Text

1 - Issuance of Amendment No. 227 TS Change Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program Based on TSTF-425
	ML21053A117
Person / Time
Site:	Wolf Creek
Issue date:	04/08/2021
From:	Samson Lee; Plant Licensing Branch IV
To:	Reasoner C; Wolf Creek
	Lee S, 301-415-3168
References
	EPID L-2020-LLA-0091
	Download: ML21053A117 (190)
	v • d • e

April 8, 2021 Mr. Cleveland Reasoner Chief Executive Officer and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION, UNIT 1 - ISSUANCE OF AMENDMENT NO. 227 RE: TECHNICAL SPECIFICATION CHANGE REGARDING RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM BASED ON TSTF-425 (EPID L-2020-LLA-0091)

Dear Mr. Reasoner:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 227 to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station, Unit 1. The amendment consists of changes to the technical specifications in response to your application dated April 27, 2020, as supplemented by letter dated October 26, 2020.

The amendment revises the technical specifications by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, Risk-Informed Technical Specification Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies. The changes are consistent with Technical Specifications Task Force (TSTF) Traveler (TSTF-425), Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-Informed TSTF] Initiative 5b.

C. Reasoner A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Samson S. Lee, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482

Enclosures:

1. Amendment No. 227 to NPF-42

2. Safety Evaluation cc: Listserv

WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 227 License No. NPF-42

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

A. The application for amendment to the Wolf Creek Generating Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated April 27, 2020, as supplemented by letter dated October 26, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, 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 Commissions regulations; D. The issuance of this 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 Commissions 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. NPF-42 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 227, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Digitally signed by Jennifer L. Dixon-Herrity Dixon-Herrity Date: 2021.04.08 13:03:18 -04'00' Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: April 8, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 227 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482 Replace the following pages of Renewed Facility Operating License No. NPF-42 and the 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.

Renewed Facility Operating License REMOVE INSERT 4 4 Technical Specifications Remove Insert Remove Insert i i 3.3-20 3.3-20 ii ii 3.3-21 3.3-21 1.1-6 1.1-6 3.3-22 3.3-22 3.1-1 3.1-1 3.3-23 3.3-23 3.1-3 3.1-3 3.3-24 3.3-24 3.1-10 3.1-10 3.3-25 3.3-25 3.1-12 3.1-12 3.3-26 3.3-26 3.1-15 3.1-15 3.3-27 3.3-27 3.1-20 3.1-20 3.3-28 3.3-28 3.1-22 3.1-22 3.3-29 3.3-29 3.2-3 3.2-3 3.3-30 3.3-30 3.2-5 3.2-5 3.3-31 3.3-31 3.2-8 3.2-8 3.3-32 3.3-32 3.2-9 3.2-9 3.3-33 3.3-33 3.2-13 3.2-13 3.3-34 3.3-34 3.3-10 3.3-10 3.3-35 3.3-35 3.3-11 3.3-11 3.3-36 3.3-36 3.3-12 3.3-12 3.3-37 3.3-37 3.3-13 3.3-13 3.3-38 3.3-38 3.3-14 3.3-14 3.3-39 3.3-39 3.3-15 3.3-15 3.3-40 3.3-40 3.3-16 3.3-16 3.3-41 3.3-41 3.3-17 3.3-17 3.3-42 3.3-42 3.3-18 3.3-18 3.3-43 3.3-43 3.3-19 3.3-19 3.3-44 3.3-44

Technical Specifications (continued)

Remove Insert Remove Insert 3.3-45 3.3-45 3.5-7 3.5-7 3.3-46 3.3-46 3.5-8 3.5-8 3.3-47 3.3-47 3.5-9 3.5-9 3.3-48 3.3-48 3.5-10 3.5-10 3.3-49 3.3-49 3.5-11 3.5-11 3.3-50 3.3-50 3.5-12 3.5-12 3.3-51 3.3-51 --- 3.5-13 3.3-52 3.3-52 3.6-6 3.6-6 3.3-53 3.3-53 3.6-11 3.6-11 3.3-54 3.3-54 3.6-12 3.6-12 3.3-55 3.3-55 3.6-13 3.6-13 3.3-56 3.3-56 3.6-14 3.6-14 3.3-57 3.3-57 3.6-15 3.6-15 3.3-58 3.3-58 3.6-17 3.6-17

--- 3.3-59 3.6-18 3.6-18

--- 3.3-60 3.6-19 3.6-19 3.4-3 3.4-3 3.6-20 3.6-20 3.4-4 3.4-4 3.6-21 3.6-21 3.4-5 3.4-5 3.7-7 3.7-7 3.4-7 3.4-7 3.7-8 3.7-8 3.4-8 3.4-8 3.7-12 3.7-12 3.4-11 3.4-11 3.7-14 3.7-14 3.4-13 3.4-13 3.7-16 3.7-16 3.4-14 3.4-14 3.7-17 3.7-17 3.4-16 3.4-16 3.7-19 3.7-19 3.4-17 3.4-17 3.7-21 3.7-21 3.4-19 3.4-19 3.7-23 3.7-23 3.4-21 3.4-21 3.7-25 3.7-25 3.4-26 3.4-26 3.7-28 3.7-28 3.4-30 3.4-30 3.7-31 3.7-31 3.4-31 3.4-31 3.7-35 3.7-35 3.4-33 3.4-33 3.7-36 3.7-36 3.4-36 3.4-36 3.7-38 3.7-38 3.4-37 3.4-37 3.7-40 3.7-40 3.4-41 3.4-41 3.7-43 3.7-43 3.4-43 3.4-43 3.7-45 3.7-45 3.5-2 3.5-2 3.7-47 3.7-47 3.5-3 3.5-3 3.8-7 3.8-7 3.5-4 3.5-4 3.8-8 3.8-8 3.5-5 3.5.5 3.8-9 3.8-9 3.5-6 3.5-6 3.8-10 3.8-10

Technical Specifications (continued)

Remove Insert Remove Insert 3.8-11 3.8-11 3.8-34 3.8-34 3.8-12 3.8-12 3.8-36 3.8-36 3.8-13 3.8-13 3.8-38 3.8-38 3.8-14 3.8-14 3.8-40 3.8-40 3.8-15 3.8-15 3.9-1 3.9-1 3.8-16 3.8-16 3.9-2 3.9-2 3.8-17 3.8-17 3.9-4 3.9-4 3.8-23 3.8-23 3.9-6 3.9-6 3.8-24 3.8-24 3.9-8 3.9-8 3.8-25 3.8-25 3.9-10 3.9-10 3.8-26 3.8-26 3.9-11 3.9-11 3.8-27 3.8-27 5.0-18 5.0-18 3.8-31 3.8-31 5.0-22 5.0-22 3.8-32 3.8-32 5.0-23 5.0-23

4 (5) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) The Operating Corporation, 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 the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations in 10 CFR Chapter I and 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 Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100%

power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 227, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Antitrust Conditions Kansas Gas & Electric Company and Kansas City Power & Light Company shall comply with the antitrust conditions delineated in Appendix C to this license.

(4) Environmental Qualification (Section 3.11, SSER #4, Section 3.11, SSER #5)*

Deleted per Amendment No. 141.

The parenthetical notation following the title of many license conditions denotes the section of the supporting Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-42 Amendment No. 227

TABLE OF CONTENTS 1.0 USE AND APPLICATION ................................................................................ 1.1-1 1.1 Definitions ................................................................................................ 1.1-1 1.2 Logical Connectors .................................................................................. 1.2-1 1.3 Completion Times .................................................................................... 1.3-1 1.4 Frequency ................................................................................................ 1.4-1 2.0 SAFETY LIMITS (SLs) .................................................................................... 2.0-1 2.1 SLs ................................................................................................... 2.0-1 2.2 SL Violations ............................................................................................ 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ............... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY .............................. 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS ....................................................... 3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) ......................................................... 3.1-1 3.1.2 Core Reactivity ................................................................................ 3.1-2 3.1.3 Moderator Temperature Coefficient (MTC) ...................................... 3.1-4 3.1.4 Rod Group Alignment Limits ............................................................ 3.1-7 3.1.5 Shutdown Bank Insertion Limits ...................................................... 3.1-11 3.1.6 Control Bank Insertion Limits ........................................................... 3.1-13 3.1.7 Rod Position Indication .................................................................... 3.1-16 3.1.8 PHYSICS TESTS Exceptions - MODE 2 ....................................... 3.1-19 3.2 POWER DISTRIBUTION LIMITS ............................................................ 3.2-1 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

(FQ Methodology) ....................................................................... 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNH) ............................ 3.2-6 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) ................................................... 3.2-9 3.2.4 QUADRANT POWER TILT RATIO (QPTR) .................................... 3.2-10 3.3 INSTRUMENTATION .............................................................................. 3.3-1 3.3.1 Reactor Trip System (RTS) Instrumentation .................................... 3.3-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrumentation .......................................................................... 3.3-22 3.3.3 Post Accident Monitoring (PAM) Instrumentation ............................ 3.3-38 3.3.4 Remote Shutdown System .............................................................. 3.3-42 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation .......................................................................... 3.3-45 Wolf Creek - Unit 1 i Amendment No. 123, 173, 183, 227

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6 Containment Purge Isolation Instrumentation ................................. 3.3-47 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation Instrumentation .......................................................... 3.3-51 3.3.8 Emergency Exhaust System (EES) Actuation Instrumentation .......................................................................... 3.3-56 3.4 REACTOR COOLANT SYSTEM (RCS) .................................................. 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ................................................... 3.4-1 3.4.2 RCS Minimum Temperature for Criticality ....................................... 3.4-5 3.4.3 RCS Pressure and Temperature (P/T) Limits .................................. 3.4-6 3.4.4 RCS Loops - MODES 1 and 2 ......................................................... 3.4-8 3.4.5 RCS Loops - MODE 3 ..................................................................... 3.4-9 3.4.6 RCS Loops - MODE 4 ..................................................................... 3.4-12 3.4.7 RCS Loops - MODE 5, Loops Filled ................................................ 3.4-15 3.4.8 RCS Loops - MODE 5, Loops Not Filled ......................................... 3.4-18 3.4.9 Pressurizer ...................................................................................... 3.4-20 3.4.10 Pressurizer Safety Valves ............................................................... 3.4-22 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) ...................... 3.4-24 3.4.12 Low Temperature Overpressure Protection (LTOP) System ........... 3.4-27 3.4.13 RCS Operational LEAKAGE............................................................ 3.4-32 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ................................. 3.4-34 3.4.15 RCS Leakage Detection Instrumentation ........................................ 3.4-38 3.4.16 RCS Specific Activity ....................................................................... 3.4-42 3.4.17 Steam Generator (SG) Tube Integrity ............................................. 3.4-44 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ............................. 3.5-1 3.5.1 Accumulators ................................................................................... 3.5-1 3.5.2 ECCS - Operating............................................................................ 3.5-4 3.5.3 ECCS - Shutdown ........................................................................... 3.5-7 3.5.4 Refueling Water Storage Tank (RWST) .......................................... 3.5-9 3.5.5 Seal Injection Flow .......................................................................... 3.5-11 3.6 CONTAINMENT SYSTEMS .................................................................... 3.6-1 3.6.1 Containment .................................................................................... 3.6-1 3.6.2 Containment Air Locks .................................................................... 3.6-2 3.6.3 Containment Isolation Valves .......................................................... 3.6-7 3.6.4 Containment Pressure ..................................................................... 3.6-14 3.6.5 Containment Air Temperature ......................................................... 3.6-15 3.6.6 Containment Spray and Cooling Systems ....................................... 3.6-16 3.6.7 Spray Additive System .................................................................... 3.6-20 Wolf Creek - Unit 1 ii Amendment No. 123, 131, 157, 164, 167 170, 183, 212, 227

Definitions 1.1 1.1 Definitions (continued)

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include, a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

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

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST (TADOT) and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

Wolf Creek - Unit 1 1.1-6 Amendment No. 123, 170, 221, 227

SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be within the limit provided in the COLR.

APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore 15 minutes SDM to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within limit. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.1-1 Amendment No. 123, 155, 227

Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 -----------------------------NOTE---------------------------------

The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within +/- 1% k/k of Once prior to predicted values. entering MODE 1 after each refueling AND

NOTE--------

Only required after 60 EFPD In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.1-3 Amendment No. 123, 227

Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit. In accordance with the Surveillance Frequency Control Program SR 3.1.4.2 Verify rod freedom of movement (trippability) by In accordance with moving each rod not fully inserted in the core the Surveillance 10 steps in either direction. Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully Prior to reactor withdrawn position, is 2.7 seconds from the criticality after beginning of decay of stationary gripper coil voltage each removal of to dashpot entry, with: the reactor head

a. Tavg 500°F; and

b. All reactor coolant pumps operating.

Wolf Creek - Unit 1 3.1-10 Amendment No. 123, 227

Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the limits In accordance with specified in the COLR. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.1-12 Amendment No. 123, 227

Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.6.2 Verify each control bank insertion is within the limits In accordance with specified in the COLR. the Surveillance Frequency Control Program SR 3.1.6.3 Verify sequence and overlap limits specified in the In accordance with COLR are met for control banks not fully withdrawn the Surveillance from the core. Frequency Control Program Wolf Creek - Unit 1 3.1-15 Amendment No. 123, 227

PHYSICS TESTS Exceptions - MODE 2 3.1.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. RCS lowest operating loop C.1 Restore RCS lowest 15 minutes average temperature not operating loop average within limit. temperature to within limit.

D. Required Action and D.1 Be in MODE 3. 15 minutes associated Completion Time of Condition C not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on Prior to initiation of power range and intermediate range channels per PHYSICS TESTS SR 3.3.1.8 and Table 3.3.1-1.

SR 3.1.8.2 Verify the RCS lowest operating loop average In accordance with temperature is 541°F. the Surveillance Frequency Control Program SR 3.1.8.3 Verify THERMAL POWER is 5% RTP. In accordance with the Surveillance Frequency Control Program SR 3.1.8.4 Verify SDM is within limits provided in the COLR. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.1-20 Amendment No. 123, 227

RCS Boron Limitations < 500°F 3.1.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.9.1 Verify RCS boron concentration is greater than the In accordance with ARO critical boron concentration. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.1-22 Amendment No. 221, 227

FQ(Z) (FQ Methodology) 3.2.1 SURVEILLANCE REQUIREMENTS

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

During power escalation following shutdown, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution measurement is obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FQC(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FQC(Z) was last verified AND In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.2-3 Amendment No. 123, 188, 227

FQ(Z) (FQ Methodology) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued) Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by 10% RTP, the THERMAL POWER at which FQW(Z) was last verified AND In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.2-5 Amendment No. 123, 227

FN H 3.2.2 SURVEILLANCE REQUIREMENTS

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

During power escalation following shutdown, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution measurement is obtained.

SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FNH is within limits specified in the COLR. Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.2-8 Amendment No. 123, 131, 188, 227

AFD (RAOC Methodology) 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)

Methodology)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

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

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY: MODE 1 with THERMAL POWER 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL 30 minutes POWER to < 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore In accordance with channel. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.2-9 Amendment No. 123, 227

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 ----------------------------NOTES--------------------------------

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

Verify QPTR is within limit by calculation. In accordance with the Surveillance Frequency Control Program SR 3.2.4.2 ------------------------------NOTE--------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after input from one Power Range Neutron Flux channel is inoperable with THERMAL POWER > 75% RTP.

Verify QPTR is within limit using core power In accordance with distribution measurement information. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.2-13 Amendment No. 123, 188, 227

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME X. Required Action and X.1.1 Initiate action to fully insert Immediately associated Completion all rods.

Time of Condition W not met. AND OR X.1.2 Initiate action to place the Immediately Rod Control System in a Two or more channels condition incapable of rod inoperable. withdrawal.

OR X.2 Initiate action to borate the Immediately RCS to greater than all rods out (ARO) critical boron concentration.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-10 Amendment No. 123, 148, 156, 188, 221, 227

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.2 -----------------------------NOTES--------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 15% RTP.

Compare results of calorimetric heat balance In accordance with calculation to power range channel output. Adjust the Surveillance power range channel output if calorimetric heat Frequency Control balance calculation results exceed power range Program channel output by more than + 2% RTP.

SR 3.3.1.3 -----------------------------NOTES--------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 50% RTP.

Compare results of the core power distribution In accordance with measurements to Nuclear Instrumentation System the Surveillance (NIS) AFD. Adjust NIS channel if absolute difference Frequency Control is 3%. Program SR 3.3.1.4 ------------------------------NOTE---------------------------------

This Surveillance must be performed on the reactor trip bypass breaker for the local manual shunt trip only prior to placing the bypass breaker in service.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.1.5 Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-11 Amendment No. 123, 156, 188, 221, 227

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.6 ------------------------------NOTE---------------------------------

Not required to be performed until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after achieving equilibrium conditions with THERMAL POWER 75 % RTP.

Calibrate excore channels to agree with core power In accordance with distribution measurements. the Surveillance Frequency Control Program SR 3.3.1.7 ------------------------------NOTES-------------------------------

1. Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3.

2. Source range instrumentation shall include verification that interlocks P-6 and P-10 are in their required state for existing unit conditions.

Perform COT. In accordance with the Surveillance Frequency Control Program SR 3.3.1.8 ------------------------------NOTE---------------------------------

This Surveillance shall include verification that interlocks P-6 and P-10 are in their required state for existing unit conditions.

Perform COT. --------NOTE--------

Only required when not performed within the Frequency specified in the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-12 Amendment No. 123, 156, 227

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.8 (continued Prior to reactor startup AND Twelve hours after reducing power below P-10 for power and intermediate instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND In accordance with the Surveillance Frequency Control Program SR 3.3.1.9 ------------------------------NOTE---------------------------------

Verification of setpoint is not required.

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

Wolf Creek - Unit 1 3.3-13 Amendment No. 123, 227

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.10 ------------------------------NOTE---------------------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.11 -----------------------------NOTES--------------------------------

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

2. This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

3. Power and intermediate range detector plateau voltage verification is not required to be performed until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after achieving equilibrium conditions with THERMAL POWER 95% RTP.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.12 Not Used.

SR 3.3.1.13 Perform COT. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-14 Amendment No. 123, 227

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.14 -----------------------------NOTE----------------------------------

Verification of setpoint is not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.1.15 -----------------------------NOTE----------------------------------

Verification of setpoint is not required.

Perform TADOT. Prior to exceeding the P-9 interlock whenever the unit has been in MODE 3, if not performed in the previous 31 days SR 3.3.1.16 ------------------------------NOTE---------------------------------

Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIMES are within limits. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-15 Amendment No.

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 6)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

1. Manual Reactor Trip 1,2 2 B SR 3.3.1.14 NA 3(b), 4(b), 5(b) 2 C SR 3.3.1.14 NA

2. Power Range Neutron Flux

a. High 1,2 4 D SR 3.3.1.1 112.3% RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16

b. Low 1(c), 2(f) 4 V SR 3.3.1.1 28.3% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 2(h), 3(i) 4 W, X SR 3.3.1.1 28.3% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16

3. Power Range Neutron Flux Rate

a. High Positive Rate 1,2 4 E SR 3.3.1.7 6.3% RTP SR 3.3.1.11 with time SR 3.3.1.16 constant 2 sec

b. High Negative 1,2 4 E SR 3.3.1.7 6.3% RTP with Rate SR 3.3.1.11 time constant SR 3.3.1.16 2 sec

4. Intermediate Range 1(c), 2(d) 2 F,G SR 3.3.1.1 35.3% RTP Neutron Flux SR 3.3.1.8 SR 3.3.1.11 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(c) Below the P-10 (Power Range Neutron Flux) interlock.

(d) Above the P-6 (Intermediate Range Neutron Flux) interlock.

(f) With keff 1.0.

(h) With keff < 1.0, and all RCS cold leg temperatures 500o F, and RCS boron concentration the rods out (ARO) critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(i) With all RCS cold leg temperatures 500o F, and RCS boron concentration the ARO critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

Wolf Creek - Unit 1 3.3-16 Amendment No. 123, 140, 221, 227

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 6)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

5. Source Range Neutron 2(e) 2 I,J SR 3.3.1.1 d 1.6 E5 cps Flux SR 3.3.1.8 SR 3.3.1.11 3(b), 4(b), 5(b) 2 J,K SR 3.3.1.1 d 1.6 E5 cps SR 3.3.1.7 SR 3.3.1.11

6. Overtemperature 'T 1,2 4 E SR 3.3.1.1 Refer to Note 1 SR 3.3.1.3 (Page 3.3-)

SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

7. Overpower 'T 1,2 4 E SR 3.3.1.1 Refer to SR 3.3.1.7 Note 2 SR 3.3.1.10 (Page SR 3.3.1.16 3.3-2)

8. Pressurizer Pressure

a. Low 1(g) 4 M SR 3.3.1.1 t 1930 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. High 1,2 4 E SR 3.3.1.1 d 2395 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

9. Pressurizer Water 1(g) 3 M SR 3.3.1.1 d 93.9% of Level - High SR 3.3.1.7 instrument span SR 3.3.1.10

10. Reactor Coolant Flow - 1(g) 3 per loop M SR 3.3.1.1 t 88.9% of Low SR 3.3.1.7 normalized SR 3.3.1.10 flow SR 3.3.1.16 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(e) Below the P-6 (Intermediate Range Neutron Flux) interlock.

(g) Above the P-7 (Low Power Reactor Trips Block) interlock.

Wolf Creek - Unit 1 3.3-17 Amendment No. 123, 221,

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 6)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

11. Not Used.

12. Undervoltage 1(g) 2/bus M SR 3.3.1.9 10355 Vac RCPs SR 3.3.1.10 SR 3.3.1.16

13. Underfrequency 1(g) 2/bus M SR 3.3.1.9 57.1 Hz RCPs SR 3.3.1.10 SR 3.3.1.16

14. Steam Generator (SG) 1,2 4 per E SR 3.3.1.1 > 22.3% of Water Level gen SR 3.3.1.7 Narrow Range Low-Low (l) SR 3.3.1.10 Instrument SR 3.3.1.16 Span

15. Not Used.

16. Turbine Trip

a. Low Fluid Oil 1(j) 3 O SR 3.3.1.10 534.20 psig Pressure SR 3.3.1.15

b. Turbine Stop 1(j) 4 P SR 3.3.1.10 1% open Valve Closure SR 3.3.1.15

17. Safety Injection (SI) 1,2 2 trains Q SR 3.3.1.14 NA Input from Engineered Safety Feature Actuation System (ESFAS)

18. Reactor Trip System Interlocks

a. Intermediate 2(e) 2 S SR 3.3.1.11 6E-11 amp Range Neutron SR 3.3.1.13 Flux, P-6

b. Low Power 1 1 per T SR 3.3.1.5 NA Reactor Trips train Block, P-7

c. Power Range 1 4 T SR 3.3.1.11 51.3% RTP Neutron Flux, P-8 SR 3.3.1.13 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(e) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(g) Above the P-7 (Low Power Reactor Trips Block) interlock.

(l) The applicable MODES for these channels are more restrictive in Table 3.3.2-1. (See Function 6.d.)

(j) Above the P-9 (Power Range Neutron Flux) interlock.

Wolf Creek - Unit 1 3.3-18 Amendment No. 123, 132, 227

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 6)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

18. (continued)

d. Power Range 1 4 T SR 3.3.1.11 53.3% RTP Neutron Flux, P-9 SR 3.3.1.13

e. Power Range 1,2 4 S SR 3.3.1.11 6.7% RTP Neutron Flux, SR 3.3.1.13 and 13.3%

P-10 RTP

f. Turbine Impulse 1 2 T SR 3.3.1.10 12.4% turbine Pressure, P-13 SR 3.3.1.13 power

19. Reactor Trip 1,2 2 trains R SR 3.3.1.4 NA Breakers (RTB) (k) 3(b), 4(b), 5(b) 2 trains C SR 3.3.1.4 NA

20. Reactor Trip Breaker 1,2 1 each per U SR 3.3.1.4 NA Undervoltage and RTB Shunt Trip Mechanisms (k) 3(b), 4(b), 5(b) 1 each per C SR 3.3.1.4 NA RTB

21. Automatic Trip Logic 1,2 2 trains Q SR 3.3.1.5 NA 3(b), 4(b), 5(b) 2 trains C SR 3.3.1.5 NA (a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(k) Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.

Wolf Creek - Unit 1 3.3-19 Amendment No. 123, 144, 227

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 6)

Reactor Trip System Instrumentation Note 1: Overtemperature T The Overtemperature T Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.3% of T span.

(1 + 1 s) 1 (1 + 4 s) 1 T T O K1 - K 2 T - T + K 3 (P - P ) - f 1 ( I)

(1 + 2 s) 1 + 3 s (1 + 5 s) (1 + 6 s)

Where: T is measured RCS T, °F.

T0 is the indicated T at RTP, °F.

s is the Laplace transform operator, sec-1.

T is the measured RCS average temperature, °F.

T is the nominal Tavg at RTP, *.

P is the measured pressurizer pressure, psig.

P is the nominal RCS operating pressure

psig.

K1 =

K2 = * /°F K3 = * /psig 1 =

sec 2 =

sec 3 =

sec 4 =

sec 5 =

sec 6 =

sec f1(I) = * { * % + (qt - qb)} when qt - qb < * % RTP 0% of RTP when * % RTP qt - qb * % RTP

{(qt - qb) - * % } when qt - qb > * % RTP where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

The values denoted with

are specified in the COLR.

Wolf Creek - Unit 1 3.3-20 Amendment No. 123, 144, 227

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)

Reactor Trip System Instrumentation Note 2: Overpower T The Overpower T Function Allowable Value shall not exceed the following Trip Setpoint by more than 2.6% of T span.

(1 + 1 s) 1 ( 7 s) 1 1 T TO K4 - K5 T - K6 T - T - f 2 ( I)

(1 + 2 s) 1 + 3 s (1 + 7 s) 1 + 6 s (1 + 6 s)

Where: T is measured RCS T, °F.

T0 is the indicated T at RTP, °F.

s is the Laplace transform operator, sec-1.

T is the measured RCS average temperature, °F.

T is the indicated Tavg at RTP (Calibration temperature for T instrumentation), * °F.

K4 =

K5 = * /°F for increasing Tavg K6 = * /°F when T > T

/°F for decreasing Tavg * /°F when T T 1 =

sec 2 =

sec 3 =

sec 6 =

sec 7 =

sec f2(I) =

The values denoted with

are specified in the COLR.

Wolf Creek - Unit 1 3.3-21 Amendment No. 123, 227

ESFAS Instrumentation 3.3.2 3.3 INSTRUMENTATION 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation LCO 3.3.2 The ESFAS instrumentation for each Function in Table 3.3.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.2-1.

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Enter the Condition Immediately with one or more required referenced in Table channels or trains 3.3.2-1 for the channel(s) inoperable. or train(s).

B. One channel or train B.1 Restore channel or train 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months inoperable. to OPERABLE status.

OR B.2.1 Be in MODE 3. 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months AND B.2.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months (continued)

Wolf Creek - Unit 1 3.3-22 Amendment No. 123, 156, 227

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One train inoperable. -------------------NOTE---------------------

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing provided the other train is OPERABLE.

C.1 ------------NOTE---------------

Only required if Function 3.a.(2) is inoperable.

Place and maintain Immediately containment purge supply and exhaust valves in closed position.

AND C.2 Restore train to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

OR C.3.1 Be in MODE 3. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months AND C.3.2 Be in MODE 5. 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months (continued)

Wolf Creek - Unit 1 3.3-23 Amendment No. 123, 156, 227

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One channel inoperable. -------------------NOTE---------------------

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of other channels.

D.1 Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR D.2.1 Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months AND D.2.2 Be in MODE 4. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months E. One Containment Pressure -------------------NOTE---------------------

channel inoperable. One additional channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

E.1 Place channel in bypass. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR E.2.1 Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months AND E.2.2 Be in MODE 4. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months (continued)

Wolf Creek - Unit 1 3.3-24 Amendment No. 123, 156, 227

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. One channel or train F.1 Restore channel or train to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months inoperable. OPERABLE status.

OR F.2.1 Be in MODE 3. 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months AND F.2.2 Be in MODE 4. 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months G. One train inoperable. -------------------NOTE---------------------

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing provided the other train is OPERABLE.

G.1 Restore train to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status OR G.2.1 Be in MODE 3. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months AND G.2.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

Wolf Creek - Unit 1 3.3-25 Amendment No. 123, 156,175, 227

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. Not Used.

I. One channel inoperable. ------------------NOTE----------------------

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing of other channels.

I.1 Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR I.2 Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months (continued)

Wolf Creek - Unit 1 3.3-26 Amendment No. 123, 156, 187, 227

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME J. One or more Main --------------------NOTE--------------------

Feedwater Pump trip One inoperable channel may be channel(s) inoperable. bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other channels.

J.1 Place channel(s) in trip. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OR J.2 Be in MODE 3. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months K. One channel inoperable. -------------------NOTE---------------------

One additional channel may be tripped for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

K.1 Place channel in bypass. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OR K.2.1 Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months AND K.2.2 Be in MODE 5. 108 hours0.00125 days 0.03 hours 1.785714e-4 weeks 4.1094e-5 months (continued)

Wolf Creek - Unit 1 3.3-27 Amendment No. 123, 132, 155, 227

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME L. One or more required L.1 Verify interlock is in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months channel(s) inoperable. required state for existing unit condition.

OR L.2.1 Be in MODE 3. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months AND L.2.2 Be in MODE 4. 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months M. One channel inoperable. M.1 Place channel in trip. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND M.2 Restore channel to During OPERABLE status. performance of next COT N. One train inoperable. -------------------NOTE---------------------

One train may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other train is OPERABLE.

N.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND N.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued)

Wolf Creek - Unit 1 3.3-28 Amendment No. 123,156, 227

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME O. One or more channels O.1 Declare associated Immediately inoperable. auxiliary feedwater pump(s) inoperable.

P. One or both train(s) P.1 Restore train(s) to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months inoperable. OPERABLE status.

OR P.2.1 Be in MODE 3. 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months AND P.2.2 Be in MODE 4. 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.

SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.2.2 Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-29 Amendment No. 123, 131, 156, 183, 227

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.3 -----------------------------NOTE--------------------------------

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.4 Perform MASTER RELAY TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.5 Perform COT. In accordance with the Surveillance Frequency Control Program SR 3.3.2.6 Perform SLAVE RELAY TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.7 -------------------------------NOTE------------------------------

Verification of relay setpoints not required.

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

Wolf Creek - Unit 1 3.3-30 Amendment No. 123, 131, 183, 227

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.8 -----------------------------NOTE--------------------------------

Verification of setpoint not required for manual initiation functions.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.2.9 -----------------------------NOTE--------------------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.2.10 -----------------------------NOTE--------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after SG pressure is 900 psig.

Verify ESF RESPONSE TIMES are within limits. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-31 Amendment No. 123, 140, 183, 227

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.11 -----------------------------NOTE--------------------------------

Verification of setpoint not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.2.12 Perform COT. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-32 Amendment No. 123, 175, 183, 184, 227

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

1. Safety Injection

a. Manual Initiation 1,2,3,4 2 B SR 3.3.2.8 NA

b. Automatic Actuation 1,2,3,4 2 trains C SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

c. Containment Pressure - 1,2,3 3 D SR 3.3.2.1 d 4.5 psig High 1 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

d. Pressurizer Pressure - 1,2,3(b) 4 D SR 3.3.2.1 t 1820 psig Low SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

e. Steam Line Pressure 1,2,3(b) 3 per steam D SR 3.3.2.1 t 571 psig(c)

Low line SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

2. Containment Spray

a. Manual Initiation 1,2,3,4 2 per train, B SR 3.3.2.8 NA 2 trains

b. Automatic Actuation 1,2,3,4 2 trains C SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

c. Containment 1,2,3 4 E SR 3.3.2.1 d 28.3 psig Pressure SR 3.3.2.5 High - 3 SR 3.3.2.9 SR 3.3.2.10 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b) Above the P-11 (Pressurizer Pressure) interlock and below P-11 unless the Function is blocked.

(c) Time constants used in the lead/lag controller are t1 t 50 seconds and t2 d 5 seconds.

Wolf Creek - Unit 1 3.3-33 Amendment No. 123, 1, 1,

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

3. Containment Isolation

a. Phase A Isolation (1) Manual Initiation 1,2,3,4 2 B SR 3.3.2.8 NA (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays (3) Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

b. Phase B Isolation (1) Manual Initiation 1,2,3,4 2 per train, B SR 3.3.2.8 NA 2 trains (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays (3) Containment 1,2,3 4 E SR 3.3.2.1 28.3 psig Pressure - SR 3.3.2.5 High 3 SR 3.3.2.9 SR 3.3.2.10

4. Steam Line Isolation

a. Manual Initiation 1,2(i), 3(i) 2 F SR 3.3.2.8 NA

b. Automatic Actuation 1,2(i), 3(i) 2 trains G SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays (SSPS) SR 3.3.2.6

c. Automatic Actuation 1,2(l), 3(l) 2 trains G SR 3.3.2.6 NA Logic (MSFIS)

d. Containment Pressure 1,2(i), 3(i) 3 D SR 3.3.2.1 18.3 psig

- High 2 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(i) Except when all MSIVs are closed and de-activated; and all MSIV bypass valves are closed and de-activated, or closed and isolated by a closed manual valve, or isolated by two closed manual valves.

(l) Except when all MSIVs are closed and de-activated.

Wolf Creek - Unit 1 3.3-34 Amendment No. 123, 136, 183, 227

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

4. Steam Line Isolation (continued)

e. Steam Line Pressure (1) Low 1,2(i),3(b)(i) 3 per steam D SR 3.3.2.1 571 psig(c) line SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10 (2) Negative Rate - 3(g)(i) 3 per steam D SR 3.3.2.1 125(h) psi High line SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

5. Turbine Trip and Feedwater Isolation

a. Automatic Actuation 1,2(j),3(j) 2 trains G SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays (SSPS) SR 3.3.2.6

b. Automatic Actuation 1,2(k),3(k) 2 trains G SR 3.3.2.6 NA Logic (MSFIS)

c. SG Water Level -High 1,2(j) 4 per SG I SR 3.3.2.1 79.7% of High (P-14) SR 3.3.2.5 Narrow Range SR 3.3.2.9 Instrument Span SR 3.3.2.10

d. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

(continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b) Above the P-11 (Pressurizer Pressure) Interlock and below P-11 unless the Function is blocked.

(c) Time constants used in the lead/lag controller are t1 50 seconds and t2 5 seconds.

(g) Below the P-11 (Pressurizer Pressure) Interlock; however, may be blocked below P-11 when safety injection on low steam line pressure is not blocked.

(h) Time constant utilized in the rate/lag controller is 50 seconds.

(i) Except when all MSIVs are closed and de-activated; and all MSIV bypass valves are closed and de-activated, or closed and isolated by a closed manual valve, or isolated by two closed manual valves.

(j) Except when all MFIVs are closed and de-activated; and all MFRVs are closed and de-activated or closed and isolated by a closed manual valve; and all MFRV bypass valves are closed and de-activated, or closed and isolated by a closed manual valve, or isolated by two closed manual valves.

(k) Except when all MFIVs are closed and de-activated.

Wolf Creek - Unit 1 3.3-35 Amendment No. 123, 126, 132, 183, 194, 227

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (a)

6. Auxiliary Feedwater

a. Manual Initiation 1,2,3 1 per pump O SR 3.3.2.8 NA

b. Automatic Actuation 1,2,3 2 trains G SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays (Solid State SR 3.3.2.6 Protection System)

c. Automatic Actuation 1,2,3 2 trains N SR 3.3.2.3 NA Logic and Actuation Relays (Balance of Plant ESFAS)

d. SG Water Level Low - 1,2,3 4 per SG D SR 3.3.2.1 22.3% of Low SR 3.3.2.5 Narrow Range SR 3.3.2.9 Instrument Span SR 3.3.2.10

e. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

f. Loss of Offsite Power 1,2,3 2 trains P SR 3.3.2.7 NA SR 3.3.2.10

g. Trip of all Main 1 2 per pump J SR 3.3.2.8 NA Feedwater Pumps

h. Auxiliary Feedwater 1,2,3 3 M SR 3.3.2.1 20.53 psia Pump Suction SR 3.3.2.9 Transfer on Suction SR 3.3.2.10 Pressure - Low SR 3.3.2.12 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

Wolf Creek - Unit 1 3.3-36 Amendment No. 123, 155, 183, 227

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE(a)

7. Automatic Switchover to Containment Sump

a. Automatic Actuation 1,2,3,4 2 trains C SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

b. Refueling Water 1,2,3,4 4 K SR 3.3.2.1 35.5% of Storage Tank (RWST) SR 3.3.2.5 instrument span Level - Low Low SR 3.3.2.9 SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection

8. ESFAS Interlocks

a. Reactor Trip, P-4(m) 1,2,3 2 per train, F SR 3.3.2.11 NA 2 trains

b. Pressurizer Pressure, 1,2,3 3 L SR 3.3.2.5 1979 psig P-11 SR 3.3.2.9 (a) The Allowable Value defines the Limiting Safety System Settings. See the Bases for the Trip Setpoints.

(m) The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trips the main turbine - MODES 1 and 2

Isolates MFW with coincident low Tavg - MODES 1 and 2

Allows manual block of the automatic reactuation of SI after a manual reset of SI - MODES 1, 2, and 3

Prevents opening of MFIVs if closed on SI or SG Water Level - High High - MODES 1, 2, and 3 Wolf Creek - Unit 1 3.3-37 Amendment No. 123, 157, 183, 227

PAM Instrumentation 3.3.3 3.3 INSTRUMENTATION 3.3.3 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3 The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required channel 30 days with one required channel to OPERABLE status.

inoperable.

B. Required Action and B.1 Initiate action in Immediately associated Completion accordance with Time of Condition A not Specification 5.6.8.

met.

(continued)

Wolf Creek - Unit 1 3.3-38 Amendment No. 123, 183, 227

PAM Instrumentation 3.3.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore all but one 7 days with two or more required channel to OPERABLE channels inoperable. status.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Table 3.3.3-1 Time of Condition C not for the channel.

met.

E. As required by Required E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.3-1. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months F. As required by Required F.1 Initiate action in Immediately Action D.1 and referenced accordance with in Table 3.3.3-1. Specification 5.6.8.

Wolf Creek - Unit 1 3.3-39 Amendment No. 123, 157, 183, 227

PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS

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

SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.

SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required In accordance with instrumentation channel that is normally energized. the Surveillance Frequency Control Program SR 3.3.3.2 -------------------------------NOTE--------------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-40 Amendment No. 123,155, 183, 227

PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 1)

Post Accident Monitoring Instrumentation CONDITION REFERENCED FROM REQUIRED FUNCTION REQUIRED CHANNELS ACTION D.1

1. Neutron Flux 2 E

2. Reactor Coolant System (RCS) Hot Leg Temperature 2 E (Wide Range)

3. RCS Cold Leg Temperature (Wide Range) 2 E

4. RCS Pressure (Wide Range) 2 E

5. Reactor Vessel Water Level 2 F

6. Containment Normal Sump Water Level 2 E

7. Containment Pressure ( Normal Range) 2 E

8. Steam Line Pressure 2 per E steam generator

9. Containment Radiation Level (High Range) 2 F

10. Not Used

11. Pressurizer Water Level 2 E

12. Steam Generator Water Level (Wide Range) 4 E

13. Steam Generator Water Level (Narrow Range) 2 per E steam generator

14. Core Exit Temperature - Quadrant 1 2(a) E

15. Core Exit Temperature - Quadrant 2 2(a) E

16. Core Exit Temperature - Quadrant 3 2(a) E

17. Core Exit Temperature - Quadrant 4 2(a) E

18. Auxiliary Feedwater Flow Rate 4 E

19. Refueling Water Storage Tank Level 2 E (a) A channel consists of two core exit thermocouples (CETs).

Wolf Creek - Unit 1 3.3-41 Amendment No. 123,183, 227

Remote Shutdown System 3.3.4 3.3 INSTRUMENTATION 3.3.4 Remote Shutdown System LCO 3.3.4 The Remote Shutdown System Functions in Table 3.3.4-1 and the required auxiliary shutdown panel (ASP) controls shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

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

Separate Condition entry is allowed for each Function and required ASP control.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Function 30 days Functions inoperable. and required ASP controls to OPERABLE status.

OR One or more required ASP controls inoperable.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Wolf Creek - Unit 1 3.3-42 Amendment No. 123, 183, 227

Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required In accordance with instrumentation channel that is normally energized. the Surveillance Frequency Control Program SR 3.3.4.2 Verify each required auxiliary shutdown panel control In accordance with circuit and transfer switch is capable of performing the Surveillance the intended function. Frequency Control Program SR 3.3.4.3 -----------------------------NOTES-------------------------------

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

2. Reactor Trip Breakers and RCP breakers are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION for each required In accordance with instrumentation channel. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-43 Amendment No. 123, 183, 227

Remote Shutdown System 3.3.4 Table 3.3.4-1 (page 1 of 1)

Remote Shutdown System Functions FUNCTION REQUIRED CHANNELS

1. Source Range Neutron Fluxa 1

2. Reactor Trip Breaker Position 1 per trip breaker

3. Pressurizer Pressure 1

4. RCS Wide Range Pressure 1

5. RCS Hot Leg Temperature 1

6. RCS Cold Leg Temperature 1

7. SG Pressure 1 per SG

8. SG Level 1 per SG

9. AFW Flow Rate 1

10. RCP Breakers 1 per pump

11. AFW Suction Pressure 1

12. Pressurizer Level 1

a. Not required OPERABLE in MODE 1 or in MODE 2 above the P-6 setpoint.

Wolf Creek - Unit 1 3.3-44 Amendment No. 123,128,183,224, 227

LOP DG Start Instrumentation 3.3.5 3.3 INSTRUMENTATION 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation LCO 3.3.5 Four channels per 4-kV NB bus of the loss of voltage Function and four channels per 4-kV NB bus of the degraded voltage Function shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4, When associated DG is required to be OPERABLE by LCO 3.8.2, "AC Sources - Shutdown."

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions --------------------NOTE--------------------

with one channel per bus The inoperable channel may be inoperable. bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

A.1 Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One or more Functions B.1 Declare associated load Immediately with two or more channels shedder and emergency per bus inoperable. load sequencer (LSELS) inoperable.

OR Required Action and associated Completion Time of Condition A not met.

Wolf Creek - Unit 1 3.3-45 Amendment No. 123, 183, 227

LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1 Not Used.

SR 3.3.5.2 -------------------------------NOTE--------------------------------

Verification of time delays is not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.5.3 Perform CHANNEL CALIBRATION with nominal Trip In accordance with Setpoint and Allowable Value as follows: the Surveillance Frequency Control

a. Loss of voltage Allowable Value 90.0V, 120V bus Program with a time delay of 1.0 + 0.15, -0.1 sec.

Loss of voltage nominal Trip Setpoint 91.28V, 120V bus with a time delay of 1.0 sec.

b. Degraded voltage Allowable Value 107.5V, 120V bus.

1. Accident time delay (SIS) 8.0 + 0.5, -0.6 sec.

2. Non-accident time delay (No SIS) 56 +8.5, -7.6 sec.

Degraded voltage nominal Trip Setpoint 108.46V, 120V bus.

SR 3.3.5.4 Verify LOP DG Start ESF RESPONSE TIMES are within In accordance with limits. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-46 Amendment No. 123, 183, 227

Containment Purge Isolation Instrumentation 3.3.6 3.3 INSTRUMENTATION 3.3.6 Containment Purge Isolation Instrumentation LCO 3.3.6 The Containment Purge Isolation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.6-1.

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. --------------NOTE-------------- A.1 Place and maintain Immediately Only applicable in containment purge supply MODE 1, 2, 3, or 4. and exhaust valves in

closed position.

One or more Functions with one or more channels or trains inoperable.

(continued)

Wolf Creek - Unit 1 3.3-47 Amendment No. 123, 183, 227

Containment Purge Isolation Instrumentation 3.3.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. ------------NOTE-------------- B.1 Place and maintain Immediately Only applicable during containment purge supply CORE ALTERATIONS or and exhaust valves in movement of irradiated closed position.

fuel assemblies within containment. OR B.2 Enter applicable Immediately One or more Functions Conditions and Required with one or more Actions of LCO 3.9.4, channels or trains "Containment inoperable. Penetrations," for containment purge supply and exhaust valves made inoperable by isolation instrumentation.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Purge Isolation Function.

SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-48 Amendment No. 123, 183, 227

Containment Purge Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.6.2 -------------------------------NOTE--------------------------------

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.3 Perform COT. In accordance with the Surveillance Frequency Control Program SR 3.3.6.4 -----------------------------NOTE----------------------------------

Verification of setpoint is not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.6.5 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.6 Verify Containment Purge Isolation ESF RESPONSE In accordance with TIMES are within limits. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-49 Amendment No. 123, 183, 227

Containment Purge Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Containment Purge Isolation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED SURVEILLANCE FUNCTION CONDITIONS REQUIRED CHANNELS REQUIREMENTS TRIP SETPOINT

1. Manual 1,2,3,4, 2 SR 3.3.6.4 NA Initiation (a),(b)

2. Automatic 1,2,3,4, 2 trains SR 3.3.6.2 NA Actuation Logic (a),(b) SR 3.3.6.6 and Actuation Relays (BOP ESFAS)

3. Containment 1,2,3,4, 1 SR 3.3.6.1 (c)

Atmosphere - (a),(b) SR 3.3.6.3 Gaseous SR 3.3.6.5 Radioactivity

4. Containment Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 3.a, for all initiation functions and requirements.

Isolation -

Phase A (a) During CORE ALTERATIONS.

(b) During movement of irradiated fuel assemblies within containment.

(c) Trip setpoint concentration value (µCi/cm3) is to be established such that the actual submersion rate would not exceed 9 mR/h in the containment building.

Wolf Creek - Unit 1 3.3-50 Amendment No. 123, 183, 227

CREVS Actuation Instrumentation 3.3.7 3.3 INSTRUMENTATION 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation LCO 3.3.7 The CREVS actuation instrumentation for each Function in Table 3.3.7-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.7-1.

ACTIONS

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with A.1 Place one CREVS train in 7 days one channel or train Control Room Ventilation inoperable. Isolation Signal (CRVIS) mode.

(continued)

Wolf Creek - Unit 1 3.3-51 Amendment No. 123, 183, 200, 221, 227

CREVS Actuation Instrumentation 3.3.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. ---------------NOTE------------- B.1.1 Place one CREVS train in Immediately Not applicable to Function the CRVIS mode.

3.

AND One or more Functions B.1.2 Enter applicable Immediately with two channels or two Conditions and Required trains inoperable. Actions of LCO 3.7.10, Control Room Emergency Ventilation System (CREVS), for one CREVS train made inoperable by inoperable CREVS actuation instrumentation.

OR B.2 Place both trains in CRVIS Immediately mode.

C. Both radiation monitoring C.1.1 Enter applicable Immediately channels inoperable. Conditions and Required Actions of LCO 3.7.10, Control Room Emergency Ventilation System (CREVS), for one CREVS train made inoperable by inoperable CREVS actuation instrumentation.

AND C.1.2 Place one CREVS train in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months CRVIS mode.

OR C.2 Place both trains in CRVIS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months mode.

(continued)

Wolf Creek - Unit 1 3.3-52 Amendment No. 123, 183, 221, 227

CREVS Actuation Instrumentation 3.3.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D .1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A, B AND or C not met in MODE 1, 2, 3, or 4. D .2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Required Action and E.1 Suspend CORE Immediately associated Completion ALTERATIONS.

Time for Condition A, B or C not met during AND movement of irradiated fuel assemblies or during E .2 Suspend movement of Immediately CORE ALTERATIONS. irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.7-1 to determine which SRs apply for each CREVS Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.7.2 Perform COT. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-53 Amendment No. 123,132,183, 200, 221, 227

CREVS Actuation Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.7.3 ------------------------------NOTE---------------------------------

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.7.4 -----------------------------NOTE----------------------------------

Verification of setpoint is not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.7.5 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.7.6 ------------------------------NOTE--------------------------------

Radiation monitor detectors are excluded from response time testing.

Verify Control Room Ventilation Isolation ESF In accordance with RESPONSE TIMES are within limits. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-54 Amendment No. 123, 183, 200, 227

CREVS Actuation Instrumentation 3.3.7 Table 3.3.7-1 (page 1 of 1)

CREVS Actuation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS REQUIREMENTS TRIP SETPOINT

1. Manual Initiation 1, 2, 3, 4, 2 SR 3.3.7.4 NA (a) and (c)

2. Automatic Actuation Logic 1, 2, 3, 4, 2 trains SR 3.3.7.3 NA and Actuation Relays (BOP (a) and (c) SR 3.3.7.6 ESFAS)

3. Control Room Radiation- 1, 2, 3, 4, 2 SR 3.3.7.1 (b)

Control Room Air Intakes (a) and (c) SR 3.3.7.2 SR 3.3.7.5 SR 3.3.7.6

4. Containment Isolation - Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 3.a, for all initiation functions and Phase A requirements.

(a) During movement of irradiated fuel assemblies.

(b) Trip Setpoint concentration value (µCi/cm3) is to be established such that the actual submersion dose rate would not exceed 2 mR/hr in the control room.

(c) During CORE ALTERATIONS.

Wolf Creek - Unit 1 3.3-55 Amendment No. 123, 183, 227

EES Actuation Instrumentation 3.3.8 3.3 INSTRUMENTATION 3.3.8 Emergency Exhaust System (EES) Actuation Instrumentation LCO 3.3.8 The EES actuation instrumentation for each Function in Table 3.3.8-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.8-1.

ACTIONS

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

1. LCO 3.0.3 is not applicable.

2. Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Place one EES train in the 7 days with one channel or train Fuel Building Ventilation inoperable. Isolation Signal (FBVIS) mode.

(continued)

Wolf Creek - Unit 1 3.3-56 Amendment No. 123, 183, 227

EES Actuation Instrumentation 3.3.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. ---------------NOTE------------- B.1.1 Place one EES train in the Immediately Not applicable to Function FBVIS mode.

3.

AND One or more Functions B.1.2 Enter applicable Immediately with two channels or two Conditions and Required trains inoperable. Actions of LCO 3.7.13, Emergency Exhaust System (EES), for one EES train made inoperable by inoperable EES actuation instrumentation.

OR B.2 Place both trains in the Immediately FBVIS mode.

C. Both radiation monitoring C.1.1 Enter the applicable Immediately channels inoperable. Conditions and Required Actions of LCO 3.7.13, Emergency Exhaust System (EES), for one EES train made inoperable by inoperable EES actuation instrumentation.

AND C.1.2 Place one EES train in the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months FBVIS mode.

OR C.2 Place both EES trains in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the FBVIS mode.

(continued)

Wolf Creek - Unit 1 3.3-57 Amendment No. 123, 183, 227

EES Actuation Instrumentation 3.3.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Suspend movement of Immediately associated Completion irradiated fuel assemblies Time for Condition A, B or in the fuel building.

C not met during movement of irradiated fuel assemblies in the fuel building.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.8-1 to determine which SRs apply for each EES Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.8.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.8.2 Perform COT. In accordance with the Surveillance Frequency Control Program SR 3.3.8.3 ------------------------------NOTE---------------------------------

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.3-58 Amendment No. 123, 183, 227

EES Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.8.4 ------------------------------NOTE---------------------------------

Verification of setpoint is not required.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.8.5 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.3-59 Amendment No. 227

EES Actuation Instrumentation 3.3.8 Table 3.3.8-1 (page 1 of 1)

EES Actuation Instrumentation APPLICABLE MODES OR SPECIFIED REQUIRED SURVEILLANCE TRIP FUNCTION CONDITIONS CHANNELS REQUIREMENTS SETPOINT

1. Manual Initiation (a) 2 SR 3.3.8.4 NA

2. Automatic Actuation Logic and (a) 2 trains SR 3.3.8.3 NA Actuation Relays (BOP ESFAS)

3. Fuel Building Exhaust Radiation -

Gaseous (a) 2 SR 3.3.8.1 (b)

SR 3.3.8.2 SR 3.3.8.5 (a) During movement of irradiated fuel assemblies in the fuel building.

(b) Trip Setpoint concentration value (µCi/cm3) is to be established such that the actual submersion dose rate would not exceed 4 mR/hr in the fuel building.

Wolf Creek - Unit 1 3.3-60 Amendment No. 227

RCS Pressure, Temperature and Flow DNB Limits 3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 --------------NOTE-------------

THERMAL POWER does not have to be reduced to comply with this Required Action.

Perform SR 3.4.1.3. Prior to THERMAL POWER exceeding 50% RTP AND Prior to THERMAL POWER exceeding 75% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER reaching 95% RTP C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to In accordance with the limit specified in the COLR. the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.4-3 Amendment No. 123, 144, 227

RCS Pressure, Temperature and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.1.2 Verify RCS average temperature is less than or equal In accordance with to the limit specified in the COLR. the Surveillance Frequency Control Program SR 3.4.1.3 Verify RCS total flow rate is 361,200 gpm and In accordance with greater than or equal to the limit specified in the the Surveillance COLR. Frequency Control Program SR 3.4.1.4 ---------------------------NOTE-----------------------------------

Not required to be performed until 7 days after 95% RTP.

Verify by precision heat balance that RCS total flow In accordance with rate is 361,200 gpm and greater than or equal to the Surveillance the limit specified in the COLR. Frequency Control Program Wolf Creek - Unit 1 3.4-4 Amendment No. 123, 144, 221, 227

RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality LCO 3.4.2 Each RCS operating loop average temperature (Tavg) shall be 551°F.

APPLICABILITY: MODE 1, MODE 2 with keff 1.0.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Tavg in one or more A.1 Be in MODE 2 with 30 minutes operating RCS loops not keff < 1.0.

within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify RCS Tavg in each operating loop 551°F. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-5 Amendment No. 123, 227

RCS P/T Limits 3.4.3 ACTIONS (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 whenever limits.

this Condition is entered.

AND Requirements of LCO not C.2 Determine RCS is Prior to entering met any time in other than acceptable for continued MODE 4 MODE 1, 2, 3, or 4. operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 ----------------------------NOTE----------------------------------

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

Verify RCS pressure, RCS temperature, and RCS In accordance with heatup and cooldown rates are within the limits the Surveillance specified in the PTLR. Frequency Control Program Wolf Creek - Unit 1 3.4-7 Amendment No. 123, 227

RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops - MODES 1 and 2 LCO 3.4.4 Four RCS loops shall be OPERABLE and in operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO not A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-8 Amendment No. 123, 227

RCS Loops - MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 Verify steam generator secondary side narrow range In accordance with water levels are 6% for required RCS loops. the Surveillance Frequency Control Program SR 3.4.5.3 Verify correct breaker alignment and indicated power In accordance with are available to the required pump that is not in the Surveillance operation. Frequency Control Program Wolf Creek - Unit 1 3.4-11 Amendment No. 123, 227

RCS Loops - MODE 4 3.4.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 -------------NOTE--------------

Only required if one RHR loop is OPERABLE.

Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B. Required loops inoperable. B.1 Suspend operations that Immediately would cause introduction OR into the RCS, coolant with boron concentration less No RCS or RHR loop in than required to meet operation. SDM of LCO 3.1.1.

AND B.2 Initiate action to restore Immediately one loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.6.2 Verify SG secondary side narrow range water levels In accordance with are 6% for required RCS loops. the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.4-13 Amendment No. 123, 145, 212, 227

RCS Loops - MODE 4 3.4.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.6.3 Verify correct breaker alignment and indicated power In accordance with are available to the required pump that is not in the Surveillance operation. Frequency Control Program SR 3.4.6.4 --------------------------NOTE------------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 4.

Verify required RHR loop locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-14 Amendment No. 123, 145, 212, 227

RCS Loops-MODE 5, Loops Filled 3.4.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR loop inoperable. A.1 Initiate action to restore a Immediately second RHR loop to AND OPERABLE status.

Required SGs secondary OR side water levels not within limits. A.2 Initiate action to restore Immediately required SG secondary side water levels to within limits.

B. Required RHR loops B.1 Suspend operations that Immediately inoperable. would cause introduction into the RCS, coolant with OR boron concentration less than required to meet No RHR loop in operation. SDM of LCO 3.1.1.

AND B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one RHR loop is in operation. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.4-16 Amendment No. 123, 145, 212, 227

RCS Loops-MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.7.2 Verify SG secondary side wide range water level is In accordance with 66% in required SGs. the Surveillance Frequency Control Program SR 3.4.7.3 Verify correct breaker alignment and indicated power In accordance with are available to the required RHR pump that is not in the Surveillance operation. Frequency Control Program SR 3.4.7.4 Verify required RHR loop locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-17 Amendment No. 123, 212, 227

RCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required RHR loops B.1 Suspend operations that Immediately inoperable. would cause introduction into the RCS, coolant with boron concentration less OR than required to meet the No RHR loop in operation SDM of LCO 3.1.1.

AND B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.8.2 Verify correct breaker alignment and indicated power In accordance with are available to the required RHR pump that is not in the Surveillance operation. Frequency Control Program SR 3.4.8.3 Verify RHR loop locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-19 Amendment No. 123, 131, 145, 212, 227

Pressurizer 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One required group of B.1 Restore required group of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer heaters pressurizer heaters to inoperable. OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not AND met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is 92%. In accordance with the Surveillance Frequency Control Program SR 3.4.9.2 Verify capacity of each required group of pressurizer In accordance with heaters is 150 kW. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-21 Amendment No. 123, 212, 227

Pressurizer PORVs 3.4.11 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. More than one block valve ---------------------NOTE-------------------

inoperable. Required Actions do not apply when block valve is inoperable solely as a result of complying with Required Actions B.2 or E.2.

F.1 Place associated PORVs 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in manual control.

AND F.2 Restore one block valve to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months OPERABLE status.

G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition F not AND met.

G.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 -------------------------------NOTE-------------------------------

Not required to be performed with block valve closed in accordance with the Required Actions of this LCO.

Perform a complete cycle of each block valve. In accordance with the Surveillance Frequency Control Program SR 3.4.11.2 Perform a complete cycle of each PORV. In accordance with the Inservice Testing Program Wolf Creek - Unit 1 3.4-26 Amendment No. 123, 212, 227

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of zero safety injection pumps are In accordance with capable of injecting into the RCS. the Surveillance Frequency Control Program SR 3.4.12.2 Verify a maximum of one ECCS centrifugal charging In accordance with pump and the normal charging pump capable of the Surveillance injecting into the RCS. Frequency Control Program SR 3.4.12.3 Verify each accumulator is isolated when In accordance with accumulator pressure is greater than or equal to the the Surveillance maximum RCS pressure for the existing RCS cold Frequency Control leg temperature allowed by the P/T limit curves Program provided in the PTLR.

SR 3.4.12.4 Verify RHR suction isolation valves are open for each In accordance with required RHR suction relief valve. the Surveillance Frequency Control Program SR 3.4.12.5 Verify required RCS vent 2.0 square inches open. In accordance with the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.4-30 Amendment No. 123, 207, 212, 227

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.12.6 Verify PORV block valve is open for each required In accordance with PORV. the Surveillance Frequency Control Program SR 3.4.12.7 Not Used.

SR 3.4.12.8 --------------------------NOTE------------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing any RCS cold leg temperature to 368°F.

Perform a COT on each required PORV, excluding In accordance with actuation. the Surveillance Frequency Control Program SR 3.4.12.9 Perform CHANNEL CALIBRATION for each required In accordance with PORV actuation channel. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-31 Amendment No. 123, 212, 227

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 -----------------------------NOTES-------------------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by In accordance with performance of RCS water inventory balance. the Surveillance Frequency Control Program SR 3.4.13.2 -----------------------------NOTE---------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is 150 In accordance with gallons per day through any one SG. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-33 Amendment No. 123, 164, 212, 227

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 --------------------------------NOTES----------------------------

1. Not required to be performed in MODES 3 and 4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each RCS PIV is equivalent to In accordance with 0.5 gpm per nominal inch of valve size up to a the Surveillance maximum of 5 gpm at an RCS pressure 2215 psig Frequency Control and 2255 psig. Program AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, and if leakage testing has not been performed in the previous 9 months AND Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following check valve actuation due to flow through the valve (continued)

Wolf Creek - Unit 1 3.4-36 Amendment No. 123, 212, 227

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.14.2 Verify RHR suction isolation valve interlock prevents In accordance with the valves from being opened with a simulated or the Surveillance actual RCS pressure signal 425 psig except when Frequency Control the valves are open to satisfy LCO 3.4.12. Program Wolf Creek - Unit 1 3.4-37 Amendment No. 123, 212, 227

RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the required In accordance with containment atmosphere particulate radioactivity the Surveillance monitor. Frequency Control Program SR 3.4.15.2 Perform COT of the required containment In accordance with atmosphere particulate radioactivity monitor. the Surveillance Frequency Control Program SR 3.4.15.3 Perform CHANNEL CALIBRATION of the required In accordance with containment sump level and flow monitoring system. the Surveillance Frequency Control Program SR 3.4.15.4 Perform CHANNEL CALIBRATION of the required In accordance with containment atmosphere particulate radioactivity the Surveillance monitor. Frequency Control Program SR 3.4.15.5 Perform CHANNEL CALIBRATION of the required In accordance with containment cooler condensate monitoring system. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.4-41 Amendment No. 123, 166, 212, 227

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 ----------------------------NOTE-----------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-133 In accordance with specific activity 500 µCi/gm. the Surveillance Frequency Control Program SR 3.4.16.2 ---------------------------NOTE------------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 In accordance with specific activity 1.0 µCi/gm. the Surveillance Frequency Control Program AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period Wolf Creek - Unit 1 3.4-43 Amendment No. 123, 170, 212, 221, 227

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify borated water volume in each accumulator is In accordance with 6122 gallons and 6594 gallons. the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is In accordance with 585 psig and 665 psig. the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each accumulator is In accordance with 2300 ppm and 2500 ppm. the Surveillance Frequency Control Program AND

NOTE-------

Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of 70 gallons that is not the result of addition from the refueling water storage tank (continued)

Wolf Creek - Unit 1 3.5-2 Amendment No. 123, 227

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify power is removed from each accumulator In accordance with isolation valve operator when RCS pressure is the Surveillance

> 1000 psig. Frequency Control Program Wolf Creek - Unit 1 3.5-3 Amendment No. 123, 182, 227

ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

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

1. In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.

2. Operation in MODE 3 with ECCS pumps made incapable of injecting pursuant to LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds 375°F, whichever comes first.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available.

Wolf Creek - Unit 1 3.5-4 Amendment No. 123, 212, 227

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position In accordance with with power to the valve operator removed. the Surveillance Frequency Control Number Position Function Program BN HV-8813 Open Safety Injection to RWST Isolation Valve EM HV-8802A Closed SI Hot Legs 2 & 3 Isolation Valve EM HV-8802B Closed SI Hot Legs 1 & 4 Isolation Valve EM HV-8835 Open Safety Injection Cold Leg Isolation Valve EJ HV-8840 Closed RHR/SI Hot Leg Recirc Isolation Valve EJ HV-8809A Open RHR to Accum Inject Loops 1 & 2 Isolation Valve EJ HV-8809B Open RHR to Accum Inject Loops 3 & 4 Isolation Valve SR 3.5.2.2 -------------------------NOTE-------------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each ECCS manual, power operated, and In accordance with automatic valve in the flow path, that is not locked, the Surveillance sealed, or otherwise secured in position, is in the Frequency Control correct position. Program SR 3.5.2.3 Verify ECCS locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the Inservice developed head. Testing Program (continued)

Wolf Creek - Unit 1 3.5-5 Amendment No. 123, 168, 227

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each ECCS automatic valve in the flow path In accordance with that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an actual Frequency Control or simulated actuation signal. Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, In accordance with each mechanical position stop is in the correct the Surveillance position. Frequency Control Program Valve Number EM-V0095 EM-V0107 EM-V0089 EM-V0096 EM-V0108 EM-V0090 EM-V0097 EM-V0109 EM-V0091 EM-V0098 EM-V0110 EM-V0092 SR 3.5.2.8 Verify, by visual inspection, each ECCS train In accordance with containment sump suction inlet is not restricted by the Surveillance debris and the suction inlet strainers show no Frequency Control evidence of structural distress or abnormal corrosion. Program Wolf Creek - Unit 1 3.5-6 Amendment No. 123, 155, 227

ECCS - Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS -Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE.

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

An RHR subsystem may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned to the ECCS mode of operation.

APPLICABILITY: MODE 4.

ACTIONS

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

LCO 3.0.4b. is not applicable to ECCS centrifugal charging pump subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. Required ECCS residual A.1 Initiate action to restore Immediately heat removal (RHR) required ECCS RHR subsystem inoperable. subsystem to OPERABLE status.

B. Required ECCS B.1 Restore required ECCS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months centrifugal charging pump CCP subsystem to (CCP) subsystem OPERABLE status.

inoperable.

C. Required Action and C.1 Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months associated Completion Time of Condition B not met.

Wolf Creek - Unit 1 3.5-7 Amendment No. 123, 227

ECCS - Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 The following SRs are applicable for all equipment In accordance with required to be OPERABLE: applicable SRs SR 3.5.2.1 SR 3.5.2.7 SR 3.5.2.3 SR 3.5.2.8 SR 3.5.2.4 Wolf Creek - Unit 1 3.5-8 Amendment No. 123, 227

RWST 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron concentration A.1 Restore RWST to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months not within limits. OPERABLE status.

OR RWST borated water temperature not within limits.

B. RWST inoperable for B.1 Restore RWST to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months reasons other than OPERABLE status.

Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Wolf Creek - Unit 1 3.5-9 Amendment No. 123, 227

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 -------------------------NOTE-------------------------------------

Only required to be performed when ambient air temperature is < 37°F or > 100°F.

Verify RWST borated water temperature is 37°F In accordance with and 100°F. the Surveillance Frequency Control Program SR 3.5.4.2 Verify RWST borated water volume is 394,000 In accordance with gallons. the Surveillance Frequency Control Program SR 3.5.4.3 Verify RWST boron concentration is 2400 ppm and In accordance with 2500 ppm. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.5-10 Amendment No. 123, 227

Seal Injection Flow 3.5.5 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Seal Injection Flow LCO 3.5.5 Reactor coolant pump seal injection flow to each RCP seal shall be within the limits of Figure 3.5.5-1.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Seal injection flow not A.1 Adjust manual seal 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months within limit. injection throttle valves to give a flow within the limits of Figure 3.5.5-1.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Wolf Creek - Unit 1 3.5-11 Amendment No. 123, 227

Seal Injection Flow 3.5.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1 -------------------------------NOTE-------------------------------

Not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at 2215 psig and 2255 psig.

Verify manual seal injection throttle valves are In accordance with adjusted to give a flow within the limits of Figure the Surveillance 3.5.5-1. Frequency Control Program Wolf Creek - Unit 1 3.5-12 Amendment No. 123, 132, 160, 227

Seal Injection Flow 3.5.5 Figure 3.5.5-1 (page 1 of 1)

Seal Injection Flow Limits Wolf Creek - Unit 1 3.5-13 Amendment No. 227

Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 ------------------------------NOTES------------------------------

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.

Perform required air lock leakage rate testing in In accordance with accordance with the Containment Leakage Rate the Containment Testing Program. Leakage Rate Testing Program SR 3.6.2.2 Verify only one door in the air lock can be In accordance with opened at a time. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.6-6 Amendment No. 123, 227

Containment Isolation Valves 3.6.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.3 Perform SR 3.6.3.6 or Once per 92 days SR 3.6.3.7 for the resilient seal purge valves closed to comply with Required Action D.1.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each containment shutdown purge valve is In accordance with sealed closed except for one purge valve in a the Surveillance penetration flow path while in Condition D of this Frequency Control LCO. Program AND Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment (continued)

Wolf Creek - Unit 1 3.6-11 Amendment No. 123, 131, 167, 223, 227

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.2 Verify each containment mini-purge valve is closed, In accordance with except when the containment mini-purge valves are the Surveillance open for pressure control, ALARA or air quality Frequency Control considerations for personnel entry, or for Program Surveillances that require the valves to be open.

SR 3.6.3.3 -----------------------------NOTE---------------------------------

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

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

SR 3.6.3.4 -----------------------------NOTE---------------------------------

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

Verify each containment isolation manual valve and Prior to entering blind flange that is located inside containment and not MODE 4 from locked, sealed, or otherwise secured and required to MODE 5 if not be closed during accident conditions is closed, except performed within for containment isolation valves that are open under the previous administrative controls. 92 days SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with operated containment isolation valve is within limits. the Inservice Testing Program (continued)

Wolf Creek - Unit 1 3.6-12 Amendment No. 123, 167, 227

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.6 -----------------------------NOTE---------------------------------

Only required to be performed when containment shutdown purge valve blind flanges are installed.

Perform leakage rate testing for containment In accordance with shutdown purge valves with resilient seals and the Surveillance associated blind flanges. Frequency Control Program AND Following each reinstallation of the blind flange SR 3.6.3.7 --------------------------------NOTE------------------------------

Only required to be performed for the containment shutdown purge valves when associated blind flanges are removed.

Perform leakage rate testing for containment In accordance with mini-purge and shutdown purge valves with resilient the Surveillance seals. Frequency Control Program AND Within 92 days after opening the valve SR 3.6.3.8 Verify each automatic containment isolation valve In accordance with that is not locked, sealed or otherwise secured in the Surveillance position, actuates to the isolation position on an Frequency Control actual or simulated actuation signal. Program Wolf Creek - Unit 1 3.6-13 Amendment No. 123, 167, 227

Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be -0.3 psig and + 1.5 psig.

APPLICABILITY: MODES 1, 2, 3, and 4.

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

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.6-14 Amendment No. 123, 167, 227

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be 120°F.

APPLICABILITY: MODES 1, 2, 3, and 4.

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

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within In accordance with limit. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.6-15 Amendment No. 123, 167, 227

Containment Spray and Cooling Systems 3.6.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two containment cooling D.1 Restore one containment 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months trains inoperable. cooling train to OPERABLE status.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition C or D AND not met.

E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Two containment spray F.1 Enter LCO 3.0.3. Immediately trains inoperable.

OR Any combination of three or more trains inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 --------------------------------NOTE------------------------------

Not required to be met for system vent flow paths opened under administrative control.

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

Wolf Creek - Unit 1 3.6-17 Amendment No. 123, 167, 212, 227

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.2 Operate each containment cooling train fan unit for In accordance with 15 minutes. the Surveillance Frequency Control Program SR 3.6.6.3 Not Used.

SR 3.6.6.4 Verify each containment spray pump's developed In accordance with head at the flow test point is greater than or equal to the Inservice the required developed head. Testing Program SR 3.6.6.5 Verify each automatic containment spray valve in the In accordance with flow path that is not locked, sealed, or otherwise the Surveillance secured in position, actuates to the correct position Frequency Control on an actual or simulated actuation signal. Program SR 3.6.6.6 Verify each containment spray pump starts In accordance with automatically on an actual or simulated actuation the Surveillance signal. Frequency Control Program SR 3.6.6.7 Verify each containment cooling train starts In accordance with automatically and minimum cooling water flow rate is the Surveillance established on an actual or simulated actuation Frequency Control signal. Program SR 3.6.6.8 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage Wolf Creek - Unit 1 3.6-18 Amendment No. 123, 167, 203, 212, 227

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.9 Verify containment spray locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.6-19 Amendment No. 123, 167, 203, 212, 227

Spray Additive System 3.6.7 3.6 CONTAINMENT SYSTEMS 3.6.7 Spray Additive System LCO 3.6.7 The Spray Additive System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spray Additive System A.1 Restore Spray Additive 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. System to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.7.1 Verify each spray additive 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 (continued)

Wolf Creek - Unit 1 3.6-20 Amendment No. 123, 167, 212, 227

Spray Additive System 3.6.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.7.2 Verify spray additive tank solution volume is 4340 In accordance with gal and 4540 gal. the Surveillance Frequency Control Program SR 3.6.7.3 Verify spray additive tank solution concentration is In accordance with 28% and 31% by weight. the Surveillance Frequency Control Program SR 3.6.7.4 Verify each spray additive automatic valve in the flow In accordance with path that is not locked, sealed, or otherwise secured the Surveillance in position, actuates to the correct position on an Frequency Control actual or simulated actuation signal. Program SR 3.6.7.5 Verify spray additive flow rate from each solution's In accordance with flow path. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.6-21 Amendment No. 123, 167, 212, 227

MSIVs and MSIV Bypass Valves 3.7.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME I. -------------NOTE-------------- I.1 Close MSIV. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Separate Condition entry is allowed for each MSIV. AND I.2 Verify MSIV is closed. Once per One or more MSIV 7 days inoperable in MODE 2 or 3.

J. Required Action and J.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition H or I AND not met.

J.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 ------------------------------NOTE--------------------------------

Only required to be performed in MODES 1 and 2.

Verify the isolation time of each MSIV is within limits. In accordance with the Inservice Testing Program SR 3.7.2.2 -------------------------------NOTE-------------------------------

Only required to be performed in MODES 1 and 2.

Verify each actuator train actuates the MSIV to the In accordance with isolation position on an actual or simulated actuation the Surveillance signal. Frequency Control Program (continued)

Wolf Creek - Unit 1 3.7-7 Amendment No. 123, 171, 174, 184, 227

MSIVs and MSIV Bypass Valves 3.7.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.2.3 Verify each MSIV bypass valve actuates to the isolation In accordance with position on an actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.7.2.4 Verify isolation time of each MSIV bypass valve is within In accordance with limit. the Inservice Testing Program Wolf Creek - Unit 1 3.7-8 Amendment No. 184, 227

MFIVs and MFRVs and MFRV Bypass Valves 3.7.3 SURVEILLANCE REQUIRMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.3.2 ------------------------------NOTE--------------------------------

Only required to be performed in MODES 1 and 2.

Verify each actuator train actuates the MFIV to the In accordance with isolation position on an actual or simulated actuation the Surveillance signal. Frequency Control Program SR 3.7.3.3 ------------------------------NOTE--------------------------------

Only required to be performed in MODES 1 and 2.

Verify each MFRV and MFRV bypass valve actuates In accordance with to the isolation position on an actual or simulated the Surveillance actuation signal. Frequency Control Program Wolf Creek - Unit 1 3.7-12 Amendment No. 123, 171, 177, 184, 227

ARVs 3.7.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. With one or more of the D.1 Initiate action to close the Immediately ARVs inoperable because associated block valve(s).

of excessive seat leakage.

AND D.2 Restore ARV(s) to 30 days OPERABLE staus.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify one complete cycle of each ARV. In accordance with the Inservice Testing Program SR 3.7.4.2 Verify one complete cycle of each ARV block valve. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-14 Amendment No. 123, 127, 155, 171, 177, 184, 227

AFW System 3.7.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A, B, or AND C not met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR Two AFW trains inoperable.

E. Three AFW trains E.1 --------------NOTE-------------

inoperable. LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

Initiate action to restore Immediately one AFW train to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 ------------------------------NOTE--------------------------------

Not required to be performed for the AFW flow control valves until the system is placed in standby or THERMAL POWER is > 10% RTP.

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

(continued)

Wolf Creek - Unit 1 3.7-16 Amendment No. 123, 171, 177, 184, 225, 227

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.5.2 ------------------------------NOTE--------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 900 psig in the steam generator.

Verify the developed head of each AFW pump at the In accordance with flow test point is greater than or equal to the required the Inservice Test developed head. Program SR 3.7.5.3 Verify each AFW automatic valve that is not locked, In accordance with sealed, or otherwise secured in position, actuates to the Surveillance the correct position on an actual or simulated Frequency Control actuation signal. Program SR 3.7.5.4 ------------------------------NOTE--------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 900 psig in the steam generator.

Verify each AFW pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.7.5.5 Verify proper alignment of the required AFW flow Prior to entering paths by verifying flow from the condensate storage MODE 2 tank to each steam generator. whenever unit has been in MODE 5 or 6 for > 30 days Wolf Creek - Unit 1 3.7-17 Amendment No. 123, 171, 177, 184, 227

CST 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the CST contained water volume is In accordance with 281,000 gal. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-19 Amendment No. 123, 171, 177, 184, 227

CCW System 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 ------------------------------NOTE--------------------------------

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

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

SR 3.7.7.2 Verify each CCW automatic valve in the flow path that In accordance with is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an actual Frequency Control or simulated actuation signal. Program SR 3.7.7.3 Verify each CCW pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-21 Amendment No. 123, 171, 177, 184, 227

ESW System 3.7.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and associated Completion B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Time of Condition A not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 ------------------------------NOTE--------------------------------

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

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

SR 3.7.8.2 Verify each ESW automatic valve in the flow path that In accordance with is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an Frequency Control actual or simulated actuation signal. Program SR 3.7.8.3 Verify each ESW pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-23 Amendment No. 123, 171, 177, 184, 227

UHS 3.7.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify water level of UHS is 1070 ft mean sea level. In accordance with the Surveillance Frequency Control Program SR 3.7.9.2 Verify plant inlet water temperature of UHS is 90°F. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-25 Amendment No. 123, 134, 171, 177, 184, 227

CREVS 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Two CREVS trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CREVS train pressurization filter unit In accordance with for 15 continuous minutes with the heaters the Surveillance operating and each CREVS train filtration filter unit for Frequency Control 15 continuous minutes. Program SR 3.7.10.2 Perform required CREVS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.10.3 Verify each CREVS train actuates on an actual or In accordance with simulated actuation signal. the Surveillance Frequency Control Program SR 3.7.10.4 Perform required unfiltered air inleakage testing of the In accordance with CRE and CBE boundaries in accordance with the the Control Room Control Room Envelope Habitability Program. Habitability Program Wolf Creek - Unit 1 3.7-28 Amendment No. 123, 134, 171, 177, 179, 184, 208, 227

CRACS 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify each CRACS train has the capability to remove In accordance with the assumed heat load. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-31 Amendment No. 123, 134, 171, 177, 184, 227

EES 3.7.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two EES trains inoperable E.1 Suspend movement of Immediately for reasons other than irradiated fuel assemblies Condition B during in the fuel building.

movement of irradiated fuel assemblies in the fuel building.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Operate each EES train for 15 continuous minutes In accordance with with the heaters operating. the Surveillance Frequency Control Program SR 3.7.13.2 Perform required EES filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.13.3 Verify each EES train actuates on an actual or In accordance with simulated actuation signal. the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.7-35 Amendment No. 123, 132, 134, 171, 177, 184, 208, 221, 227

EES 3.7.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.13.4 Verify one EES train can maintain a negative In accordance with pressure 0.25 inches water gauge with respect to the Surveillance atmospheric pressure in the auxiliary building during Frequency Control the SIS mode of operation. Program SR 3.7.13.5 Verify one EES train can maintain a negative In accordance with pressure 0.25 inches water gauge with respect to the Surveillance atmospheric pressure in the fuel building during the Frequency Control FBVIS mode of operation. Program Wolf Creek - Unit 1 3.7-36 Amendment No. 123, 132, 134, 171, 177, 184, 227

Fuel Storage Pool Water Level 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Fuel Storage Pool Water Level LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool water A.1 -------------NOTE--------------

level not within limit. LCO 3.0.3 is not applicable.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft In accordance with above the top of the irradiated fuel assemblies seated the Surveillance in the storage racks. Frequency Control Program Wolf Creek - Unit 1 3.7-38 Amendment No. 123, 134, 171, 177, 184, 227

Fuel Storage Pool Boron Concentration 3.7.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify the fuel storage pool boron concentration is In accordance with within limit. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-40 Amendment No. 123, 134, 171, 177, 184, 227

Secondary Specific Activity 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Secondary Specific Activity LCO 3.7.18 The specific activity of the secondary coolant shall be 0.10 µCi/gm DOSE EQUIVALENT I-131.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months limit.

AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify the specific activity of the secondary coolant is In accordance with 0.10 µCi/gm DOSE EQUIVALENT I-131. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-43 Amendment No. 123, 134, 171, 177, 184, 227

SSIVs 3.7.19 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.19.1 Verify each automatic SSIV in the flow path is in the In accordance with correct position. the Surveillance Frequency Control Program SR 3.7.19.2 Verify the isolation time of each automatic SSIV is In accordance with within limit. the Inservice Testing Program SR 3.7.19.3 Verify each automatic SSIV in the flow path actuates In accordance with to the isolation position on an actual or simulated the Surveillance actuation signal. Frequency Control Program Wolf Creek - Unit 1 3.7-45 Amendment No. 184, 227

Class 1E Electrical Equipment Air Conditioning (A/C) System 3.7.20 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.20.1 Verify each Class 1E electrical equipment A/C train In accordance with actuates on an actual or simulated actuation signal. the Surveillance Frequency Control Program SR 3.7.20.2 Verify each Class 1E electrical equipment A/C train In accordance with has the capability to remove the assumed heat load. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.7-47 Amendment No. 219, 227

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

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 V and 4320 V, the Surveillance and frequency 59.4 Hz and 60.6 Hz. Frequency Control Program (continued)

Wolf Creek - Unit 1 3.8-7 Amendment No. 123, 163, 204, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.3 -----------------------------NOTES----------------------------------

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 60 minutes at a load 5650 kW and the Surveillance 6201 kW. Frequency Control Program SR 3.8.1.4 Verify each fuel oil transfer pump starts on low level in In accordance with the associated day tank standpipe. 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 each fuel oil transfer system operates to transfer In accordance with fuel oil from the storage tank to the day tank. the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.8-8 Amendment No. 123, 163, 204, 227

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: the Surveillance Frequency Control

a. In 12 seconds, voltage 3950 V and frequency Program 59.4 Hz; and

b. Steady state voltage 3950 V and 4320 V, and frequency 59.4 Hz and 60.6 Hz.

SR 3.8.1.8 Not Used.

SR 3.8.1.9 Not Used.

SR 3.8.1.10 -------------------------------NOTE----------------------------------

If performed with DG synchronized with offsite power, it shall be performed at a power factor 0.9. 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 does not trip and voltage is maintained In accordance with 4992 V and frequency is maintained 65.4 Hz during the Surveillance and following a load rejection of 5650 kW and 6201 Frequency Control kW. Program (continued)

Wolf Creek - Unit 1 3.8-9 Amendment No. 123, 154, 161, 163, 204, 206,215, 227

AC Sources - Operating 3.8.1 SURVEILLANCE 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 normally be performed in MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify on an actual or simulated loss of offsite power In accordance with signal: the Surveillance Frequency Control

a. De-energization of emergency buses; Program

b. Load shedding from emergency buses;

c. DG auto-starts from standby condition and:

1. energizes permanently connected loads in 12 seconds,

2. energizes auto-connected shutdown loads through the shutdown sequencer,

3. maintains steady state voltage 3950 V and 4320 V,

4. maintains steady state frequency 59.4 Hz and 60.6 Hz, and

5. supplies permanently connected and auto-connected shutdown loads for 5 minutes.

(continued)

Wolf Creek - Unit 1 3.8-10 Amendment No. 123, 154, 163, 204, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.12 ------------------------------NOTES---------------------------------

1. All DG starts may be preceded by a prelube period.

2. This Surveillance shall not normally be performed in MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify on an actual or simulated Engineered Safety In accordance with Feature (ESF) actuation signal each DG auto-starts the Surveillance from standby condition and: Frequency Control Program

a. In 12 seconds after auto-start and during tests, achieves voltage 3950 V and frequency 59.4 Hz;

b. Achieves steady state voltage 3950 V and 4320 V, and frequency 59.4 Hz and 60.6 Hz;

c. Operates for 5 minutes;

d. Permanently connected loads remain energized from the offsite power system; and

e. Emergency loads are auto-connected and energized through the LOCA sequencer from the offsite power system.

(continued)

Wolf Creek - Unit 1 3.8-11 Amendment No. 123, 154, 161, 163, 204, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.13 Verify each DG's automatic trips are bypassed on In accordance with actual or simulated loss of voltage signal on the the Surveillance emergency bus concurrent with an actual or simulated Frequency Control ESF actuation signal except: Program

a. Engine overspeed;

b. Generator differential current;

c. Low lube oil pressure;

d. High crankcase pressure;

e. Start failure relay; and

f. High jacket coolant temperature.

(continued)

Wolf Creek - Unit 1 3.8-12 Amendment No. 123, 154, 163, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.14 ------------------------------NOTES---------------------------------

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

2. If performed with DG synchronized with offsite power, it shall be performed at a power factor 0.9. 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 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 6300 kW and 6821 kW; Frequency Control and Program

b. For the remaining hours of the test loaded 5650 kW and 6201 kW.

SR 3.8.1.15 -----------------------------NOTES----------------------------------

1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded 5650 kW and 6201 kW. Momentary transients outside of load range do not invalidate this test.

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

Verify each DG starts and achieves: In accordance with the Surveillance

a. In 12 seconds, voltage 3950 V and frequency Frequency Control 59.4 Hz; and Program

b. Steady state voltage 3950 V and 4320 V, and frequency 59.4 Hz and 60.6 Hz.

(continued)

Wolf Creek - Unit 1 3.8-13 Amendment No. 123, 154, 161, 163, 204, 215, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.16 ------------------------------NOTE-----------------------------------

This Surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify each DG: In accordance with the Surveillance

a. Synchronizes with offsite power source while Frequency Control loaded with emergency loads upon a simulated Program restoration of offsite power;

b. Transfers loads to offsite power source; and

c. Returns to ready-to-load operation.

SR 3.8.1.17 ------------------------------NOTE-----------------------------------

This Surveillance shall not normally be performed in MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify, with a DG operating in test mode and connected In accordance with to its bus, an actual or simulated Safety Injection signal the Surveillance overrides the test mode by: Frequency Control Program

a. Returning DG to ready-to-load operation; and

b. Automatically energizing the emergency load from offsite power.

(continued)

Wolf Creek - Unit 1 3.8-14 Amendment No. 123, 154, 163, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENT (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.18 -------------------------------NOTE----------------------------------

This Surveillance shall not normally be performed in MODE 1 or 2. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify interval between each sequenced load block is In accordance with within +/- 10% of design interval for each LOCA and the Surveillance shutdown sequence timer. Frequency Control Program (continued)

Wolf Creek - Unit 1 3.8-15 Amendment No. 123, 154, 163, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.19 -------------------------------NOTES--------------------------------

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

2. This Surveillance shall not normally be performed in MODE 1 or 2. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify on an actual or simulated loss of offsite power In accordance with signal in conjunction with an actual or simulated Safety the Surveillance Injection 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 12 seconds,

2. energizes auto-connected emergency loads through load sequencer,

3. achieves steady state voltage 3950 V and 4320 V,

4. achieves steady state frequency 59.4 Hz and 60.6 Hz, and

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

(continued)

Wolf Creek - Unit 1 3.8-16 Amendment No. 123, 154, 163, 204, 227

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.20 ------------------------------NOTE-----------------------------------

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

Verify when started simultaneously from standby In accordance with condition, each DG achieves: the Surveillance Frequency Control

a. In 12 seconds, voltage 3950 V and frequency Program 59.4 Hz; and

b. Steady state voltage 3950 V and 4320 V, and frequency 59.4 Hz and 60.6 Hz.

SR 3.8.1.21 ----------------------------NOTE-------------------------------------

The continuity check may be excluded from the actuation logic test.

Perform ACTUATION LOGIC TEST for each train of the In accordance with load shedder and emergency load sequencer. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.8-17 Amendment No. 123, 161, 163, 204, 227

Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each fuel oil storage tank contains 85,300 gal In accordance with of fuel. the Surveillance Frequency Control Program SR 3.8.3.2 Verify lubricating oil inventory is 750 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 pressure in two starting air receivers is 435 In accordance with psig or pressure in one starting air receiver is 610 the Surveillance psig for each DG starting air subsystem. 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 Wolf Creek - Unit 1 3.8-23 Amendment No. 123, 163, 227

DC Sources - Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources - Operating LCO 3.8.4 The Train A and Train B DC electrical power subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One DC electrical power A.1 Restore DC electrical 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months subsystem inoperable. power subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is 128.4 V on float In accordance with charge. the Surveillance Frequency Control Program (continued)

Wolf Creek - Unit 1 3.8-24 Amendment No. 123, 132, 163, 227

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 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 is:

Connections 60 cells 59 cells 58 cells inter-cell 33 E-6 ohms 30 E-6 ohms 27 E-6 ohms inter-tier, 150 E-6 ohms 150 E-6 ohms 150 E-6 ohms inter-bank, terminal field jumper NA 150 E-6 ohms 150 E-6 ohms 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 could degrade battery performance. Frequency Control Program SR 3.8.4.4 Remove visible terminal corrosion, verify battery cell In accordance with to cell and terminal connections are clean and tight, the Surveillance and are coated with anti-corrosion material. Frequency Control Program SR 3.8.4.5 Verify battery connection resistance is: In accordance with the Surveillance Connections 60 cells 59 cells 58 cells Frequency Control Program inter-cell 33 E-6 ohms 30 E-6 ohms 27 E-6 ohms inter-tier, 150 E-6 ohms 150 E-6 ohms 150 E-6 ohms inter-bank, terminal field jumper NA 150 E-6 ohms 150 E-6 ohms (continued)

Wolf Creek - Unit 1 3.8-25 Amendment No. 123, 163, 192, 227

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.4.6 Verify each battery charger supplies In accordance with 300 amps at 128.4 V for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . the Surveillance Frequency Control Program 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, 3, or 4.

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 (continued)

Wolf Creek - Unit 1 3.8-26 Amendment No. 123, 163, 192, 227

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.4.8 ---------------------------NOTE-----------------------------------

This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify battery capacity is 85% of the manufacturer's In accordance with rating when subjected to a performance discharge the Surveillance test or a modified performance discharge test. Frequency Control Program AND 18 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 Wolf Creek - Unit 1 3.8-27 Amendment No. 123, 163, 227

Battery Cell Parameters 3.8.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Declare associated Immediately associated Completion battery inoperable.

Time of Condition A not met.

OR One or more batteries with average electrolyte temperature of the representative cells

< 60°F.

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

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)

Wolf Creek - Unit 1 3.8-31 Amendment No. 123, 163, 227

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 AND Once within 7 days after a battery discharge

< 110 V AND Once within 7 days after a battery overcharge

> 150 V SR 3.8.6.3 Verify average electrolyte temperature of In accordance with representative cells is 60 °F. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.8-32 Amendment No. 123, 163, 227

Inverters - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Inverters - Operating LCO 3.8.7 The required Train A and Train B inverters shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required inverter A.1 --------------NOTE-------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, "Distribution Systems -

Operating" with any vital bus de-energized.

Restore inverter to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage and alignment to In accordance with required AC vital buses. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.8-34 Amendment No. 123, 163, 227

Inverters - Shutdown 3.8.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage and alignments to In accordance with required AC vital buses. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.8-36 Amendment No. 123, 163, 227

Distribution Systems - Operating 3.8.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One DC electrical power D.1 Restore DC electrical 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months distribution subsystem power distribution inoperable. subsystem to OPERABLE AND status.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Two trains with inoperable F.1 Enter LCO 3.0.3. Immediately distribution subsystems that result in a loss of safety function.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to AC, In accordance with DC, and AC vital bus electrical power distribution the Surveillance subsystems. Frequency Control Program Wolf Creek - Unit 1 3.8-38 Amendment No. 123, 163, 227

Distribution Systems - Shutdown 3.8.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate actions to restore Immediately required AC, DC, and AC vital bus electrical power distribution subsystems to OPERABLE status.

AND A.2.5 Declare associated Immediately required residual heat removal subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to In accordance with required AC, DC, and AC vital bus electrical power the Surveillance distribution subsystems. Frequency Control Program Wolf Creek - Unit 1 3.8-40 Amendment No. 123, 163, 227

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of all filled portions of the Reactor Coolant System and the refueling canal, that have direct access to the reactor vessel, shall be maintained within the limit specified in the COLR.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend CORE Immediately within limit. ALTERATIONS.

AND A.2 Suspend positive reactivity Immediately additions.

AND A.3 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified In accordance with in the COLR. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.9-1 Amendment No. 123, 155, 227

Unborated Water Source Isolation Valves 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Unborated Water Source Isolation Valves LCO 3.9.2 Each valve used to isolate unborated water sources, BG-V0178 and BG-V0601, shall be secured in the closed position.

APPLICABILITY: MODE 6.

ACTIONS

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

Separate Condition entry is allowed for each unborated water source isolation valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE--------------- A.1 Suspend CORE Immediately Required Action A.3 must ALTERATIONS.

be completed whenever Condition A is entered. AND A.2 Initiate actions to secure Immediately One or more valves not valve in closed position.

secured in closed position.

AND A.3 Perform SR 3.9.1.1. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify each valve that isolates unborated water In accordance with sources, BG-V0178 and BG-V0601, is secured in the the Surveillance closed position. Frequency Control Program Wolf Creek - Unit 1 3.9-2 Amendment No. 123, 227

Nuclear Instrumentation 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.9.3.2 --------------------------NOTE------------------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.9-4 Amendment No. 123, 227

Containment Penetrations 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS.

required status.

AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify each required containment penetration is in the In accordance with required status. the Surveillance Frequency Control Program SR 3.9.4.2 --------------------------NOTE------------------------------------

Only required for an open equipment hatch.

Verify the capability to install the equipment hatch. In accordance with the Surveillance Frequency Control Program SR 3.9.4.3 Verify each required containment purge isolation In accordance with valve actuates to the isolation position on an actual or the Surveillance simulated actuation signal. Frequency Control Program Wolf Creek - Unit 1 3.9-6 Amendment No. 123, 131, 135, 146, 227

RHR and Coolant Circulation - High Water Level 3.9.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.4 Close all containment 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetrations providing direct access from containment atmosphere to outside atmosphere.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating In accordance with reactor coolant at a flow rate of 1000 gpm. the Surveillance Frequency Control Program SR 3.9.5.2 Verify required RHR loop locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.9-8 Amendment No. 123, 212, 227

RHR and Coolant Circulation - Low Water Level 3.9.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Initiate action to restore Immediately one RHR loop to operation.

AND B.3 Close all containment 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetrations providing direct access from containment atmosphere to outside atmosphere.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify one RHR loop is in operation and circulating In accordance with reactor coolant at a flow rate of 1000 gpm. the Surveillance Frequency Control Program SR 3.9.6.2 Verify correct breaker alignment and indicated power In accordance with available to the required RHR pump that is not in the Surveillance operation. Frequency Control Program SR 3.9.6.3 Verify RHR loop locations susceptible to gas In accordance with accumulation are sufficiently filled with water. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.9-10 Amendment No. 123, 212, 227

Refueling Pool Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Refueling Pool Water Level LCO 3.9.7 Refueling pool water level shall be maintained 23 ft above the top of reactor vessel flange.

APPLICABILITY: During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling pool water level A.1 Suspend movement of Immediately not within limit. irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify refueling pool water level is 23 ft above the In accordance with top of reactor vessel flange. the Surveillance Frequency Control Program Wolf Creek - Unit 1 3.9-11 Amendment No. 123, 227

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program (continued)

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits,

2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. water and sediment content within the limits for ASTM 2D fuel oil;

b. Other properties for ASTM 2D fuel oil are analyzed within 31 days following sampling and addition to storage tanks; and

c. Total particulate concentration of the fuel oil is 10 mg/l when tested in accordance with ASTM D-2276, Method A, at a Frequency in accordance with the Surveillance Frequency Control Program.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

5.5.14 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

(continued)

Wolf Creek - Unit 1 5.0-18 Amendment No. 123, 138, 164, 227

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE, CRE boundary, control building envelope (CBE), and CBE boundary.

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

c. Requirements for (i) determining the unfiltered air inleakage past the CRE and CBE boundaries 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.

The following are exceptions to Section C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1. The Tracer Gas Test based on the Brookhaven National Laboratory Atmospheric Tracer Depletion (ATD) Method is used to determine the unfiltered air inleakage past the CRE and CBE boundaries. The ATD Method is described in WCNOC letters dated February 21, 2005 (WO 05-0003), June 29, 2007 (WM 07-0057), and September 28, 2007 (ET 07-0045).

d. Measurement, at designated locations, of the CRE pressure relative to the outside atmosphere during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency in accordance with the Surveillance Frequency Control Program. The results shall be trended and used as part of the periodic assessment of the CRE boundary.

(continued)

Wolf Creek - Unit 1 5.0-22 Amendment No. 123, 142, 152, 164, 179, 221, 227

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 Control Room Envelope Habitability Program (continued)

e. The quantitative limits on unfiltered air inleakage into the CRE and CBE.

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 DBA 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 habitability, determining CRE and CBE unfiltered inleakage, and measuring CRE pressure and assessing the CRE and CBE as required by paragraphs c and d, respectively.

5.5.19 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 Technical Specifications Initiative 5b, 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.

Wolf Creek - Unit 1 5.0-23 Amendment No. 123, 142, 152, 164, 179, 227

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 227 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482

1.0 INTRODUCTION

By application dated April 27, 2020 (Reference 1), as supplemented by letter dated October 26, 2020 (Reference 2), the Wolf Creek Nuclear Operating Corporation (the licensee), requested changes to the Wolf Creek Generating Station, Unit 1 (Wolf Creek), Technical Specifications (TSs).

The proposed changes would revise the TSs by relocating specific surveillance requirement (SR) frequencies to a licensee-controlled program in accordance with Nuclear Energy Institute (NEI) 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies (Reference 3). The U.S.

Nuclear Regulatory Commission (NRC or the Commission) staff reviewed and approved NEI 04-10, Revision 1, by letter dated September 19, 2007 (Reference 4). The requested changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee ControlRITSTF [Risk-Informed TSTF] Initiative 5b (Reference 5),

which provides for the application of NEI 04-10, Revision 1.

The supplemental letter dated October 26, 2020, 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 on June 16, 2020 (85 FR 36436).

2.0 REGULATORY EVALUATION

2.1 Background

The licensee proposed to modify the Wolf Creek TSs by relocating specific surveillance frequencies to a licensee-controlled program (i.e., the Surveillance Frequency Control Program (SFCP)) in accordance with NEI 04-10, Revision 1. The licensee stated that the proposed changes are consistent with the adoption of NRC-approved TSTF-425, Revision 3. The Federal Register notice published on July 6, 2009 (74 FR 31996), announced the availability of Enclosure 2

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 Inservice Testing Program or the Primary Containment Leakage Rate Testing Program);

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 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 the specific surveillance frequencies documented in the license amendment request 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) 3.6 Containment Systems 3.7 Plant Systems 3.8 Electrical Power Systems 3.9 Refueling Operations 5.5 Programs and Manuals The licensee proposed to add the SFCP to TS Section 5.0, Administrative Controls, Subsection 5.5, Programs and Manuals. The SFCP describes the requirements for the program to control changes to the relocated surveillance frequencies. The proposed changes to the Administrative Controls section of the TSs to incorporate the SFCP include a specific reference to NEI 04-10, Revision 1, as the basis for making any changes to the surveillance frequencies once they are relocated out of the TSs. The TS Bases for each affected surveillance would be revised to state that the surveillance frequency is controlled under the SFCP.

In a letter dated September 19, 2007 (Reference 4), the NRC staff approved NEI 04-10, Revision 1, as acceptable for referencing in licensing actions, to the extent specified and under the limitations delineated in NEI 04-10, Revision 1, and in the NRC staffs safety evaluation (SE) for NEI 04-10, Revision 1.

The licensee proposed other changes and deviations from TSTF-425, which are discussed in Section 3.2 of this SE.

2.2 Applicable Commission Policy Statements In the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, 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 two years later, the NRC provided additional detail concerning the use of PRA in the Use of Probabilistic Risk Assessment in Nuclear Regulatory Activities; Final Policy Statement, 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 NRCs 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 Commissions 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 NRCs deterministic approach and supports the NRCs 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 Commissions 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 Wolf Creek TSs.

Section 50.36(c)(3) of 10 CFR states, [s]urveillance 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.

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, Domestic Licensing of Production and Utilization Facilities, 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.

2.4 Applicable Guidance Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 6), 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, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications (Reference 7), describes an acceptable risk informed approach specifically for assessing proposed TS changes.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 8), 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.

NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases (References 9 and 10, respectively), contain the improved STS for Westinghouse plants. The improved STS were developed based on the criteria in the Final 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).

TSTF-425, Revision 3, involves the relocation of most time-based surveillance frequencies to a licensee-controlled program, SFCP, and adds the SFCP to the administrative controls section of TS. TSTF-425, Revision 3, implements the staff-approved NEI 04-10, Revision 1.

3.0 TECHNICAL EVALUATION

The licensees adoption of TSTF-425, Revision 3, provides for Wolf Creek administrative relocation of applicable surveillance frequencies and provides for the addition of the SFCP to the Administrative Controls section of the TSs. TSTF-425, Revision 3, also provides for the application of NEI 04-10, Revision 1, for any changes to surveillance frequencies within the SFCP. The Federal Register 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 licensees application for the changes proposed in TSTF-425, Revision 3, included documentation regarding the PRA technical

adequacy consistent with the guidance provided in RG 1.200, Revision 2, Section 4.2. In accordance with NEI 04-10, Revision 1, PRA and non-PRA methods are used, in combination with plant performance data and other considerations, to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is in accordance with guidance provided in RG 1.174, Revision 3, and RG 1.177, Revision 1, in support of changes to surveillance test intervals (STIs). 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.

3.1 RG 1.177 Five Key Safety Principles RG 1.177, Revision 1, identifies five key principles required for risk-informed changes to TSs.

Each of these principles is addressed by the industry methodology document, NEI 04-10, Revision 1.

3.1.1 The Proposed Change Meets Current Regulations The regulations in 10 CFR 50.36(c)(3) require that TSs will 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. NEI 04-10, Revision 1, provides guidance for relocating the surveillance frequencies from the TSs to a licensee-controlled program by providing an NRC-approved methodology for control of the surveillance frequencies. The surveillances themselves would remain in the TSs, as required by 10 CFR 50.36(c)(3).

This change is consistent with other NRC-approved TS changes in which the surveillance frequencies are relocated to licensee-controlled documents, such as surveillances performed in accordance with the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program. Further, the NEI 04-10, Revision 1, guidance provides for monitoring the performance of SSCs for which surveillance frequencies are decreased to assure that the reduced testing does not adversely impact the SSCs. Thus, this proposed change meets the current regulations for monitoring surveillance test failures and implementing corrective actions to address such failures, in accordance with 10 CFR 50.65 and 10 CFR Part 50, Appendix B, Criterion XVI.

Therefore, the proposed change meets the first key 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 Consistency with the defense-in-depth philosophy 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 common cause failure mechanisms is assessed.

Independence of barriers is not degraded.

Defenses against human errors are preserved.

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

TSTF-425, Revision 3, requires 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 core damage frequency (CDF) and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. The guidance in RG 1.174, Revision 3, and RG 1.177, Revision 1, for changes to CDF and LERF is achieved by evaluation 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 increased likelihood of CCFs. Both the quantitative risk analysis and the qualitative considerations assure a reasonable balance of defense-in-depth is maintained to provide reasonable assurance of protection of public health and safety, satisfying the second key 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 with the principle 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 authorized for use by the NRC) will continue to be met as described in the plant licensing basis (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.

Thus, safety margins are maintained by the proposed methodology and therefore, the third key principle of RG 1.177, Revision 1, is satisfied.

3.1.4 When Proposed Changes Result in an Increase in CDF or Risk, the Increases Should Be Small and Consistent with the Intent of the Commissions Safety Goal Policy Statement The guidance in RG 1.177, Revision 1, provides a framework for evaluation of the risk 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. TSTF-425, Revision 3, requires application of NEI 04-10, Revision 1, in the SFCP.

The guidance in NEI 04-10, Revision 1, satisfies the intent of RG 1.177, Revision 1, requirements for evaluation of the change in risk and for assuring that such changes are small.

Thus, the licensees proposed relocation of specific surveillance frequencies to the SFCP in accordance with NEI 04-10, Revision 1, and TSTF-425, Revision 3, satisfies the fourth key principle of RG 1.177, Revision 1.

3.1.4.1 Technical Acceptability of PRAs The NRC staffs review of the technical acceptability of the licensees PRAs supporting this application is consistent with the safety implications of the proposed TS change, and the role the PRA plays in justifying the change. That is, the more the potential change in risk or the greater the uncertainty in that risk that results from the requested TS change, or both, the greater the rigor that must go into ensuring the acceptability of the PRA.

The licensee used RG 1.200, Revision 2, to address the plant PRA technical acceptability for this application. RG 1.200, Revision 2, provides regulatory guidance for assessing the technical acceptability of a PRA and endorses (with clarifications and qualifications) the use of the following:

1. American Society of Mechanical Engineers/American Nuclear Society (ASME/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 (hereafter referred to as the ASME/ANS PRA Standard) (Reference 11);

2. NEI 00-02, Revision 1, Probabilistic Risk Assessment (PRA) Peer Review Process Guidance (Reference 12); and

3. NEI 05-04, Revision 2, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard (Reference 13).

The licensee has performed an assessment of the PRA models used to support the SFCP against the guidance provided in RG 1.200, Revision 2, to assure that the PRA models, using plant specific data and models, are capable of determining the change in risk due to changes to surveillance frequencies of SSCs. NEI 04-10 states that Capability Category II (CC-II) of the ASME/ANS PRA Standard should be met, and any identified deficiencies to the CC-II supporting requirements of the ASME/ANS PRA Standard are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including the use of sensitivity studies, as appropriate. This level of PRA acceptability is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP and is consistent with Regulatory Position 2.3.1, Technical Adequacy of the PRA, of RG 1.177, Revision 1.

Internal Events and Internal Flooding PRA Upon implementation of the SFCP, the licensee will determine whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the internal events and internal flooding PRA. The NRC staff reviewed Table 3-1, WCGS [Wolf Creek] Open PRA Peer Review Findings, of Attachment II to the license amendment request (LAR), which summarized the peer review team assessment of the Wolf Creek PRA models that do not conform to CC-II of the ASME/ANS PRA Standard supporting requirements. The NRC staffs assessment of these open finding level facts and observations (F&Os), the impacted supporting requirements, and the licensees resolutions concluded that they are addressed and dispositioned for this application per the NEI 04-10 guidance, as discussed below.

F&O 3-8, associated with Supporting Requirements AS-C3, HR-I3, IE-D3, SC-C3, SY-C3, and QU-F4, was generated because of a lack of a clear method for identification and characterization of key plant-specific assumptions and sources of uncertainty. In the LAR, the licensee stated that plant-specific sources of uncertainty were being collected and characterized in the individual PRA notebooks to resolve this issue. In the response to NRC Request for Additional Information (RAI) No. 1, by letter dated October 26, 2020, the licensee stated that this action will be complete prior to implementation of the SFCP. The NRC staff finds this acceptable because the key plant-specific assumptions and sources of uncertainty will be identified and characterized in accordance with the ASME/ANS PRA Standard before the SFCP is in effect, and because NEI 04-10, Revision 1, includes guidance on the performance of sensitivity studies related to key assumptions and causes of uncertainty as discussed in Section 3.1.4.5 of this SE.

F&O 4-10, associated with Supporting Requirement LE-C13, was generated because the documentation for how steam generator tube ruptures (SGTRs) were modeled does not provide sufficient technical basis to justify the credit taken. Additionally, the simplified approach for interfacing system loss-of-coolant accident (ISLOCA) releases does not discuss any consideration of potential scrubbing credit. The licensee stated that the Wolf Creek LERF PRA model only treats those SGTRs where steam generator isolation has failed as generating a large and early release, and that with successful steam generator isolation, the containment is not completely bypassed and therefore, these SGTRs do not result in a large and early release.

Further, the licensee stated that in the case of failed steam generator isolation, fission product scrubbing via secondary side inventory is not realistic for most scenarios because of the uncertainty of the leak location and the availability of a sufficient water pool above the leak to scrub fission products from a potential release.

Regarding not taking credit for potential ISLOCA scrubbing, the licensee stated that crediting scrubbing for ISLOCA sequences requires complex modeling of the release pathway through the auxiliary building in order to track fission product plate out prior to offsite release, and that this is beyond the state of practice in the industry.

The SFCP process for extending an STI is not affected by the issue discussed in F&O 4-10 because the STI evaluation determines the change in risk, and the uncredited actions will affect similarly the original and the extended interval risk with little or no impact on the change in risk.

Therefore, the NRC staff finds that the change in risk would not be expected to be significantly impacted by modeling SGTR or ISLOCA scrubbing, and therefore, the licensees disposition of F&O 4-10 is acceptable for this application.

F&O 6-8, associated with Supporting Requirement SY-C2, was generated because walkdowns and interviews were performed but not documented. In the LAR, the licensee stated that the majority of the system engineer interviews have been completed and documented in the corresponding systems analyses notebooks.

In the response to NRC RAI No. 2 in the supplemental letter dated October 26, 2020, the licensee stated that the outstanding interview has taken place, and that the interview confirmed the relevant modeling assumptions and conclusions. The licensee further stated that the documentation of the interview would be incorporated into the model of record. The NRC staff finds this acceptable because the goal of the F&O (confirming relevant modeling assumptions and conclusions with knowledgeable experts) has been achieved, and this result will be documented in the PRA model documentation going forward.

F&O AS-B3-01, associated with Supporting Requirement AS-B3, was generated because feed and bleed scenarios involving open power operated relief valves did not consider the potential for sump strainer blockage. In the LAR, the licensee stated that sump blockage is accounted for in loss-of-coolant accident and reactor coolant pump seal leakage events. The licensee further stated that a sensitivity study was performed to determine the impact of not limiting sump blockage to loss-of-coolant accident type events, and the results did not impact this application.

Based on these results, the NRC staff finds that F&O AS-B3-01 would be expected to have a negligible impact on STI calculations.

Conclusion Based on its review, the NRC staff concludes that the licensees internal events PRA, including internal flooding, is technically acceptable to support the evaluation of changes proposed to surveillance frequencies within the SFCP using the process in NEI 04-10, Revision 1, and is consistent with Regulatory Position 2.3.1 of RG 1.177, Revision 1.

3.1.4.2 Scope of the PRA Upon implementation of the SFCP, the licensee must evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10, Revision 1, to determine its potential impact on risk, due to impacts from internal events, fires, seismic, other external events, and from shutdown conditions. Consideration is made of both CDF and LERF metrics.

Wolf Creek has full-power internal event and internal flood PRA models. These models received peer reviews as discussed in Section 3.1.4.1 of this SE. In accordance with NEI 04-10, Revision 1, the licensee will use these models to perform quantitative evaluations to support the development of changes to surveillance frequencies in the SFCP. The NRC staff finds that the use of these models is acceptable because the NRC-approved methodology in NEI 04-10, Revision 1, allows for more refined analysis to be performed to support changes to surveillance frequencies in the SFCP.

Wolf Creek does not have a PRA model for internal fires. However, the licensee performed a Fire Induced Vulnerability Evaluation (FIVE) for Wolf Creek in response to Individual Plant Examination of External Events, Generic Letter 88-20, Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)

(Reference 14). The licensee performed its FIVE for Wolf Creek in accordance with Electric Power Research Institute (EPRI) Topical Report (TR)-100370, Fire-Induced Vulnerability

Evaluation (FIVE) (Reference 15). The licensee will assess the impacts on fire risk of an STI extension using a qualitative or a bounding approach supplemented with insights from the IPEEE FIVE analysis and the in-development fire PRA model. Based on its review, the NRC staff finds that the licensees approach is acceptable for this application because it is consistent with the guidance in NEI 04-10, Revision 1.

The licensee does not have PRA models for seismic events, high winds, and external flooding events. These events were assessed in the licensees IPEEE, Generic Letter 88-20, Supplement 4. In Section 5, External Events Considerations, of Attachment II to the LAR, the licensee stated that Wolf Creek is currently developing a seismic PRA, but it has neither been completed nor peer reviewed. A seismic margins assessment (SMA) was performed with a review level earthquake at 0.3 gravity peak ground acceleration during the IPEEE. The licensee proposed to assess SSCs impacted by frequency changes in the proposed program against the SMA consistent with the endorsed guidance in NEI 04-10.

The SMA in the licensees IPEEE concluded that only a few instances are required for a detailed review to screen equipment anchorages to a high confidence of a low probability of failure of at least 0.3 gravity peak ground acceleration. The licensees recent seismic hazard and screening report (Reference 16) shows that the ground motion response spectrum for the Wolf Creek site is higher than the safe shutdown earthquake at a frequency of about 5 hertz or higher. In response to RAI No. 5, by letter dated in October 26, 2020, the licensee stated that sources of insights used in its SFCP for a qualitative evaluation of seismic risk will include the SMA from the IPEEE, the more recent Expedited Seismic Evaluation Process Report, and the seismic PRA model when it is complete. Based on its review, the NRC staff finds that the use of the SMA and insights from recent assessments for seismic events are consistent with the guidance in NEI 04-10, Revision 1.

In Section 5.0, External Events Considerations, of Attachment II to the LAR, the licensee stated that Wolf Creek has developed a high winds PRA model and conducted an external events screening assessment in accordance with Parts 6 and 7 of the ASME/ANS PRA Standard. However, given that the high winds PRA model still has outstanding F&Os that need to be addressed, the licensee proposed to utilize the IPEEE analysis for high winds and tornadoes for the STI change evaluations. In response to RAI No. 6, by letter dated in October 26, 2020, the licensee stated that sources of insights used in its SFCP for a qualitative evaluation of high winds and tornado risk will include the IPEEE and the high winds PRA when it is complete. Based on its review, the NRC staff finds that the use of the IPEEE analysis and insights from recent assessments for high winds and tornadoes are consistent with the guidance in NEI 04-10, Revision 1.

In Section 5.0 of Attachment II to the LAR, the licensee stated that other external hazards, including external flooding, were determined in the IPEEE to be negligible contributors to overall plant risk. In the licensees recent flood hazard reevaluation report (Reference 17), the licensee determined that the current design basis does not bound the reevaluated local intense precipitation hazard. In response to RAI No. 7, by letter dated in October 26, 2020, the licensee stated that sources of insights used in its SFCP for a qualitative evaluation of external flooding risk will include the IPEEE and the more recent flood hazard reevaluation, and the external events screening assessment. Based on its review, the NRC staff finds that the use of the IPEEE analysis and insights from recent assessments for external flooding are consistent with the guidance in NEI 04-10, Revision 1.

Based on the supplemental letter dated October 26, 2020, the licensee stated that it will address the impact of all external hazards on a specific STI extension using a qualitative approach that follows the methodology of NEI 04-10, Revision 1, as endorsed by the NRC staff. The licensee stated that its approach will be contained in the Wolf Creek SFCP procedures. The licensee explained that for the unscreened external hazards, qualitative insights will be developed from available sources and the licensees program will require that these qualitative risk insights, the basis for qualitative conclusions and the uncertainties or limitations associated with those insights relevant to the specific STI extension, be presented to the independent decision-making panel (IDP) for consideration.

The NRC staff notes that in accordance with NEI 04-10, Revision 1, the licensee can perform an initial qualitative screening analysis, and, if the qualitative information is insufficient to provide confidence that the net impact of the STI change would be negligible, a bounding analysis will be performed. The licensee stated that its approach will use a qualitative process, and bounding analyses where appropriate, for assessing the risk impact of extending the surveillance frequency on SSCs for non-PRA modeled hazards and shutdown events.

Based on its review of the LAR, as supplemented, and the endorsed guidance in NEI 04-10, Revision 1, the NRC staff finds that (1) the licensees approach for considering impacts from internal fires, seismic hazards, high winds and tornadoes, external flooding, and other external hazards on STI extensions is consistent with the NRC staff endorsed guidance in NEI 04-10, Revision 1, and (2) the qualitative risk insights from these external hazards will be presented to the IDP for consideration. Therefore, the NRC staff concludes that internal fires, seismic hazards, high winds and tornadoes, external flooding, and other external hazards are appropriately considered in the licensees proposed SFCP.

The licensee stated that for assessing the shutdown risk, it will use the shutdown risk management program for implementation of Nuclear Management and Resources Council 91-06, Guidelines for Industry Actions to Assess Shutdown Management (Reference 18), for the proposed changes to surveillance frequencies under the SFCP. This is an acceptable approach in accordance with NEI 04-10, Revision 1.

Thus, the NRC staff finds that the licensees evaluation methodology ensures that the scope of 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 Consistent with NEI 04-10, Revision 1, upon implementation of the SFCP, the licensee stated that it will determine 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 common cause failure 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 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 the guidance provided in RG 1.200, Revision 2, and by sensitivity studies identified in NEI 04-10, Revision 1.

The licensee stated that it will perform quantitative evaluations of the impact of selected testing strategy (i.e., staggered testing or sequential testing) consistently with the guidance of NUREG/CR-6141, Handbook of Methods for Risk-Based Analyses of Technical Specifications (Reference 19) and NUREG/CR-5497, Common-Cause Failure Parameter Estimations (Reference 20), as discussed in NEI 04-10, Revision 1.

Thus, the NRC staff finds that through the application of NEI 04-10, Revision 1, the licensees PRA 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 the licensees PRAs assume all failures to be time-related because the breakdown between standby time-related contribution and a cyclic demand-related contribution is unknown. 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 the guidance in RG 1.177, Revision 1, Section 2.3.3, Probabilistic Risk Assessment Modeling, which permits separation of the failure rate contributions into demand and standby for evaluation of Supporting Requirements. According to the guidance in NEI 04-10, Revision 1, 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 and will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The process requires consideration of qualitative sources of information with regards 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 process is not reliant upon risk analyses as the sole basis for the proposed changes.

The potential beneficial risk impacts of reduced surveillance frequency, including reduced downtime, lesser potential for restoration errors, reduction of potential for test caused transients, and reduced test-caused wear of equipment, are identified qualitatively, but are conservatively not required to be quantitatively assessed.

Thus, the NRC staff finds that through the application of NEI 04-10, Revision 1, the licensee has employed reasonable assumptions with regard to extensions of STIs and is 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 NEI 04-10, Revision 1, provides that sensitivity studies be performed to assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact to the frequency of initiating events, and of any identified deviations from CC-II of the ASME/ANS PRA Standard, as endorsed in RG 1.200, Revision 2. 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. Guidance in Step 5 of NEI 04-10, Revision 1, specifies risk

sensitivity studies to be conducted by changing the unavailability terms for PRA basic events that correspond to SSCs being evaluated.

Consistent with NEI 04-10, Revision 1, monitoring and feedback of SSC performance once the revised surveillance frequencies are implemented will also be performed. Therefore, the NRC staff finds that through the application of NEI 04-10, Revision 1, the licensee has appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, 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.

3.1.4.6 Acceptance Guidelines The licensee states that it 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 the guidance provided in NEI 04-10, Revision 1, as required in the SFCP. Each individual change to surveillance frequency must show a risk impact below 10-6 per year for change to CDF, and below 10-7 per year for change to LERF.

These are consistent with the limits of RG 1.174, Revision 3, for very small changes in risk.

Where the RG 1.174, Revision 3, limits are not met, the process either considers revised surveillance frequencies, which are consistent with RG 1.174, Revision 3, or the process terminates without permitting the proposed changes. Where quantitative results are unavailable to permit comparison to acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to the surveillance frequency is negligible or zero. Otherwise, bounding quantitative analyses are required, which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174, Revision 3, 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 below 10-5 per year for change to CDF, and below 10-6 per year for change to LERF, and the total CDF and total LERF must be reasonably shown to be less than 10-4 per year and 10-5 per year, respectively. These values are consistent with the limits of RG 1.174, Revision 3, for acceptable changes in risk, as referenced by RG 1.177, Revision 1, for changes to surveillance frequencies.

Consistent with the NRCs safety evaluation for NEI 04-10, Revision 1, dated September 19, 2007, the SFCP provides that the licensee must calculate the change in risk from a baseline model utilizing failure probabilities based on the surveillance frequencies prior to implementation of the SFCP, compared to a revised model with failure probabilities based on changed surveillance frequencies. The NRC staff notes that the licensees SFCP includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (less than 5 x 10-8 CDF and 5 x 10-9 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 3, 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 compared to numerical acceptance guidelines. Post-implementation performance monitoring and feedback are also required to assure continued reliability of the components.

Based on its review, the NRC staff finds that the licensees application of NEI 04-10, Revision 1, provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4, Acceptance Guidelines for Technical Specification Changes, of RG 1.177, Revision 1.

Therefore, the NRC staff concludes that the proposed licensee methodology satisfies the fourth key safety principle of RG 1.177, Revision 1, by assuring any increase in risk is small consistent with the intent of the Commissions Safety Goal Policy Statement.

3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensees adoption of TSTF-425, Revision 3, requires application of NEI 04-10, Revision 1, in the SFCP. NEI 04-10, Revision 1, provides for performance monitoring of SSCs whose surveillance frequency has been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback include consideration of maintenance rule monitoring of equipment performance. In the event of a degradation of SSC performance, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may apply as part of the maintenance rule requirements.

Therefore, the NRC staff finds that 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, Maintenance Rule Control, 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 Deviations From TSTF-425 and Other Changes In Sections 2.2.1 and 2.2.2 of the LAR, as supplemented, the licensee identified variations and technical changes from the TSTF-425 template. The NRC staffs evaluation of those changes is discussed below.

The licensee stated that Wolf Creek SRs have numbers that differ from the corresponding Westinghouse Standard Technical Specifications (NUREG-1431)

TSTF-425 SRs. The NRC staff finds that the different SR numbering is acceptable because they are editorial in nature and do not substantively change the TS requirements.

The licensee stated that for NUREG-1431 SRs not contained in the Wolf Creek TSs, the corresponding mark-ups included in TSTF-425 for these SRs are not applicable to Wolf Creek. The NRC staff finds this variation acceptable because it is editorial in nature and does not substantively change the TS requirements.

The TSTF-425 TS Section 5.5.18 insert for the new SFCP references NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies. The licensee is adopting this new program as TS Section 5.5.19 and references NEI 04-10, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies. The NRC staff finds this variation acceptable because it is editorial in nature and continues to meet the intent of TSTF-425, Revision 3.

The licensee proposed to replace the TS Bases insert, The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program, which is provided in TSTF-425, Revision 3. The licensee proposed that the new text reads, The Surveillance Frequency is controlled under the Surveillance Frequency Control Program, or The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program, as appropriate. In a letter dated April 14, 2010 (Reference 21), the NRC staff agreed that the insert applies to surveillance frequencies that are relocated and subsequently evaluated and changed in accordance with the SFCP but does not apply to frequencies relocated to the SFCP, but not changed. The NRC staff finds the licensee proposed variation acceptable because it is editorial in nature and continues to meet the intent of TSTF-425, Revision 3.

The licensee identified several Wolf Creek plant-specific SRs with fixed periodic frequencies that are not contained in NUREG-1431, and therefore, are not included in TSTF-425. The licensee assessed these SRs and determined that the relocation of the frequencies for these SRs is consistent with TSTF-425. The licensee requested that these surveillances be controlled under the SFCP.

The SFCP provides administrative controls such that surveillances related to testing, calibration, and inspection are conducted at a frequency to assure that the necessary quality of the systems and components is maintained, the facility operation will be within safety limits, and that the limiting conditions for operation will be met. The SFCP provides that changes to the frequencies be evaluated using the methodology and probabilistic risk guidelines provided in NEI 04-10, Revision 1, as approved by NRC letter dated September 19, 2007. The NEI 04-10, Revision 1, methodology includes qualitative considerations, risk analyses, sensitivity studies, and bounding analyses, as necessary, and recommends monitoring of the performance of SSCs for which frequencies are changed to assure that reduced testing does not adversely impact the SSCs.

For example, the licensee included Wolf Creek TS 5.5.13 in the scope of this amendment. This TS has a periodic frequency that was not identified for relocation in TSTF-425, Revision 3. TS 5.5.13.c is revised as follows (deleted text in strikeout and added text in italics):

Total particulate concentration of the fuel oil is 10 mg/l [milligrams per liter] when tested every 31 days in accordance with ASTM [American Society of Testing & Materials] D-2276, Method, at a Frequency in accordance with the Surveillance Frequency Control Program.

The NRC staff evaluated the proposed deviations for the identified plant-specific TS SRs listed in Attachment 7 of the LAR and determined that the SR frequencies do not meet any of the exclusion criteria in TSTF-425, Revision 3, and that the frequencies are fixed periodic frequencies. Therefore, the NRC staff finds that relocation of these plant-specific TS SR frequencies to the SFCP is acceptable.

The licensee proposed to revise TS 5.5.18.d. The proposed revision would delete of 18 months on a STAGGERED TEST BASIS and replace it with in accordance with the Surveillance Frequency Control Program. TSTF-425 includes the relocation of the frequency for the NUREG-1431, SR 3.7.10.4, associated with verifying that one Control

Room Emergency Ventilation System (CREVS) train can maintain a positive pressure relative to adjacent area(s). The licensee proposed to adopt the frequency change identified for the NUREG-1431, SR 3.7.10.4, in TSTF-425 as the Wolf Creek TS 5.5.18.d frequency.

The NRC staff finds this variation acceptable because the frequency for Wolf Creek TS SR 3.7.10.4 has been moved to TS 5.5.18.d with the facilitys adoption of TSTF-448, Control Room Habitability (Reference 22). The frequency located in TS 5.5.18.d is a periodic frequency, does not meet the scope exclusion criteria, and is consistent with the intent of TSTF-425. SR 3.7.10.4 was revised under TSTF-448, to perform control room envelope unfiltered air in-leakage testing in accordance with the Control Room Envelope Habitability Program. The licensee adopted TSTF-448 and designated the Control Room Envelope Habitability Program as TS 5.5.18.

Due to the relocation of SR frequencies and replacing the frequencies with the statement, In accordance with the Surveillance Frequency Control Program, there are multiple SRs that moved to the next page in the TSs. The licensee made various formatting changes, such as inserting new pages due to text rollover and changes to the table of contents. The NRC staff reviewed the changes and determined that they are editorial in nature and are therefore acceptable.

3.3 Addition of Surveillance Frequency Control Program to TS Section 5 The licensee proposed including the SFCP and specific requirements into Wolf Creek TS Section 5.5.19, Programs and Manuals, 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 Technical Specifications Initiative 5b, 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 proposed program is consistent with the model application of TSTF-425 and, therefore, the NRC staff concludes that it is acceptable.

3.4 Summary and Technical Conclusions The NRC staff has reviewed the licensees proposed relocation of certain surveillance frequencies to a licensee-controlled document and its proposed control of changes to surveillance frequencies in accordance with a new program, the SFCP, identified in the administrative controls of the TSs. The SFCP and the new TS Section 5.5.19 reference NEI 04-10, Revision 1, 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 the TS to a licensee-controlled document, provided those frequencies are changed in accordance with NEI 04-10, Revision 1, which is specified in the administrative controls of the TS.

The licensees proposed adoption of TSTF-425, Revision 3, and the risk-informed methodology of NEI 04-10, Revision 1, as referenced in the Administrative Controls section of the TSs, satisfies the key principles of risk-informed decisionmaking applied to changes to TS as delineated in RG 1.177, Revision 1, and RG 1.174, Revision 3, 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 Commissions Safety Goal Policy Statement; and The impact of the proposed change is performance monitoring using measurement strategies.

The NRC staff finds that with the proposed relocation of surveillance frequencies to an owner-controlled document and administratively controlled in accordance with the TS SFCP, the licensee continues to meet the regulatory requirement of 10 CFR 50.36, specifically, 10 CFR 50.36(c)(3), Surveillance Requirements.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Kansas State official was notified of the proposed issuance of the amendment on December 18, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes SRs. 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, published in the Federal Register on June 16, 2020 (85 FR 36436), and there has been no public comment on such finding. 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 Commissions 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. Smith, S. L., Wolf Creek Nuclear Operating Corporation, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-482: Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425), dated April 27, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20119A873).

2. Smith, S. L., Wolf Creek Nuclear Operating Corporation, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-482: Response to Request for Additional Information Re Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TSTF-425), dated October 26, 2020 (ADAMS Accession No. ML20300A569).

3. Nuclear Energy Institute, Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies, NEI 04-10, Revision 1, dated April 2007 (ADAMS Accession Number ML071360456).

4. Nieh, H. K., U.S. Nuclear Regulatory Commission, letter to B. Bradley, Nuclear Energy Institute, Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04 10, Revision 1, Risk-Informed Technical Specification Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies (TAC NO. MD6111), dated September 19, 2007 (ADAMS Accession No. ML072570267).

5. Technical Specifications Task Force, letter and enclosure to U.S. Nuclear Regulatory Commission, Transmittal of TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b, dated March 18, 2009 (ADAMS Accession No. ML090850642).

6. 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 3, dated January 2018 (ADAMS Accession No. ML17317A256).

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

8. 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, dated March 2009 (ADAMS Accession No. ML090410014).

9. U.S. Nuclear Regulatory Commission, Standard Technical Specifications -

Westinghouse Plants, NUREG-1431, Revision 4, Volume 1, Specifications, dated April 2012 (ADAMS Accession No. ML12100A222).

10. U.S. Nuclear Regulatory Commission, Standard Technical Specifications -

Westinghouse Plants, NUREG-1431, Revision 4, Volume 2, Bases, dated April 2012 (ADAMS Accession No. ML12100A228).

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

PRA Standard, ASME/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, dated February 2009, New York, NY.

12. Nuclear Energy Institute, Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, NEI 00-02, Revision 1, dated May 2006 (ADAMS Accession No. ML061510623).

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

14. U.S. Nuclear Regulatory Commission, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Generic Letter No. 88-20, Supplement 4, dated June 28, 1991 (ADAMS Accession No. ML031150485).

15. Electric Power Research Institute, Fire-Induced Vulnerability Evaluation (FIVE), TR-100370 September 1993.

16. Smith R. A., Wolf Creek Nuclear Operating Corporation, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-482 - Wolf Creek Nuclear Operating Corporations Seismic Hazard and Screen Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 31, 2014 (ADAMS Accession No. ML14097A020).

17. Broschak, J. P., Wolf Creek Nuclear Operating Corporation, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-482: Wolf Creek Nuclear Operating Corporations Flood Hazard Reevaluation Report in Response to NRC Request for Information Regarding Recommendation 2.1, Flooding, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 10, 2014 and Flood Hazard Reevaluation Report, Revision 0, dated February 10, 2014 (ADAMS Accession No. ML14077A280 and ML14077A281, respectively).

18. Nuclear Management Resources Council Guidelines for Industry Actions to Assess Shutdown Management, NUMARC 91-06, dated December 1991 (ADAMS Accession No. ML14365A203).

19. Brookhaven National Laboratory, Avaplan Oy, and Science Applications International Corporation, Handbook of Methods for Risk-Based Analyses of Technical Specifications, NUREG/CR-6141, dated December 1994 (ADAMS Accession No. ML093090361).

20. U.S. Nuclear Regulatory Commission, NUREG/CR-5497, Common-Cause Failure Parameter Estimations, October 1998 (ADAMS Accession No. ML070580356).

21. Bowman, E. E., U.S. Nuclear Regulatory Commission, letter to Technical Specifications Task Force, Notification of Issue with NRC-Approved Technical Specifications Task Force (TSTF) Traveler TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b, dated April 14, 2010 (ADAMS Accession No. ML100990099).

22. Singal, B. K., U.S. Nuclear Regulatory Commission, letter to Muench, R. A., Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station - Issuance of Amendment Re: Application to Revise Technical Specifications Regarding Control Room Envelope Habitability in Accordance with TSTF-448, Revision 3 (Tac No. MD7791), dated December 24, 2008 (ADAMS Accession No. ML083390833).

Principal Contributors: C. Moulton D. Wu J. Circle T. Sweat Date: April 8, 2021

ML21053A117 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC NAME SLee PBlechman VCusumano DATE 03/29/2021 03/29/2021 02/17/2021 OFFICE NRR/DRA/APLB/BC(A) NRR/DEX/EEEB/BC NRR/DEX/EMIB/BC NAME SVasavada BTitus ABuford (TScarbrough for)

DATE 12/11/2020 01/22/2021 02/11/2021 OFFICE NRR/DSS/SCPB/BC NRR/DSS/SNSB/BC NRR/DNRL/NCSG/BC NAME BWittick SKrepel SBloom DATE 02/02/2021 01/22/2021 01/22/2021 OFFICE OGC NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME AGhoshNaber JDixon-Herrity SLee DATE 03/24/2021 04/08/2021 04/08/2021

v t e Letter Sequence Approval
EPID:L-2020-LLA-0091, TSTF-425 (Approved, Closed)
Initiation Request Acceptance Administration Questions 17 September 2020 Answers 26 October 2020 Results Approval Other:
MONTHYEAR2020-05-22 [Table View]

ML21053A117

Text

SUBJECT:

Dear Mr. Reasoner:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

2.1 Background

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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