ML091700296

Issuance of Amendment 183, Revise Technical Specification 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, to Extend the Surveillance Frequency on ESFAS Slave Relays from 92 Days to 18 Months

ML091700296
Person / Time
Site:	Wolf Creek
Issue date:	07/30/2009
From:	Balwant Singal Plant Licensing Branch IV
To:	Muench R Wolf Creek
Singal, Balwant, 415-3016, NRR/DORL/LPL4
References
TAC MD9470
Download: ML091700296 (45)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 30, 2009 Mr. Rick A. Muench President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION - ISSUANCE OF AMENDMENT RE:

TECHNICAL SPECIFICATION 3.3.2, "ENGINEERED SAFETY FEATURE ACTUATION SYSTEM (ESFAS) INSTRUMENTATION" (TAC NO. MD9470)

Dear Mr. Muench:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 183 to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated August 14, 2008.

The amendment revises Technical Specification (TS) 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," to extend the Surveillance Frequency on selected ESFAS slave relays from 92 days to 18 months.

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

Sincerely, b~ I \.U ~~ Ic&,,,,-\~

Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482

Enclosures:

1. Amendment No. 183 NPF-42

2. Safety Evaluation cc wlencls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 183 License No. NPF-42

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

A. The application for amendment to the Wolf Creek Generating Station (the facility)

Renewed Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated August 14, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

-2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-42 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 183, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

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

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: Jul y 30, 2009

ATIACHMENT TO LICENSE AMENDMENT NO. 183 RENEWED FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Replace the following pages of the Renewed Facility Operating License No. NPF-42 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.

The corresponding overleaf pages are provided to maintain document completeness.

Renewed Facility Operating License REMOVE INSERT 4 4 Technical Specifications REMOVE INSERT REMOVE INSERT i i 3.3-44 3.3-43 ii ii 3.3-45 3.3-44 3.3-29 3.3-29 3.3-46 3.3-45 3.3-30 3.3-47 3.3-46 3.3-31 3.3-30 3.3-48 3.3-47 3.3-32 3.3-31 3.3-49 3.3-48 3.3-33 3.3-32 3.3-50 3.3-49 3.3-34 3.3-33 3.3-51 3.3-50 3.3-35 3.3-34 3.3-52 3.3-51 3.3-36 3.3-35 3.3-53 3.3-52 3.3-37 3.3-36 3.3-54 3.3-53 3.3-38 3.3-37 3.3-55 3.3-54 3.3-39 3.3-38 3.3-56 3.3-55 3.3-40 3.3-39 3.3-57 3.3-56 3.3-41 3.3-40 3.3-58 3.3-57 3.3-42 3.3-41 3.3-59 3.3-58 3.3-43 3.3-42

4 (5) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 183, and the Environmental Protection Plan contained in Appendix 8, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

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

Deleted per Amendment No. 141.

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

Renewed License No. NPF-42 Amendment No. 183

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 -1\i10DE2 3.1-19 3.2 POWER DISTRIBUTION LIMITS 3.2-1 3.2.1 Heat Flux Hot Channel Factor (Fa(l))

(Fa Methodology) 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F~H) 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 3.3-21 3.3.3 Post Accident Monitoring (PAM) Instrumentation............................. 3.3-36 3.3.4 Remote Shutdown System 3.3-40 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation 3.3-43 Wolf Creek - Unit 1 Amendment No. 123, 173, 183

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6 Containment Purge Isolation Instrumentation 3.3-45 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation Instrumentation 3.3-49 3.3.8 Emergency Exhaust System (EES) Actuation Instrumentation.. 3.3-54 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.4-1 3.4.2 RCS Minimum Temperature for Criticality 3.4-5 3.4.3 RCS Pressure and Temperature (PfT) Limits 3.4-6 3.4.4 RCS Loops - MODES 1 and 2............ 3.4-8 3.4.5 RCS Loops - MODE 3....................................................................... 3.4-9 3.4.6 RCS Loops - MODE 4 3.4-12 3.4.7 RCS Loops - MODE 5, Loops Filled 3.4-14 3.4.8 RCS Loops - MODE 5, Loops Not Filled 3.4-17 3.4.9 Pressurizer 3.4-19 3.4.10 Pressurizer Safety Valves 3.4-21 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)....................... 3.4-23 3.4.12 Low Temperature Overpressure Protection (LTOP) System 3.4-26 3.4.13 RCS Operational LEAKAGE 3.4-31 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage 3.4-33 3.4.15 RCS Leakage Detection Instrumentation 3.4-37 3.4.16 RCS Specific Activity......................................................................... 3.4-41 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 3.5-3 3.5.3 ECCS - Shutdown 3.5-6 3.5.4 Refueling Water Storage Tank (RWST)........................................... 3.5-8 3.5.5 Seal Injection Flow 3.5-10 3.6 CONTAINMENT SYSTEMS 3.6-1 3.6.1 Containment...................................................................................... 3.6-1 3.6.2 Containment Air Locks...................................................................... 3.6-2 3.6.3 Containment Isolation Valves 3.6-7 3.6.4 Containment Pressure 3.6-14 3.6.5 Containment Air Temperature 3.6-15 3.6.6 Containment Spray and Cooling Systems 3.6-16 3.6.7 Spray Additive System 3.6-19 Wolf Creek - Unit 1 ii Amendment No. 123, 131, 157, 164, 167 4-7Q, 183

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

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

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS SR 3.3.2.4 Perform MASTER RELAY TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.2.5 Perform COT. 184 days SR 3.3.2.6 Perform SLAVE RELAY TEST. 18 months SR 3.3.2.7 --------------------------------NOTE-----------------------------

Verification of relay setpoints not required.

Perform TADOT. 18 months (continued)

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

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

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

Verification of setpoint not required for manual initiation functions.

Perform TADOT. 18 months SR 3.3.2.9 ------------------------------NOTE-------------------------------

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

Perform CHANNEL CALIBRATION. 18 months 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. 18 months on a STAGGERED TEST BASIS SR 3.3.2.11 -----------------------------NOTE-------------------------------

Verification of setpoint not required.

Perform TADOT. 18 months SR 3.3.2.12 Perform COT. 31 days Wolf Creek - Unit 1 3.3-30 Amendment No. 123, 131, 183

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

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

1. Safety Injection

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

b. Automatic Actuation 1.2,3,4 2 trains C SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

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

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

e. Steam Line Pressure 1,2,3(b) 3 per steam D SR 3.3.2.1  ;:.571 psig(C)

Low line SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

2. Containment Spray

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

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

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

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

(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 t, ;:. 50 seconds and t2 s 5 seconds.

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

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

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

3. Containment Isolation

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

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

4. Steam Line Isolation

a. Manual Initiation 2 F SR 3.3.2.8 NA

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

c. Automatic Actuation 2 trains G SR 3.3.2.6 NA Logic (MSFIS)

d. Containment Pressure 3 D SR 3.3.2.1  :;;18.3 psig

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

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

(i) Except when all MSIVs are closed.

Wolf Creek - Unit 1 3.3-32 Amendment No. 42-J, 175, 183

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

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

4. Steam Line Isolation (continued)

e. Steam Line Pressure (1) Low 3 per steam D SR 3.3.2.1 ~ 571 psig(C) line SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10 (2) Negative Rate 3 per steam D SR 3.3.2.1 ,,; 125(h) psi High line SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

5. Turbine Trip and Feedwater Isolation

a. Automatic Actuation 1,20},30} 2 trains G SR 3.3.2.2 NA Logic and Actuation SR 3.3.2.4 Relays (SSPS) SR 3.3.2.6

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

c. SG Water Level -High 1,20} 4 per SG SR 3.3.2.1 ,,;79.7% of High (P-14) SR 3.3.2.5 Narrow Range SR 3.3.2.9 Instrument Span SR 3.3.2.10

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

(continued)

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

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

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

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

(h) Time constant utilized in the ratellag controller is ~ 50 seconds.

(i) Except when all MSIVs are closed.

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

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

Wolf Creek - Unit 1 3.3-33 Amendment No. 123, 132, 175, 177, 183

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

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

6. Auxiliary Feedwater

a. Manual Initiation 1,2,3 1 per pump 0 SR 3.3.2.8 NA

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

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

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

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

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

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

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

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

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

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

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

7. Automatic Switchover to Containment Sump

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

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

Safety Injection

8. ESFAS Interlocks

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

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

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

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

---

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

Separate Condition entry is allowed for each Function.

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

inoperable.

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

met.

(continued)

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

PAM Instrumentation 3.3.3 ACTIONS (continued)

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

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

met.

E. As required by Required E.1 Be in MODE 3. 6 hours6.944444e-5 days <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 5.6.8.

Wolf Creek - Unit 1 3.3-37 Amendment No. 123, 157, 183 I

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 for each required 31 days instrumentation channel that is normally energized.

SR 3.3.3.2 --------------------------------NOT E-------------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. 18 months Wolf Creek - Unit 1 3.3-38 Amendment No. .:t-2-d, 183 I

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

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

1. Neutron Flux 2 E

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

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

4. RCS Pressure (Wide Range) 2 E

5. Reactor Vessel Water Level 2 F

6. Containment Normal Sump Water Level 2 E

7. Containment Pressure ( Normal Range) 2 E

8. Steam Line Pressure 2 per E steam generator

9. Containment Radiation Level (High Range) 2 F

10. Not Used

11. Pressurizer Water Level 2 E

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

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

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

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

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

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

18. Auxiliary Feedwater Flow Rate 4 E

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

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

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

OR One or more required ASP controls inoperable.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <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 /> Wolf Creek - Unit 1 3.3-40 Amendment No. 123, 155, 183 I

Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized.

SR 3.3.4.2 Verify each required auxiliary shutdown panel control 18 months circuit and transfer switch is capable of performing the intended function.

SR 3.3.4.3 ------------------------------NOTES-------------------------------

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

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

Perform CHANNEL CALIBRATION for each required 18 months instrumentation channel.

Wolf Creek - Unit 1 3.3-41 Amendment No. ~,183 I

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

Remote Shutdown System Functions FUNCTION REQUIRED CHANNELS

1. Source Range Neutron Flux a

2. Reactor Trip Breaker Position 1 per trip breaker

3. Pressurizer Pressure

4. RCS Wide Range Pressure

5. RCS Hot Leg Temperature

6. RCS Cold Leg Temperature

7. SG Pressure 1 per SG

8. SG Level 1 per SG

9. AFW Flow Rate

10. RCP Breakers 1 per pump

11. AFW Suction Pressure

12. Pressurizer Level

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

Wolf Creek - Unit 1 3.3-42 Amendment No.~, 183 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-kV NB bus of the loss of voltage Function and four channels per 4-kV NB bus of the degraded voltage Function shall be OPERABLE.

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

ACTIONS

---

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETIONTIME A, One or more Functions ---------------------NOT E-------------------

with one channel per bus The inoperable channel may be inoperable. 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.

A,1 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 B.1 Declare associated load Immediately with two or more channels shedder and emergency per bus inoperable. load sequencer (LSELS) inoperable.

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

Wolf Creek - Unit 1 3.3-43 Amendment No. ~,183 I

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

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

Verification of time delays is not required.

Perform TADOT. 31 days SR 3.3.5.3 Perform CHANNEL CALIBRATION with nominal Trip 18 months Setpoint and Allowable Value as follows:

a. Loss of voltage Allowable Value ~ 82.5V, 120V bus with a time delay of 1.0 + 0.2, -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 ~ 105.9V, 120V bus with a time delay of 119 +/- 11.6 sec.

Degraded voltage nominal Trip Setpoint 106.9V, 120V bus with a time delay of 119 sec.

SR 3.3.5.4 Verify LOP DG Start ESF RESPONSE TIMES are 18 months on a within limits. STAGGERED TEST BASIS Wolf Creek - Unit 1 3.3-44 Amendment No. 123, 128, 183 I

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

APPLICABILITY: According to Table 3.3.6-1.

ACTIONS

---

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

Separate Condition entry is allowed for each Function.

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

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

(continued)

Wolf Creek - Unit 1 3.3-45 Amendment No. ~,183 I

Containment Purge Isolation Instrumentation 3.3.6 ACTIONS (continued)

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

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

SURVEILLANCE REQUIREMENTS

---

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

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

SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.6.2 -------------------------------NOTE-------------------------------

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS (continued)

Wolf Creek - Unit 1 3.3-46 Amendment No. ~,183 I

Containment Purge Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.6.3 Perform COT. 92 days SR 3.3.6.4 ------------------------------NOTE---------------------------------

Verification of setpoint is not required.

Perform TADOT. 18 months SR 3.3.6.5 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.6 Verify Containment Purge Isolation ESF RESPONSE 18 months on a TIMES are within limits. STAGGERED I TEST BASIS Wolf Creek - Unit 1 3.3-47 Amendment No.~, 183 I

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

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

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

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

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

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

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

Isolation Phase A (a) During CORE ALTERATIONS.

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

(c) Trip setpoint concentration value (J.1Ci/cm 3 ) is to be established such that the actual submersion rate would not exceed mR/h in the containment building.

Wolf Creek - Unit 1 3.3-48 Amendment No. ~,183 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 3.3.7-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.7-1.

ACTIONS

---

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

Separate Condition entry is allowed for each Function.

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

(continued)

Wolf Creek - Unit 1 3.3-49 Amendment No. ~, 183 I

CREVS Actuation Instrumentation 3.3.7 ACTIONS (continued)

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

3.

---

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

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

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

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

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

(continued)

Wolf Creek - Unit 1 3.3-50 Amendment No.~, 183 I

CREVS Actuation Instrumentation 3.3.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D .1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A, B AND or C not met in MODE 1, 2, 3,or4. 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 Condition A, B or C not met in MODE 5 or AND 6, or during movement of irradiated fuel assemblies. E.2 Suspend movement of Immediately irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS

---

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

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

SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.7.2 Perform COT. 92 days (continued)

Wolf Creek - Unit 1 3.3-51 Amendment No. ~,183 I

CREVS Actuation Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS (continued)

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

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS SR 3.3.7.4 -----------------------------NOTE---------------------------------

Verification of setpoint is not required.

Perform TADOT. 18 months SR 3.3.7.5 Perform CHANNEL CALIBRATION. 18 months Wolf Creek - Unit 1 3.3-52 Amendment No. ~,183 I

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

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

1. Manual Initiation 1, 2, 3, 4, 5, 6, 2 SR 3.3.7.4 NA and (a)

2. Automatic Actuation Logic 1, 2, 3, 4, 5,6, 2 trains SR 3.3.7.3 NA and Actuation Relays (BOP and (a)

ESFAS)

3. Control Room Radiation 1, 2, 3, 4, 5,6, 2 SR 3.3.7.1 (b)

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

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

(a) During movement of irradiated fuel assemblies.

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

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

---

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

Separate Condition entry is allowed for each Function.

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

(continued)

Wolf Creek - Unit 1 3.3-54 Amendment No. ~, 183 I

EES Actuation Instrumentation 3.3.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIIIIIE B. ---------------N0 TE------------ B.1.1 Place one EES train in the Immediately Not applicable to Function FBVIS mode.

3.

---

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

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

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

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

(continued)

Wolf Creek - Unit 1 3.3-55 Amendment No. ~,183 I

EES Actuation Instrumentation 3.3.8 ACTIONS (continued)

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

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

SURVEILLANCE REQUIREMENTS

---

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

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

SURVEILLANCE FREQUENCY SR 3.3.8.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.8.2 Perform COT. 92 days SR 3.3.8.3 -------------------------------NOTE--------------------------------

The continuity check may be excluded.

Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS (continued)

Wolf Creek - Unit 1 3.3-56 Amendment No.~, 183 I

EES Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS (continued)

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

Verification of setpoint is not required.

Perform TADOr. 18 months SR 3.3.8.5 Perform CHANNEL CALIBRATION. 18 months Wolf Creek - Unit 1 3.3-57 Amendment No. ~,183 I

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

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

1. Manual Initiation (a) 2 SR 3.3.8.4 NA

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

3. Fuel Building Exhaust Radiation Gaseous (a) 2 SR 3.3.8.1 (b)

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

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

Wolf Creek - Unit 1 3.3-58 Amendment No. ~, 183 I

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 183 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482

1.0 INTRODUCTION

By application dated August 14, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082350072), Wolf Creek Nuclear Operating Corporation (the licensee) proposed a license amendment to change the technical specifications (TSs, Appendix A to Renewed Facility Operating License No. NPF-42) for Wolf Creek Generating Station (WCGS).

The proposed amendment would extend the surveillance test interval (STI) for the slave relays used in the engineered safety feature actuation system (ESFAS) from 92 days to 18 months.

The licensee justified its request to extend the STI based on the information contained in the Westinghouse Electric Corporation report, Westinghouse Commercial Atomic Power (WCAP) 13878-P-A, "Reliability Assessment of Potter & Brumfield MDR [Motor-driven Rotary] Series Relays," Revision 2, issued August 2000 (Proprietary).

In its letter dated August 14, 2008, the licensee proposes to make the following changes to the TS:

• Revise the table of contents to reflect the repagination resulting from the proposed change.

• Revise Surveillance Requirement (SR) 3.3.2.6 by deleting the note to the SR and by changing the surveillance test frequency from 92 days to 18 months.

• Delete SRs 3.3.2.13 and 3.3.2.14. In addition, delete the term "continued" on page 3.3-30 and place a double line in the table to indicate the end of the SR table.

Enclosure 2

-2

• Revise Table 3.3.2-1, "Engineered Safety Feature Actuation System Instrumentation," as follows:

Function 1.b., Safety Injection - Automatic Actuation Logic and Actuation Relays: Delete SR 3.3.2.13 as a required surveillance.

Function 3.a.(2), Containment Isolation - Phase A Isolation - Automatic Actuation Logic and Actuation Relays: Delete SR 3.3.2.13 as a required surveillance.

Function 5.a., Turbine Trip and Feedwater Isolation - Automatic Actuation Logic and Actuation Relays (SSPS): Delete SR 3.3.2.14 as a required surveillance.

Function 7.a., Automatic Switchover to Containment Sump - Automatic Actuation Logic and Actuation Relays: Revise SR 3.3.2.13 to SR 3.3.2.6.

2.0 REGULATORY EVALUATION

In Section 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations (10 CFR), the U.S. Nuclear Regulatory Commission (NRC) established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plant's TSs.

In 10 CFR 50.55a, "Codes and standards," paragraph (h)(2) states the following:

Protection systems. For nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements stated in either [Institute of Electrical and Electronic Engineers Standard] IEEE Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," or in IEEE Std. 603-1991, "Criteria for Safety Systems for Nuclear Power Generating Stations," and the correction sheet dated January 30, 1995. For nuclear power plants with construction permits issued before January 1, 1971, protection systems must be consistent with their licensing basis or may meet the requirements of IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.

The licensee meets the requirements of IEEE Std. 279 for WCGS because its construction permit was issued in 1977.

-3 In 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," the following General Design Criteria (GOG) pertain to the proposed TS changes:

• GDC 20, "Protection system functions," states the following:

The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

• GDC 21, "Protection system reliability and testability," states the following:

The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protective system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

• GDC 29, "Protection against anticipated operational occurrences," states the following:

The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

The licensee will continue to meet the requirements of IEEE Std. 279 and GDCs 20,21, and 29 because the proposed change does not affect the design capability, function, operation, or testing method of the slave relays.

3.0 TECHNICAL EVALUATION

The solid-state protection system (SSPS) performs the design logic for most engineered safety feature (ESF) equipment actuations and provides the status, permissive, and annunciator output signals to the control room. The actuation of ESF components is accomplished through master and slave relays. The SSPS energizes the master relays appropriate for the condition of the unit. Each master relay then energizes one or more slave relays, which then causes the actuation of the end devices. The master and slave relays are tested routinely to ensure their operability.

-4 Surveillance testing can identify relay failures before the relay is required to perform its intended function. However, relay testing has the potential to cause inadvertent ESF equipment actuation. Extending the STI reduces the number of surveillances performed on the relays and thus reduces the potential for unnecessary ESF equipment actuations.

In NRC Generic Letter (GL) 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation," dated September 27, 1993 (ADAMS Accession No. ML031070342), the NRC documented the results of a study of surveillance testing required by the TSs and recommended that, in certain cases, reducing the amount of testing at power will improve safety, decrease equipment degradation, and relieve licensee burden. Extending the STI for slave relays is consistent with the NRC's recommendation in GL 93-05.

Currently, at WCGS and other Westinghouse plants, slave relays for ESF equipment actuation are tested quarterly with the exception of some relays that the NRC has previously approved to be tested every 18 months. Westinghouse had submitted topical report WCAP-13878-P-A, Revision 2, which the NRC staff accepted by letter dated July 12, 2000 (ADAMS Accession No. ML003731486). (The NRC staff had accepted Revision 1 ofWCAP-13878-P-A by letters dated May 31, 1996, and October 26, 1998.)

The licensee's proposed change to extend the STI for slave relays is based on topical report WCAP-13878-P-A. In its letter dated May 31, 1996 (ADAMS Legacy Accession No.

0906110426), the NRC staff identified in its approval of the topical report certain plant-specific information that must be submitted. The licensee provided the following information in its submittal:

• Confirm the applicability of the WCAP-13878-P-A analyses for its plant.

The licensee has identified that the Potter & Brumfield (P&B) MDR slave relay models 4103-1 and 4121-1 are used at WCGS for ESFAS applications that require testing in accordance with the TS. These relays are bounded by WCAP-13878-P-A and have similar environmental conditions as those discussed in the topical report.

Therefore, NRC staff concludes that the licensee has adequately demonstrated the applicability of the topical report to WCGS.

• Ensure that its procurement program for P&B MDR is adequate for detecting the types of failures that are discussed in Office for Analysis and Evaluation of Operational Data Special Study Report No. AEOD/S93-06, "Potter & Brumfield Model MDR Rotary Relay Failures," dated September 5, 1990, and Supplement 1 to that report, dated November 27, 1991, as well as in a report from the San Onofre Nuclear Generating Station, dated July 21, 1995, relating to the requirements of 10 CFR Part 21, "Reporting of Defects and Noncompliance."

In its letter dated August 14, 2008, the licensee stated in its submittal that all the P&B MDR relays currently installed at WCGS meet the applicable requirements and that it currently procures the qualified replacement relays from Westinghouse. Moreover, the current manufacturer of MDR relays, Tyco Electronics Corporation, is surveyed periodically under a joint survey program conducted by the Nuclear Utility Procurement

-5 Issues Committee to ensure that standards of control are met in design, procurement, materials, manufacturing process, inspection, test and measurement, and test equipment. Based on the above, the NRC staff concludes that the licensee has adequately addressed its concern about refurbished relays.

• Ensure that all pre-1992 P&B MDR relays operated either in a normally energized mode or at a 20-percent duty cycle have been removed from ESFAS applications.

In its letter dated August 14, 2008, the licensee stated in its submittal that slave relays at WCGS are normally deenergized except for one relay that is normally energized (K637).

Also, the licensee has replaced all normally energized or 20-percent duty cycle relays with the post-1992 relays. Based on the above, the NRC staff concludes that the licensee has adequately addressed this item for its application.

• Ensure that the contact loading analysis for the P&B MDR relays has been performed to determine the acceptability of these relays.

The licensee has performed a technical review to determine the adequacy of contact loading of the P&B MDR relays that are subject to TS surveillance and has determined that the slave contacts are adequate for their application. Therefore, the NRC staff concludes that the licensee has adequately addressed the regulatory concerns regarding this matter.

• Reevaluate the adequacy of the extended STI if two or more P&B MDR ESFAS subgroup relays fail in a 12-month period.

The Maintenance Rule program provides the monitoring performance results of the P&B MDR surveillance test results. The Maintenance Rule program implements the requirements of 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," and provides instructions for the initiation, analysis, retrieval, trending, and periodic reporting of data-relative performance indicators of plant systems and components. This program also provides guidance to determine the cause for failures to meet performance criteria and for repetitive failures.

The functional failure guidance for the ESFAS specifies that a functional failure is any failure that results in a complete loss of train actuation. The performance criteria for the ESFAS is less than or equal to one functional failure of a train of actuation every 18 months. The failure of an MDR slave relay would be considered a functional failure and would result in an evaluation of the failure and of the adequacy of the extended STI.

Because the slave-relay testing is done at the 18-month refueling outage, the NRC staff concludes this acceptable.

Based on the above, the NRC staff concludes that WCAP-13878-P-A applies to WCGS, and therefore, can be used as a basis for extending the STI for P&B MDR slave relays. The staff evaluated the following proposed changes to WCGS TS:

• Revise the table of contents to reflect the repagination resulting from the proposed change.

-6 This is an editorial change and does not affect the requirement of the TS.

Therefore, the NRC staff concludes this change is acceptable.

• Revise SR 3.3.2.6 by deleting the note to the SR and by changing the surveillance test frequency from 92 days to 18 months.

The note to the SR exempts certain slave relays from the 92-day test interval because they were tested at the 18-month interval. Because the STI for this test is changed from 92 days to 18 months, there is no applicability or need for the note. Also, because the licensee has justified the application of topical report WCAP-13878-P-A to WCGS and cited applicable regulatory precedents for its use, the NRC staff concludes the 18-month STI acceptable.

• Delete SRs 3.3.2.13 and 3.3.2.14. In addition, delete the term "continued" on page 3.3-30 and place a double line in the table to indicate the end of the SR table.

SRs 3.3.2.13 and 3.3.2.14 pertain to the relays that were previously tested at the 18-month interval. Because the changes proposed to SR 3.3.2.6 cover this requirement, there is no need to include redundant test requirements. Therefore, the NRC staff concludes this change is acceptable. The other changes are editorial in nature and do not affect the TS requirements. Therefore, the NRC staff also concludes these changes are acceptable.

• Revise Table 3.3.2-1, "Engineered Safety Feature Actuation System Instrumentation," as follows:

Function 1.b., Safety Injection - Automatic Actuation Logic and Actuation Relays: Delete SR 3.3.2.13 as a required surveillance.

Function 3.a.(2), Containment Isolation - Phase A Isolation - Automatic Actuation Logic and Actuation Relays: Delete SR 3.3.2.13 as a required surveillance.

Function 5.a., Turbine Trip and Feedwater Isolation - Automatic Actuation Logic and Actuation Relays (SSPS): Delete SR 3.3.2.14 as a required surveillance.

Function 7.a., Automatic Switchover to Containment Sump - Automatic Actuation Logic and Actuation Relays: Revise SR 3.3.2.13 to SR 3.3.2.6.

The licensee will delete SRs 3.3.2.13 and 3.3.2.14 from Table 3.3.2-1 and replace them with SR 3.3.2.6. SR 3.3.2.13 or SR 3.3.2.14 has been deleted from these functions, and SR 3.3.2.6 has been added to the last function because all the functions listed above, except for the last function, already include SR 3.3.2.6. Because this change is consistent with topical report WCAP-13878-P-A, the NRC staff concludes the proposed conforming changes are acceptable.

-7 Based on the above, the NRC staff concludes that the proposed changes to the TS are in accordance with the NRC-approved topical report WCAP-13878-P-A and, therefore, are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Kansas State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 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 the Federal Register on October 7,2008 (73 FR 58679). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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.

Principal Contributor: H. Garg Date: Jul y 30, 2009

ML091700296 *SE input memo OFFICE NRR/LPL4/PM NRRlLPL4/PM NRRlLPL4/LA DE/EICB/BC DIRSIITSB/BC OGC NRRlLPL4/BC NRRlLPLR/PM NAME NDiFrancesco BSingal JBurkhardt WKemper* RElliott LBSubin MMarkley BSingal DATE 6/29/09 7/1109 6/29/09 5/8109 7/8/09 7/21/09 7/30109 7/30109