ML111661877

Issuance of Amendment No. 202, Revise Technical Specifications to Adopt TSTF-425 Rev. 3, Relocate Surveillance Frequencies to Licensee Control-Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b

ML111661877
Person / Time
Site:	Callaway
Issue date:	07/29/2011
From:	Thadani M Plant Licensing Branch IV
To:	Heflin A Union Electric Co
Gibson, Lauren, NRR/DORL/LPL4, 415-1056
References
TAC ME4506
Download: ML111661877 (223)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 July 29, 2011 Mr. Adam C. Heflin Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. 80x 620 Fulton, MO 65251 SU8JECT: CALLAWAY PLANT, UNIT 1 -ISSUANCE OF AMENDMENT RE: ADOPTION OF TSTF-425, REVISION 3, "RELOCATE SURVEILLANCE FREQUENCIES TO LICENSEE CONTROL - RITSTF INITIATIVE 58" (TAC NO. ME4506)

Dear Mr. Heflin:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 202 to Facility Operating License No. NPF-30 for the Callaway Plant, Unit 1.

The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated August 5, 2010, as supplemented by letters dated March 23, May 3, and July 25, 2011.

The amendment revises the TSs by relocating specific surveillance frequencies to a licensee controlled program with the guidance of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies." The amendment adopts NRC-approved Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF

[Risk-Informed TSTF] Initiative 5b." When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TSs.

A. Heflin - 2 A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Mohan C. Thadani, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosures:

1. Amendment No. 202 to NPF-30

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. 50-483 AMENDMENT TO FACILITY OPERAliNG LICENSE Amendment No. 202 License No. NPF-30

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

A. The application for amendment by Union Electric Company (UE, the licensee),

dated August 5,2010, as supplemented by letters dated March 23, May 3, and July 25, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regUlations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

-2

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 Facility Operating License No. NPF-30 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. 202 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This amendment is effective as of its date of issuance, and shall be implemented within 180 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License No. NPF-30 and Technical Specifications Date of Issuance: Jul y 29, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 202 FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of the Facility Operating License No. NPF-30 and 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.

Facility Operating License REMOVE INSERT

-3 Technical Specifications REMOVE INSERT REMOVE INSERT TOe 1 to Toe 4 Toe 1 to Toe 4 3.5-2 to 3.5-12 3.5-2 to 3.5-13 3.1-1 3.1-1 3.6-6 3.6-6 3.1-3 3.1-3 3.6-13 to 3.6-17 3.6-13 to 3.6-17 3.1-10 3.1-10 3.6-19 to 3.6-21 3.6-19 to 3.6-22 3.1-12 3.1-12 3.7-8 3.7-8 3.1-14 to 3.1-15 3.1-14 to 3.1-15 3.7-11 3.7-11 3.1-20 to 3.1-22 3.1-20 to 3.1-23 3.7-17 to 3.7-18 3.7-17 to 3.7-18 3.2-3 3.2-3 3.7-20 3.7-20 3.2-5 3.2-5 3.7-22 3.7-22 3.2-8 3.2-8 3.7-25 3.7-25 3.2-9 3.2-9 3.7-27 3.7-27 3.2-13 3.2-13 3.7-31 3.7-31 3.3-12 to 3.3-77 3.3-12 to 3.3-83 3.7-34 3.7-34 3.4-2 to 3.4-3 3.4-2 to 3.4-3 3.7-38 3.7-38 3.4-5 to 3.4-6 3.4-5 to 3.4-6 3.7-40 3.7-40 3.4-9 3.4-9 3.7-42 3.7-42 3.4-11 3.4-11 3.7-45 3.7-45 3.4-14 3.4-14 3.7-47 3.7-47 3.4-16 3.4-16 3.8-6 to 3.8-39 3.8-6 to 3.8-41 3.4-18 3.4-18 3.9-2 3.9-2 3.4-24 3.4-24 3.9-4 3.9-4 3.4-28 to 3.4-29 3.4-28 to 3.4-29 3.9-6 3.9-6 3.4-31 3.4-31 3.9-8 3.9-8 3.4-34 to 3.4-35 3.4-34 to 3.4-35 3.9-10 3.9-10 3.4-39 3.4-39 3.9-12 to 3.9-13 3.9-12 to 3.9-14 3.4-41 to 3.4-44 3.4-41 to 3.4-45 5.0-21 5.0-21

-3 (4) UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) UE, 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 license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth 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 UE 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. 202 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Environmental Qualification (Section 3.11! SSER #3)**

Deleted per Amendment No. 169.

• Amendments 133, 134, & 135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.

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

Amendment No. 202

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 3.1.9 RCS Boron Limitations < 500°F ....................................................... 3.1-22 3.2 POWER DISTRIBUTION LIMITS .........................................................3.2-1 3.2.1 Heat Flux Hot Channel Factor (Fa(Z>>

(Fa Methodology) ......................................................................3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor ................................... 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 .......................................................................... 3.3-27 3.3.3 Post Accident Monitoring (PAM) Instrumentation ............................3.3-51 3.3.4 Remote Shutdown System ............................................................. 3.3-56 CALLAWAY PLANT Amendment 202

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation .......................................................................... 3.3-59 3.3.6 Containment Purge Isolation Instrumentation ................................. 3.3-61 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation Instrumentation ......................................................... 3.3-67 3.3.8 Emergency Exhaust System (EES) Actuation Instrumentation ......................................................................... 3.3-73 3.3.9 Boron Dilution Mitigation System (BDMS) ...................................... 3.3-79 3.4 REACTOR COOLANT SYSTEM (RCS) ................................................ 3.4-1 3.4.1 RCSPressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ................................................... 3.4-1 3.4.2 RCS Minimum Temperature for Criticality ....................................... 3.4-3 3.4.3 RCS Pressure and Temperature (prr) Limits .................................. 3.4-4 3.4.4 RCS Loops - MODES 1 and 2 ......................................................... 3.4-6 3.4.5 RCS Loops - MODE 3 ..................................................................... 3.4-7 3.4.6 RCS Loops - MODE 4 ..................................................................... 3.4-10 3.4.7 RCS Loops - MODE 5, Loops Filled ................................................ 3.4-12 3.4.8 RCS Loops - MODE 5, Loops Not Filled ......................................... 3.4-15 3.4.9 Pressurizer ....................................................................................... 3.4-17 3.4.10 Pressurizer Safety Valves ................................................................ 3.4-19 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) ...................... 3.4-21 3.4.12 Cold Overpressure Mitigation System (COMS) ............................... 3.4-25 3.4.13 RCS Operational LEAKAGE ............................................................ 3.4-30 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ................................ 3.4-32 3.4.15 RCS Leakage Detection Instrumentation ........................................ 3.4-36 3.4.16 RCS Specific Activity ....................................................................... 3.4-40 3.4.17 Steam Generator (SG) Tube Integrity .............................................. 3.4-43 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 ..................................................................... 3.6-3 3.6.3 Containment Isolation Valves ......................................................... 3.6-7 CALLAWAY PLANT 2 Amendment 202

TABLE OF CONTENTS 3.6 CONTAINMENT SYSTEMS (continued) 3.6.4 Containment Pressure ..................................................................... 3.6-16 3.6.5 Containment Air Temperature ......................................................... 3.6-17 3.6.6 Containment Spray and Cooling Systems ...................................... 3.6-18 3.6.7 Recirculation Fluid pH Control (RFPC) System ............................... 3.6-21 3.7 PLANT SYSTEMS ................................................................................. 3.7-1 3.7.1 Main Steam Safety Valves (MSSVs) ............................................... 3.7-1 3.7.2 Main Steam Isolation Valves (MSIVs). Main Steam Isolation Valve Bypass Valves (MSIVBVs), and Main Steam Low Point Drain Isolation Valves (MSLPDIVs) ................................. 3.7-5 3.7.3' Main Feedwater Isolation Valves (MFIVs), Main Feedwater Regulating Valves (MFRVs). and Main Feedwater Regulating Valve Bypass Valves (MFRVBVs) .............................................. 3.7-9 3.7.4 AtmospheriC Steam Dump Valves (ASDs) ...................................... 3.7-12 3.7.5 Auxiliary Feedwater (AFW) System ................................................. 3.7-15 3.7.6 Condensate Storage Tank (CST) .................................................... 3.7-19 3.7.7 Component Coaling Water (CCW) System ..................................... 3.7-21 3.7.8 Essential Service Water System (ESW) .......................................... 3.7-23 3.7.9 Ultimate Heat Sink (UHS) ................................................................ 3.7-26 3.7.10 Control Room Emergency Ventilation System (CREVS) ................. 3.7-28 3.7.11 Control Room Air Conditioning System (CRACS) ........................... 3.7-32 3.7.12 Not Used .......................................................................................... 3.7-35 3.7.13 Emergency Exhaust System (EES) ................................................. 3.7-36 3.7.14 Not Used .......................................................................................... 3.7-39 3.7.15 Fuel Storage Pool Water Level ........................................................ 3.7-40 3.7.16 Fuel Storage Pool Boron Concentration .......................................... 3.7-41 3.7.17 Spent Fuel Assembly Storage ......................................................... 3.7-43 3.7.18 Secondary Specific Activity .............................................................. 3.7-45 3.7.19 Secondary System Isolation Valves (SSIVs) ................................... 3.7-46 3.8 ELECTRICAL POWER SYSTEMS ........................................................ 3.8-1 3.8.1 AC Sources - Operating ................................................................... 3.8-1 3.8.2 AC Sources - Shutdown .................................................................. 3.8-18 3.8.3 Diesel Fuel Oil, Lube 011, and Starting Air ....................................... 3.8-22 3.8.4 DC Sources - Operating .................................................................. 3.8-25 3.8.5 DC Sources - Shutdown .................................................................. 3.8-28 3.8.6 Battery Cell Parameters ................................................................... 3.8-30 3.8.7 Inverters - Operating ........................................................................ 3.8-34 3.8.8 Inverters - Shutdown ........................................................................ 3.8-36 3.8.9 Distribution Systems - Operating ..................................................... 3.8-38 CALLAWAY PLANT 3 Amendment 202

TABLE OF CONTENTS 3.8 ELECTRICAL POWER SYSTEMS (continued) 3.8.10 Distribution Systems - Shutdown ..................................................... 3.8-40 3.9 REFUELING OPERATIONS ................................................................. 3.9-1 3.9.1 Boron Concentration ........................................................................ 3.9-1 3.9.2 Unborated Water Source Isolation Valves ....................................... 3.9-3 3.9.3 Nuclear Instrumentation ................................................................... 3.9-5 3.9.4 Containment Penetrations ............................................................... 3.9-7 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level ................................................... 3.9-9 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level .................................................... 3.9-11 3.9.7 Refueling Pool Water Level ............................................................. 3.9-14 4.0 DESIGN FEATURES ............................................................................... 4.0-1 4.1 Site Location .................................................................................... 4.0-1 4.2 Reactor Core ................................................................................... 4.0-1 4.3 Fuel Storage .................................................................................... 4.0-1 5.0 ADMINISTRATIVE CONTROLS ........................................................... 5.0-1 5.1 Responsibility ................................................................................... 5.0-1 5.2 Organization .................................................................................... 5.0-2 5.3 Unit Staff Qualifications ................................................................... 5.0-4 5.4 Procedures ...................................................................................... 5.0-5 5.5 Programs and Manuals ..................................................................... 5.0-6 5.6 Reporting Requirements .................................................................. 5.0-22 5.7 High Radiation Area ......................................................................... 5.0-27 CALLAWAY PLANT 4 Amendment 202

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

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

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

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within limits. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.1-1 Amendment No. 202

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 % L'lk/k of Once prior to predicted values. entering MODE 1 after each refueling

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NOTE-------

Only required after 60 EFPD In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.1-3 Amendment No. 202

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 moving each rod not fully inserted in the core with the

~ 10 steps in either direction. Surveillance Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully Prior to reactor withdrawn position, is:5 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.

CALLAWAY PLANT 3.1-10 Amendment No. 202

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 specified in the COLR. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.1-12 Amendment No. 202

Control Bank Insertion Limits 3.1.6 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. Control bank sequence or B.1.1 Verify SDM to be within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> overlap limits not met. the limits provided in the COLR.

OR B.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit.

AND B.2 Restore control bank 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> sequence and overlap to within limits.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank position is within Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the limits specified in the COLR. prior to achieving criticality SR 3.1.6.2 Verify each control bank insertion is within the limits In accordance specified in the COLR. with the Surveillance Frequency Control Program (Continued)

CALLAWAY PLANT 3.1-14 Amendment No. 202

Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.6.3 Verify sequence and overlap limits specified in the In accordance COLR are met for control banks not fully withdrawn with the from the core. Surveillance Frequency Control Program CALLAWAY PLANT 3.1-15 Amendment No. 202

PHYSICS TESTS Exceptions - MODE 2 3.1.8 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION 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 SR PHYSICS TESTS 3.3.1.7, 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 temperature is ~ 541°F. with the Surveillance Frequency Control Program SR 3.1.8.3 Verify THERMAL POWER is :s; 5% RTP. In accordance with the Surveillance Frequency Control Program (contInued)

CALLAWAY PLANT 3.1-20 Amendment No. 202

PHYSICS TESTS Exceptions - MODE 2 3.1.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.8.4 Verify SDM is within limits provided in the COLR. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.1-21 Amendment No. 202

RCS Boron Limitations < SOO°F 3.1.9 3.1 REACTIVITY CONTROL SYSTEMS 3.1.9 RCS Boron Limitations < SOO°F LCO 3.1.9 The boron concentration of the Reactor Coolant System (RCS) shall be greater than the all rods out (ARO) critical boron concentration.

APPLICABILITY: MODE 2 with keff < 1.0 with any RCS cold leg temperature < SOO°F and with Rod Control System capable of rod withdrawal, MODE 3 with any RCS cold leg temperature < SOO°F and with Rod Control System capable of rod withdrawal, MODES 4 and S with Rod Control System capable of rod withdrawal.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A RCS boron concentration A. 1 Initiate boration to Immediately not within limit. restore RCS boron concentration to within limit.

OR A2 Initiate action to place Immediately the Rod Control System in a condition incapable of rod withdrawal.

OR A3 ------------ NOTE ----------

Not applicable in MODES 4 and S.

Initiate action to increase Immediately all RCS cold leg temperatures to S SOO°F.

CALLAWAY PLANT 3.1-22 Amendment No. 202 I

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 ARO critical boron concentration. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.1-23 Amendment No. 202

Fa(Z) (Fa Methodology) 3.2.1 SURVEILLANCE REQUIREMENTS

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NOT E ----------------------------------------------------------

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

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify F8(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75% RTP Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions after exceeding, by

~ 10% RTP, the THERMAL POWER at which F8(Z) was last verified In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.2-3 Amendment No. 202

Fa(Z) (Fa Methodology) 3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued) Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions after exceeding, by

~ 10% RTP, the THERMAL POWER at which was last verified In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.2-5 Amendment 202

SURVEILLANCE REQUIREMENTS

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NOT E ----------------------------------------------------------.

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

SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify F t1NH is within limits specified in the COLR. Once after each refueling prior to THERMAL POWER exceeding 75% RTP In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.2-8 Amendment No. 202

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.

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NOT E -------------------------------------------

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 COMPLETION CONDITION REQUIRED ACTION 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 channel. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.2-9 Amendment No. 202

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

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER s 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER

> 75% RTP.

Verify QPTR is within limit using power distribution In accordance measurement information. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.2-13 Amendment No. 202

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

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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 SR 3.3.1.2 --------------------------------- NOTE ---------------------------

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

Compare results of calorimetric heat balance In accordance calculation to power range channel output. Adjust with the power range channel output if calorimetric heat Surveillance balance calculation results exceed power range Frequency channel output by more than +2% RTP. Control Program SR 3.3.1.3 --------------------------------- NOTE -------.-------------------

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

Compare results of the incore power distribution In accordance measurements to Nuclear Instrumentation System with the (NIS) AFD. Adjust NIS channel if absolute difference Surveillance is ~2%. Frequency Control Program (continued)

CALLAWAY PLANT 3.3-12 Amendment No. 202

RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY 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 SR 3.3.1.6 --------------------------------- NOTE ---------------------------

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

Calibrate excore channels to agree with incore power In accordance distribution measurements. with the Surveillance Frequency Control Program (continued)

CALLAWAY PLANT 3.3-13 Amendment No. 202

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS continued)

SURVEILLANCE FREQUENCY SR 3.3.1.7 --------------------------------- NOTE ---------------------------

1. Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 CALLAWAY PLANT 3.3-14 Amendment No. 202

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 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 Prior to reactor startup 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reducing power below P-10 for power and intermediate instrumentation Four hours after reducing power below P-6 for source range instrumentation CALLAWAY PLANT 3.3-15 Amendment No. 202

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.8 (continued)

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 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 (contlnueo)

CALLAWAY PLANT 3.3-16 Amendment 202

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.11 --------------------------------- NOTE ---------------------------

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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after achieving equilibrium conditions with THERMAL POWER 2 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 SR 3.3.1.14 --------------------------------- NOTE ---------------------------

Verification of setpoint is not required.

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

CALLAWAY PLANT 3.3-17 Amendment No. 202

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE I I=REQUENCY 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 CALLAWAY PLANT 3.3-18 Amendment No. 202

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (PAGE 1 OF 8)

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

1. Manual Reactor 1,2 2 B SP 3 3.1 1t1 NA Trip 3(b), 4(b), Sib) 2 C SF': 3.3.1 ~4 NA

2. Power Range Neutron Flux

a. High 1,2 4 D SF< 11 S; 112.3% RTP SR 3.31 SR 3.3.1 '1 SR3.3.111 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 SR3.3.116 2(h),3(i) 4 Y,Z SR3.3.1.1  :::;28.3% RTP SR 331.8 SR 3.3.1!

SR 33.1.1

3. Power Range 1,2 4 E SR :33.1.7 6.3 % RTP Neutron Flux SR 3 3.1 1 with time Rate - High SH 3.31.16 constant Positive Rate 2 sec

4. Intermediate 1(e),2(d) 2 F, G SR 3 3.1.1  :::; 35.3% RTP Range Neutron SR 3318 Flux SF{ 3.31.11 (a) The Allowable Value defines the limiting safety system setting except for Trip Functions 14.a and 14.b (the Nominal Trip Setpoint defines the limiting safety system setting for these Trip Functions). See the Bases for the Nominal Trip Setpoints.

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

(c) Below the P-l0 (Power Range Neutron Flux) interlock.

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

(f) With kef! ~ 1.0.

(h) With kef! < 1.0, and all RCS cold leg temperatures ~ 500°F, and RCS boron concentration s; the 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?: 500°F, and RCS boron concentration s; the ARO critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted CALLAWAY PLANT 3,3-19 Amendment No. 202 I

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (PAGE 2 OF 8)

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

5. Source Range 2(e) 2 I, J SR3.3.1.1 1.6 E5 cps Neutron Flux SR:'U.1.8 SR:U 1.11 3(b), 4(b), 5(b) 2 J, K SR 3.3.1.1 ~ 1.6 E5 cps SR 3.~l.1 7 Sl~ 3.3.1.11

6. Overtemperature 1,2 4 E SR 3.31.1 Refer to Note 1 ilT SR 3.~j.1.3 (Page 3.3-23)

SR 3.3.1.6 SR 3.317 SR 3.3.1.10 SRJ.3116

7. Overpower ilT 1,2 4 E SR 3.1.1 Refer to Note 2 Sf~ 3.3.1.7 (Page 3.3-24)

SR 3.3.1.10 SR 3.3.1 16

8. Pressurizer Pressure

a. Low 1(g) 4 M SR 3.:1.1.1 21874 psig SR 3.3.1.7 SR 3.3.1.10 SR33.116

b. High 1,2 4 E Sf'\ 331 1  :> 2393 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.1G (a) The Allowable Value defines the limiting safety system setting except for Trip Functions 14.a and 14.b (the Nominal Trip Setpoint defines the limiting safety system setting for these Trip Functions). See the Bases for the Nominal 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.

CALLAWAY PLANT 3.3-20 Amendment No. 202 I

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (PAGE 3 OF 8)

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

9. Pressurizer Water 3 M SR 3.3.1 1 s; 93.8% of Level- High SR3.3.1.7 instrument SR 3.3.1.10 span

10. Reactor Coolant 3 per loop M SR 3.3.1.1 ~ 88.8% of Flow - Low SR 3.3.17 indicated loop SR ~U.1.10 flow SR 3.3.116

11. Not Used

12. Undervoltage 2/bus M SR 3.3.1.9 ~ 10105 Vac RCPs SR 3.3.1.10 SR 3.3.116

13. Underfrequency 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) Water Level Low-Low(l)

a. Steam 1,2 4 per SG E SR 3 3.1 1 ~ 20.6% (q) of Generator SR 3.3.1.7 Narrow Range Water Level SR 3.3.1.10 Instrument Low-Low SR 3.3.1.16 Span (Adverse Containment Environment)

(a) The Allowable Value defines the limiting safety system setting except for Trip Functions 14.a and 14.b (the Nominal Trip Setpoint defines the limiting safety system setting for these Trip Functions). See the Bases for the Nominal Trip Setpoints.

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

(I) The applicable MODES for these channels in Table 3.3.2-1 are more restrictive.

(m) Not used.

(q) 1. If the as-found instrument channel setpoint is conservative with respect to the Allowable Value, but outside its as-found test acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value that is within the as-left setpoint tolerance band on either side of the Nominal Trip Setpoint, or to a value that is more conservative than the Nominal Trip Setpoint; otherwise, the channel shall be declared inoperable. The Nominal Trip Setpoints and the methodology used to determine the as-found test acceptance criteria band and the as-left setpoint tolerance band shall be specified in the Bases CALLAWAY PLANT 3.3-21 Amendment No. 202 I

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (PAGE 4 OF 8)

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

14. Steam Generator (SG) Water Level Low-Low(l)

b. Steam 4 per SG E SR3.3.1.1 ~ 16.6% (q) of Generator SR 3.3.17 Narrow Range Water Level SR 3.3.1 10 Instrument Low-Low SR 3.3.1 16 Span (Normal Containment Environment)

c. Not used.

d. Containment 1,2 4 x SR:U.11 ~ 2.0 psig Pressure SR :U.1.7 Environmental SR 3.3.1.10 Allowance SR 3.3.1 1G Modifier

15. Not Used (a) The Allowable Value defines the limiting safety system setting except for Trip Functions 14.a and 14.b (the Nominal Trip Setpoint defines the limiting safety system setting for these Trip Functions). See the Bases for the Nominal Trip Setpoints.

(I) The applicable MODES for these channels in Table 3.3.2-1 are more restrictive.

(n) Not used.

(0) Not used.

(p) Except when the Containment Pressure - Environmental Allowance Modifier channels in the same protection sets are tripped.

(q) 1. If the as-found instrument channel setpoint is conservative with respect to the Allowable Value, but outside its as found test acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value that is within the as-left setpoint tolerance band on either side of the Nominal Trip Setpoint, or to a value that is more conservative than the Nominal Trip Setpoint; otherwise, the channel shall be declared inoperable. The Nominal Trip Setpoints and the methodology used to determine the as-found test acceptance criteria band and the as-left setpoint tolerance band shall be specified in the Bases CALLAWAY PLANT 3.3-22 Amendment No. 202 I

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (PAGE 5 OF 8)

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

16. Turbine Trip

a. Low Fluid 3 o SR 3.:1.1.10 539.42 psig Oil Pressure SR 33.1.15

b. Turbine 4 P SR3.3.1.10 ~ 1% open Stop Valve sr~ :331 15 Closure

17. Safety Injection 1,2 2 trains Q SR 3.3.1 14 NA (SI) Input from Engineered Safety Feature Actuation System (ESFAS)

18. Reactor Trip System Interlocks

a. Intermediate 2 S SR 3.3.1.11 6E-11 amp Range SR 3.3.113 Neutron Flux, P-6

b. Low Power 1 per train T SR 33.1.5 NA Reactor Trips Block, P-7

c. Power 4 T SR 3.3.111  ::;; 51,3% RTP Range SR 3.3.1 13 Neutron Flux, P-8

d. Power 4 T SR 3.3.1.11 <:; 53,3% RTP Range SR:U.113 Neutron Flux, P-9 (a) The Allowable Value defines the limiting safety system setting except for Trip Functions 14.a and 14.b (the Nominal Trip Setpoint defines the limiting safety system setting for these Trip Functions). See the Bases for the Nominal Trip Setpoints, (e) Below the P-6 (Intermediate Range Neutron Flux) interlock.

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

CALLAWAY PLANT 3.3-23 Amendment No. 202 I

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (PAGE 6 OF 8)

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

18. Reactor Trip System Interlocks

e. Power Range 1,2 4 S SR:U.l.11  ? 6.7% RTP Neutron Flux, SH:U1.13 and P-10 ~ 12.4% RTP

f. Turbine 2 T SR :)3.1.10 $12.4%

Impulse SR 3.3.1.13 turbine power Pressure, P-13

19. Reactor Trip 1,2 2 trains R SR 3.3.1.4 NA Breakers (RTBs)(k) 3(b), 4(b), 5(b) 2 trains C SR 331.4 NA

20. Reactor Trip 1,2 1 each per U SR :331.4 NA Breaker RTB Undervoltage and Shunt Trip 3(b), 4(b), 5(b) 1 each per C SR :J.3.1.4 NA Mechanisms(k) RTB

21. Automatic Trip 1,2 2 trains Q SR 3.3.15 NA Logic 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 except for Trip Functions 14.a and 14.b (the Nominal Trip Setpoint defines the limiting safety system setting for these Trip Functions). See the Bases for the Nominal 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.

CALLAWAY PLANT 3.3-24 Amendment No. 202 I

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (page 7 of8)

Reactor Trip System Instrumentation Note 1: Overtemperature ~ T The Overtemperature ~T Function Allowable Value shall not exceed the following setpoint by more than 1.23% of ~ T span (1.85% RTP).

flT(1+'1 )( 1 -1~flTo{K1-K2[(1+'4S)T S

1 rJ+K 3 (p-P') f 1 (fll)}

(1 + '2S) 1 + '3SJ (1 + '5S) 1 + '6S Where: ~T is measured RCS ~ T, of.

~ To is the indicated t. T at RTP, of.

s is the Laplace transform operator, sec* 1.

T is the measured RCS average temperature, OF.

T' is the nominal Tavg at RTP, $ oOF.

P is the measured pressurizer pressure, psig.

P' is the nominal RCS operating pressure = 0 psig.

K1 = 0 K2 = °/oF K3 */psig

'1 ~

• sec '2 $ 0 sec '3

• sec

'4 ~

• sec '5 $ 0 sec '6

• sec

• {*% + (q! - qb)} when qt - qb <: * %RTP 0% ofRTP when' %RTP $ ql qb $ * %RTP

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

The values denoted with' are specified in the COLR.

CALLAWAY PLANT Amendment 202 I

RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (page 8 of 8)

Reactor Trip System Instrumentation Note 2: Overpower t,T The Overpower t, T Function Allowable Value shall not exceed the following setpoint by more than 1.21 % of t, T span (1.82% RTP).

Where: t, T is measured RCS t, T, oF.

t,To is the indicated t, T at RTP, oF.

s is the Laplace transform operator, sec* 1 .

T is the measured RCS average temperature,°F.

T" is the nominal Tavg at RTP, S 'oF.

Ks =',OF for increasing Tavg K6 =',OF when T> T"

',OF for decreasing Tavg ',OF when T s T"

'1 ;:>: , sec '2 s' sec '3 =' sec

'6 =' sec '7;:>: , sec The values denoted with' are specified in the COLR.

CALLAWAY PLANT 3.3-26 Amendment 202 I

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.

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

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

OR B.2.1 Be in MODE 3. 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> 81:iD.

B.2.2 Be in MODE 5. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> (continued)

CALLAWAY PLANT 3.3-27 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

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

One train may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

OR C.3.1 Be in MODE 3. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> AND C.3.2 Be in MODE 5. 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> l continued}

CALLAWAY PLANT 3.3-28 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

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

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

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

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

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

CALLAWAY PLANT 3.3-29 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

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

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

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

G.1 Restore train to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

OR G.2.1 Be in MODE 3. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> AND G.2.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> H. One or more trains ------------------- NOTE ------------------

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

H.1 Declare associated Immediately Pressurizer PORV(s) inoperable.

(contlnueo)

CALLAWAY PLANT Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

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

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

1.1 Place channel in trip. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR 1.2 Be in MODE 3. 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> J. One channel inoperable. ------------------ NOTE ------------------

The inoperable channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of other channels.

J.1 Place channel in trip. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OR J.2 Be in MODE 3. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> (ContinUed)

CALLAWAY PLANT 3.3-31 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

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

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

K.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

OR K.2.1 Be in MODE 3. 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> AND K.2.2 Be in MODE 5. 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> L. One or more required L.1 Verify interlock is in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> channel{s) inoperable. required state for existing unit condition.

OR L.2.1 Be in MODE 3. 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> AND L.2.2 Be in MODE 4. 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> (continued)

CALLAWAY PLANT 3.3-32 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME M. Two channels inoperable. M.1 Place channels in trip. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND OR AFW actuation on Trip of all M.2 Be in MODE 3. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Main Feedwater Pumps maintained from one actuation train.

N. One or more Containment N.1 Place channel(s) in trip. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Pressure - Environmental Allowance Modifier OR channel(s) inoperable.

N.2.1 Be in MODE 3. 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> AND N.2.2 Be in MODE 4. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> O. One channel inoperable. 0.1 Place channel in trip. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND 0.2 Restore channel to During OPERABLE status. performance of the next required COT (continUed)

CALLAWAY PLANT 3.3-33 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

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

AND P.2 Declare associated Immediately steam generator blowdown and sample line isolation valve(s) inoperable.

Q One train inoperable. ------------------- NOTE ------------------

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

Q.1 Restore train to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

OR Q.2.1 Be in MODE 3. 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> AND Q.2.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

CALLAWAY PLANT 3.3-34 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME R. One or both train(s) R.1 Restore train(s) to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable. OPERABLE status.

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

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

S.1 Restore train to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OPERABLE status.

OR S.2.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND S.2.2 Be in MODE 4. 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> CALLAWAY PLANT 3.3-35 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS

---

NOT E ----------------------------------------------------------

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 SR 3.3.2.3 --------------------------------- NOTE ---------------------------

The continuity check may be excluded from the BOP ESFAS test.

Perform ACTUATION LOGIC TEST. I n accordance with the Surveillance Frequency Control Program SR 3.3.2.4 Perform MASTER RELAY TEST. In accordance with the Surveillance Frequency Control Program (continued)

CALLAWAY PLANT 3.3-36 Amendment No. 202

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.5 Perform COT. In accordance with the Surveillance Frequency Control Program SR 3.3.2.6 --------------------------------- NOT E ---------------------------

Not applicable to slave relays K602, K620, K622, K624, K630, K740, K741 , and K750.

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 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 (continued)

CALLAWAY PLANT 3.3-37 Amendment No. 202

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SU RVEI LLANCE FREQUENCY SR 3.3.2.9 --------------------------------- NOT E ---------------------------

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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after SG pressure is

~ 900 psig.

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

CALLAWAY PLANT 3.3-38 Amendment No. 202

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.13 --------------------------------- NOTE ---------------------------

Only applicable to slave relays K602, K622, K624 ,

K630, K740, and K741.

Perform SLAVE RELAY TEST. In accordance with the Surveillance Frequency Control Program Prior to entering MODE 4 when in MODE 5 or 6

> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if not performed within the previous 92 days SR 3.3.2.14 --------------------------------- NOTE ---------------------------

Only applicable to slave relays K620 and K750.

Perform SLAVE RELAY TEST. In accordance with the Surveillance Frequency Control Program Prior to entering MODE 3 when in MODE 5 or 6

> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if not performed within the previous 92 days CALLAWAY PLANT 3.3-39 Amendment No. 202

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 11)

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 1,2,3,4 2 B SR 3,3.2 8 NA Initiation

b. Automatic 1,2,3,4 2 trains C SR 3.3,2.2 NA Actuation Sf~ 3.3,2.4 Logic and SR 3.3,2,(i Actuation SR 3,3,2,1]

Relays (SSPS)

c. Containment 1,2,3 3 D SR :\3,2.1 ~ 4,5 psig Pressure SR ~1325 High 1 SR :\.32.9 SR 33.2.10 d, Pressu rizer 1,2,3(b) 4 D SR 3321 2: 1834 psig Pressure SR 3.3.2,5 Low SR 3,3.2,9 SR 3.3,2,10 e, Steam Line 1,2,3(b) 3 per steam D SR 3,32.1 2: 610 pSig(C)(S)

Pressure line SR 3.3.2.5 Low SR 3.3.2,9 SR 3.3,2,10

2. Containment Spray

a. Manual 1,2,3,4 2 per train, B SR 3,3,2 8 NA Initiation 2 trains

b. Automatic 1,2,3,4 2 trains C SR 3,3,2 2 NA Actuation SR :U,2.4 Logic and SR 3.3.2.0 Actuation Relays (SSPS)

(a) The Allowable Value defines the limiting safety system setting except for Functions 1 ,e, 4,e,(1), 5,c, 5,e,(1), 5.e.(2),

6.d.(1), and 6.d,(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal 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 '1 2: 50 seconds and '2 s 5 seconds.

(s) 1. If the as-found instrument channel setpoint is conservative with respect to the Allowable Value, but outside its as-found test acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value that is within the as-left setpoint tolerance band on either side of the Nominal Trip Setpoint, or to a value that is more conservative than the Nominal Trip Set point; otherwise, the channel shall be declared inoperable. The Nominal Trip Setpoints and the methodology used to determine the as-found test acceptance criteria band and the as-left setpoint tolerance band shall be specified in the Bases.

CALLAWAY PLANT 3.3-40 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 11)

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

2. Containment Spray c, Containment 1.2.3 4 E SR  :>3.2.1 s 28.3 psig Pressure SR 3,3.2.5 High - 3 SR 3.3.2,9 SR 3.3,2,10

3. Containment Isolation

a. Phase A Isolation (1) Manual 1.2.3,4 2 B SR 3,3.2.8 NA Initiation (2) Automatic 1.2.3,4 2 trains C SR 3.3,2.2 NA Actuation SR 3.3.2.4 Logic and SR 3 3.2.6 Actuation SR 3.3.2.13 Relays (SSPS)

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

Injection

b. Phase B Isolation (1) Manual 1,2,3,4 2 per train, B SR 3,3.2 8 NA Initiation 2 trains (2) Automatic 1,2,3,4 2 trains C SR 3,3,2.2 NA Actuation SR 3,3,24 Logic and SR 3,3,2.6 Actuation Relays (SSPS)

(3) Contain- 1,2,3 4 E SR 3.3.2.1 s 28.3 psig ment SR 3.3.2.5 Pressure SR 3.3.2.9 High - 3 SR3.32.10 (a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(1). 5.c, 5.e.(1), 5.e,(2},

6,d,(1). and 6.d,(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

CALLAWAY PLANT 3.3-41 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 11)

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

a. Manual Initiation 1,2(i)(k), 3(i)(k) 2 F SR :U.2.R NA

b. Automatic 1,2(i)(k), 3(i)(k) 2 trains G SR 3.3.2.2 NA Actuation Logic SF{ 3.:3.2 4 and Actuation SF< :U.2.6 Relays (SSPS)

c. Automatic 1, 2(k),3(k) 2 trains(o) S SR :U23 NA Actuation Logic and Actuation Relays (MSFIS)

d. Containment 1,2(i)(k), 3(i)(k) 3 D SR ~'l.3.2.1 ~ 18.3 psig Pressure - High 2 SR 3.3.2.5 SR 3.3.2.D SR 3 ;3210 (a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(1), 5.c, 5.e.(1), 5.e.(2).

6.d.(1), and 6.d.(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

(i) Except when:

1. All MSIVBVs are:

1.a Closed and de-activated, or 1.b Closed and isolated by a closed manual valve, or 1.c Isolated by two closed manual valves.

AND

2. All MSLPDIVs are:

2.a Closed and de-activated, or 2.b Closed and isolated by a closed manual valve, or 2.c Isolated by two closed manual valves.

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

(0) Each train requires a minimum of two programmable logic controllers to be OPERABLE.

CALLAWAY PLANT 3.3-42 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 11)

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

4. Steam Une Isolation

e. Steam Une Pressure (1) Low 1,2 (i)(k), 3(b)(i)(k) 3 per steam D SR:3 321 ~ 610 pSig(C)(S) line SR 33.25 SR 33.29 SR 332.10 (2) Negative 3(g)(i)(k) 3 per steam D SR 3.321 5124 psi(h)

Rate - High line SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10 (a) The Allowable Value defines the limiting safety system setting except for Functions 1 ,e, 4.e.(1), 5.c, 5.e.(1), 5.e.(2),

6.d,(1), and 6.d.(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal 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 leadllag controller are '1 ~ 50 seconds and '2 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 ratellag controller is ~ 50 seconds.

(i) Except when:

1. All MSIVBVs are:

1.a Closed and de-activated, or 1.b Closed and isolated by a closed manual valve, or 1,c Isolated by two closed manual valves.

AND

2. All MSLPDIVs are:

2.a Closed and de-activated, or 2.b Closed and isolated by a closed manual valve, or 2.c Isolated by two closed manual valves.

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

(s) 1. If the as-found instrument channel setpoint is conservative with respect to the Allowable Value, but outside its as-found test acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value that is within the as-left setpoint tolerance band on either side of the Nominal Trip Setpoint, or to a value that is more conservative than the Nominal Trip Setpoint; otherwise, the channel shall be declared inoperable. The Nominal Trip Setpoints and the methodology used to determine the as-found test acceptance criteria band and the as-left setpoint tolerance band shall be specified in the Bases.

CALLAWAY PLANT 3.3-43 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 of 11)

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

5. Turbine Trip and Feedwater Isolation

a. Automatic 1(j) ,2 (j) ,3 (j) 2 trains G SR :)32 2 NA Actuation Logic SR :13.2.4 and Actuation SR :U.2,6 Relays (SSPS) SR 3.3.2.14

b. Automatic 2 trains(O) S SR 3.3.2.3 NA Actuation Logic and Actuation Relays (MSFIS) c, SG Water Level 4 per SG SR 3.3.2.1  :; 91.4%(s) of High High (P-14) SR 3.3.2,5 Narrow Range SR 3,3.2,9 Instrument SR 3.32.10 Span

d. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements. The Applicability exceptions of Footnote (j) also apply to Function 5.d.

(a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(I), 5.c, 5.e.(1). 5.e.(2),

6.d.(I), and 6.d.(2} (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

Except when:

1 All MFIVs are closed and de-activated:

AND

2. All MFRVs are:

2.a Closed and de-activated, or 2.b Closed and isolated by a closed manual valve; AND

3. All MFRVBVs are:

3.a Closed and de-activated, or 3.b Closed and isolated by a closed manual valve. or 3.c Isolated by two closed manual valves.

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

(o) Each train requires a minimum of two programmable logic controllers to be OPERABLE.

(s) 1. If the as-found instrument channel setpoint is conservative with respect to the Allowable Value, but outside its as-found test acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value that is within the as-left setpoint tolerance band on either side of the Nominal Trip Setpoint, or to a value that is more conservative than the Nominal Trip Setpoint; otherwise.

the channel shall be declared inoperable. The Nominal Trip Setpoints and the methodology used to determine the as-found test acceptance criteria band and the as-left setpoint tolerance band shall be specified in the Bases.

CALLAWAY PLANT 3.3-44 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 11)

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

5. Turbine Trip and Feedwater Isolation

e. Steam Generator Water Level Low-Low(q)

(1) Steam 1m, 2m, 3m 4 per SG D SR:U.2.1 Z 20.6%(5) of Generator SR 3.3.2.5 Narrow Range Water Level SR 3.3.2.9 Instrument Low-Low SR 3.3.2.10 Span (Adverse Containment Environment)

(2) Steam 1m(r), 2(j)(r), 3(j)(r) 4 per SG D SR 3.3.2.1 z 16.6%(5) of Generator SR :33.2.5 Narrow Range Water Level SR 3.3.2.9 Instrument Low-Low SR 3.3.2.10 Span (Normal Containment Environment)

(a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(1}, 5.c, 5.e.(1}, 5.e.(2},

6.d.(1}, and 6.d.(2} (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

(j) Except when:

1. All MFIVs are closed and de-activated; AND

2. All MFRVs are:

2.a Closed and de-activated, or 2.b Closed and isolated by a closed manual valve; AND

3. All MFRVBVs are:

3.a Closed and de-activated, or 3.b Closed and isolated by a closed manual valve, or 3.c Isolated by two closed manual valves.

(q) Feedwater isolation only.

(r) Except when the Containment Pressure - Environmental Allowance Modifier channels in the same protection sets are tripped.

(s) 1. If the as-found instrument channel setpoint is conservative with respect to the Allowable Value, but outside its as-found test acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value that is within the as-left setpoint tolerance band on either side of the Nominal Trip Setpoint, or to a value that is more conservative than the Nominal Trip Setpoint; otherwise, the channel shall be declared inoperable. The Nominal Trip Setpoints and the methodology used to determine the as-found test acceptance criteria band and the as-left setpoint tolerance band shall be specified in the Bases.

CALLAWAY PLANT 3.3-45 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 7 of 11)

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

5. Turbine Trip and Feedwater Isolation

e. Steam Generator Water Level Low-Low(q)

(3) Not used.

(4) Containment 1m, 2m, 3m 4 N SR 3 3.2.1 s 2.0 psig Pressure Sf~ :33.2.5 Environmental SR:3 3.2.9 Allowance SR :U.2.10 Modifier (a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(1), 5.c, 5.e.( 1), 5.e.(2),

6.d.( 1), and 6.d.(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

0) Except when:

1. All MFIVs are closed and de-activated; AND

2. All MFRVs are:

2.a Closed and de-activated, or 2.b Closed and isolated by a closed manual valve; AND

3. All MFRVBVs are:

3.a Closed and de-activated, or 3.b Closed and isolated by a closed manual valve, or 3.c Isolated by two closed manual valves.

(q) Feedwater isolation only.

CALLAWAY PLANT 3.3-46 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 8 of 11)

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/pump P SR 3.32.8 NA

b. Automatic 1,2,3 2 trains G SR 3.3.22 NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays (SSPS)

c. Automatic 1,2,3 2 trains Q SR 3323 NA Actuation Logic and Actuation Relays (BOP ESFAS)

d. SG Water Level Low-Low (1) Steam 1,2,3 4 perSG D SR :U.2 1 ~ 20.6%(s) of Generator SR 3:)25 Narrow Range Water Level SR 3.3.2.9 Instrument Low-Low SR 3.3.2.10 Span (Adverse Containment Environment)

(2) Steam 4 per SG D SR 3:; 21 ~ 16.6%(5) of Generator SR:U;)5 Narrow Range Water Level SR 3.32.9 Instrument Low-Low SR 3.3.2.10 Span (Normal Containment Environment)

(a) The Allowable Value defines the limiting safety system setting except for Functions 1 .e, 4.e.(1), 5.c, 5.e.(1), 5.e.(2),

6.d.(1), and 6.d.(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

(r) Except when the Containment Pressure - Environmental Allowance Modifier channels in the same protection sets are tripped.

(s) 1. Ifthe as-found instrument channel setpoint is conservative with respect to the Allowable Value, but outside its as-found test acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

2. The instrument channel setpoint shall be reset to a value that is within the as-left setpOint tolerance band on either side of the Nominal Trip Setpoint, or to a value that is more conservative than the Nominal Trip Setpoint; otherwise, the channel shall be declared inoperable. The Nominal Trip Setpoints and the methodology used to determine the as-found test acceptance criteria band and the as-left setpoint tolerance band shall be specified in the Bases.

CALLAWAY PLANT 3.3-47 Amendment No. 202 ,

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page90f11)

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

6. Auxiliary Feedwater

d. SG Water Level Low-Low (3) Not used.

(4) Containment 1,2,3 4 N SR 3.3.2.1 <; 2.0 psig Pressure - SR 3.3.2.5 Environmental SF'< 3.3.2.9 Allowance SR 3.3.2.10 Modifier

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

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

g. Trip of all Main 1(v),2(n),(v) 4(u),(w) J,M SR 3.3.2.8 NA Feedwater Pumps

h. Auxiliary 1,2,3 3 0 SR 3.3.2.1 2! 20.64 psia Feedwater Pump SR 3:329 Suction Transfer SR 3.:32.10 on Suction SR 3.3.2.12 Pressure - low (a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(1), 5.c, 5.e.(1), 5.e.(2),

6.d.(1), and 6.d.(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

(n) Trip function may be blocked just before shutdown of the last operating main feedwater pump and restored just after the first main feedwater pump is put into service following performance of its startup trip test.

(u) During startup of the second main feedwater pump, the following exception applies: The requirement for four OPERABLE channels is met if two required channels are OPERABLE on the associated main feedwater pump in operation supplying feedwater to the SGs and two required channels are in the tripped condition on the second main feedwater pump.

(v) During removal of the first of two operating main feedwater pumps from service, the following exception applies:

(1) LCO 3.0.3 is not applicable for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the channels associated with the first main feedwater pump, OR (2) The requirement for four OPERABLE channels is met if two required channels are OPERABLE on the associated main feedwater pump in operation supplying feedwater to the SGs and two required channels on the main feedwater pump to be removed from service are in the tripped condition.

(w) During removal of the first of two operating main feedwater pumps from service, the following exception applies: The requirement for four OPERABLE channels is met if two required channels are OPERABLE on the associated main feedwater pump in operation supplying feedwater to the SGs and two required channels on the main feedwater pump to be removed from service are in the tripped condition.

CALLAWAY PLANT 3.3-48 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 10 of 11)

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

7. Automatic Switch over to Containment Sump

a. Automatic 1,2,3,4 2 trains C SR 33.2.2 NA Actuation Logic SR :U2A and Actuation SR 3.3.213 Relays (SSPS)

b. Refueling Water 1,2,3,4 4 K Sf" 3.3.2.1 235.2%

Storage Tank SR 3.3.25 (RWST) Level SH 3 329 Low Low SR 3.321 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection

8. ESFAS Interlocks

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

b. Pressurizer 1,2,3 3 L 3.32.5 ~ 1981 psig Pressure, P-11 3.3.2.9

9. Automatic Pressurizer PORV Actuation

a. Automatic 1,2,3 2 trains H SR 3.3.2.2 NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.14 Relays (SSPS)

b. Pressurizer 1,2,3 4 D SR 3.3.2.1 ,.;2350 psig Pressure - High SR 33.2.5 3.3.2.9 (a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(1), 5.c, 5.e.(1), 5.e.(2),

6.d.(1), and 6.d.(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

CALLAWAY PLANT 3.3-49 Amendment No. 202 I

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 11 of 11 )

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

10. Steam Generator Blowdown and Sample Line Isolation

a. Manual Initiation 1(1), 2(t), 3(1) 2 trains P SR :U,2,a NA (1 per MDAFW pump)

b. Automatic 1(1),2(1),3(1) 2 trains Q SR 3.3.2.3 NA Actuation Logic and Actuation Relays (BOP ESFAS)

c. Safety Injection 1(1),2(1),3(1) Refer to Function 1 (Safety Injection) for initiation functions and requirements.

d. Loss of Offsite 1(1),2(1),3(t) 2 trains R SR :U.2,7 NA Power (a) The Allowable Value defines the limiting safety system setting except for Functions 1.e, 4.e.(1), 5.c, 5.e.(1), 5.e.(2),

6,d,(1), and 6.d.(2) (the Nominal Trip Setpoint defines the limiting safety system setting for these Functions). See the Bases for the Nominal Trip Setpoints.

(t) Except when all Steam Generator Slowdown and Sample Line Isolation Valves are:

1, Closed and de-activated, or

2. Closed and isolated by a closed manual valve, or

3. Isolated by a combination of closed manual valve{s) and closed de-activated automatic valve(s).

CALLAWAY PLANT 3.3-50 Amendment No. 202 I

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.

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

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)

CALLAWAY PLANT 3.3-51 Amendment No. 202 I

PAM Instrumentation 3.3.3 ACTIONS (continued)

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

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

E. As required by Required E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action D.1 and referenced in Table 3.3.3-1. AND E.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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.S.

CALLAWAY PLANT 3.3-52 Amendment No. 202 ,

PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS

---

NOT E ----------------------------------------------------------

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. I n accordance with the Surveillance Frequency Control Program SR 3.3.3.2 --------------------------------- NOT E ---------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.3-53 Amendment No. 202

PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2)

Post Accident Monitoring Instrumentation CONDITION REFERENCED FROM REQUIRED REQUIRED FUNCTION CHANNELS ACTION D.1 2 E

1. Neutron Flux

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

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

4. RCS Pressure (Wide Range) 2 E

5. Reactor Vessel Level Indicating System 2 F (RVLlS)

6. Containment Normal Sump Water Level 2 E

7. Containment Pressure (Normal Range) 2 E

8. Steam Line Pressure 2 per steam E 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 steam E generator (continued)

CALLAWAY PLANT 3.3-54 Amendment No. 202 I

PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2)

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

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).

CALLAWAY PLANT 3.3-55 Amendment No. 202 I

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.

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

One or more required ASP controls inoperable.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> CALLAWAY PLANT 3.3-56 Amendment No. 202 I

Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required In accordance instrumentation channel that is normally energized. with the Surveillance Frequency Control Program SR 3.3.4.2 --------------------------------- NOTE ---------------------------

Only required to be performed in MODES 1 and 2 for the turbine-driven AFW pump.

Verify each required auxiliary shutdown panel control In accordance circuit and transfer switch is capable of performing the with the intended function. Surveillance Frequency Control Program SR 3.3.4.3 --------------------------------- NOTE --------------------------

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

2. Reactor trip breaker and RCP breaker position indications are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION for each required In accordance instrumentation channel. with the i Surveillance Frequency Control Program CALLAWAY PLANT 3.3-57 Amendment No. 202

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

Remote Shutdown System Functions

1. Source Range Neutron Flux(a) 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 Breaker Position 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.

CALLAWAY PLANT 3.3-58 Amendment No. 202 I

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 4.16-kV NB bus of the loss of voltage Function and four channels per 4.16-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.

COMPLETION CONDITION REQUIRED ACTION TIME A. One or more Functions with A.1 ------------ NOTE ----------

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

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

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

CALLAWAY PLANT 3.3-59 Amendment No. 202 I

LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1 Tie breakers between 480 Vac buses NG01 and In accordance NG03 and between 480 Vac buses NG02 and NG04 with the shall be verified open. Surveillance Frequency Control Program 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 Setpoint and Allowable Value as follows: with the Surveillance

a. Loss of voltage Allowable Value 83 +0, -8.3V Frequency (120V Bus) with a time delay of 1.0 + 0.2, Control Program

-0.5 sec.

Loss of voltage nominal Trip Setpoint 83V

( 120V Bus) with a time delay of 1.0 sec.

b. Degraded voltage Allowable Value 107.47 +/- 0.38V (120V Bus) with a time delay of 119 +/- 11.6 sec.

Degraded voltage nominal Trip Setpoint 107.47V (120V Bus) with a time delay of 119 sec.

SR 3.3.5.4 Verify LOP DG Start ESF RESPONSE TIMES are In accordance within limits. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.3-60 Amendment No. 202

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

---

NOT E ----------------------------------------------------------

Separate Condition entry is allowed for each Function.

COMPLETION CONDITION REQUIRED ACTION TIME A. One radiation monitoring A.1 Restore the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> channel inoperable. channel to OPERABLE status.

(continued)

CALLAWAY PLANT 3.3-61 Amendment No. 202 I

Containment Purge Isolation Instrumentation 3.3.6 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. ------------- NOTE ------------ B.1 Place and maintain Immediately Only applicable in MODE 1, containment purge 2,3,or4. supply and exhaust valves in closed position.

One or more F unctions with one or more manual channels or automatic actuation trains inoperable.

Both radiation monitoring channels inoperable.

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

(cOntinUed)

CALLAWAY PLANT 3.3-62 Amendment No. 202 I

Containment Purge Isolation Instrumentation 3.3.6 ACTIONS (continued)

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

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

CALLAWAY PLANT 3.3-63 Amendment No. 202 I

Containment Purge Isolation Instrumentation 3.3.6 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 SR 3.3.6.2 --------------------------------- NOT E ---------------------------

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 CALLAWAY PLANT 3.3-64 Amendment No. 202

Containment Purge Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 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 TIMES are within limits. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.3-65 Amendment No. 202

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 REQUIRED SURVEILLANCE NOMINAL FUNCTION CONDITIONS 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 3.36.2 NA Actuation SR 3.3.66 Logic and Actuation Relays (BOP ESFAS)

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

Purge SR 3.3.6.3 Exhaust SR 3.3.6.5 Radiation Gaseous

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

Phase A (a) During CORE ALTERATIONS.

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

(c) Set to ensure ODCM limits are not exceeded.

CALLAWAY PLANT 3,3-66 Amendment No. 202 I

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 33.7-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.7-1.

ACTIONS

---

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

Separate Condition entry is allowed for each Function.

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

(Continued)

CALLAWAY PLANT 3.3-67 Amendment No. 202 I

CREVS Actuation Instrumentation 3.3.7 ACTIONS (continued)

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

Function 3.

AND One or more Functions with B.1.2 Enter applicable Immediately two channels or two trains Conditions and Required 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 Immediately CRVIS mode.

~contlnuea)

CALLAWAY PLANT 3.3-68 Amendment No. 202 I

CREVS Actuation Instrumentation 3.3.7 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in CRVIS mode.

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

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Conditions 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 <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. Required Action and E.1 Suspend CORE Immediately associated Completion ALTERATIONS.

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

CALLAWAY PLANT 3.3-69 Amendment No. 202 I

CREVS Actuation Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS

---

NOT E ----------------------------------------------------------

Refer to Table 3.3.1-1 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 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 --------------------------------- NOT E ---------------------------

Verification of setpoint is not required.

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

CALLAWAY PLANT 3.3-70 Amendment No. 202

CREVS Actuation Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 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 RESPONSE TIMES are within limits with the Surveillance Frequency Control Program CALLAWAY PLANT 3.3-71 Amendment No. 202

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

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

1. Manual 1,2,3,4, 2 SR 33.7.4 NA Initiation (a), and (c)

2. Automatic 1,2,3,4, 2 trains SR :3.3.7.3 NA Actuation (a), and (c)

Logic and Actuation (a) 2 trains SR 33.7.6 NA Relays (BOP ESFAS)

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

Radiation and (a) SR 3.3.7.2 Control Room SR 3.3.7.5 Air Intakes (a) 2 SR 3.3.7.6 (b)

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

Phase A

5. Fuel Building Refer to LCO 3.3.8, "EES Actuation Instrumentation," for all initiation functions and Exhaust requirements.

Radiation-Gaseous (a) During CORE ALTERATIONS or during movement of irradiated fuel assemblies within containment.

(b) Nominal Trip Setpoint concentration value (flCi/cm 3 ) shall be established such that the actual submersion dose rate would not exceed 2 mRlhr in the control room.

(c) During movement of irradiated fuel assemblies in the fuel building.

CALLAWAY PLANT 3.3-72 Amendment No. 202 I

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

---

NOT ES --------------------------------------------------------

1. LCO 3.0.3 is not applicable.

2. Separate Condition entry is allowed for each Function.

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

AND A.2 Place one CREVS train 7 days in Control Room Ventilation Isolation Signal (CRVIS) mode.

(contlnuea)

CALLAWAY PLANT 3.3-73 Amendment No. 202 I

EES Actuation Instrumentation 3.3.8 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. ------------- NOTE ------------ B.1.1 Place one EES train in Immediately Not applicable to the FBVIS mode and Function 3. one CREVS train in the CRVIS mode.

One or more Functions AND with two channels or two trains inoperable. B.1.2 Enter applicable Immediately Conditions and Required Actions of LCO 3.7.10, "Control Room Emergency Ventilation System (CREVS)," for one CREVS train made inoperable and enter applicable Conditions and Required Actions of LCO 3113, "Emergency Exhaust System (EES),"

for one EES train made inoperable by inoperable EES actuation instrumentation.

B.2 Place both EES trains in Immediately the FBVIS mode and both CREVS trains in the CRVIS mode.

CALLAWAY PLANT 3.3-74 Amendment No. 202 I

EES Actuation Instrumentation 3.3.8 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME 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 and enter applicable 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the FBVIS mode and one CREVS train in the CRVIS mode.

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

D. Required Action and D.1 Suspend movement of Immediately associated Completion irradiated fuel Time for Conditions A, B, assemblies in the fuel or C not met during building.

movement of irradiated fuel assemblies in the fuel building.

CALLAWAY PLANT 3.3-75 Amendment No. 202 I

EES Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS

---

NOT E ----------------------------------------------------------

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

SU RVEI LLANCE 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 SR 3.3.8.4 --------------------------------- NOTE ---------------------------

Verification of setpoint is not required.

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

CALLAWAY PLANT 3.3-76 Amendment No. 202

EES Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.8.5 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.3-77 Amendment No. 202

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

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

1. Manual (a) 2 SR 3.3.8.4 NA Initiation

2. Automatic (a) 2 trains SR 3.:UU NA Actuation Logic and Actuation Relays (BOP ESFAS)

3. Fuel (a) 2 SR 3.3.8.1 (b)

Building SR 3.3.8.2 Exhaust SR 3.3.8.5 Radiation

- Gaseous (a) During movement of irradiated fuel assemblies in the fuel building.

(b) Nominal Trip Setpoint concentration value (I!Ci/cm 3) shall be established such that the actual submersion dose rate would not exceed 4 mRlhr in the fuel building.

CALLAWAY PLANT 3.3-78 Amendment No. 202 I

BDMS 3.3.9 3.3 INSTRUMENTATION 3.3.9 Boron Dilution Mitigation System (BDMS)

LCO 3.3.9 Two trains of the BDMS shall be OPERABLE and one RCS loop shall be in operation.

APPLICABILITY: MODES 2 (below P-6 (Intermediate Range Neutron Flux) interlock), 3, 4, and 5.

---

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

The boron dilution flux multiplication signal may be blocked:

1. During subcritical physics testing;

2. During control bank movement in MODE 2 (below P-6 (Intermediate Range Neutron Flux) interlock);

3. During control bank movement in MODE 3;

4. During shutdown bank movement in MODE 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One train inoperable. A.1 Restore train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

(continued)

CALLAWAY PLANT 3.3-79 Amendment No. 202 I

BDMS 3.3.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Two trains inoperable. B.1 **********. NOTE _._-.__**__*

Plant temperature changes are allowed provided the temperature Required Action and change is accounted for associated Completion Time in the calculated SDM.

of Condition A not met.

Suspend operations Immediately involving positive reactivity additions.

AND B.2 Perform SR 3.1.1.1. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND B.3.1 Close and secure 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> unborated water source isolation valves.

AND B.3.2 Verify unborated water Once per 31 days source isolation valves are closed and secured.

(continued)

CALLAWAY PLANT 3.3-80 Amendment No. 202 I

BDMS 3.3.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. No RCS loop in operation. C.1 Close and secure 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> unborated water source isolation valves.

AND C.2 Verify unborated water Once per 31 days source isolation valves are closed and secured.

CALLAWAY PLANT 3.3-81 Amendment No. 202 I

BDMS 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.9.2 --------------------------------- NOTE ---------------------------

Only required to be performed in MODE 5.

Verify BGV0178 is secured in the closed position. In accordance with the Surveillance Frequency Control Program SR 3.3.9.3 --------------------------------- NOT E ---------------------------

Not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power below P-6 interlock.

Perform COT and verify nominal flux multiplication In accordance setpoint of 1.7. with the Surveillance Frequency Control Program SR 3.3.9.4 --------------------------------- NOTE ---------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.3-82 Amendment No. 202

BDMS 3.3.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.9.5 Verify the centrifugal charging pump suction valves In accordance from the RWST open and the CVCS volume control with the tank discharge valves close in less than or equal to Surveillance 30 seconds on a simulated or actual actuation signal. Frequency Control Program SR 3.3.9.6 Verify one RCS loop is in operation. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.3-83 Amendment No. 202

RCS Pressure, Temperature, and Flow ONB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to In accordance the limit specified in the COLR. with the Surveillance Frequency Control Program SR 3.4.1.2 Verify RCS average temperature is less than or equal In accordance to the limit specified in the COLR. with the Surveillance Frequency Control Program SR 3.4.1.3 Verify RCS total flow rate is ~ 382,630 gpm. In accordance with the Surveillance Frequency Control Program SR 3.4.1.4 --------------------------------- NOTE ---------------------------

Calculated rather than verified by precision heat balance when performed prior to THERMAL POWER exceeding 75% RTP.

Verify by precision heat balance that RCS total flow Once after each rate is ~ 382,630 gpm. refueling prior to THERMAL POWER exceeding 75%

RTP In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT Amendment No. 202

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 operating RCS loop average temperature (Tavg) shall be ~ 551°F.

APPLICABILITY: MODE 1, MODE 2 with keff ~ 1.0.

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

within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE I FREQUENCY SR 3.4.2.1 Verify RCS Tavg in each operating loop ~ 551°F. In accordance

• with the I Surveillance I Frequency

~ontrol Program CALLAWAY PLANT 3.4-3 Amendment No. 202

RCS PIT Limits 3.4.3 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION 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.

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AN D Requirements of LCO not C.2 Determine RCS is Prior to entering met any time in other than acceptable for continued MODE4 MODE 1,2,3, or4. operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 --------------------------------- NOT E ---------------------------

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 heatup and cooldown rates are within the limits with the specified in the PTLR. Surveillance Frequency Control Program CALLAWAY PLANT 3.4-5 Amendment No. 202

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 COMPLETION CONDITION REQUIRED ACTION TIME A. Requirements of LCO not A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 CALLAWAY PLANT 3.4-6 Amendment No. 202

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 water levels are ~ 7% for required RCS loops. with the Surveillance Frequency Control Program SR 3.4.5.3 Verify correct breaker alignment and indicated power In accordance are available to the required pump that is not in with the operation. I Surveillance Frequency Control Program I

CALLAWAY PLANT 3.4-9 Amendment No. 202

RCS Loops - MODE 4 3.4.6 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. Required loops inoperable. B.1 Suspend operations that Immediately would cause introduction 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.

I 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 are ~ 7% for required RCS loops. with the Surveillance Frequency Control Program SR 3.4.6.3 Verify correct breaker alignment and indicated power In accordance are available to the required pump that is not in with the operation. Surveillance Frequency Control Program CALLAWAY PLANT 3.4-11 Amendment No. 202

RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one RHR loop is in operation. In accordance with the Surveillance Frequency Control Program SR 3.4.7.2 Verify SG secondary side wide range water level is In accordance

~ 86% in required SGs. with the Surveillance Frequency Control Program SR 3.4.7.3 Verify correct breaker alignment and indicated power In accordance are available to the required RHR pump that is not in with the operation. Surveillance Frequency Control Program CALLAWAY PLANT 3.4-14 Amendment No. 202

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

COMPLETION CONDITION REQUIRED ACTION TIME B. Required RHR loops B.1 Suspend operations that Immediately inoperable. would cause introduction into the RCS, coolant with boron concentration less than required to No RHR loop in operation. meet 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 are available to the required RHR pump that is not in with the operation. Surveillance Frequency Control Program CALLAWAY PLANT 3.4-16 Amendment No. 202

Pressurizer 3.4.9 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition B not AND met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is s; 92%. In accordance with the Surveillance Frequency Control Program SR 3.4.9.2 Verify capacity of each required group of backup In accordance pressurizer heaters is :?: 150 kW. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.4-18 Amendment No. 202

Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 --------------------------------- NOT E --------------------------

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 CALLAWAY PLANT 3.4-24 Amendment No. 202

COMS 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of zero safety injection pumps are In accordance capable of injecting into the RCS. with the Surveillance Frequency Control Program SR 3.4.12.2 Verify a maximum of one centrifugal charging pump In accordance is capable of injecting into the RCS. with the Surveillance Frequency Control Program SR 3.4.12.3 Verify each accumulator is isolated when In accordance accumulator pressure is greater than or equal to the with the maximum RCS pressure for the existing RCS cold Surveillance leg temperature allowed by the PIT limit curves Frequency provided in the PTLR. Control Program SR 3.4.12.4 Verify RHR suction isolation valves are open for each In accordance required RHR suction relief valve. with 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)

CALLAWAY PLANT 3.4-28 Amendment No. 202

COMS 3.4.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.12.6 Verify PORV block valve is open for each required In accordance PORV. with 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing any RCS cold leg temperature to ~ 275°F.

Perform a COT on each required PORV, excluding In accordance actuation. with the Surveillance Frequency Control Program SR 3.4.12.9 Perform CHANNEL CALIBRATION for each required In accordance PORV actuation channel. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.4-29 Amendment No. 202

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 -------------------------- NOT ES ------------------------------

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

2. Not applicable to primary to secondary LEAKAGE Verify RCS operational LEAKAGE is within limits by In accordance performance of RCS water inventory balance. with the Surveillance Frequency Control Program SR 3.4.13.2 -------------------------- NOTE -----------------------------

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

Verify primary to secondary LEAKAGE is s 150 In accordance gallons per day through anyone SG. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.4-31 Amendment No. 202

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 --------------------------------- NOT ES -------------------------

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 Inservice maximum of 5 gpm at an RCS pressure 2 2215 psig Testing Program and ~ 2255 psig.

In accordance with the Surveillance Frequency Control Program 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 (continued)

CALLAWAY PLANT 3.4-34 Amendment No. 202

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR3.4.14.1 (continued) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following check valve actuation due to flow through the valve SR 3.4.14.2 Verify RHR suction isolation valve interlock prevents In accordance the valves from being opened with a simulated or with the actual RCS pressure signal ~ 425 psig except when Surveillance the valves are open to satisfy LCO 3.4.12. Frequency Control Program CALLAWAY PLANT 3.4-35 Amendment 202

RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the required In accordance containment atmosphere particulate radioactivity with the monitor. Surveillance Frequency Control Program SR 3.4.15.2 Perform COT of the required containment In accordance atmosphere particulate radioactivity monitor. with the Surveillance Frequency Control Program SR 3.4.15.3 Perform CHANNEL CALIBRATION of the required In accordance containment sump level and flow monitoring system. with the Surveillance Frequency Control Program SR 3.4.15.4 Perform CHANNEL CALIBRATION of the required In accordance containment atmosphere particulate radioactivity with the monitor. Surveillance Frequency Control Program SR 3.4.15.5 Perform CHANNEL CALIBRATION of the required In accordance containment cooler condensate monitoring system. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.4-39 Amendment No. 202

RCS Specific Activity 3.4.16 ACTIONS (continued)

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

C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR DOSE EQUIVALENT 1-131

> 60 J..lCi/gm.

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 specific activity::; 225 mCi/gm. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.4-41 Amendment No. 202

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.2 --------------------------------- NOTE ---------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance specific activity::; 1.0 ~Ci/gm. with the Surveillance Frequency Control Program Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of ~ 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period CALLAWAY PLANT 3.4-42 Amendment No. 202

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

All SG tubes satisfying the tube repair criteria shall be plugged in accordance with Steam Generator Program.

APPLICABILITY: MODES 1 2, 3, and 4.

ACTIONS

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

Separate Condition entry is allowed for each SG tube.

COMPLETION CONDITION REQUIRED ACTION TIME A. One or more SG tubes A.1 Verify tube integrity of 7 days satisfying the tube repair the affected tube(s) is criteria and not plugged in maintained until the next accordance with the Steam refueling outage or Generator Program. inspection.

AND A.2 Plug the affected tube(s) Prior to entering in accordance with the MODE 4 following Steam Generator the next refueling Program. outage or SG tu be inspection (continued)

CALLAWAY PLANT 3.4-43 Amendment No. 202 I

SG Tube Integrity 3.4.17 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SG tube integrity not maintained.

CALLAWAY PLANT 3.4-44 Amendment No. 202 ,

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam In accordance with Generator Program. the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program. a SG tube inspection CALLAWAY PLANT 3.4-45 Amendment No. 202 I

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 I n accordance

~ 6061 gallons and ~ 6655 gallons. with the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is In accordance

~ 602 psig and ~ 648 psig. with the Surveillance Frequency Control Program (continued)

CALLAWAY PLANT 3.5-2 Amendment No. 202

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.4 Verify boron concentration in each accumulator is In accordance 2: 2300 ppm and:::; 2500 ppm. with the Surveillance Frequency Control Program AND

---

NOTE -------

Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of 2: 70 gallons that is not the result of addition from the refueling water storage tank SR 3.5.1.5 Verify power is removed from each accumulator In accordance isolation valve operator when RCS pressure is with the

> 1000 psig. Surveillance Frequency Control Program CALLAWAY PLANT 3.5-3 Amendment No. 202

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.

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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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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, "Cold Overpressure Mitigation System," is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or until the temperature of all RCS cold legs exceeds 375°F, whichever comes first.

APPLICABILITY: MODES 1, 2, and 3.

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

AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> CALLAWAY PLANT 3.5-4 Amendment No. 202 I

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 power to the valve operator removed. with the Surveillance Number PO§jitioo Functioo Frequency BNHV8813 Open Safety Injection to RWST Control Program Isolation Valve EMHV8802A Closed SI Hot Legs 2 & 3 Isolation Valve EMHV8802B Closed SI Hot Legs 1 & 4 Isolation Valve EMHV8835 Open Safety Injection Cold Leg Isolation Valve EJHV8840 Closed RHRlSI Hot Leg Recirc Isolation Valve EJHV8809A Open RHR to Accum Inject Loops 1 & 2 Isolation Valve EJHV8809B Open RHR to Accum Inject Loops 3 & 4 Isolation Valve SR 3.5.2.2 Verify each ECCS manual, power operated, and In accordance automatic valve in the flow path, that is not locked, with the sealed, or otherwise secured in position, is in the Surveillance correct position. Frequency Control Program SR 3.5.2.3 Verify ECCS piping is full of water. In accordance with 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)

CALLAWAY PLANT 3.5-5 Amendment No. 202

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 that is not locked, sealed, or otherwise secured in with the position, actuates to the correct position on an actual Surveillance or simulated actuation signal. Frequency Control Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an In accordance actual or simulated actuation signal. with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, In accordance each mechanical position stop is in the correct with the position. Surveillance Valve Number Frequency Control Program EMV0095 EMV0107 EMV0089 EMV0096 EMV0108 EMV0090 EMV0097 EMV0109 EMV0091 EMV0098 EMV0110 EMV0092 SR 3.5.2.8 Verify, by visual inspection, each ECCS train In accordance containment sump suction inlet is not restricted by with the debris and the suction inlet strainers show no Surveillance evidence of structural distress or abnormal corrosion. Frequency Control Program CALLAWAY PLANT 3.5-6 Amendment No. 202

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.

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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

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

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

COMPLETION CONDITION REQUIRED ACTION 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 Centrifugal B.1 Restore required ECCS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Charging Pump subsystem Centrifugal Charging inoperable. Pump subsystem to OPERABLE status.

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

CALLAWAY PLANT 3.5-7 Amendment No. 202 I

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 CALLAWAY PLANT 3.5-8 Amendment No. 202 I

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 COMPLETION CONDITION REQUIRED ACTION TIME A. RWST boron concentration A.1 Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reasons other than OPERABLE status.

Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> CALLAWAY PLANT 3.5-9 Amendment No. 202 I

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

Only required to be performed when ambient air temperature is < 3rF or > 100°F.

Verify RWST borated water temperature is ~ 37°F In accordance and s 100°F. with the Surveillance Frequency Control Program SR 3.5.4.2 Verify RWST borated water volume is ~ 394,000 In accordance gallons. with the Surveillance Frequency Control Program SR 3.5.4.3 Verify RWST boron concentration is ~ 2350 ppm and In accordance s 2500 ppm. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.5-10 Amendment No. 202

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 (RCP) seal injection flow shall be within the limits of Figure 3.5.5-1.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. Seal injection flow not within A.1 Adjust manual seal 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit. injection throttle valves such that the RCP seal injection flow is within the limits of Figure 3.5.5-1.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> CALLAWAY PLANT Amendment No. 202 I

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the Reactor Coolant System pressure stabilizes at

~ 2215 psig and ~ 2255 psig.

Verify manual seal injection throttle valves are In accordance adjusted to give a flow within the limits of with the Figure 3.5.5-1. Surveillance Frequency Control Program CALLAWAY PLANT 3.5-12 Amendment No. 202

Seal Injection Flow 3.5.5 Delta P from Charging Header to RCS vs. Total Seal Injection Line Flow 210

_2004~~'~~+-'-~{'+'-~'~~~'~""~~"~"'~~~"'~'~'

E!

f,/)

@:. 190

J 1/1 f 180-t~-~*~

0 f,/)

~ 170

160 1~.,,***'t

*:4"'

.lil&

0 150 ~"~"f"t.,--1

4'\1

..c:

~ 140 Ci 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 Total Seal Injection Line Flow IGPMI Figure 3.5.5-1 (page 1 of 1)

Delta P from Charging Header to RCS VS. Total Seal Injection Line Flow CALLAWAY PLANT Amendment No. 202 ,

Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOT ES -------------------------

SR 3.6.2.1 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 opened at a In accordance time. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.6-6 Amendment No. 202

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each containment shutdown purge valve is In accordance sealed closed or closed and blind flange installed with the except for one purge valve in a penetration flow path Surveillance while in Condition D of this LCO. Frequency Control Program Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment SR 3.6.3.2 Verify each containment mini-purge valve is closed, In accordance except when the containment mini-purge valves are with the open for pressure control, ALARA or air quality Surveillance considerations for personnel entry, or for Frequency Surveillances that require the valves to be open. Control Program 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 blind flange that is located outside containment and with the not locked, sealed, or otherwise secured and required Surveillance to be closed during accident conditions is closed, Frequency except for containment isolation valves that are open Control Program under administrative controls.

(COntinued)

CALLAWAY PLANT 3.6-13 Amendment No. 202

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 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 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 shutdown purge valves with resilient seals and with the associated blind flanges. Surveillance Frequency Control Program Following each reinstallation of the blind flange (continUed)

CALLAWAY PLANT 3.6-14 Amendment No. 202

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.7 --------------------------------- NOT E ---------------------------

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 mini-purge and shutdown purge valves with resilient with the seals. Surveillance Frequency Control Program Within 92 days after opening the valve SR 3.6.3.8 Verify each automatic containment isolation valve that In accordance is not locked, sealed or otherwise secured in position, with the actuates to the isolation position on an actual or Surveillance simulated actuation signal. Frequency Control Program CALLAWAY PLANT 3.6-15 Amendment No. 202

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

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.6-16 Amendment No. 202

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

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within In accordance limit. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.6-17 Amendment No. 202

Containment Spray and Cooling Systems 3.6.6 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition C not AND met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. Two containment spray E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> trains inoperable.

AND OR E.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Two containment cooling trains inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power In accordance operated, and automatic valve in the flow path that is with the not locked, sealed, or otherwise secured in position is Surveillance in the correct position. Frequency Control Program SR 3.6.6.2 Operate each containment cooling train fan unit for In accordance 2: 15 minutes. with the Surveillance Frequency Control Program CALLAWAY PLANT Amendment No. 202

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.3 Verify each containment cooling train cooling water In accordance flow rate is 2:: 2200 gpm. with the Surveillance Frequency Control Program 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 flow path that is not locked, sealed, or otherwise with the secured in position, actuates to the correct position on Surveillance an actual or simulated actuation signal. Frequency Control Program SR 3.6.6.6 Verify each containment spray pump starts In accordance automatically on an actual or simulated actuation with the signal. Surveillance Frequency Control Program SR 3.6.6.7 Verify each containment cooling train starts In accordance automatically and minimum cooling water flow rate is with the established on an actual or simulated actuation Surveillance signal. Frequency Control Program SR 3.6.6.8 Verify each spray nozzle is unobstructed. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.6-20 Amendment No. 202

Recirculation Fluid pH Control System 3.6.7 3.6 CONTAINMENT SYSTEMS 3.6.7 Recirculation Fluid pH Control (RFPC) System LCO 3.6.7 The RFPC System shall be OPERABLE.

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

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

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.7.1 Verify the integrity of the RFPC System. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.6-21 Amendment No. 202

Recirculation Fluid pH Control System 3.6.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.7.2 Verify the RFPC System ensures an equilibrium sump In accordance pH;:: 7.1. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.6-22 Amendment No. 202

MSIVs, MSIVBVs, and MSLPDIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE I FREQUENCY SR 3.7.2.1 Verify isolation time of each MSIV is within limits. In accordance with the Inservice Testing Program SR 3.7.2.2 Verify each MSIV, each MSIVBV, and each MSLPDIV In accordance actuates to the isolation position on an actual or with the simulated actuation signal. Surveillance Frequency Control Program SR 3.7.2.3 Verify isolation time of each MSIVBV and MSLPDIV is In accordance with within limits. the Inservice Testing Program CALLAWAY PLANT 3.7-8 Amendment No. 202

MFIVs, MFRVs, and MFRVBVs 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 --------------------------------- NOT E ---------------------------

Only required to be performed in MODES 1 and 2.

Verify the closure time of each MFRV and MFRVBV is In accordance with within limits. the Inservice Testing Program SR 3.7.3.2 --------------------------------- NOTE ---------------------------

For the MFRVs and MFRVBVs, only required to be performed in MODES 1 and 2.

Verify each MFIV, MFRV and MFRVBV actuates to In accordance the isolation position on an actual or simulated with the actuation signal. Surveillance Frequency Control Program SR 3.7.3.3 Verify the closure time of each MFIV is within limits. In accordance with the Inservice Testing Program CALLAWAY PLANT 3.7-11 Amendment 202

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 --------------------------------- NOTE ---------------------------

Only required to be performed for the AFW flow control valves when the system is placed in automatic control or when THERMAL POWER is > 10% RTP.

Verify each AFW manual, power operated, and In accordance automatic valve in each water flow path, and in both with the steam supply flow paths to the steam turbine driven Surveillance pump, that is not locked, sealed, or otherwise secured Frequency in position, is in the correct position. Control Program SR 3.7.5.2 --------------------------------- NOTE ---------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 2 900 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 sealed, or otherwise secured in position, actuates to with the the correct position on an actual or simulated Surveillance actuation signal. Frequency Control Program (continued)

CALLAWAY PLANT 3.7-17 Amendment No. 202

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.5.4 --------------------------------- NOTE ---------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 2 900 psig in the steam generator.

Verify each AFW pump starts automatically on an In accordance actual or simulated actuation signal. with 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 CALLAWAY PLANT 3.7-18 Amendment No. 202

CST 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the CST contained water volume is In accordance

. 281,000 gal. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-20 Amendment No. 202

CCWSystem 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 automatic valve in the flow path servicing safety with the related equipment, that is not locked, sealed, or Surveillance otherwise secured in position, is in the correct Frequency position. Control Program SR 3.7.7.2 Verify each CCW automatic valve in the flow path that In accordance is not locked, sealed, or otherwise secured in with the position, actuates to the correct position on an actual Surveillance or simulated actuation signal. Frequency Control Program SR 3.7.7.3 Verify each CCW pump starts automatically on an In accordance actual or simulated actuation signal. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-22 Amendment No. 202

ESW 3.7.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 --------------------------------- NOTE ---------------------------

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

Verify each ESW manual, power operated, and In accordance automatic valve in the flow path servicing safety with the related equipment, that is not locked, sealed, or Surveillance otherwise secured in position, is in the correct Frequency position. Control Program SR 3.7.8.2 Verify each ESW automatic valve in the flow path that In accordance is not locked, sealed, or otherwise secured in with the position, actuates to the correct position on an actual Surveillance or simulated actuation signal. Frequency Control Program SR 3.7.8.3 Verify each ESW pump starts automatically on an In accordance actual or simulated actuation signal. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-25 Amendment No. 202

UHS 3.7.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify water level of UHS is ~ 831.25 ft mean sea In accordance level. with the Surveillance Frequency Control Program SR 3.7.9.2 Verify average water temperature of UHS is ~ 90°F. In accordance with the Surveillance Frequency Control Program SR 3.7.9.3 Operate each cooling tower fan for ~ 15 minutes in In accordance both the fast and slow speed. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-27 Amendment No. 202

CREVS 3.7.10 SURVEILLANCE REQUIREMENTS SU RVEI LLANCE FREQUENCY SR 3.7.10.1 Operate each CREVS train pressurization filter unit In accordance for ~ 10 continuous hours with the heaters operating with the and each CREVS train filtration filter unit for Surveillance

~ 15 minutes. Frequency Control 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 simulated actuation signal. with 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. Envelope Habitability Program CALLAWAY PLANT 3.7-31 Amendment No. 202

CRACS 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify each CRACS train has the capability to remove In accordance the assumed heat load. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-34 Amendment No. 202

Emergency Exhaust System 3.7.13 SURVEILLANCE REQUIREMENTS SU RVEI LLANCE FREQUENCY SR 3.7.13.1 Operate each EES train for ~ 10 continuous hours In accordance with the heaters operating. with 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 simulated actuation signal. with the Surveillance Frequency Control Program SR 3.7.13.4 Verify one EES train can maintain a negative pressure In accordance

~ 0.25 inches water gauge with respect to with the atmospheric pressure in the auxiliary building during Surveillance the SIS mode of operation. Frequency Control Program SR 3.7.13.5 Verify one EES train can maintain a negative pressure In accordance

~ 0.25 inches water gauge with respect to with the atmospheric pressure in the fuel building during the Surveillance FBVIS mode of operation. Frequency Control Program CALLAWAY PLANT 3.7-38 Amendment No. 202

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 the storage racks.

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

ACTIONS COMPLETION CONDITION REQUIRED ACTION 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 above the storage racks. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-40 Amendment No. 202

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 within limit. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-42 Amendment No. 202

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 5 0.10 IlCi/gm DOSE EQUIVALENT 1-131.

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

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

AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify the specific activity of the secondary coolant is In accordance 50.10 IlCilgm DOSE EQUIVALENT 1-131. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.7-45 Amendment No. 202

SSIVs 3.7.19 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.19.1 Verify the isolation time of each automatic SSIV is In accordance with within limits. the Inservice Testing Program SR 3.7.19.2 Verify each automatic SSIV in the flow path actuates In accordance to the isolation position on an actual or simulated with the actuation signal. Surveillance Frequency Control Program CALLAWAY PLANT 3.7-47 Amendment No. 202

AC Sources - Operating 3.8.1 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A, B, C, 0, E, AND or F not met.

G.2 Bein MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> H. Three or more AC sources H.1 Enter LCO 3.0.3. Immediately inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE ~REQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power In accordance availability for each required offsite circuit. with the Surveillance Frequency Control Program (contlnueo)

CALLAWAY PLANT 3.8-6 Amendment No. 202

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 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 achieves steady state voltage ~ 3740 V and ~ 4320 V, with the and frequency ~ 58.8 Hz and ~ 61.2 Hz. Surveillance Frequency Control Program CALLAWAY PLANT 3.8-7 Amendment No. 202

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 operates for;?: 60 minutes at a load;?: 5580 kW and with the

6201 kW. Surveillance Frequency Control Program SR 3.8.1.4 Verify each fuel oil transfer pump starts on low level in In accordance the associated day tank standpipe. with the Surveillance Frequency Control Program SR 3.8.1.5 Check for and remove accumulated water from each In accordance day tank. with the Surveillance Frequency

! Control Program SR 3.8.1.6 Verify each fuel oil transfer system operates to In accordance transfer fuel oil from storage tank to the day tank. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.8-8 Amendment No. 202

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 achieves in s; 12 seconds, voltage ~ 3740 V and with the s; 4320 V, and frequency ~ 58.8 Hz and s; 61.2 Hz. Surveillance Frequency Control Program SR 3.8.1.8 Not used.

SR 3.8.1.9 Not used.

SR 3.8.1.10 Verify each DG operating at a power factor $ 0.9 and In accordance

~ 0.8 does not trip and voltage is maintained s; 4784 V with the and frequency is maintained s; 65.4 Hz during and Surveillance following a load rejection of ~ 5580 kWand Frequency s; 6201 kW. Control Program (contInued)

CALLAWAY PLANT 3.8-9 Amendment No. 202

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 signal: with the Surveillance

a. De-energization of emergency buses; Frequency Control Program

b. Load shedding from emergency buses;

c. DG auto-starts from standby condition and:

1. energizes permanently connected loads in s 12 seconds,

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

3. maintains steady state voltage;;:: 3740 V and s 4320 V,

4. maintains steady state frequency

58.8 Hz and s 61.2 Hz, and

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

5 minutes.

CALLAWAY PLANT 3.8-10 Amendment No. 202

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 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 safety injection signal In accordance each DG auto-starts from standby condition and: with the Surveillance

a. In s 12 seconds after auto-start and during Frequency tests, achieves voltage:::>: 3740 V and s 4320 V; Control Program

b. In s 12 seconds after auto-start and during tests, achieves frequency:::>: 58.8 Hz and s 61.2 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 load sequencer from the offsite power system.

CALLAWAY PLANT 3.8-11 Amendment No. 202

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 actual or simulated loss of voltage signal on the with the emergency bus concurrent with an actual or Surveillance simulated safety injection signal except: Frequency Control 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.

CALLAWAY PLANT 3.8-12 Amendment No. 202

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.14 --------------------------- NOTE S -------------------------------

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

2. The DG may be loaded to:2: 5580 kW and

~ 6201 kW for the entire test period if auto connected design loads are less than 6201 kW.

Verify each DG operating at a power factor ~ 0.9 and In accordance

2: 0.8 operates for :2: 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

• with the

• Surveillance

a. For :2: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded :2: 6600 kW and i Frequency

~ 6821 kW; and Control Program

b. For the remaining hours of the test loaded

2: 5580 kW and ~ 6201 kW.

SR 3.8.1.15 --------------------------- NOTE S -------------------------------

1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated :2: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded :2: 5580 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 ~ 12 seconds, In accordance voltage :2: 3740 V, and ~ 4320 V and frequency with the

2: 58.8 Hz and ~ 61.2 Hz.

• Surveillance Frequency Control Program CALLAWAY PLANT 3.8-13 Amendment No. 202

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.16 ------------------------------ NOT E ------------------------------

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

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

b. Transfers loads to offsite power source; and

c. Returns to ready-to-Ioad 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 In accordance connected to its bus, an actual or simulated Safety with the Injection signal overrides the test mode by: Surveillance Frequency

a. Returning DG to ready-to-Ioad operation; and Control Program

b. Automatically energizing the emergency load from offsite power.

~comlnuea)

CALLAWAY PLANT 3.8-14 Amendment No. 202

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOT E ------------------------------

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 within +/- 10% of design interval for each LOCA and with the shutdown load sequencer. Surveillance Frequency Control Program CALLAWAY PLANT 3.8-15 Amendment 202

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOT ES -------------------------------

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 signal in conjunction with an actual or simulated with the Safety Injection signal: Surveillance Frequency

a. De-energization of emergency buses; Control Program

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 LOCA load sequencer,

3. achieves steady state voltage ~ 3740 V and 54320 V,

4. achieves steady state frequency

~ 58.8 Hz and ~ 61.2 Hz, and

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

~ 5 minutes.

CALLAWAY PLANT 3.8-16 Amendment 202

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 condition, each DG achieves, in ~ 12 seconds, with the voltage:2': 3740 V and ~ 4320 V, and frequency Surveillance

58.8 Hz and ~ 61.2 Hz. Frequency Control Program 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 In accordance the load shedder and emergency load sequencer. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.8-17 Amendment No. 202

AC Sources - Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources - Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Class 1E AC electrical power distribution subsystem required by LCO 3.8.10, "Distribution Systems - Shutdown"; and

b. One diesel generator (DG) capable of supplying one train of the onsite Class 1E AC electrical power distribution subsystems required by LCO 3.8.10; and

c. The shutdown portion of one Load Shedder and Emergency Load Sequencer (LSELS) associated with the required DC and AC electrical power distribution train.

APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.

ACTIONS

---

NOT E ----------------------------------------------------------

LCO 3.0.3 is not applicable.

COMPLETION CONDITION REQUIRED ACTION TIME A. One required offsite circuit ------------------- NOTE ------------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.10, with the required train de-energized as a result of Condition A.

A.1 Declare affected required Immediately feature(s) with no offsite power available inoperable.

(continued)

CALLAWAY PLANT 3.8-18 Amendment No. 202 I

AC Sources - Shutdown 3.8.2 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A One required offsite circuit A.2.1 Suspend CORE Immediately inoperable. ALTERATIONS.

(continued)

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

AND A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

A.2.4 Initiate action to restore Immediately required offsite power circuit to OPERABLE status.

(continued)

CALLAWAY PLANT 3.8-19 Amendment 202 I

AC Sources - Shutdown 3.8.2 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. One required DG B.1 Suspend CORE Immediately inoperable. ALTERATIONS.

AND B.2 Suspend movement of Immediately irradiated fuel assemblies.

AND B.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

AND B.4 Initiate action to restore Immediately required DG to OPERABLE status.

C. One required LSELS C.1 Declare the affected DG Immediately (shutdown portion) and offsite circuit inoperable. inoperable.

CALLAWAY PLANT 3.8-20 Amendment No. 202 I

AC Sources - Shutdown 3.8.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 --------------------------------- NOTE --------------------------

The following SRs are not required to be performed:

SR 3.8.1.3, 3.8.1.10, SR 3.8.1.11, SR 3,8.1.14 through SR 3.8.1.16, and SR 3,8.1.18, (Shutdown Load Sequencer only).

For AC sources required to be OPERABLE, the In accordance with following SRs are applicable: applicable SRs SR 3.8.1.1 SR 3.8.1.11 SR 3.8,1.2 SR 3.8.114 SR 3.8.1.3 SR 3,8.1.15 SR 3.8.1.4 SR 3.8.1.16 SR 3.8.1.5 SR 3.8.1.18 (shutdown load SR 3.8.1.6 sequencer only)

SR 3,8.1.7 SR 3.8.1.21 (shutdown load SR 3.8.1.10 sequencer only)

CALLAWAY PLANT 3.8-21 Amendment No. 202 I

Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8 ELECTRICAL POWER SYSTEMS 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air LCO 3.8.3 The stored diesel fuel oil, lube oil, and starting air subsystem shall be within limits for each required diesel generator (DG).

APPLICABILITY: When associated DG is required to be OPERABLE.

ACTIONS

---

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

Separate Condition entry is allowed for each DG.

COMPLETION CONDITION REQUIRED ACTION TIME A. One or more DGs with fuel A.1 Restore fuel oil level to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> level < 80,900 gal and within limits.

> 69,800 gal in storage tank.

B. One or more DGs with lube B.1 Restore lube oil 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> oil inventory < 750 gal and inventory to within limits.

> 686 gal.

C. One or more DGs with C.1 Restore fuel oil total 7 days stored fuel oil total particulates within limit.

particulates not within limit.

D. One or more DGs with new D.1 Restore stored fuel oil 30 days fuel oil properties not within properties to within limits. limits.

(continued)

CALLAWAY PLANT 3.8-22 Amendment No. 202 I

Diesel Fuel Oil, Lube Oil. and Starting Air 3.8.3 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME One or more DGs with two E.1 Restore two starting air 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> starting air receivers in receivers with pressure service with pressure 2:: 435 psig.

< 435 psig and 2:: 250 psig.

QB One or more DGs with only E.2 Restore one starting air 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> one starting air receiver in receiver with pressure service with pressure 2:: 610 psig.

< 610 psig and 2:: 300 psig.

F. Required Action and F.1 Declare associated DG Immediately associated Completion inoperable.

Time not met.

QB One or more DGs diesel fuel oil, lube oil, or starting air subsystems not within limits for reasons other than Condition A, B, C, D.

or E.

CALLAWAY PLANT 3.8-23 Amendment No. 202 I

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 ~ 80,900 gal In accordance of fuel. with 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 psig or pressure in one starting air receiver is ~ 610 with the psig, for each DG starting air subsystem. Surveillance Frequency Control Program SR 3.8.3.5 Check for and remove accumulated water from each In accordance fuel oil storage tank. with the Surveillance Frequency Control Program SR 3.8.3.6 Not used.

CALLAWAY PLANT 3.8-24 Amendment No. 202

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

B. Require Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is ~ 130.2 V on float In accordance charge. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.8-25 Amendment No. 202

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 connectors. with the Surveillance Frequency Control Program Verify battery connection resistance is ~ 69E-6 ohm for cell to cell connections and ~ 69E-6 ohm for terminal connections.

SR 3.8.4.3 Verify battery cells, cell plates, and racks show no In accordance visual indication of physical damage or abnormal with the deterioration that could degrade battery performance. Surveillance Frequency Control Program SR 3.8.4.4 Remove visible terminal corrosion, verify battery cell In accordance to cell and terminal connections are clean and tight, with the and are coated with anti-corrosion material. Surveillance Frequency Control Program SR 3.8.4.5 Verify battery connection resistance is ~ 69E-6 ohm In accordance for cell to cell connections, and ~ 69E-6 ohm for with the terminal connections. Surveillance Frequency Control Program SR 3.8.4.6 Verify each battery charger supplies c 300 amps at In accordance c 130.2 V for c 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. with the Surveillance Frequency Control Program (contlnueo)

CALLAWAY PLANT 3.8-26 Amendment No. 202

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 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, or4.

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

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

Verify battery capacity is ~ 80% of the manufacturer's In accordance rating when subjected to a performance discharge with the test or a modified performance discharge test. Surveillance Frequency Control Program 18 months when battery shows degradation or has reached 85% of expected life CALLAWAY PLANT 3.8-27 Amendment No. 202

DC Sources - Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources - Shutdown LCO 3.8.5 The Train A or Train B DC electrical power subsystem shall be OPERABLE to support one train of the DC electrical power distribution subsystems required by LeO 3.8.10, "Distribution Systems - Shutdown."

APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.

ACTIONS

---

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

LCO 3.0.3 is not applicable.

COMPLETION CONDITION REQUIRED ACTION TIME A. Required DC electrical A.1 Declare affected required Immediately power subsystem feature(s) inoperable.

inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

(continued)

CALLAWAY PLANT 3.8-28 Amendment No. 202 I

DC Sources - Shutdown 3.8.5 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. Required DC electrical A.2.3 Suspend operations Immediately power subsystem involving positive inoperable. reactivity additions that (continued) could result in loss of required SDM or boron concentration.

A.2.4 Initiate action to restore Immediately required DC electrical power subsystem to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1 --------------------------------- NOTE ---------------------------

The following SRs are not required to be performed:

SR 3.8.4.6, SR 3.8.4.7, and SR 3.8.4.8.

For DC sources required to be OPERABLE, the In accordance with following SRs are applicable: applicable SRs SR 3.8.4.1 SR 3.8.4.4 SR 3,8.4,7 SR 3.8.4.2 SR 3.8.4.5 SR 3.8.4.8 SR 3.8.4.3 SR 3.8.4.6 CALLAWAY PLANT 3.8-29 Amendment 202 I

Battery Cell Parameters 3.8.6 3.8 ELECTRICAL POWER SYSTEMS 3.8.6 Battery Cell Parameters LCO 3.8.6 Battery cell parameters for Train A and Train B batteries shall be within the limits of Tabe 3.8.6-1.

APPLICABILITY: When associated DC electrical power subsystems are required to be OPERABLE.

ACTIONS

---

NOTE ----------------------------------------------------------.

Separate Condition entry is allowed for each battery.

COMPLETION CONDITION REQUIRED ACTION TIME A One or more batteries with A1 Verify pilot cells 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> one or more battery cell electrolyte level and float parameters not within voltage meet Category A or B limits. Table 3.8.6-1 Category C limits.

AND A2 Verify battery cell 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> parameters meet Table 3.8.6-1 Category C AND limits.

Once per 7 days thereafter AND A3 Restore battery cell 31 days parameters to Category A and B limits of Table 3.8.6-1.

(continued)

CALLAWAY PLANT 3.8-30 Amendment No. 202 I

Battery Cell Parameters 3.8.6 ACTIONS (continued)

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

Time of Condition A not met.

One or more batteries with average electrolyte temperature of the representative cells < 60 OF.

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

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

CALLAWAY PLANT 3.8-31 Amendment No. 202

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 Category B limits. with the Surveillance Frequency Control Program Once within 7 days after a battery discharge

< 110 V Once within 7 days after a battery overcharge

> 150 V SR 3.8.6.3 Verify average electrolyte temperature of In accordance representative cells is 2: 60°F. with the Surveillance Frequency Control Program CALLAWAY PLANT Amendment No. 202

Battery Cell Parameters 3.8.6 TABLE 3.8.6-1 (PAGE 1 OF 1)

BATTERY CELL PARAMETERS REQUIREMENTS CATEGORY A: CATEGORYC:

LIMITS FOR EACH CATEGORY B: ALLOWABLE LIMITS PARAMETER DESIGNATED LIMITS FOR EACH FOR EACH PILOT CELL CONNECTED CELL CONNECTED CELL Electrolyte Level > Minimum level > Minimum level Above top of plates, indication mark, and indication mark, and and not overflowing

<::; 114 inch above  :<::; 114 inch above maximum level maximum level indication mark(a) indication mark(a)

Float Voltage ~ 2.13 V ~ 2.13 V > 2.07 V Specific Gravity(b)(c) ~ 1.200 ~ 1.195 Not more than 0.020 below average of all AND connected cells Average of all AND connected cells

> 1.205 Average of all connected cells

~ 1.195 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided it is not overflowing.

(b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float charge.

(c) A battery charging current of < 2 amps when on float charge is acceptable for meeting specific gravity limits following a battery recharge, for a maximum of 7 days. When charging current is used to satisfy specific gravity requirements, specific gravity of each connected cell shall be measured prior to expiration of the 7 day allowance.

CALLAWAY PLANT 3.8-33 Amendment No. 202 I

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 COMPLETION CONDITION REQUIRED ACTION 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> CALLAWAY PLANT 3.8-34 Amendment No. 202 I

Inverters - Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage, and alignment to In accordance required AC vital buses. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.8-35 Amendment No. 202

Inverters - Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Inverters - Shutdown LCO 3.8.8 The Train A or Train B inverters shall be OPERABLE to support one train of the onsite Class 1E AC vital bus electrical power distribution subsystems required by LeO 3.8.10, "Distribution Systems - Shutdown."

APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.

ACTIONS

---

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

LCO 3.0.3 is not applicable.

COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required A.1 Declare affected required Immediately inverters inoperable. feature(s) inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies .

( continued)

CALLAWAY PLANT 3.8-36 Amendment No. 202 I

Inverters - Shutdown 3.8.8 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required A.2.3 Suspend operations Immediately inverters inoperable. involving positive (continued) reactivity additions that could result in loss of required SDM or boron concentration.

A.2.4 Initiate action to restore Immediately required inverters to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct inverter voltage, and alignments to In accordance required AC vital buses. with the Surveillance Frequency Control Program CALLAWAY PLANT Amendment 202

Distribution Systems - Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems - Operating LCO 3.8.9 Train A and Train B AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE.

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

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One AC electrical power A.1 Restore AC electrical 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> distribution subsystem power distribution inoperable. subsystem to OPERABLE status.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from discovery of failure to meet LCO B. One AC vital bus B.1 Restore AC vital bus 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> subsystem inoperable. subsystem to OPERABLE status.

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from discovery of failure to meet LCO (continued)

CALLAWAY PLANT 3.8-38 Amendment No. 202 I

Distribution Systems - Operating 3.8.9 ACTIONS (continued)

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

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

SURVEILLANCE REQUIREMENTS SU RVEI LLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to In accordance required AC, DC, and AC vital bus electrical power with the distribution subsystems. Surveillance Frequency Control Program CALLAWAY PLANT 3.8-39 Amendment No. 202

Distribution Systems - Shutdown 3.8.10 3.8 ELECTRICAL POWER SYSTEMS 3.8.10 Distribution Systems - Shutdown LCO 3.8.10 The necessary portion of the Train A or Train B AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE to support one train of equipment required to be OPERABLE.

APPLICABILITY: MODES 5 and 6, During movement of irradiated fuel assemblies.

ACTIONS

---

NOT E ----------------------------------------------------------

LCO 3.0.3 is not applicable.

COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required AC, A.1 Declare associated Immediately DC, or AC vital bus supported required electrical power distribution feature(s) inoperable.

subsystems inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

(continued)

CALLAWAY PLANT 3.8-40 Amendment No. 202 I

Distribution Systems - Shutdown 3.8.10 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more required AC, A.2.3 Suspend operations Immediately DC, or AC vital bus involving positive electrical power distribution reactivity additions that subsystems inoperable. could result in loss of (continued) required SDM or boron concentration.

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 required AC, DC, and AC vital bus electrical power with the distribution subsystems. Surveillance Frequency Control Program CALLAWAY PLANT 3.8-41 Amendment 202

Boron Concentration 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit. In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.9-2 Amendment No. 202

Unborated Water Source Isolation Valves 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify each valve that isolates unborated water In accordance sources is secured in the closed position. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.9-4 Amendment No. 202

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 CALLAWAY PLANT 3.9-6 Amendment No. 202

Containment Penetrations 3.9.4 SURVEILLANCE REQUIREMENTS SLI RVEI LLANCE FREQUENCY SR 3.9.4.1 Verify each required containment penetration is in the In accordance required status. with 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 valve actuates to the isolation position on a manual with the actuation signal. Surveillance Frequency Control Program CALLAWAY PLANT 3.9-8 Amendment No. 202

RHR and Coolant Circulation - High Water Level 3.9.5 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. RHR loop requirements not A.3 Initiate action to satisfy Immediately met. RHR loop requirements.

(continued)

AND AA Close all containment 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 reactor coolant at a flow rate of 2 1000 gpm. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.9-10 Amendment 202

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

COMPLETION CONDITION REQUIRED ACTION TIME B. No RHR loop in operation. B.1 Suspend operations that Immediately would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LeO 3.9.1.

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

AND B.3 Close all containment 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 reactor coolant at a flow rate of?: 1000 gpm. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.9-12 Amendment No. 202

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

SURVEILLANCE FREQUENCY SR 3.9.6.2 Verify correct breaker alignment and indicated power In accordance available to the required RHR pump that is not in with the operation.

CALLAWAY PLANT 3.9-13 Amendment No. 202

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 :2 23 ft above the top of reactor vessel flange.

APPLICABILITY: During movement of irradiated fuel assemblies within containment.

ACTIONS COMPLETION CONDITION REQUIRED ACTION 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 :2 23 ft above the In accordance top of reactor vessel flange. with the Surveillance Frequency Control Program CALLAWAY PLANT 3.9-14 Amendment No. 202

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 Control Room Envelope Habitability Program (continued)

e. The quantitative limits on unfiltered air inleakage into 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 -now rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air leakage 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 CRE and CBE as required by paragraphs c and d, respectively.

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

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

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

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

CALLAWAY PLANT 5.0-21 Amendment 202

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 202 TO FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By letter dated August 5, 2010 (Reference 1), as supplemented by letters dated March 23, May 3, and July 25, 2011 (References 2, 3, and 4, respectively), Union Electric Company (UE, the licensee) proposed changes to the Technical Specifications (TSs) for Callaway Plant, Unit 1.

The supplemental letters dated March 23, May 3, and July 25, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 11, 2011 (76 FR 1649).

The amendment would revise the TSs by relocating specific surveillance frequencies to a licensee-controlled program with the guidance of Nuclear Energy Institute (NEI) 04-10, "Risk Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies" (Reference 5). The requested change is the adoption of NRC-approved TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b" (Reference 6). When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls section of the TSs. All surveillance frequencies can be relocated except:

Frequencies that reference other approved programs for the specific interval (such as the In-Service 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");

Enclosure 2

-2

• 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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").

A new program is added to the administrative controls of TS Section 5 as TS 5.5.8, "Surveillance Frequency Control Program." The new program is called the SFCP and describes the requirements for the program to control changes to the relocated surveillance frequencies.

The TS Bases for each of the affected surveillance requirements are revised to state that the frequency is set in accordance with the SFCP. Some surveillance requirements Bases do not contain a discussion of the frequency. In these cases, the Bases describing the current frequency were added to maintain consistency with the Bases for similar surveillances. These instances are noted in the markup along with the source of the text. The proposed licensee changes to the Administrative Controls of the TSs to incorporate the SFCP include a specific reference to Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," Revision 1 (Reference 5) as the basis for making any changes to the surveillance frequencies once they are relocated out of the TSs.

By letter dated September 19,2007, the NRC staff approved NEI 04-10, Revision 1 (Reference 7), as acceptable for referencing in licensing actions to the extent specified and under the limitations delineated in NEI 04-10, and the safety evaluation providing the basis for NRC acceptance of NEI 04-10.

2.0 REGULATORY EVALUATION

In the "Final Policy Statement: Technical Specifications for Nuclear Power Plants" published in the Federal Register on 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 the Standard Technical Specifications (STS). In discussing the use of PSA in Nuclear Power Plant Technical Specifications, the Commission wrote in part: 1 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 10 CFR 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 1 The bracketed text is in the Federal Register.

-3 made * *

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

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

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

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

Approximately 2 years later, the NRC provided additional detail concerning the use of PRA in the "Final Policy Statement: Use of Probabilistic Risk Assessment in Nuclear Regulatory Activities" published in the Federal Register on August 16, 1995 (60 FR 42622). The Commission, in discussing the deterministic and probabilistic approach to regulation, and the Commission's extension and enhancement of traditional regulation, wrote in part:

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

The Commission provided its new policy, stating:

Although PRA methods and information have thus far been used successfully in nuclear regulatory activities, there have been concerns that PRA methods are not consistently applied throughout the agency, that sufficient agency PRAlstatistics expertise is not available, and that the Commission is not deriving full benefit from the large agency and industry investment in the developed risk assessment methods. Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency. This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the scope of PRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data.

Implementation of the policy statement will improve the regulatory process in three areas: Foremost, through safety decision making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees.

-4 Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA:

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

(2) PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with

[Title 10 of the Code of Federal Regulations (10 CFR)] 50.109 (Backfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

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

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

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, TS 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; (4) design features; and (5) administrative controls.

The regulations in 10 CFR 50. 36(c)(3) , "Surveillance requirements," state that, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

These categories will remain in the TSs. The new TS SFCP provides the necessary administrative controls to require that surveillances relocated to the SFCP are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that

-5 facility operation will be within safety limits, and that the limiting conditions for operation will be met. Changes to surveillance frequencies in the SFCP are made using the methodology contained in NEI 04-10, including qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, and recommended monitoring of structures, systems, and components (SSCs), and are required to be documented. Furthermore, changes to frequencies are subject to regulatory review and oversight of the SFCP implementation through the rigorous NRC review of safety-related SSC performance provided by the Reactor Oversight Program.

Licensees are required by TSs to perform surveillance test, calibration, or inspection on specific safety-related system equipment (e.g., reactivity control, power distribution, electrical, and instrumentation) to verify system operability. Surveillance frequencies, currently identified in TSs, are based primarily upon deterministic methods such as engineering judgment, operating experience, and manufacturer's recommendations. The licensee's use of NRC-approved methodologies identified in NEI 04-10 provides a way to establish risk-informed surveillance frequencies that complement the deterministic approach and support the NRC's traditional defense-in-depth philosophy.

The licensee's SFCP ensures that surveillance requirements specified in the TSs are performed at intervals sufficient to assure the above regulatory requirements are met. Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants" (the Maintenance Rule), and Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 (corrective action program), require licensee monitoring of surveillance test failures and implementing corrective actions to address such failures. One of these actions may be to consider increasing the frequency at which a surveillance test is performed. In addition, the SFCP implementation guidance in NEI 04-10 requires monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs. These requirements, and the monitoring required by NEI 04-10, ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions taken.

NRC Regulatory Guide (RG) 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 8), describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This regulatory guide also provides risk acceptance guidelines for evaluating the results of such evaluations.

NRC RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications" (Reference 9), describes an acceptable risk-informed approach specifically for assessing proposed permanent TS changes.

NRC RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Reference 10), 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 decision making for light-water reactors.

- 6

3.0 TECHNICAL EVALUATION

3.1 Proposed Changes The licensee's adoption of TSTF-425, Revision 3, for Callaway provides for administrative relocation of applicable surveillance frequencies, and for the addition of the SFCP to the administrative controls of TSs. TSTF-425, Revision 3, also requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425, Revision 3, included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200, Revision 1 (Reference 10). In accordance with NEI 04-10, 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 (Reference 8) and RG 1.177 (Reference 9) in support of changes to surveillance test intervals.

In its letter dated August 5,2010, as supplemented by letter dated May 3, 2011 (References 1 and 3, respectively), the licensee proposed to remove the following surveillance frequencies and relocate them to the licensee-controlled program:

3.1.1.1 3.3.2.3 3.3.8.1 3.4.12.2 3.5.5.1 3.7.9.3 3.8.3.4 3.1.2.1 3.3.2.4 3.3.8.2 3.4.12.3 3.6.2.2 3.7.10.1 3.8.3.5 3.1.4.1 3.3.2.5 3.3.8.3 3.4.12.4 3.6.3.1 3.7.10.3 3.8.4.1 3.1.4.2 3.3.2.6 3.3.8.4 3.4.12.5 3.6.3.2 3.7.11.1 3.8.4.2 3.1.5.1 3.3.2.7 3.3.8.5 3.4.12.6 3.6.3.3 3.7.13.1 3.8.4.3 3.1.6.2 3.3.2.8 3.3.9.1 3.4.12.8 3.6.3.6 3.7.13.3 3.8.4.4 3.1.6.3 3.3.2.9 3.3.9.2 3.4.12.9 3.6.3.7 3.7.13.4 3.8.4.5 3.1.8.2 3.3.2.10 3.3.9.3 3.4.13.1 3.6.3.8 3.7.13.5 3.8.4.6 3.1.8.3 3.3.2.11 3.3.9.4 3.4.13.2 3.6.4.1 3.7.15.1 3.8.4.7 3.1.8.4 3.3.2.12 3.3.9.5 3.4.14.1 3.6.5.1 3.7.16.1 3.8.4.8 3.1.9.1 3.3.2.13 3.3.9.6 3.4.14.2 3.6.6.1 3.7.18.1 3.8.6.1 3.2.1.1 3.3.2.14 3.4.1.1 3.4.15.1 3.6.6.2 3.7.19.2 3.8.6.2 3.2.1.2 3.3.3.1 3.4.1.2 3.4.15.2 3.6.6.3 3.8.1.1 3.8.6.3 3.2.2.1 3.3.3.2 3.4.1.3 3.4.15.3 3.6.6.5 3.8.1.2 3.8.7.1 3.2.3.1 3.3.4.1 3.4.1.4 3.4.15.4 3.6.6.6 3.8.1.3 3.8.8.1 3.2.4.1 3.3.4.2 3.4.2.1 3.4.15.5 3.6.6.7 3.8.1.4 3.8.9.1 3.2.4.2 3.3.4.3 3.4.3.1 3.4.16.1 3.6.6.8 3.8.1.5 3.8.10.1 3.3.1.1 3.3.5.1 3.4.4.1 3.4.16.2 3.6.7.1 3.8.1.6 3.9.1.1 3.3.1.2 3.3.5.2 3.4.5.1 3.5.1.1 3.6.7.2 3.8.1.7 3.9.2.1 3.3.1.3 3.3.5.3 3.4.5.2 3.5.1.2 3.7.2.2 3.8.1.10 3.9.3.1 3.3.1.4 3.3.5.4 3.4.5.3 3.5.1.3 3.7.3.2 3.8.1.11 3.9.3.2 3.3.1.5 3.3.6.1 3.4.6.1 3.5.1.4 3.7.5.1 3.8.1.12 3.9.4.1 3.3.1.6 3.3.6.2 3.4.6.2 3.5.1.5 3.7.5.3 3.8.1.13 3.9.4.2 3.3.1.7 3.3.6.3 3.4.6.3 3.5.2.1 3.7.5.4 3.8.1.14 3.9.4.3 3.3.1.8 3.3.6.4 3.4.7.1 3.5.2.2 3.7.6.1 3.8.1.15 3.9.5.1 3.3.1.9 3.3.6.5 3.4.7.2 3.5.2.3 3.7.7.1 3.8.1.16 3.9.6.1 3.3.1.10 3.3.6.6 3.4.7.3 3.5.2.5 3.7.7.2 3.8.1.17 3.9.6.2 3.3.1.11 3.3.7.1 3.4.8.1 3.5.2.6 3.7.7.3 3.8.1.18 3.9.7.1 3.3.1.13 3.3.7.2 3.4.8.2 3.5.2.7 3.7.8.1 3.8.1.19 3.3.1.14 3.3.7.3 3.4.9.1 3.5.2.8 3.7.8.2 3.8.1.20 3.3.1.16 3.3.7.4 3.4.9.2 3.5.4.1 3.7.8.3 3.8.1.21 3.3.2.1 3.3.7.5 3.4.11.1 3.5.4.2 3.7.9.1 3.8.3.1 3.3.2.2 3.3.7.6 3.4.12.1 3.5.4.3 3.7.9.2 3.8.3.2

-7 The table of contents was modified to reflect the changes above. This affected pages 1, 2, 3, and 4. In addition, the licensee proposed to modify the Administrative Controls section of the Callaway TSs by adding new TS 5.5.18, which would state, 5.5.18 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

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

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

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

3.2 RG 1.177 Five Key Safety Principles RG 1.177 identifies five key safety principles required for risk-informed changes to TSs. Each of these principles is addressed by the industry methodology document, NEI 04-10.

3.2.1 The Proposed Change Meets Current Regulations The regulations in 10 CFR 50.36(c)(3) require that TSs will include surveillances. NE104-10 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 In-service Testing Program or the Primary Containment Leakage Rate Testing Program. Thus, this proposed change meets the first key safety principle of RG 1.177 by complying with current regulations.

- 8 3.2.2 The Proposed Change Is Consistent With the Defense-in-Depth Philosophy Consistency with the defense-in-depth philosophy, the second key safety principle of RG 1.177, 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 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 General Design Criteria in 10 CFR Part 50, Appendix A, is maintained, TSTF-425, Revision 3, requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP, NEI 04-10 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 of RG 1.174 and RG 1.177 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 common cause failures. Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of common cause failures. Both the quantitative risk analysis and the qualitative considerations assure a reasonable balance of defense-in-depth is maintained to ensure protection of public health and safety, satisfying the second key safety principle of RG 1.177.

3.2.3 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

-9 conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis, or, if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the Updated Final Safety Analysis Report and bases to TS), since 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 the third key safety principle of RG 1.177 is satisfied.

3.2.4 When Proposed Changes Result in an Increase in Core Damage Freguency or Risk. the Increases Should Be Small and Consistent With the Intent of the Commission's Safety Goal Policy Statement RG 1.177 provides a framework for evaluating the risk impact of proposed changes to surveillance frequencies. This requires the 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 in the SFCP. NEI 04-10 satisfies the intent of RG 1.177 requirements for evaluating the change in risk, and for assuring that such changes are small.

3.2.4.1 Quality of the PRA The quality of the Callaway PRA is compatible 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 from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA.

The licensee used RG 1.200 to address the technical adequacy of the Callaway PRA.

RG 1.200 is NRC's developed regulatory guidance, which in revision one endorsed with comments and qualifications the use of the American Society of Mechanical Engineers (ASME)

RA-Sb-2005, "Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 11), NEI 00-02, Revision 1, "PRA Peer Review Process Guidelines" (Reference 12), and NEI 05-04, Revision 0, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard" (Reference 13). The licensee has performed an assessment of the PRA models used to support the SFCP against the requirements of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability category II of ASME RA-Sb-2005 was applied as the standard, and any identified deficiencies to those requirements are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including by the use of sensitivity studies where appropriate. (The NRC staff notes that in Revision 2, RG 1.200 endorsed with comments and qualifications an updated combined standard which includes requirements for fire, seismic, and other external events PRA models. The existing internal events standard was subsumed

- 10 into the combined standard, but the technical requirements are essentially unchanged. Since NEI 04-10, Revision 1, specifically identified the use of RG 1.200, Revision 1, to assess the internal events standard, the licensee's approach is reasonable and consistent with the approved methodology.)

The licensee identified that there are five open significant facts and observations (F&Os) remaining from its 2000 Industry PRA Peer Review. In 2006, a consultant review of the Callaway internal events PRA model was conducted using the ASME PRA Standard (Reference 13), which identified a number of supporting requirements which were not met at capability category II. The licensee provided a summary description and disposition of these items in Attachment 2, Tables 1 and 2, of its letter dated August 5,2010 (Reference 1), and provided additional information in its letter dated March 23, 2011 (Reference 3), in response to the NRC's staff request for additional information dated December 17, 2010 (Reference 14).

The staff reviewed the licensee's assessment of each of these identified deficiencies. The staff assessed these findings to determine whether they may be addressed and dispositioned for each surveillance frequency evaluation per the NEI 04-10 methodology. The staff's assessment is provided below.

2000 Peer Review Findings IE-7: Two issues identified for interfacing systems loss-of-coolant accident (ISLOCA) events were that only containment bypass locations were addressed, and that quantification of these events did not correlate variables for basic events. The licensee does not agree with the first peer review issue and cited supporting requirement IE-A2 of the American Nuclear Society (ANS)/ASME PRA Standard which requires inclusion of events resulting in " ... uncontrolled loss of core coolant outside containment. .. The licensee further identified that low-pressure piping II failures inside containment are adequately addressed by considering potential impacts of the piping failure on emergency core cooling system functionality. The NRC staff concludes that the licensee's dispOSition of this issue is acceptable.

For the second issue regarding correlation of ISLOCA variables, the licensee has developed a revised ISLOCA model which addresses this issue, and stated that the updated analYSis would be applied for any evaluation of ISLOCA risk. The licensee confirmed that this revised ISLOCA model is completed and available to use for sensitivity studies.

ST-1: State-of-the-art analyses were not used for determining the probability of failure of low pressure piping for ISLOCA events. The licensee has developed a revised ISLOCA model which addresses this issue, and stated that the updated analysis would be applied for any evaluation of ISLOCA risk. The licensee confirmed that this revised ISLOCA model is completed and available to use for sensitivity studies.

TH-3: The peer review recommended development of guidance for success criteria, which is considered a documentation issue which has no impact on this application.

L2-1: Containment isolation failure and internal flood initiators are not addressed in the LERF calculations. The licensee stated that these items would be addressed by developing separate PRA analyses if required for evaluation of a speCific surveillance frequency change as a sensitivity analysis per NEI 04-10.

- 11 L2-3: LERF calculations use split fractions to determine overall probability of a plant damage state, with no obvious relation of the split fractions to elementary phenomena or system failures.

The licensee stated that split fractions involving mitigating system failures were explicitly modeled, and that other split fractions are based on plant design features which do not usually change.

2006 Consultant Review Findings IE-3: A system-by-system review to identify initiating events was not found in the documentation. The licensee stated that this review has now been completed and documented with no new initiators identified. This deficiency can be addressed per the methodology of NEI04-10.

IE-4: Events occurring while the unit was not at power were not addressed in the initiating events analysis. The licensee stated that such events have now been considered, and no new initiators were identified. This deficiency can be addressed per the methodology of NEI 04-10.

IE-6: Operating experience for Callaway was not reviewed for initiating event precursors. The licensee stated that Callaway's experience has now been reviewed, and no new initiators were identified. This deficiency can be addressed per the methodology of NEI 04-10.

IE-?: Uncertainty bounds are not provided for initiating events. The licensee identified that uncertainty bounds have since been determined for the updated model, and so are available as needed to support this application. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

IE-8: Recovery events for loss of component cooling water and loss of service water are credited without sufficient analysis or data. The licensee discussed the basis for these events, which was a review of applicable cutsets by an individual with both PRA and system engineering experience and judgment as to the timeframes for repairs. A sensitivity study was also performed which showed that CDF increased by less than 2 percent if no credit is taken for any repairs. This deficiency is not significant, and can be addressed per the methodology of NEI04-10.

IE-lO: Plant availability is not factored into the initiating event frequencies. The licensee identified that the updated model has accounted for plant availability, and discussed the results of a sensitivity study on the loss-of-offsite power initiator which showed only a small impact on the results. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

IE-12: A comparison of the frequency of support system initiating event frequencies generated by fault tree models to generic data sources was not documented. The licensee stated that such comparisons for all initiators has been completed and documented, and no outlier frequencies were identified. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

IE-13: A deficiency related to the age and documentation of the ISLOCA analysis was dispositioned by the licensee with a revised analysis for the updated model.

- 12 IE-14: The initiating events documentation resides over several documents, which is not conducive to updates or peer reviews. The licensee stated that this is a documentation issue which has no effect on the application.

AS-2: The manual generation of transfer sequences, rather than using an event-tree structure, introduces the possibility of errors, although no errors were identified. The licensee identified an extensive review of these sequences with no errors identified, and therefore has no impact on this application.

AS-1/AS-3/AS-5/AS-7: Initiating event impacts may not be correctly captured. The licensee provided a detailed description of the deficiencies in the model, its investigation of the scope of the deficiencies, and sensitivity studies to bound the potential impact. The licensee has concluded that where there are actual model logic errors, the impact is not significant based on sensitivity studies showing only a 1 percent increase in CDF and on the fact that the errors do not appear in cutsets reviewed for the baseline PRA model. In other cases involving support system impacts for initiating events involving service water and component cooling water initiators, and room cooling for the switchgear rooms, the licensee's investigation found that the existing model was correct or conservative, and that the finding was not valid.

AS-4: The reactor coolant pump (RCP) sealloss-of-coolant accident (LOCA) model is out of date. The licensee provided a comparison of the probability of leakage for the existing model (based on WCAP-1 0541, "Reactor Coolant Pump Seal Performance Following a Loss of All AC

[Alternating Current) Power" (Reference 15) and the currently accepted model (Westinghouse Owners' Group WOG2000). A sensitivity study to the baseline PRA model to changes in the RCP seal failure probability was also provided which demonstrated the relative insensitivity of CDF. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

SC-1: Success criteria are not documented in a single location. The licensee identified this finding as a documentation issue which has no effect on the application.

SC-2: A check on the reasonableness of the success criteria was not documented. The licensee identified comparisons with other similar plant designs, and stated that this finding is a documentation issue which has no effect on the application.

SY-1: Main feedwater system dependencies on instrument air are not modeled, and the specific data used for undeveloped events in the instrument air model could not be verified. The licensee provided a sensitivity study using data from NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants" (Reference 16), for the instrument air system failure probability, and increasing the signal failure probabilities in undeveloped events by 10 percent. The result showed only a 0.59 percent increase in CDF, demonstrating that the results are not sensitive to this deficiency. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

SY-2: Common cause failure probabilities require updating, and battery chargers and breakers should be included in the scope of common cause failures modeled. The licensee identified sensitivity studies which demonstrate that the impact of these findings is not significant.

Specifically, the missing common cause failure modes were added to the model with no

- 13 discernable change in the results, and the existing common cause failure events were increased by 10 percent with only a 3.54 percent increase in CDF. The NRC staff agrees that the impact of this finding is minimal and, therefore, this deficiency can be addressed per the methodology of NEI 04-10.

HR-1/HR-2/HR-3: These findings identify deficiencies in the documentation of human reliability analyses, including documentation of "reasonableness" checks and key sources of uncertainty.

The licensee identified this finding as a documentation issue which has no effect on the application.

DA-1: Documentation of data collection is lacking. The licensee identified this finding as a documentation issue which has no effect on the application.

DA-2: Component groupings did not consider the characteristics of usage. The licensee stated that the updated PRA model corrected this deficiency, and component grouping are similar to the existing model. Therefore, the impact is not expected to be significant and, therefore, this deficiency can be addressed per the methodology of NEI 04-10.

DA-3: No justification was provided for the probabilities applied to three specific basic events.

The licensee identified a sensitivity analysis which doubled the probabilities resulting in a 0.03 percent increase in CDF, and further stated the issues were documentation related only.

Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

IF-1/IF-2/IF-3/IF-4/IF-5/IF-6: The internal flood analysis has findings in the areas of treatment of operator responses, use of plant-specific information in pipe-break frequencies, screening criteria, and lack of treatment of LERF. The licensee is developing a revised internal flooding model which addresses these deficiencies. Portions of this model, especially initiating event frequencies, are completed and can be directly used in the existing internal flood model to disposition items related to initiating events (IF-1, IF-3, IF-4, and IF-5) until the new internal flooding model is fully complete and able to be used.

Item IF-2 deals with quantitative screening criteria, and the licensee identified that no areas were screened, but instead had been conservatively evaluated. Internal flood areas will be reviewed for each specific surveillance frequency evaluation to determine if additional flood areas need to be evaluated. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

Item IF-6 addresses the lack of calculation of LERF for internal flooding events, and the licensee has identified that its existing CDF results will be evaluated using a conditional probability of a large early release to evaluate the increase in LERF for internal floods. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

QU-1: The "state-of-knowledge" correlation is not included in the current PRA. The licensee stated that the correlation can be addressed for a given surveillance interval evaluation as necessary. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

- 14 QU-2: Support system initiating events may have missing dependencies. The licensee stated that the peer review team identified that the supporting requirement from the standard (QU-B9) was met at capability category II and, therefore, this item is not an actual gap.

QU-5: A review of non-significant sequences and cutsets was not documented. The licensee stated that all sequences and non-significant cutsets are periodically reviewed as part of the normal PRA model update process or for specific applications and, therefore, this issue is a documentation issue which has no effect on the application.

QU-8: Separate quantification processes are applied to internal events, ISLOCA, and internal flooding, and should be considered to be integrated into a single process. The licensee stated that the risk contribution from each source is able to be determined, and so there is no effect on the application.

QU-9/QU-10/QU-11/QU-12: These findings identify deficiencies in the documentation of model quantification, including documentation of key assumptions sources of uncertainty, limitations, and definitions of "significant." The licensee identified this finding as a documentation issue which has no effect on the application.

LE-1: Some LERF contributors, identified in Table 4.5.9-3 of the standard, may not be addressed in the PRA model. The licensee identified that the finding was related to the lack of treatment of undetected containment structural integrity issues, and also identified a minimal impact on LERF. Other deficiencies related to the split fractions for induced steam generator tube rupture and high-pressure melt ejection phenomena were determined to be considered in the current LERF model. Therefore, this deficiency can be addressed per the methodology of NEI04-10.

LE-2: Uncertainty and sensitivity analyses are not included. The evaluation of uncertainty for each surveillance interval extension application can be accomplished as needed; therefore, this deficiency can be addressed per the methodology of NEI 04-10.

LE-3: Plant-specific analyses of induced steam generator tube rupture and secondary side isolation capability is not addressed. The licensee stated that this failure mode is not typically significant and would be addressed by sensitivity studies as needed. Therefore, this deficiency can be addressed per the methodology of NEI 04-10.

Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the NRC staff concludes that the level of PRA quality, combined with the proposed evaluation and disposition of the identified deficiencies, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177.

3.2.4.2 Scope of the PRA The licensee is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10 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. In cases where a PRA of

- 15 sufficient scope or where quantitative risk models were unavailable, the licensee uses bounding analyses, or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero.

The licensee proposes to use the Fire Induced Vulnerability Evaluation (FIVE) methodology, which was completed for the Individual Plant Examination of External Events (IPEEE).

Similarly, the IPEEE Seismic Margins Assessment will be used to provide seismic insights.

Other external hazards were assessed as insignificant during the IPEEE assessment.

Therefore, the risk contribution from these sources will be assessed either qualitatively or by bounding analyses for evaluation of surveillance frequency changes.

The NRC staff concludes that the licensee's evaluation methodology is sufficient to ensure 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 of RG 1.177.

3.2.4.3 PRA Modeling The licensee 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 the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy consistent with guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10.

The licensee 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 17),

and NUREG/CR-5497, "Common-Cause Failure Parameter Estimations" (Reference 18), as discussed in NEI 04-10.

Based on the above, the NRC staff concludes that, through the application of NEI 04-10, the Callaway 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 of RG 1.177.

3.2.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in the Callaway PRA include a standby time-related contribution and a cyclic demand-related contribution. NEI 04-10 criteria adjust the time-related failure contribution of SSCs affected by the proposed change to a surveillance frequency. This is consistent with RG 1.177, Section 2.3.3, which permits separation of the failure-rate contributions into demand and standby for evaluation of surveillance requirements. If the available data do not support distinguishing between the time-related and demand-related failures, then the change to a 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

- 16 be confirmed by the required monitoring and feedback implemented after the change in a surveillance frequency is implemented. The process requires consideration of qualitative sources of information with regard to potential impacts of test frequency on SSC performance, including industry and plant-specific operating experience, vendor recommendations, industry standards, and code-specified test intervals. Thus, the 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. The NRC staff concludes that, through the application of NEI 04-10, the licensee has employed reasonable assumptions with regard to extensions of surveillance test intervals, and is consistent with Regulatory Position 2.3.4 of RG 1.177.

3.2.4.S Sensitivity and Uncertainty Analyses NEI 04-10 requires sensitivity studies 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 capability category II of ASME PRA Standard (ASME RA-Sb-200S) (Reference 11). 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. Required monitoring and feedback of SSC performance, once the revised surveillance frequencies are implemented, will also be performed. The NRC staff concludes that, through the application of NEI 04-10, the licensee has appropriately considered the possible impact of PRA model uncertainty and sensitivity to key assumptions and model limitations, and is consistent with Regulatory Position 2.3.S of RG 1.177.

3.2.4.6 Acceptance Guidelines The licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using the guidance contained in NRC-approved NEI 04-10 in accordance with the TS SFCP. Each individual change to a surveillance frequency must show a risk impact below 1E-6 per year for a change to CDF, and below 1E-7 per year for a change to LERF. These are consistent with the limits of RG 1.174 for very small changes in risk.

Where the RG 1.174 limits are not met, the process either considers revised surveillance frequencies which are consistent with RG 1.174 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 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 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 'I E-S per year for a change to CDF, and below 1E-6 per year for a change to LERF, and the total CDF and total

- 17 LERF must be reasonably shown to be less than 1E-4 per year and 1E-5 per year.

respectively. These are consistent with the limits of RG 1.174 for acceptable changes in risk. as referenced by RG 1.177 for changes to surveillance frequencies. The NRC staff interprets this assessment of cumulative risk as a requirement to 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 staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with insignificant risk increases (less than 5E-8 CDF and 5E-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 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. The licensee's application of NEI 04-10 provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177. Therefore, the NRC staff concludes that the licensee's proposed methodology satisfies the fourth key safety principle of RG 1.177 by assuring any increase in risk is small and consistent with the intent of the Commission's Safety Goal Policy Statement.

3.2.5 The Impact of the Proposed Change Should Be Monitored USing Performance Measurement Strategies The licensee's adoption of TSTF-425, Revision 3, requires application of NEI 04-10 in the SFCP. NEI 04-10 requires performance monitoring of SSCs whose surveillance frequencies have 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 includes consideration of Maintenance Rule monitoring of equipment performance. In the event of 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. The performance monitoring and feedback specified in NEI 04-10 is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 of RG 1.177. Thus, the fifth key safety principle of RG 1.177 is satisfied.

3.3 Addition of Surveillance Frequency Control Program to Administrative Controls The licensee has included the SFCP and specific requirements into the Administrative Controls, TS Section 5.5.18, Surveillance Frequency Control Program, as follows:

This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications

- 18 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 the Surveillance Requirements for which the Frequency is controlled by the program.

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

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

The NRC staff concludes that the licensee's proposed program is consistent with the model application of TSTF-425, Revision 3, and, therefore, is acceptable.

3.4 Summary and Conclusions The NRC staff has reviewed the licensee's proposed relocation of certain surveillance frequencies to a licensee-controlled document, and controlling changes to surveillance frequencies in accordance with a new program, the SFCP, identified in the administrative controls of TSs. The SFCP and TS Section 5.5.18 reference NEI 04-10, 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 TSs to a licensee-controlled document, provided those frequencies are changed in accordance with NEI 04-10 which is specified in the Administrative Controls of the TSs.

The licensee's proposed adoption of TSTF-425, Revision 3, and risk-informed methodology of NEI 04-10, as referenced in the Administrative Controls of the TSs, satisfies the key principles of risk-informed decision making applied to changes to TSs as delineated in RG 1.177 and RG 1.174, in that:

• The proposed change meets current regulations;

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

• The proposed change maintains sufficient safety margins;

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

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

- 19 The NRC staff concludes that with the proposed relocation of surveillance frequencies to a licensee-controlled document and administratively controlled in accordance with the TS SFCP, the licensee continues to meet the regulatory requirement of 10 CFR 50.36 and, specifically, 10 CFR 50.36(c)(3), surveillance requirements.

4.0 REGULATORY COMMITMENTS In its letter dated August 5, 2010, the licensee made the following regulatory commitments:

• The existing Bases information describing the basis for the Surveillance Frequency will be relocated to the licensee-controlled Surveillance Frequency Control Program.

• [The licensee] proposes ... an Independent Decision-making Panel of qualified individuals with appropriate experience for recommending the acceptability of proposed surveillance frequency changes, in lieu of the site Maintenance Rule Expert Panel. This panel will be comprised of individuals whose experience levels are equal to or exceed the requirements of those on the Maintenance Rule Expert Panel. The individuals who will make up this panel will be designated by the senior management team that provides process oversight. The designated individuals will have expertise in the areas of probabilistic risk assessment, operations, maintenance, engineering, quality assurance, operating experience, and licensing. At least three individuals will have a minimum of 5 years experience at Callaway Plant or similar nuclear plants, and at least one individual will have worked on the modeling and updating of the PRA for Callaway Plant or similar plants for a minimum of 3 years. This level of experience and expertise will ensure that recommendations are well-considered and safety-focused. When developing potential changes, the panel will be augmented by the Surveillance Test Coordinator and at least one subject matter expert on the structure, system or component being evaluated.

• The Callaway Plant On-Site Review Committee (ORC) will review probabilistic and deterministic assessments to determine if a sufficient basis exists to support Surveillance Test Interval (STI) change proposals and to approve or disapprove proposed STI changes.

• During the interim period, i.e., from approval of this license amendment request until the availability of PRA Update 5, each of the F/Os (Table 1 of Attachment 2 [of the licensee's letter dated August 5,2010 (ULNRC-05725)]) and F&Os (Table 2 of Attachment 2 [of the licensee's letter dated August 5,2010 (ULNRC-05725)]) to Capability Category II of the Standard will be considered for each STI under evaluation, and, when appropriate, sensitivity studies will be performed to address selected F/Os or F&Os.

- 20 The licensee stated a completion date of October 27,2011, for the four commitments. This date was contingent upon NRC approval by April 30, 2011, and 180 days for implementation. The NRC staff acknowledges that this completion date will be adjusted by the licensee to account for NRC approval having been granted after April 30, 2011; therefore, these regulatory commitments are acceptable.

5.0 STATE CONSULTATION

In accordance with the NRC's regulations, the Missouri State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in Federal Register on January 11, 2011 (76 FR 1649). 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.

7.0 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) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1. Maglio, S., AmerenUE, letter to U.S. Nuclear Regulatory Commission, "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (LDCN 10-0020)," dated August 5,2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102250056).

2. Maglio, S., AmerenUE, letter to U.S. Nuclear Regulatory Commission, "License Amendment Application for a Technical Specification Change That Would Relocate Specific Surveillance Frequencies to a Licensee Controlled Program (LDCN 10-0020)

(TAC No. ME4506)," dated March 23,2011 (ADAMS Accession No. ML110830010).

- 21

3. Maglio, S., AmerenUE, letter to U.S. Nuclear Regulatory Commission, "License Amendment Application Supplement for a Technical Specification Change That Would Relocate Specific Surveillance Frequencies to a Licensee Controlled Program (LDCN 10-0020) (TAC No. ME4506)," dated May 3,2011 (ADAMS Accession No. ML111240412).

4. Graessle, L. H., AmerenUE, letter to U.S. Nuclear Regulatory Commission, "License Amendment Application Supplement for a Technical Specification Change That Would Relocate Specific Surveillance Frequencies to a Licensee Controlled Program (LDCN 10-0020) (TAC No. ME4506)," dated July 25,2011 (ADAMS Accession No. ML112061672).

5. Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," dated March 18, 2009 (ADAMS Accession No. ML090850642).

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

7. 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).

8. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis," November 2002 (ADAMS Accession No. ML023240437).

9. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.177, "An Approach for Plant Specific, Risk-Informed Decisionmaking: Technical Specifications," August 1998 (ADAMS Accession No. ML003740176).

10. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," January 2007 (ADAMS Accession No. ML070240001).

11. American Society of Mechanical Engineers (ASME) PRA Standard ASME RA-Sb-2005 ,

"Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Application."

12. Nuclear Energy Institute, NEI 00-02, Revision 1, "Probabilistic Risk Assessment (PRA)

Peer Review Process Guidance," May 2006 (ADAMS Accession No. ML061510621).

- 22

13. Nuclear Energy Institute, Revision 0, NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard," August 2006.

14. Polickoski, J. T., U.S. Nuclear Regulatory Commission, e-mail to T. B. Ellwood and S.

Maglio, Union Electric Company, Callaway Plant, Unit 1, Email, Request for Additional Information, RE: License Amendment Request for Technical Specification Change Regarding Risk Informed Justification (TAC No. ME4506)," dated December 17, 2010 (ADAMS Accession No. ML103510674).

15. Westinghouse Electric Corporation, "Reactor Coolant Pump Seal Performance Following Loss of All AC Power," WCAP-1 0541.

16. U.S. Nuclear Regulatory Commission, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," NUREG/CR-6928, February 2007 (ADAMS Accession No. ML070650650).

17. U.S. Nuctear Regulatory Commission, "Handbook of Methods for Risk-Based Analyses of Technical Specifications," NUREG/CR-6141, December 1994 (non-publicly available).

18. U.S. Nuclear Regulatory Commission, "Common-Cause Failure Parameter Estimations,"

NUREG/CR-5497, October 1998 (non-publicly available).

Principal Contributor: A. Howe Date: July 29, 2011

A. Heflin -2 A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, IRA by Lauren Kate Gibson fori Mohan C. Thadani, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosures:

1. Amendment No. 202 to NPF-30

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

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