ML050350255

Technical Specifications, Amendment 157 Elimination of Requirements for Hydrogen Recombiners and Hydrogen Monitors

ML050350255
Person / Time
Site:	Wolf Creek
Issue date:	01/31/2005
From:	Donohew J NRC/NRR/DLPM/LPD4
To:	Muench R Wolf Creek
Donohew J N, NRR/DLPM,415-1307
Shared Package
ML050380495	List: ML050350255 ML050380482
References
TAC MC3862
Download: ML050350255 (9)

Text

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 ............................... 3.0-3 3.1 REACTIVITY CONTROL SYSTEMS .............................................. 3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) .............................................. 3.1-1 3.1.2 Core Reactivity .............................................. 3.1-2 3.1.3 Moderator Temperature Coefficient (MTC) ...................................... 3.1-4 3.1.4 Rod Group Alignment Limits .............................................. 3.1-7 3.1.5 Shutdown Bank Insertion Limits ....................................... ....... 3.1-11 3.1.6 Control Bank Insertion Limits .............................................. 3.1-13 3.1.7 Rod Position Indication .............................................. 3.1-16 3.1.8 PHYSICS TESTS Exceptions - MODE 2 ........................................ 3.1-19 3.2 POWER DISTRIBUTION LIMITS .............................................. 3.2-1 3.2.1 Heat Flux Hot Channel Factor (Fe(Z))

(F0 Methodology) .......... 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FXH) ............................ 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 ............................. 3.3-37 3.3.4 Remote Shutdown System ................................. 3.3-41 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation............................................................................ 3.3-44 Wolf Creek - Unit I i Amendment No. 123

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6 Containment Purge Isolation Instrumentation ................ .................. 3.3-46 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation Instrumentation .. 3.3-50 3.3.8 Emergency Exhaust System (EES) Actuation Instrumentation........................................................................... 3.3-55 3.4 REACTOR COOLANT SYSTEM (RCS) ........................... 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ................................................... 3.4-1 3.4.2 RCS Minimum Temperature for Criticality ........................................ 3.4-5 3.4.3 RCS Pressure and Temperature (P[T) Limits ................. ................. 3.4-6 3.4.4 RCS Loops - MODES 1 and 2.................................................... 3.4-8 3.4.5 RCS Loops - MODE 3................. ................... 3.4-9 3.4.6 RCS Loops - MODE 4................................................... 3.4-12 3.4.7 RCS Loops - MODE 5, Loops Filled ................................................. 3.4-14 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 3.5 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 ..................................... 3.6-15 3.6.5 Containment Air Temperature .............. ....................... 3.6-16 3.6.6 Containment Spray and Cooling Systems ..................................... 3.6-17 3.6.7 Spray Additive System ..................................... 3.6-20 Wolf Creek - Unit 1 I! Amendment No. 123, 131, 157

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 I with two or more required channel to OPERABLE channels inoperable. status.

I D. Required Action and D.1 Enter the Condition Immediately I associated Completion referenced in Table 3.3.3-1 Time of Condition C not for the channel. I 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 I 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 /> I F. As required by Required F.1 Initiate action in Immediately Action D.1 and referenced in Table 3.3.3-1.

accordance with Specification 5.6.8.

I Wolf Creek - Unit I 3.3-38 Amendment No. 423,157

PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS

--- ----------- \I - ------------ __ _ ___

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 ___NOTE-_

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. 18 months Wolf Creek - Unit 1 3.3-39 Amendment No. 123

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 I

1. Neutron Flux 2 E I

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

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

4. RCS Pressure (Wide Range) 2 E I

5. Reactor Vessel Water Level 2 F I

6. Containment Normal Sump Water Level 2 E I

7. Containment Pressure (Normal Range) 2 E I

B. Steam Line Pressure 2 per E steam generator I

9. Containment Radiation Level (High Range) 2 F I

.,0. Not Used I

11. Pressurizer Water Level 2 E I

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

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

I Core Exit Temperature - Quadrant 1 E

14. 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 I

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

Wolf Creek - Unit I 3.3-40 Amendment No. Qg, 157

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 S e ieNOTEio-y o ae dP t Separate Condition entry isallowed for each Function and required ASP control. I 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 3.3-41 Amendment No. 421,155

Spray Additive System 3.6.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.7.2 Verify spray additive tank solution volume is > 4340 184 days gal and

• 4540 gal.

SR 3.6.7.3 Verify spray additive tank solution concentration is 184 days

Ž 28% and 31 % by weight.

SR 3.6.7.4 Verify each spray additive automatic valve in the flow 18 months path that is not locked, sealed, or otherwise secured In position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.7.5 Verify spray additive flow rate from each solution's 5 years flow path.

Wolf Creek- Unit I 3.6-21 Amendment No. 123

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

9. WCAP 10965-P-A, uANC: A Westinghouse Advanced Nodal Computer Code."

10. WCAP-1 261 0-P-A, "VANTAGE+ Fuel Assembly Reference Core Report."

11. WCAP-8745-P-A, 'Design Bases for the Thermal Power AT and Thermal Overtemperature AT Trip Functions."

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

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

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.4.3, 'RCS Pressure and Temperature (PIT) Limits,"

and

2. Specification 3.4.12, 'Low Temperature Overpressure Protection System."

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

(continued)

Wolf Creek - Unit 1 5.0-29 Amendment No. 423113,- 2, 144

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

1. NRC letter dated December 2, 1999, "Wolf Creek Generating Station, Acceptance for Referencing of Pressure Temperature Limits Report (TAC No. MA4572)," and

2. WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January, 1996.

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

5.6.7 Not Used.

5.6.8 PAM Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.9 Not Used.

5.6.10 Steam Generator Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a report within 12 months following completion of the inspection. This Special Report shall include:

1) Number and extent of tubes inspected, (continued)

Wolf Creek - Unit 1 5.0-30 Amendment No. 423,130,21, 157