Text

Issuance of Amendment No. 258 Adoption of TSTF-425, Revision 3,
	ML17061A050
Person / Time
Site:	Cooper
Issue date:	03/31/2017
From:	Thomas Wengert; Plant Licensing Branch IV
To:	Higginbotham K; Nebraska Public Power District (NPPD)
	Wengert T, DORL/LPL-IV, 415-4037
References
	CAC MF7498
	Download: ML17061A050 (28)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 31, 2017 Mr. Kenneth Higginbotham Vice President-Nuclear and CNO Nebraska Public Power District 72676 648A Avenue Brownville, NE 68321

SUBJECT:

COOPER NUCLEAR STATION - ISSUANCE OF AMENDMENT RE:

ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE TRAVELER TSTF-425, REVISION 3 (CAC NO. MF7498)

Dear Mr. Higginbotham:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 258 to Renewed Facility Operating License No. DPR-46 for the Cooper Nuclear Station.

The amendment consists of changes to the technical specifications (TSs) in response to your application dated March 22, 2016, as supplemented by two letters dated December 7, 2016.

The amendment revises the TSs by relocating specific surveillance frequencies to a licensee-controlled program consistent with the NRG-approved Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-informed TSTF] Initiative 5b." The TSTF-425 revised pages also reflect issuance of two amendments issued before the review of this TSTF-425 amendment was completed.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

-rf,,~ *,~~~CJ ,*cf-*

Thomas J. Wengert, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Amendment No. 258 to DPR-46

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 258 Renewed License No. DPR-46

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Nebraska Public Power District (the licensee),

dated March 22, 2016, as supplemented by two letters dated December 7, 2016, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-46 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 258, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~ ({),)

Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-46 and the Technical Specifications Date of Issuance: March 31, 2017

ATTACHMENT TO LICENSE AMENDMENT NO. 258 COOPER NUCLEAR STATION RENEWED FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following page of the Renewed Facility Operating License No. DPR-46 with the enclosed revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Renewed Facility Operating License REMOVE INSERT Replace the following pages of Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Technical Specifications REMOVE INSERT iii iii 1.1-5 1.1-5 3.1-10 3.1-10 3.1-13 3.1-13 3.1-17 3.1-17 3.1-19 3.1-19 3.1-21 3.1-21 3.1-22 3.1-22 3.1-26 3.1-26 3.2-1 3.2-1 3.2-2 3.2-2 3.2-4 3.2-4 3.3-3 3.3-3 3.3-4 3.3-4 3.3-5 3.3-5 3.3-11 3.3-11 3.3-12 3.3-12 3.3-16 3.3-16 3.3-17 3.3-17 3.3-18 3.3-18 3.3-21 3.3-21 3.3-24 3.3-24 3.3-26 3.3-26

REMOVE INSERT 3.3-27 3.3-27 3.3-30 3.3-30 3.3-36 3.3-36 3.3-45 3.3-45 3.3-50 3.3-50 3.3-55 3.3-55 3.3-56 3.3-56 3.3-59 3.3-59 3.3-62 3.3-62 3.3-65 3.3-65 3.3-68 3.3-68 3.4-3 3.4-3 3.4-5 3.4-5 3.4-7 3.4-7 3.4-9 3.4-9 3.4-11 3.4-11 3.4-13 3.4-13 3.4-16 3.4-16 3.4-18 3.4-18 3.4-20 3.4-20 3.4-22 3.4-22 3.4-23 3.4-23 3.5-3 3.5-3 3.5-4 3.5-4 3.5-5 3.5-5 3.5-6 3.5-6 3.5-9 3.5-9 3.5-10 3.5-10 3.5-12 3.5-12 3.5-13 3.5-13 3.6-2 3.6-2 3.6-7 3.6-7 3.6-12 3.6-12 3.6-13 3.6-13 3.6-14 3.6-14 3.6-15 3.6-15 3.6-16 3.6-16 3.6-17 3.6-17 3.6-19 3.6-19 3.6-21 3.6-21 3.6-22 3.6-22 3.6-24 3.6-24 3.6-26 3.6-26 3.6-29 3.6-29 3.6-30 3.6-30 3.6-32 3.6-32 3.6-33 3.6-33 3.6-35 3.6-35 3.6-39 3.6-39

REMOVE INSERT 3.6-42 3.6-42 3.7-2 3.7-2 3.7-4 3.7-4 3.7-5 3.7-5 3.7-7 3.7-7 3.7-10 3.7-10 3.7-12 3.7-12 3.7-13 3.7-13 3.7-15 3.7-15 3.8-5 3.8-5 3.8-6 3.8-6 3.8-7 3.8-7 3.8-8 3.8-8 3.8-9 3.8-9 3.8-15 3.8-15 3.8-17 3.8-17 3.8-18 3.8-18 3.8-23 3.8-23 3.8-24 3.8-24 3.8-27 3.8-27 3.8-30 3.8-30 3.9-2 3.9-2 3.9-3 3.9-3 3.9-4 3.9-4 3.9-7 3.9-7 3.9-8 3.9-8 3.9-11 3.9-11 3.9-14 3.9-14 3.10-5 3.10-5 3.10-8 3.10-8 3.10-12 3.10-12 3.10-14 3.10-14 3.10-15 3.10-15 3.10-17 3.10-17 3.10-22 3.10-22 3.10-23 3.10-23 5.0-18 5.0-18 5.0-19 5.0-19 5.0-20 5.0-20 5.0-21 5.0-21 5.0-22 5.0-22 5.0-23 5.0-23 5.0-24

(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2419 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 258, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Cooper Nuclear Station Safeguards Plan," submitted by letter dated May 17, 2006.

NPPD shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The NPPD CSP was approved by License Amendment No. 238 as supplemented by changes approved by License Amendments 244 and 249.

(4) Fire Protection NPPD shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated April 24, 2012 (and supplements dated July 12, 2012, January 14, 2013, February 12, 2013, March 13, 2013, June 13, 2013, December 12, 2013, January 17, 2014, February 18, 2014, and April 11, 2014), and as approved in the safety evaluation dated April 29, 2014.

Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if Amendment No. 258

TABLE OF CONTENTS 3.8 ELECTRICAL POWER SYSTEMS .............................................................. 3.8-1 3.8.1 AC Sources - Operating ......................................................................... 3.8-1 3.8.2 AC Sources - Shutdown ......................................................................... 3.8-10 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air .............................................. 3.8-13 3.8.4 DC Sources - Operating ......................................................................... 3.8-16 3.8.5 DC Sources - Shutdown ......................................................................... 3.8-20 3.8.6 Battery Cell Parameters ......................................................................... 3.8-22 3.8.7 Distribution Systems - Operating ............................................................ 3.8-26 3.8.8 Distribution Systems - Shutdown ........................................................... 3.8-29 3.9 REFUELING OPERATIONS ........................................................................ 3.9-1 3.9.1 Refueling Equipment Interlocks .............................................................. 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock .................................................. 3.9-3 3.9.3 Control Rod Position .............................................................................. 3.9-4 3.9.4 Control Rod Position Indication .............................................................. 3.9-5 3.9.5 Control Rod OPERABILITY - Refueling ................................................. 3.9-7 3.9.6 Reactor Pressure Vessel (RPV) Water Level ......................................... 3.9-8 3.9.7 Residual Heat Removal (RHR) - High Water Level ................................ 3.9-9 3.9.8 Residual Heat Removal (RHR) - Low Water Level ................................. 3.9-12 3.10 SPECIAL OPERATIONS ............................................................................3.10.1 3.10.1 lnservice Leak and Hydrostatic Testing Operation ............................... 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing ................................................ 3.10-4 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ................................... 3.10-6 3.10.4 Single Control Rod Withdrawal - Cold Shutdown .................................. 3.10.9 3.10.5 Single Control Rod Drive (CRD)

Removal - Refueling ................................................................... 3.10-13 3.10.6 Multiple Control Rod Withdrawal - Refueling ........................................ 3.10-16 3.10.7 Control Rod Testing - Operating .......................................................... 3.10-18 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling ................................... 3.10-20 4.0 DESIGN FEATURES ................................................................................... 4.0-1 4.1 Site Location .......................................................................................... 4.0-1 4.2 Reactor Core ......................................................................................... 4.0-1 4.3 Fuel Storage .......................................................................................... 4.0-2 5.0 ADMINISTRATION CONTROLS .................................. :.............................. 5.0-1 5.1 Responsibility ......................................................................................... 5.0-1 5.2 Organization .......................................................................................... 5.0-2 5.3 Unit Staff Qualifications .......................................................................... 5.0-4 5.4 Procedures ............................................................................................ 5.0*5 5.5 Programs and Manuals .......................................................................... 5.0-6 5.6 Reporting Requirements ....................................................................... 5.0-20 5.7 High Radiation Area ............................................................................... 5.0-24 Cooper iii Amendment No. 258

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SOM) b. The moderator temperature is 2: 68°F, corresponding to (continued) the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME RESPONSE TIME consists of two components:

a. The time from initial movement of the main turbine stop valve or control valve until 80% of the turbine bypass capacity is established; and

b. The time from initial movement of the main turbine stop valve or control valve until initial movement of the turbine bypass valve.

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.

Cooper 1.1-5 Amendment No. 258

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. In accordance with the Surveillance Frequency Control Program SR 3.1.3.2 (Deleted)

SR 3.1.3.3 ----~~~-~--~~-----------NOl"E----~-~~~~--~--------~~

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one notch. In accordance with the Surveillance Frequency Control Program SR 3.1.3.4 Verify each control rod scram time from fully In accordance with withdrawn to notch position 06 is s 7 seconds. SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

Cooper 3.1-10 Amendment No. 258

Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify, for a representative sample, each tested In accordance with control rod scram time is within the limits of Table the Surveillance 3.1.4-1 with reactor steam dome pressure ~ 800 psig. Frequency Control Program SR 3.1.4.3 Verify each affected control rod scram time is within Prior to declaring the limits of Table 3.1.4-1 with any reactor steam control rod dome pressure. OPERABLE after work on control rod orCRD

System that could affect scram time SR 3.1.4.4 Verify each affected control rod scram time is within Prior to exceeding the limits of Table 3.1.4-1 with reactor steam dome 40% RTP after pressure <:: 800 psig. fuel movement within the affected core cell Prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time Cooper 3.1-13 Amendment No. 258

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control rod C.1 Verify the associated Immediately upon scram accumulators control rods are fully discovery of inoperable with reactor inserted. charging water steam dome pressure < 900 header pressure psig. AND < 940 psig C.2 Declare the associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months control rod inoperable.

D. Required Action 8.1 or C.1 D.1 ------~----~NOl"E~~~~-----

and associated Completion Not applicable if all l"ime not met. inoperable control rod scram accumulators are associated with fully inserted control rods.

Place the reactor mode Immediately switch in the shutdown

.position.

SURVEILLANCE REQUIREMENl"S SURVEILLANCE .FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator pressure In accordance with is ~ 940 psig. the Surveillance Frequency Control Program Cooper 3.1-17 Amendment No. 258

Rod Pattern Control 3.1.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Nine or more OPERABLE B.1 --------~-----NOTE---~---~-

control rods not in Rod worth minimizer compliance with BPWS. (RWM) may be bypassed as allowed by LCO 3.3.2.1.

Suspend withdrawal of Immediately control rods.

B.2 Place the reactor mode 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch in the shutdown position.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply with In accordance with BPWS. the Surveillance Frequency Control Program Cooper 3.1-19 Amendment No. 258

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium pentaborate In accordance with solution is within the limits of Figure 3.1.7-1. the Surveillance Frequency Control Program SR 3.1.7.2 Verify temperature of sodium pentaborate solution is In accordance with within the limits of Figure 3.1.7-2. the Surveillance Frequency Control Program SR 3.1.7.3 Verify temperature of pump suction piping is within In accordance with the limits of Figure 3.1.7-2. the Surveillance Frequency Control Program SR 3.1.7.4 Verify continuity of explosive charge. In accordance with the Surveillance Frequency Control Program SR 3.1.7.5 Verify the concentration of boron in solution is .within In accordance with the limits of Figure 3.1. 7-1. the Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or boron is added to solution (continued}

Cooper 3.1-21 Amendment No. 258

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY (continued) Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual valve in the flow In accordance with path that is not locked, sealed, or otherwise secured the Surveillance in position, is in the correct position or can be aligned Frequency Control to the correct position. Program SR 3.1.7.7 Verify each pump develops a flow rate ~ 38.2 gpm at In accordance a discharge pressure~ 1300 psig. with the In service Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem from pump In accordance with into reactor pressure vessel. the Surveillance Frequency Control Program SR 3.1.7.9 Verify all heat traced piping between storage tank In accordance with and pump suction is unblocked. the Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-2 Cooper 3.1-22 Amendment No. 258

SDV Vent and Drain Valves 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 ~~~--~----~---~~~---NC>TE----~----------~~--~~-~~

Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2.

Verify each SDV vent and drain valve is open. In accordance with the Surveillance Frequency Control Program SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully In accordance with closed and fully open position. the Surveillance Frequency Control Program SR 3.1.8.3 Verify each SDV vent and drain valve: In accordance with the Surveillance

a. Closes in s 30 seconds after receipt of an Frequency Control actual or simulated scram signal; and Program

b. ()pens when the actual or simulated scram signal is reset.

Cooper 3.1-26 Amendment No. 258

APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LCO 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER ~ 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits. limits.

B. Required Action and 8.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time < 25% RTP.

not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the Once within 12 limits specified in the COLR. hours after ~ 25%

RTP In accordance with the Surveillance Frequency Control Program Cooper 3.2-1 Amendment No. 258

MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY: THERMAL POWER 2 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within limits. A.1 Restore MCPR(s) to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits.

8. Required Action and 8.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion nme < 25% RTP.

not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within 12 limits specified in the COLR. hours after 2 25%

RTP In accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.2-2 Amendment No. 258

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER ~ 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within limits. A.1 Restore LHGR(s) to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits.

B. Required Action and B.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time < 25% RTP.

not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the limits Once within 12 specified in the COLR. hours after ~ 25%

RTP In accordance with the Surveillance Frequency Control Program Cooper 3.2-4 Amendment No. 258

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

--~-----~-~-~~----~~~~~~~~-----~~~~NC>TES--~-~--------~------~~--~--~-----~~

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR .3.3.1.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.2 ----~--~~~~---~-~~---NC>TE--~~~--~~---~-~-~---

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL PC>WER ~ 25% RTP.

Verify the absolute difference between the average In accordance with power range monitor (APRM) channels and the the Surveillance calculated power is s 2% RTP plus any gain Frequency Control adjustment required by LC() 3.4.1, "Recirculation Program Loops C>perating" while operating at;;;:: 25% RTP.

SR 3.3.1.1.3 --~--~~--~-~~--~----~~-NC>TE-~~~-------~--------~~--

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .after entering MC>DE 2.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.4 Perform a functional test of each RPS channel test In accordance with switch. the Surveillance Frequency Control Program (continued)

Cooper 3.3-3 Amendment No. 258

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels overlap. withdrawing SRMs from the fully inserted position SR 3.3.1.1.6 --------------------NOTE-------------------

Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.7 Adjust the channel to conform to a calibrated flow In accordance with signal. the Surveillance Frequency Control Program SR 3.3.1.1.8 Calibrate the local power range monitors. In accordance with the Surveillance Frequency Control Program SR 3. 3. 1. 1. 9 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.10 ---------------------------NOTES---------------------------

1. Neutron detectors and recirculation loop flow transmitters are excluded.

2. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.3-4 Amendment No. 258

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.12 -----------------------NOTES-----------------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.13 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.14 Verify Turbine Stop Valve - Closure and Turbine In accordance with Control Valve Fast Closure, Trip Oil Pressure - Low the Surveillance Functions are not bypassed when THERMAL Frequency Control POWER is ;;:: 29.5% RTP. Program SR 3.3.1.1.15 ------------------------------NOTE-----------------------------

Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits. In accordance with the Surveillance Frequency Control Program Cooper 3.3-5 Amendment No. 258

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS

--~-----------------------------------------1\J()TE-------------------~---------~-------------

Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable M()DE or other specified condition.

SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.2.2 --------------------------N()TES---------------------------

1. ()nly required to be met during C()RE ALTERATl()NS.

2. One SRM may be used to satisfy more than one of the following.

Verify an OPERABLE SRM detector is located in: In accordance with the Surveillance

a. The fueled region; Frequency Control Program

b. The core quadrant where C()RE ALTERATl()NS are being performed, when the associated SRM is included in the fueled region; and

c. A core quadrant adjacent to where CORE AL TERATIONS are being performed, when the associated SRM is included in the fueled region.

SR 3.3.1.2.3 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.3-11 Amendment No. 258

SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.2.4 -----------------NOTE----------------------*

Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.

Verify count rate is ~ 3.0 cps with a signal to noise 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during ratio<?: 2:1. CORE ALTERATIONS AND In accordance with the Surveillance Frequency Control Program SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST and In accordance with determination of signal to noise ratio. the Surveillance Frequency Control Program SR 3.3.1.2.6 -------------------------NOTE:----------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after IRMs on Range 2 or below.

Perform CHANNE:L FUNCTIONAL TEST and In accordance with determination of signal to noise ratio. the Surveillance Frequency Control Program SR 3.3.1.2. 7 -------------------------NOTES--------------------------

1. Neutron detectors are excluded.

2. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after IRMs on Range 2 or below.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Cooper 3.3-12 Amendment No. 258

Control Rod Block Instrumentation 3.3.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor Mode E.1 Suspend control rod Immediately Switch - Shutdown Position withdrawal.

channels inoperable.

E.2 Initiate action to fully insert all Immediately insertable control rods in core cells containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS

NOTES----------------------------------

1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.

2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains control rod block capability.

SURVEILLANCE FREQUENCY SR 3.3.2.1.1

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.3-16 Amendment No. 258

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.2 --------------------NOTE-----------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at s 9.85% RTP in MODE 2.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2. 1.3 -----------------NOTE--*---------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is s 9.85% RTP in MODE 1.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.4 --------------NOTE Neutron detectors are excluded.

Verify the RBM: In accordance with the Surveillance

a. Low Power Range - Upscale Function is not Frequency Control bypassed when THERMAL POWER is<!: 27.5% Program and < 62.5% RTP and a peripheral control rod is not selected.

b. Intermediate Power Range - Upscale Function is not bypassed when THERMAL POWER is

<!: 62.5% and < 82.5% RTP and a peripheral control rod is not selected.

c. High Power Range - Upscale Function is not bypassed when THERMAL POWER is~ 82.5%

RTP and a peripheral control rod is not selected.

(continued)

Cooper 3.3-17 Amendment No. 258

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.5 ----~--~~-~-----~--~N()TE-~--~------~-----~----------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATl()N. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.6 Verify the RWM is not bypassed when THERMAL In accordance with P()WER is~ 9.85% RTP. the Surveillance Frequency Control Program SR 3.3.2.1. 7 ------~--~-----~------~-----N()TE--------------~----~-----~~

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.8 Verify control rod sequences input to the RWM are in Prior to declaring conformance with BPWS. RWM ()PERABLE following loading of sequence into RWM Cooper 3.3-18 Amendment No. 258

Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 SURVEILLANCE REQUIREMENTS .

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

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided feedwater and main turbine high water level trip capability'is maintained.

SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The Allowable In accordance with Value shall be s 54.0 inches. the Surveillance Frequency Control Program SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST In accordance with including valve actuation. the Surveillance Frequency Control Program Cooper 3.3-21 Amendment No. 258

PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK on each required PAM In accordance with Instrumentation channel. the Surveillance Frequency Control Program SR 3.3.3.1.2 Perform CHANNEL CALIBRATION of the Primary In accordance with Containment Hz and Oz Analyzers. the Surveillance Frequency Control Program SR 3.3.3.1.3 Perform CHANNEL CALIBRATION of each required In accordance with PAM Instrumentation channel except for the Primary the Surveillance Containment Hz and Oz Analyzers. Frequency Control Program Cooper 3.3-24 Amendment No. 258

Alternate Shutdown System 3.3.3.2 3.3 INSTRUMENTATION 3.3.3.2 Alternate Shutdown System LCO 3.3.3.2 The Alternate Shutdown System Functions shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

-~--~~~~~~~---------~---~--~~----~~---N()"fE~~~~~~-~~-~.;._~~-~-~~----~-----~--~~~-

Separate Condition entry is allowed for each Function.

-M--~--------__.;..---~~~~~--------------------------~---------------------------~-----------------------------

CONDITION REQUIRED ACTION COMPLETION TIME A One or more required A.1 Restore required Function 30 days Functions inoperable. to ()PERABLE status.

8. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required In accordance with instrumentation channel that is normally energized. the Surveillance Frequency Control Program (continued)

Cooper 3.3-26 Amendment No. 258

Alternate Shutdown System 3.3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.3.2.2 Verify each required control circuit and transfer switch In accordance with is capable of performing the intended function. the Surveillance Frequency Control Program SR 3.3.3.2.3 Perform CHANNEL CALIBRATION for each required In accordance with instrumentation channel. the Surveillance Frequency Control Program Cooper 3.3-27 Amendment No. 258

ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS

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

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains ATWS-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program

SR 3.3.4.1.2 Perform CHANNEL CALIBRATION. The Allowable In accordance with Values shall be: the Surveillance Frequency Control

a. Reactor Vessel Water Level - Low Low Program (Level 2): ~ -42 inches; and

b. Reactor Pressure - High: s; 1072 psig.

SR 3.3.4.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST In accordance with including breaker actuation. the Surveillance .

Frequency Control Program Cooper 3.3-30 Amendment No. 258

ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 3.c and 3.f; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions other than 3.c and 3.f provided the associated Function or the redundant Function maintains ECCS initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.4 Perform CHANNEL CALIBRATl()N. In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-36 Amendment No. 258

RCIC System Instrumentation 3.3.5.2 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.5.2-1 to determine which SRs apply for each RCIC Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Function 2; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.2 Perform CHANNEL FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.4 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2.5 Perform L()GIC SYSTEM FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-45 Amendment No. 258

Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS

NOTES--------------------------------

1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.4 ----------------------------NOTE------------------------

For Function 2.d, radiation detectors are excluded.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.5 Calibrate each radiation detector. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-50 Amendment No. 258

Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 Declare associated 1 *hour secondary containment isplation valves inoperable.

AND C.2.1 Place the associated 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months standby gas treatment (SGT) subsystem(s} in operation.

OR C.2.2 Declare associated SGT 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months subsystem(s) inoperable.

SURVEILLANCE REQUIREMENTS

--~--~-----------~-----~-~------~~---~----------~--NOTES----~-----~----~-----~-------------~------~-~--~-

1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains secondary containment isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. lri accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.3-55 Amendment No. 258

Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMTNS (continued)

SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-56 Amendment No. 258

LLS Instrumentation 3.3.6.3 SURVEILLANCE REQUIREMENTS

~~-~-~~~~~-~----~-~-~~-~--~---~----~-l\l()TES~--~~--~-~~---~-~~-----~~~---~--~------

1. Refer to Table 3.3.6.3-1 to determine which SRs apply for each Function.

2. When a channel is placed in an inoperable status solely for performance*of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains LLS initiation c~pability.

SURVEILLANCE FREQUENCY SR 3.3.6.3.1 Perform CHANNEL FUNCTl()NAL TEST for portion In accordance with of the channel outside primary containment. the Surveillance Frequency Control Program SR 3.3.6.3.2 ~~--~---~---~--~~N()TE-----~-~-----~~-~-~-----

()nly required to be performed prior to entering M()DE 2 during each scheduled outage > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when entry is made into primary containment.

Perform CHANNEL FUNCTl()NAL.TEST for portions In accordance with of the channel inside primary containment. the Surveillance Frequency Control Program SR 3.3.6.3.3 Perform CHANNEL FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.3.4 Perform CHANNEL CALI BRATl()N. In accordance with the Surveillance

Frequency Control Program SR 3.3.6.3.5 Perform L()GIC SYSTEM FUNCTl()NAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-59 Amendment No. 258

CREF System Instrumentation 3.3.7.1 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.7.1-1 to determine which SRs apply for each CREF Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains CREF initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.7.1.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3. 7.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-62 Amendment No. 258

LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS

~-------------~~~-~---~~----------~--~----NC>TE~--~--~-~~-----~~-----~-~---~---~~--~---~----

1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided the associated Function maintains DG initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.2 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.3-65 Amendment No. 258

RPS Electric Power Monitoring 3.3.8.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to fully insert Immediately associated Completion Time all insertable control rods of Condition A or 8 not met in core cells containing in MODE 5 with any control one or more fuel rod withdrawn from a core assemblies.

cell containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.2.1 Perform CHANNEL CALIBRATION. The Allowable In accordance with Values shall be: the Surveillance Frequency Control

a. Overvoltage s 131 V with time delay set to Program s 3.8 seconds.

b. Undervoltage ~ 109 V, with time delay set to s 3.8 seconds.

c. Underfrequency ~ 57.2 Hz, with time delay set to s 3.8 seconds.

SR 3.3.8.2.2 Perform a system functional test. In accordance with the Surveillance Frequency Control Program Cooper 3.3-68 Amendment No. 258

Recirculation Loops Operating 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 ---~--------~~~-~--~~---NOT"E~-~-~--~-~--~--------~

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are in operation.

Verify recirculation loop flow mismatch with both In accordance with recirculation loops in operation is: the Surveillance Frequency Control

a. s 10% of rated core flow when operating at Program

< 70% of rated core flow; and

b. s 5% of rated core flow when operating at

~ 70% of rated core flow.

SR 3.4.1.2 Verify core flow as a function of T"HERMAL POWER In accordance with is not in the Stability Exclusion Region of the the Surveillance power/flow map specified in the COLR. Frequency Control Program Cooper 3.4-3 Amendment No. 258

Jet Pumps 3.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 --~~--~~~~--~------~--N()TE:S-~~~~~------~~~~----

1. Not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after associated recirculation loop is in operation.

2. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after> 25% RTP.

Verify at least one of the following criteria (a orb) is In accordance with satisfied for each operating recirculation loop: the Surveillance Frequency Control

a. Recirculation pump flow to speed ratio differs Program by s 5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by s 5% from established patterns.

b. Each jet pump diffuser to lower plenum differential pressure differs by s 20% from established patterns.

Cooper 3.4-5 Amendment No. 258

SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the SRVs In accordance with and SVs are as follows: the lnservice Testing Program Number of Setpoint SRVs (psig) 2 1080 +/- 32.4 3 1090 +/- 32.7 3 1100 +/- 33.0 Number of Setpoint SVs (psig) 3 1240 +/- 37.2 Following testing, lift settings shall be within +/- 1%.

SR 3.4.3.2 --~~-~-------~----~~--~N()TE-------~~-----------~--~-~-

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify each SRV opens when manually actuated. In accordance with the Surveillance Frequency Control Program Cooper 3.4-7 Amendment No. 258

RCS Operational LEAKAGE 3.4.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Verify source of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met. AND OR C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Pressure boundary LEAKAGE exists.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and In accordance with unidentified LEAKAGE increase are within limits. the Surveillance Frequency Control Program Cooper 3.4-9 Amendment No. 258

RCS leakage Detection Instrumentation 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time '

of Condition A or B not met. AND C.2 Be in MODE4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. All required leakage D.1 Enter LCO 3.0.3. Immediately detection systems inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required drywall In accordance with atmospheric monitoring channel. the Surveillance Frequency Control Program SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of In accordance with required leakage detection instrumentation. the Surveillance Frequency Control Program SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required In accordance with leakage detection instrumentation. the Surveillance Frequency Control Program Cooper 3.4-11 Amendment No. 258

RCS Specific Activity 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.2.2.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 ----~--------~-----~~~~~NOTE-~~~~--~----~----~--~-

Only required to be performed in MODE 1.

~--------------------------------:.---------

Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance with specific activity is s 0.2 µCi/gm. the Surveillance Frequency Control Program Cooper 3.4-13 Amendment No. 258

RHR Shutdown Cooling System - Hot Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 --------------------NOTE-----------------------

Not required to be met until 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor steam dome pressure is less than the shutdown cooling permissive pressure.

Verify one RHR shutdown cooling subsystem or In accordance with recirculation pump is operating. the Surveillance Frequency Control Program Cooper 3.4-16 Amendment No. 258

RHR Shutdown Cooling System - Cold Shutdown 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

8. No RHR shutdown cooling B.1 Verify reactor coolant 1 hourfrom subsystem in operation. circulating by an alternate discovery of no method. reactor AND coolant circulation No recirculation pump in AND operation.

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND B.2 Monitor reactor coolant Once per hour temperature.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR shutdown cooling subsystem or In accordance with recirculation pump is operating. the Surveillance Frequency Control Program Cooper 3.4-18 Amendment No. 258

RCS PIT Limits 3.4.9

-~_QT IONS {continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------------NOTE--------- C.1 Initiate action to restore Immediately Required Action C.2 shall parameter(s) to within be completed if this limits.

Condition is entered.

~---------------- AN[)

Requirements of the LCO C.2 [)etermine RCS is Prior to entering not met in other than acceptable for operation. MODE 2 or 3.

MODES 1, 2, and 3.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 -----------------------------NOTE----------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify: In accordance with the Surveillance

a. RCS pressure and RCS temperature are Frequency Control within the applicable limits specified in the Program curves in the PTLR; and

b. RCS heatup and cooldown rates are within limits specified in the PTLR.

(continued)

Cooper 3.4-20 Amendment No. 258

RCS PIT Limits 3.4.9 SURVEILLA,~N_C_E_R_E_Q_U_IR_E_:_uE_:_:_:_1L_(:_:_:_t:_~-e-d)_ _ _ _ _ _ _-,=r---~~~--~QUENCY SR 3.4.9.5 ----------------------NOTE---------------

Only required to be performed when tensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange In accordance with temperatures are within the limits specified in the the Surveillance PTLR. Frequency Control Program

~,--,,_, ___ _

SR 3.4.9.6 -------------------------------NOTE--------------------------------

Not required to be performed until 30 minutes after RCS temperature::;; 80°F in MODE 4.

Verify reactor vessel flange and head flange In accordance with temperatures are within the limits specified in the the Surveillance PTLR. Frequency Control Program

+-----------------

SR 3.4.9.7 ------------------------NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature s 90°F in MODE 4.

Verify reactor vessel flange and head flange In accordance with temperatures are within the limits specified in the the Surveillance PTLR. Frequency Control Program Cooper 3.4-22 Amendment No. 258

Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3 .4.1 O The reactor steam dome pressure shall be s 1020 psig.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit. dome pressure to within limit.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is s 1020 psig. In accordance with the Surveillance Frequency Control Program Cooper 3.4-23 Amendment No. 258

ECCS- Operating 3.5.1 ACTIONS {continued)

---~"

.. _,,,.,=~'

CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition C, D, E, or F AND not met.

G.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR dome pressure to $ 150 psig.

Two or more ADS valves inoperable.

H. Two or more low pressure H.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than condition A.

OR HPCI System and one or more ADS valves inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each EGGS injection/spray subsystem, the In accordance with piping is filled with water from the pump discharge the Surveillance valve to the injection valve. Frequency Control Program (continued)

Cooper 3.5-3 Amendment No. 258

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued}

SURVEILLANCE FREQUENCY SR 3.5.1.2 --------------NOTE--------.:----------

Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal _with reactor steam dome pressure less than the shutdown cooling permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

~----:"-----------------------------

Verify each ECCS injection/spray subsystem manual, In accordance with power operated, and automatic valve in the flow path, the Surveillance that is not locked, sealed, or otherwise secured in Frequency Control position, is in the correct position. Program SR 3.5.1.3 Verify ADS pneumatic supply header pressure is ~ 88 In accordance with psig. the Surveillance Frequency Control Program SR 3.5.1.4 Verify the RHR System cross tie shutoff valve is In accordance with closed. the Surveillance Frequency Control Program SR 3.5.1.5 --------------------NOTE-----------------------

Not required to _be performed if performed within the previous 31 days.

Verify each recirculation pump discharge valve cycles Once each startup through one complete cycle of full travel or is de- prior to exceeding energized in the closed position. 25% RTP (continued)

Cooper 3.5-4 Amendment No. 258

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.6 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the lnservice corresponding to the specified reactor pressure. Testing Program SYSTEM HEAD NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray 2: 4720 gpm 1 2: 113 psig LPCI 2: 15,000 gpm 2 2: 20 psig SR 3.5.1.7 ----------------NOTE-----------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow* are adequate to perform th~ test.

Verify, with reactor pressure s 1020 and 2: 920 psig, In accordance with the HPCI pump can develop a flow rate 2: 4250 gpm the Surveillance against a system head corresponding to reactor Frequency Control pressure. Program SR 3.5.1.8 -----------..:---------NOTE----------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressures 165 psig, the HPCI In accordance with pump can develop a flow rate 2: 4250 gpm against a the Surveillance system head corresponding to reactor pressure. Frequency Control Program (continued)

Cooper 3.5-5 Amendment No. 258

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 ----------~-----------NO"fES---------------~-------

1. For HPCI only, not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

2. Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem actuates In accordance with on an actual or simulated automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.10 ------------------~-------NOl"E------------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or simulated In accordance with automatic initiation signal. the Surveillance Frequency Control Program SR 3.5.1.11 ----------------------------NO"fE-----------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify each ADS valve opens when manually In accordance with actuated.

the Surveillance Frequency Control Program Cooper 3.5-6 Amendment No. 258

ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required ECCS injection/spray In accordance with subsystem, the suppression pool water level is~ 12 ft the Surveillance 7 inches. Frequency Control Program SR 3.5.2.2 Verify, for each required ECCS injection/spray In accordance with subsystem, the piping is filled with water from the the Surveillance pump discharge valve to the injection valve. Frequency Control Program SR 3.5.2.3 -~------------~~-l'J()l"E---------~-----------

()ne LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

Verify each required ECCS injection/spray subsystem In accordance with manual, power operated, and automatic valve in the the Surveillance flow path, that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position. Program (continued)

Cooper 3.5-9 Amendment No. 258

ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each required ECCS pump develops the In accordance specified flow rate against a system head with the corresponding to the specified reactor pressure. lnservice Testing SYSTEM HEAD Program NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF cs ~4720 gpm 1 ~ 113 psig LPCI ~ 7700 gpm 1 ~ 20 psig SR 3.5.2.5 -------------------~-~-NOTE-----------------~----

Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray subsystem In accordance with actuates on an actual or simulated automatic initiation the Surveillance signal. Frequency Control Program Cooper 3.5-10 Amendment No. 258

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with water In accordance with from the pump discharge valve to the injection valve. the Surveillance Frequency Control Program SR 3.5.3.2 Verify each RCIC System manual, power operated, In accordance with and automatic valve in the flow path, that is not the Surveillance locked, sealed, or otherwise secured in position, is in Frequency Control the correct position. Program SR 3.5.3.3 --~~~--~---~~~-~~~-N()l"E:--~----~-~--~~---~~--

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1020 psig and ~ 920 In accordance with psig; the RCIC pump can develop a flow rate <:: 400 the Surveillance gpm against a system head corresponding to reactor Frequency Control pressure. Program SR 3.5.3.4 ~~--~~------~~--~--~NC>TE:--~--~-----~-~~-~-~--

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressures 165 psig, the RCIC In accordance with pump can develop a flow rate ~ 400 gpm the Surveillance against a system head corresponding to reactor Frequency Control pressure. Program (continued)

Cooper 3.5-12 Amendment No. 258

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.3.5 ------~----~~~~----~--NCJTES-~-----~--~---~-~-~----

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test

2. Vessel injection may be excluded.

Verify the RCIC System actuates on an actual or In accordance with simulated automatic initiation signal. the Surveillance Frequency Control Program Cooper 3.5-13 Amendment No. 258

Primary Containment 3.6.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and leakage In accordance with rate testing except for primary containment air lock the Primary testing, in accordance with the Primary Containment Containment Leakage Rate Testing Program. Leakage Rate Testing Program SR 3.6.1.1.2 Verify drywell to suppression chamber bypass In accordance with leakage is equivalent to a hole< 1.0 inch in diameter. the Surveillance Frequency Control Program AND

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

()nly required after two consecutive tests fail and continues until two consecutive tests pass 9 months Cooper 3.6-2 Amendment No. 258

Primary Containment Air Lock 3.6.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1 --~-~~--~----~--~---~-N()l"ES---~---~-~-~~~~~---~--

1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.1.

Perform required primary containment air lock In accordance with leakage rate testing in accordance with the the Primary Primary Containment Leakage Rate Testing Containment Program. Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary containment air In accordance with lock can be opened at a time. the Surveillance Frequency Control Program Cooper 3.6-7 Amendment No. 258

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 ---------.------NQTES------------------

1. Not required to be met when the 24 inch primary containment purge and vent valves are open in one supply line and one exhaust line for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.

2. When the purging or venting in accordance with Note 1 is through the Standby Gas Treatment (SGT) System, both SGT subsystems shall be OPERABLE, and only one SGT subsystem shall be operating.

Verify each 24 inch primary containment purge and In accordance with vent valve is closed. the Surveillance Frequency Control

.Program SR 3.6.1.3.2 ----------------------NOTES---------'-------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls. '

Verify each primary containment isolation manual In accordance with valve and blind flange that is located outside primary the Surveillance containment and not locked, sealed, or otherwise Frequency Control secured and is required to be closed during accident Program conditions is closed.

(continued)

Cooper 3.6-12 Amendment No. 258

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.3 -~~~~~--~~~--------~N()TES-~--~---------~~-~~-

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment manual isolation Prior to entering valve and blind flange that is located inside MODE 2 or 3 from primary containment and not locked, sealed, or MODE 4 if primar}'

otherwise secured and is required to be closed containment was during accident conditions is closed. de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing incore probe (TIP) In accordance with shear isolation valve explosive charge. the Surveillance Frequency Control Program SR 3.6. 1.3.5 Verify the isolation time of each power operated, In accordance

automatic PCIV, except for MSIVs, is within limits. with the lnservice

Testing Program (continued)

Cooper 3.6-13 Amendment No. 258

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is ~ 3 seconds In accordance and ::::; 5 seconds. with the lnservice Testing Program SR 3.6.1.3. 7 Verify each automatic PCIV actuates to the isolation In accordance with position on an actual or simulated isolation signal. the Surveillance Frequency Control Program SR 3.6.1.3.8 Verify a representative sample of reactor In accordance with instrumentation line EFCVs actuate to the isolation the Surveillance position on an actual or simulated instrument line Frequency Control break. Program SR 3.6.1.3.9 Remove and test the explosive squib from each In accordance with shear isolation valve of the TIP System. the Surveillance Frequency Control Program SR 3.6.1.3.10 Verify leakage rate through each Main Steam line is In accordance with

106 scfh when tested at~ 29 psig. the Primary Containment Leakage Rate Testing Program (continued)

Cooper 3.6-14 Amendment No. 258

PC IVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.3.11 Verify each inboard 24 inch primary containment In accordance with purge and vent valve is blocked to restrict the the Surveillance maximum valve opening angle to 60°. - Frequency Control Program SR 3.6.1.3.12 Verify leakage rate through the Main Steam Pathway In accordance with is s 212 scfh when tested at ~ 29 psig. the Primary Containment Leakage Rate Testing Program Cooper 3.6-15 Amendment No. 258

Drywell Pressure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure LCO 3.6.1.4 Drywell pressure shall be s 0. 75 psig.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not within A.1 Restore drywell pressure 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months limit. to within limit.

B. Required Action and B.1 Be in MODE3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify drywell pressure is within limit. In accordance with the Surveillance Frequency Control Program Cooper 3.6-16 Amendment No. 258

Drywell Air Temperature 3.6.1.5 3.6 CONTAINMENT SYSTEMS 3.6.1.5 Drywell Air Temperature LCO 3.6.1.5 Drywell average air temperature shall be :s; 150°F.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air A:1 Restore drywell average 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within limit. air temperature to within limit.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify drywell average air temperature is within limit. In accordance with the Surveillance Frequency Control Program Cooper 3.6-17 Amendment No. 258

LLS Valves 3.6.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 -------------------------NOTE-----------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify each LLS valve opens when manually In accordance with actuated. the Surveillance Frequency Control Program SR 3.6.1.6.2 ---------------------------NOTE---------------------------

Valve actuation may be excluded.

Verify the LLS System actuates on an actual or In accordance with simulated automatic initiation signal. the Surveillance Frequency Control Program

---*---w~-------------~---"----=---

Cooper 3.6-19 Amendment No. 258

Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two lines with one or more D.1 Restore all vacuum 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months reactor building-to- breakers in one line to suppression chamber OPERABLE status.

vacuum breakers inoperable for opening.

E. Required Action and E.1 Be in MODE3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Associated Completion Time not met. AND E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.7.1 -~-~---~~-~~~----~~~NOTES-~----~--~-----~~~---

1. Not required to be met for vacuum breakers that are open during Surveillances.

2. Not required to be met for vacuum breakers open when performing their intended function.

Verify each vacuum breaker is closed. In accordance with the Surveillance Frequency Control Program SR 3.6.1.7.2 Perform a functional test of each vacuum breaker. In accordance with the Surveillance Frequency Control Program (continued)

Cooper 3.6-21 Amendment No. 258

Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.6.1.7.3 Verify the full open setpoint of each vacuum In accordance with breaker is s 0.5 psid. the Surveillance Frequency Control Program Cooper 3.6-22 Amendment No. 258

Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.8.1 --------------~---~----------N()l"E-------~-----------~--------

Not required to be met for vacuum breakers that are open during Surveillances.

Verify each vacuum breaker is closed. In accordance with the Surveillance Frequency Control Program SR 3.6.1.8.2 Perform a functional test of each required vacuum In accordance with breaker. the Surveillance Frequency Control Program SR 3.6.1.8.3 Verify the opening setpoint of each required vacuum In accordance with breaker is s 0.5 psid. the Surveillance Frequency Control Program Cooper 3.6-24 Amendment No. 258

RHR Containment Spray 3.6.1.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.9.1 Verify each RHR containment spray subsystem In accordance with manual, power operated, and automatic valve in the the Surveillance flow path that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position or can be Program aligned to the correct position.

SR 3.6.1.9.2 Verify each required RHR pump develops a flow rate In accordance with of> 7700 gpm through the associated heat the lnservice exchanger while operating in the suppression pool Testing Program cooling mode.

SR 3.6.1.9.3 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage Cooper 3.6-26 Amendment No. 258

Suppression Pool Average Temperature 3.6.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Suppression pool average E.1 Depressurize the reactor 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months temperature> 120°F. vessel to< 200 psig.

AND E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature is within In accordance with the applicable limits. the Surveillance Frequency Control Program 5 minutes when performing testing that adds heat to the suppression pool 3.6-29 Amendment No. 258

Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be ~ 12 ft 7 inches and s 12 ft 11 inches.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression pool 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months level not within limits. water level to within limits.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion lime not met. AND B.2 Be in MODE4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.2.1 Verify suppression pool water level is within limits. In accordance with the Surveillance Frequency Control Program Cooper 3.6-30 Amendment No. 258

RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling subsystem In accordance with manual, power operated, and automatic valve in the the Surveillance flow path that is not locked, sealed, or otherwise Frequency Control secured in position, is in the correct position or can Program be aligned to the correct position.

SR 3.6.2.3.2 Verify each RHR pump develops a flow rate > 7700 In accordance gpm through the associated heat exchanger while with the lnservice operating in the suppression pool cooling mode. Testing Program Cooper 3.6-32 Amendment No. 258

Primary Containment Oxygen Concentration 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Oxygen Concentration LCO 3.6.3.1 The primary containment oxygen concentration shall be < 4.0 volume percent.

APPLICABILITY: MODE 1 during the time period:

a. From 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is > 15% RTP following startup, to

b. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing THERMAL POWER to < 15% RTP prior to a reactor shutdown.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A Primary containment oxygen A.1 Restore oxygen 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months concentration not within concentration to within limit. limit.

B. Required Action and 8.1 Reduce THERMAL 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months associated Completion Time POWER to ::; 15% RTP.

not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1.1 Verify primary containment oxygen concentration is In accordance with within limits. the Surveillance Frequency Control Program Cooper 3.6-33 Amendment No. 258

Secondary Containment 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Initiate action to suspend Immediately OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 2: 0.25 inch In accordance with of vacuum water gauge. the Surveillance Frequency Control Program SR 3.6.4.1.2 Verify all secondary containment equipment hatches In accordance with are closed and sealed. the Surveillance Frequency Control Program

SR 3.6.4.1.3 Verify one secondary containment access door in In accordance with each access opening is closed. the Surveillance Frequency Control Program SR 3.6.4.1.4 Verify each SGT subsystem can maintain 2: 0.25 inch In accordance with of vacuum water gauge in the secondary containment the Surveillance for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at a flow rate :s: 1780 cfm. Frequency Control Program Cooper 3.6-35 Amendment No. 258

SCI Vs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 ------~-----~~~~--------N()TES---~------~-~--~~----~

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment isolation manual In accordance with valve and blind flange that is not locked, sealed, or the Surveillance otherwise secured and is required to be closed during Frequency Control accident conditions is closed. Program SR 3.6.4.2.2 Verify the isolation time of each power operated In accordance automatic SCIV is within limits. with the lnservice Testing Program SR 3.6.4.2.3 Verify each automatic SCIV actuates to the isolation In accordance with position on an actual or simulated actuation signal. the Surveillance Frequency Control Program Cooper 3.6-39 Amendment No. 258

SGT System 3.6.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) E.2 Initiate action to suspend Immediately OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for<::: 10 continuous In accordance with hours with heaters operating. the Surveillance Frequency Control Program SR 3.6.4.3.2 Perform required SGT filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or In accordance with simulated initiation signal. . the Surveillance Frequency Control Program SR 3.6.4.3.4 . Verify the SGT units eras~ tie damper is in the correct In accordance with position, and each SGT room air supply check valve the Surveillance and SGT dilution air shutoff valve can be opened. Frequ~ncy Control Program Cooper 3.6-42 Amendment No. 258

RHRSWB System 3.7.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met. AND OR C.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Both RHRSWB subsystems inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify each RHRSWB manual and power operated In accordance with valve in the flow path, that is not locked, sealed, or the Surveillance otherwise secured in position, is in the correct Frequency CoAtrol position or can be aligned to the correct position. Program Cooper 3.7-2 Amendment No. 258

SW System and UHS 3.7.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met. AND OR B.2 Be in MODE4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Both SW subsystems inoperable.

UHS inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the river water level is ~ 865 ft mean sea level. In accordance with the Surveillance Frequency Control Program SR 3.7.2.2 Verify the average water temperature of UHS is In accordance with s 95°F. the Surveillance Frequency Control Program (continued)

Cooper 3.7-4 Amendment No. 258

SW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.2.3 ~~--~--~~~---~~--~---f\J()TE-~---~-------~-~~---~~

Isolation of flow to individual components does not render SW System inoperable.

Verify each SW subsystem manual, power operated, In accordance with and automatic valve in the flow paths servicing safety the Surveillance related systems or components, that is not locked, Frequency Control sealed, or otherwise secured in position, is in the Program correct position.

SR 3.7.2.4 Verify each SW subsystem actuates on an actual or In accordance with simulated initiation signal. the Surveillance Frequency Control Program Cooper 3.7-5 Amendment f\Jo. 258

REC System 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 ~~--~~--~---~--~~-~-N()TES---~----~----~----~------~

1. SR 3.0.1 is not applicable when both Service Water backup subsystems are OPERABLE.

2. REC system leakage beyond limits by itself is only a degradation of the REC system and does not result in the REC system being inoperable.

Verify the REC system leakage is within limits. In accordance with the Surveillance Frequency Control Program SR 3.7.3.2 Verify the temperature of the REC supply water is In accordance with s 100°F. the Surveillance Frequency Control Program SR 3.7.3.3 ----~-----------~--~--------N()l"E-----~--------~-------~~-~-

lsolation of flow to individual components does not render REC System inoperable.

Verify each REC subsystem manual, power operated, In accordance with and automatic valve in the flow paths servicing safety the Surveillance related cooling loads, that is not locked, sealed, or Frequency Control otherwise secured in position, is in the correct Program position.

SR 3.7.3.4 Verify each REC subsystem actuates on an actual or In accordance with simulated initiation signal. the Surveillance Frequency Control Program Cooper 3.7-7 Amendment No. 258

CREF System 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREF System for <:: 15 minutes. In accordance with the Surveillance Frequency Control Program SR 3.7.4.2 Perform required CREF filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP.

SR 3.7.4.3 Verify the CREF System actuates on an actual or In accordance with simulated initiation signal. the Surveillance Frequency Control Program SR 3.7.4.4 Perform required CRE unfiltered air inleakage testing In accordance with in accordance with the Control Room Envelope the Control Room Habitability Program. Envelope Habitability Program Cooper. 3.7-10 Amendment No. 258

Air Ejector Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 ---------------------------NOTE-------------------------------

Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.

Verify the gross gamma activity rate of the noble In accordance with gases is s 1.0 Ci/second. the Surveillance Frequency Control Program Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a

~ 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level.

Cooper 3.7-12 Amendment No. 258

Spent Fuel Storage Pool Water Level 3.7.6

3. 7 PLANT SYSTEMS

3. 7.6 Spent Fuel Storage Pool Water Level LCO 3. 7.6 The spent fuel storage pool water level shall be ~ 21 ft 6 inches over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage pool A.1 ----~--~---NOl"E----~~--~-

water level not within limit. LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the spent fuel storage pool water level is <:: 21 ft In accordance with 6 inches over the top of irradiated fuel assemblies the Surveillance seated in the spent fuel storage pool racks. Frequency Control Program Cooper 3.7-13 Amendment No. 258

Main Turbine Bypass System 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify operation of each main turbine bypass valve. In accordance with the Surveillance Frequency Control Program SR 3.7.7.2 Perform a system functional test. In accordance with the Surveillance Frequency Control Program SR 3.7.7.3 Verify the TURBINE BYPASS SYSTEM RESPONSE In accordance with TIME is within limits. the Surveillance Frequency Control Program Cooper 3.7-15 Amendment No. 258

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power In accordance with availability for each offsite circuit. the Surveillance Frequency Control Program SR 3.8.1.2 -~--~-~------~-~--N()TE:S-~---~-------------------

1. Performance of SR 3.8.1. 7 satisfies this SR.

2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.

3. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1. 7 must be met.

Verify each DG starts from standby conditions and In accordance with achieves steady state voltage<:!: 3950 Vand s 4400 V the Surveillance and frequency <:!: 58.8 Hz and s 61.2 Hz. Frequency Control Program (continued)

Cooper 3.8-5 Amendment No. 258

AC Sources - Operating 3.8.1 SURVEILL~]'-J9!; 8!=_9U l8J::MENI$Jgi;>~!!_n_u,,ed}

SURVEILLANCE FREQUENCY SR 3.8.1.3 ---~~~---------~~~~---N()l"ES-~--~-~---~-~-~--~-

1. DG loadings may include gradual loading as recommended by the manufacturer.

2. Momentary transients outside the load range do not invalidate this test.

3. This Surveillance shall be conducted on only one DG at a time.

4. This SR shall be preceded by and immediately follow, without shutdown, a successful performance of SR 3.8.1.2 or SR 3.8.1.7.

Verify each DG is synchronized and loaded and In accordance with operates for 2:: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at a load <::: 3600 kW and the Surveillance s 4000 kW. Frequency Control Program SR 3.8.1.4 Verify each day tank contains <::: 1500 gal of fuel oil. In accordance with the Surveillance Frequency Control Program SR 3.8.1.5 Check for and remove accumulated water from each In accordance with day tank. the Surveillance Frequency Control Program SR 3.8.1.6 Verify the fuel oil transfer system operates to In accordance with automatically transfer fuel oil from storage tanks to the Surveillance the day tanks. Frequency Control Program

*-~'-"*-*"~*------------------~-------

(continued)

Cooper 3.8-6 Amendment No. 258

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.7 -------------------NOTE-----------------

All DG starts may be preceded by an engine prelube period.

Verify each DG starts from standby condition and In accordance with achieves, in s 14 seconds, voltage <!: 3950 V and the Surveillance frequency <!: 58.8 Hz, and after steady state Frequency Control conditions are reached, maintains voltage <!: 3950 V Program and s 4400 V and frequency <!: 58.8 Hz and s 61.2 Hz.

SR 3.8.1.8 -----------------NOTE---------------------

This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR.

Verify automatic and manual transfer of unit power In accordance with supply from the normal offsite circuit to the alternate the Surveillance offsite circuit. Frequency Control Program (continued)

Cooper 3.8-7 Amendment No. 258

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.9 -------------------------------NOTES------------------

1. Momentary transients outside the load and power factor ranges do not invalidate this test.

2. This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR.

3. If performed with DG synchronized with offsite power, the surveillance shall be performed at a power factors 0.89. However, if grid conditions do not permit, the power factor limit is not required to be met. Under this condition the power factor shall be maintained as close to the limit as practicable.

Verify each DG operates for~ 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s: In accordance with the Surveillance

a. For ~ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded*~ 4200 kW and s 4400 Frequency Control kW; and Program

b. For the remaining hours of the test loaded

~ 3600 kW ands 4000 kW.

+---~~---------

SR 3.8.1.10 -------------------~------NOTES--------~---------------

Th is Surveillance shall not be performed in MODE 1, 2 or 3. However, credit may be taken for unplanned events that satisfy this SR.

Verify interval between each sequenced load is within In accordance with

+/- 10% of nominal timer setpoint. the Surveillance Frequency Control Program (continued)

Cooper 3.8-8 Amendment No. 258

AC Sources - Operating 3.8.1

.§l:J~~~!l::~~NC:E REQUIREMENTS {continued)

SURVEILLANCE FREQUENCY SR 3.8.1.11 ------------------------NOTES------------------------------

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.

Verify, on an actual or simulated loss of offsite power

In accordance with signal in conjunction with an actual or simulated the Surveillance ECCS initiation signal: Frequency Control Program

a. De-energization of emergency buses;

b. Load shedding from emergency buses; and

c. DG auto-starts from standby condition and:

1. energizes permanently connected loads in s 14 seconds,

2. energizes auto-connected emergency loads through the timed logic sequence,

3. maintains steady state voltage 2: 3950 V ands 4400 V,

4. maintains steady state frequency 2: 58.8 Hz ands 61.2 Hz, and

5. supplies permanently connected and auto-connected emergency loads for 2: 5 minutes.

Cooper 3.8-9 Amendment No. 258

Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify the fuel oil storage tanks contain a combined In accordance with volume of;:::: 49,500 gal of fuel. the Surveillance Frequency Control Program SR 3.8.3.2 Verify lube oil inventory is ;:::: 504 gal. In accordance with the Surveillance Frequency Control Program SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are In accordance with tested in accordance with, and maintained within the the Diesel Fuel Oil limits of, the Diesel Fuel Oil Testing Program. Testing Program SR 3.8.3.4 Verify each DG has a minimum of one air start In accordance with receiver with a pressure 2: 200 psig. the Surveillance Frequency Control Program SR 3.8.3.5 Check for and remove accumulated water from each In accordance with fuel oil storage tank. the Surveillance Frequency Control Program Cooper 3.8-15 Amendment No. 258

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage on float charge is: In accordance with the Surveillance

a. ;?; 125.9 V for the 125 V batteries; and Frequency Control Program

b. ~ 260.4 V for the 250 V batteries.

SR 3.8.4.2 Verify no visible corrosion at battery terminals and In accordance with connectors. the Surveillance Frequency Control OR Program Verify battery connection resistance meets the limits specified in Table 3.8.4-1.

SR 3.8.4.3 Verify battery cells, cell plates, and racks show no In accordance with visual indication of physical damage or abnormal the Surveillance deterioration that degrades battery performance. Frequency Control Program SR 3.8.4.4 Remove visible corrosion and verify battery cell to cell In accordance with and terminal connections are coated with anti- the Surveillance corrosion material. Frequency Control Program SR 3.8.4.5 Verify battery connection resistance meets the limits In accordance with specified in Table 3.8.4-1. the Surveillance Frequency Control Program SR 3.8.4.6 Verify: In accordance with the Surveillance

a. Each required 125 V battery charger supplies Frequency Control 2: 200 amps at 2: 125 V for ~ 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and Program

b. Each required 250 V battery charger supplies

~ 200 amps at;?; 250 V for~ 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

(continued)

Cooper 3.8-17 Amendment No. 258

DC Sources - Operating 3.8.4

__SURVEILLAN~E REQUIREMENTS (continued)

___ -------*----- SURVEILLANCE ----~----i---~~~UENCY SR 3.8.4. 7 -------------------------------NOTES----------------------

1. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4. 7.

2. This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.

Verify battery capacity is adequate to supply, and In accordance with maintain in OPERABLE status, the required the Surveillance emergency loads for the design duty cycle when Frequency Control subjected to a battery service test. Program SR 3.8.4.8 -----------------------------NOTE--------------------------------

This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.

Verify battery capacity is ~ 90% of the manufacturer's 1 In accordance with rating when subjected to a performance discharge I the Surveillance test or a modified performance discharge test. 1 Frequency Control Program AND 12 months when battery shows degradation or has reached 85% of expected life with capacity < 100%

of manufacturer's rating AND 24 months when battery has reached 85% of the expected life with capacity

~ 100% of manufacturer's rating Cooper 3.8-18 Amendment No. 258

Battery Cell Parameters 3.8.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.

B. Required Action and B.1 Declare associated battery Immediately associated Completion Time inoperable.

of Condition A not met.

OR One or more batteries with average electrolyte temperature of the representative cells not within limits.

OR One or more batteries with one or more battery cell parameters not within Category C limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet Table 3.8.6-1 In accordance with Category A limits. the Surveillance Frequency Control Program (continued)

Cooper 3.8-23 Amendment No. 258

Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 In accordance with Category B limits. the Surveillance Frequency Control Program Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after battery discharge < 105 V for a 125 V battery or < 210 V for a 250 V battery AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after battery overcharge

> 140 Vfor a 125 V battery or

> 280 Vfora 250 V battery SR 3.8.6.3 Verify average electrolyte temperature of In accordance with representative cells is 2! 70°F. the Surveillance Frequency Control Program Cooper 3.8-24 Amendment No. 258

Distribution Systems - Operating 3.8.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

c. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met. AND C.2 Be in MODE4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. One or more 250 V DC D.1 Declare associated Immediately electrical power distribution supported feature( s) subsystems inoperable. inoperable.

E. Two or more electrical power E.1 Enter LCO 3.0.3. Immediately distribution subsystems inoperable that result in a loss offunction.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct breaker alignments and voltage to In accordance with required AC and DC, electrical power distribution the Surveillance subsystems. Frequency Control Program Cooper 3.8-27 Amendment No. 258

Distribution Systems - Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate actions to restore Immediately required AC and DC electrical power distribution subsystems to OPERABLE status.

A.2.5 Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and voltage to In accordance with required AC and DC electrical power distribution the Surveillance subsystems. Frequency Control Program Cooper 3.8-30 Amendment No. 258

Refueling Equipment Interlocks 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of In accordance with the following required refueling equipment interlock the Surveillance inputs: Frequency Control Program

a. All-rods-in,

b. Refuel platform position,

c. Refuel platform fuel grapple, fuel loaded,

d. Refuel platform fuel grapple not full up,

e. Refuel platform frame mounted hoist, fuel loaded,

f. Refuel platform monorail mounted hoist, fuel loaded, and

g. Service platform hoist, fuel loaded.

\

Cooper 3.9-2 Amendment No. 258

Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.

APPLICABILITY: MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one-rod-out A.1 Suspend control rod Immediately interlock inoperable. withdrawal..

A.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in refuel position. In accordance with the Surveillance Frequency Control Program SR 3.9.2.2 -----------------------------NOTE-----------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Cooper 3.9-3 Amendment No. 258

Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.

APPLICABILITY: When loading fuel assemblies into the core.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control rods A.1 Suspend loading fuel Immediately not fully inserted. assemblies into the core.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted. In accordance with the Surveillance Frequency Control Program Cooper 3.9-4 Amendment No. 258

Control Rod OPERABILITY - Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY - Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.

APPLICABILITY: MODE 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to fully insert Immediately control rods inoperable. inoperable withdrawn control rods.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 --~----------~-----------~--NOTE---~--~---------~----~---~-

Not required to be performed until 7 days after the control rod is withdrawn.

Insert each withdrawn control rod at least one notch. In accordance with the Surveillance Frequency Control Program SR 3.9.5.2 Verify each withdrawn control rod scram accumulator In accordance with pressure is~ 940 psig. the Surveillance Frequency Control Program Cooper 3.9-7 Amendment No. 258

RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure Vessel (RPV) Water Level LCO 3.9.6 RPV water level shall be~ 21 ft above the top of the RPV flange.

APPLICABILITY: During movement of irradiated fuel assemblies within the RPV, During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not within A.1 Suspend movement of fuel Immediately limit. assemblies and handling of control rods within the RPV.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is ~ 21 ft above the top of the In accordance with RPVflange. the Surveillance Frequency Control Program Cooper 3.9-8 Amendment No. 258

RHR - High Water Level 3.9.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem is In accordance with operating. the Surveillance Frequency Control Program Cooper 3.9-11 Amendment No. 258

RHR - Low Water Level 3.9.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem is In accordance with operating. the Surveillance Frequency Control Program Cooper 3.9-14 Amendment No. 258

Reactor Mode Switch Interlock Testing 3.10.2 ACTIONS -~------~.--~----------~-------

(contin~:;DIT~~ _R_:-:-~-~-:h-E~=~~:ode COMPLETION TIME A. A-.3-_-1 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch in the shutdown position.

OR A.3.2 --------------NOTE-------------

Only applicable in MODE 5.

Place the reactor mode 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch in the refuel position.

SURVEILLANCE SURVEILLANCE SR 3.10.2.1 Verify all control rods are fully inserted in core cells In accordance with containing one or more fuel assemblies. the Surveillance Frequency Control Program SR 3.10.2.2 Verify no CORE ALTERATIONS are in progress. In accordance with Cooper 3.10-5 Amendment No. 258

Single Control Rod Withdrawal - Hot Shutdown 3.10.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.3.1 Perform the applicable SRs for the required LCOs. According to the applicable SRs SR 3.10.3.2 ----~~~~~---~--~--~~NOT"E---~--~-~~----~-----~-

Not required to be met if SR 3.10.3.1 is satisfied for LCO 3.10.3.d.1 requirements.

Verify all control rods, other than the control rod being In accordance with withdrawn, in a five by five array centered on the the Surveillance control rod being withdrawn, are disarmed. Frequency Control Program SR 3.10.3.3 Verify all control rods, other than the control rod being In accordance with withdrawn, are fully inserted. the Surveillance Frequency Control Program Cooper 3.10-8 Amendment No. 258

Single Control Rod Withdrawal - Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.10.4.2 ----------------NOTE-----'--*-------------

Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c.1 requirements.

Verify all control rods, other than the control rod being In accordance with withdrawn, in a five by five array centered on the the Surveillance control rod being withdrawn, are disarmed. Frequency Control Program SR 3.10.4.3 Verify all control rods, other than the control rod being In accordance with withdrawn, are fully inserted. the Surveillance Frequency Control Program SR 3.10.4.4 ------------------NOTE------------------

Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.b.1 requirements.

Verify a control rod withdrawal block is inserted. In accordance with the Surveillance Frequency Control Program Cooper 3.10-12 Amendment No. 258

Single CRD Removal - Refueling 3.10.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Initiate action to fully insert Immediately all control rods.

A.2.2 Initiate action to satisfy the Immediately requirements of this LCO.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all control rods, other than the control rod In accordance with withdrawn for the removal of the associated CRD, are the Surveillance fully inserted. Frequency Control Program SR 3.10.5.2 Verify all control rods, other than the control rod In accordance with withdrawn for the removal of the associated CRD, in the Surveillance a five by five array centered on the control rod Frequency Control withdrawn for the removal of the associated CRD, are Program disarmed.

SR 3.10.5.3 Verify a control rod withdrawal block is inserted. In accordance with the Surveillance Frequency Control Program SR 3.10.5.4 Perform SR 3.1.1.1. According to SR 3.1.1.1 (continued)

Cooper 3.10-14 Amendment No. 258

Single CRD Removal - Refueling 3.10.5 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.10.5.5 Verify no other CORE ALTERATIONS are in In accordance with progress. the Surveillance Frequency Control Program Cooper 3.10-15 Amendment No. 258

Multiple Control Rod Withdrawal - Refueling 3.10.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Initiate action to fully insert Immediately all control rods in core cells containing one or more fuel assemblies.

A.3.2 Initiate action to satisfy the Immediately requirements of this LCO.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.6.1 Verify the four fuel assemblies are removed from core In accordance with cells associated with each control rod or CRD the Surveillance removed. Frequency Control Program SR 3.10.6.2 Verify all other control rods in core cells containing In accordance with one or more fuel assemblies are fully inserted. the Surveillance Frequency Control Program SR 3.10.6.3 ---------------------------NOTE------------------------------

Only required to be met during fuel loading.

Verify fuel assemblies being loaded are in In accordance with compliance with an approved spiral reload sequence. the Surveillance Frequency Control Program Cooper 3.10-17 Amendment No. 258

SOM Test- Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.10.8.2 -----------NOTE--------------

Not required to be met if SR 3.10.8.3 satisfied.

Perform the MODE 2 applicable SRs for LCO 3.3.2.1, According to the Function 2 of Table 3.3.2.1-1. applicable SRs SR 3.10.8.3 ---------------NOTE---------------

Not required to be met if SR 3.10.8.2 satisfied.

Verify movement of control rods is in compliance with During control rod the approved control rod sequence for the SOM test movement

by a second licensed operator or other qualified member of the technical staff.

SR 3.10.8.4 Verify no other CORE ALTERATIONS are in In accordance with progress. the Surveillance Frequency Control Program SR 3.10.8.5 Verify each withdrawn control rod does not go to the Each time the withdrawn overtravel position.

control rod is withdrawn to "full out" position Prior to satisfying LCO 3.10.8.c requirement after work on control rod orCRD System that could affect coupling (continued)

Cooper 3.10-22 Amendment No. 258

SOM Test- Refueling 3.10.8 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.10.8.6 Verify CRD charging water header pressure ~ 940 In accordance with psig. the Surveillance Frequency Control Program Cooper 3.10-23 Amendment No. 258

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Control Room Envelope Habitability Program (continued) personnel receiving radiation exposures in excess of either (a) 5 rem whole body or its equivalent to any part of the body for the duration of the loss-of-coolant accident, or (b) 5 rem total effective dose equivalent (TEDE) for the duration of the fuel handling accident. The program shall include the following elements:

a. The definition of the CRE and CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0. No exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0, are proposed.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by the CREF System, operating at the flow rate required by the Ventilation Filter Testing Program, at a Frequency of 24 months. The results shall be trended and used as part of the periodic assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitabiiity, determining CRE unfiltered air inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and *d, respectively.

(continued) I Cooper 5.0-18 Amendment No. 258

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

Cooper 5.0-19 Amendment No. 258

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 (Deleted) 5.6.2 Annual Radiological Environmental Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Assessment Manual (ODAM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODAM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in Regulatory Guide 4.8, December 1975.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted in a supplementary report as soon as possible.

(continued)

Cooper 5.0-20 Amendment No. 258

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effl.uent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.8.1.

5.6.4 (Deleted) 5.6.5 Core Operating Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and 3. 7. 7.

2. The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.7.7.

3. The Linear Heat Generation Rates for Specifications 3.2.3 and 3. 7.7.

4. The three Rod Block Monitor Upscale Allowable Values for Specification 3.3.2.1.

5. The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel" (Revision specified in the COLR).

(continued)

Cooper 5.0-21 Amendment No. 258

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

2. NEDE-23785-1-P-A, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident", Volume Ill, Revision 1, October 1984.

3. NED0-31960 and NED0-31960 Supplement 1, "BWR Owner's Group Long-Term Stability Solutions Licensing Methodology" (the approved Revision at the time the reload analysis is performed).

c. The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

{PTLR)

a. Reactor pressure and temperature limit for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Limiting Conditions for Operations Section 3.4.9, "RCS Pressure and Temperature (PfT) Limits."

2. Surveillance Requirements Section 3.4.9, "RCS Pressure and Temperature (PfT) Limits."

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

(continued)

Cooper 5.0-22 Amendment No. 258

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT PTLR) (continued)

1. BWROG-TP-11-022-A, Revision 1 (SIR-05-044, Revision 1-A), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,"

dated August 2013.

2. BWROG-TP-11-023-A, Revision 0 (0900876.401, Revision 0-A), "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations," dated May 2013.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

(continued)

Cooper 5.0-23 Amendment No. 258

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS

5. 7 Radiation Area

5. 7.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.1601 of 10 CFR Part 20, each high radiation area in which the deep dose equivalent in excess of 100 mrem but less than 1000 mrem in one hour (measurement made at 12 inches from source of radiation) shall be barricaded (barricade will impede physical movement across the entrance or access to the high radiation area; i.e., doors, yellow and magenta rope, turnstile) and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Special Work Permit (SWP). Radiation protection personnel or personnel escorted by radiation protection personnel shall be exempt from the SWP issuance requirement during the performance of their assigned duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A monitoring device which continuously indicates the radiation dose rate in the area.

b. A monitoring device which continuously integrates the radiation dose in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rates in the area have been established and personnel have been made knowledgeable of them.

c. A radiation protection qualified individual (i.e., qualified in radiation protection procedures), with a dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic dose rate monitoring at the frequency specified by Health Physics supervision.

5. 7 .2 In addition to the requirements of Specification 5. 7 .1, areas accessible to personnel with dose rates such that a major portion of the body could receive in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a deep dose equivalent in excess of 1000 mrem (measurement made at 12 inches from source of radiation) shall be provided with locked doors to prevent unauthorized entry. Doors shall remain locked except during periods of access by personnel under an approved SWP which shall specify the dose rates in the immediate work area. For individual high radiation areas accessible to personnel that are located within large areas, such as the containment, or areas where no enclosure exists for purposes of locking and no enclosure can be reasonably constructed around the individual areas, then that area shall be barricaded and conspicuously posted. Area radiation monitors that have been set to alarm if radiation levels increase, provide both a visual and an audible signal to alert personnel in the area of the increase. These monitors may be used to meet Specification 5. 7 .1.a provided that the dose rates and alarms have been established by radiation protection personnel. Stay times or continuous surveillance, direct or remote (such as use of closed circuit TV cameras), may be made by personnel qualified in radiation protection procedures to provide additional positive exposure control over the activities within the area.

Cooper 5.0-24 Amendment No. 258 I

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 258 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1.0 INTRODUCTION

By application dated March 22, 2016 (Reference 1), as supplemented by two letters dated December 7, 2016 (References 2 and 3), Nebraska Public Power District (NPPD or the licensee), requested changes to the Technical Specifications (TSs) for Cooper Nuclear Station (CNS).

The proposed changes would revise the CNS TSs to adopt the U.S. Nuclear Regulatory Commission (NRC)-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF [Risk-Informed TSTF] Initiative Sb" (Reference 4) for CNS.

The two supplemental letters dated December 7, 2016, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on May 24, 2016 (81 FR 32807).

2.0 REGULATORY EVALUATION

2.1 Description of the Proposed Changes The licensee proposed to modify the CNS TSs by relocating specific surveillance frequencies to a licensee-controlled program (i.e., the Surveillance Frequency Control Program (SFCP) in accordance with Nuclear Energy lnsitute (NEI) 04-10, Revision 1 (Reference 5). The licensee stated that the proposed change is consistent with the adoption of NRG-approved TSTF-425, Revision 3. When implemented, TSTF-425, Revision 3, relocates most periodic frequencies of TS surveillances to the SFCP, and provides requirements for the new SFCP in the Administrative Controls section of the TSs. All surveillance frequencies can be relocated except the following:

Frequencies that reference other approved programs for the specific interval, such as the lnservice Testing Program or the Primary Containment Leakage Rate Testing Program; Enclosure 2

Frequencies that are purely event-driven (e.g., "each time the control rod is withdrawn to the 'full out' position");

Frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching:::: [greater than or equal to] 95% RTP [rated thermal power]"); and

Frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g., "drywell to suppression chamber differential pressure decrease").

The licensee proposed to relocate specific surveillance frequencies from the following TS Sections to the SFCP:

3. 1 Reactivity Control System 3.2 Power Distribution Limits 3.3 Instrumentation 3.4 Reactor Coolant System (RCS) 3.5 Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System 3.6 Containment Systems

3. 7 Plant Systems 3.8 Electrical Power Systems 3.9 Refueling Operations 3.1 O Special Operations The licensee proposed to add the SFCP to CNS TS, Section 5.0, "Administrative Controls."

Proposed TS 5.5.14 describes the requirements for the SFCP to control changes to the relocated surveillance frequencies to ensure that surveillances are performed at intervals to ensure limiting conditions for operation are met. The TS Bases for each affected surveillance would be revised to state that the surveillance frequency is controlled under the SFCP and were included in the application for information only. The proposed changes to the Administrative Controls section of the TSs include a specific reference to NEI 04-10, Revision 1, as the basis for making any changes to the surveillance frequencies when they are relocated out of the TSs.

In letter dated September 19, 2007 (Reference 6), the NRC staff approved Topical Report NEI 04-10, Revision 1, as an acceptable methodology for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, Revision 1, and the safety evaluation (SE) providing the basis for NRC acceptance of NEI 04-10, Revision 1.

The licensee proposed other changes and deviations from TSTF-425, which are discussed in Section 3.3 of this SE.

2.2 Applicable Commission Policy Statements In the "Final Policy Statement: Technical Specifications Improvements for Nuclear Power Plants," dated July 22, 1993 (58 FR 39132), the NRC addressed the use of probabilistic safety analysis (PSA, currently referred to as probabilistic risk assessment or PRA) in STS. In this 1993 publication, the NRC states, in part:

The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria [of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36] to be deleted from Technical Specifications based solely on PSA (Criterion 4). However, if the results of PSA indicate that Technical Specifications can be relaxed or removed, a deterministic review will be performed ....

The Commission Policy in this regard is consistent with its Policy Statement on "Safety Goals for the Operation of Nuclear Power Plants," 51 FR 30028, published on August 21, 1986. The Policy Statement on Safety Goals states in part, "* *

probabilistic results should also be reasonably balanced and supported through use of deterministic arguments. In this way, judgments can be made * *

about the degree of confidence to be given these [probabilistic]

estimates and assumptions. This is a key part of the process for determining the degree of regulatory conservatism that may be warranted for particular decisions.

This defense-in-depth approach is expected to continue to ensure the protection of public health and safety." ...

The Commission will continue to use PSA, consistent with its policy on Safety Goals, as a tool in evaluating specific line-item improvements to Technical Specifications, new requirements, and industry proposals for risk-based Technical Specification changes ....

Approximately 2 years later, the NRC provided additional detail concerning the use of PRA in the "Final Policy Statement: Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities," dated August 16, 1995 (60 FR 42622). In this publication, the NRC states, in part:

The Commission believes that an overall policy on the use of PRA methods in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that would promote regulatory stability and efficiency. In addition, the Commission believes that the use of PRA technology in NRC regulatory activities should be increased to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach ....

PRA addresses a broad spectrum of initiating events by assessing the event frequency. Mitigating system reliability is then assessed, including the potential for multiple and common cause failures. The treatment therefore goes beyond the single failure requirements in the deterministic approach. The probabilistic approach to regulation is, therefore, considered an extension and enhancement of traditional regulation by considering risk in a more coherent and complete manner....

Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency. This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the

scope of PRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data ....

Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA:

(1) The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

(2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

(3) PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

(4) The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.

2.3 Applicable Regulations In 10 CFR 50.36, "Technical specifications," the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. These categories will remain in the CNS TSs.

Section 50.36(c)(3) of 10 CFR states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The FR notice published on July 6, 2009 (74 FR 31996), which announced the availability of TSTF-425, Revision 3, states that the addition of the SFCP to the TSs provides the necessary administrative controls to require that surveillance frequencies relocated to the SFCP are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The FR notice also states that changes to surveillance frequencies in the SFCP are made using the methodology contained in NEI 04-10,

Revision 1, including qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, and recommended monitoring of structures, systems, and components (SSCs), and are required to be documented.

Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants" (i.e., the Maintenance Rule), and 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," require licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. Such failures can result in the licensee increasing the frequency of a surveillance test. In addition, by having the TSs require that changes to the frequencies listed in the SFCP be made in accordance with NEI 04-10, Revision 1, the licensee will be required to monitor the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs.

2.4 Applicable NRC Regulatory Guides and Review Plans Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2 (Reference 7), describes an acceptable risk-informed approach for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 1 (Reference 8), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference 9), describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decisionmaking for light-water reactors (LWRs).

NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance" (Reference 10) provides general guidance for evaluating the technical basis for proposed risk-informed changes. Guidance on evaluating PRA technical adequacy is provided in SRP, Chapter 19, Section 19.1, Revision 3, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load" (Reference 11 ). More specific guidance related to risk-informed TS changes is provided in SRP, Chapter 16, Section 16.1, Revision 1, "Risk-Informed Decisionmaking: Technical Specifications" (Reference 12), which includes changes to surveillance test intervals (STls) (i.e.,

surveillance frequencies) as part of risk-informed decision making. Section 19.2 of the SRP references the same criteria as RG 1.174, Revision 2, and RG 1.177, Revision 1, and states that a risk-informed application should be evaluated to ensure that the proposed changes meet the following key principles:

The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change;

The proposed change is consistent with the defense-in-depth philosophy;

The proposed change maintains sufficient safety margins;

When proposed changes result in an increase in core damage frequency (CDF) or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement;

The impact of the proposed change should be monitored using performance measurement strategies.

NUREG-1433 "Standard Technical Specifications, General Electric BWR/4 Plant," Volume 1, Specifications and Volume 2, Bases, Revision 4.0 (Reference 13), contains the improved STS for General Electric BWR/4 plants. The improved STS were developed based on the criteria in the Final Commission Policy Statement of Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132), which was subsequently codified by changes to 10 CFR 50.36 (60 FR 36953).

3.0 TECHNICAL EVALUATION

The licensee's adoption of TSTF-425, Revision 3, provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the Administrative Controls section of TSs. The changes to the Administrative Controls section of the TSs will also require the application of NEI 04-10, Revision 1, for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes described in TSTF-425, Revision 3, included documentation regarding the PRA technical adequacy consistent with RG 1.200, Revision 2. NEI 04-10, Revision 1, states that PRA methods are used with plant performance data and other considerations to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is consistent with guidance provided in RG 1.174, Revision 2, and RG 1.177, Revision 1, in support of changes to STls.

3.1 Key Principles RG 1.177, Revision 1, identifies five key safety principles required for risk-informed changes to TSs. Each of these principles is addressed by NEI 04-1 O, Revision 1. Sections 3.1.1 through 3.1.5 of this section contain a discussion of the five principles, including the NRC staff's evaluation of how the licensee's license amendment request (LAR) satisfies each principle.

3.1.1 The Proposed Change Meets Current Regulations Section 50.36(c)(3) of 10 CFR requires that TSs include surveillances, which are "requirements relating to test, calibration, or inspection to assure that necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The licensee is required by its TSs to perform surveillance tests, calibration, or inspection on specific safety-related equipment (e.g., reactivity control, power distribution, electrical, and instrumentation) to verify system operability. Surveillance frequencies are based primarily upon deterministic methods, such as engineering judgment, operating experience, and manufacturer's recommendations. The licensee's use of NRG-approved methodologies identified in NEI 04-10, Revision 1, provides a way to establish

risk-informed surveillance frequencies that complements the deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

The SRs remain in the TSs, as required by 10 CFR 50.36(c)(3). This change is analogous with other NRG-approved TS changes in which the SRs are retained in TSs, but the related surveillance frequencies are relocated to licensee-controlled documents, such as surveillances performed in accordance with the lnservice Testing Program and the Primary Containment Leakage Rate Testing Program. Thus, this proposed change complies with 10 CFR 50.36(c)(3) by retaining the requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The regulatory requirements in 10 CFR 50.65 and 10 CFR Part 50, Appendix B, and the monitoring required by NEI 04-10, Revision 1, ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions taken. The licensee's SFCP ensures that SRs specified in the TSs are performed at intervals sufficient to assure that the above regulatory requirements are met. Based on the foregoing, the NRC staff concludes that the proposed change meets the first key safety principle of RG 1.177, Revision 1, by complying with current regulations.

3.1.2 The Proposed Change Is Consistent with the Defense-in-Depth Philosophy The defense-in-depth philosophy (i.e., the second key safety principle of RG 1.177, Revision 1) is maintained if:

A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation;

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided;

System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers). (Because the scope of the proposed methodology is limited to revision of surveillance frequencies, the redundancy, independence, and diversity of plant systems are not impacted.);

Defenses against potential common cause failures (CCFs) are preserved, and the potential for the introduction of new CCF mechanisms is assessed;

Independence of barriers is not degraded;

Defenses against human errors are preserved;

The intent of the General Design Criteria in 10 CFR Part 50, Appendix A, is maintained.

The changes to the Administrative Controls section of the TSs will require the application of NEI 04-10, Revision 1, for any changes to surveillance frequencies within the SFCP.

NEI 04-10, Revision 1, uses both the GDF and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. In accordance with RG 1.174, Revision 2, and RG 1.177, Revision 1, changes to GDF and LERF are evaluated using a comprehensive risk analysis, which assesses the impact of proposed changes, including contributions from human errors and CCFs. Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of CCFs. The NRG staff concludes that both the quantitative risk analysis and the qualitative considerations provide reasonable assurance that defense-in-depth is maintained to ensure protection of public health and safety, satisfying the second key safety principle of RG 1.177, Revision 1.

3.1.3 The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that will be conducted by the licensee under the SFCP when frequencies are revised will assess the impact of the proposed frequency change to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the proposed surveillance test frequency change is not in conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis; or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

The design, operation, testing methods, and acceptance criteria for SSCs specified in applicable codes and standards (or alternatives approved for use by the NRG) will continue to be met as described in the plants' licensing bases, including the Updated Final Safety Analysis Report and TS Bases, because these are not affected by changes to the surveillance frequencies.

Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. On this basis, the NRG staff concludes that safety margins are maintained by the proposed methodology and, therefore, the third key safety principle of RG 1.177, Revision 1, is satisfied.

3.1.4 When Proposed Changes Result in an Increase in GDF or Risk, the Increases Should Be Small and Consistent with the Intent of the Commission's Safety Goal Policy Statement RG 1.177, Revision 1, provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies, which requires identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency, and performance of sensitivity and uncertainty evaluations. The changes to the Administrative Controls section of the TSs will require application of NEI 04-10, Revision 1, in the SFCP. NEI 04-10, Revision 1, satisfies the intent of RG 1.177, Revision 1, guidance for evaluation of the change in risk, and for assuring that such changes are small by providing the technical methodology to support risk-informed TSs for control of surveillance frequencies.

3.1.4.1 PRA Technical Adequacy The technical adequacy of the licensee's PRA must be commensurate with the safety significance of the proposed TS change and the role the PRA plays in justifying the change.

That is, the greater the change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the technical adequacy of the PRA.

RG 1.200 (Reference 9) provides regulatory guidance for assessing the technical adequacy of a PAA. The current revision (i.e., Revision 2) of this RG endorses, with clarifications and qualifications, the use of the following:

(1) American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) RA-Sa-2009, "Addenda to ASME RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (i.e., the PAA Standard) (Reference 14),

(2) NEI 00-02, "PAA Peer Review Process Guidance" (Reference 15), and (3) NEI 05-04, "Process for Performing Internal Events PAA Peer Reviews Using the ASME/ANS PAA Standard," Revision 2 (Reference 16).

The licensee performed an assessment of the PAA models used to support the SFCP using the guidance of RG 1.200, Revision 2, to ensure that the PAA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability Category (CC) II of the NRG-endorsed PAA standard is the target capability level for supporting requirements for the internal events PAA (IEPRA) for this application. Any identified deficiencies to those requirements are further assessed to determine any impacts to proposed decreases to surveillance frequencies, including the use of sensitivity studies where appropriate, in accordance with NEI 04-10, Revision 1.

In Attachment 2 of its letter dated March 22, 2016, the licensee stated that a full scope peer review by the Boiling Water Reactor Owners Group (BWROG) of the CNS IEPRA was performed in June 2008 using RG 1.200, Revision 1 (Reference 17), and PAA standard ASME RA-Sc-2007 Addenda to "ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 18). The licensee also performed a self-assessment of the IEPRA in 2014 to update the peer review results from Revision 1 to Revision 2 of RG 1.200. In March 2011, the BWROG performed a peer review of the CNS PAA fire events model using RG 1.200, Revision 2, NEI 07-12 (Reference 19), and the 2009 Addendum to the ASME/ANS RA-Sa-2009 PAA Standard.

Findings from the June 2008 full scope peer review and March 2011 peer review of the fire PAA were also provided in Attachment 2, Tables 2-1 and 2-2 of the LAR. The tables identify PRA standard supporting requirements determined by the peer review not to meet CC II and provide an assessment of the impact of these findings (referred to in the LAA as "gaps") for this application. In the LAA, the licensee explains that the impact of each "gap" will be reviewed as part of STI change evaluations.

The NRC staff reviewed the 2008 full scope peer review findings on the IEPRA, and the 2011 peer review findings for the fire PAA. The NRC staff assessed these "gaps" and findings to identify whether any incompleteness could impact this application, and to ensure that any deficiencies in meeting CC II can be addressed for the SFCP per the NRG-approved NEI 04-10, Revision 1 methodology.

The NRC staff noted several findings identified in Attachment 2, Tables 2-1 and 2-2 of the LAA that are dispositioned as having little or no impact on STI evaluations without explanation of this determination. As described below, the NRC staff requested further information via requests for

additional information (RAls) regarding "gaps" that appeared to represent incompleteness in the internal events and fire PRA modeling.

The BWROG PRA Peer Review created Facts and Observations (F&O) QU-FS-01, which corresponds to ASME/ANS RA-Sa-2007 standard element QU-FS. In F&O QU-FS-01, the peer review team noted that using initiating events frequency estimates in the PRA instead of initiating event fault trees rendered the PRA incomplete. The licensee dispositioned the F&O by stating that application-specific guidelines have been revised to ensure limitations in the quantification are documented. In an e-mail dated October 27, 2016 (Reference 20), the NRC staff requested that the licensee explain how the CNS PRA models will address this concern in Step 8 of the NEI 04-10, Revision 1 guidance to assure accuracy in calculations of net change in CDF and LERF for evaluations of STls. In its response to APLA RAl-1, by letter dated December 7, 2016 (Reference 2), the licensee stated that the initiating frequency estimates have been replaced with fault trees during a routine update to the CNS IEPRA. Based on the licensee's incorporation of initiating event fault trees into the PRA, the NRC staff concludes that this deficiency in the PRA has been removed and the RAI response is acceptable.

The BWROG PRA Peer Review created F&O HR-G7, which corresponds to ASME/ANS RA-Sa-2009 standard element HR-G7, due to the method used for assessing dependencies associated with cutsets containing multiple human errors. The licensee dispositioned the F&O by stating that each human error probability (HEP) combination identified in the finding was evaluated after the original HEP cutset calculation. In the e-mail dated October 27, 2016, the NRC staff requested that the licensee explain how the CNS PRA models address the quantification of "dependent" vs. "independent" HEPs in sequence cutsets containing multiple human errors in accordance with RG 1.200, Revision 2, and ASME/ANS RA-Sa-2009. In its response to APLA RAl-2 in the letter dated December 7, 2016 (Reference 2), the licensee stated that the human reliability dependency analyses, along with HEP floor value assignments, apply an endorsed method consistent with NUREG-1921,

"[Electric Power Research Institute] EPRl/NRC-RES Fire Human Reliability Analysis" (Reference 21 ), and that this method will continue to be used in support of the proposed STI change evaluations. The licensee further stated that discussions detailed in the Office of Nuclear Reactor Regulation's SE for Transition to National Fire Protection Association (NFPA)

NFPA 805 (Reference 22) address the licensee's human reliability analysis and application of HEP floor values consistent with the guidance in NUREG-1921. The NRC staff concludes that the licensee's response to the RAI is acceptable because using the guidance in NUREG-1921 is an acceptable method of assigning an HEP, and the methodology was used to consider HEPs in the licensee's PRA.

The CNS fire PRA (FPRA) Peer Review identified F&Os associated with supporting requirements SY-A2, SY-C2, SY-A3, and DA-C2, due to significant system modeling that had been performed. The licensee dispositioned the F&Os by stating that new components and fire-induced impacts should be considered. In the e-mail dated October 27, 2016, the NRC staff requested that the licensee confirm that the enhanced feedwater system modeling for the FPRA has been incorporated into the IEPRA to reflect the as-built as-operated plant. In its response to APLA RAl-3(a), by letter dated December 7, 2016 (Reference 2), the licensee stated that the elements of feedwater system modeling were contained in the current internal events model.

The licensee further stated that new components have been evaluated in Nuclear Engineering Design Calculation (NEDC)09-079 for applicability with respect to the IEPRA, and that this evaluation determined that new components added by the FPRA address fire specific failure modes. Based on the foregoing, the NRC staff concludes the licensee's response to APLA RAl-3 is acceptable because new models, as applicable, have been included in the IEPRA.

Fire ignition frequencies and non-suppression probabilities were previously developed in the NUREG/CR-6850/EPRI 1011989 and revised in Supplement 1 to NUREG/CR-6850/EPRI 1019259. For supporting requirement QU-E3, Table 2-2 of Reference 1, the CNS disposition stated in part, "generic fire frequencies are directly based on assumptions in NUREG/CR-6850 (including FAQ-48 enhancements)." NUREG-2169, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database (FEDB), dated January 2015 (Reference 23), provides updated fire ignition frequency estimates using the most current FEDS data while applying methodology enhancements. In the e-mail dated October 27, 2016 (Reference 20), the NRC staff requested that the licensee explain which updated fire ignition frequencies will be used for the STI evaluation and whether the values will come from the NRG-endorsed guidance NUREG-2169.

In its response to APLA RAl-5, by letter dated December 7, 2016 (Reference 2), the licensee stated that the CNS PRA program has included the updated fire ignition values from the NRG-endorsed guidance NUREG-2169. Based on the licensee's incorporation of the updated fire ignition frequencies in to the PRA, the NRC staff concludes that this deficiency in the PRA has been removed and the licensee's response is acceptable.

The CNS FPRA Peer Review identified an F&O associated with supporting requirement PRM-89 scenario quantification using PRAQuant. The licensee disposition states, in part, that "PRAQuant solves each fire scenario by setting all internal events initiators to 0.0 and setting fire initiators and those basic events representing components impacted by the fire to 1.0." In the e-mail dated October 27, 2016, the NRC staff requested that the licensee explain whether the quantification of the FPRA included both the random (non-fire) and fire-induced failure probabilities of fire-affected basic events that could lead to double-counting failures. In its response to APLA RAl-6, by letter dated December 7, 2016 (Reference 2), the licensee stated that the CNS FPRA quantification and LERF include both random (non-fire) and fire-induced failure probabilities, and that there is negligible impact from the random failure when the FPRA is quantified. The staff concludes that the disposition is acceptable for this application because the fire-induced failure probabilities are generally orders of magnitude higher than the random failure probabilities and the random failure would therefore contribute negligibly to the fire risk.

The NRC staff notes that this revision of the CNS PRA had been previously reviewed by the NRC and found to be technically adequate to support the risk-informed application to adopt NFPA 805 as incorporated by reference into 10 CFR 50.48(c) (Reference 22). Based on the review of the information provided to support the NFPA 805 application, and the information provided in the March 22, 2016, LAR, and supplemental letter dated December 7, 2016 (Reference 2), the NRC staff concludes that the review of the PRA is consistent with Regulatory Position 2.3.1, "Technical Adequacy of the PRA," of RG 1.177, Revision 1 (Reference 8). As summarized in this SE, the NRC staff concludes that any deficiencies identified during the review of the PRA have been resolved to support the evaluation of changes proposed to surveillance frequencies within the SFCP.

3.1.4.2 Scope of the PRA The proposed changes to the Administrative Controls section of the TSs would require the licensee to evaluate each proposed change to a relocated surveillance frequency using NEI 04-10, Revision 1, to determine its potential impact on CDF and LERF from internal events, fires, seismic, other external events, and shutdown conditions. In cases where a PRA of sufficient scope or quantitative risk models are unavailable, the licensee uses bounding

analyses, or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be insignificant.

The licensee has at-power internal events, internal flooding, and fire PRA models. In accordance with NEI 04-10, Revision 1, as required by proposed TS 5.5.14, the licensee will use these PRA models to perform quantitative evaluations to support the development of changes to surveillance frequencies in the SFCP. This is acceptable because the NRC-approved methodology in NEI 04-10, Revision 1, allows for more refined analysis to be performed supporting changes to surveillance frequencies in the SFCP.

For other hazard groups for which a PRA model does not exist, a qualitative or bounding analysis, consistent with NEI 04-10, Revision 1, is performed to provide justification for the acceptability of the proposed test interval change. Based on the application of NRC-approved NEI 04-10, Revision 1, as required by proposed TS 5.5.14, the NRC staff concludes that the licensee's evaluation methodology is sufficient to ensure the risk contribution of each surveillance frequency change is properly identified for evaluation and is consistent with Regulatory Position 2.3.2, "Scope of the Probabilistic Risk Assessment for Technical Specification Change Evaluations," of RG 1.177, Revision 1.

3.1.4.3 PRA Modeling The licensee's methodology includes the determination of whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out.

The methodology adjusts the failure probability of the impacted SSCs, including any impacted CCF modes, based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy, consistent with guidance contained in RG 1.200, Revision 2, and by sensitivity studies identified in NEI 04-10, Revision 1.

By letter dated September 19, 2007 (Reference 6), the NRC staff approved Topical Report NEl-04-10, Revision 1, which describes an acceptable methodology to evaluate changes in surveillance frequency. The NRC staff concludes that the CNS PRA modeling is consistent with the guidance in NEl-04-10, Revision 1, and therefore the modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance frequency, and is consistent with Regulatory Position 2.3.3, "Probabilistic Risk Assessment Modeling," of RG 1.177, Revision 1.

3.1.4.4 Assumptions for Time Related Failure Contributions The failure probabilities of SSCs modeled in PRAs may include a standby time-related

contribution and a cyclic demand-related contribution. In Attachment 2, Section 2.4, "Identification of Key Assumptions," of the LAR dated March 22, 2016, the licensee states that the determination of standby failure rates are a key source of uncertainty and therefore, sensitivity studies will be performed on standby failure rates for STI evaluations. The NEI 04-10, Revision 1 criteria adjust the time-related failure contribution of SSCs affected by the proposed change to a surveillance frequency. This is consistent with RG 1.177, Revision 1, Section 2.3.3, which permits separation of the failure rate contributions into demand and standby for evaluation of SRs. If the available data do not support distinguishing between the

time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The SSC failure rate per unit time is assumed to be unaffected by the change in test frequency, such that the failure probability is assumed to increase linearly with time. This assumption will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The NEI 04-10, Revision 1, process requires consideration of qualitative sources of information with regard to potential impacts of test frequency on SSC performance, including industry and plant-specific operating experience, vendor recommendations, industry standards, and code-specified test intervals. Thus, the NRC staff concludes that the licensee's process is not reliant upon risk analyses as the sole basis for the proposed changes because the licensee has, and will, apply the associated guidance in NRG-approved NEI 04-10, Revision 1.

The potential benefits of a reduced surveillance frequency, including reduced downtime and reduced potential for restoration errors, test-caused transients, and test-caused wear of equipment, are identified qualitatively, but are not quantitatively assessed. The NRC staff concludes that the licensee applied NRG-approved NEI 04-10, Revision 1, to employ reasonable assumptions with regard to extensions of STls, and the requested changes are consistent with Regulatory Position 2.3.4, "Assumptions in Completion Time and Surveillance Frequency Evaluations, of RG 1.177, Revision 1.

3.1.4.5 Sensitivity and Uncertainty Analyses The proposed amended TSs would require that changes to the frequencies listed in the SFCP be made in accordance with NEI 04-10, Revision 1. Therefore, the licensee will be required to have sensitivity studies that assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact on the frequency of initiating events, and any identified deviations from CC II of the PRA standard. Where the sensitivity analyses identify a potential impact on the proposed change, revised surveillance frequencies are considered, along with any qualitative considerations that may bear on the results of such sensitivity studies. In accordance with NEI 04-10, Revision 1, as required by proposed TS 5.5.14, the licensee will also perform monitoring and feedback of SSC performance, once the revised surveillance frequencies are implemented. Therefore, the NRC staff concludes that the licensee appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, and the LAR is consistent with Regulatory Position 2.3.5, "Sensitivity and Uncertainty Analyses Relating to Assumptions in Technical Specification Change Evaluations," of RG 1.177, Revision 1, because the licensee has, and will, apply the associated guidance in NRG-approved NEI 04-10, Revision 1.

3.1.4.6 Acceptance Guidelines In accordance with NEI 04-10, Revision 1, as required by proposed TS 5.5.14, the licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using NEI 04-10, Revision 1, in accordance with the TS SFCP. Each individual change to surveillance frequency must show a risk impact below 1E-6 per year for change to CDF, and below 1E-7 per year for change to LERF. These changes to CDF and LERF are consistent with the acceptance criteria of RG 1.174, Revision 2 (Reference 7), for very small changes in risk. Where the RG 1.174, Revision 2, acceptance criteria are not met, the process in NEI 04-10, Revision 1, either considers revised surveillance frequencies that are

consistent with RG 1.174, Revision 2, or the process terminates without permitting the proposed changes. Where quantitative results are unavailable for comparison with the acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible or insignificant. Otherwise, bounding quantitative analyses are required that demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174, Revision 2, acceptance guidelines for very small changes in risk. In addition to assessing each individual SSC surveillance frequency change, the cumulative impact of all changes must result in a risk impact less than 1E-5 per year for change to CDF, and less than 1E-6 per year for change to LERF. The total CDF and total LERF must be reasonably shown to be less than 1E-4 per year and 1E-5 per year, respectively.

These values are consistent with the acceptance criteria of RG 1.174, Revision 2, as referenced by RG 1.177, Revision 1 (Reference 8), for changes to surveillance frequencies.

Consistent with the NRC staff's SE dated September 19, 2007, for NEI 04-10, Revision 1 (Reference 6), the TS SFCP will require the licensee to calculate the total change in risk (i.e., the cumulative risk) by comparing a baseline model that uses failure probabilities based on surveillance frequencies prior to being changed per the SFCP, to a revised model that uses failure probabilities based on the changed surveillance frequencies. The NRC staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (i.e., less than 5E-8 per year for CDF and 5E-9 per year for LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies.

The quantitative acceptance guidance of RG 1.174, Revision 2, is supplemented by qualitative information to evaluate the proposed changes to surveillance frequencies, including industry and plant-specific operating experience, vendor recommendations, industry standards, the results of sensitivity studies, and SSC performance data and test history. The final acceptability of the proposed change is based on all of these considerations and not solely on the PRA results. Post-implementation performance monitoring and feedback are also required to ensure continued reliability of the components. The licensee's application of NRG-approved NEI 04-10, Revision 1, provides acceptable methods for evaluating the risk increase associated with proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177, Revision 1. Therefore, the NRC staff concludes that the proposed methodology satisfies the fourth key safety principle of RG 1.177, Revision 1, by assuring that any increase in risk is small, consistent with the intent of the Commission's Safety Goal Policy Statement.

3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's proposed TS 5.5.14 requires application of NEI 04-10, Revision 1 (Reference 5),

in the SFCP. NEI 04-10, Revision 1, requires performance monitoring of SSCs whose surveillance frequencies have been revised as part of a feedback process to ensure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of Maintenance Rule (i.e., 10 CFR 50.65) monitoring of equipment performance. In the event of SSC performance degradation, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions that may be required by the Maintenance Rule. The performance monitoring and feedback specified in NEI 04-10, Revision 1, is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 of RG 1.177, Revision 1. Thus, the NRC staff concludes that the fifth key safety principle of RG 1.177, Revision 1, is satisfied.

3.2 Addition of Surveillance Frequency Control Program to Administrative Controls Section of TSs The licensee proposed including the SFCP and specific requirements into the CNS TSs, Section 5.5.14, as follows:

Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure that the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

The NRC staff concludes that the proposed TS 5.5.14, which requires an acceptable program to control surveillance frequencies to ensure that limiting conditions for operations are met and includes necessary program and applicability requirements, and its requirements set forth above are consistent with the model application of TSTF-425, and are therefore, acceptable.

3.3 Deviations and Variations from TSTF-425 and Other Changes 3.3.1 Variation in TS Surveillance Requirement Numbering The CNS TS SRs that have surveillance numbers identical to the corresponding NUREG-1433 SRs are not deviations from TSTF-425. The CNS TSs also contain SRs with numbers that differ from the corresponding NUREG-1433 SR numbers, although the SR requirements are identical or similar to those in NUREG-1433, Revision 4. The NRC staff reviewed these SRs and has determined that the SR numbering variations are editorial and non-substantive deviations from TSTF-425 with no impact on the NRC staff's model SE dated July 6, 2009 (74 FR 31996). Therefore, the NRC staff concludes that the deviations are acceptable.

3.3.2 Surveillance Requirement 3.3.1.2.4 In an RAI dated October 27, 2016 (Reference 20), the NRC staff questioned whether the entire frequency of SR 3.3.1.2.4 could be relocated to the licensee's proposed SFCP, since the first part of the frequency, "12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS," appeared to meet one of the four exclusion criteria of Part 2.0 of the approved TSTF-425, Revision 3, traveler as a

" ... frequency that is related to a specific condition .... " SR 3.3.1.2.4 concerns the state of the source range nuclear instruments that are monitored from the control room. In its response to

STSB RAl-1, by letter dated December 7, 2016 (Reference 2), the licensee proposed a new TS page 3.3-12 that revised only the second part of SR 3.3.1.2.4. Since the revised markup only changes a frequency that is periodic, it meets the criteria of the approved TSTF-425, Revision 3, traveler. Therefore, the NRC staff concludes that the proposed revision of SR 3.3.1.2.4 is acceptable.

3.3.3 Comparison of SR Changes between the CNS TSs and NUREG-1433 TSs The NRC staff noted that NUREG-1433 contains additional SRs that are not in the CNS TSs, and, therefore, the corresponding surveillances in TSTF-425 are not applicable to CNS. The NRC staff reviewed the variations and determined that, because the SRs do not apply to CNS, these deviations from TSTF-425 have no impact on the NRC staff's model SE dated July 6, 2009 (7 4 FR 31996) and are therefore acceptable.

The licensee also proposed changes to certain CNS plant-specific SRs that are not in NUREG-1433 and are not included in the associated changes provided in TSTF-425 to reflect relocation of the plant-specific surveillances that involve fixed periodic frequencies. The NRC staff reviewed the proposed plant-specific SRs and concludes that relocation of plant-specific SR frequencies is consistent with TSTF-425, Revision 3, and with the NRC staff's model SE dated July 6, 2009, including the scope exclusions identified in Section 1.0, "Introduction," of the model SE. Therefore, this variation is acceptable.

3.3.4 Proposed Format Change for SR 3.1.4.4 The licensee proposed a correction to the TS format on TS page 3.1-13. In the SR 3.1.4.4 FREQUENCY column, the licensee proposed that the word "AND" be underlined. The NRC staff evaluated this change and concludes that it is editorial and non-substantive in nature. It is also consistent with the existing format of CNS TS and is consistent with NUREG-1433, Revision 4. Therefore, the NRC staff concludes that this change is acceptable.

3.3.5 Surveillance Requirement 3.8.4.7 3.3.5.1 Licensee's Proposed Change The licensee proposed the following deletion in SR 3.8.4.7, as indicated by stricken text:

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

1. The modified performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7 once per 60 months.

2. This Surveillance shall not be performed in M()DE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.

Verify battery capacity is adequate to supply, and maintain in ()PERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.

Frequency: every 24 Months In accordance with the Surveillance Frequency Control Program The licensee provided the following justification for the change in Attachment 1, Subsection 2.2.3, of the LAR:

CNS TS SR 3.8.4.7, Note 1, is being revised to delete the frequency of "once per 60 months." This change is consistent with the latest revision of NUREG-1433.

CNS TS SR 3.8.4.7 reflects the surveillance requirement of STS SR 3.8.4.3 which contains a similar Note. However, the STS 3.8.4.3 Note 1 does not contain the reference to the Frequency of the modified performance discharge test. This change allows the modified performance discharge test in CNS TS SR 3.8.4.8 to be performed in lieu of the service test in CNS SR 3.8.4.7 at the Frequency established in the SFCP.

3.3.5.2 NRC Staff Evaluation of Proposed Change The NRC staff reviewed the licensee's proposed change in TS SR 3.8.4.7, Note 1, to delete the frequency of "once per 60 months" along with the justification provided in the LAR. The staff found that the licensee characterized the change as administrative in nature when, in fact, the language proposed to be deleted constitutes a technical change. Based on the regulation, 10 CFR 50.36(c)(3), SRs relate to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Therefore, a technical change in the SR must assure the level of quality of the batteries is maintained in order to support its safety function. The "once per 60 months" frequency is a part of the condition that would allow the licensee to perform a modified performance test in lieu of a service test.

By e-mail dated November 29, 2016 (Reference 24), the NRC staff requested that the licensee provide the technical basis for the change. Additionally, the staff requested confirmation that the modified performance discharge test completely encompasses the load profile of the battery service test, and that the intent of the service test is to verify the battery capacity to supply the design basis load profile.

By letter dated December 7, 2016 (Reference 3), the licensee provided the following response:

Consistent with Institute of Electrical and Electronics Engineers (IEEE) 450-1995

["IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Storage Batteries for Stationary Applications"], a modified performance discharge test encompasses the duty cycle of the service test.

Surveillance Requirements 3.8.4.7 and 3.8.4.8 specify that the service test and modified performance test are encompassed by the duty cycle requirements specified in design calculations. As stated in IEEE 450-1995 Section 5.4, "A modified performance test can be used in lieu of a service test at any time."

The modified performance discharge test has an initial load for one minute that exceeds the maximum load in design calculations followed by a load determined from the battery design that also exceeds the design basis load profile.

Therefore, either a service test or modified performance discharge test verify battery capacity to support the design basis load profile.

The licensee's Updated Safety Analysis Report, Section Vlll-6, Subsection 6.5, "Inspection and Testing, references IEEE Standard 450-1995, and therefore commits to the standard as a part of the CNS licensing basis. The NRC staff has endorsed IEEE Standard 450 through RG 1.129, "Maintenance, Testing, and Replacement of Vented Lead-Acid Storage Batteries for Nuclear Power Plants". Therefore, the NRC staff has found that the modified performance test can be used in lieu of a service test or a performance test with an exception noted in the RG.

Specifically, the exception in Regulatory Position 5 of the RG states that the modified performance test/service test should be performed with intervals not to exceed 24 months if the battery does not show signs of degradation. Since the licensee's response to the staff's RAI confirmed that a modified performance discharge test encompasses the duty cycle of the service test, the staff concludes that the response is acceptable.

Based on the evaluation discussed above, the NRC staff determined that the proposed change to SR 3.8.4.7 Note 1 regarding the technical change stemming from the deletion of the frequency of "once per 60 months" to the modified performance discharge test is consistent with the requirements in 10 CFR 50.36(c)(3). Therefore, the staff concludes that the proposed change is acceptable and consistent with NRC regulations.

3.3.6 Repagination of TS Pages The page numbers for TS pages 5.0-20 through 5.0-23 have been revised due to the repagination of the CNS TSs for this license amendment. The NRC staff reviewed these changes and concluded that they are formatting changes that do not alter any TS requirements and, therefore, are acceptable.

3.3.7 Additional Changes to TS Pages Revised by Previous Amendments Revised TS pages 3.4-20, 3.4-22, and 3.4-23 also include the revisions, unrelated to this license amendment, that were approved on July 25, 2016, in CNS Amendment No. 256, concerning the relocation of the reactor coolant system pressure-temperature limits from the TS limiting condition for operation to a licensee-controlled document (Reference 25). Also, TS page 1.1-5 includes a revision, unrelated to this license amendment, which was approved on July 25, 2016, in CNS Amendment No. 254 (Reference 26). These revisions were made by the NRC staff for administrative efficiency to reflect the TS changes associated with Amendment Nos. 254 and 256, which were issued while the TSTF-425 related changes were undergoing NRC staff review.

3.4 Technical Evaluation Summary The NRC staff has reviewed the licensee's proposed relocation of certain surveillance frequencies in TS Sections 3.1 through 3.10 to a licensee-controlled document, and controlling changes to surveillance frequencies in accordance with a new program, the SFCP, required the proposed addition of TS 5.5.14 to the Administrative Controls section of TSs. The NRC staff confirmed that this amendment does not relocate surveillance frequencies that reference other approved programs for the specific interval, are purely event-driven, are event-driven but have a time component for performing the surveillance on a one-time basis once the event occurs, or are related to specific conditions. The SFCP and TS Section 5.0, Subsection 5.5.14 references NEI 04-10, Revision 1 (Reference 5), which provides a risk-informed methodology using plant-specific risk insights and performance data to revise surveillance frequencies within the SFCP. This methodology supports relocating surveillance frequencies from TSs to a licensee-controlled document, provided those frequencies are changed in accordance with the

NRG-approved NEI 04-10, Revision 1, which is specified in the Administrative Controls section of the TSs (i.e., TS 5.5.14).

The licensee proposed to relocate specific surveillance frequencies from the following TS Sections to the SFCP:

3.1 Reactivity Control System 3.2 Power Distribution Limits 3.3 Instrumentation 3.4 Reactor Coolant System (RCS) 3.5 Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System 3.6 Containment Systems

3. 7 Plant Systems 3.8 Electrical Power Systems 3.9 Refueling Operations 3.10 Special Operations The licensee's proposed adoption of TSTF-425, Revision 3, and risk-informed methodology of NRG-approved NEI 04-10, Revision 1, as referenced in the Administrative Controls section of TSs, satisfies the key principles of risk-informed decisionmaking applied to changes to TSs, as delineated in RG 1.174 and RG 1.177, in that:

The proposed change meets current regulations;

The proposed change is consistent with defense-in-depth philosophy;

The proposed change maintains sufficient safety margins;

Increases in risk resulting from the proposed change are small and consistent with the Commission's Safety Goal Policy Statement; and

The impact of the proposed change is monitored with performance measurement strategies.

Paragraph 50.36(c) of 10 CFR discusses the categories that will be included in TSs.

Paragraph 50.36(c)(3) of 10 CFR discusses the specific category of SRs and states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The NRC staff finds that, with the proposed relocation of certain surveillance frequencies from the TS sections identified above to a licensee-controlled document and administratively controlled in accordance with the TS SFCP, the licensee continues to meet the requirements in 10 CFR 50.36(c)(3). For administrative efficiency, the NRC staff also revised some TSTF-425 related TS pages to reflect the changes associated with Amendment Nos. 254 and 256, which were issued while the TSTF-425 related changes were undergoing staff review. Based on the above, the proposed TS revisions to adopt TSTF-425, with the variations noted, are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment on March 2, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Registeron May 24, 2016 (81 FR 32807).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

1. Limpias, 0. A., Nebraska Public Power District, letter to U.S. Nuclear Regulatory Commission,

Subject:

"Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program, dated March 22, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16110A425).

2. Higginbotham, K., Nebraska Public Power District, letter to U.S. Nuclear Regulatory Commission,

Subject:

"Response to Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Revision 3, Cooper Nuclear Station, Docket No. 50-298, DPR-46, dated December 7, 2016 (ADAMS Accession No. ML16355A013).

3. Higginbotham, K., Nebraska Public Power District, letter to U.S. Nuclear Regulatory Commission,

Subject:

"Response to Nuclear Regulatory Commission Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Revision 3, Cooper Nuclear Station, Docket No. 50-298, DPR-46, dated December 7, 2016 (ADAMS Accession No. ML16355A010).

4. Technical Specifications Task Force, letter to U.S. Nuclear Regulatory Commission,

Subject:

Transmittal of TSTF, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," dated March 18, 2009 (ADAMS Package Accession No. ML090850642).

5. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession No. ML071360456).

6. Nieh, H. K., U. S. Nuclear Regulatory Commission, letter to Mr. Biff Bradley, Nuclear Energy Institute,

Subject:

"Final Safety Evaluation for Nuclear Energy Institute (NEI)

Topical Report (TR) 04-10, Revision 1, 'Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies (TAC No. MD6111),'"

dated September 19, 2007 (ADAMS Accession No. ML072570267).

7. U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Regulatory Guide 1.174, Revision 2, May 2011 (ADAMS Accession No. ML100910006).

8. U.S. Nuclear Regulatory Commission, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Regulatory Guide 1.177, Revision 1, May 2011 (ADAMS Accession No. ML100910008).

9. U.S. Nuclear Regulatory Commission, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Regulatory Guide 1.200, Revision 2, March 2009 (ADAMS Accession No. ML090410014).

10. U.S. Nuclear Regulatory Commission, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance,"

NUREG-0800, Section 19.2, June 2007 (ADAMS Accession No. ML071700658).

11. U.S. Nuclear Regulatory Commission, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load," NUREG-0800, Section 19.1, Revision 3, September 2012 (ADAMS Accession No. ML12193A107).

12. U.S. Nuclear Regulatory Commission, "Risk-Informed Decision Making: Technical Specifications," NUREG-0800, Section 16.1, Revision 1, March 2007 (ADAMS Accession No. ML070380228).

13. U.S. Nuclear Regulatory Commission, "Standard Technical Specifications, General Electric BWR/Plants, NUREG-1433, "Volume 1, Specifications and Volume 2, Bases, Revision 4, April 2012 (ADAMS Accession Nos. ML12104A192 and ML12104A193, respectively).

14. American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)

RA-Sa-2009, "Addenda to ASME RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,

February 2009, New York, NY.

15. NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance,"

Revision 1, May 2006 and NEI 00-02 Appendix D, "Self Assessment Process for Addressing ASME PRA Standard RA-Sb-2005, as endorsed by NRC Regulatory Guide 1.200," October 2006 (ADAMS Accession Nos. ML061510619 and ML063390593, respectively).

16. NEI 05-04 "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," NEI 05-04, Revision 2, November 2008 (ADAMS Accession No. ML083430462).

17. U.S. Nuclear Regulatory Commission, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Regulatory Guide 1.200, Revision 1, January 2007 (ADAMS Accession No. ML070240001 ).

18. American Society of Mechanical Engineers (ASME) PRA Standard ASME RA-Sc-2007 Addenda to, "ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated July 2007, New York, NY.

19. NEI 07-12, Revision 1, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," Nuclear Energy lnsitute, Revision 1, June 2010 (ADAMS Accession No. ML102230070).

20. Wengert, T., U.S. Nuclear Regulatory Commission, e-mail to Shaw, J. D., Nebraska Public Power District,

Subject:

"Cooper Nuclear Station - Formal Request for Additional Information Concerning License Amendment Request to Adopt TSTF-425 Revision 3 (CAC MF7498)," dated October 27, 2016 (ADAMS Accession No. ML16301A204).

21. U.S. Nuclear Regulatory Commission/Electric Power Research Institute, "EPRl/NRC-RES Fire Human Reliability Analysis Guidelines," NUREG-1921 (EPRI 1023001), Final Report, July 2012 (ADAMS Accession No. ML12216A104)

22. Sebrosky, J. M., U.S. Nuclear Regulatory Commission, letter to Mr. Oscar A. Limpias, Nebraska Public Power District,

Subject:

"Cooper Nuclear Station - Issuance of Amendment Regarding Transition to a Risk-informed, Performance-based Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC NO. ME8551)," dated April 29, 2014 (ADAMS Accession No. ML14055A023).

23. U.S. Nuclear Regulatory Commission and Electric Power Research lnsitute, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database," NUREG-2169 (EPRI 3002002936), January 2015 (ADAMS Accession No. ML15016A069).

24. Wengert, T., U.S. Nuclear Regulatory Commission, e-mail to Shaw, J. D., Nebraska Public Power District,

Subject:

"Cooper Nuclear Station - Formal Request for Additional Information Concerning License Amendment Request to Adopt TSTF-425 Revision 3 (CAC. MF7498)," dated November 29, 2016 (ADAMS Accession No. ML16335A015).

25. Wengert, T. J., U.S. Nuclear Regulatory Commission, letter to Mr. Oscar A. Limpias, Nebraska Public Power District,

Subject:

"Cooper Nuclear Station - Issuance of Amendment to Relocate the Pressure-Temperature Curves to a Pressure-Temperature Limits Report (CAC NO. MF6582)," dated July 27, 2016 (ADAMS Accession No. ML16158A022).

26. Wengert, T. J., U.S. Nuclear Regulatory Commission, letter to Mr. Oscar A. Limpias, Nebraska Public Power District,

Subject:

"Cooper Nuclear Station - Issuance of Amendment Re: Adoption of Technical Specification Task Force Change Traveler, TSTF-535, Revision 0 (CAC No. MF7442)," dated July 25, 2016 (ADAMS Accession No. ML16119A433).

Principal Contributors: A. Driver, NRR T. Martinez-Navedo, NRR P. Snyder, NRR Date: March 31, 2017.

ML17061A050; BWI: ML17061A059 *via memorandum **via e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC* NRR/DRA/APLA/BC*

NAME TWengert PBlechman AKlein SRosenberg DATE 3/17/17 3/13/17 2/3/17; reconcur 3/16/17 2/23/17 OFFICE NRR/DE/EICB/BC** NRR/DSS/SBPB/BC* NRR/DE/EPNB/BC* NRR/DE/ESGB/BC(A)*

NAME MWaters RDennig DAiiey AJohnson DATE 3/16/17 3/16/17 3/17/17 3/16/17 OFFICE NRR/DE/EEEB NRR/DE/EEEB NRR/DE/EEEB NRR/DE/EEEB NAME TMartinez-Navedo (non-concur) SRay (non-concur) RMathew (non-concur) SMatharu (non-concur)

DATE 3/16/17 3/16/17 3/16/17 3/16/17 OFFICE NRR/DE/EEEB/BC OGC NLO NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME JZimmerman (non-concur) MYoung w/revisions RPascarelli TWengert DATE 3/17/17 3/30/17 3/31/17 3/31/17

v t e Letter Sequence Approval
CAC:MF7498, (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions 27 October 2016 29 November 2016 Answers 7 December 2016 Results Approval Other: ML17079A256
MONTHYEAR2016-03-22 [Table View]

ML17061A050

Text

SUBJECT:

Dear Mr. Higginbotham:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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