ML20207J962
| ML20207J962 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/18/1998 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20207J946 | List: |
| References | |
| PROC-981218, NUDOCS 9903160446 | |
| Download: ML20207J962 (100) | |
Text
i 5
D TECHNICAL REQUIREMENTS MANUAL REVISION:
1 w
2 TECHNICAL REQUIREMENTS MANUAL 9
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WOLF CREEK GENERATING STATION UNIT 1
DOCKET NO.
STN 50-482
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Susunary Statsmant:
The contents of the Technical Requiresnents Manual (TRM) are requirements and surveillances which do not meet the four criteria for inclusion in the Technical Specifications as specified in 10CFR50. 36 (c) (2) (ii).
These requirements are not of immediate concern for public health and safety, however, they shall ba implemented as prescribed in the TRM.
BOOTE: Any revisions to this document require a USQD.
DC30 12/18/98 O
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WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL O
Cs m List of Effective Pages O
Revision 1 l
Page Revision Table / Cover Sheet 1
.D List of Effective Pages 1
D Table of Contents
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93 16.0-1 1
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16.0-vii 1
IA 16.0 General Requirements 16.0-1 1
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9 16.6-3 0
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16.7 (3/4.7) Plant Systems 16.7-1 1
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16.9 (3/4.9) Refueling Operations 16.9-1 0
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16.10 (3/4.10) Special Test Exceptions I
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WOLF CREEK 3
TECHNICAL REQUIREMENTS MANUAL G-AN TABLE OF CONTENTS fL'f O 16.0 GENERAL REQUIREMENTS.............................................
16.0-iv 16.0.0 CROSS REFERENCE...............................................
16.0-iv s
16.0.1 PURPOSE.......................................................
16.0-1 3
16.0.2 DEFINITIONS...................................................
16.0-1 16.0.3 GENERAL OPERATIONAL REQUIREMENTS..............................
16.0-4 v.
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16.0.4 GENERAL SURVEILLANCE REQUIREMENTS.............................
16.0-4 16.1 (3/4.1) REACTIVITY CONTROL SYSTEMS.............................. 16.1-L M
16.1.1 INTENTIONALLY B?.ANK...........................................
16.1'-1 i
16.1.2 (3/4.1.2) BORATION SYSTEMS....................................
16.1-1 16.1.3(3/4.1.3) MOVEABLE CONTROL ASSEMBLIES.......................... 16.1-10 0
- 16. 2 INTENTIONALLY RLANK............................................. 16. 2-1 16.3(3/4.3) INSTRUbENTATION.......................................... 16. 3 -1 16.3.1(3/4.3.3) MONITORING INSTRUMENTATION........................... 16. 3-1 16.3.2(3/4.3.4) TURBINE OVERSPEED PROTECTION.........................
16.3-16 16.3.3(3/4.3.1) REACTOR TRIP SYSTEM INSTRUMENTATION (Rs.SPONSE TIMES). 16.3-18 16.3.4(3/4.3.2) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (RESPONSE TIMES).....................
16.3-20 16.4(3/4.4) REACTOR COOLANT SYSTEM...................................
I'6.4-1 16 4.1(3/4.4.2) SAFETY VALVES.
..................................... 16.4-1 16.4.2(3/4.4.7)
CHEMISTRY............................................
16.4*2 16.4.3(3/4.4.9) PRESSURE / TEMPERATURE LIMITS..........................
16.4-6 16.4.4(3/4.4.10) STRUCTURAL INTEGRITY................................
16.4-7 16.4.5(3/4.4.11) REACTOR COOLANT SYSTEM VENTS........................
16.4-9 16.5(3/4.5) EMERGENCY CORE COOLING....................................
16.5-1 16.5.1 m 4.5.1)
ACCUMULATORS.........................................
16.5-1 16.6(3/4.6) PRIM'.RY CONTAINNENT......................................
16.6-1 16.6.1(3/4.6.1) CONTAINMENT LEAKAGE..................................
16.6-1 16.7(3/4.7) PLANT FYSTEMS............................................
16.7-1 16.7.1(3/4.7.2) STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...... 16.7-1
}
16.7.2(3/4.7.8)
SNUBBERS.............................................
16.7-2 j
16.7.3(3/4.7.9) SEALED SOURCE CONTAMINATION..........................
16.7-12 16.7.4(3/4.7.12) AREA TEMPERATURE MONITORING.........................
16.7-14 16.8(3/4.8) ELECTRICAL PONER SYSTEMS.................................
16.8-1 16.8.1(3/4.8.4) ELECTRICAL EQUIPMENT PROTECTIVE DEVICES..............
16.8-1 16.0.2(3/4.8.4.8) SURVEILLANCE REQUIREMENTS (DIESEL GENERATOR).......
16.8-4 16.9(3/4.9) REFUELING OPERATIONS.....................................
16.9-1 16.9.1(3/4.9.5)
COMMUNICATIONS.......................................
16.9-1 16.9.213/4.9.6) REFUELING MACHINE....................................
16.9-2 16.9.3(3/4.9.7) CRANE TRAVEL - SPENT FUEL STORAGE FACILITY........... 16. 9-4 16.9.4 (3/4.9.10) WATER LEVEL - REACTOR VESSEL....................... 16. 9-5 16.10(3/4.10) SPECIAL TEST EXCEPTIONS................................
16.10-1 16.10.1(3/4.10.5)' POSITION INDICATION SYSTEM - SHUTDOWN..............
16.10-1 16.11(3/4.11) RADIOhCTIVE EFFLUENTS..................................
16.11-1 16.11.1 LIQUID HOLDUP TANKS.........................
................ 16.11-1 1
16.11.2 EXPLOSIVE GAS MIXTURE........................................
16.11-2 16.11.3 GAS STORAGE TANKS............................................
16.11-3
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T-1 WOLF CREEK TECHNICAL REQUIREMENTS MANUAL O*T TABLE OF CONTENTS (Continued)
O LIST OF FIGURES Number Title Page TIGURE 16.7-1(4.7-1)
SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST 16.7-11
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88 TABLE OF CONTENTS (Continued)
O LIST OF TABLES 3
Number Title Page D
TABLE 16.3-1 SEISiIC MONITOILING INSTIGENTATION.............. 16.3-4 3
TASIE 1G.3-2 SEISMIC MONITORING INSTBttBENTATION SURVEILLANCE N]
REQUIREbSDITS....................................
16.3-5 TABLE 16.3-3 MTE00lOI4GICAL ENITORING INST 3pSMTATIOtt....... 16. 3-7
~4 TABLE 16.3-4 MTEOROIAGICAL BEttITORING INST 3tLSSNTATION SURVEILIABICE REQUIREMNTS....................... 16. 3-8 I
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TABLE 16.3-5 ACCIDENT N00tITORING INST 3ttSBNTATIOtt............. 16. 3-10 TABLE 16.3-6 ACCIDENT N00tITORING INST 3ttBENTATIOtt SURVEILLANCE IIEQUIREBSalTS....................... 16. 3-11 TABLE 16.3-7 EEPLOSIVE GAS NattII.MLING INSTRUDENTATION........ 16.3-14 TABIE 16.3-5 EXPIASIVE GAS MONITORING INSTRISSNTATIOtt SURVEILLANCE ItEQUIRDS3ITS....................... 16. 3-15 TABIE 16.3-9 REACTOR TRIP SYSTEM INSTRIBSNTATIOst ItESPOttSE TIES.................................. 16. 3-19 TABI2 16.3-10 EneGINEERED SANETY FEATU3tES ItESPOttSE TIMES....... 16.3-21 TARIE 16.4-3 ItP WTOR COOLANT SYSTEM CHEMISTRY LIMITS......... 16. 4 ~4 TABLE 16.4-4 REf CTOR COOIANT SYSTEM CHEMISTRY SURVEILLANCE 3tEWIREMENTS.................................... 16. 4 - 5 TABIA 16.7-1(4.4-2) ENUBEER VISUAL INSPECTIOtt INTERVAL..........,... 16.7-9 TABLE 16.7 2 AREA TEMPERATUltE M0tt!TORING..................... 16.7-15 iii Rev. O i
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TECHNICAL REQUIREMENTS MANUAL C
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16.0.0 CROSS REFERENCE D
The following provides a cross-reference list between the old technical 1
g specification LCO and Surveillance numbers and the new Technical l
Requirements Manual section.
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OLD TECHNICAL SPECIFICATION NUMBER Technical Requirements Manual l 9
BORATION SYSTEMS
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Flow Path - Shutdown I
3.1.2.1...........................................
16.1.2.1 4.1.2.1...........................................
16.1.2.1.1 j
B3/4.1/2..........................................
16.1.2.1.2 Flow Path - Operating 3.1.2.2...........................................
~6.1.2.2 4.1.2.2...........................................
16.1.2.2.1 B3/4.1/?..........................................
16.1.2.2.2 Charging Pump - Shutdcrn l
3.1.2.3...........................................
16.1.2.3 4.1.2.3.1.........................................
16.1.2.3.1 B3/4.1/2..........................................
16.1.2.3.2 Charging Pump - Operating 3.1.2.4...........................................
16.1.2.4 4.1.2.4..........................................
16.1.2.4.1 B3/4.1/2..........................................
16.1.2.4.2 Borated Water Source - Shutdown 3.1.2.5...........................................
16.1.2.5 2
4.1.2.5...........................................
16.1.2.5.1 f'~N B3/4.1/2..........................................
16.1.2.5.2
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Borated Water Sources - Operating N~ /
3.1.2.6...........................................
16.1.2.6 j
4.1.2.6...........................................
16.1.2.6.1 B3/4.1/2..........................................
16.1.2.6.2 MOVEABLE CONTROL ASSEMBLIES Position Indication System - Shutdown 3.1.3.3...........................................
16.1.3.1 4.1.3.3...........................................
16.1.3.1.1 j
B3/4.1.3..........................................
16.1.3.1.2 Rod Drop Time j
3.1.3.4...........................................
16.1.3.2 4.1.3.4...........................................
16.1.3.2.1 B3/4.1.3..........................................
16.1.3.2.2 NONITORING INSTRUMENTATION Movable Incore Detectors 3.3.3.2...........................
16.3.1.1 4.3.3.2...........................................
16.3.1.1.1 B3/4.3.3.2........................................
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OLD TECHNICAL SPECIFICATION NUMBER Technical Requirements Manual l 0
Seismic Instrumentation -(NOTE - This section was relocated in an earlier amendment).
.D JF 3.3.3.3...........................................
16.3.1.2 g
4.3.3.3.1.........................................
16.3.1.2.1.a 4.3.3.3.2.........................................
16.3.1.2.1.b
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B3/4.3.3..........................................
16.3.1.2.2 Table 3.3-7.......................................
Table 16.3-1 M
Table 4.3-4.......................................
Table 16.3-2
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Meteorological Instramentation 3.3.3.4...........................................
16.3.1.3 y~
4.3.3.4...........................................
16.3.1.3.1 B3/4.3.4..........................................
16.3.* 3.2 Table 3.3-8.......................................
Tabit 16.3-3 Table 4.3-5.......................................
Table 16.3-4 Accident Monitoring Instrumentation-3.3.3.6...........................................
16.3.1.4 4.3.3.6...........................................
16.3.1.4.1 B3/4.3.6..........................................
16.3.1.4.2 Table 3.3-10......................................
Table 16.3-5 Table 4.3-7.......................................
Table 16.3-6 Loose-Part Detection System 3.3.3.9...........................................
16.3.1.5 4.3.3.9...........................................
16.3.1.5.1 B3/4.3.9..........................................
16.3.1.5.2 Explosive Gas Monitoring Instrumentation 3.3.3.11..........................................
16.3.1.6 4.3.3.11..........................................
16.3.1.6.1 (No Bases Section)................................
16.3.1.6.2 Table 3.3-13......................................
Table 16.3-7 Table 4.3-9.......................................
Table 16.3-8 TURBINE OVERSPEED PROTECTION 3.3.4.............................................
16.3.2.1 4.3.4.1...........................................
16.3.2.1.la 4.3.4.2...........................................
16.3.2.1.ib B3/4.3.4..........................................
16.3.2.1.2 REACTOR TRIP SYSTEM 3.3.1..............................................
16.3.3.1 4.3.1.2............................................
16.3.3.2 3.3.2..............................................
16.3.4.1 4.3.2.2............................................
36.3.4.1.1 Table 3.3-2........................................
Table 16.3-9 B 3/4.3.2..........................................
16.3.4.1.2 ENGINEERED SAFETY FEATURES Tably 3.3-5......................
Table 16.3-10 SAFETY VALVES 3.4.2.1.............
16.4.1.1 4.4.2.1...........................................
16.4.1.1.1 B3/4.4............................................
16.4.1.1.2 v
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/~'N OLD TECHNICAL SPECIFICATION NUMBER Technical Requirements Manual l
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CHEMISTRY 3.4.7.............................................
16.4.2.1 4
4.4.7.............................................
16.4.2.1.1 2
B3/4.4.7..........................................
16.4.2.1.2 y,
Table 3.4-2.......................................
Table 16.4-3 Table 4.4-3.......................................
Table 16.4-4
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PRESSURIZER V
3.4.9.2...........................................
16.4.3.1
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4.4.9.2...........................................
16.4.3.1.1 B3/4.4.9..........................................
16.4.3.1.2 sc STRUCTURAL INTEGRITY 3.4.10............................................
16.4.4.1 4.4.10............................................
16.4.4.1.1 B3/4'.4.10.........................................
16.4.4.1.2 I
REACTOR COOLANT SYSTEM VENTS 3.4.11............................................
16.4.5.1 4.4.11............................................
16.4.5.1.1 B3/4.4.11.........................................
16.4.5.1.2 CONTAINMENT LEAKAGE 3.6.1.2...........................................
16.6.1.1 4.6.1.2...........................................
16.6.1.1.1 B3/4.6.1.2........................................
16.6.1.1.2 CONTAINMENT VESSEL STRUCTURAL INTEGRITY 3.6.1.6...........................................
16.6.1.2 4.6.1.6.1.........................................
16.6.1.2.1.e 4.6.1.6.2.........................................
16.6.1.2.1.?
4.6.1.6.3..........................................
16.6.1.2.1.c B3/4.6.1.6........................................
16.6.1.2.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 3.7.2.............................................
16.7.1.1 4.7.2...........................................
16.7.1.1.1 B3/4.7.2..........................................
16.7.1.1.2 SNUBBERS 3.7.8.............................................
16.7.2.1 4.7.8............................................
16.7.2.1.1 B3/4.7.8..........................................
16.7.2.1.2 Table 4.7-2.......................................
Table 16.7-1 Figure 4.7-1......................................
Figure 16.7-1 SEALED SOURCE CONTAMINATION 3.7.9..................
16.7.3.1 4.7.9.1...........................................
16.7.3.1.1.a 4.7.9.2...........................................
16.7.3.1.1.b 4.7.9.3...........................................
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WOLF CREEK TECHNICAL REQUIREMENTS MANUAL
~ h OLD TECHNICAL SPECIFICATION NUMBER Techr.ical Requirements Manual l 0
' AREA TEMPERATURE MONITORING z:
3.7.12............................................
16.7.4.1
- J.
4.7.12............................................
16.7.4.1.1
.A B3/4.7.12.........................................
16.7.4.1.2
~ Table 3.7-4.......................................
Table 16.7-2 3
k ELECTRICAL EQUIPMENT PROTECTIVE DEVICES
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3.8.4.1...........................................
16.8.1.1 Y
4.8.4.1...........................................
16.8.1.1.1 B3/4.8.4..........................................
16.8.1.1.2 ColeE7NICATIONS
~0 3.9.5.............................................
16.9.1.1 4.9.5.............................................
16.9.1.1.1 83/4.9.5..........................................
16.9.1.1.2 REFUELING MACRINE 3.9.6.............................................
16.9.2.1 4.9.6.1...........................................
16.9.2.1.1.a 4.9.6.2...........................................
16.9.2.1.1.b B3/4.9.6..............................
16.9.2.1.2 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY 3.9.7.............................................
16.9.3.1 4.9.7.............................................
16.9.3.1.1 B3/4.9.7..........................................
16.9.3.1.2 WATER LEVEL - REACTOR VESSEL - Control Rods N
3.9.10.2...........................................
16.9.4.1 4.9.10.2'................
16.9.4.1.1 B5/4.9.10.........................................
16.9.4.1.2 POSITION INDICATION SYSTEM - SHUTDOWN 3.10.5............................................
16.10.1.1 4.10.5............................................
16.10.1.1.1 B3/4.10.5.........................................
16.10.1.1.2 LIQUID HOLDUP TANKS 3.11.1.4..........................................
16.11.1.1 4.11.1.4..........................................
16.11.1.1.1 B3/4.11.1.4.......................................
16.11.1.1.2 EXPLOSIVE GAS MIXTURE 3.11.2.5..........................................
16.11.,2.1 4.11.2.5..........................................
16.11.2.1.1 B3.4.11.2.5....................................,...
16.11.2.1.2 GAS STORAGE TANKS 3.11.2.6..........................................
16.11.3.1 4.11.2.6..........................................
16.11.3.1.1 B3/4.11.2.6.......................................
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TECHNICAL REQUIREMENTS MANUAL p;n hU 15.0.1 PURPOSE The purpose of this manual is to specify requirements which have been removed 2
from the Wolf Creek Technical Specifications as a result of the implementation 3
of Wolf Creek Technical Specifications Amendment 89, which approved proposed g
changes to the scope and contents of the Wolf Creek Technical Specifications in acccrdance with the NRC's Final Policy Statement on Technical Specifications j
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Improvements for Nuclear Power Rear tcra, 58 FR 39102, dated July 22, 1993.
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16.0.2 DEFINITIONS h
ACTION 1
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16.0.2.1 ACTION shall be that part of an Operational Requirement which prescribes remedial measures required under designated conditions.
ACTUATION LOGIC TEST 16.0.2.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.
ANALOG CHANNEL OPERATIONAL TEST 16.0.2.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify. OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.
CHANNEL CALIBRATION 16.0.2.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and securacy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 16.0.2.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
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' CORE ALTERATION L C" l
16.0.2.6 CORE ALTERATION shall be the movement of any fuel, sources, or
.1 reactivity control components within the reactor vessel with the vessel head l
D removed and fuel in the. vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe position.
I OPERABLE - OPERABILITY
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16.0.2.7 A system, subsystem, train, component or device shall be OPERABLE or Q
have OPERABILITY when it is capable of performing its specified function (s),
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and when all necessary attendant instrumentation, controls, electrical power, l
cooling or seal water, lubrication or other auxiliary equipment that are l
required for the system, subsystem, train, component, or device to perform its M
function (s) are also capable of performing their related support function (s).
OPERATIONAL MODE - MODE 16.0.2.8 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Technical Specification Table 1.2.
RATED THERMAL POWER 16.0.2.9 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3565 MWt.
SITE BOUNDARY 16.0.2.10 The SITE BOUNDARY shall be that line beyond which the land is A
neither owned, nor leased, nor otherwise controlled by the licensee.
SLAVE RELAY TEST e
16.0.2.11 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.
SOURCE CHECK I
16.0.2.12 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity, 1
STAGGERED TEST BASIS 16.0.2.13 A STAGGERED TEST BASIS shall consist of a.
A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.
The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
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TECHNICAL REQUIREMENTS MANUAL G
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16.0.2.14 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST 2
16.0.2.15 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating
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the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include 4
adjustment, as necessary, of the Trip Actuating Device such that it actuates at g
the required Setpoint within the required accuracy.
~
WASTE GAS HOLDUP SYSTEM OPERATION i?
o 16.0.2.16 The Waste Gas Holdup System is in operation when there is any gas input to the system or when a recombiner is in operation. This includes
" degassing operation" which is defined as the operation of switching from a H2 Blanket to a N2 Blanket (or vice versa) in the VCT while in Modes 3, 4, and 5.
If the Waste Gas Holdup System has no inputs and the recombiners are not running, then the system is not in operation. The operation of.the Waste Gas Holdup System on recirc only, then the system is not considered to be in operation. If the Waste Gas Holdup System is not in operation, per this definition, then the Waste Gas Analyzers do not have to be in operation.
l O
16.0-3 Rev. O i
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F rI WOLF CREEK I>
TECHNICAL REQUIREMENTS MANUAL C
7 16.0.3 GENERAL OPERATIONAL REQUIREMENTS
_ p) D
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16.0.3.1 The limiting conditions for operation (LCO) contained in the following operational requirements shall be met for the modes and operational 3
conditions specified under the applicability section except as noted under section 16.0.3.2, 3
M 16.0.3.2 Upon making the determination that the LCO cannot be met, the
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guidance contain1d in the ACTION section of the operational requirement shall be implemented. Completion of the prescribed actions is not required if 9
compliance with the LCO is restored within the time limits allowed in the g
ACTIONS section.
~
16.0.3.3 When an LCO is not met, except as provided in the associated ACTION M
guidance, action shall be implemented in a timely ranner to place the unit in a safe condition as determined by plant management. Where corrective measures are taken that allow operation under the guidance contained in the ACTION section, the action may be implemented within the specified time limits as determined by plant management.
16.0.3.4 Entry into a higher Operational Mode or condition covered by these operational requirements shall not be made if the LCO and associated ACTION cannot be met.
Plant management may allow ontry into an Operational Mode or condition while relying on the guidance of the ACTION section to meet the LCO.
.16.0.4 GENERAL SURVEILLANCE REQUIREMENTS 16.0.4.1 Surveillance Requirements shall be met during the Operational Modes or other conditions specified for the individual LCO's unless stated otherwise
{
in the individual surveillance requirement.
(m) 16.0.4.2 Each Surveillance requirement shall be performed within the s'
specified surveillance interval plus 25% unless otherwise approved by plant management.
16.0.4.3 Failure to perform a Surveillance Requirement within the allcwed time inter ~tl defined in 16.0.4.2, and without prior approval by plant j
management, constitutes noncompliance with the LCO.
However, application of the ACTION guidance may be delayed up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit completion of the surveillance requirements. Surveillance Requirements do not have to be performed on inoperable equipment.
16.0.4.4 Entry into an Operational Mode or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the LCO have been performed with in the time limits allowed by 16.0.4.2.
This shall not prevent passage through or to Operational Modes as required to comply with ACTION guidance.
16.0.4.5 Surveillance Requirements for inservice inspection and tes+ing of ASME Code Class 1, 2, and 3 components shall be applicable at follows:
a.
Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and 16.0-4 Rev. 0 (g
v
-i WOLF CREEK D,
TECHNICAL REQUIREMENTS MANUAL 0
lm')m Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, U
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Section 50.55a(g).
l SurveillanceintebvalsspecifiedinSectionXIoftheASMEBoilerand b.
a Pressure Vessel Code and applicable Addenda for the Inservice inspection and testing activit'*s required by the ASME Boiler and Pressure Code and a}
applicable Addena.
.all be applicable as follow in these requirements:
\\
Weekly At least once per 7 days Monthly At least once por 31 days 9
Quarterly or every three months At least once per 92 days
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Semiannually or every six months At least once per 184 days Every nine months At least once per 276 days Yearly or annually At least once per 366 days
,g c.
The provisions of requirement 16.0.4.2 are applicable to the above required frequencies for performing inservice inspection and testing activities; d.
Performance of the above inservice inspection and testing activities anall be in a addition to other specified surveillance requirements; and NothingintheASMEBoilerandPrc[sureVesselCodeshallbeconstruedto e.
supersede the requirements of any TRM Requirement.
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16.0-5 Rev. 1 l
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I H
WOLF CREEK E
TECHNICAL REQUIREMENT 3 MANUAL D
ON 16.1 (3/4.1) kEACTIVITT CONTm0L SYSTEbs JO.
16.1.1-INTENTIOMh!.LY BIANK
.D 16.1.2 (3/4.1.2) BORATION SYSTEMS
.5 FLOW PATH - SHUTDOWN g
LIMITING CONDITION FOR OPERATION
'd
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16.1.2.1 (3.1.2.2)
As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency g
power source:
a.
A flow path from the Boric Acid Storage System via a boric acid transfer pump and a centrifugal charging pump to the Reactor Coclant System if the Boric Acid Storage System is OPERABLE as given in Section 16.1.2.5a. for MODES 5 and 6 or as given in Section 16.1.2.6a for MODE 4; or' b.
The flow path from the refueling water storage tank via a centrifugal charging pump
- to the Reactor Coolant System if the r<
refueling water storage tank is OPERABLE as given in Section
./
16.1.2.5b for MODES 5 and 6 or as given in Section 16.1.2.6b for MODE 4.
APPLICABILITY: MODES 4, 5, and 6.
ACTION:
O()
With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involvin'g CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 16.1.2.1.1 (4.1.2.1)
At least one of the above required flow paths shall be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, saaled, or otherwise secured in position, is in its correct position.
BASES 16.1.2.1.2 The Boration Systems ensures that negative reactivity control is available during each mode of facility operation. The components required to l
perform this function includes (1) borated water sources, (2) centrifugal l
charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.
In Mode 6 with the reactor vessel head removed, an OPERABLE SI pump can be used to substitute for an inoperable centrifugal charging pump.
16.1-1 Rev. O i
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1 WOLF CRECK D
TECHNICAL REQUIREMENTS MANUAL G
sm
[
]U With the RCS aversge temperature equal to or greater than 350*F a minimum of
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two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to 3
provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after 3
xenon decay and cooldown to 200*F.
The maximum expected boration capability y;
requirement occurs at EOL from full power equilibrium xenon conditions and requires 17,658 gallons of 7000 ppm borated water from the boric acid storage s
i tanks or 83,754 gallons of 2400 ppm borated water from the RWST. With the RCS average temperature less than 350', only one boron injection flow path is f
required.
s With the RCS temperature below 200*F, one Boration System is acceptable without single failure consideration on the basis of the stable reactivity 3
condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.
The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR suction relief valve.
The boron capability required below 200'* is sufficient to provide a SHUTDOWN MARGIN of 1.3% Ak/k af ter xenon decay and cooldown from 200'F to 140' F.
This condition requires either 2968 gallons of 7000 ppm borated water from the boric acid storage tanks or 14,071 gallons of 2400 ppm borated water from the RWST.
,n
(
The contained water volume li.aits include allowance for water not A
available because of discharge line location and other physical characteristics. In the case of the boric acid tanks, all of the contained volume is considered usable. The required usable volume may be contained in either or both of the boric acid tanks.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
The OPERABILITY of cae Boration System during REFUELING ensures that this system is available for reactivity control while in MODE 6.
If the required centrifugal charging pump is not operable, an SI pump can be used for boration during Mode 6 with the reactor vessel head removed. The operation of SI pump in this instance is justified based on the similarity of the SI pump and CCP in d
providing the safety functions to the plant as well as the capacity of the two pumps when taking suction from the refueling water storage tank.
When determining compliance with action statement requirements, addition to the RCS of borated water with a concentration greater than or equal to the minimum required RWST concentration shall not be considered to be a positive reactivity change.
16.1-2 Rev. O j
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WOLF CREEK I
TECHNICAL REQUIREMENTS MANUAL D-
-REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING A
LIMITING CONDITION FOR OPERATION
.m 16.1.2.2 (3.1.2.2)
At least two of the following three boron injection low
\\
paths shall be OPERABLE:
a.
The flow path from the Boric Acid Storage System via a boric acid g
transfer pump and a centrifugal charging pump to the Reactor Coolant System.
U b.
A flow path from the refueling water storage tank via BNLCV112D ar centrifugal charging pumps to the Reactor Coolant System.
I c.
A flow path from the refueling water storage tank via BNLCV112E and charging pumps to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, and 3*.
ACTION:
With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow Paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, nk,f SURVEILLANCE REQUIREMENTS 16.1.2.2.1 (4.1.2.2)
At least two of the ebove required flow paths shall be demonstrated OPERABLE:
a.
At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; b.
At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and c.
At least once per 18 months by verifying that the flow path required by Section 16.1.2.2a delivers at least 30 gpm to the Reactor Coolant System.
1 i
i 16.1-3 Rev. O
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4 WOLF CREEK I
TECHNICAL REQUIREMENTS MANUAL D
s m Cs/
U BASES 16.1.2.2.2 See Section 16.1.2.1.2 h
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- The provisions of General Operational Requirement 16.0.3.4 and General Surveillance Requirement 16.0.4.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Technical Specification 4.5.3.2 provided that the centrifugal charging pump is restored L
to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of J
(
the RCS cold logs exceeding 375'F, whichever comes first.
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16.1-4 Rev. O w
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3 WOLF CREEK TECHNICAL REQUIREMENTS MANUAL D
REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN E
LIMITING CONDITION EUR OPERATION
..a._--_--
16.1.2.3 (3.1.2.3)
One centrifugal charging pump
- in the boron injection flow
\\
path required by Section 16.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.
O g
APPLICABILITY: MODES 4, 5, and 6
~
ACTION:
With no centrifugal c brging pump OPERABLE or capable of being powered from an OPERABLE emergen y power source, suspend all operations involving CORE I
q ALTERATIONS or positive reactivity changes, j
SURVEILLANCE REQUIREMENTS 16.1.2.3.1 (4.1.2.3.1)
The above required centrifugal charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that the pump develops a differential pressure of greater than or equal to 2400 psid when tested pursuant to General Surveillance Requirement 16.0.4.5.
BASES 16.1.2.3.2 See Section 16.1.2.1.2 In Mode 6 with the reactor vessel head removed, an OPERABLE SI pump can be used to substitute for an inoperable centrifugal charging pump in the flow path from the Refueling Water Storage Tank per Section 16.1.2.1.b.
16.1-5 Rev. 1 l
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WOLF CREEK 3
TECHNICAL REQUIREMENTS MANUAL g
{
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REACTIVITY CONTROL SYSTEMS O
CHARGING PUMPS - OPERATING
.u 3
LIMITING CONDITION FOR OPERATION 16.1.2.'4 (3.1.2.4)
At least two centrifugal charging pumps shall be OPERABLE.
7*
APPLICABILITY: MODES 1, 2, and 3.*
O
\\
ACTION:
With only one centrifugal charging pump OPERABLE, restore at least two g
centrifugal charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least f
HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200*F within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s: restore at least two charging pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
4*
.ti SURVEILLANCE REQUIREMENTS 16.1.2.4.1 (4.1.2.4)
At least two centrifugal charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that the pump develops a differential pressure of greater than cr ecual to 2400 psid when tested pursuant to General Surveillance Requirement 16.0.4.5.
BASES 16.1.2.4.2 See Section 16.1.2.1.2
- The provisions of General Operational Requirement 16.0.3.4 and General Surveillance Requirement 16.0.4.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Technical Specification 4.5.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.
16.1-6 Rev. O
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3 WOLF CREEK TECHNICAL REQUIREMENTS MANUAL 0
h REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN A
LIMITING CONDITION FOR OPERATION D
N 16.1.2.5 (3.1.2.5)
As a minimum, one of the following borated water sources
\\
shall be OPERABLE:
9 a.
A Boric Acid Storage System with:
\\
1)
A minimum contained borated water volume of 2968 gallons, 2)
Between 7000 and 7700 ppm of boron, and
- 3) A minimum solution temperature of 65'F.
b.
The refueling water storage tenk (RHST) with:
1)
A minimum contained borated water volume of 55,416 gallons, 2)
A minimum boron concentration of 2400 ppm, and 3)
A minimum solution temperature of 37'F.
APPLICABILITY: MODES 5 and 6.
ACTION:
N With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 16.1.2.5.1 (4.1.2.5) The above required borated water source shall be demonstrated OPERABLE:
a.
At least once per 7 days by:
- 1) Verifying the boron concentration of the water,
- 2) Verifying the contained borated water volume, and
- 3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 37'F.
BASES 16.1.2.5.2 See Section 16.1.2.1.2 16.1-7 Rev. O O
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WOLF CREEK g
TECHNICAL REQUIREMENTS MANUAL
,: m
/7 REACTIVlTY CONTROL SYSTEMS t
U BORATED WATER SOURCES - OPERATING J
LIMITING CONDITION FOR OPERATION 2
16.1.2.6 (3.1.2.6)
As a minimum, the following borated water sources shall be
\\
OPERABLE as required by TRM Section 16.1.2.2 for MODES 1, 2, and 3 and one of the following borated water sources shall be OPERABLE as required by TRM M
Section 16.1.2.1 for MODE 4:
\\
a.
A Boric Acid Storage System with:
1)
A minimum contained borated water volume of 17,658 gallons, 2)
Between 7000 and 7700 ppm of boron, and 3)
A minimum solution temperature of 65'F.
b.
The refueling water storage tank (RWST) with:
1)
A minimum contained borated water volume of 394,000 gallons 2)
Between 2400 and 2500 ppm of boron, 3)
A minimum solution temperature of 37'r, and 4)
A maximum solution temperature of 100*F.
r~%
APPLICABILITY: MODES 1, 2, 3, and 4.
D ACTION:
a.
With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources in MODE 1, 2 or 3, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak /k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With the RWST inoperable in MODE 1, 2, or 3, due to the boron concentration not being within the specified limits, restore the tank to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c.
With the RWST inoperable in MODE 1, 2, or 3, for reasons other than the boron concentration not being within the specified limits, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
16.1-8 Rev. 1 l
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1
N WOLF CREEK TECHNICAL REQUIREMENTS FANUAL D
' N d.
With no borated water source OPERABLE in MODE 4, restore one borated 7}
water source to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
0 SURVEILLANCE REQUIREMENTS 16.1.2.6.1 (4.1.2.6)
Each required borated water source shall be demonstrated
\\
OPERABLE:
h a.
At least once per 7 days by:
1)
Verifying the boron concentration in the water, t
M 2)
Verifying the contained borated water volume of the water source, and 3)
Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.
b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when the outside air temperature is either less than 37'r or greater than 100*F.
BASES 16.1.2.6.2 See Section 16.1.2.1.2 e
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16.1-9 Rev. O
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wotr CREEK 1)
TECHNICAL REQUIREMENTS MANUAL
'D OM 16.1.3 (3/4.1.3) MNEABLE CONTROL ASSDSLIES U
POSITION INDICATION SYSTEM-SHUTDOWN LIMITING CONDITION FOR OPERATION A
16.1.3.1 (3.1.3.3)
One digital rod position indicator (excluding demand N
position indication) shall be OPERABLE and capable of determining the control
\\
rod position within i 12 steps for each shutdown or control rod not fully inserted.
1 M
1 APPLICABILITY: MODES 3*l, 4*t, and 5*#.
g t
ACTION:
- fs With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.
SURVEILLANCE REQUIREMENTS 16.1.3.1.1 (4.1.3.3)
Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicator agrees with the demand position indicator within 12 steps when exercised over the full range of rod travel at least once per 18 months.
BASES 16.1.3.1.2 See Technical Specification Bases 3/4.1.3.
f-s
'With the Reactor Trip System breakers in the closed position.
ISee Special Test Exception in Section 16.10.1.
16.1-10 Rev. 0 O
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3 WOLF CREEK I
TECHNICAL RDQUIREMENTS MANUAL D
O'N REACTIVITY CONTROL SYSTEMS
')
ROD DROP TIME 3
LIMITING CONDITION FOR OPERATION D
X 16.1.3.2 (3.1.3.4)
The individual full-length shutdown and control rod drop
\\
time from the physical fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to p,)
dashpot entry with:
Tavg greater than or equal to 551*F, and a.
t$
b.
All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2.
l ACTIONt With the rod drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
SURVEILLANCE REQUIREMENTS 16.1.3.2.1 (4.1.3.4)
See Technical Specification 4.1.3.1.3 BASES N
16.1.3.2.2 See Technical Specification Bases 3/4.1.3 s
16.1-11 Rev. O O
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TECHNICAL REQUIREMENTS MANUAL p.
'M 16.2 INTENTIONALLY BLANK 3
'D 2
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WOLF CREEK E
TECHNICAL REQUIREMENTS MANUAL D
g 16.3 ' (3/4.3) INSTRIBSNTATION 16.3.1 (3/4.3.3) MONITORING INSTRUMENTATION MOVABLE INCORE DETECTORS D
f LIMITING CONDITION FOR OPERATION 16.3.1.1 (3.3.3.2)
The Movable Incore Dete: tion System shall be OPERABLE with:
h)
'a.
At least 75% of the detector thimbles, T
b.
A minimum of two detector thimbles per core quadrant, and L0 c
Sufficient movable detectors, drive, and readout equipment to map d
these thimbles.
APPLICABILITY: When tha Movable Incore Detection System is used for:
a.
Recalibration of the Excore Neutron Flux Detection System, b.
Monitoring the QUADRANT POWER TILT RATIO, or c.
Measurement of Eg(X,Y,Z) and FAH(X,Y)
ACTION:
a.
With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions.
f
b.
The provisions of General Operational Requirements 16.0.3.3 and
(
16.0.3.4 are not applicable.
_ SURVEILLANCE REQUIREMENTS 16.3.1.1.1 (4.3.3.2)
The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by normalizing each detector output when requirsJ for:
a.
. Recalibration of the Excore Neutron Flux Detection System, or b.
Monitoring the QUADRANT POWER TILT RATIO, or c.
Measurement of Fg(X,Y,Z) and FAH(X'Y) j BASES 16.3.1.1.2 The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from 16.3-1 Rev. 0
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H 3
WOLF CREEK TECHNICAL REQUIREMENTS MANUAL O #*
use of this system accurately represent the spatial neutron flux distribution U
of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.
M M
E For the purpose of measuring rg (X,Y,Z) or FAH (XeY) a full incore flux 3
map is used. Quarter-core flux mapt. as defined in WCAP-8648, June 1976, may IO be used in recalibration of the Excore Neutron Flux Detection System, and full
\\
incore flux maps or synenetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Neutron Flux channel is
,9 inoperable.
\\
so o
4 i
i
.s l
1
)
l l
1 I
16.3-2 Rev. O O
7 _.
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3 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL "a
M INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING OONDITION FOR OPERATION M
16.3.1.2 (3.3.3.3)
The seismic monitoring instrumentation shown in Table 16.3-
\\
1 shall be OPERABLE.
h APPLICABILITY: At all times.
ACTION:
N a.
With one or more of the above required seismic monitoring instruments inoperable, restore the instrument (s) to OPERABLE status within 30 days.
b.
The provisions of General Operational Requirements 16.0.3.3 and 16.0.3.4 are not applicable.
SURVEILLANCE REQUIREMENTS
- 16. 3.1. 2.1. a ( 4. 3. 3. 3.1)
Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CA1IBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 16.3-2.
16.3.1.2.1.b (4.3.3.3.2)
Each of the above required seismic monitoring instruments actuated during a seismic event c reater than or equal to 0.02 g O
shall be restored to OPERABLE status within ?= hours and a CRANNEL CALIBRATION performed within 10 days following the seisto event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion.
BASES 16.3.1.2.2 The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.
This capability is reqaired to permit comparison of the measured respe+.se to that used in the design basis for the facility to determine if plant snutdown is required pursuant to Appendix A of 10 CFR Part 100.
The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974, i
16,3-3 Rev. O i
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WO; IEK E
TECHNICAL RE00adEMENTS MANUAL C1
/' F3.
(
?
Table 16.3-1
' SEISMIC MONITORING INSTRUMENTATION L
3 MINIMUM MEASUREMENT INSTRUMENTS 3
INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
\\
^
1.
Triaxial Peak Recording M
Accelerographs a.
Radwaste Base Slab i 5.0g 1
b.
Control Room i 5.0g 1
c.
ESW Pump Facility i 5.0g 1
d.
Ctat Structure i 5.0g 1
e.
Auxiliary Bldg. SI Pump i 5.0g 1
Suetions f.
SGB Piping i 5.0g 1
g.
SGC Support i 5.0g 1
2.
Triaxial Time History and Response Spectrum Recording rystem, Monitoring the Following Accelerometers (Active) a.
Ctat. Base Slab i 1.0g I
b.
Ctat. Oper. Floor i 1.0g 1
T c.
Reactor Support i 1.0g 1
d.
Aux. Bldg. Base Slab i 1.0g 1
e.
Aux. Bldg. Control Room i 1.0g 1
Air Filter f.
Free Field i 0.5g 1
3.
Triaxial Response-Spectrum Recorder (Passive)
Ctat. Base Slab i 1.09 1
4.
Triaxial Seismic ACCELERATION Switches LEVEL North East Vertical a.
OBE Free Field 0.06g 0.06g 0.06g 1
b.
SSE Free Field 0.15g 0.15g 0.16g 1
c.
Seismic Trigger 0.02g 0.02g 0.02g 1
l l
[
16.3-4 Rev. 1 l
O
H WOLF CREEK
[
TECHNICAL REQUIREMENTS MANUAL Li
(\\ _,/ )
Table 16.3-2
. SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS D
ANALOG y
CHANNEL N
CHANNEL CHANNEL OPERATIONAL
\\
INSTRUMENTS AND SENSOR LOCATIONS CHFCK CALIBRATION TEST 1.
Triaxial Peak Recording g
Accelerographs
~
a.
Radwaste Base Slab N.A.
R N.A.
M b.
Control Room N.A.
R N.A.
c.
ESW Pump Facility N.A.
R N.A.
d.
Ctat Structure N.A.
R N.A.
e.
Auxiliary Bldg. SI Pump N.A.
R N.A.
Suction f.
SGB Piping N.A.
R N.A.
g.
SGC Support N.A.
R N.A.
2.
Tr' t 'il Time History and Response Spectrum Recording System, Monitoring the Following Accelerometers (Active) a.
Ctat. Base Slab M
R SA b.
Ctat. Oper. Floor M
R SA n
c.
Reactor Support M
R SA**
d.
Aux. Bldg. Base Slab M
R SA**
e.
Aux. Bldg. Control Room M
R SA**
Filters f.
Free Field M
R SA**
3.
Triaxial Response-Spectrum Recorder (Passive)
Ctmt. Base Slab N.A.
R N.A.*
4.
Triaxial Seismic Switches a.
OBE Free Field M
R SA b.
SSE Free Field M
R SA c.
System Trigger M
R SA
- Checking at the Main Control Board Annunciation for contact closure output in the Control Room shall be performed at least once per 184 days.
- The Bi-stable Trip Setpoint need not be determined during the performance of an ANALOG CHANNEL OPERATIONAL TEST.
16.3-5 Rev. O i
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WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL 0
OM INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION g
.3 33 16.3.1.3 (3.3.3.4)
The meteorological monitoring instrumentation channels in
\\
Table 16.3-3 shall be OPERABLE.
h APPLICABILITY: At all times.
\\
ACTION:
M a.
With one or more ".equired meteorological monitoring channels inoperable restore the instrument (s) to OPERABLE status within 7 days.
b.
The provisions of General Operational Requirements 16.0.3.3 and 16.0.3.4 are not applicable.
SURVEILLANCE REQUIREMENTS 16.3.1.3.1 (4. 3. 3. 4 ) Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CRANNEL CHECK and CRANNEL CALIBRATION at the frequencies.shown in Table 16.3-4.
BASES
i 16.3.1.3.2 The OPERABILITY of the meteorological instrumentation ensures that
/
sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.
)
1 16.3-6 Rev. 0
)
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WOLF CREEK
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TECHNICAL REQUIREMENTS MANUAL D
m U
TABLE 16.3-3 METEOROLOGICAL MONITORIN3 INSTRUMENTATION MINIMUM O
INSTRUMENT LOCATION OPERABLE l0 h
1.
Wind Speed Nominal Elev. 10m 1
y Nominal Elev. 60m 1
2.
Wind Direction Nominal Elev. 10m 1
'O Nominal Elev. 60m 1
3.
Air Temperature - A Nominal Elev. 10m-60m 1
16.3-7 Rev. 0
r :-
H WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL D
m O
TABLE 16.3-4 METEOROLOGICAL MONITORING INSTRUMENTATION 3
SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK
_ CAL.IBRATION 1.
Wind Speed 9
g a.
Nominal Elev. 10m D
SA b.
Nominal Elev. 60m D
SA 2.
Wind Direction a.
Nominal Elev. 10m D
SA b.
Nominal Elev. 60m D
SA 3.
Air Temperature -
AT a.
Nominal Elev. 10-D SA 6mm O
l 4
16.3-8 Rev. O e
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WOLF CREEK 3,
TECHNICAL REQUIREMENTS MANUAL
,[
INSTRUMENTATION
\\
'3 ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION M
16.3.1.4 (3.3.3.6)
The accident monitoring instrumentation channels shown in
\\
Table 16.3-5 shall be OPERABLE.
h APPLICABILITY: MODES 1, 2, and 3.
ACTION:
M a.
With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 16.3-5, restore the inoperable channel (s) to OPERABLE status within 30 days.
b.
With the number of OPERABLE accident monitoring instrumentation channels, except the unit vent-high range noble gas monitor, less than the Minimum Channels OPERABLE requirements of Table 16.3-5, restore one channel to OPERABLE status within 7 days; otherwise, be in at least HCT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With the number of OPERABLE channels for the unit vent-high range
-noble gas monitor less than the Minimum Channels OPERABLE requirements of Table 16.3-5, initiate an alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and restore the inoperable channel to OPERABLE status within 7 days.
(O
- d.
The provisions of General Operational Requirement 16.0.3.4 are not d
applicabic SURVEILLANCE REQUIREMENTS 16.3.1.4.1 (4.3.3.6)
Each accident monitoring instrumentation channel shall be demonstreted OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 16.3-6.
BASES 16.3.1.4.2 See Technical Specification Bases 3/4.3.3.6.
16.3-9 Rev. 0
4 WOLF CREEK I
TECHNICAL REQUIREMENTS MANUAL D
TABLE 16.3-5 ACCIDENT MONITORING INSTRUMENTATION
.D TOTAL MINIMUM
^
NO. OF CHANNELS INSTRUMENT CHANNELS OPERABLE
\\
1.
Containment Pressure - Extended 2
1
- 3 Range
\\
2.
PORV Position Indicator
- 1/ Valve 1/ Valve 3.
PORV Block Valve Position 1/ Valve 1/ Valve Indicator **
4.
Safety Valve Position Indicator 1/ Valve 1/ Valve
.5.
Unit Vent - High Range Noble Gas N.A.
1 Monitor
- Not applicable if the associated block valve is in the closed position.
- Not applicable if the block valve is verified in the closed position and l
power is removed.
16.3-10 Rev. 1 l
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NOLF CREEK I
TECHNICAL REQUIREMENTS MANUAL
~r 9
TABLE 16.3-6 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
.3 CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
'\\
1.
Containment Pressure - Extended M
R Range
.9 2.
PORV Position Indicator
- M N.A.
3.
PORV Block Valve Position M
N.A.
M
-Indicator **
4.
Safety Valve Position Indicator M
N.A 5.
Unit Vent - High Range Noble Gas M
R Monitor
- Not applicable if the associated block valve is in the closed position.
- Not applicable if the block valve is verified in the closed positon and power is removed.
9 16.3-11 Rev. O O
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'4 WOLT CREEK D
TECHNICAL REQUIREMENTS MANUAL O
A 'T INSTRUMENTATION t
JC LOOSE-PART DETECTION SYSTEM LIMITING CONDITION EUR OPERATION D
D) 16.3.1.5 (3.3.3.9)
The Loose-Part Detection System shall be OPERABLE.
\\
APPLICABILITY: MODES 1 and 2.
Y ACTION:
~~
a.
With all Loose-Part Detection System channels in one or more T
collection regions inoperable restore the inoperable region (s) to OPERABLE status within 30 days.
b.
The provisions of General Operational Requirement 16.0.3.4 is not applicable.
SURVEILLANCE REQUIREMENTS 16.3.1.5.1 (4.3.3.9)
Each channel of the Loose-Prrt Detection System shall be demonstrated OPERABLE by performance of:
a.
A CRANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
An ANALOG CRANNEL OPERATIONAL TEST except for verification of Setpoint at least once per 31 days, and c.
A CHANNEL CALIBRATION at least once per 18 months.
G BASES
,m 16.3.1.5.2 The OPERABILITY of the loose-part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components. The allowable out-of-service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May,
- 1981, 16.3-12 Rev. 1 l
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e Ff 3
WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL U
, f*
lA INSTRUMENTATION
\\.
EXPLOSIVE GAS MONITORING INSTRUMENTATION J
3 LIMITING CONDITION FOR OPERATION 1
M 16.3.1.6 (3.3.3.11)
The explosive gas monitoring instrumentation channels
\\
shown in Table 16.3-7 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Section 16.11.2 are not exceeded.
9 APPLICABILITY: As shown in Table 16.3-7.
ACTION:
~
D) a.
With an explosive gas monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, declare the channel inoperable and take the ACTION shown in Table 16.3-7.
b.
With less than the minimum number of explosive gas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 16.3-7 Restore the inoperable instrumentation to OPERABLE status within 30 days, and, if unsuccessful, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.
c.
The provisions of General Operational Requirements 16.0.3.3 and 16.0.3.4 are not applicable.
O, SURVEILLANCE REQUIREMENTS 16.3.1.6.1 (4.3.3.11)
Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 16.3-8.
BASES 16.3.1.6.2 Intentienally Blank 16.3-13 Rev. O O
m
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WOLF CREEK E
TECHNICAL REQUIREMENTS MANUAL 3
TABLE 16.3-7 EXPLOSIVE GAS MONITORING INSTRUMENTATION IA MINIMUM 3
INSTRUMENT CHANNELS APPLICABILITY ACTION N-OPERABLE
.\\
1.
WASTE GAS HOLDUP SYSTEM
(
~
A)
Explosive Gas Monitoring-
{
\\
System a.
Hydrogen Monitor 1/Recombiner 1
b.
Oxygen Monitor 2/Recombiner 2
- During WASTE GAS HOLDUP SYSTEM operation.
ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner.
ACTION 2 - With the Outlet Oxygen Monitor channel inoperable, operation of the system may continue provided grab samples are taken and analyzed at
/)'s least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both oxygen channels or both the inlet oxygen and inlet hydrogen channels inoperable, suspend oxygen supply to the
(,
recombiner.. Addition of waste gas.to the system may continue provided grab samples are taken and analyzed at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations.
16.3-14 Rev. 0 a.
e W.?
4 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL G
_,/ O TABLE 16.3-8 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
..D CHANNEL MODSS FOR WHICH y
CHANNEL CHANNEL OPERATIONAL SURVEILLANCE U
INSTRUMENT CHECK CALIBRATION TEST IS REQUIRED
\\
1.
(Not Used)
M
\\
2 WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System a.
Inlet Hydrogen Monitor D
Q(1)
H 1
b.
Outlet Hydrogen Monitor D
Q(1)
M c.
Inlet Oxygen Monitor D
O(2)
M de Outlet Oxygen Monitor D
Q(3)
M TABLE NOTATIONS
- During WASTE GAS HOLDUP SYSTEM operation.
(1)
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
a.
One volume percent hydrogen, balance nitrogen and b.
Four volume percent hydrogen, balance nitrogen.
(2)
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
a.
One volume percent oxygen, balance nitrogen, and b.
Four volume percent oxygen, balance nitrogen.
(3)
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
a.
10 ppm by volume oxygen, balance nitrogen, and b.
80 ppm by volume oxygen, balance nitrogen.
16.3-15 Rev. O
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WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL O
P 16.3.2 (3/4.3.4) TURBINE OVERSPEED PROTECTION
(,,,
O LIMITING CONDITION FOR OPERATION 16.3.2.1 (3.3.4)
At least one Turbine Overspeed Protection System shall be OPERABLE.
APPLICABILITY: HODES 1, 2,* and 3*.
\\
ACTION:
M a.
With one stop valve or one governor valve per high pressure turbine g
steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within
, I' 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or close at least one valve in the affected steam lines or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With the above reqaired Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> isolate the turbine from the steam supply.
SURVEILLANCE REQUIREMENTS
- 16. 3. 2.1. la (4. 3. 4.1)
The provisions of General Surveillance Requirement 16.0.4.4 are not applicable.
16.3.2.1.lb (4.3.4.2)
The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:
a.
At least once per 92 days by dire ct cbservation of the movement of each of the following valves through at least one complete cycle from e-g the running position:
8 1)
Four high pressure turbine stop valves,
\\s /
2)
Four high pressure main turbine governor valves 3)
Six low pressure turbine reheat stop valves, and 4)
Six low pressure turbine reheat intercept valves.
b.
At least once per 18 months by performance of a CRANNEL CALIBRATION on the Turbine Overspeed Protection Systems; and c.
At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or corrosion.
16.3-16 Rev. 0 1
4 I
1 i
1 i '
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WOLF CREEK l
y TECHNICAL REQUIREMENTS MANUAL
.e O' 'E
/
BASES
(
U 16.3.2.1.2 This operational requirement is provided to ensure that tJe turbine
~
overspeed protection instrumentation and the turbine speed control val >es are
.D OPERABLE and will protect the turbine from excessive overspeed. Although the 4
orientation of the turbine is such that the number of potentially damaging missiles which could impact and damage safety-related components, equiptnen* or g
structures is minimal, protection from excessive turbine overspeed is requ2 red.
~
h*-
"\\
.0
- Not applicable in MODE 2 or 3 with all main steam line isolation valves and associated bypass valves in the closed position and all other steam flow paths to the turbine isolated.
1 O
V i
16.3-17 Rev. O I
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WOLF CREEK D
TECHNICAL REQUIREMENTS FANUAL D
INSTRUMENTATION 16.3.3 (3/4.3.1) REAC105t TRI? SYSTEM INSTRUMENTATION l
i LIMITING CONDITION FOR OPERATION
.D DJ
\\
16.3.3.1 (3.3.1)
[j See Technical Specification 3.3.1 SURVEILLANCE REQUIREMENTS 16.3.3.2 (4.3.1.2)
See Technical Specification 4.3.1.1 and 4.3.1.2.
The i
REACTOR TRIP SYSTEM RESPONSE TIML limits are given in Table 16.3-9.
BASES 16.3.3.3 See Technical Specification Bases 3/4.3.1 and 3/4.3.2.
l O
i i
I 1
16.3-18 Rev. 1 l
O
E$
3 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL D
{
f As' O TABLE 16.3-9 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES 711NCTIONAL UNIT RESPONSE TIME 1.
Manual Reactor Trip N.A.
- 2..
Power Range, Neutron Flux
$ 0.5 second*
3.
Power Range, Neutron Flux, b,
High Positive Rate N.A.
_"\\
4.
Power Range, Neutr(A Flux, High Negative Rate 5,0.5 second*
5.
Intermediate Range, Neutron Flux N.A.
6.
Source Range, Neutron Flux N.A.
"1.
Overtemperature AT
$ 6.0 seconde
- 8.
Overpower AT f,6.0 seconds
- 9.
Pressuriser Pressure-Low 1 2.0 seconds 10.
Pressurizer Pressure-High
< 2.0 seconds 11.
Pressurizer water Level-High N.A.
l 12.
Reactor Coolant Flow-Low a.
Single Loop (Above P-8)
'< 1.0 second
\\
b.
Two 1. oops (Above P-7 and below P-9) 2 1.0 second l
13.
Steam Generator water Level-Low-Low 1 2.0 seconds l
14.
Undervoltage-Reactor Coolant Pumps 1 1.5 seconds 15.
Underfrequency-Reactor Coolant Pumps 5 0.6 second 16.
Turbine Trip a.
Low Fluid Oil Pressure N.A.
b.
Turbine Stop Valve Closure N.A.
17 Safety Injection Input fro:a ESF N.A.
18.
Reactor Trip Systen Interlocks N.A.
19.
Reactor Trip Breakers N.A.
l 20.
Automatic Trip and Interlock Logic N.A.
- Neutron detectors are seempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
1 16.3-19 Rev. 1 l
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~
F N
1 WOLF CREEK D
TECHNICAL REQUIREMENTS. MANUAL 0
O?;
INSTRUMENTATION QU 16.3.4 (3/4.3.2) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION l
h j
LIMITING CONDITION FOR OPERATION
{4 16.3.4.1 (3.3.2)
See Technical Specification 3.3.2.
[
SURVEILLANCE REQUIREMENTS h
16.3.4.1.1 (4.3.2.2)
See Technical Specification 4.3.2.1 and 4.3.2.2.
The ENGINEERED SAFETY FEATURES RESPONSE TIME limits are given in Table 16.3-10.
r BASES
,f,.
....m I
./
16.3.4.1.2 See Technical Specification Bases 3/4.3.1 and 3/4.3.2.
ESF response times specified in Table 16.3-10 which include sequential operation of RWST and VCT valves (Notes 3 and 4) are based on values assumed in the non LOCA safety analyses. These analyses take credit for injection of borated water from the RWST.
Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening j
of the RWST charging pump suction valves. When the sequential operation of the RWST and VCT valves is not included in the response times (Note 7) the values specified are based on the LOCA analyses. The LOCA analyses take credit for injection flow regardless of the source. Verification of the response times specified in Table 16.3-10 will assure that the assumptions used for the LOCA and non LOCA analyses with respect to operation of the VCT and RWST valves are valid.
16.3-20 Rev. 1 l
.g
7-H 3
WOLF' CREEK 2)
TECHNICAL REQUIREMENTS MANUAL
^
'N O TABLE 16.3-10 (Sheet 1 of 3)
- g U
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS U
- 1. Manual'Initistion 0
a.
Safety Injection (ECCS)
N.A.
- g b.
Containment Spray N.A.
c.
Phase *A* Isolation N.A.
h d.
Phase "B" Isolation N.A.
y-e.
Containment Purge Isolation N.A.
g f.
Steam Line Isolation N.A.
g.
Feedwater Isolation N.A.
\\
h.
Auxiliary IDedwater N.A.
I.
Essential Service Water N.A.
g j Containment Cooling N.A.
s.
k.
Control Room Isolation N.A.
1.
Reactor Trip N.A.
m.
Emergency Diesel Generators N.A.
n.
Component Coolang Water N.A.
o.
Turbine Trip N.A.
- 2. Containment Pressure-Hioh-1 I7I/27(43 a.
Safety Injection (ECCal 3 29 la Resctor Trip
<2
- 2) Feedwater Isolation 57
- 3) Phase "A" Isolation
< 1.5(5)
- 4) Auxiliary Feedwater 560 llI
- 5) Essential Service Water
$ 60
- 6) Containment cooling
$ 60(13
- 7) Component Cooling Water N.A.
- 8) Emergency Diesel Generators 3 14(6)
- 9) Turbine Trip N.A.
- 3. Pressurizer Pressure-Low a.
Safety Injection (ECCS)
$ 29III/21I43
- 1) Peactor Trip
$2
- 2) Feedwater Isolation
<7
- 3) Phase "A" Isolation 2(5)
~~ 60
- 4) Auxiliary Teodwater Ill
- 5) Essential Service Water
$ 6D
- 6) Containment Cooling
< 60lII
- 1) Component Cooling Water U.A.
- 8) Emergency Diesel Generators 5 14(6)
- 9) Turbine Trip N.A.
- 4. Steam Line Pressure-Low
- a. Safety Injection (ECCS)
$ 39I33/27(43 il Reactor Trip
<2
- 2) Feedwater Isolation 27
- 3) Phase "A" Isolation 2(5)
- 4) Auxiliary Teodwater 5 60
- 5) Essential Service Water
< 60ll) 61 containment Cooling
[60Ill
- 7) Component Cooling Water N.A.
- 8) Emergency Diese1' Generators 5 14(6)
- 9) Turbine Trip W.A.
b.
Steam Line Isolation
$ 2(5) 16.3-21 Rev. 0
/%
r1 r4 2
woLr caEcx D
TECHNICAL REQUL'EMENTS MANUAL 0
"'g(fl TABLE 16.3-10 (Sheet 2 of 31 d
Vg ENGINEERED SAFETY PEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 5. Contairunent Pressure-Hich-3 a.
$ 32III/2()(2)
N b.
Phase "B" Isolation g
-< 31.5 t
l 3
- 6. Containment Pressure-High-2 Steam Line Isolation i 2(5)
- 7. Steam Line Pressure-Necative Rate-High g
steam Line Isolation f 2(5)
- 8. Steam Generator Water Level-Hiah-High a.
< 2.5 b.
Feedwater Isolation 77 I
- 9. Steam Generator Water Level - Low-Low a.
Start Motor-Driven Auxiliary Feedwater Pumps 5 60 b.
Start Turbine-Driven Auxiliary Teodwater Pumps 5 60
- 10. Loss-of-Offsite Power Start Turbine-Driven Auxiliary Feedwater N.A.
g Pumps
- 11. Trip of All Main Peedwater Pumps Start Motor-Driven N.A Auxiliary Feedwater Pumps 12.
Auxiliary r edwater Pump Suetion e
Pressure-Low Transfer to Essential Service Water N.A.
l 13.
RWST Level-Low-Low Coincident with Safety Injection Automatic Switchover to Containment i 60 l
- 14. Loss of Power a.
4 kV Sus Undervoltage-1 14 Loss of Voltage b.
4 kV Bus Undervoltage-5 144 Grid Degraded voltage i
- 15. Phase "A' Isolation a.
Control Room Isolation N.A.
b.
Containment Purge Isolation 5 2(5) 16.3-22 Rev. 1 l
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TECHNICAL REQUIREMENTS MANUAL D
% P1 TABLE 16.3-10 (Sheet 3 of 31 D
\\
TABLE NOTATIONS p
(1) Diesel generator starting and sequence loading delays included.
,S (2) Diesel generator starting delay g included. Offsite power h) available.
\\
(3) Diesel generator starting and sequence loading delay included. RHR
~7 pumps M included. Sequential transfer of charging pump suction y
from the VCT to the LWST (RNST valves open, then VCT valves close) is included.
(4) Diesel generator starting and sequence loading delays not included.
M Offsite power available. RHR pumps M included. Sequential-transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is included.
(5) Does not include valve closure time.
(6) Includes time for diesel to reach full speed.
(7) Diesel generator starting and aequence loading delays included.
sequential transfer of charging pump suction from the VCT to the RNST (RNST valves open, then VCT valves close) is ng included.
Response time assumes only opening of RWST valves.
~+
s 16.3-23 Rev. O D
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WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL 0
' OL 16.4 (3/4.4) REACTOR COCUdfT SYSTEM V
'16.4.1 (3/4.4.2) SANETY VALVES SHUTDOWN D
j LIMITING CONDITION FOR OPERATION M
i
\\
16.4.1.1 (3.4.2.1)
A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig + 1%.1
- J
\\
APPLICABILITY: MODES 4 and 5.2 3
ACTION:
With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR 1
loep into operation in the shutdown cooling mode.
I SURVEILLANCE REQUIREMENTS 16.4.1.1.1(4.4.2.1)
No additional requirements other than those required by General Surveillance Requirement 16.0.4.5.
BASES 16.4.1.1.2 The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam. The relief O
capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.
An acceptable method to meet the requirement of Specification 3.4.9.3 for the RCS depressurized through a 2 square inch or larger vent, is to remove a pressurizer Code safety valve. Removing a safety valve provides an RCS vent path much larger than the two square inches required to provide RCS overpressure protection during cold shutdown. Thus, when in Mode 5, it would be acceptable to have all pressurizer Code safety valves removed simultaneously for testing and maintenance, provided these valves are all restored to OPERABLE status prior to entering Mode 4.
1 The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
2When the ROS is not vented to atmosphere by a vent opening 22 square inches.
16.4-1 Rev. 1 l
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WOLF CREEK 3
TECHNICAL REQUIREMENTS MANUAL D
[D-REACTOR COOLANT SYSTEM J
16.4.2 (3/4.4.7) CMENISTRY LIMITING CONDITION FOR OPERATION 1
3 16.4.2.1 (3.4.7)
The Reactor Coolant System chemistry shall be maintained M
within the limits specified in Table 16.4-3.
\\
APPLICABILITY: At all times.
ACTION:
\\
MODES 1, 2, 3, and 4:
M With any one or more chemistry parameter in excess of its Steady-a.
1 State Limit but within its Transient Limit, restore the parameter to
{
within its Steady-State Limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.
With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
At All Other Times:
With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or in excess of its Transient Limit, reduce the pressurizer pressure to f-s less than or equal to 500 psig, if applicable, and perform an engineering j
(
evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4.
SURVEILLANCE REQUIREMENTS 16.4.2.1.1 (4.4.7) The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 16.4-4.
BASES 16.4.2.1.2 The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the 16.4-2 Rev. O AU i
)
F H
3 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL l
O
~("'N Z oxygen, chloride, and fluoride limits are time and temperature dependent.
( )O Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant g
effect of the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the S
Transient Limits provides time for taking corrective actions to restore the N
contaminant concentrations to within the Steady-State Limits.
Y The Surveillance Requirements provide adequate assurance th'at concentrations in
,.,)
excess of the limits will be detected in sufficient time to take corrective action.
g TO k
}
16.4-3 Rev. 0 I
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1
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3 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL 0
m C'
TABLE 16.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS
- D
\\
STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT id
- g Dissolved Oxygen
- 5 0.10 ppm s 1.00 ppm Chloride s 0.15 ppm s 1.50 ppm Fluoride s 0.15 ppm 5 1.50 ppm
't.
- Limit not applicable with T less than or equal to 250*F.
avg 9
16.4-4 Rev. O u
f[
l H
3 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL C
I II T
O TABLE 16.4-4 REACTOR COOLANT SYSTEM S
CHEMISTRY SURVEILLANCE REQUIREMENTS h
.4 g
SAMPLE AND l
PARAMSTER ANALYSIS FREQUENCY C
g Dissolved Oxygen
- At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
~
Chloride At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> DC Fluoride At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> j
- Not required with.T less than or equal to 250*F.
ayg 16.4-5 Rev. 0 v
1
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WOLF CREEK V
TECHNICAL REQUIREMENTS MANUAL D
M 16.4.3 (3/4.4.9) PRESSURE / TEMPERATURE LIMITS O
PRESSURIZER LIMITING CONDITION FOR OPERATION D
E 16.4.3.1 (3.4.9 2) The pressurizer temperature shall be limited to:
3 a.
A maximum heatup of 100'F in any 1-hour period, s) b.
A maximum cooldown of 200*F in any 1-hour period, and
)
c.
A maximum spray water temperature differential of 583'F.
l M
APPLICABILITY: At all times.
\\
ACTION:
With the pressurizer temperature limits in excess of any of the above limits restore the temperature to within the limits within 30 minutes perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 16.4.3.1.1 (4.4.9.2)
The pressurizer temperatures shall be determined to be f-s\\
within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be (s) within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.
BASES 16.4.3.1.2 The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure vessel Code,Section III, Appendix G.
The pressurizer heatup and cool rates shall not exceed 100*E/h and 700*/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 583*F.
Although the pressurizer operates in temperature ranges above th'ose for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
Also see Technical Specification Bases 3/4.4.9.
16.4-6 Rev. 1 l
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TECHNICAL REQUIREMENTS MANUAL 3
p)
,b 16.4.4 (3/4.4.10) STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION D
M 16.4.4.1 (3.4.10)
The structural integrity of ASME Code Class 1, 2 and 3
\\
components shall be maintained in accordance with Section 16.4.4.1.1 (4.4.10).
APPLICABILITY: All MODES.
ACTION:
l With the structural integrity of any ASME Code Class 1 component (s) a.
not conforming to the above requirements, restore the structural integrity of the affected component (s) to within ).ts limit or isolate the affected component (s) prior to increasing the Reactor Coolant System te.aperature more than 50*F above the minimum temperature required by NDT considerations, b.
With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
With the structural integrity of any ASME Code Class 3 component (s) c.
not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or p
isolate the affected component (s) from service.
d.
The provisions of General Operational Requirement 16.0.3.4 are not applicable.
SURVEILLANCE REQUIREMENTS n it 16.4.4.1.1 (4.4.10)
In addition to the requirements'of General Surveillance Requirement 16.0.4.5, each reactor coolant pump flywheel mhall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.
(See Technical Specification 6.8.5).
In lieu of Position C.4.b(1) and C.4.b(2), conduct a qualified in-place DT examination over the volume from the inner bore of the flywheel to the circle of gne-half the cuter radius or conduct a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels once every ten years coinciding with the inservice inspection schedule as required by ASME Section XI.
- The volumetric examination and surface examination of the Reactor Coolant pump "D" motor flywheel for the First 10-year Inservice Inspection Interval may be delayed one cycle to coinciJe with the Fall 1997 refueling outage.
16.4-7 Rev. 0 pV
F 4
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TECHNICAL REQUIREMENTS HANUAL l
D j O f4 BASES i
)y
%.J 16.4.4.1.2 The inservice inspection and testing programs for ASME Code Class
~
1, 2, and 3 components ensure that the structural integrity and operational 3
readiness of these components will be maintained at an acceptable level J
throughout the life of the plant. These programs are in accordance with 2
Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda
\\
as required by 10 CFR Part 50.55a.
j g
Components of the Reactor Coolant System were designed to provide access to permit inservice inspection in accordance with Section XI of the ASME Boiler
\\
and Pressure Vessel Code, 1974' Edition and Addenda through Summer 1975.
M m
i i
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l i
1
(~'
(j 16.4-6 Rev. 0
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WOLF CREEK 3)
TECHNICAL REQUIREMENTS MANUAL D
~'N A 16.4.5 (3/4.4.11) REACTOR COOLANT SYSTEM VENTS
)
LIMITING CONDITION FOR OPERATION s
.~
16.4.5 (3.4.11) At least one reactor vessel head vent path consisting of at U
least two valves in series powered from emergency busses shall be OPERABLE and Y
closed.
M
\\
APPLICABILITY: MODES 1, 2,
3, and 4.
L q
ACTION:
s 4
With the above reactor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed I
with power removed from the valve actuator of all the valves in the ineperable vent paths restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 16.4.5.1.1 (4.4.11)
Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:
a.
Verifying all manual isolation valves in each vent path are locked in the open position, b.
Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or
-~s REFUELING, and
\\ss c.
Verifying flow through the reactor vessel head vent paths during venting during COLD SHUTDOWN or REFUELING.
BASES 16.4.5.1.2 Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel head vent path ensures the capability exists to perform this function.
The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure vent valve power supply or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System ve7ts are consistent with the requirements of Item II.B.1 of NUREG-0737,
" Clarification of TMI Action Plan Requirements," November 1980.
16.4-9 Rev. O b
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WOLF CREEK
.I TECHNICAL REQUIREMENTS MANUAL D
f\\M y/O 16.5 (3/4.5) DERGmeCY CORE COOLING 16.5.1. (3/4.5.1) AcctMUIATORs 16.5.1.1. (3.5.1) LIMITING CONDITIONS FOR OPERATION M
\\
16.5.1.1 (3.5.1)
Each Accumulator shall have a minimum of one pressure and one level instrument channel OPERABLE.
h?
\\
APPLICABILITY: Modes 1 and 2.
Mode 3 with RCS pressure >1000 psig.
M i
/
ACTION:
With cne or more accumulators with less than one pressure or one water level instrumes.t channel OPERABLE, restore at least one channel to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or declare the affected accumulator (s) inoperable.
SURVEILLANCE REQUIREMENTS 16.5.1.1.1 (4.5.1.2)
Each accumulator water level and pressure channel shall be demonstrated OPERABLE at least once per 18 months by the performance of a CHANNEL CALIBRATION.
O BASES 16.5.1.1.2 Technical Specification 4.5.1.2, which required the performance of e channel calibration of each accumulator water level and pressure channel once per 18 months, was relocated to the Updated Safety Analysis Report in Amendment 119.to the WCGS Technical Specifications, and later relocated to Revision 0 of the TRM along with the rest of USAR Chapter 16.
This was accomplished in accordance with the recommendations of Generic Letter 93-05 and NUREG-1366.
These recommendations were based on the recognition that accumulator instrumentation operability is not directly related to the capability of accumulators to perform their safety function. The LCO, Applicability and Action Statements were added to Pevision 1 to clarify that an accumulator need not be declared inoperable if one of the redundant instrumentation channe,t j
becomes inoperable and to specify the action to take when an accumulator has no l
OPERABLE instrumentation channels.
16.5-1 Rev. 1 l
7-,
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1 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL C~s M
16.6 (3/4.6) PRIMhRY CONTADGSNT
- k. J 16.6.1 (3/4.6.1) CONTAIleENT 22AKA3E LIMITING CONDITION FOR OPERATION l
L 2
16.6.1.1 (3.6.1.2) Containment leakage rates shall be limited in accordance
.4 with the Containment Leakage Rate Testing Program
\\
APPLICABILITY: MODES 1, 2, 3, 4, and 5 l
\\
ACTION:
l C0 a.
If Reactor Coolant System temperature is at or below 200*F, restore the overall integrated leakage rate to less than 0.75 L and the a
combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 La prior to increasing the Reactor Coolant System temperature above 200*F.
b.
If the Reactor Coolant System temperature is above 200 degrees F, perform the Action of Technical Specification 3.6.1.1.
l l
l O
16.6-1 Rev. 1 l
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C.
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WOLF CREEK 2)
TECHNIO.AL REQUIREMENTS MANUAL O
rmM SURVEILLANCE REQUIREMENTS ig Gi 16.6.1.1.1 (4.6.1.2)
The containment leakage rates shall be demonstrated in
~
accordance with the Containment Leakage Rate Testing Program as follows:
.D
.D A Type A test shall be performed at a frequency of at least l
a.
so once per 10 years based on acceptable performance history. Acceptable g
performance history is defined as completion of two consecutive periodic Type A tests (>24 months elapsc time) where the calculated g
performance leakage rate was less than 1.0 La.
An extencion of up to g
15 months may be allowed on a limited basis for scheduling purposes only.
[0 b.
If the As-found Type A test results are not acceptable (i.e., >1.0 L.),
then acceptable performance should be re-established by performing a Type A test within 48 months following the unsuccessful Type A test.
Following a successful As-found Type A test, the testing frequency may
{
be returned to once per 10 years i
Type B and C tests shall be conducted with gas at a pressure not less c.
than P, 48 psig, at intervals specified by the Containment Leakage a
Rate Testing Program except for tests involving:
i 1)
Air locks, 2)
Purge supply and exhaust isolation valves with resilient material
)
seals, and 3)
Valves pressurized with fluid from a seal system.
16.6-2 Rev. O fh 0
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WOLF CREEK E
TECHNICAL REQUIREMENTS MANUAL D
DY' 4
- C' CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
D S
d.
Air locks shall be tested and demonstrated OPERABLE by the a
requirements of Technical Specification 4.6.1.3;
\\
s.
Purge supply and exhaust isolation valves with resilient material 9
seals shall be tested and demonstrated OPERABLE by the requirements of Technical Specification 4.6.1.7.2 and 4.6.1.7.4, as applicable;
\\
M BASES 16.6.1.1.2 The limitations on containment leakage rates specified in the Containment Leakage Rate Testing Program ensure that the total contajnment leakage volume will not exceed the value assumed in the safety analysis at the peak accident pressure, P*
a The testing for measuring leakage rates as specified in the Containment Leakage Rate Testing Program are consistent with Regulatory Guide 1.163, September 1995, and the requirements of Option B of 10 CFR 50, Appendix J.
(
l 16.6-3 Rev. O o
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1 M
3L WOLF CREEK I
TECHNICAL REQUIREMENTS MANUAL
.D
'M CONTAINMENT SYSTEMS I
D CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION p3 16.6.1.2 (3.6.1.6)
The structural integrity of the containment vessel shall be
.\\
maintained at a level consistent with the acceptance criteria in Section
- 16. 6.1.2.1 (4. 6.1. 6).
APPLICABILITY: MODES 1, 2, 3, and 1.
\\
ACTION:
M a.
With the abnormal degradation indicated by the conditions in Section 16.6.1.2.1.a.4 (4. 6.1. 6. la. 4 ), restore the tendons to the required level of integrity or verify that containment integrity is maintained within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 15 days in accordance with Technical Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, b.
With the indicated abnormal degradation of the structural integrity other than ACTION a. at a level below the acceptance criteria of Section 16. 6.1. 2.1. a ( 4. 6.1. 6), restore the containment vessel to the required le-el of integrity or verify that containment integrity is maintained within 15 days and perform an engineering evaluation of I
the contair aent and provide a Special Report to the Commission within 30 c ays in accordance with Technical Specification 6.9.2 or
/, s) be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD
\\s,/
SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The provisions of General Operational Requirement 16.0.3.4 are not applicable.
SURVEILLANCE REQUIREMENTS
- 16. 6.1. 2.1. a (4. 6.1. 6.1). Containment Vessel Tendons. The structural integrity of the prestressing tendons of the containment vessel shall be demonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test and at 5-year intervals thereafter. The structural integrity of the tendons shall be demonstrated by:
(a) Determining that a random but representative sample of at least 11 tendons (4 inverted U and 7 hoop) each have an observed lift-off force within the predicted limits established for each tendon. For each subsequent inspection one tendon from each group (1 inverted U and I hoop) shall be kept unchanged to develop a history and to correlate the observed data. The procedure of inspection and the j
tendon acceptance cri+eria shall be as follows:
16.6-4 Rev. O v
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TECHNICAL, REQUIREMENTS MANUAL
- D O5 CONTAINMENT SYSTEMS
~~Y U SURVEILLANCE REQUIREMENTS (Continued) g
(,
4 1.
If the measured prestressing force of the selected tendon in a 4
group lies above the prescribed lower limit, the lift-off test to is considered to be a positive indication of the sample tendon's g
acceptability, 2.
If the measured prestressing force of the selected tendon in a group lies between the prescribed lower limit and 90% of the A.
prescribed lower limit, two adjacent (accessible) tendons, one 4
on each side of this tendon shall be checked for their prestressing y
forces. If the prestressing forces of these two tendons are above 95% of the prescribed lower limits for the tendons, all three tendons shall be restored to the required level of integrity, and the tendon group shall be considered as acceptable.
If the measured prestressing force of any two tendons falls below 95% of the prescribed lower limits of the tendons, additional lift-off testing shall be done to detect the cause and extent of such occurrence. The condition shall be considered as an indication of abnormal degradation of the containment structure, 3.
If the measured prestressing force of any tendon lies below 90%
of the prescribed lower limit, the defective tendon shall be completely detensioned and additional lift-off testing shall be done so as to determine the cause and extent of such 1
occurrence. The condition shall be considered as an indication j
of abnormal degradation of the contaxnment structure, (mV) 4.
If the average of all measured prestressing forces for each group (corrected for average condition) is found to be less than the minimum required prestress level at the anchorage locations for that group, the condition shall be considered as abnormal degradation of the containment structure, 5.
If from conse:utive surveillances the meast.ced prestressing i
forces for the same tendon or tendons in a proup indicate a i
trend of prestress loss larger than expected and the resulting j
prestressing forces will be less than the minimum required for the group before the next scheduled surveillance, additional lift-off testing shall be done so as to determine the cause and 1
extent of such occurrence. The condition shall be considered as an indication of abnormal degradation of the containment l
l structure, and l
6.
Unless there is abnormal degradation of the containment vessel during the first three inspections, the sample population for subsequent inspections shall include at least 6 tendons (3 hoop, 1
3 inverted U).
16.6-5 Rev. 0
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TECHNICAL REQUIREMENTS MANUAL
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CONTAINMENT SYSTEMS
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SURVEILLANCE REQUIREMENTS (Continued) 1 (b) Performing tendon detensioning, inspections, and material tests on I
h a previously stressed tendon from each group. A randomly selected i
E tendon from each group shall be completely detensioned in order to j
l0 identify broken or damaged wires and determining that over the
\\
entire length of the removed wire sample (which shall include the broken wire if so identified) that:
M 1.
The tendon wires are free of corrosion, cracks, and damage, and 2.
A minimum tensile strength of 240 kai (guaranteed ultimate I0 strength of the tendon material) exists for at least three wire samples (one from each end and one at mid-length) cut from each r
removed wire.
Failure to meet the requirements of Section 16.6.1.2.1.a (b)
(4. 6.1. 6.1.b) shall be considered as an indication of abnormal degradation of the containment structure.
(c)
Performing tendon retensioning of those tendons detensioned for i
inspection to at least the force level recorded prior to detensioning or the predicted value, whichevtar is greater, with the tolerance within minus zero to plus 6%, but not to exceed 70% of the guaranteed ultimate tensile strength of the tendons. During retensioning of these tendons the changes in load and elongation shall be measured 1
simultaneously at a minimum of three approximately equally spaced levels of force between zero and the seating force. If the
/- s\\
elongation corresponding to a specific load differs by more than 10%
from that recorded during the insta11' tion, an investigation shall a
k be made to ensure that the difference is not related to wire 1
failures or slip of wires in anchorages. This condition shall be considered as an indication of abnormal degradation of the containment structure, d.
Verifying the OPERABILITY of the sheathing filler grease by assur,ang:
1.
There are no changes in the presence or physical appearance of the sheathing filler-grease including the presence of free water, 2.
- 1. mount of grease replaced does n'ot exceed 5% of thc net duct volume, wher, injected at + 10% of the specified installation
- pressure, 3.
Minimum grease coverage exists for the different parts of the anchorage system, 4.
During general visual examination of the containment external surface, that grease leakage that could affect containment integrity is not present, and 16.6-6 Rev. 0 1
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TECHNICAL REQUIREMENTS MANUAL G
M CONTAINMENT SYSTEMS (Q'J O SURVEILLANCE REQUIREMENTS (Continued)
J 5.
The chemical properties of the filler material are within the A
tolerance limits specified as follows:
4 J
\\
Water Content 0 - 10% by dry weight Chlorides 0 - 10 ppm g
Nitrates 0 - 10 ppm Sulfides 0 - 10 ppm Reserved Alkalinity
>0 h3 Failure to meet the requirements of Section 16.6.1.2.1.a (d)
(4.6.1.6.1.d) shall be considered as an indication of abnormal degradation of the containment structure.
- 16. 6.1. 2.1. b ( 4. 6.1. 6. 21 End Anchorages and Adjacent Concrete Surfaces. As an assurance of the structural integrity of the containment vessel, tendon anchorage assembly hardware (such as bearing plates, stressing washers, wedges, and buttonheads) of all tendons selected for inspection shall be visually examined. Tendon anchorages selected for inspection shall be visually examined to the extent practical without dismantling the load bearing components of the anchorages. Bottom grease caps of all vertical tendons shall be visually inspected to detect grease leakage or grease cap deformations. The surrounding concrete shall also be checked visually for indication of any abnormal condition. The frequency of this surveillance shall be in accordance with Section 16.6.1.2.1 (4.6.1.6.1).
Significant grease leakage, grease cap deformation or abnormal conct.ie condition shall be considered as an indication of abnormal degradation of the containment structure.
p_,
kjl
- 16. 6. a.2.1. c ( 4. 6.1. 6. 3) Containment Vessel Surfaces. The exterior surface of m
the containment shall be visually examined to detect areas of large spall, severe scaling, D-cracking in an area of 25 sq. ft.'or more, other surface deterioration or dicintegration, or grease leakage, each of which shall be considered as evidence of abnormal degradation of structural integrity cf the containment. This inspection shall be performed in accordance with the Containment Leakage Rate Testing Program.
(See Technical Specification 6.8.5)
BASES 16.6.1.2.2 This limitation ensures that the structural integrity of the containment will be maintained in accordance with safety analysis requirements t
for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 48.9 psig in the event l
of a steam line break accident. The measurement of containment tendon lift-off
. force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed exterior surfaces of the containment, and l
16.6-7 Rev. 1 l
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TECHNICAL REQUIREMENTS MANUAL Co 4
01?
the Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.
The Surveillance Requirements for demonstrating the containment's 3
structural integrity are in compliance with the recommendations of proposed Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in S
Prestressed Concrete Containment Structures," April 1979, and proposed D3 Regulatory Guide 1.35.1, " Determining Prestreesing Forces for Inspection of
\\
Prestressed Concrete Containments," April 1979.
g The required Special Reports from any engineering evaluation of containment g
abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection proc <. dure, the tolerance on cracking, the results of the engineering evaluation 6
and the corrective actions taken.
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16.7 (3/4.7) PLANT SYSTEMS 16.7.1 (3/4.7.2)
STEAM GENERATOR PRESSURE / TEMPERATURE LDf1TATION LIMITING QNDITION FOR OPERATION D
16.7.1.1 (3.7.2)
The temperatures of both the reactor and secondary coolants M
in the steam generator shall be greater than 70*F when the pressure of either
\\
coolant in the steam generator is greater than 200 psig.
h APPLICABILITY: At all tires.
ACTION:
W With the requirements of the above sectiott not satisfied:
Reduce the steam generator pressure of the applicable side to less a.
than or equal to 200 psig within 30 minutes, and b.
Perform an engineering evaluation'to determine the effect of the everpressurization on the structural integrity of the steam generator.
Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above WO'F.
SUAVEILLANCE REQUIREMENTS 16.7.1.1.1 (4.7.2)
The pressure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the temperature of either the reactor or secondary coolant is less than 70*F.
h BASES 16.7.1.1.2 The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximura allowable fracture toughness stress limits. The limitations of 70'F and 200 psig are based on a steam generet or RTNDT of 60'F and are sufficient to l prevent brittle fracture.
16.7-1 Rev. O
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TECHNICAL REQUIREMENTS MANUAL D
OM PLANT SYSTEMS U
s, 16.7.2 (3/4.7.8) SNUBBEPS LIMITING CONDITION FOR OPERATION D
I' 16.7.2.1 (3.7.8)
All snubbers shall be OPERABLE. The only snubbers excluded from the requirement are those installed on nonsafety-related systems and then A
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only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
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APPLICABILITY: MODES 1, 2, 3, and 4.
MODES 5 and 6 for snubbers located on
\\
systems required OPERABLE in those MODES.
M
-ACTION:
4 With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per TRM 16.7.2.1.1.g on the attached component or declare the l
attached system inoperable and follow the appropriate ACTION statement for that system.
SURVEILLANCE REQUIREMENTS 16.7.2.1.1 (4.7.8)
Each snubber shall be demonstrated OPERABLE by performance of the,following augmented inservice inspection program and the requirements of General Surveillance Requirement 16.0.4.5.
I a.
Inspect' on Types
[ }
As use.d in this specification type of snubber shall mean snubbers
(_,f of the same design and manufacturer, irrespective of capacity.
b.
Visual Inspections Snubbers are categorized 3: Anaccessible or accessible during reactor operation.. Each of these catenories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 16. 7-1 (4. 7-2). The visual inspection interval for each type snubber shall be determined based upon the criteria provided in Table 16.7-1 (4.7-2) and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before Operating License Amendment 44.
16.7-2 Rev. 1 l
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TECHNICAL REQUIREMENTS MANUAL D
AN PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) k c.
Visual Inspection Acceptance Criteria T
M Visual inspections shall verify that:
(1) there are no visible
\\
indications of damage or impaired OPERABILITY, and (2) attachments to the foundation or supporting structure are functional, and l
3)
(3) fasteners for attachment of the snubber to the component and to g
tne snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of M
establishing the next visual inspection interval, provided that:
(1) the cause of the rejection is clearly established and remedied l
for that particular snubber and for other snubbers irrespective of i
type that may be generically susceptibles or (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Section 16.7.2.1.lf (4.7.8f).
l l
A review and evaluation shall be performed and documented to determine system operability with an unacceptable snubber. If operability cannot be justified, the system shall be declared inoperable and the ACTION requirements shall be met, d.
Transient Event Inspection An inspection shall be performed of all mechanical l
snubbers attached to sections of systems that have experienced unexpected potentially damaging transients as determined from a review of operational data and a visual inspection of the systems
(
within 6 months following such an event.
In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following (1) manually induced snubber movements (2) evaluation of in-place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel, i
16.7-3
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TECHNICAL REQUIREMENTS MANUAL D
O I*
PLANT SYSTE*fS (jC SURVEILLANCE REQUIREMENTS (Continued)
~
e.
Functional Tests During the first refueling shutdown and at least once per 18 months 2
thereafter during shutdown, a representative sample of snubbers of M
each type shall be tested using one of the following sample plans.
\\
The sample plan shall be selected prior to tne test period and cannot be changed during the test period. The NRC Regional Administrator g
shall be notified in writing of the sample plan selected for each
. '\\
snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:
M
- 1) At least 10% of the total of each type of snubber shall be functionally _ tested either in place or in a bench test.
For each snubber of a type that does not meet the functional test acceptance criteria of Section 16.7.2.1.lf (4.7.Bf), an
]
additional 10% of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested, or
- 2) A representative sample of each type of snubber shall be functionally tested in accordance with rigure 16.7-1 (4.7-1).
j "C" is the total number of snubbers of a type found not meeting the acceptance requirements of Section 16.7.2.1.lf (4.7.8f).
The cumulative number of snubbers of a type tested is denoted by "N".
At the end of each day's testing, the new values of "N" and "C" I
(previous day's total plus current day's increments) shall be plotted on the rigure 16.7-1 (4. 7-1). If at any time the point p
plotted falls in the " Reject" region, all snubbers of that type j
shall be functionally tested. If at any time the point plotted v
falls in the " Accept" region, testing of snubbers of that type-may be terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or the " Reject" region or all the snubbers of that type have been tested; or
- 3) An initial representative sample of 55 snubbers shall be functionally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial smnple size multiplied by the factor, 1 + C/2 where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. The results from this sample plan shall be I
plotted using an " Accept" line which follows the equation N =
55 (1 + C/2). Each snubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the " Accept" line, testing of that type of snubber may be i
terminated. If the point plotted falls above the " Accept" line testing must continue until the point falls in the " Accept" region or all the snubbers of that type have been tested.
j
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TECilNICAL REQUIREMENTS MANUAL C) p M.
PLANT SYSTEMS
\\vhO SURVEILLANCE REQUII'.EMENTS (Continued)
O e.
Functional Tests (Continued)
.C d
Testing equipment failure during functional testing may invalidate
\\
that day's testing and allow that day's testing to resume anew at a later time provided all snubbers tested with the failed equipment g
during the day of equipment failure are retested. The representative
\\
sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning the testing. The review shall ensure, as far as 3) practicable, that they are representative of the various configurations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same location as sr.ubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional test results shall be reviewed at that time to determine if additional camp'.es should be limited to the type of snubber which has failed the functional testing.
f.
Functional Test Acceptance Criteria The snubber functional test shall verify that:
- 1) Activation (restraining action) is achieved within the specified range in both tension and compression;
)
(9
/
(
)
- 2) DELETED, and l
3)
For mechanical anubbers, the force required to initiate or main-tain motion of the snubber is within the specified range in both directions of travel.
Testing methods may be used to measure parameters indirectly or pcrameters other than those specified if those results can be correlated to the specified parameters through established methods.
g.
Service Life Monitorino Program An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be j
subject to the same failure mode.
l 16.7-5 Rev. 1 l
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TECHNICAL REQUIREMENTS MANUAL D
f ( ITI PLANT SYSTEMS i
D
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SURVEILLANCE REQUIREMENTS (Continued) f g.
Service Life Monitoring Program (Continued) k f0 For the snubbers found inoperable, an engineering evaluation shall l
\\
be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluatian shall be to g
determine if the components to which the inoperable tnubbers are
-\\
attached were adversely affected by the inoperability of the snubbers in order to ensure that the component' remains capable of meeting the designed service.
W C
If any snubber selected for functional testing either fails to i
lock up or fails to move, i.e.,
frozen-in-place, the cause will be l
evaluated and, if caused by manufacturer or design deficiency, all l
snubbers of the same type subject to the same defect shall be functionally tested. This testing requirement shall be independent of the requirements stated in Section 16.7.2.1.le (4.7.8e). for snubbers not meeting the functional test acceptance criteria, h.
Functional Testing of Repaired and Replaced Snubbers
- p. '
Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent
,-s
/
service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.
monitored to ensure that the service life is not exceeded between surveillance inspections. The maximum expected service life for various seals, springs, and other critical parts shall be determined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history. Critf. cal parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts replacements shall be documented and the clocumentation shall be retained in accordance with Technical Specification 6.10.2.
(See Technical Specification 6.8.5) 16.7-6 Rev. 1 l
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TECHNICAL REQUIREMENTS MANUAL
{N BASES o
q g,
16.7.2.1.2 All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is I
maintained during and following a seismic or other event initiating dynamic 4
loads.
E
\\
Snubbers are classified and grouped by design and manufacturer but not by size.
For example, mechanical snubbers utilizing the same design features of the 2-g kip,10-kip, and 100-kip capacity tranufactured by Company "A" are of the same y-type. The same design mechanical snubbers manufactured by Company "B" Mr the purposes of this Technical Requirement would be of a different type, f am
{
either manufacturer. Snubbers may also be classified and grouped by iy inaccessible or accessible for visual inspection purposes. Therefore., each snubber type may be grouped for inspection in accordance with accessibility.
a j
A list of is?dividual snubbers with detailed information of snubber location and size and of systems affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50.
The accessibility of each snubber shall be determined and approved by the Plant Safety Review Committee. The determination shall be based upon the existing radiation levels and the i
expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g.,
temperature, atmosphere, location etc.), and the recommendations of Regulatory Guides 8.8 and 2.10, The add' 4.on or deletion of any mechanical snubber shall
{
be made in accordance with Section 50.59 of 10 CFR Part 50.
The visual inspection frequency is based upon maintaining a constant level of snubber protection during an earthquake or severe transient. Therefore, the n
required inspection i.iterval varies inversely with the observed snubber
(
failures and is determined by the number of inoperable snubbers found during an b'
inspection of each type.
In order to establish the inspection frequency for each type of snubber on a safety-related system, it was assumed that the frequency of snubber failures and initiating events is constant with time and that the failure of any snubber could cause the system to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.
The acceptance criteria are to be used in the visual inspection to determine OPERASILITY of the snubbers. Since the visual inspections are augmented by l
functional testing program, the visual inspection need not be a hands on inspection, but shall require visual scrutiny sufficient to assure that fasteners or mountings for connecting the snubbers to supports or foundations shall have no visible bolts, pins or fasteners 16.7-7 Rev.'1 l
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TECHNICAL REQUIREMENTS MANUAL m
missing, or other visible signs of physical damage such as cracking or O
loosening.
To provide assurance of snubber functional reliability, one of three f
functional testing methods are used with the stated acceptance criteria:
1.
M Functionally test 10% of a type of snubber with an additional.104 tested for each functional testing failure, or
\\
2.
Functionally test a sample size and determine sample acceptance or S
rejection using Figure 16. 7-1 (4.7-1), or 3.
Functionally test a representative sample size and determine sample g
acceptance or rejection using the stated equation.
Figure 16.7-1 (4.7-1) was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by Acheson J.
Duncan.
Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the completion of their fabrication or at a subsequent date.
Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the i
exemptions.
The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber,' seal replaced, spring replaced, in high radiation area, in high temperature area,-
O
.etc.).
The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.
1 i
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n)
O l
TABLE 16.7-1 (4.7-2) l SNUBBER VISUAL INSPECTION INTERVAL l
NUMBER OF UNACCEPTABLE SNUBBERS i
k Population Column A Column B Column C l
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per Category Extend Interval Repeat Interval Reduce Interval (Notes 1 and 2)
(Notes 3 and 6)
(Notes 4 and 6)
(Notes 5 and 6) l i
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g.
1 0
0 1
80 0
0 2
p 100 0
1 4
150 0
3 8
200 2
5 13
)(
300 5
12 25
.h 400 8
18 36 500 12 24 48 750 20 40 78 1000 or greater 29 56 109 Note 1: The next visual inspection interval for a snubber category shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during that interval.
Snubbers may be ' categorized, based upon their accessibility during power operation, as accessible or inaccessible. These categories may be examined separately or jointly. However, categories must be determined and documented before any inspection and that determination shall be the basis upon which to determine the next inspection interval for that category.
Note 2: Interpolation between population per category and the number of unacceptable snubbers is peratissible. Use next lower integer for the value of the limit for Columns A, B, and C if that integer includes a fractional value of unacceptable snubbers as determined by interpolation.
1 1
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-D TECHNICAL REQUIREMENTS MANUAL U
O' A Note 3: If the number of unacceptable snubbers is equal to or less than the D
number in Column A, the next inspection interval may be twice the l
previous interval but not greater than 48 months.
l 4
If the number of unacceptable snubbers is equal to or less than the Note 4:
number in Column B but greater than the number in Column A, the next l
4 inspection interval shall be the same as the previous interval, 1
YN
\\
Note 5: If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be pj two-thirds of the previous interval. However, if the number of i
.g unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be
~
reduced proportionally by interpolation, that is, the previous interval shall be reduced by a factor that is one-third of the j
ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Column B and C.
Note 6: The provisions of TRM 16.0.4.2 are applicable for l
all inspection intervals up to and including 48 months.
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0:
,0 9
2
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8
/
REJECT 6
j r
C 5
r D+
7 CONTINUE TESTING r
O l
2 r
ACCEPT 1
7 0
10 20 30 40 50 80 M
80 M
1M N
Figure 16,7-1 (4.7-1) l SAMPLE PIAN 2) FOR SNUBBER IVNCTIONAL TEST i
16.7-11 Rev. 1 l
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pR PLANT SYSTEMS
(
D 16.7.3 (3/4.7.9) SEALED SOURCE CONTAMINATION D
LIMITING CONDITION FOR OPERATION
.0 E
16.7.3.1 (3.7.9)
Each sealed source containing radioactive material either in
\\
excess of 100 microcuries of beta and/or gamma-emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal 9
to 0.005 microcurie of removable contamination.
y APPLICABILITY: At all times.
M ACTION:
With a sealed source having removable contamination in excess of the a.
above limits, immediately withdraw the sealed source from use and either:
1.
Decontaminate and repair the sealed source, or 2.
Dispose of the sealed source in accordance with Commission Regulations.
b.
The provisions of General Operational Requirements 16.0.3.3 and 16.0.3.4 are not applicable.
SURVEILLANCE REQUIREMENTS
(h 16.7.3.1.1.a (4.7.9.1)
Test Requirements - Each sealed source shall be tested
,,)
for leakage and/or contamination by:
a.
The licensee, or b.
Other persons specifically authorized by the Commission or an Agreement State.
The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample.
- 16. 7. 3.1.1.b ( 4. 7. 9. 2) Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.
i a.
Sources in use - At least once per 6 months for all sealed sources
{
containing radioactive materials:
- 1) With a half-life greater than 30 days (excluding Hydrogen 3) and 2)
In any form other than gas.
16.7-12 Rev. 0
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i eM PLANT SYSTEMS
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(
SURVEILLANCE REQUIREMENTS (Continued)
A b.
Stored sources not in use - Each sealed source and fission detector E
shall be tested prior to use or transfer to another licensee unless
)
C tested within the previous 6 months. Sealed sources and fission
\\
detectors transferred without a certificate indicating the last test j
date shall be tested prior to being placed into use; and
\\
Startup sources and fission detectors - Each sealed startup source c.
and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair y) or maintenance to the source.
16.7.3.1.1.c (4.7.9.3)
Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage testa reveal the presence of greater than or equal to 0.005 microcurie of removable contamination.
BASES 16.7.3.1.2 The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.
Sealed sources are classified into three groups according to their use, with
('"'j}
Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not.
Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism, i
16.7-13 Rev. 0
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l TECHNICAL REQUIREMENTS MANUAL
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G fB PLANT SYSTEMS 16.7.4 (3/4.7.12) AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION
.o.
I 16.7.4.1 (3.7.12)
The temperature limit of each area given in Table 16.7-2 I
\\
shall not be exceeded for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or by more than 30'F.
.1-l APPLICABILITY: Whenever the equipment in an affected area is required to be
{
h ACTION:
W L.
a.
With one or more areas exceeding the temperature limit (s) shown in D
1 Table 16.7-2 for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, prepare an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of
)
General Operational Rcquirements 16.0.3.3 and 16.0.3.4 are not applicable.
b.
With one or more areas exceeding the temperature limit (s) shown in Table 16.7-2 by more than 30'F, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either restore the area (s) to within the temperature limit (s) or declare the equipment in the affected area (s) inoperable, j
SURVEILLANCE REQUIREMENTS 16.7.4.1.1 (4.7.12)
The temperature in each of the areas shown in Table 16.7-2 shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
BASES i
16.7.4.1.2 The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may cegrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of +3*F.
16.7-14 Rev. O OV
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TECHNICAL REQUIREMENTS MANUAL D
Of.
TABLE 16.7-2
)
it
)D I
AREA TEMPERATURE MONITORING 3
MAXIMUM TEMPERATURE AREA LIMIT (
'F )
'l)
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1.
ESW Pump Room A.
119 4
2.
ESW Pump Room B 119
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3.
Auxiliary Teodwater Pump Room A 119 f0 4.
Auxiliary Feedwater Pump Room B 119 5.
Turbine Driven Auxiliary Feedwater 147 Pump Room 6.
ESF Switchgear Room I 87
(
7.
ESF Switchgear Room II 87 8.
RHR Pump Room A 119 9.
RHR Pump Room B 119 10.
CTMT Spray Pump Room A 119 11.
CTMT Spray Pump Room B 119 12.
Safety Injection Pump Room A 119 13.-
Safety Injection Pump Room b 119 14.
Centrifugal Charging Pump Room A 119 15.
Centrifugal Charging Pump Room B 119 16.
Electrical Penetration Room A 101 17.
Electrical Penetration Room B 101 l
18.
Component Cooling Water Room A 119 j
19.
Component Cooling Water Room B 119 1
20.
Diesel Generator Room A 119 21.
Diesel Generator Room B 119 22.
Control Room 84 i
16.7-15 Rev. 0
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TECHNICAL REQUIREMENTS MANUAL O
[] i.
16.8 (3/4.8) ELECTRICAL POWER SYSTEMS T
VU 16.8.1 (3/4.8.4) EIECTRICAL EQUIPDENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION X
\\
16.8.1.1 (3.8.4.1)
For each containment penetration provided with a penetration conductor overcurrent protective device (s), each device shall be y
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APPLICABILITY: MODES 1, 2, 3, and 4.
r.
[4 ACTION:
7 With one or more of the above required containment penetration conductor overcurrent protective device (s) inoperable:
a.
Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated backup circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, declare the affected system or component inop erable, and verify the backup circuit breaker to be tripped or the inoperable circuit breaker racked out, or removed, at least once per 7 days thereafter; the provisions of General Operational Requirement 16.0.3.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, their inoperable circuit i
breakers racked out, or removed, or b.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD
(
SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 16.8.1.1.1 ( 4. 8. 4.1)
Protective devices required to be OPERABLE as containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE.
a.
At least once per 18 months:
- 1) By verifying that the 13.8 kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers, and performing the following:
a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and 16.8-1 Rev. O
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WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL 0
s%
ELECTRICAL POWER SYSTEMS U
SURVEILLANCE REQUIREMENTS (Continued) m S
c) For each circuit breaker found inoperable during these a
functional tests, an additional representative sample of M
at least 10% of all the circuit breakers of the inoperable
\\
type shall also be functionally tested until no more l
failures are found or all circuit breakers of that type l
g have been functionally tested.
\\
- 2) By selecting and functiorally testing a representative s.mple of at least 10% of each type of lower voltage circuit breakers.
N Circuit breake s selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers shall consist of injecting a current in excess of the breakers nominal Setpoint and measuring the response time. The measured response time will be compared to the manufacturer's data to I
ensure that it is less than or equal to a value specified by the manufacturer. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.
b.
At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures O
prepared in conjunction with its manufacturer's recommendations.
BASES 16.8.1.1.2 Containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance.
The Surveillance Requirements applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker. Each manufacturer's molded case and metal case circuit breakers are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers are tested. If a wide variety exists within any manufacturer's brand of circuit breakers, it is necessary to divide that manufacturer's breakers into groups and treat each group as a separate type of breaker for surveillance purposes.
A list of containment penetration conductor overcurrent protective devices whose circuit limiting fault current exceeds the penetration rating, with 16.8-2 Rev. 0 O
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WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL D
(
)3 information of location and size and equipment powered by the protected circuit, is available at the plant site in accordance with Section 50.71(c) of 10 CFR Part 50.
The addition or deletion of any containment penetration g
conductor overcurrent protective device would be made in accordance with Section 50.59 of 10 CFR Part 50.
3 03
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9
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ro -
I
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i 16.8-3 Rev. O O
H 3:
wotr CREEx D
TECHNICAL REQUIREMENTS MANUAL O
T (g'
ELECTRICAL POWER SYSTEMS
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D 16.8.2 (3/4.8.4.8) SURVEILLANCE REQUIREMr.NTS (DISSEL GENERATOR) 3 16.8.2 (3/4.8.4.8.1.1.2)
Each diesel generator shall be demonstrated OPERABLE:
h XI a.
(g.) At least once per 18 months, during shutdown by:
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1)
Verifying that the auto-connected loads to each diesel g
generator do not exceed 6201 kWs b
N
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i 16.8-4 Rev. 1 l
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1 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL C
.F 16.9 (3/4.9) REFUELING OPEJtATIONS l
16.9.1 (3/4.9.5) CCBGWNICATIONS LIMITING CONDITION FOR OPERATION
.0 16.9.1.1 (3.9.5) Direct communications shall be maintained between the control l
X room and personnel at the refueling station.
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l APPLICABILITY: During CORE ALTERATIONS.
l fJ ACTION:
g
~
When direct communications between the control room and personnel at the M
refueling station cannot be maintained, suspend all CORE ALTERATIONS.
SURVEILLANCE REQUIREMENTS 16.9.1.1.1 ( 4. 9. 5 ) Direct communications between the control zoom and personnel at the refueling station shall be demonstrated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, BASES 16.9.1.1.2 The requirement for communications capability ensures that refueling station personnel can ce promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.
16.9-1 Rev. O t
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TECHNICAL REQUIREMENTS MANUAL l
0 REFUELING OPERATIONS 16.9.2 (3/4.9.6) REFUILING MACHINE LIMITING CONDITION FOR OPERATION J
16.9.2.1 (3.9.6)
The refueling machine shall be used for movement of drive M
rods or fuel assemblies and shall be OPERABLE with:
\\
a.
The refueling machine used for movement of fuel assemblies having:
j M
l 1)
A minimum capacity of 4800 pounds, i
g 2)
Automatic overload cutoffs with the following Setpoints:
- 0 a)
Primary - less than or equal to 250 pounds above the indicated suspended weight for wet conditions and less than or equal to 350 pounds above the indicated suspended weight for dry conditions, and b)
Secondary - less than or equal to 150 pounds above the primary overload cutoff.
3)
An automatic load reduction trip with a Setpoint of less than or equal to 250 pounds below the surpended weight for wet con-ditions er dry condi'. ions.**
b.
The auxiliary hoist used for latching and unlatching drive rods and thimble plug handling operations having:
( ';
1)
A minimum capacity of 3000 pounds, and
~
2)
A 1000-pound load indicator which shall be used to monitor lifting loads for these operation.
APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor vessel.
1 ACTION:
i With the requirements for refueling machine and/or auxiliary hoist OPERABILITY not satisfied, suspend use of any inoperable refueling me: hine crane and/or auxiliary hoist from operations involving the 1
movement of drive rods and fuel assemblies within the reactor vessel.
SURVEILLANCE REQUIREMENTS i
16.9.2.1.1.a (4.9.6.1) The refueling machine used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the movement of fuel assemblics in the reactor vessel by i
16.9-2 Rev. O O
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TECHNICAL REQUIREMENTS MANUAL D
Il REFUELING OPERATIONS vo SURVEILLANCE REQUIREMENTS (Continued) f performing a load test of at least 125% of the secondary automatic overload cutoff and demonstrating an automatic load cutoff when the refueling machine W
load exceeds the Setpoints of Section 16.9.2.la.2 ) and by demonstrating an
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automatic load reduction trip when the load reduction exceeds the Setpoint of Section 16.9.2.la.3.
Q
- 16. 9. 2.1.1. b ( 4. 9. 6. 2 ) Each auxiliary hoist and associated load indicator used
\\.
for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the movement of drive rods within the N
reactor vessel by performing a load test of a'. least 1250 pounds.
y BASES 16.9.2.1.2 The OPERABILITY requirements for the refueling machine and auxiliary hoist ensure that: (1) manipulator cranes > ill be used for movement of drive rods and fuel assembli<a, (2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the cort internals and reactor vessel are protected from wscessive lifting force in the event they are inadvertently engaged during lifting operations.
When a fuel assembly bottom nozzle is less than or equal to two (2) inches above the full down position in the core, upender, and ECCA char.ge fixture during raising of the fuel assembly, primary cutoff will be automatically bypassed.
j
- When a fuel assembly bottom nozzle is less than or equal to two (2) inches above the full down position in the core, upender, and RCCA change fixture during lowering of the fuel assembly, the automatic underload cutoff will be automatically bypassed.
16.9-3 Rev. O v
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3 WOLF CREEK D
TECHNICAL REQUIREMENTS MANUll D
M REFUELING OPERATIONS
[
7
(
16.9.3 (3/4.9.7) CRANE TRAVEL - SPENT FUEL STORAGE FACILITY LIMITING CONDITION FOR OPERATION J
16.9.3.1 (3.9.7)
Loads in excess of 2250 pounds shall be prohibited from M
travel over fuel assemblies in the spent fuel storage facility.
\\
l APPLICABILITY: With fuel assemblies in the spent fuel storage facility.
4 ACTION:
a.
With the requirements of the above specification not satisfied,
]
M place the crane load in a safe condition.
1 b.
The provisions of General Operational Requirements 16.0.3.3 and 16.0.3,4 are not applicable.
1 SURVEILLANCE REQUIREMENTS I
16.9.3.1.1 (4.9.7)
Crane interlocks and physical stops which prevent crane travel with loads in excess of 2250 pounds over fuel assemblies stull be demonstrated OPERABLE within 7 days prior to cr.ne use and at least once per 7 I
days thereafter during crane operation.
BASES.
16.9.3.1.2 The restriction on movement of loads in excess of the nominal O
weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool areas ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.
16.9-4 Rev. O i
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WOLF CREEK D
TECHNICAL REQUIREMENTS MANUAL D
[] A 16.9.4 (3/4.9.10) MATER LEVEL - REACTOR VE5SEL D
CON?ROL RODS g
LIMITING CONDITION FOR OPERATION
. ~.
S 16.9.4.1 (3.9.10.2) At least 23 feet of water shall be maintained over the top (0
of the irradiated fuel assemblies within the reactor pressure vessel.
\\
APPLICABILITY: During movement of control rods within the reactor pressure 7y vessel while in MODE 6.
ACTION:
i M
With the requiremente of the above specification not satisfied, suspend all l
g operations involvirg movement of control rods within the pressure vessel.
^5 SURVEILLRNCE REQUIREMENTS l
16.9.4.1.1 (4.9.10.2)
The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during movement of control rods within the reactor vessel.
i BASES 16.9.4.1.2 See Technical Specification Bases 3/4.9.10.
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16.9-5 Rev. 0
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TECHNICAL REQUIREMENTS MANUAL D
m 16.10 (3/4.10)
SPECIAL TEST EXCEPTIN S s..) o 16.10.1 (3/4.10,5) POSITION INDICATION SYSTEM - SNUTDOHN LIMITING CONDITION It)R OPERATION l
5 16.10.1.1 (3.10.5) The limitations of Section 16.1.3.1 may be suspended during l
J) the performance of individual full-length shutdown and control rod drop time i
\\
measurements provided only one shutdown or control bank is withdrawn from the fully inserted position at a time.
M APPLICABILITY: HODES 3, 4, and 5 during performance of red drop time g
measurements and during surveillance of digital rod position indicators for
~
~.9 ACTION:
With the Position Indication System inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers.
SURVEILLANCE REQUIREMENTS l
16.10.1.1.1 (4.10.5) The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:
a.
Within 12 steps when the rods are etationary, and b.
Within 24 steps during rod motion.
BASES
._4 16.10.1.1.2 This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements.
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16.10-1 Rev. O S
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TECHNICAL REQUIREMENTS MANUAL 0
DR 16.11 (3/4.11) RADIQACTIVE EFFLUENTS
/
\\ )3 16.11.1 LIQU"ID HOIAUP TANKS LIMITING CONDITION FOR OPERATION
.C 16.11.1.1 (3.11.1.4)
The quantity of radioactive material contained in each of M
the following unprotected outdoor tanks shall be limited to less than or equal
\\
to 150 Curies, excluding tritium and dissolved or entrained ncble gases.
Q a.
Reactor Makeup Water Storage Tank, b.
Refueling Water Storage Tank, g
c.
Condensate Storage Tank, arad
~
d.
Outside temporary tanks, excluding demineralizer vessels and liners M
being used to solidify radioactive wastes.
1 APPLICABILITY: At all times.
ACTION:
a.
With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events leading to this condition in the next Annual Radioactive Effluent Release Report, pursuaht to Technical Specification 6.9.1.7.
b.
The provisions of General Operational Requirements 16.0.3.3 and i
16.0.3.4 are not applicable.
SURVEILLANCE REQUIREMENTS V
16.11.1.1.1 (4.11.1.4)
The quantity of radioactive natarial contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added and within 7 days following any-addition of radioactive material to the tank.
(See Technical Specification 6.8.5)
BASES 16.11.1.1.2 The tanks listed in this specification include all those outdoor radwaste tanks that are not' surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.
Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
16.11-1 Rev. O O
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TECHNICAL REQUIREMENTS MANUAL 16.11.2 EXPIDSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION g
16.11.2.1 (3.11.2.5)
The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 3% by volume whenever the 0
hydrogen concentration exceeds 4% by volume.
W
\\
APPLICABILITY: At all times.
j
()
ACTION:
b a.
With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM
~
greater than 3% by volume but less than or equal to 4% by volume, l
b reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
/?
b.
With the concentration of oxygen in the WA:TE GAS HOLDUP SYSTEM greater than 4% by volume and the hydrogen concentration greater
{
than 4% by volume, immediately suspend all additions of waste gases 1
to the system and reduce the concentration of oxygen to less than or equal to 4% by volume, then take ACTION a. above.
c.
The provisions of General Operational Requirements 16.0.3.3 and 16.0.3.4 are not applicable.
i SURVEILLANCE REOUIREMENTS m _ _ _ _ __
16.11.2.1.1 (4.11.2.5) The concentrations of hydrogen and oxygen !,
.ne WASTE CAS HOLDUP SYSTEM shall be determined to be within the above limits by l
/~'
continuously monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the l (
hydrogen and oxygen monitors required OPERABLE by Table 16.3-7 of Section N-16.3.1.6.
l (See Technical Specification 6.8.5)
BASES 16.11.2.1.2 This specification is provided to ensure that the concentration of i
potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is l
maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen.
Maintaining the concentration of hydrogen and exygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50, i
16.11-2 Rev. O l
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WOLF CREEK l
D TECHNICAL REQUIREMENTS MANUAL l
0
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RADIOACTIVE EFFLUENTS O
?
J 16.11.3 GAS STORAGE TANKS l
LIMITING CONDITION FOR OPERATION l
L' A
16.11.3.1 (3.11.2.6)
The quantity of radioactivity contained in each gas N
storage tank shall be limited to less than or equal to 2.5 x 105 Curies of
\\
noble gases (considered as Xe-133 equivalent).
S APPLICABILITY: At all times.
\\
ACTION:
N a.
With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, reduce the tank contents to within the limit, and describe the events leading to this condition in the next Annual Radioactive Effluent Release Report, pursuant to Technical Specification 6.9.1.7.
b.
The provisions of General Operational Requirements 16.0.3.3 and 16.0.3.4 are not applicable.
SURVEILLANCE REQUIREMENTS 16.12.3.1.1 (4.11.2.6) The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 7 days when radioactive materials are being added and within 7 days
(""'
following any addition of radioactive material to the tank.
(See Technical Specification 6.8.5)
BASES I
I II I
ll l 16.11.3.1.2 The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. Restricting the quantity of radioactivity contained in each gas storage thak provides assurance that ir ^5e event of an j
uncontrolled release of the tank's contents, the resulting w-bcdy exposure i
t to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will noc exceed 0.5 rem.
This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981.
Swapping between gas storage tanks is allowed without sampling if the tank being swapped to was previously demonstrated to have a curie content below 2 x 105 curies.
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1 16.11-3 Rev. 0 V)
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