Envelope Habitability Program..............................................,,................ 5.5-15

5.6 Reporting Requirements................................................................................................... 5.6-1 5.6.1 Deleted.......................................................................................................................... 5.6-1 5.6.2 Annual Radiological Environmental Operating Report.................................................... 5.6-1,

5.6.3 Radioactive Effluent Release Report............................................................................... 5.6-2 5.6.4 Deleted.............................................. ;........................................................................... 5.6-2 5.6.5 CORE OPERATING LIMITS REPORT (COLR).............................................................. 5.6-2 5.6.6 Post Accident Monitoring Report.................................................................................... 5.6-4 5.6. 7 Steam Generator Tube Inspection Report...................................................................... 5.6-4 5.7 High Radiation Area.............................................. ;........................................................... 5.7-1 Cook Nuclear Plant Unit 1 Page 5 of 5 Amendment No. 287, 298, ~. 320

1.1 Definitions ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Cook Nuclear Plant Unit 1 -

Definitions 1.1 The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint 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 previously reviewed and approved by the NRC..

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 1 O. CFR 50.55a(f).

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 riot 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 LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

'1.1-3 Amendment No. 28-7, 298

Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE Verify each pressurizer safety valve is OPERABLE in accordance with the llNSERVICE TESTING!

IPROGRAMilnservice Testing Program. Following testing, lift settings shall be within +/- 1 %.

Cook Nuclear Plant Unit 1

. 3.4.10-2 FREQUENCY In accordance with the I nservice Testing Program INSERVICEI rrESTINGI PROGRAM!

Amendment No. 287

ACTIONS (continued)

CONDITION REQUIRED ACTION A.2 Isolate the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.

B. Required Action and B.1 Be in MODE 3.

associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5.

C. RHR System interlock C.1 Isolate the affected function inoperable.

penetration by use of one closed manual or deactivated automatic valve.

SURVEILLANCE REQUIREMENTS SR 3.4.14.1 SURVEILLANCE

NOTE------------------------------

On ly required to be performed in MODES 1 and 2.

Verify leakage from each RCS PIV is equivalent to

0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure;

: 2215 psig and ::::;; 2255 psig.

RCS PIV Leakage 3.4.14 COMPLETION TIME 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 6 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 4 hours FREQUENCY In accordance with the lnservice Testing Program INSERVICEI ITESTINGI PROGRAM Cook Nuclear Plant Unit 1 3.4.14-2 Amendment No. 2-87, 329

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.

Number Position Function 1-IM0-261 Open SI suction line 1-IM0-262 Open Mini flow line 1-IM0-263 Open Mini flow line 1-IM0-315 Closed Low head SI to hot leg 1-IM0-325 Closed Low head SI to hot leg 1-IM0-390 Open RWSTto RHR 1-ICM-305 Closed Sump line 1-ICM-306 Closed Sump line SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.3 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.

SR 3.5.2.4 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.5 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 1 3.5.2-2 ECCS - Operating 3.5.2 FREQUENCY 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 31 days In accordance with the IASeFViGe

+estiA§ PFe§FaFR INSERVICE!

TESTING!

PROGRAM[

24 months 24 months Amendment No. 287

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.4 Verify the isolation time of each automatic power In accordance operated containment isolation valve is within limits.

with the lnservice T,..,-.a.~--

n--~...,.,~

-.~*-

INSERVICEI TESTING!

PROGRAM!

SR 3.6.3.5 Verify each automatic containment isolation valve 24 months that is not locked, sealed, or otherwise secured in position, actuates to.the isolation position on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 1 3.6.3-5 Amendment No. 287

3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System Containment Spray System 3.6.6 LCO 3.6.6 Two containment spray trains shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A.1 Restore containment spray 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train inoperable.

train to OPERABLE status.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND B.2 Be in MODE 5.

84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.6.2 Verify each containment spray pump's developed In accordance head at the flow test point is greater than or equal to with the IAsei:vise the required developed head.

+estiA§ PFe§FaFA INSERVICEI lfESTINGI PROGRAM!

Cook Nuclear Plant Unit 1 3.6.6-1 Amendment No. 287

ACTIONS (continued)

CONDITION REQUIRED ACTION B. Required Action and B.1 Be in MODE 3.

associated Completion Time of Condition A not AND met.

One or more steam generators with

4 MSSVs inoperable.

B.2 SURVEILLANCE REQUIREMENTS Be in MODE 4.

SURVEILLANCE MSSVs 3.7.1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours FREQUENCY SR 3.7.1.1

NOTE------------------------------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV is OPERABLE in accordance with the llNSERVICE TESTING!

IPROGRAMJlnservice Testing Programwith the lift setpoint per Table 3.7.1-2. Following testing, lift setting shall be within +/-1 %.

Cook Nuclear Plant Unit 1 3.7.1-2 In accordance with the lnservice Testing Program IN SERVICE[

TESTING!

PROGRAM!

Amendment No. 287

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.2.1

N 0 TE------------------------------

Only required to be performed in MODES 1 and 2.

Verify the isolation time of each SGSV is within limits.

SR 3.7.2.2

NOTE------------------------------

Only required to be performed in MODES 1 and 2.

Verify each SGSV actuates to the isolation position on an actual or simulated actuation signal.

SGSVs 3.7.2 FREQUENCY In accordance with the I nservice T--~'-- D----..._,....,

~,..,*-

INSERVICEI TESTING!

PROGRAM!

24 months Cook Nuclear Plant Unit 1 3.7.2-2 Amendment No. 2-87, 318

ACTIONS (continued)

CONDITION D. Required Action and associated Completion Time not met.

REQUIRED ACTION D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE MFIVs and MFRVs 3.7.3 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours FREQUENCY SR 3.7.3.1 Verify the isolation time of each MFIV is within limits.

In accordance with the lnservice T--.a.:-.- M---.---....

--,11 I~ I

-.~* -*

INSERVICEI TESTING/

PROGRAM!

SR 3.7.3.2 Verify the isolation time of each MFRV is within In accordance limits.

with the lnservice

+estin§ PFe§Fam INSERVICEI rrESTINGI PROGRAM/

SR 3.7.3.3 Verify each MFIV and MFRV actuates to the 24 months isolation position on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 1 3.7.3-2 Amendment No. 28-7, 318

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.5.1

N()TE------------------------------

AFW train(s) may be considered ()PERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

Verify each required AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2

N()TE------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after;;:: 850 psig in the steam generator.

Verify the developed head of each required AFW pump at the flow test point is greater than or equal to the required developed head.

SR 3.7.5.3

N()TES----------------------------

1.

AFW train(s) may be considered ()PERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

2.

()nly required to be met in M()DES 1, 2, and 3.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 1 3.7.5-3 AFW System 3.7.5 FREQUENCY 31 days In accordance with the IRseFVise

+estiR§ PFe§Fam IN SERVICE!

TESTING!

PR()GRAMi 24 months Amendment No. 287

5.5 Programs.and Manuals Programs and Manuals 5.5 5.5.5 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel.

A qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle testing or penetrant testing, or combination of the two tests) of exposed surfaces of the removed flywheels shall be conducted once every 10years.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequen~y.

- _______ -~ ___ _

5.5.6

!DELETED llnservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves.

a.

Testing Frequencies specified in the ASME Operation and Maintenance Standards and Guides (OM Codes) and applicable Addenda are as follows:

ASME OM Codes and applicable Addenda terminology for.

inservice testing activities Quarterly or every 3 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 92 days At least once per 366 days At least once per 731 days

b.

The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities.

c.

The provisions of SR 3.0.3 are applicable to inservice testing activities.

d.

Nothing in the ASME OM Codes shall be construed to supersede the requirements of. any TS.

Cook Nuclear Plant Unit 1 5.5:-4 Amendment No. 287 to AEP-NRC-2016-29 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes

UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility............... *.*................................................................................................... 5.1-1 5.2 Organization.................................................................... :....................... :......................... 5.2-1 5.2.1 Onsite and Offsite Organizations.................................................................................... 5.2-1 5.2.2 Unit Staff.~................................................................................................'...................... 5.2-1 5.3 Unit Staff Qualifications................................... ~......................... ;....................................... 5.3-1 5.4 Proced.ures....................................................................................................................... 5.4-1 5.5 Programs and Manuals..................................................................................................... 5.5-1 5.5.1 Offsite Dose Calculation Manual (ODCM)...................................................................... 5.5-1 5.5.2 Leakage Monitoring Program................................................ *......................................... 5.5-2 5.5.3 Radioactive Effluent Controls Program........................................................................... 5.5-2

5.5.4 Component Cyclic or Transient Limits........................................................... :................ 5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program........,............................................ 5.5-4 5.5.6 lnsePJice Testing ProgramlDeletedl................................................................................. 5.5-4 5.5.7 Steam Generator (SG) Program..................................................................................... 5.5-5 5.5.8 Secondary Water Chemistry Program............................................................................ 5.5-8 5.5.9 Ventilation Filter Testing Program (VFTP)...................................................................... 5.5-8 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program.............................. 5.5-11 5.5.11 Diesel Fuel Oil Testing Program......*.............................................................................. 5.5-12 5.5.12 Technical Specifications (TS) Bases Control Program................................................... 5.5-12 5.5.13 Safety Function Determination Program (SFDP)............................................................. 5.5-13 5.5.14 Containment Leakage Rate Testing Program................................................................. 5.5-14 5.5.15 Battery Monitoring and Maintenance Program............................................................... 5.5-14 5.5.16 Control Room Envelope Habitability Program................................................................ 5."5-15 5.6 Reporting Requirements,.................................................................................................. 5.6-1 5.6.1 Deleted.......................................................................................................................... 5.6-1 5.6.2 Annual Radiological Environmental Operating Report.................................................... 5.6-1 5.6.3 Radioactive Effluent Release Report....... _.......................................... :............................ 5.6-2

.5.6.4 Deleted-............................................*............................................................................. 5.6-2

5.6.5 CORE OPERATING LIMITS REPORT (COLR).............................................................. 5.6-2 5.6.6 Post Accident Monitoring Report.................................................................................... 5.6-4 5.6. 7 Steam Generator Tube Inspection Report...................................................................... 5.6-4 5.7 High Radiation Area........................................................ ~............................. :................... 5.7-1 Cook Nuclear Plant Unit 2 Page 5 of 5 Amendment No. 299, 219, ~. JG2, 304

1.1 Definitions ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Cook Nuclear Plant Unit 2 Definitions 1.1

. The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exc~eds its actuation setpoint 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 previously reviewed and approved by the NRC.

The INSERVICE TESTING PROGRAM is the licensee

. program that fulfills the requirements of 10 CFR 50.55A(f).

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 LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component bQdy, pipe wall, or vessel wall.

1.1-3 Amendment No. 299, 279

-Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE Verify each pressurizer safety valve is OPERABLE.

in accordance with the lnservice Testing Program

/INSERVICE TESTING PROGRAM!. Following testing, lift settings shall be within +/- 1 %.

Cook Nuclear Plant Unit 2 3.4.10-2 FREQUENCY In accordance with the lnservice Testing Program INSERVICEI rrESTINGI PROGRAM!

Amendment No. 269

ACTIONS (continued)

CONDITION A.2 B. Required Action and B.1 associated Completion Time of Condition A not AND met.

B.2 REQUIRED ACTION Isolate the high pressure portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.

Be in MODE 3.

Be in MODE 5.

RCS PIV Leakage 3.4.14 COMPLETION TIME 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 6 hours 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. RHR System interlock C.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months function inoperable.

penetration by use of one closed manual or deactivated automatic valve.

SURVEILLANCE REQUIREMENTS SR 3.4.14.1 SURVEILLANCE

NOTE------------------------------

Only required to be performed in MODES 1 and 2.

Verify leakage from each RCS PIV is equivalent to

0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure;;: : 2215 psig and:::; 2255 psig.

Cook Nuclear Plant Unit 2 3.4.14-2 FREQUENCY In accordance with the lnservice Testing Program INSERVICEI TESTING!

PROGRAM!

Amendment No. 269

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.

Number Position Function 2-IM0-261 Open SI suction line 2-IM0-262 Open Mini flow line 2-IM0-263 Open Mini flow line 2-IM0-315 Closed Low head SI to hot leg 2-IM0-325 Closed Low head SI to hot leg 2-IM0-390 Open RWSTto RHR 2-ICM-305 Closed Sump line 2-ICM-306 Closed Sump line SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the

correct position.

SR 3.5.2.3 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.

. SR 3.5.2.4 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.5 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 2 3.5.2-2 ECCS - Operating 3.5.2 FREQUENCY 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 31 days In accordance with the IRseFVise

+estiR§ PFe§Fam IN SERVICE\\

TESTING\\

PROGRAM!

24 months 24 months Amendment No. 269

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.4 Verify the isolation time of each automatic power-In accordance _-

operated containment isolation valve is within limits.

with the lnservice -

T'"'~~=-~ n~~~~....,....,

INSERVICE/

TESTING!

PROGRAM!

SR 3.6.3.5 Verify each automatic containment isolation valve 24 months that is not locked, sealed, or otherwise *secured in position, actuates to the isolation position on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 2 3.6.3-5 Amendment No. 269

3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System Containment Spray System 3.6.6 LCO 3.6.6 Two containment spray trains shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A One containment spray A.1 Restore containment spray 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train inoperable.

train to OPERABLE status.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND B.2 Be in MODE 5.

84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.6.2 Verify each containment spray pump's developed In accordance.

head at the flow test point is greater than or equal to with the IRsePJiGe the required developed head.

+estiR§ PFe§Fam INSERVICEI TESTINGJ PROGRAM!

Cook Nuclear Plant Unit 2 3.6.6-1 Amendment No. 269

ACTIONS (continued)

CONDITION REQUIRED ACTION B. Required Action and B.1 Be in MODE 3.

associated Completion Time of Condition A not AND met.

One or more steam generators with

~ 4 MSSVs inoperable.

B.2 SURVEILLANCE REQUIREMENTS Be in MODE 4.

SURVEILLANCE MSSVs 3.7.1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours FREQUENCY SR 3.7.1.1

NOTE--------------:.. ______________ :_

Only required to be performed in MODES 1 and 2.

Verify each required MSSV is OPERABLE in accordance with the lnservice Testing Program llNSERVICE TESTING PROGRAMJ.with the lift setpoint per Table 3. 7.1-2. Following testing, lift setting shall be within +/-1 %.

Cook Nuclear Plant Unit 2 3.7.1-2 In accordance with the lnservice Testing Program INSERVICE/

ifESTINGI PROGRAM!

Amendment No. 269

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.2.1

N 0 TE------------------------------

Only required to be performed in MODES 1 and 2.

Verify the isolation time of each SGSV is within limits.

SR 3.7.2.2

~------------------NOTE------------------------------

Only required to be performed in MODES 1 and 2.

Verify each SGSV actuates to the isolation position on an actual or simulated actuation signal.

SGSVs 3.7.2 FREQUENCY In accordance with the lnservice T.-.-*:-- n---'""'"'"'

INSERVICEI TESTING!

PROGRAM!

24 months Cook Nuclear Plant Unit 2 3.7.2-2 Amendment No. ~. 301

ACTIONS (continued)

CONDITION D. Required Action and associated Completion Time not met.

REQUIRED ACTION D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

SURVEILLANCE REQUIREMENTS SURVEILLANCE MFIVs and MFRVs 3.7.3 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours FREQUENCY SR 3.7.3.1 Verify the isolation time of each MFIV is within limits.

In accordance with the lnservice T.--~=-- D------

~*

INSERVICEI ifESTINGI PROGRAM!

SR 3.7.3.2 Verify the isolation time of each MFRV is within In accordance limits.

with the lnservice

+estin§ Pre§ram INSERVICEI TESTING!

PROGRAM!

SR 3.7.3.3 Verify each MFIV and MFRV actuates to the 24 months

'isolation position on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 2 3.7.3-2 Amendment No. 269, 301

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.5.1

~-N()TE------------------------------

AFW train(s) may be considered ()PERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

Verify each required AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2

N()TE------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after <:: 850 psig in the steam generator.

Verify the developed head of each required AFW pump at the flow test point is greater than or equal to the required developed head.

SR 3.7.5.3

~-----------------------N()TES--~-------------------------

1.

AFW train(s) may be considered ()PERABLE during alignment and operation for steam generator level control, if it is, capable of being manually realigned to the AFW mode of operation.

2.

Only required to be met in M()DES 1, 2, and 3.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

Cook Nuclear Plant Unit 2 3.7.5-3 AFW System 3.7.5 FREQUENCY 31 days In accordance with the IRsePJise

+estiR§ PFe§Fam INSERVICEI rrESTING!

PR()GRAMI 24 months Amendment No. 269

5.5 Programs and Manuals Programs and Manuals 5.5 5.5.5 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel.

A qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle testing or penetrarit testing, or combination of the two tests) of exposed surfaces of the removed flywheels shall be conducted once every 10 years.

The provisions of SR 3.0.2 and SR 3.0.3, are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency.

5.5.6

!DELETED! lnservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves.

a.

Testing Frequencies specified in the ASME Operation and Maintenance Standards and Guides (OM Codes) and applicable Addenda are as follmvs:

ASME OM Codes and applicable Addenda terminology for inservice testing activities Quarterly or every 3 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 92 days At least once per 366 days At least once per 731 days

b.

The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities.

c.

The provisions of SR 3.0.3 are applicable to inservice testing activities.

d.

Nothing in the ASME OM Codes shall be construed to supersede the requirements of any TS.

Cook Nuclear Plant Unit 2 5.5-4 Amendment No. 269 to AEP-NRC-2016-29 Donald C. Cook Nuclear Plant Unit 1 Technical Specification Bases Pages Marked to Show Proposed Changes

{For Information Only)

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.9 and a I at all times, unless otherwise stated. SR 3.0.2 and 3.0.3 a I in Cha ter 5 onl SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems_and c.omponents, an_d_.. *~~- *- __ ---. -*-

that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes 'a failure to meet an LCO. Surveillances may be performed by means of

any series of sequential, overlapping, or tptal steps provided the entire Surveillance is performed within the specified Frequency. Additionally, the definitions related to instrument testing (e.g., CHANNEL CALIBRATION) specify that these tests are performed by means of any series of sequential, overlapping, or total steps.

Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a.. The systems or components are known to be inoperable, although still meeting the SRs; or

b.

The requirements of the Surveillance(s) are known not to be met between required Surveillance performances..

Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated.

LCO are not applicable, unless otherwise specified. The SRs associated with a test exception are only applicable when the test exception is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR.

Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.

Cook Nuclear Plant Unit 1 B 3.0-14 Revision No. 49

BASES SR Applicability B 3.0 SR 3.0.1 (continued)

SR 3.0.2 Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

An example of this process is Auxiliary feedwater (AFW) pump turbine maintenance during refueling that requires testing at steam pressures

~ 850 psig. However, if other appropriate testing is satisfactorily completed, the AFW System can be considered OPERABLE. This allows startup and other necessary testing to proceed until the plant reaches the steam pressure required to perform the testing.

l SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per... " interval.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).

When a Section 5.5, "Programs and Manuals," specification states that the provision of SR 3.0.2 are applicable, a 25% extension of the testing interval, whether stated in the specification or incorporated by reference, is permitted.

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.

These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. AA--e[§xample~

of where SR 3.0.2 does not apply are*

the Containment Leakage Rate Testing Program re uired b 10 CFR 50, A endix J, and the inservice.

Cook Nuclear Plant Unit 1 B 3.0-15 Revision No. 49

/

SR Applicability B 3.0 testing of pumps and valves in accordance with applicable American Societ of Mechanical Engineers 0 eration and Maintenance Code, as re uired b 10 CFR 50.55a.

hese program establishes testing requirements and Frequencies in accordance with the requirements of regulations. The TS canno, in and of themselves, extend a test interval specified in the regulations direct! orb referenc.

Cook Nuclear Plant Unit 1 B 3.0-16 Revision No. 49

BASES SR 3.0.2 (continued)

SR 3.0.3 SR Applicability B 3.0 As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per... " basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to.this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specifi~d.

SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

When a Section 5.5, "Programs and Manuals," specification states that the provisions of SR 3.0.3 are applicable, it permits the flexibility to defer declaring the testing requirement not met in accordance with SR 3.0.3 when the testing has not been completed within the testing interval (including the allowance of SR 3.0.2 if invoked by the Section.5.5 specification).

This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements. When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Cook Nuclear Plant Unit 1 B 3.0-17 Revision No. 49

BASES SURVEILLANCE REQUIREMENTS REFERENCES SR 3.4.10.1 Pressurizer Safety Valves B 3.4.10 SRs are specified in the [INSERVICE TESTING PROGRAMJlnservice Testing Program. Pressurizer safety valves are to be tested ih accordance with the requirements of the ASME OM Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No additional requirements are specified.

The pressurizer safety valve setpoint is +/- 3% for OPERABILITY; however, the valves are reset to +/- 1 % during the Surveillance to allow for drift.

1.

ASME, Boiler and Pressure Vessel Code, Section llL ________ _

2.

UFSAR, Chapter 14.

3.

WCAP-7769, Rev. 1, June 1972.

4.

ASME, Operation and Maintenance Standards and Guides (OM Codes).

Cook Nuclear Plant Unit 1 B 3.4.10-4 Revision No. 2

BASES RCS PIV Leakage B 3.4.14 ACTIONS (continued)

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time after exceeding the limit considers the time required to complete the Required Action and the low probability of a second valve failing during this time period.

8.1and8.2 If any Required Action and associated Completion Time of Condition A is not met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This Action may

-reduce the leakage and also reduces the potential for a LQCA o~tside_the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

The inoperability of the RHR interlock renders the RHR suction isolation valves incapable of preventing inadvertent opening of the valves at RCS pressures in excess of the RHR systems design pressure. If the RHR interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one closed manual or deactivated power operated valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Required Action accomplishes the purpose of the function.

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV is required to verify that leakage is below the specified limit and to identify each leaking valve.

The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

The Frequency required by the llNSERVICE TESTING PROGRAMJ lnservice Testing Program is within frequency allowed by the American Society of Mechanical Engineers (ASME) OM Code (Ref. 9), and is based on the need to perform such Surveillances under the conditions that apply during an outage and the Cook Nuclear Plant Unit 1 B 3.4.14-4 Revision No. 0

BASES ECCS - Operating B 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.3 Verifying that each ECCS pump's developed head at the flow test point is greater than or equal to the required developed head ensures that ECCS pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of ECCS pump performance required by the ASME OM Code (Ref. 10). Since the ECCS pumps cannot be tested with flow through the normal ECCS flow paths, they are tested on recirculation flow (RHR and SI pumps) or normal charging flow path (centrifugal charging pumps). This test confirms one point on the pump design curve and is indicative of overall perf9!rnanc~: _________ _

Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance. The Fre uency of this SR is in accordance with the INSERVICE TESTING PROGRAM lnservice Testing Program.

SR 3.5.2.4 and SR 3.5.2.5 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are lpcked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a unit outage and the potential for unplanned unit transients if the Surveillances were performed with the reactor at power. The 24 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.

SR 3.5.2.6 Proper throttle valve position is necessary for proper ECCS performance.

These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This *surveillance verifies the mechanical stop of each listed ECCS throttle valve is in the correct position. The 24 month Frequency is based on the same reasons as those stated in SR 3.5.2.4 and SR 3.5.2.5.

Cook Nuclear Plant Unit 1 B 3.5.2-8 Revision No. 24

BASES Containment Isolation Valves B 3.6.3 APPLICABLE SAFETY ANALYSES (continued)

LCO APPLICABILITY while purging, and limiting containment purge operation to using no more than one supply path and one exhaust path at a time. The containment purge valves have been demonstrated capable of closing ag~inst the dynamic forces associated with a LOCA and are assured of receiving a containment ventilation isolation signal.

The Containment Isolation Valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Containment isolation valves form a part of the containment boundary.

The containment isolation valves' safety function is related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during a OBA.

The automatic power operated isolation valves are required to have isolation times within limits and to actuate on an automatic isolation signal. The valves covered by this LCO are listed in the UFSAR Ref. 3) and the associated stroke times are listed in he INSERVICE TESTING

!PROGRAM! lnservice Testing Program.

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves/devices are those listed in Reference 3.

The containment isolation valve leakag~ rates are addressed by LCO 3.6.1, "Containment," as Type C testing.

This LCO provides assurance that the containment isolation valves and purge valves will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.

In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation valves during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations.';

Cook Nuclear Plant Unit 1

. B 3.6.3-3 Revision No. 0

, BASES Containment Isolation Valves B 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.

3.4 REFERENCES

Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.

The Frequency of this SR is in accordance with the llNSERV!CEI

!TESTING PROGRAMJ In.service Testing Program.

SR 3.6.3.5 Automatic containment Isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a OBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power~ Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standppint.

1.

UFSAR, Section 14.3.4.

2.

UFSAR, Section 14.2.6.

3.

UFSAR, Section 5.4.1 and Table 5.4-1.

Cook Nuclear Plant Unit 1 B 3.6.3-9 Revision No. 0

BASES Containment Spray System B 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

REFERENCES incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the llNSERVICE TESTING PROGRAM!

lnservice Testing Program.

SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power. Operating experience has shown these components us_ually pass the Surveillances when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

These Surveillances include a Note that states that in MODE 4, only the manual portion of the actuation signal is required. This is acceptable since the automatic portion of the actuation signal is not required to be OPERABLE by ITS 3.3.2, "Engineered Safety Features Actuation System (ESFAS) Instrumentation."

SR 3.6.6.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. This SR ensures that each spray nozzle is unobstructed.and that spray coverage of the containment during an accident is not degraded. The event based surveillance frequency following maintenance that could result in nozzle blockage was chosen because this passive portion of the system is not susceptible to service induced degradation.

1.

UFSAR, Section 1.4.7.

2.

UFSAR, Section 14.3.4.

3.

10 CFR 50.49.

4.

10 CFR 50, Appendix K.

5.

ASME, Operation and Maintenance Standards and Guides (OM Codes).

Cook Nuclear Plant Unit 1 B 3.6.6-6 Revision No. 32

BASES LCO (continued)

APPLICABILITY ACTIONS MSSVs B 3.7.1 The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances, to relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the llNSERVICE TESTING PROGRAM!

lnservice Testing Program.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB, or Main Steam System integrity.

In MODES 1, 2, and 3, five MSSVs per steam generator are required to be OPERABLE to prevent Main Steam System overpressurization.

In MODES 4 and 5, there are no credible transients requiring the MSSVs.

The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

The ACTIONS Table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

A.1 and A.2 With one or more inoperable MSSVs on one or more steam generators, Required Action A.1 requires an appropriate reduction in reactor power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . However, with a reactor power reduction alone there may be insufficient total steam flow capacity provided by the remaining OPERABLE MSSVs to preclude overpressurization in the event of a turbine trip without steam dump. Therefore, a Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is allowed in Required Action A.2 to reduce the setpoints. The Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is based on a reasonable time to correct the MSSV inoperability, the time required to perform the power reduction, operating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period.

The maximum THERMAL POWER corresponding to the heat removal capacity of the remaining OPERABLE MSSVs is determined via a conservative heat balance calculation as described in the attachment to Reference 5, with an appropriate allowance for Nuclear Instrumentation System trip channel uncertainties.

Cook Nuclear Plant Unit 1 B 3.7.1-3 Revision No. 0

BASES MSSVs B 3.7.1 ACTIONS (continued)

SURVEILLANCE REQUIREMENTS Required Action A.2 is modified by a Note, indicating that the Power Range Neutron Flux-High reactor trip setpoint reduction is only required in MODE 1. In MODES 2 and 3 the Reactor Trip System trips specified in LCO 3.3.1, "Reactor Trip System Instrumentation," provide sufficient protection.

The allowed Completion Times are reasonable based on operating experience to accomplish the Required Actions in an orderly manner without challenging unit systems.

B.1 and B.2 If any Required Action and associated Completion Time is not met, or if one or more steam generators have ;::: 4 inoperable M$SVs, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SR 3.7.1.1 This SR verifies the OPERABILITY of the MSSVs b the verification of each MSSV lift setpoint in accordance with the INSERVICE TESTING IPROGRAMI lnservice Testing Program. The ASME Code,Section XI (Ref. 6), requires that safety and relief valve tests be performed in accordance with ANSl/ASME OM-1-1987 (Ref. 7).

The ASME Code specifies the activities and frequencies necessary to satisfy the requirements. Table 3.7.1-2 allows a+/- 3% setpoint tolerance for OPERABILITY; however, the valves are reset to +/- 1 % during the Surveillance to allow for drift. The lift settings, according to Table 3. 7.1-2, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure. If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

Cook Nuclear Plant Unit 1 B 3.7.1-4 Revision No. 0

BASES SGSVs B 3.7.2 ACTIONS (continued)

SURVEILLANCE REQUIREMENTS achieve this status, the unit must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Condition C would be entered. The Completion Time is reasonable, based on operating experience, to reach MODE 2 in an orderly manner and without challenging unit systems.

C.1 and C.2 Condition C is modified by a Note indicating that separate Condition entry is allowed for each SGSV.

Since the SGSVs are required to be OPERABLE in MODES 2 and 3, the inoperable SGSVs must be closed. When closed, the SGSVs are already in the position required by the assumptions in the safety analysis.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is consistent with that allowed in Condition A.

For inoperable SGSVs that are closed, the inoperable SGSVs must be verified on a periodic basis to be closed. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering judgment, in view of SGSV status indications available in the control room, and other administrative controls, to ensure that these valves are in the closed position.

D:1 and D.2

. If the SGSVs are not closed within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply.. To

.achieve this status, the unit must be placed at least in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from MODE 2 conditions in an orderly manner and without challenging unit systems.

SR 3.7.2.1 This SR verifies that SGSV closure time is within the limit given in.

Reference 4 and is within that assumed in the accident analyses. The.

valve(s) may also be tested to more restrictive requirements in accordance with the llNSERVICE TESTING PROGRAMJ lnservice Testing Program. The SR is normally performed upon returning the unit to operation following a refueling outage. The SGSVs should not be tested at power, since a unit trip could occur. As the SGSVs are not tested at power, they are exempt from the ASME OM Code (Ref. 5) requirements during operation in MODE 1 or 2.

Cook Nuclear Plant Unit 1 B 3.7.2-3 Revision No. 34

BASES SGSVs.

B 3.7.2 SURVEILLANCE REQUIREMENTS (continued)

REFERENCES The Frequency is in accordance with the llNSERVICE TESTING!

!PROGRAM! lnservice Testing Program.

This test is conducted in MODE 3 with the uni_t at operating temperature and pressure. This SR is modified by a Note that allows entry into and operation*in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

SR 3.7.2.2 This SR verifies that each SGSV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The Frequency of -

SGSV testing is every 24 months. The 24 month Frequency for testing is based on equipment reliability. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.

1.

UFSAR, Section 10.2.

2.

UFSAR, Section 14.2.5.

3.

10 CFR 100.11.

4.

Technical Requirements Manual

5.

ASME, Operations and Maintenance Standards and Guides (OM Codes).

Cook Nuclear Plant Unit 1 B 3.7.2-4 Revision No. 34

BASES MFIVs and MFRVs B 3.7.3 ACTIONS (continued)

With both the MFIV and MFRV inoperable in the same flow path, there is no redundant system to operate automatically and perform the required safety function. Under these conditions, the affected flow path must be isolated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This action returns the system to the condition where at least one valve in each flow path is performing the required safety function. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable, based on operating experience, to complete the actions required to close fhe MFIV or MFRV, or otherwise isolate the affected flow path.

D.1 and D.2 If any Required Action and associated Completion Time is not met, the unit must be placed iri a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVf::ILLANCE SR 3.7.3.1 and SR 3.7.3.2 REQUIREMENTS REFERENCES These SRs verify that the closure time of each MFIV and MFRV is within the limit given in Reference 2 and is within that assumed in the accident and transient analyses. The valve(s) may also be.tested to more restrictive requirements in accordance with the llNSERVICE TESTING!

!PROGRAM! lnservice Testing Program.

The Frequency for this SR is in accordance with the llNSERVICEI

!TESTING PROGRAM! lnservice Testing.Program.

SR 3.7.3.3 This SR verifies that each MFIV and MFRV can close on an actual or simulated actuation signal. Thi~ Surveillance is normally performed upon returning the unit to operation following a refueling outage.

The Frequency for this SR is every 24 months. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.

1; UFSAR, Section 10.5.1.2.

2.

Technical Requirements Manual Cook Nuclear Plant Unit 1 B 3.7.3-4 Revision No. 34

BASES AFW System B 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.5.2 Verifying that each required AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 2).

Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance. Performance of inservice testing a discussed in the ASME OM Code (Ref. 2) and the INSERVICE TESTING PROGRAM (only required at 3 month intervals) satisfies this requirement.

This SR,is modified by a Note indicating that the SR should be deferred for the turbine driven AFW pump until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test at eritry into MODE 3. At 850 psig, there is sufficient pressure to perform the test.

SR *3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.

This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

The 24 month Frequency is based on the need to perform this Surveillan~e under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is acceptable based on operating experience and the design reliability of the equipment.

The SR is modified by two Notes. Note 1 *states that one or more AFW trains may be considered OPERABLE during alignment and operation *for

steam generator level control, if it is capable of being manually (i.e.,

remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable. This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable;.

Cook Nuclear Plant Unit 1 B 3.7.5-8 Revision No. 0 to AEP-NRC-2016-29 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Bases Pages Marked to Show Proposed Changes (For Information Only)

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in* accordance with SR 3.0:2, constitutes a failure to meet an LCO. Surveillances may be performed by means of any series of sequential, overlapping, or total steps provided the entire Surveillance is performed within the specified Frequency. Additionally, the definitions related to instrument testing (e.g., CHANNEL CALIBRATION) speCify that these tests are performed by means of any series of sequential, overlapping, or total steps..

Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a.

The systems or components are known to be inoperable, although still meeting the SRs; or

b.

The requirements of the Surveillance(s) are known not to be met between required Surveillance performances.

Surveillances do not have to be performed when the unit is in a MOOE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a test exception are only applicable when the test exception is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR.

Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.

Cook Nuclear Plant Unit 2 B 3.0-14 Revision No. 48

BASES SR Applicability B 3.0 SR 3.0.1 (continued)

SR 3.0.2 Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

An example of this process is Auxiliary feedwater (AFW) pump turbine maintenance during refueling that requires testing at steam pressures

~ 850 psig. However, if other appropriate testing is satisfactorily

completed, the AFW System can be considered OPERABLE. This allows startup and other necessarY testing to proceed until the plant reaches the steam pressure required to perform the testing.

SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per... " interval.

SR 3.0.2 permits a 25% extension bf the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).

When a Section 5.5, "Programs and Manuals," specification states that the provisions of SR 3.0.2 are applicable, a 25% extension of the testing interval, whether stated in the specification or incorporated by reference, is permitted.

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.

These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. AA--e[§xampl~

of where SR 3.0.2 does not apply ar *

the Containment Leakage Rate Testing Program re uired b 10 CFR 50, A endix J, and the inservic Cook Nuclear Plant Unit 2 B3.0-15 Revision No. 48

Cook Nuclear Plant Unit 2 83.0-16 SR Applicability B 3.0 Revision No. 48

BASES SR Applicability B 3.0 SR 3.0.2 (continued)

SR 3.0.3 As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per... " basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes

. the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.

SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A*

delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater, applies from t~e point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

When a Section 5.5 "Programs and Manuals," specification states that the provisions of SR 3.0.3 are applicable, it permits the flexibility to defer declaring the testing requirement not met in accordance with SR 3.0.3 when the testing has not been completed within the testing interval (including the allowance of SR 3.0.2 if invoked by the Section 5.5 specification).

This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements. When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 1 O CFR 50, Cook Nuclear Plant Unit 2 B 3.0-17 Revision No. 48

BASES Pressurizer Safety Valves B 3.4.10 SURVEILLANCE SR 3.4.10.1 REQUIREMENTS REFERENCES SRs are specified in the JINSERVICE TESTING PROGRAMjlnservice Testing Program. Pressurizer safety valves are to be tested in accordance with the requirements of the ASME OM Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No additional requirements are specified.

The pressurizer safety valve setpoint is +/- 3% for OPERABILITY; however, the valves are reset to +/- 1 % during the Surveillance to allow for drift.

1.

ASME, Boiler and Pressure Vessel Code, Section Ill.

2.

UFSAR, Chapter 14.

3.

WCAP-7769, Rev. 1, June 1972.

4, ASME, Operation and Maintenance Standards and Guides (OM Codes).

Cook Nuclear Plant Unit 2 B 3.4.10-4 Revision No. 2 I

BASES RCS PIV Leakage B 3.4.14 ACTIONS (continued)-

SURVEILLANCE REQUIREMENTS The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time after exceeding the limit considers the time

required to complete the Required Action and the low probability of a second valve failing during this time period.

8.1 and 8.2 If any Required Action and associated Completion Time of Condition A is not met, the unit must be brought to a MODE in which the.requirement does not apply. To achieve this status, the unit must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

The inoperability of the RHR interlock renders the RHR suction isolation valves incapable of preventing inadvertent opening of the valves at RCS pressures in excess of the RHR systems design pressure. If the RHR interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one closed manual or deactivated power operated valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Required Action accomplishes the purpose of the function.

SR 3.4.14.1 Performance of leakage testing on each RCS PIV is required to verify that leakage is below the specified limit and to identify each leaking valve.

The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the~ other valve in series meets the*

leakage requirement. In this situation, the protection provided by redundant valves would be lost.

The Frequency required by the llNSERVICE TESTING PROGRAMJ lnservioe Testing Program is within frequency allowed by the American Society of Me~hanical Engineers (ASME) OM Code (Ref. 9), and is based on the need to perform such Surveillances* under the conditions that apply during an outage and the

_ Cook Nuclear Plant Unit 2 B 3.4.14-4 Revision No. 0

BASES ECCS - Operating B 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

REFERENCES

incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the llNSERVICE TESTING!

IPROGRAMilnservice Testing Program.

SR 3.5.2.4 and SR 3.5.2.5 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a unit outage and the potential for unplanned unit transients if the Surveillances were performed with the reactor at power. The 24 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.

SR 3.5.2.6 Proper throttle valve position is necessary for proper ECCS performance.

These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance verifies the mechanical stop of each listed ECCS throttle valve is in the correct position. The 24 month Frequency is based on the same reasons as those stated in SR 3.5.2.4 and SR 3.5.2.5.

SR 3.5.2.7 Periodic inspections of the containment sump suction inlets ensure that '

they are unrestricted and stay in proper operating condition. This Surveillance verifies that the sump suction inlets are not restricted by debris and the suction inlet strainers show no evidence of structural distress, such as openings or gaps, which would allow debris to bypass the strainers. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage, on the need to have access to the location. This Frequency has been found to be sufficient to detect abnormal degradation and is confirmed by operating experience.

1.

UFSAR, Section 1.4.7.

2.

10 CFR 50.46.

3.

UFSAR, Section 14.3.1.

Cook Nuclear Plant Unit 2

. B 3.5.2-8 Revision No. 16

BASES Containment Isolation Valves B 3.6.3 APPLICABLE SAFETY ANALYSES (continued)

LCO APPLICABILITY while purging, and limiting containment purge operation to using no more than one supply path and one exhau*st path at a time. The containment purge valves have been demonstrated capable of closing against the dynamic forces associated with a LOCA and are assured of receiving a containment ventilation isolation signal.

The Containment Isolation Valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Containment isolation valves form a part of the containment boundary.

The containment isolation valves' safety function is related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during a OBA.

The automatic power operated isolation valves are required to have

. isolation times within limits and to actuate on an automatic isolation signal. The valves covered by this LCO are listed in the UFSAR Ref. 3) and the associated stroke times are listed in the INSERVICE TESTING IPROGRAM!lnsePlice Testing Program.

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves/devices are those listed in Reference 3.

The containment isolation valve leakage rates are addressed by LCO 3.6.1, "Containment," as Type C testing.

This LCO provides assurance that the containment isolation valves and purge valves will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.

In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation valves during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations."

Cook Nuclear Plant Unit 2 B 3.6.3-3 Revision No. 0

BASES Containment Isolation Valves B 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.

3.4 REFERENCES

Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safet analyses.

The Frequency of this SR is in accordance with the INSERVICE lfESTING PROGRAMllnservice Testing Program.

SR 3.6.3.5 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a OBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

1.

UFSAR, Section 14.3.4.

2.

UFSAR, Section 14.2.6.

3.

UFSAR, Section 5.4.1. and Table 5.4-1.

Cook Nuclear Plant Unit 2 B 3.6.3-9 Revision No. 0

BASES Containment Spray System B 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

REFERENCES incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the llNSERVICE TESTING!

IPROGRAM!lnservice Testing Program.

SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray.

valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillances when performed at the 24 month Freguency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

These Surveillances include a Note that states that in MODE 4, only the manual portion of the actuation signal is required. This is acceptable since the automatic portion of the actuation signal is not required to be OPERABLE by ITS 3,3.2, "Engineered Safety Features Actuation System (ESFAS) Instrumentation."

SR 3.6.6.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown th~ough test connections. This SR ensures that each*spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded. The event based surveillance frequency following maintenance* that could result in nozzle blockage was chosen because this passive portion of the system is not susceptible to service induced degradation.

1.

UFSAR, Section 1.4.7.

2.

UFSAR, Section 14.3.4.

3.

10 CFR 50.49.

4.

10 CFR 50, Appendix K.

5.

ASME, Operation and Maintenance Standards and Guides (OM Codes).

Cook Nuclear Plant Unit 2 B 3.6.6-6 Revision No. 31

BASES LCO (continued)

APPLICABILITY ACTIONS MSSVs B 3.7.1 The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances, to relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the llNSERVICE TESTING!

IPROGRAMilnservice Testing Program.

This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could

result in a challenge.to the RCPB, or Main Steam System integrity.

In MODES 1, 2, and 3, five MSSVs per steam generator are required to be OPERABLE to prevent Main Steam System overpressurization.

In MODES 4 and 5, there are no credible transients requiring the MSSVs.

The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

The ACTIONS Table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

A.1 and A.2 With one or more inoperable MSSVs on one or more steam generators, Required Action A.1 requires an appropriate reduction in reactor power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . However; with a reactor power reduction alone there may be insufficient total steam flow capacity provided by the remaining OPERABLE MSSVs to preclude overpressurization in the event of a turbine trip without steam dump. Therefore, a Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is allowed in Required Action A.2 to reduce the setpoints. The Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is based on a reasonable time to correct the MSSV inoperability, the time required to perform the power reduction, op~rating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period.

The maximum THERMAL POWER corresponding to the heat rem.oval capacity of the remaining OPERABLE MSSVs is determined via a conservative heat balance calculation as described in the attachment to Reference 5, with an appropriate allowance for Nuclear Instrumentation System trip channel uncertainties.

Cook Nuclear Plant Unit 2 B 3.7.1-3 Revision No. 0

BASES MSSVs B 3.7.1 ACTIONS (continued)

SURVEILLANCE REQUIREMENTS Required Action A.2 is modified by a Note, indicating that the Power Range Neutron Flux-High reaetor trip setpoint reduction is only required in MODE 1. In MODES 2 and 3 the Reactor Trip System trips specified in LCO 3.3.1, "Reactor Trip System Instrumentation," provide sufficient protection.

The allowed Completion Times are reasonable based on operating experience to accomplish the Required Actions in an orderly manner without challenging unit systems.

B.1 and B.2 If any Required Action and associated Completion Time is not met, or if one or more steam generators have ;;::: 4 inoperable MSSVs, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in.

MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SR 3.7.1.1 This SR verifies the OPERABILITY of the MSSVs b the verification of each MSSV lift setpoint in accordance with the INSERVICE TESTING IPROGRAMllnservice Testing Program. The ASME Code,Section XI (Ref. 6), requires that safety and relief valve tests be performed in

.accordance with ANSl/ASME OM-1-1987 (Ref. 7).

The ASME Code specifies the activities and frequencies necessary to satisfy the requirements. Table 3.7.1-2 allows a+/- 3% setpoint tolerance for OPERABILITY; however, the valves are reset to+/- 1% during the Surveillance to allow for drift. The lift settings, according to Table 3.7:1-2, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift. pressure. If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

Cook Nuclear Plant Unit 2 B 3.7.1-4 Revision No. 0

BASES SGSVs B 3.7.2 ACTIONS (continued)

SURVEILLANCE REQUIREMENTS a_chieve this status, the unit must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Condition C would be entered. The Completion Time is reasonable, based on operating experience, to reach MODE 2 in an orderly manner and without challenging unit systems.

c~ 1 and C.2 Condition C is modified by a Note indicating that separate Condition entry is allowed for each SGSV.

Since the SGSVs are required to be OPERABLE iri MODES 2 and 3, the inoperable SGSVs must be closed. When closed, the SGSVs are already in the position required by the assumptions in the safety analysis.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is consistent with that allowed in Condition A.

For inoperable SGSVs that are closed, the inoperable SGSVs must be verified o"n a periodic basis to be closed. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering judgment, in view of SGSV status indications available in the control room, and other

administrative controls, bensure that these valves are in the closed position.

D.1 and D.2 If the SGSVs are not closed within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To*

achieve this status, the unit must be placed at least in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from MODE 2 conditions in an orderly manner and without challenging unit systems.

SR 3.7.2.1 This SR verifies that SGSV closure time is within the limit given in Reference 4 and is within that assumed in the accident analyses. The valve(s) may also be tested to more restrictive requirements in accordance with the llNSERVICE TESTING PROGRAMjlnservice Testing Program. This SR is normally performed upon returning the unit to operation following a refueling outage. The SGSVs should not be tested at power, since a unit trip could occur. As the SGSVs are not tested at power, they are exempt from the ASME OM Code (Ref. 5) requirements during operation in MODE 1 or 2.

Cook Nuclear Plant Unit 2 B 3.7.2-3 Revision No. 33

BASES SGSVs B 3.7.2 SURVEILLANCE REQUIREMENTS (continued)

REFERENCES The Frequency is in accordance with the llNSERVICE TESTING!

IPROGRAMilnservice Testing Program.

This test is conducted in MODE 3 with the unit at operating temperature and pressure. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

SR 3.7.2.2 This SR verifies that each SGSV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The Frequency of SGSV testing is every 24 months. The 24 month Frequency for testing is based on equipment reliability. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, this Frequ*ency is acceptable from a reliability standpoint.

1.

UFSAR, Section 10.2.

2.

UFSAR, Section 14.2.5.

3.

10 CFR 100.11.

4.

Technical Requirements Manual

5.

ASME, Operations and Maintenance Standards and Guides (OM Codes).

Cook Nuclear Plant Unit 2 B 3.7.2-4 Revision No. 33

BASES MFIVs and MFRVs B 3.7.3 ACTIONS (continued)

SURVEILLANCE REQUIREMENTS REFERENCES With both the MFIV and MFRV inoperable in the same flow path, there is no redundant system to operate automatically and perform the required safety function. Under these conditions, the affected flow path must be isolated within ~ hours. This action returns the system to the condition where at least one valve in each flow path is performing the required safety function. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable, based on operating experience, to complete the actions required to close the MFIV or MFRV, or otherwise isolate the affected flow path.

D.1 and D.2 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SR 3.7.3.1 and SR 3.7.3.2 These SRs verify that the closure time of each MFIV and MFRV is within the limit given in Reference 2 and is within that assumed in the accident and transient analyses. The valve(s) may also be tested to more restrictive requirements in accordance with the llNSERVICE TESTING!

IPROGRAMllnservice Testing Program.

The Frequency for this SR is in accordance with the llNSERVICEI lfESTING PROGRAMllnservice Testi!lg Program.

SR 3.7.3.3 This SR verifies that each MFIV and MFRV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following. a refueling outage.

The Frequency for this SR is every 24 months. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.

1.

UFSAR, Section 10.5.1.2.

2.

Technical Requirements Manual Cook Nuclear Plant Unit 2 B 3.7.3-4 Revision No. 33

BASES AFW System B 3.7.5 SURVEILLANCE REQUIREMENTS (continued) initiation without declaring the train(s) inoperable. Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown

. operations for steam generator level control, and these manual operations are an accepted function of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

SR 3.7.5.2 Verifying that each required AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded to an unacceptable level

during the cycle. Flow and differential head are normal tests of Gentrifugal pump performance required by the ASME OM Code (Ref. 2).

Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY and detect incipient failures by in_dicating abnormal performance. Performance of inservice testing a discussed in the ASME OM Code (Ref. 2) and the INSERVICE TESTING PROGRAM (only required at 3 month intervals) satisfies this requirement.

This SR is modified by a Note indicating that the SR should be deferred for the turbine driven AFW pump until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient pressu.re to perform the test.

SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.

This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 24 month Frequency is acceptable based on operating experience and the design reliability of the equipment.

Cook Nuclear Plant Unit 2 B 3.7.5-8 Revision No. O to AEP-NRC-2016-29 Description and Assessment of the Proposed Alternative to the ASME Code Request in Accordance with 10 CFR 50.55a(z)(2)

Alternative Due to Hardship Without a Compensating Increase in Quality and Safety

1.0 DESCRIPTION

The request is to adopt a proposed alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code by adoption of approved Code Case OMN-20, "lnservice Test Frequency."

2.0 ASSESSMENT

Technical Evaluation of the Proposed Alternative to the OM Code Section lnservice Testing (IST) of Division 1 of the OM Code, which is incorporated by reference in 10 CFR 50.55a(a), specifies component test frequencies based either on elapsed time periods (e.g., quarterly, 2 years) or on the occurrence of a plant condition or event (e.g., cold shutdown, refueling outage).

ASME Code Case OMN-20, "lnservice Test Frequency," has been approved for use by the ASME OM committee as an alternative to the test frequencies for pumps and valves specified in ASME OM Division: 1 Section IST 2009 Edition through OMa-2011 Addenda, and all earlier editions and addenda of ASME OM Code.

Code Case OMN-20 is not referenced in the latest revision of Regulatory Guide 1.192 (August 2014) as an acceptable OM Code Case to comply with 10 CFR 50.55a(f) requirements as allowed by 1 O CFR 50:55a(b)(6). The proposed alternative is to use Code Case OMN-20 to extend or. reduce the IST frequency requirements for the 5th 10 year IST interval or until OMN-20 is incorporated into the next revision of Regulatory Guide 1.192.

ASME Code Components Affected The Code Case applies to pumps and valves specified in ASME OM Division: 1 Section IST 2009 Edition through OMa-2011 Addenda and all earlier editions and addenda of ASME OM Code.

Frequency extensions may also be applied to accelerated test frequencie~. (e.g., pumps in Alert Range) as specified in OMN-20.

For pumps and valves with test periods of 2 years or less, the test frequency allowed by OMN-20 and the current Technical Specification (TS} IST Program (as modified by Surveillance Requirement (SR) 3.0.2 and Enforcement Guidance Memorandum (EGM) 2012-001) are the same.

For pumps and valves with test frequencies greater than 2 years, OMN-20 allows the test frequency to* be extended by 6 months.

The current TS IST Program does not allow extension of test frequencies that are greater than 2 years.

to AEP-NRC-2016-29 Page2

Applicable Code Edition and Addenda

ASME Code Case OMN-20 applies to ASME OM Division: 1 Section IST 2009 Edition through OMa-2011 Addenda and all earlier editions and addenda of ASME OM Code.

The Donald C. Cook Nuclear Plant (CNP) Code Edition and Addenda that are applicable to the program interval is ASME OM Code 2004 Edition through the 2006 Addenda. The CNP current interval ends June 30, 2026.

Applicable Code Requirement

This request is made in accordance with 10 CFR 50.55a(z)(2}, and proposes an alternative to the requirements of 10 CFR 50.55a{f}, which requires pumps and valves to meet the test requirements set forth in specific documents incorporated by reference in 10 CFR 50.55a(a). ASME Code Case OMN-20 applies to Division 1, Section IST of the ASME OM Code and associated addenda incorporated by reference in 1 O CFR 50.55a(a).

Reason for Request

The IST Program controls specified in Section 5.5.6 of TS provide: a) a table specifying certain IST frequencies; b) an allowance to apply SR 3.0.2 to inservice tests required by the OM Code and with frequencies of two years or less; c) an allowance to apply SR 3.0.3 to inservice tests required by the OM Code; and d) a statement that, "Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS." In Regulatory Issue Summary (RIS) 2012-10, "NRC Staff Position on Applying Surveillance Requirement 3.0.2 and 3.0.3 to Administrative Controls Program Tests," and Enforcement Guidance Memorandum (EGM) 2012-001, "Dispositioning Noncompliance with Administrative Controls Technical Specifications Programmatic Requirements that Extend Test Frequencies and Allow Performance of Missed Tests," the Nuclear Regulatory Commission (NRC) stated that items b, c, and d of the TS IST Program were inappropriately added to the TS and may not be applied (although the EGM allows licensees to continue to apply those paragraphs pending a generic resolution of the issue).

IN RIS 2012-10 and EGM 2012-001, the NRC stated that the current TS allowance to apply SR 3.0.2 and SR 3.0.3 to the IST Program would no longer be permitted. In response, OMN-20, which provides allowances similar to SR 3.0.2, was approved and is proposed to be used as an alternative to the test periods specified in the OM code. The proposed alternative substitutes an approved Code Case for the existing TS requirements that the NRC has determined are not legally acceptab,le as a TS allowance. This proposed alternative provides an equivalent level of safety as the existing TS allowance, while maintaining consistency with 1 O CFR 50.55a and the ASME OM Code.

Proposed Alternative and Basis for Use The proposed alternative is OMN-20, "lnservice Test Frequency," which addresses testing periods for pumps and valves specified in ASME OM Division 1, Section IST, 2009 Edition through OMa-2011 Addenda, and all earlier editions and addenda of ASME OM Code.

to AEP-NRC-2016-29 Page 3 This request is being made in accordance with 10 CFR 50.55a(z)(2), in that the existing requirements are considered a hardship without a compensating increase in quality and safety for the following reasons:

1)

For IST periods up to and including 2 years, Code Case OMN-20 provides an allowance to extend the IST testing periods by up to 25%.

The period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of ttie test would cause an unacceptable increase in the plant risk profile due to transient coriditions or other ongoing surveillance, test or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified. The test period extension and the statements regarding the appropriate use of the period extension are equivalent to the.existing TS SR 3.0.2 allowance and the statements regarding its use in the SR 3.0.2 Bases. Use of the SR 3.0.2 period extension has been a practice in the nuclear industry for many decades and elimination of this allowance would place a hardship on CNP when there is no evidence that the period.

extensions affect component reliability.

2)

For IST periods of greater than 2 years, OMN-20 allows an extension of up to 6 months.

The ASME OM Committee determined that such an extension is appropriate. The 6-month extension will have a minimal impact on.component reliability considering that the most probable result of performing any inservice test is satisfactory verification of the test acceptance criteria. As such, pumps and valves will continue to be adequately assessed for operational readiness when tested in accordance with the requirements specified in 10 CFR 50.55a(f) with the frequency extensions allowed by Code Case OMN-20.

3)

As stated in EGM 2012-001, if an lnservice Test is not performed within its frequency, SR 3.0.3 will not be applied. The effect of a missed lnservice Test on the Operability of TS equipment will be assessed under the licensee's Operability Determination Program.

Duration of Proposed Alternative The proposed alternative is. requested for the current 1 O year IST interval or until Code Case OMN-20 is incorporated into a future revision of Regulatory Guide 1.192, referenced by a future revision of 10 CFR 50.55a, whichever occurs first.

Precedents The NRC approved the use of OMN-20 for North Anna on March 27, 2014 (NRC Agencywide Documents Access and Management System, Accession Number ML14084A407).

v t e Letter Sequence Request
CAC:MF8152, (Open)
Initiation Request Acceptance... Administration Questions Answers Results Approval... Other: ML17096A627
MONTHYEAR2016-07-21 [Table View]

AEP-NRC-2016-29, Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3 and to Request an Alternative to the ASME Code

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Enclosures:

1.0 DESCRIPTION

2.0 ASSESSMENT

3.0 REGULATORY ANALYSIS

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3.4 REFERENCES

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2.0 ASSESSMENT

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Applicable Code Requirement

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AEP-NRC-2016-29, Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3 and to Request an Alternative to the ASME Code

Text

Enclosures:

1.0 DESCRIPTION

2.0 ASSESSMENT

3.0 REGULATORY ANALYSIS

3.4 REFERENCES

3.4 REFERENCES

1.0 DESCRIPTION

2.0 ASSESSMENT

Applicable Code Edition and Addenda

Applicable Code Requirement

Reason for Request

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