ML15077A066

Wolf Creek - Changes to Technical Specification Bases, Revisions 61 Through 66

ML15077A066
Person / Time
Site:	Wolf Creek
Issue date:	03/11/2015
From:	Koenig S R Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 15-0023
Download: ML15077A066 (54)

Text

WeLF CREEK'NUCLEAR OPERATING CORPORATION Steven R. KoenigManager Regulatory AffairsMarch 11, 2015RA 15-0023U. S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Changes to Technical Specification Bases -Revisions 61 through 66Gentlemen:

The Wolf Creek Generating Station (WCGS) Unit 1 Technical Specifications (TS), Section5.5.14, "Technical Specifications (TS) Bases Control Program,"

provide the means for makingchanges to the Bases without prior Nuclear Regulatory Commission (NRC) approval.

Inaddition, TS Section 5.5.14 requires that changes made without NRC approval be provided tothe NRC on a frequency consistent with 10 CFR 50.71(e).

The Enclosure provides thosechanges made to the WCGS TS Bases (Revisions 61 through 66) under the provisions to TSSection 5.5.14 and a List of Effective Pages. This submittal reflects changes from January 1,2014 through December 31, 2014.This letter contains no commitments.

contact me at (620) 364-4041.

If you have any questions concerning this matter, pleaseSRK/rltEnclosure cc: M. L. Dapas (NRC), w/eC. F. Lyon (NRC), w/eN. F. O'Keefe (NRC), w/eSenior Resident Inspector (NRC), w/eP.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831An Equal Opportunity Employer M/F/HC/VET Enclosure to RA 15-0015Wolf Creek Generating StationChanges to the Technical Specification Bases(28 pages)

RTS Instrumentation B 3.3.1BASESAPPLICABLE SAFETY ANALYSES, LCO, andAPPLICABILITY

12. Undervoltape Reactor Coolant Pumps (continued) connected in parallel with the 13.8 kV power supply to each RCPmotor at the motor side of the supply breaker.

Each PT secondary side is connected to an undervoltage relay and time delay relay,as well as a separate underfrequency relay. The undervoltage relays provide output signals to the SSPS which trips the reactor, ifpermissive P-7 issatisfied (i.e. greater than 10% of rated thermalpower), when the voltage at one out of two RCP motors on bothbuses drops below 10578 Vac. The time delay relay preventsspurious trips caused by transient voltage perturbations.

This tripFunction will generate a reactor trip before the Reactor CoolantFlow -Low Trip Setpoint is reached.The LCO requires two Undervoltage RCP channels per bus to beOPERABLE, for a total of four channels.

The Trip Setpoint is_> 10,578 Vac.In MODE 1 above the P-7 setpoint, the Undervoltage RCP tripmust be OPERABLE.

Below the P-7 setpoint, all reactor trips onloss of RCP due to undervoltage are automatically blocked sincethe core is not producing sufficient power to generate DNBconditions.

Above the P-7 setpoint, the reactor trip onUndervoltage RCPs is automatically enabled.13. Underfrequency Reactor Coolant PumpsThe Underfrequency RCP reactor trip Function ensures thatprotection is provided against violating the DNBR limit due to aloss of flow in two or more RCS loops from a major networkfrequency disturbance.

An underfrequency condition will slowdown the pumps, thereby reducing their coastdown time following a pump trip. An adequate coastdown time is required so thatreactor heat can be removed immediately after reactor trip. Thereis one potential transformer (PT), with a primary to secondary ratioof 14400:120, connected in parallel with the 13.8 kV power supplyto each RCP motor at the motor side of the supply breaker.

EachPT secondary side is connected to an undervoltage relay and timedelay relay, as well as a separate underfrequency relay. Theunderfrequency relays provide output signals to the SSPS whichtrips the reactor, if permissive P-7 issatisfied (i.e. of greater than10% of rated thermal power), when the frequency of one out oftwo RCP motors on both buses drops below 57.15 Hz. The timedelay set on the underfrequency relay prevents spurious tripsWolf Creek -Unit 1B 3.3.1-19Revision 66 RTS Instrumentation B 3.3.1BASESAPPLICABLE

13. Underfrequency Reactor Coolant Pumps (continued)

SAFETY ANALYSES, LCO, and caused by transient frequency perturbations.

This trip FunctionAPPLICABILITY will generate a reactor trip before the Reactor Coolant Flow -LowTrip Setpoint is reached.The LCO requires two Underfrequency RCP channels per bus tobe OPERABLE, for a total of four channels.

The Trip Setpoint is_ 57.15 Hz.In MODE 1 above the P-7 setpoint, the Underfrequency RCPtrip must be OPERABLE.

Below the P-7 setpoint, all reactor tripson loss of RCP due to underfrequency are automatically blockedsince the core is not producing sufficient power to generate DNBconditions.

Above the P-7 setpoint, the reactor trip onUnderfrequency RCPs is automatically enabled.14. Steam Generator Water Level -Low LowThe SG Water Level -Low Low trip Function ensures thatprotection is provided against a loss of heat sink and actuates theAFW System prior to uncovering the SG tubes. The SGs are theheat sink for the reactor.

In order to act as a heat sink, the SGsmust contain a minimum amount of water. A narrow range low lowlevel in any SG is indicative of a loss of heat sink for the reactor.The level transmitters provide input to the SG Level ControlSystem. Therefore, the actuation logic must be able to withstand an input failure to the control system, which may then require theprotection function actuation, and a single failure in the otherchannels providing the protection function actuation.

ThisFunction also performs the ESFAS function of starting the AFWpumps on low low SG level.The LCO requires four channels of SG Water Level -Low Low perSG to be OPERABLE because these channels are sharedbetween protection and control.

The Trip Setpoint for the SGWater Level Low -Low is _> 23.5% of narrow range instrument span.In MODE I or 2, when the reactor requires a heat sink, the SGWater Level -Low Low trip must be OPERABLE.

The normalsource of water for the SGs is provided by the Main Feedwater (MFW) pumps (not safety related).

The MFW pumps are only inoperation in MODE 1 or 2 above the point of adding heat. TheWolf Creek -Unit 1B 3.3.1-20Revision 66 RCS Loops -MODE 5, Loops FilledB 3.4.7BASESLCO b. Core outlet temperature is maintained at least 1 0°F below(continued) saturation temperature, so that no vapor bubble may form andpossibly cause a natural circulation flow obstruction.

Note 2 allows one RHR loop to be inoperable for a period of up to2 hours, provided that the other RHR loop is OPERABLE and inoperation.

This permits periodic surveillance tests to be performed on theinoperable loop during the only time when such testing is safe andpossible.

Note 3 requires that the secondary side water temperature of each SG be_< 50°F above each of the RCS cold leg temperatures before the start of areactor coolant pump (RCP) with any RCS cold leg temperature

< 3680F.This restriction is to prevent a low temperature overpressure event due toa thermal transient when an RCP is started.Note 4 provides for an orderly transition from MODE 5 to MODE 4 duringa planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation.

This Note provides for thetransition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.RHR pumps are OPERABLE if they are capable of being powered andare able to provide forced flow if required.

When both RHR loops (ortrains) are required to be OPERABLE, the associated Component CoolingWater (CCW) train is required to be capable of performing its relatedsupport function(s).

The heat sink for the CCW System is normallyprovided by the Service Water System or Essential Service Water (ESW)System, as determined by system availability.

In MODES 5 and 6, oneDiesel Generator (DG) is required to be OPERABLE per LCO 3.8.2, "ACSources -Shutdown."

The same ESW train is required to be capable ofperforming its related support function(s) to support DG OPERABILITY.

AService Water train can be utilized to support RHR OPERABILITY if theassociated ESW train is not capable of performing its related supportfunction(s).

A SG can perform as a heat sink via natural circulation whenit has an adequate water level and is OPERABLE.

APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation ofthe reactor coolant to remove decay heat from the core and to provideproper boron mixing. One loop of RHR provides sufficient circulation forthese purposes.

However, one additional RHR loop is required to beOPERABLE, or the secondary side wide range water level of at least twoSGs is required to be _> 66%.Operation in other MODES is covered by:LCO 3.4.4, "RCS Loops-MODES 1 and 2";Wolf Creek -Unit 1B 3.4.7-3Revision 63 RCS Loops -MODE 5, Loops FilledB 3.4.7BASESAPPLICABILITY (continued)

LCO 3.4.5, "RCS Loops-MODE 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.5, "Residual Heat Removal (RHR) and CoolantCirculation

-High Water Level" (MODE 6); andLCO 3.9.6, "Residual Heat Removal (RHR) and CoolantCirculation

-Low Water Level" (MODE 6).ACTIONSA.1 and A.2If one RHR loop is inoperable and the required SGs have secondary sidewide range water levels < 66%, redundancy for heat removal is lost.Action must be initiated immediately to restore a second RHR loop toOPERABLE status or to restore the required SG secondary side waterlevels. Either Required Action A.1 or Required Action A.2 will restoreredundant heat removal paths. The immediate Completion Time reflectsthe importance of maintaining the availability of two paths for heatremoval.B.1 and B.2If no RHR loop is in operation, except during conditions permitted byNotes 1 and 4, or if no loop is OPERABLE, all operations involving introduction into the RCS, coolant with boron concentration less thanrequired to meet the minimum SDM of LCO 3.1.1 must be suspended andaction to restore one RHR loop to OPERABLE status and operation mustbe initiated.

To prevent inadvertent criticality during a boron dilution, forced circulation from at least one RCP is required to provide propermixing. Suspending the introduction into the RCS, coolant with boronconcentration less than required to meet the minimum SDM of LCO 3.1.1is required to assure continued safe operation.

With coolant addedwithout forced circulation, unmixed coolant could be introduced to thecore, however coolant added with boron concentration meeting theminimum SDM maintains acceptable margin to subcritical operations.

The immediate Completion Times reflect the importance of maintaining operation for heat removal.SURVEILLANCE SR 3.4.7.1REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the required loop is inoperation.

Verification may include flow rate, temperature, or pump statusmonitoring, which help ensure that forced flow is providing heat removal.The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient considering other indications andalarms available to the operator in the control room to monitor RHR loopperformance.

Wolf Creek -Unit 1B 3.4.7-4Revision 42 1 LTOP SystemB 3.4.12B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.12 Low Temperature Overpressure Protection (LTOP) SystemBASESBACKGROUND The LTOP System controls RCS pressure at low temperatures so theintegrity of the reactor coolant pressure boundary (RCPB) is notcompromised by violating the pressure and temperature (P/T) limits of10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPBcomponent for demonstrating such protection.

The PTLR provides themaximum allowable actuation logic setpoints for the power operated reliefvalves (PORVs) and the maximum RCS pressure for the existing RCScold leg temperature during cooldown,

shutdown, and heatup to meet theReference 1 requirements during the LTOP MODES.The reactor vessel material is less tough at low temperatures than atnormal operating temperature.

As the vessel neutron exposureaccumulates, the material toughness decreases and becomes lessresistant to pressure stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is increased only astemperature is increased.

The potential for vessel overpressurization is most acute when the RCS iswater solid, occurring only while shutdown; a pressure fluctuation canoccur more quickly than an operator can react to relieve the condition.

Exceeding the RCS P/T limits by a significant amount could cause brittlecracking of the reactor vessel. LCO 3.4.3, "RCS Pressure andTemperature (P/T) Limits,"

requires administrative control of RCSpressure and temperature during heatup and cooldown to preventexceeding the PTLR limits.This LCO provides RCS overpressure protection by having a minimumcoolant input capability and having adequate pressure relief capacity.

Limiting coolant input capability requires both safety injection pumps andone Emergency Core Cooling System (ECCS) centrifugal charging pumpto be incapable of injection into the RCS and isolating the accumulators.

The normal charging pump (NCP), in addition to one ECCS centrifugal charging pump flow, has been included in the analysis of design basismass input overpressure transient.

The pressure relief capacity requireseither two redundant RCS relief valves or a depressurized RCS and anRCS vent of sufficient size. One RCS relief valve or the open RCS vent isthe overpressure protection device that acts to terminate an increasing pressure event.Wolf Creek -Unit 1B 3.4.12-1Revision 61 LTOP SystemB 3.4.12BASESBACKGROUND With minimum coolant input capability, the ability to provide core coolant(continued) addition is restricted.

The LCO does not require the makeup controlsystem deactivated or the safety injection (SI) actuation circuits blocked.Due to the lower pressures in the LTOP MODES and the expected coredecay heat levels, the makeup system can provide adequate flow via themakeup control valve. If conditions require the use of more than oneECCS centrifugal charging pump for makeup in the event of loss ofinventory, either the NCP or other ECCS pumps can be made available through manual actions.The LTOP System for pressure relief consists of two PORVs with reducedlift settings, or two residual heat removal (RHR) suction relief valves, orone PORV and one RHR suction relief valve, or a depressurized RCS andan RCS vent of sufficient size. Two RCS relief valves are required forredundancy.

One RCS relief valve has adequate relieving capability toprevent overpressurization for the required coolant input capability.

PORV Requirements As designed for the LTOP System, each PORV is signaled to open if theRCS pressure approaches a limit determined by the LTOP actuation logic.The LTOP actuation logic monitors both RCS temperature and RCSpressure and determines when a condition not acceptable with respect tothe PTLR limits is approached.

The wide range RCS temperature indications are auctioneered to select the lowest temperature signal.The lowest temperature signal is processed through a function generator that calculates a pressure limit for that temperature.

The calculated pressure limit is then compared with the indicated RCS pressure from awide range pressure channel.

If the indicated pressure meets or exceedsthe calculated value, a PORV is signaled to open.The PTLR presents the PORV setpoints for LTOP. The setpoints arenormally staggered so only one valve opens during a low temperature overpressure transient.

Having the setpoints of both valves within thelimits in the PTLR ensures that the Reference 1 limits will not beexceeded in any analyzed event.When a PORV is opened in an increasing pressure transient, the releaseof coolant will cause the pressure increase to slow and reverse.

As thePORV releases

coolant, the RCS pressure decreases until a resetpressure is reached and the valve is signaled to close. The pressurecontinues to decrease below the reset pressure as the valve closes.Wolf Creek -Unit 1B 3.4.12-2Revision 61 LTOP SystemB 3.4.12BASESBACKGROUND (continued)

RHR Suction Relief Valve Requirements During LTOP MODES, the RHR System is operated for decay heatremoval and low pressure letdown control.

Therefore, the RHR suctionisolation valves are open in the piping from the RCS hot legs to the inletsof the RHR pumps. While these valves are open the RHR suction reliefvalves are exposed to the RCS and are able to relieve pressure transients in the RCS.The RHR suction isolation valves must be open to make the RHR suctionrelief valves OPERABLE for RCS overpressure mitigation.

The RHRsuction relief valves are spring loaded, bellows type water relief valveswith pressure tolerances and accumulation limits established bySection III of the American Society of Mechanical Engineers (ASME)Code (Ref. 3) for Class 2 relief valves.RCS Vent Requirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in anRCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must becapable of relieving the flow resulting from the limiting LTOP mass or heatinput transient, and maintaining pressure below the P/T limits. Therequired vent capacity may be provided by one or more vent paths.APPLICABLE SAFETY ANALYSESSafety analyses (Ref. 4) demonstrate that the reactor vessel is adequately protected against exceeding the Reference 1 P/T limits. In MODES 1, 2,and 3, the pressurizer safety valves will prevent RCS pressure fromexceeding the Reference 1 limits. In MODE 3 (with any RCS cold legtemperature

< 368°F) and below, overpressure prevention falls to twoOPERABLE RCS relief valves or to a depressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure reliefcapability.

The actual temperature at which the pressure in the P/T limit curve fallsbelow the pressurizer safety valve setpoint increases as the reactorvessel material toughness decreases due to neutron embrittlement.

Eachtime the PTLR curves are revised, the LTOP System must be re-evaluated to ensure its functional requirements can still be met using theRCS relief valve method or the depressurized and vented RCS condition.

Wolf Creek -Unit 1B 3.4.12-3Revision 0

LTOP SystemB 3.4.12BASESAPPLICABLE The PTLR contains the acceptance limits that define the LTOPSAFETY ANALYSES requirements.

Any change to the RCS must be evaluated against the(continued)

Reference 9 analyses to determine the impact of the change on the LTOPacceptance limits.Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:Mass Input Type Transients

a. Inadvertent safety injection; orb. Charging/letdown flow mismatch.

Heat Input Type Transients

a. Inadvertent actuation of pressurizer heaters;b. Loss of RHR cooling; orc. Reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

The following are required with exception described below during theLTOP MODES to ensure that mass and heat input transients do notoccur, which either of the LTOP overpressure protection means cannothandle:a. Rendering both safety injection pumps and one ECCS centrifugal charging pump incapable of injection (there are no limitations on theuse of the NCP during the LTOP MODES);b. Deactivating the accumulator discharge isolation valves in theirclosed positions or by venting the affected accumulator; andc. Precluding start of an RCP if secondary temperature is more than50°F above primary temperature in any one loop. LCO 3.4.5, "RCSLoops -MODE 3," LCO 3.4.6, "RCS Loops -MODE 4," andLCO 3.4.7, "RCS Loops -MODE 5, Loops Filled,"

provide thisprotection.

Operation below 350°F but greater than 3250F with all ECCS centrifugal charging and safety injection pumps OPERABLE is allowed for up to 4hours. During low pressure, low temperature operation all automatic safety injection actuation signals except Containment Pressure

-High areWolf Creek -Unit 1B 3.4.12-4Revision 61 LTOP SystemB 3.4.12BASESAPPLICABLE SAFETY ANALYSES(continued) blocked.

In normal conditions a single failure of the ESF actuation circuitry will result in the starting of at most one train of safety injection (one centrifugal charging pump, and one safety injection pump). Fortemperatures above 3250F, an overpressure event occurring as a result ofstarting two pumps can be successfully mitigated by operation of bothPORV's without exceeding Appendix G limit. Given the short timeduration that this condition is allowed and the low probability of a singlefailure causing an overpressure event during this time, the single failure ofa PORV is not assumed.

Initiation of both trains of safety injection duringthis 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel are not considered to be credibleaccidents.

Although LTOP is required to be OPERABLE when RCS temperature isless than 3680F, operation with all ECCS centrifugal charging pumps andboth safety injection pumps OPERABLE is acceptable when RCStemperature is greater than 3500F. Should an inadvertent safety injection occur above 3500F, a single PORV has sufficient capacity to relieve thecombined flow rate of all ECCS pumps and the NCP. Above 3500F, twoRCPs and all pressurizer safety valves are required to be OPERABLE.

Operation of an RCP eliminates the possibility of a 50°F difference existing between indicated and actual RCS temperature as a result ofheat transport effects.

Considering instrument uncertainties only, anindicated RCS temperature of 350°F is sufficiently high to allow full RCSpressurization in accordance with Appendix G limitations.

Should anoverpressure event occur in these conditions, the pressurizer safetyvalves provide acceptable and redundant overpressure protection.

The Reference 9 analyses demonstrate that either one RCS relief valve orthe depressurized RCS and RCS vent can maintain RCS pressure belowlimits when only one ECCS centrifugal charging pump (in addition to theNCP) is actuated.

However, the LCO allows only one ECCS centrifugal charging pump OPERABLE and the NCP functional during the LTOPMODES. Since neither one RCS relief valve nor the RCS vent can handlethe pressure transient caused by accumulator injection, when RCStemperature is low, the LCO also requires accumulator isolation whenaccumulator pressure is greater than or equal to the maximum RCSpressure for the existing RCS cold leg temperature allowed in the PTLR.The isolated accumulators must have their discharge valves closed andthe valve power supply breakers fixed in their open positions.

Fracture mechanics analyses established the temperature of LTOPApplicability at 3680F.Wolf Creek -Unit 1B 3.4.12-5Revision 61 LTOP SystemB 3.4.12BASESAPPLICABLE SAFETY ANALYSES(continued)

PORV Performance The fracture mechanics analyses show that the vessel is protected whenthe PORVs are set to open at or below the limit shown in the PTLR. Thesetpoints are derived by analyses that model the performance of theLTOP System, assuming the mass injection transient of one ECCScentrifugal charging pump and the NCP injecting into the RCS and theheat injection transient of starting an RCP with the RCS 50°F colder thanthe secondary coolant.

These analyses consider pressure overshoot andundershoot beyond the PORV opening and closing, resulting from signalprocessing and valve stroke times. The PORV setpoints at or below thederived limit ensures the Reference 1 P/T limits will be met.The PORV setpoints in the PTLR will be updated when the revised P/Tlimits conflict with the LTOP analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due toneutron embrittlement caused by neutron irradiation.

Revised limits aredetermined using neutron fluence projections and the results ofexaminations of the reactor vessel material irradiation surveillance specimens.

The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"

discuss these examinations.

The PORVs are considered active components.

Thus, the failure of onePORV is assumed to represent the worst case, single active failure.RHR Suction Relief Valve Performance The RHR suction relief valves do not have variable pressure andtemperature lift setpoints like the PORVs. Analyses show that one RHRsuction relief valve with a setpoint at or between 436.5 psig and463.5 psig will pass flow greater than that required for the limiting LTOPtransient while maintaining RCS pressure less than the P/T limit curve.As the RCS P/T limits are decreased to reflect the loss of toughness in thereactor vessel materials due to neutron embrittlement, the RHR suctionrelief valves must be analyzed to still accommodate the design basistransients for LTOP.The RHR suction relief valves are considered active components.

Thus,the failure of one valve is assumed to represent the worst case singleactive failure.Wolf Creek -Unit 1B 3.4.12-6Revision 56 LTOP SystemB 3.4.12BASESAPPLICABLE RCS Vent Performance SAFETY ANALYSIS(continued)

With the RCS depressurized, analyses show a vent size of 2.0 squareinches is capable of mitigating the limiting LTOP transient.

The capacityof a vent this size is greater than the flow of the limiting transient for theLTOP configuration, one ECCS centrifugal charging pump and the NCPinjecting into the RCS, maintaining RCS pressure less than the maximumpressure on the P/T limit curve.The RCS vent size will be re-evaluated for compliance each time the P/Tlimit curves are revised based on the results of the vessel materialsurveillance.

The RCS vent is passive and is not subject to active failure.The LTOP System satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO requires that the LTOP System is OPERABLE.

The LTOPSystem is OPERABLE when the maximum coolant input or heat inputbounded by that assumed in the analyses and required pressure reliefcapabilities are OPERABLE.

Violation of this LCO could lead to the lossof low temperature overpressure mitigation and violation of theReference 1 limits as a result of an operational transient.

To limit the coolant input capability, the LCO requires that a maximum ofzero safety injection pumps, one ECCS centrifugal charging pump and theNCP be capable of injecting into the RCS, and all accumulator discharge isolation valves be closed and immobilized (when accumulator pressure isgreater than or equal to the maximum RCS pressure for the existing RCScold leg temperature allowed in the PTLR).The LCO is modified by four Notes. Note 1 allows two ECCS centrifugal charging pumps to be made capable of injecting into the RCS for < 1 hourfor pump swap operations.

One hour provides sufficient time to safelycomplete the actual transfer and to complete the administrative controlsand surveillance requirements associated with the swap. The intent is tominimize the actual time that more than one ECCS centrifugal chargingpump is physically capable of injection.

This is accomplished by rackingout the breaker for one pump or employing two independent means toprevent a pump start in accordance with SR 3.4.12.2.

Note 2 recognizes the Applicability overlap between LCO's 3.4.12 and3.5.2 and states that two safety injection pumps and two ECCS centrifugal charging pumps may be made capable of injecting into the RCS:Wolf Creek -Unit 1B 3.4.12-7Revision 61 LTOP SystemB 3.4.12BASESLCO (a) In MODE 3 with any RCS cold leg temperature

< 3680F and ECCS(continued) pumps OPERABLE pursuant to LCO 3.5.2, "ECCS-Operating",

and(b) For up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 4 from MODE 3 or thetemperature of one or more RCS cold legs decreases below 325°F,whichever comes first.Note 3 states that one or more safety injection pumps may be madecapable of injecting into the RCS in MODES 5 and 6 when the RCS waterlevel is below the top of the reactor vessel flange for the purpose ofprotecting the decay heat removal function.

Note 4 states that the accumulator may be unisolated when theaccumulator pressure is less than the maximum RCS pressure for theexisting RCS cold leg temperature as allowed by the P/T limit curvesprovided in the PTLR. The accumulator discharge isolation valveSurveillance is not required under these pressure and temperature conditions.

The elements of the LCO that provide low temperature overpressure mitigation through pressure relief are:a. Two OPERABLE PORVs; orA PORV is OPERABLE for LTOP when its block valve is open, itslift setpoint is set to the limit required by the PTLR and testingproves its ability to open at this setpoint, and motive power isavailable to the two valves and their control circuits.

b. Two OPERABLE RHR suction relief valves; orAn RHR suction relief valve is OPERABLE for LTOP when its RHRsuction isolation valves are open, its setpoint is at or between436.5 psig and 463.5 psig, and testing has proven its ability to openat this setpoint.

c. One OPERABLE PORV and one OPERABLE RHR suction reliefvalve; ord. A depressurized RCS and an RCS vent.An RCS vent is OPERABLE when open with an area of _2.0 square inches.Wolf Creek -Unit 1B 3.4.12-8Revision 1

LTOP SystemB 3.4.12BASESACTIONS G.1(continued)

The RCS must be depressurized and a vent must be established within8 hours when:a. Both required RCS relief valves are inoperable; orb. A Required Action and associated Completion Time of Condition A,B, D, E, or F is not met; orc. The LTOP System is inoperable for any reason other thanCondition A, B, C, D, E, or F.The vent must be sized >_ 2.0 square inches to ensure that the flowcapacity is greater than that required for the worst case mass inputtransient reasonable during the applicable MODES. This action is neededto protect the RCPB from a low temperature overpressure event and apossible brittle failure of the reactor vessel.The Completion Time considers the time required to place the plant in thisCondition and the relatively low probability of an overpressure eventduring this time period due to increased operator awareness ofadministrative control requirements.

SURVEILLANCE SR 3.4.12.1, SR 3.4.12.2.

and SR 3.4.12.3REQUIREMENTS To minimize the potential for a low temperature overpressure event bylimiting the mass input capability, a maximum of zero safety injection pumps, one ECCS centrifugal charging pump and the NCP are verified tobe capable of injecting into the RCS and the accumulator discharge isolation valves are verified closed and with power removed from the valveoperator.

Verification that each accumulator is isolated is only required whenaccumulator pressure is greater than or equal to the maximum RCSpressure for the existing RCS cold leg temperature allowed by the P/Tlimit curves provided in the PTLR.The safety injection pumps and one ECCS centrifugal charging pump arerendered incapable of injecting into the RCS through removing the powerfrom the pumps by racking the breakers out under administrative control.An alternate method of cold overpressure protection may be employedusing at least two independent means to render a pump incapable ofinjecting into the RCS such that a single failure or single action will notWolf Creek -Unit 1B 3.4.12-11 Revision 61 LTOP SystemB 3.4.12BASESSURVEILLANCE SR 3.4.12.1, SR 3.4.12.2, and SR 3.4.12.3 (continued)

REQUIREMENTS result in an injection into the RCS. This may be accomplished by placingthe pump control switch in pull to lock and closing at least one valve in thedischarge flow path, or by closing at least one valve in the discharge flowpath and removing power from the valve operator, or by closing at leastone manual valve in the discharge flow path under administrative control.Providing pumps are rendered incapable of injecting into the RCS, theymay be energized for purposes such as testing or for filling accumulators.

The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering other indications andalarms available to the operator in the control room, to verify the requiredstatus of the equipment.

SR 3.4.12.4Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction isolation valves are open and bytesting it in accordance with the Inservice Testing Program.

ThisSurveillance is only required to be performed if the RHR suction reliefvalve is being used to meet this LCO.The RHR suction isolation valves are verified to be opened every 72hours. The Frequency is considered adequate in view of otheradministrative controls such as valve status indications available to theoperator in the control room that verify the RHR suction isolation valvesremain open.The ASME Code (Ref. 8), test per Inservice Testing Program verifiesOPERABILITY by proving proper relief valve mechanical motion and bymeasuring and, if required, adjusting the lift setpoint.

SR 3.4.12.5The RCS vent of >_ 2.0 square inches is proven OPERABLE by verifying its open condition either:a. Once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for a valve that is not locked, sealed, orotherwise secured in the open position.

Wolf Creek -Unit 1B 3.4.12-12 Revision 32 RCS Leakage Detection Instrumentation B 3.4.15BASESACTIONS A.1 and A.2 (continued)

With the required Containment Sump Level and Flow Monitoring Systeminoperable, no other form of sampling can provide the equivalent information;

however, the containment atmosphere particulate radioactivity monitor will provide indications of changes in leakage.Together with the atmosphere

monitor, the periodic surveillance for RCSwater inventory

balance, SR 3.4.13.1, must be performed at an increased frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to provide information that is adequate to detectleakage.

A Note is added allowing that SR 3.4.13.1 is not required to beperformed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation (stableRCS pressure, temperature, power level, pressurizer and makeup tanklevel, makeup and letdown, and RCP seal injection and return flows). The12 hour allowance provides sufficient time to collect and process allnecessary data after stable plant conditions are established.

Restoration of the required Containment Sump Level and Flow Monitoring System to OPERABLE status within a Completion Time of 30 days isrequired to regain the function after the system's failure.

This time isacceptable, considering the Frequency and adequacy of the RCS waterinventory balance required by Required Action A.1. The Completion Timeis modified by a Note indicating that the 30 days is extended until startupfrom a plant shutdown or startup from Refueling Outage 20.B.1.1, B.1.2, B.2.1 and B.2.2With the containment atmosphere particulate radioactivity monitoring instrumentation channel inoperable, alternative action is required.

Eithersamples of the containment atmosphere must be taken and analyzed forparticulate radioactivity or water inventory

balances, in accordance withSR 3.4.13.1, must be performed to provide alternate periodic information.

Alternatively, continued operation is allowed if the containment air coolercondensate monitoring system is OPERABLE, provided grab samples aretaken or water inventory balances are performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.With a sample obtained and analyzed or water inventory balanceperformed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor may be operated for up to 30 daysto allow restoration of the required containment atmosphere particulate radioactivity monitor.The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval provides periodic information that is adequate todetect leakage.

A Note is added allowing that SR 3.4.13.1 is not requiredto be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation (stable RCS pressure, temperature, power level, pressurizer and makeuptank level, makeup and letdown, and RCP seal injection and return flows).The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process allnecessary data after stable plant conditions are established.

The 30 dayCompletion Time recognizes at least one other form of leakage detection is available.

Wolf Creek -Unit 1B 3.4.15-5Revision 65 RCS Leakage Detection Instrumentation B 3.4.15BASESACTIONS C.1 and C.2(continued)

With the required containment cooler condensate monitoring systeminoperable, alternative action is again required.

Either SR 3.4.15.1 mustbe performed or water inventory

balances, in accordance with SR3.4.13.1, must be performed to provide alternate periodic information.

Provided a CHANNEL CHECK is performed every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or a waterinventory balance is performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, reactor operation maycontinue while awaiting restoration of the containment cooler condensate monitoring system to OPERABLE status.The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval provides periodic information that is adequate to detectRCS LEAKAGE.

A Note is added allowing that SR 3.4.13.1 is not required tobe performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation (stableRCS pressure, temperature, power level, pressurizer and makeup tank level,makeup and letdown, and RCP seal injection and return flows.) The 12 hourallowance provides sufficient time to collect and process all necessary dataafter stable plant conditions are established.

D.1 and D.2With the required containment atmosphere particulate radioactivity monitor and the required Containment Cooler Condensate Monitoring System inoperable, the means of detecting leakage is the Containment Sump Level and Flow Monitoring System. This Condition does notprovide all the required diverse means of leakage detection.

TheRequired Action is to restore either of the inoperable required monitoring methods to OPERABLE status within 30 days to regain the intendedleakage detection diversity.

The 30 day Completion Time ensures that theplant will not be operated in a reduced configuration for a lengthy timeperiod.Refer to LCO 3.3.6, "Containment Purge Isolation Instrumentation,"

upona loss of the required containment atmosphere radioactivity monitor toensure LCO requirements are met.E.1 and E.2If a Required Action of Condition A, B, C or D cannot be met, the plantmust be brought to a MODE in which the requirement does not apply. Toachieve this status, the plant must be brought to at least MODE 3 within6 hours and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Timesare reasonable, based on operating experience, to reach the requiredplant conditions from full power conditions in an orderly manner andwithout challenging plant systems.Wolf Creek -Unit 1B 3.4.15-6Revision 31 Containment Spray and Cooling SystemsB 3.6.6B 3.6 CONTAINMENT SYSTEMSB 3.6.6 Containment Spray and Cooling SystemsBASESBACKGROUND The Containment Spray and Containment Cooling system providescontainment atmosphere cooling to limit post accident pressure andtemperature in containment to less than the design values. Reduction ofcontainment pressure and the iodine removal capability of the sprayreduces the release of fission product radioactivity from containment tothe environment, in the event of a Design Basis Accident (DBA), to withinlimits. The Containment Spray and Containment Cooling system isdesigned to meet the requirements of 10 CFR 50, Appendix A, GDC 38,"Containment Heat Removal,"

GDC 39, "Inspection of Containment HeatRemoval Systems,"

GDC 40, "Testing of Containment Heat RemovalSystems,"

GDC 41, "Containment Atmosphere Cleanup,"

GDC 42,"Inspection of Containment Atmosphere Cleanup Systems,"

and GDC 43,"Testing of Containment Atmosphere Cleanup Systems,"

and GDC 50,"Containment Design Bases" (Ref. 1).The Containment Cooling System and Containment Spray System areEngineered Safety Feature (ESF) systems.

They are designed to ensurethat the heat removal capability required during the post accident periodcan be attained.

The Containment Spray System and the Containment Cooling System provides a redundant method to limit and maintain postaccident conditions to less than the containment design values.Containment Spray SystemThe Containment Spray System consists of two separate trains of equalcapacity, each capable of meeting the design bases. Each train includesa containment spray pump, spray headers,

nozzles, valves, and piping.Each train is powered from a separate ESF bus. The refueling waterstorage tank (RWST) supplies borated water to the Containment SpraySystem during the injection phase of operation.

In the recirculation modeof operation, containment spray pump suction is transferred from theRWST to the containment recirculation sumps.The Containment Spray System provides a spray of borated water mixedwith sodium hydroxide (NaOH) from the Spray Additive System into theupper regions of containment to reduce the containment pressure andtemperature and to reduce fission products from the containment atmosphere during a DBA. The RWST solution temperature is animportant factor in determining the heat removal capability of theContainment Spray System during the injection phase. In therecirculation mode of operation, heat is removed from the containment recirculation sump water by the residual heat removal heat exchangers.

Each train of the Containment Spray System provides adequate sprayIWolf Creek- Unit 1B 3.6.6-1Revision 42 Containment Spray and Cooling SystemsB 3.6.6BASESBACKGROUND Containment Spray System (continued) coverage to meet the system design requirements for containment heatremoval.

The Spray Additive System injects an NaOH solution into thespray. The resulting alkaline pH of the spray enhances the ability of thespray to scavenge fission products from the containment atmosphere.

The NaOH added in the spray also ensures an alkaline pH for the solutionrecirculated in the containment recirculation sump. The alkaline pH of thecontainment sump water minimizes the evolution of iodine and minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.The Containment Spray System is actuated either automatically by acontainment High-3 pressure signal or manually.

An automatic actuation opens the containment spray pump discharge valves, starts the twocontainment spray pumps and begins the injection phase. A manualactuation of the Containment Spray System requires the operator tosimultaneously actuate two separate switches on the main control boardto begin the same sequence.

The injection phase continues until anRWST level Low-Low alarm is received.

The Low-Low level alarm for theRWST signals the operator to manually align the system to therecirculation mode. The Containment Spray System in the recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of theContainment Spray System in the recirculation mode is controlled by theoperator in accordance with the emergency operating procedures.

Containment Cooling SystemTwo trains of containment

cooling, each of sufficient capacity to supply100% of the design cooling requirement, are provided.

Each train of twofan units is supplied with cooling water from a separate train of essential service water (ESW). Air is drawn into the coolers through the fan anddischarged to the steam generator compartments, pressurizer compartment, and instrument tunnel, and outside the secondary shield inthe lower areas of containment.

During normal operation, all four fan units are normally operating.

Thefans are normally operated at high speed with Service Water supplied tothe cooling coils. The Containment Cooling System, operating inconjunction with the Containment Ventilation and Air Conditioning

systems, is designed to limit the ambient containment air temperature during normal unit operation to less than the limit specified in LCO 3.6.5,"Containment Air Temperature."

This temperature limitation ensures thatthe containment temperature does not exceed the initial temperature conditions assumed for the DBAs.Wolf Creek -Unit 1B 3.6.6-2Revision 63 MFIVs and MFRVs and MFRV Bypass ValvesB 3.7.3BASESSURVEILLANCE SR 3.7.3.1REQUIREMENTS This SR verifies that the closure time of each MFIV, MFRV, and MFRVbypass valve is within limits (Figure B 3.7.3-1 for the MFIVs and < 15seconds for the MFRV and MFRV bypass valves) when tested pursuant tothe Inservice Testing Program.

The MFIV, MFRV, and MFRV bypassvalve closure time is assumed in the accident and containment analyses.

For the MFRVs, this Surveillance is normally performed upon returning theunit to operation following a refueling outage. The Surveillance may beperformed as required for post-maintenance testing of the MFRVs underappropriate conditions during applicable MODES. In particular, theMFRVs should normally not be tested at power since even a partial strokeexercise increases the risk of a valve closure with the unit generating power. However, when the plant is operating using the MFRV bypassvalves (at low power levels during MODE 1), the surveillance for theMFRVs may be performed for post-maintenance testing during suchconditions without increasing plant risk.For the MFRV bypass valves, this Surveillance is performed routinely during plant operation (or as required for post-maintenance testing),

but itmay also be required to be performed upon returning the unit to operation following a refueling outage.If it is necessary to adjust stem packing to stop packing leakage and if arequired stroke test is not practical in the current plant MODE, it should beshown by analysis that the packing adjustment is within torque limitsspecified by the manufacturer for the existing configuration of packing,and that the performance parameters of the valve are not adversely affected.

A confirmatory test must be performed at the first available opportunity when plant conditions allow testing.

Packing adjustments beyond the manufacturer's limits may not be performed without (1) anengineering analysis and (2) input from the manufacturer, unless tests canbe performed after the adjustments.

(Reference 3)The Frequency for this SR is in accordance with the Inservice TestingProgram.

Operating experience has shown that these components usually pass the Surveillance when performed at the Inservice TestingProgram Frequency.

This SR is modified by a Note that allows entry intoand operation in MODE 3 prior to performing the SR. This allows a delayof testing until MODE 3, to establish conditions consistent with thoseunder which the acceptance criterion was generated.

Test conditions arewith the unit at normal operating temperature and pressure, as discussed in Reference 2.Wolf Creek -Unit 1B 3.7.3-9Revision 66 MFIVs and MFRVs and MFRV Bypass ValvesB 3.7.3BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.7.3.2This SR verifies that each actuator train can close its respective MFIV onan actual or simulated actuation signal. The manual close hand switch inthe control room provides an acceptable actuation signal. ThisSurveillance is normally performed upon returning the plant to operation following a refueling outage in conjunction with SR 3.7.3.1.

However, it isacceptable to perform this Surveillance individually.

This SR is modifiedby a Note that allows entry into and operation in MODE 3 prior toperforming the SR. This allows a delay of testing until MODE 3, toestablish conditions consistent with those under which the acceptance criterion was generated The Frequency of MFIV testing is every 18 months. The 18 monthFrequency for testing is based on the refueling cycle. Operating experience has shown that these components usually pass theSurveillance when performed at the 18 month Frequency.

Therefore, thisFrequency is acceptable from a reliability standpoint.

SR 3.7.3.3This SR verifies that each MFRV and MFRV bypass valve is capable ofclosure on an actual or simulated actuation signal. The actuation ofsolenoids locally at the MFRVs and MFRV bypass valves constitutes anacceptable simulated actuation signal. This Surveillance is normallyperformed upon returning the unit to operation following a refueling outagein conjunction with SR 3.7.3.1.

However, it is acceptable to perform thisSurveillance individually.

The Frequency of MFRV and MFRV bypass valve testing is every 18months. The 18 month Frequency for testing is based on the refueling cycle. This Frequency is acceptable from a reliability standpoint.

This SRis modified by a Note that allows entry into and operation in MODE 3 priorto performing the SR. This allows a delay of testing until MODE 3, toestablish conditions consistent with those under which the acceptance criterion was generated.

REFERENCES

1. USAR, Section 10.4.7.2. ASME Code for Operation and Maintenance of Nuclear PowerPlants.3. NUREG-1482, Revision 1, "Guidelines for Inservice Testing atNuclear Power Plants."Wolf Creek -Unit 1B 3.7.3-10Revision 66 CREVSB 3.7.10B 3.7 PLANT SYSTEMSB 3.7.10 Control Room Emergency Ventilation System (CREVS)BASESBACKGROUND The CREVS provides a protected, controlled temperature environment from which occupants can control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke.The CREVS consists of two independent, redundant trains thatrecirculate, cool, pressurize, and filter the air in the control room envelope(CRE) and control building envelope (CBE) that limits the inleakage ofunfiltered air. Each CREVS train consists of a recirculation system trainand a pressurization system train. The air conditioning portion of eachtrain consists of a fan, a self-contained refrigeration system, and aprefilter.

The filtration portion of each system consists of a high efficiency particulate air (HEPA) filter, an activated charcoal absorber section forremoval of gaseous activity (principally iodines),

and a second HEPAfollows the absorber section to collect carbon fines. Each pressurization system train consists of ductwork to bring air from outside the building, amoisture separator, an electric heater, a HEPA, an activated charcoaladsorber, and a second HEPA. Ductwork, valves or dampers, doors,barriers, and instrumentation also form part of the system.The CREVS is an emergency system which may also operate duringnormal unit operations.

Upon receipt of the actuating signal, normal airsupply and exhaust to the CRE is isolated, and a portion of the ventilation air is recirculated through the filtration system train(s),

and thepressurization system is started.

The filtration system prefilters removeany large particles in the air, and the pressurization system moistureseparator removes any entrained water droplets

present, to preventexcessive loading of the HEPA filters and charcoal adsorbers.

Continuous operation of each pressurization train for at least 15 minutesper month, with the heaters functioning, reduces moisture buildup on theHEPA filters and adsorbers.

The heaters are important to theeffectiveness of the charcoal adsorbers.

Actuation of the CREVS by a Control Room Ventilation Isolation Signal(CRVIS),

places the system in the emergency mode of operation.

Actuation of the system to the emergency mode of operation closes theunfiltered outside air intake and unfiltered exhaust dampers, and alignsthe system for recirculation.

A portion of the recirculation of the air withinthe CRE flows through the redundant filtration system trains of HEPA andthe charcoal adsorbers.

The CRVIS also initiates pressurization andfiltered ventilation of the air supply to the CRE.Wolf Creek -Unit 1B 3.7.10-1Revision 64 CREVSB 3.7.10BASESBACKGROUND Outside air is filtered, diluted with air from the electrical equipment and(continued) cable spreading rooms, and added to the air being recirculated from theCRE. Pressurization of the CRE prevents infiltration of unfiltered air fromthe surrounding areas of the building.

The air entering the CBE during normal operation is continuously monitored by radiation and smoke detectors.

A high radiation signalinitiates the CRVIS; the smoke detectors provide an alarm in the controlroom. A CRVIS is initiated by the radiation monitors (GKRE0004 andGKRE0005),

fuel building ventilation isolation signal, containment isolation phase A, containment atmosphere radiation monitors (GTRE0031 andGTRE0032),

containment purge exhaust radiation monitors (GTRE0022 and GTRE0033),

or manually.

A single CREVS train operating in the CREVS alignment established bysurveillance procedures will pressurize the control room to >_ 0.25 incheswater gauge. The CREVS operation in maintaining the CRE habitable isdiscussed in the USAR, Section 6.4 and 9.4 (Ref. 1).Either of the pressurization and recirculation trains provide the requiredfiltration and pressurization to the CRE. Normally open isolation dampersare arranged in series pairs so that the failure of one damper to shut willnot result in a breach of isolation.

The CREVS is designed in accordance with Seismic Category I requirements.

The CREVS is designed to maintain a habitable environment in the CREfor 30 days of continuous occupancy after a Design Basis Accident (DBA)without exceeding a 5 rem whole body dose or its equivalent to any part ofthe body (Ref. 2).By operation of the control room pressurization trains and the control roomfiltration units, the CREVS pressurizes, recirculates and filters air withinthe CRE as well as the CBE that generally surrounds the CRE. Theboundaries of these two distinct but related volumes are credited in theanalysis of record for limiting the inleakage of unfiltered outside air.The station CRE design is unique. The Control Building by and largesurrounds the CRE. The Control Building is also designed to be at apositive pressure with respect to its surrounding environment although notpositive with respect to the CRE. In the emergency pressurization andfiltration mode, the control room air volume receives air through a filtration system that takes a suction on the Control Building.

The Control Buildingin turn receives filtered air from the outside environment.

Wolf Creek -Unit 1B 3.7.10-2Revision 41 CREVSB 3.7.10BASESACTIONS D.1, D.2.1, and D.2.2 (continued)

An alternative to Required Action D.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation ofthe CRE. This places the unit in a condition that minimizes the accidentrisk. This does not preclude the movement of fuel to a safe position.

E.1 and E.2During movement of irradiated fuel assemblies, with two CREVS trainsinoperable or with one or more CREVS trains inoperable due to aninoperable CRE or CBE boundary, action must be taken immediately tosuspend activities that could result in a release of radioactivity that mightrequire isolation of the CRE. This places the unit in a condition thatminimizes the accident risk. This does not preclude the movement of fuelto a safe position.

F.1If both CREVS trains are inoperable in MODE 1, 2, 3, or 4, for reasonsother than an inoperable CRE and CBE boundary (i.e., Condition B), theCREVS may not be capable of performing the intended function and theunit is in a condition outside the accident analyses.

Therefore, LCO 3.0.3must be entered immediately.

SURVEILLANCE SR 3.7.10.1REQUIREMENTS Standby systems should be checked periodically to ensure that theyfunction properly.

As the environment and normal operating conditions onthis system are not too severe, testing each train once every month, byinitiating from the control room, flow through the HEPA filters and charcoaladsorber of both the filtration and pressurization

systems, provides anadequate check of this system. Monthly heater operations dry out anymoisture accumulated in the charcoal from humidity in the ambient air.Each pressurization system train must be operated for > 15 continuous minutes with the heaters energized.

Each filtration system train need onlybe operated for _> 15 minutes continuously to demonstrate the function ofthe system. The 15-minute run time is based on Position C.6.1 ofReference

9. The 31 day Frequency is based on the reliability of theequipment and the two train redundancy.

Wolf Creek -Unit 1B 3.7.10-7Revision 64 CREVSB 3.7.10BASESSURVEILLANCE SR 3.7.10.2REQUIREMENTS (continued)

This SR verifies that the required CREVS testing is performed inaccordance with the Ventilation Filter Testing Program (VFTP). TheCREVS filter tests use the procedure guidance in Regulatory Guide 1.52,Rev. 2 (Ref. 3) in accordance with the VFTP. The VFTP includes testingthe performance of the HEPA filter, charcoal absorber efficiency, minimum flow rate, and the physical properties of the activated charcoal.

Specific test Frequencies and additional information are discussed indetail in the VFTP.SR 3.7.10.3This SR verifies that each CREVS train starts and operates on an actualor simulated CRVIS. The actuation signal includes Control RoomVentilation or High Gaseous Radioactivity.

The CREVS trainautomatically switches on an actual or simulated CRVIS into a CRVISmode of operation with flow through the HEPA filters and charcoaladsorber banks. The Frequency of 18 months is consistent with a typicaloperating cycle. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 monthFrequency.

Therefore, the Frequency is acceptable from a reliability standpoint.

SR 3.7.10.4This SR verifies the OPERABILITY of the CRE and CBE boundaries credited in the accident analysis by testing for unfiltered air inleakage pastthe credited envelope boundaries and into the CRE. The details of thetesting are specified in the Control Room Envelope Habitability Program.The CRE is considered habitable when the radiological dose to CREoccupants calculated in the licensing basis analyses of DBAconsequences is no more than 5 rem whole body or its equivalent to anypart of the body and the CRE occupants are protected from hazardous chemicals and smoke. For WCGS, there is no CREVS actuation forhazardous chemical releases or smoke and there are no Surveillance Requirements that verify OPERABILITY for hazardous chemicals orsmoke. This SR verifies that the unfiltered air inleakage into the CRE andCBE boundaries is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.

When unfiltered air inleakage isgreater than the assumed flow rate, Condition B must be entered.Required Action B.3 allows time to restore the CRE or CBEWolf Creek -Unit 1B 3.7.10-8Revision 41 CREVSB 3.7.10BASESSURVEILLANCE SR 3.7.10.4 (continued)

REQUIREMENTS boundary to OPERABLE status provided mitigating actions can ensurethat the CRE remains within the licensing basis habitability limits for theoccupants following an accident.

Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 4) which endorses, withexceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 5). Thesecompensatory measures may also be used as mitigating actions asrequired by Required Action B.2. Temporary analytical methods may alsobe used as compensatory measures to restore OPERABILITY (Ref. 6).Options for restoring the CRE or CBE boundary to OPERABLE statusinclude changing the licensing basis DBA consequence

analysis, repairing the boundary, or a combination of these actions.

Depending upon the nature of the problem and the corrective action, a full scopeinleakage test may not be necessary to establish that the envelopeboundary has been restored to OPERABLE status.REFERENCES

1. USAR, Section 6.4 and 9.4.2. USAR, Chapter 15, Appendix 15A.3. Regulatory Guide 1.52, Rev. 2.4. Regulatory Guide 1.196.5. NEI 99-03, "Control Room Habitability Assessment,"

June 2001.6. Letter from Eric J. Leeds (NRC) to James W. Davis (NEI) datedJanuary 30, 2004, "NEI Draft White Paper, Use of Generic Letter91-18 Process and Alternative Source Terms in the Context ofControl Room Habitability."

(ADAMS Accession No. ML040300694).

7. USAR Section 2.2.8. Regulatory Guide 1.78, Rev. 0.9. Regulatory Guide 1.52, Rev. 3.Wolf Creek -Unit 1B 3.7.10-9Revision 64 CRACSB 3.7.11BASESACTIONS C.1, C.2.1, and C.2.2 (continued) operation immediately.

This action ensures that the remaining train isOPERABLE, that no failures preventing automatic actuation will occur,and that active failures will be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might requireisolation of the control room. This places the unit in a condition thatminimizes accident risk. This does not preclude the movement of fuel to asafe position.

D.1 and D.2In MODE 5 or 6, or during movement of irradiated fuel assemblies, withtwo CRACS trains inoperable, action must be taken immediately tosuspend activities that could result in a release of radioactivity that mightrequire isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to asafe position.

E.1If both CRACS trains are inoperable in MODE 1, 2, 3, or 4, the CRACSmay not be capable of performing its intended function.

Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.11.1REQUIREMENTS Testing of the CRACS condenser heat exchangers under designconditions is impractical.

This SR verifies that the heat removal capability of the CRACS air conditioning units is adequate to remove the heat loadassumed in the control room during design basis accidents.

This SRconsists of verifying the heat removal capability of the condenser heatexchanger (either through performance testing or inspection),

ensuringthe proper operation of major components in the refrigeration cycle,verification of unit air flow capacity, and water flow measurement (Reference 2). The 18 month Frequency is appropriate since significant degradation of the CRACS is slow and is not expected over this timeperiod.Wolf Creek -Unit 1B 3.7.11-3Revision 63 CRACSB 3.7.11BASESREFERENCES

1. USAR, Section 9.4.1.2. NRC letter dated May 28, 2014, "Wolf Creek Generating Station -Interpretation of Technical Specification Surveillance Requirement 3.7.11.1, "Verify each CRACS train has the capability to remove theassumed heat load" (TAC NO. MF3665)."

Wolf Creek -Unit 1B 3.7.11-4Revision 63 EESB 3.7.13BASESACTIONS D.1 and D.2When Required Action A.1 cannot be completed within the associated Completion Time during movement of irradiated fuel assemblies in thefuel building, the OPERABLE Emergency Exhaust System train must bestarted in the FBVIS mode immediately or fuel movement suspended.

This action ensures that the remaining train is OPERABLE, that noundetected failures preventing system operation will occur, and that anyactive failure will be readily detected.

If the system is not placed in operation, this action requires suspension offuel movement, which precludes a fuel handling accident.

This does notpreclude the movement of fuel assemblies to a safe position.

E.. 1If the fuel building boundary is inoperable such that a train of theEmergency Exhaust System operating in the FBVIS mode cannotestablish or maintain the required negative

pressure, action must be takento restore an OPERABLE fuel building boundary within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24hour Completion Time is reasonable based on the low probability of aDBA occurring during this time period and the availability of theEmergency Exhaust System to provide a filtered release (albeit withpotential for some unfiltered fuel building leakage).

F.1During movement of irradiated fuel assemblies in the fuel building, whentwo trains of the Emergency Exhaust System are inoperable for reasonsother than an inoperable fuel building boundary (i.e., Condition E), or ifRequired Action E.1 cannot be completed within the associated Completion Time action must be taken to place the unit in a condition inwhich the LCO does not apply. Action must be taken immediately tosuspend movement of irradiated fuel assemblies in the fuel building.

Thisdoes not preclude the movement of fuel to a safe position.

SURVEILLANCE SR 3.7.13.1REQUIREMENTS Standby systems should be checked periodically to ensure that theyfunction properly.

As the environmental and normal operating conditions on this system are not severe, testing each train once every month, byinitiating from the control room flow through the HEPA filters and charcoaladsorbers, provides an adequate check on this system.Wolf Creek -Unit 1B 3.7.13-5Revision 57 EESB 3.7.13BASESSURVEILLANCE SR 3.7.13.1 (continued)

REQUIREMENTS Monthly heater operation dries out any moisture accumulated in thecharcoal from humidity in the ambient air. Systems with heaters must beoperated for _> 15 continuous minutes with the heaters energized.

Operating heaters would not necessarily have the heating elementsenergized continuously for 15 minutes, but will cycle depending on thetemperature.

The 31 day Frequency is based on the known reliability ofthe equipment and the two train redundancy available.

This SR can besatisfied with the Emergency Exhaust System in the SIS or FBVIS lineupduring testing.

The 15-minute run time is based on Position C.6.1 ofReference 10.SR 3.7.13.2This SR verifies that the required Emergency Exhaust System filter testingis performed in accordance with the Ventilation Filter Testing Program(VFTP). The Emergency Exhaust System filter tests are based on theguidance in References 6 and 7 in accordance with the VFTP. The VFTPincludes testing HEPA filter performance, charcoal absorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal.

Specific test frequencies and additional information arediscussed in detail in the VFTP.SR 3.7.13.3This SR verifies that each Emergency Exhaust System train starts andoperates on an actual or simulated actuation signal. The 18 monthFrequency is consistent with References 6 and 7. Proper completion ofthis SR requires testing the system in both the SIS (auxiliary buildingexhaust) and the FBVIS (fuel building exhaust) modes of operation.

During emergency operations the Emergency Exhaust System willautomatically start in either the SIS or FBVIS lineup depending on theinitiating signal. In the SIS lineup, the fans operate with dampers alignedto exhaust from the auxiliary building and prevent unfiltered leakage.

Inthis SIS lineup, each train is capable of maintaining the auxiliary buildingat a negative pressure at least 0.25 inches water gauge relative to theoutside atmosphere.

In the FBVIS lineup, which is initiated upondetection of high radioactivity by the fuel building exhaust gaseousradioactivity

monitors, the fans operate with the dampers aligned toexhaust from the fuel building to prevent unfiltered leakage.

In the FBVISlineup, each train is capable of maintaining the fuel building at a negativepressure at least 0.25 inches water gauge relative to the outsideatmosphere.

Normal exhaust air from the fuel building is continuously Wolf Creek -Unit 1B 3.7.13-6Revision 64 EESB 3.7.13BASESSURVEILLANCE SR 3.7.13.3 (continued)

REQUIREMENTS monitored by radiation detectors.

One detector output will automatically align the Emergency Exhaust System in the FBVIS mode of operation.

This surveillance requirement demonstrates that each Emergency Exhaust System unit can be automatically started and properly configured to the FBVIS or SIS alignment, as applicable, upon receipt of an actual orsimulated SIS signal and an FBVIS signal. It is not required that eachEmergency Exhaust System unit be started from both actuation signalsduring the same surveillance test provided each actuation signal is testedindependently within the 18 month test frequency.

SR 3.7.13.4This SR verifies the integrity of the auxiliary building enclosure.

Theability of the auxiliary building to maintain negative pressure with respectto potentially uncontaminated adjacent areas is periodically tested toverify proper function of the Emergency Exhaust System. During the SISmode of operation, the Emergency Exhaust System is designed tomaintain a slight negative pressure in the auxiliary

building, to preventunfiltered leakage.

The Emergency Exhaust System is designed tomaintain a negative pressure

>_ 0.25 inches water gauge with respect toatmospheric pressure at a flow rate specified in the VFTP. TheFrequency of 18 months is consistent with the guidance provided inNUREG-0800, Section 6.5.1 (Ref.8).An 18 month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 9.SR 3.7.13.5This SR verifies the integrity of the fuel building enclosure.

The ability ofthe fuel building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify properfunction of the Emergency Exhaust System. During the FBVIS mode ofoperation, the Emergency Exhaust System is designed to maintain aslight negative pressure in the fuel building, to prevent unfiltered leakage.The Emergency Exhaust System is designed to maintain a negativepressure

_> 0.25 inches water gauge with respect to atmospheric pressureat a flow rate specified in the VFTP. The Frequency of 18 months isconsistent with the guidance provided in NUREG-0800, Section 6.5.1(Ref.8).An 18 month Frequency (on a STAGGERED TEST BASIS) is consistent with Reference 9.Wolf Creek -Unit 1B 3.7.13-7Revision 64 1 EESB 3.7.13BASESREFERENCES 1.2.3.4.5.6.7.8.9.10.USAR, Section 6.5.1.USAR, Section 9.4.2 and 9.4.3.USAR, Section 15.7.4.Regulatory Guide 1.25, Rev. 0 (Safety Guide 25).10 CFR 100.ASTM D 3803-1989.

ANSI N510-1980.

NUREG-0800, Section 6.5.1, Rev. 2, July 1981.Regulatory Guide 1.52, Rev. 2.Regulatory Guide 1.52, Rev. 3.Wolf Creek -Unit IB 3.7.13-8Revision 64 SSIVsB 3.7.19BASESBACKROUND For each or any of the four feedwater lines, a positive displacement (continued) metering pump delivers the chemicals from a supply tank into theassociated feedwater line via an injection flow path that includes anautomatic air-operated globe isolation valve, a check valve, and a manualvalve prior to entering into the feedwater system.The Steam Generator Chemical Injection System is used to maintainproper system pH and scavenge oxygen present in the steam generators to minimize corrosion during plant shutdown conditions.

The system addshydrazine and amine mixture to the steam generator and is normally not inuse during plant power operation, except during plant conditions in hotstandby or cold layup. The Steam Generator Chemical Injection Systemis infrequently used during the Applicability of this Specification.

The manual valve located in each chemical injection flow path ismaintained locked closed until the system is used. When the system isused, the manual valve is opened under administrative controls.

Thecontrols include the presence of a dedicated operator who has constantcommunication with the control room while the flow path is open.Therefore, crediting the locked closed manual valve in the chemicalinjection flow path for isolation is warranted when it is only opened underadministrative controls.

The main steam and related secondary side lines are automatically isolated upon receipt of an SLIS or feedwater isolation signal (FWIS).The diverse parameters sensed to initiate an SLIS are low steam linepressure, high negative steam pressure rate, and high containment pressure (Hi-2).A FWIS is generated by a safety injection signal (SIS), reactor trip with lowTave, steam generator water level high-high, or steam generator waterlevel low-low.

The diverse parameters sensed to initiate an SIS are lowsteam line pressure, low pressurizer

pressure, and high containment pressure (Hi-I).The steam generator blowdown and sample isolation (AFAS) isolates thesteam generator blowdown and sample lines. A steam generator blowdown and sample isolation (AFAS) is generated by a SIS, motor-driven AFAS, or undervoltage on switchgear 4.16 kV buses NBO1 orNB02.Descriptions of SSIVs are found in the USAR, Section 10.4.7 (Ref. 1),Section 10.4.8 (Ref. 2), and Section 10.3 (Ref. 3).Wolf Creek -Unit 1B 3.7.19-3Revision 54 SSIVsB 3.7.19BASESAPPLICABLE The accident analysis assume that the steam generators are isolatedSAFETY ANALYSES after receiving an isolation signal as discussed in the Background section.Further discussion can be found in the USAR, Chapters 6 and 15.The SSIVs function to ensure the primary success path for steam line andfeed line isolation and for delivery of required auxiliary feedwater flow and,therefore, satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO provides assurance that SSIVs will isolate the plant's secondary side, following a main feed line or main steam line break and ensures therequired flow of auxiliary feedwater to the intact steam generators.

Theautomatic secondary system isolation valves are considered OPERABLEwhen their isolation times are within limits and they are capable of closingon an isolation actuation signal. OPERABILITY of the automatic SSIVsalso requires the OPERABILITY of the auxiliary relays downstream of theBalance of Plant Engineered Safety Features Actuation System (ESFAS)cabinets (the auxiliary relays in the system cabinets are considered to bepart of the end devices covered by this LCO).The locked closed manual valves in the chemical injection flow path areconsidered OPERABLE when they are locked closed. Locked closedmanual SSIVs include steam generator chemical injection isolation valves(AEV0128,

AEV0129, AEV0130, and AEV0131).

Automatic secondary system isolation valves include the SGBIVs(BMHV0001,

BMHV0002, BMHV0003, and BMHV0004) and the SGBSIVs(BMHV0019,

BMHV0020, BMHV0021,

BMHV0022, BMHV0065,

BMHV0066, BMHV0067,

BMHV0068, BMHV0035,

BMHV0036, BMHV0037, and BMHV0038),

and the main steam low point drainisolation valves (ABLVO07,

ABLVO08, ABLVO09, and ABLV01 0).APPLICABILITY The SSIVs must be OPERABLE in MODES 1, 2, and 3, when there issignificant mass and energy in the Reactor Coolant System (RCS) andsteam generators.

When the SSIVs are closed and de-activated, orclosed and isolated by a closed manual valve, or the flow path is isolatedby a combination of closed manual valve(s) and closed de-activated automatic valve(s),

they are performing the specified safety function ofisolating the plant's secondary side. The combination provides a meansof dual isolation that cannot be affected by a single active failure thusassuring the safety function is met. An air-operated SSIV is de-activated when power and air are removed from its actuation solenoid valves, and asolenoid-operated SSIV is de-activated when power is removed from itsassociated solenoid valve.In MODES 4, 5, and 6, the steam generator energy is low. Therefore, theSSIVs are not required for isolation of potential high energy secondary system pipe breaks in these MODES.Wolf Creek -Unit 1B 3.7.19-4Revision 61 SSIVsB 3.7.19BASESACTIONS The ACTIONS are modified by a Note to provide clarification that, for thisLCO, separate Condition entry is allowed for each SSIV. This isacceptable, since the Required Actions for each Condition provideappropriate compensatory actions for each inoperable SSIV. Complying with the Required Actions may allow for continued operation, andsubsequent inoperable SSIVs are governed by subsequent Condition entry and application of associated Required Actions.A second Note has been added to allow SSIVs to be unisolated intermittently under administrative controls.

These administrative controlsconsist of stationing a dedicated operator at the valve controls, who is incontinuous communication with the control room. In this way, the SSIVcan be rapidly isolated when the need for secondary system isolation isindicated.

A.1 and A.2With one or more SSIVs inoperable, action must be taken to restore theaffected valves to OPERABLE status, or to close or isolate inoperable valves within 7 days. When these valves are closed or isolated, they areperforming their specified safety function.

The 7 day Completion Time takes into account the low probability of anevent occurring during this time period that would require isolation of theplant's secondary side. The 7 day Completion Time is reasonable, basedon operating experience.

Inoperable SSIVs that are closed or isolated must be verified on a periodicbasis that they are closed or isolated.

This is necessary to ensure that theassumptions in the accident analyses remain valid. The 7 day Completion Time is reasonable based on engineering

judgment, in view of valvestatus indications in the control room, and other administrative

controls, toensure that these valves are in the closed position or isolated.

If the inoperable SSIV is both closed and de-activated, or both closed andisolated by a closed manual valve, or the affected SSIV flow path isisolated by two closed manual valves, or two closed de-activated automatic valves, or one closed manual valve in combination with oneclosed de-activated automatic valve, the LCO does not apply asdiscussed in the Applicability.

The combination provides a means of dualisolation that cannot be affected by a single active failure thus assuringthe safety function is met. For example, BMHV0065 is determined to beinoperable.

If BMHV0065 is closed or is open and isolated by BMV0009,then Required Action A.2 must be performed.

If BMHV0065 is closed andBMV009 is closed, then the LCO is considered met since BMHV0065does not meet the Applicability statement.

Wolf Creek -Unit 1B 3.7.19-5Revision 61 SSIVsB 3.7.19BASESACTIONS B.1 and B.2(continued)

If the Required Action and associated Completion Time of Condition A isnot met, the unit must be placed in a MODE in which the LCO does notapply. To achieve this status, the unit must be placed at least in MODE 3within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach therequired unit conditions in an orderly manner and without challenging unitsystems.SURVEILLANCE SR 3.7.19.1REQUIREMENTS This SR verifies the proper alignment for required automatic SSIVs in theflow path that are used to isolate the plant's secondary side. The SSIV isallowed to be in a nonaccident position provided the valve willautomatically reposition within the proper stroke time. This SR does notrequire any testing or valve manipulation.

Rather, it involves verification, through a system walkdown (which may include the use of local or remoteindicators),

that valves capable of being mispositioned are in the correctposition.

This SR does not apply to the locked closed manual valves inthe chemical injection flow path since these valves were verified to be inthe correct position upon locking.The 31 day Frequency is based on engineering

judgment, is consistent with the procedural controls governing valve operation, and ensurescorrect valve positions.

SR 3.7.19.2This SR verifies that the isolation time of each required automatic SSIV iswithin limits when tested pursuant to the Inservice Testing Program.

Thespecific limits are documented in the Inservice Testing Program.

TheSSIV isolation times are less than or equal to those assumed in theaccident and containment analyses.

The SR is performed only forrequired SSIVs. This Surveillance does not include verifying a closuretime for the steam generator chemical injection isolation valves. Anexception is made for the steam generator chemical addition injection isolation valves which are not included in the Inservice Testing Program.These valves are passive and contain a locking device and a check valvein their flow path.Wolf Creek -Unit IB 3.7.19-6Revision 54 RHR and Coolant Circulation

-Low Water LevelB 3.9.6B 3.9 REFUELING OPERATIONS B 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation

-Low Water LevelBASESBACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heatand sensible heat from the Reactor Coolant System (RCS), as requiredby GDC 34, to provide mixing of borated coolant, and to prevent boronstratification (Ref. 1). Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat istransferred to the Component Cooling Water System. The coolant isthen returned to the RCS via the RCS cold leg(s). Operation of theRHR System for normal cooldown decay heat removal is manuallyaccomplished from the control room. The heat removal rate is adjustedby controlling the flow of reactor coolant through the RHR heatexchanger(s) and the bypass lines. Mixing of the reactor coolant ismaintained by this continuous circulation of reactor coolant through theRHR System.APPLICABLE SAFETY ANALYSESIf the reactor coolant temperature is not maintained below 2000F, boilingof the reactor coolant could result. This could lead to a loss of coolant inthe reactor vessel. Additionally, boiling of the reactor coolant could leadto boron plating out on components near the areas of the boiling activity.

The loss of reactor coolant and the subsequent plate out of boron willeventually challenge the integrity of the fuel cladding, which is a fissionproduct barrier.

Two trains of the RHR System are required to beOPERABLE, and one train in operation, in order to prevent thischallenge.

Although the RHR System does not meet a specific criterion of the NRCPolicy Statement, it was identified in 10 CFR 50.36(c)(2)(ii) as animportant contributor to risk reduction.

Therefore, the RHR System isretained as a Specification.

In MODE 6, with the water level <23 ft above the top of the reactorLCOIn MODE 6, with the water level < 23 ft above the top of the reactorvessel flange, both RHR loops must be OPERABLE.

Additionally, one loop of RHR must be in operation in order to provide:a. Removal of decay heat;b. Mixing of borated coolant to minimize the possibility of criticality; andWolf Creek -Unit 1B 3.9.6-1Revision 0

RHR and Coolant Circulation

-Low Water LevelB 3.9.6BASESLCO(continued)

c. Indication of reactor coolant temperature.

An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flowpath and to determine the RCS temperature.

The flow path starts in oneof the RCS hot legs and is returned to the RCS cold legs. AnOPERABLE RHR loop must be capable of being realigned to provide anOPERABLE flow path.When both RHR loops (or trains) are required to be OPERABLE, theassociated Component Cooling Water (CCW) train is required to beOPERABLE.

The heat sink for the CCW System is normally provided bythe Service Water System or Essential Service Water (ESW) System, asdetermined by system availability.

In MODES 5 and 6, one DieselGenerator (DG) is required to be OPERABLE per LCO 3.8.2, "AC Sources-Shutdown."

The same ESW train is required to be capable ofperforming its related support function(s) to support DG OPERABILITY.

However, a Service Water train can be utilized to support CCW/RHROPERABILITY if the associated ESW train is not capable of performing itsrelated support function(s).

APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loopmust be in operation in MODE 6, with the water level < 23 ft above thetop of the reactor vessel flange, to provide decay heat removal.Requirements for the RHR System in other MODES are covered byLCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5,Emergency Core Cooling Systems (ECCS). RHR loop requirements inMODE 6 with the water level > 23 ft are located in LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation

-High Water Level."Since LCO 3.9.6 contains Required Actions with immediate Completion Times related to the restoration of the degraded decay heat removalfunction, it is not permitted to enter this LCO from either MODE 5 orfrom LCO 3.9.5, "RHR and Coolant Circulation

-High Water Level,"unless the requirements of LCO 3.9.6 are met. This precludes diminishing the backup decay heat removal capability when the RHRSystem is degraded.

ACTIONS A.1 and A.2If less than the required number of RHR loops are OPERABLE, actionshall be immediately initiated and continued until the RHR loop isrestored to OPERABLE status and to operation in accordance with theLCO or until __ 23 ft of water level is established above the reactorWolf Creek -Unit 1B 3.9.6-2Revision 63 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -Title Page Technical Specification Cover PageTitle PageTAB -Table of Contentsi 34 DRR 07-1057 7/10/07ii 29 DRR 06-1984 10/17/06iii 44 DRR 09-1744 10/28/09TAB -B 2.0 SAFETY LIMITS (SLs)B 2.1.1-1 0 Amend. No. 123 12/18/99B 2.1.1-2 14 DRR 03-0102 2/12/03B 2.1.1-3 14 DRR 03-0102 2/12/03B 2.1.1-4 0 Amend. No. 123 2/12/03B 2.1.2-1 0 Amend. No. 123 12/18/99B 2.1.2-2 12 DRR 02-1062 9/26/02B 2.1.2-3 0 Amend. No. 123 12/18/99TAB -B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILTY B 3.0-1 34 DRR 07-1057 7/10/07B 3.0-2 0 Amend. No. 123 12/18/99B 3.0-3 0 Amend. No. 123 12/18/99B 3.0-4 19 DRR 04-1414 10/12/04B 3.0-5 19 DRR 04-1414 10/12/04B 3.0-6 19 DRR 04-1414 10/12/04B 3.0-7 19 DRR 04-1414 10/12/04B 3.0-8 19 DRR 04-1414 10/12/04B 3.0-9 42 DRR 09-1009 7/16/09B 3.0-10 42 DRR 09-1009 7/16/09B 3.0-11 34 DRR 07-1057 7/10/07B 3.0-12 34 DRR 07-1057 7/10/07B 3.0-13 34 DRR 07-1057 7/10/07B 3.0-14 34 DRR 07-1057 7/10/07B 3.0-15 34 DRR 07-1057 7/10/07B 3.0-16 34 DRR 07-1057 7/10/07TAB -B 3.1B 3.1.1-1B 3.1.1-2B 3.1.1-3B 3.1.1-4B 3.1.1-5B 3.1.2-1B 3.1.2-2B 3.1.2-3B 3.1.2-4B 3.1.2-5B 3.1.3-1B 3.1.3-2B 3.1.3-3B 3.1.3-4REACTIVITY CONTROL SYSTEMS000190000000000Amend. No. 123Amend. No. 123Amend. No. 123DRR 04-1414Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 12312/18/9912/18/9912/18/9910/12/0412/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/99Wolf Creek -Unit 1Revsion 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.1B 3.1.3-5B 3.1.3-6B 3.1.4-1B 3.1.4-2B 3.1.4-3B 3.1.4-4B 3.1.4-5B 3.1.4-6B 3.1.4-7B 3.1.4-8B 3.1.4-9B 3.1.5-1B 3.1.5-2B 3.1.5-3B 3.1.5-4B 3.1.6-1B 3.1.6-2B 3.1.6-3B 3.1.6-4B 3.1.6-5B 3.1.6-6B 3.1.7-1B 3.1.7-2B 3.1.7-3B 3.1.7-4B 3.1.7-5B 3.1.7-6B 3.1.8-1B 3.1.8-2B 3.1.8-3B 3.1.8-4B 3.1.8-5B 3.1.8-6REACTIVITY CONTROL SYSTEMS0000480048000000000000000484848000151505(continued)

Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 10-3740Amend. No. 123Amend. No. 123DRR 10-3740Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 10-3740DRR 10-3740DRR 10-3740Amend. No. 123Amend. No. 123Amend. No. 123DRR 03-0860DRR 03-0860Amend. No. 123DRR 00-142712/18/9912/18/9912/18/9912/18/9912/28/1012/18/9912/18/9912/28/1012/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/28/1012/28/1012/28/1012/18/9912/18/9912/18/997/10/037/10/0312/18/9910/12/00TAB -B 3.2B 3.2.1-1B 3.2.1-2B 3.2.1-3B 3.2.1-4B 3.2.1-5B 3.2.1-6B 3.2.1-7B 3.2.1-8B 3.2.1-9B 3.2.1-10B 3.2.2-1B 3.2.2-2B 3.2.2-3B 3.2.2-4B 3.2.2-5B 3.2.2-6POWER DISTRIBUTION LIMITS480484848484848294848048484848DRR 10-3740Amend. No. 123DRR 10-3740DRR 10-3740DRR 10-3740DRR 10-3740DRR 10-3740DRR 10-3740DRR 06-1984DRR 10-3740DRR 10-3740Amend. No. 123DRR 10-3740DRR 10-3740DRR 10-3740DRR 10-374012/28/1012/18/9912/28/1012/28/1012/28/1012/28/1012/28/1012/28/1010/17/0612/28/1012/28/1012/18/9912/28/1012/28/1012/28/1012/28/10Wolf Creek -Unit IiiRevision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.2 POWER DISTRIBUTION LIMITS (continued)

B 3.2.3-1 0 Amend. No. 123 12/18/99B 3.2.3-2 0 Amend. No. 123 12/18/99B 3.2.3-3 0 Amend. No. 123 12/18/99B 3.2.4-1 0 Amend. No. 123 12/18/99B 3.2.4-2 0 Amend. No. 123 12/18/99B 3.2.4-3 48 DRR 10-3740 12/28/10B 3.2.4-4 0 Amend. No. 123 12/18/99B 3.2.4-5 48 DRR 10-3740 12/28/10B 3.2.4-6 0 Amend. No. 123 12/18/99B 3.2.4-7 48 DRR 10-3740 12/28/10TAB -B 3.3 INSTRUMENTATION B 3.3.1-1 0B 3.3.1-2 0B 3.3.1-3 0B 3.3.1-4 0B 3.3.1-5 0B 3.3.1-6 0B 3.3.1-7 5B 3.3.1-8 0B 3.3.1-9 0B 3.3.1-10 29B 3.3.1-11 0B 3.3.1-12 0B 3.3.1-13 0B 3.3.1-14 0B 3.3.1-15 0B 3.3.1-16 0B 3.3.1-17 0B 3.3.1-18 0B 3.3.1-19 66B 3.3.1-20 66B 3.3.1-21 0B 3.3.1-22 0B 3.3.1-23 9B 3.3.1-24 0B 3.3.1-25 0B 3.3.1-26 0B 3.3.1-27 0B 3.3.1-28 2B 3.3.1-29 1B 3.3.1-30 1B 3.3.1-31 0B 3.3.1-32 20B 3.3.1-33 48B 3.3.1-34 20B 3.3.1-35 19B 3.3.1-36 20B 3.3.1-37 20B 3.3.1-38 20B 3.3.1-39 25Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 00-1427Amend. No. 123Amend. No. 123DRR 06-1984Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 14-2329DRR 14-2329Amend. No. 123Amend. No. 123DRR 02-0123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 00-0147DRR 99-1624DRR 99-1624Amend. No. 123DRR 04-1533DRR 10-3740DRR 04-1533DRR 04-1414DRR 04-1533DRR 04-1533DRR 04-1533DRR 06-080012/18/9912/18/9912/18/9912/18/9912/18/9912/18/9910/12/0012/18/9912/18/9910/17/0612/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9911/6/1411/6/1412/18/9912/18/992/28/0212/18/9912/18/9912/18/9912/18/994/24/0012/18/9912/18/9912/18/992/16/0512/28/102/16/0510/13/042/16/052/16/052/16/055/18/06Wolf Creek -Unit 1iiiRe~vision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.3 INSTRUMENTATION (continued)

B 3.3.1-40 20B 3.3.1-41 20B 3.3.1-42 20B 3.3.1-43 20B 3.3.1-44 20B 3.3.1-45 20B 3.3.1-46 48B 3.3.1-47 20B 3.3.1-48 48B 3.3.1-49 20B 3.3.1-50 20B 3.3.1-51 21B 3.3.1-52 20B 3.3.1-53 20B 3.3.1-54 20B 3.3.1-55 25B 3.3.1-56 66B 3.3.1-57 20B 3.3.1-58 29B 3.3.1-59 20B 3.3.2-1 0B 3.3.2-2 0B 3.3.2-3 0B 3.3.2-4 0B 3.3.2-5 0B 3.3.2-6 7B 3.3.2-7 0B 3.3.2-8 0B 3.3.2-9 0B 3.3.2-10 0B 3.3.2-11 0B 3.3.2-12 0B 3.3.2-13 0B 3.3.2-14 2B 3.3.2-15 0B 3.3.2-16 0B 3.3.2-17 0B 3.3.2-18 0B 3.3.2-19 37B 3.3.2-20 37B 3.3.2-21 37B 3.3.2-22 37B 3.3.2-23 37B 3.3.2-24 39B 3.3.2-25 39B 3.3.2-26 39B 3.3.2-27 37B 3.3.2-28 37B 3.3.2-29 0B 3.3.2-30 0B 3.3.2-31 52DRR 04-1533DRR 04-1533DRR 04-1533DRR 04-1533DRR 04-1533DRR 04-1533DRR 10-3740DRR 04-1533DRR 10-3740DRR 04-1533DRR 04-1533DRR 05-0707DRR 04-1533DRR 04-1533DRR 04-1533DRR 06-0800DRR 14-2329DRR 04-1533DRR 06-1984DRR 04-1533Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 01-0474Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 00-0147Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 08-0503DRR 08-0503DRR 08-0503DRR 08-0503DRR 08-0503DRR 08-1096DRR 08-1096DRR 08-1096DRR 08-0503DRR 08-0503Amend. No. 123Amend. No. 123DRR 11-07242/16/052/16/052/16/052/16/052/16/052/16/0512/28/102/16/0512/28/102/16/052/16/054/20/052/16/052/16/052/16/055/18/0611/6/142/16/0510/17/062/16/0512/18/9912/18/9912/18/9912/18/9912/18/995/1/0112/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/994/24/0012/18/9912/18/9912/18/9912/18/994/8/084/8/084/8/084/8/084/8/088/28/088/28/088/28/084/8/084/8/0812/18/9912/18/994/11/11Wolf Creek -Unit 1ivRevision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.3 INSTRUMENTATION (continued)

B 3.3.2-32 52B 3.3.2-33 0B 3.3.2-34 0B 3.3.2-35 20B 3.3.2-36 20B 3.3.2-37 20B 3.3.2-38 20B 3.3.2-39 25B 3.3.2-40 20B 3.3.2-41 45B 3.3.2-42 45B 3.3.2-43 20B 3.3.2-44 20B 3.3.2-45 20B 3.3.2-46 54B 3.3.2-47 43B 3.3.2-48 37B 3.3.2-49 20B 3.3.2-50 20B 3.3.2-51 43B 3.3.2-52 43B 3.3.2-53 43B 3.3.2-54 43B 3.3.2-55 43B 3.3.2-56 43B 3.3.2-57 43B 3.3.3-1 0B 3.3.3-2 5B 3.3.3-3 0B 3.3.3-4 0B 3.3.3-5 0B 3.3.3-6 8B 3.3.3-7 21B 3.3.3-8 8B 3.3.3-9 8B 3.3.3-10 19B 3.3.3-11 19B 3.3.3-12 21B 3.3.3-13 21B 3.3.3-14 8B 3.3.3-15 8B 3.3.4-1 0B 3.3.4-2 9B 3.3.4-3 15B 3.3.4-4 19B 3.3.4-5 1B 3.3.4-6 9B 3.3.5-1 0B 3.3.5-2 1B 3.3.5-3 1DRR 11-0724Amend. No. 123Amend. No. 123DRR 04-1533DRR 04-1533DRR 04-1533DRR 04-1533DRR 06-0800DRR 04-1533Amend. No. 187 (ETS)Amend. No. 187 (ETS)DRR 04-1533DRR 04-1533DRR 04-1533DRR 11-2394DRR 09-1416DRR 08-0503DRR 04-1533DRR 04-1533DRR 09-1416DRR 09-1416DRR 09-1416DRR 09-1416DRR 09-1416DRR 09-1416DRR 09-1416Amend. No. 123DRR 00-1427Amend. No. 123Amend. No. 123Amend. No. 123DRR 01 -1235DRR 05-0707DRR 01 -1235DRR 01-1235DRR 04-1414DRR 04-1414DRR 05-0707DRR 05-0707DRR 01-1235DRR 01-1235Amend. No. 123DRR 02-1023DRR 03-0860DRR 04-1414DRR 99-1624DRR 02-0123Amend. No. 123DRR 99-1624DRR 99-16244/11/1112/18/9912/18/992/16/052/16/052/16/052/16/055/18/062/16/053/5/103/5/102/16/052/16/052/16/0511/16/119/2/094/8/082/16/052/16/059/2/099/2/099/2/099/2/099/2/099/2/099/2/0912/18/9910/12/0012/18/9912/18/9912/18/999/19/014/20/059/19/019/19/0110/12/0410/12/044/20/054/20/059/19/019/19/0112/18/992/28/027/10/0310/12/0412/18/992/28/0212/18/9912/18/9912/18/99Wolf Creek -Unit 1VRevision66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.3 INSTRUMENTATION (continued)

B 3.3.5-4 1 DRR 99-1624 12/18/99B 3.3.5-5 0 Amend. No. 123 12/18/99B 3.3.5-6 22 DRR 05-1375 6/28/05B 3.3.5-7 22 DRR 05-1375 6/28/05B 3.3.6-1 0 Amend. No. 123 12/18/99B 3.3.6-2 0 Amend. No. 123 12/18/99B 3.3.6-3 0 Amend. No. 123 12/18/99B 3.3.6-4 0 Amend. No. 123 12/18/99B 3.3.6-5 0 Amend. No. 123 12/18/99B 3.3.6-6 0 Amend. No. 123 12/18/99B 3.3.6-7 0 Amend. No. 123 12/18/99B 3.3.7-1 0 Amend. No. 123 12/18/99B 3.3.7-2 57 DRR 13-0006 1/16/13B 3.3.7-3 57 DRR 13-0006 1/16/13B 3.3.7-4 0 Amend. No. 123 12/18/99B 3.3.7-5 0 Amend. No. 123 12/18/99B 3.3.7-6 57 DRR 13-0006 1/16/13B 3.3.7-7 0 Amend. No. 123 12/18/99B 3.3.7-8 0 Amend. No. 123 12/18/99B 3.3.8-1 0 Amend. No. 123 12/18/99B 3.3.8-2 0 Amend. No. 123 12/18/99B 3.3.8-3 57 DRR 13-0006 1/16/13B 3.3.8-4 57 DRR 13-0006 1/16/13B 3.3.8-5 0 Amend. No. 123 12/18/99B 3.3.8-6 24 DRR 06-0051 2/28/06B 3.3.8-7 0 Amend. No. 123 12/18/99TAB -B 3.4B 3.4.1-1B 3.4.1-2B 3.4.1-3B 3.4.1-4B 3.4.1-5B 3.4.1-6B 3.4.2-1B 3.4.2-2B 3.4.2-3B 3.4.3-1B 3.4.3-2B 3.4.3-3B 3.4.3-4B 3.4.3-5B 3.4.3-6B 3.4.3-7B 3.4.4-1B 3.4.4-2B 3.4.4-3B 3.4.5-1B 3.4.5-2B 3.4.5-3B 3.4.5-4REACTOR COOLANT SYSTEM (RCS)0101000000000000000290053290Amend. No. 123DRR 02-0411DRR 02-0411Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 06-1984Amend. No. 123Amend. No. 123DRR 11-1513DRR 06-1984Amend. No. 123.12/18/99 4/5/024/5/0212/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9910/17/0612/18/9912/18/997/18/1110/17/0612/18/99Wolf Creek -Unit IviRevtision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.5-5 12B 3.4.5-6 12B 3.4.6-1 53B 3.4.6-2 29B 3.4.6-3 12B 3.4.6-4 12B 3.4.6-5 12B 3.4.7-1 12B 3.4.7-2 17B 3.4.7-3 63B 3.4.7-4 42B 3.4.7-5 12B 3.4.8-1 53B 3.4.8-2 62B 3.4.8-3 42B 3.4.8-4 42B 3.4.9-1 0B 3.4.9-2 0B 3.4.9-3 0B 3.4.9-4 0B 3.4.10-1 5B 3.4.10-2 5B 3.4.10-3 0B 3.4.10-4 32B 3.4.11-1 0B 3.4.11-2 1B 3.4.11-3 19B 3.4.11-4 0B 3.4.11-5 1B 3.4.11-6 0B 3.4.11-7 32B 3.4.12-1 61B 3.4.12-2 61B 3.4.12-3 0B 3.4.12-4 61B 3.4.12-5 61B 3.4.12-6 56B 3.4.12-7 61B 3.4.12-8 1B 3.4.12-9 56B 3.4.12-10 0B 3.4.12-11 61B 3.4.12-12 32'B 3.4.12-13 0B 3.4.12-14 32B 3.4.13-1 0B 3.4.13-2 29B 3.4.13-3 29B 3.4.13-4 35B 3.4.13-5 35B 3.4.13-6 29(continued)

DRR 02-1062DRR 02-1062DRR 11-1513DRR 06-1984DRR 02-1062DRR 02-1062DRR 02-1062DRR 02-1062DRR 04-0453DRR 14-1572DRR 09-1009DRR 02-1062DRR 11-1513DRR 14-1103DRR 09-1009DRR 09-1009Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 00-1427DRR 00-1427Amend. No. 123DRR 07-0139Amend. No. 123DRR 99-1624DRR 04-1414Amend. No. 123DRR 99-1624Amend. No. 123DRR 07-0139DRR 14-0346DRR 14-0346Amend. No. 123DRR 14-0346DRR 14-0346DRR 12-1792DRR 14-0346DRR 99-1624DRR 12-1792Amend. No. 123DRR 14-0346DRR 07-0139Amend. No. 123DRR 07-0139Amend. No. 123DRR 06-1984DRR 06-1984DRR 07-1553DRR 07-1553DRR 06-19849/26/029/26/027/18/1110/17/069/26/029/26/029/26/029/26/025/26/047/1/147/16/099/26/027/18/114/20/147/16/097/16/0912/18/9912/18/9912/18/9912/18/9910/12/0010/12/0012/18/992/7/0712/18/9912/18/9910/12/0412/18/9912/18/9912/18/992/7/072/27/142/27/1412/18/992/27/142/27/1411/7/122/27/1412/18/9911/7/1212/18/992/27/142/7/0712/18/992/7/0712/18/9910/17/0610/17/069/28/079/28/0710/17/06Wolf Creek -Unit 1viiRevision66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.14-1 0 Amend. No. 123 12/18/99B 3.4.14-2 0 Amend. No. 123 12/18/99B 3.4.14-3 0 Amend. No. 123 12/18/99B 3.4.14-4 0 Amend. No. 123 12/18/99B 3.4.14-5 32 DRR 07-0139 2/7/07B 3.4.14-6 32 DRR 07-0139 2/7/07B 3.4.15-1 31 DRR 06-2494 12/13/06B 3.4.15-2 31 DRR 06-2494 12/13/06B 3.4.15-3 33 DRR 07-0656 5/1/07B 3.4.15-4 33 DRR 07-0656 5/1/07B 3.4.15-5 65 DRR 14-2146 9/30/14B 3.4.15-6 31 DRR 06-2494 12/13/06B 3.4.15-7 31 DRR 06-2494 12/13/06B 3.4.15-8 31 DRR 06-2494 12/13/06B 3.4.16-1 31 DRR 06-2494 12/13/06B 3.4.16-2 31 DRR 06-2494 12/13/06B 3.4.16-3 31 DRR 06-2494 12/13/06B 3.4.16-4 31 DRR 06-2494 12/13/06B 3.4.16-5 31 DRR 06-2494 12/13/06B 3.4.17-1 29 DRR 06-1984 10/17/06B 3.4.17-2 58 DRR 13-0369 02/26/13B 3.4.17-3 52 DRR 11-0724 4/11/11B 3.4.17-4 57 DRR 13-0006 1/16/13B 3.4.17-5 57 DRR 13-0006 1/16/13B 3.4.17-6 57 DRR 13-0006 1/16/13B 3.4.17-7 58 DRR 13-0369 02/26/13TAB -B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)B 3.5.1-1 0 Amend. No. 123B 3.5.1-2 0 Amend. No. 123B 3.5.1-3 0 Amend. No. 123B 3.5.1-4 0 Amend. No. 123B 3.5.1-5 1 DRR 99-1624B 3.5.1-6 1 DRR 99-1624B 3.5.1-7 16 DRR 03-1497B 3.5.1-8 1 DRR 99-1624B 3.5.2-1 0 Amend. No. 123B 3.5.2-2 0 Amend. No. 123B 3.5.2-3 0 Amend. No. 123B 3.5.2-4 0 Amend. No. 123B 3.5.2-5 41 DRR 09-0288B 3.5.2-6 42 DRR 09-1009B 3.5.2-7 42 DRR 09-1009B 3.5.2-8 38 DRR 08-0624B 3.5.2-9 38 DRR 08-0624B 3.5.2-10 41 DRR 09-0288B 3.5.2-11 41 DRR 09-0288B 3.5.3-1 56 DRR 12-1792B 3.5.3-2 56 DRR 12-1792B 3.5.3-3 56 DRR 12-1792B 3.5.3-4 56 DRR 12-179212/18/9912/18/9912/18/9912/18/9912/18/9912/18/9911/4/0312/18/9912/18/9912/18/9912/18/9912/18/993/20/097/16/097/16/095/1/085/1/083/20/093/20/0911/7/1211/7/1211/7/1211/7/12Wolf Creek -Unit 1viiiRevision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

B 3.5.4-1 0 Amend. No. 123 12/18/99B 3.5.4-2 0 Amend. No. 123 12/18/99B 3.5.4-3 0 Amend. No. 123 12/18/99B 3.5.4-4 0 Amend. No. 123 12/18/99B 3.5.4-5 0 Amend. No. 123 12/18/99B 3.5.4-6 26 DRR 06-1350 7/24/06B 3.5.5-1 21 DRR 05-0707 4/20/05B 3.5.5-2 21 DRR 05-0707 4/20/05B 3.5.5-3 2 Amend. No. 132 4/24/00B 3.5.5-4 21 DRR 05-0707 4/20/05TAB -B 3.6 CONTAINMENT SYSTEMSB 3.6.1-1 0B 3.6.1-2 0B 3.6.1-3 0B 3.6.1-4 17B 3.6.2-1 0B 3.6.2-2 0B 3.6.2-3 0B 3.6.2-4 0B 3.6.2-5 0B 3.6.2-6 0B 3.6.2-7 0B 3.6.3-1 0B 3.6.3-2 0B 3.6.3-3 0B 3.6.3-4 49B 3.6.3-5 49B 3.6.3-6 49B 3.6.3-7 41B 3.6.3-8 36B 3.6.3-9 36B 3.6.3-10 8B 3.6.3-11 36B 3.6.3-12 36B 3.6.3-13 50B 3.6.3-14 36B 3.6.3-15 39B 3.6.3-16 39B 3.6.3-17 36B 3.6.3-18 36B 3.6.3-19 36B 3.6.4-1 39B 3.6.4-2 0B 3.6.4-3 0B 3.6.5-1 0B 3.6.5-2 37B 3.6.5-3 13B 3.6.5-4 0B 3.6.6-1 42B 3.6.6-2 63Amend. No. 123Amend. No. 123Amend. No. 123DRR 04-0453Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 11-0014DRR 11-0014DRR 11-0014DRR 09-0288DRR 08-0255DRR 08-0255DRR 01-1235DRR 08-0255DRR 08-0255DRR 11-0449DRR 08-0255DRR 08-1096DRR 08-1096DRR 08-0255DRR 08-0255DRR 08-0255DRR 08-1096Amend. No. 123Amend. No. 123Amend. No. 123DRR 08-0503DRR 02-1458Amend. No. 123DRR 09-1009DRR 14-157212/18/9912/18/9912/18/995/26/0412/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/991/31/111/31/111/31/113/20/093/11/083/11/089/19/013/11/083/11/083/9/113/11/088/28/088/28/083/11/083/11/083/11/088/28/0812/18/9912/18/9912/18/994/8/0812/03/0212/18/997/16/097/1/14Wolf Creek -Unit 1ixRevision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.6 CONTAINMENT SYSTEMS (continued)

B 3.6.6-3 37 DRR 08-0503 4/8/08B 3.6.6-4 42 DRR 09-1009 7/16/09B 3.6.6-5 0 Amend. No. 123 12/18/99B 3.6.6-6 18 DRR 04-1018 9/1/04B 3.6.6-7 0 Amend. No. 123 12/18/99B 3.6.6-8 32 DRR 07-0139 2/7/07B 3.6.6-9 58 DRR 13-0369 2/26/13B 3.6.7-1 0 Amend. No. 123 12/18/99B 3.6.7-2 42 DRR 09-1009 7/16/09B 3.6.7-3 0 Amend. No. 123 12/18/99B 3.6.7-4 29 DRR 06-1984 10/17/06B 3.6.7-5 42 DRR 09-1009 7/16/09TAB -B 3.7 PLANT SYSTEMSB 3.7.1-1B 3.7.1-2B 3.7.1-3B 3.7.1-4B 3.7.1-5B 3.7.1-6B 3.7.2-1B 3.7.2-2B 3.7.2-3B 3.7.2-4B 3.7.2-5B 3.7.2-6B 3.7.2-7B 3.7.2-8B 3.7.2-9B 3.7.2-10B 3.7.2-11B 3.7.3-1B 3.7.3-2B 3.7.3-3B 3.7.3-4B 3.7.3-5B 3.7.3-6B 3.7.3-7B 3.7.3-8B 3.7.3-9B 3.7.3-10B 3.7.3-11B 3.7.4-1B 3.7.4-2B 3.7.4-3B 3.7.4-4B 3.7.4-5B 3.7.5-10000323244444444444444444444443750373737373737666637111919154Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 07-0139DRR 07-0139DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 09-1744DRR 08-0503DRR 11-0449DRR 08-0503DRR 08-0503DRR 08-0503DRR 08-0503DRR 08-0503DRR 08-0503DRR 14-2329DRR 14-2329DRR 08-0503DRR 99-1624DRR 99-1624DRR 04-1414DRR 04-1414DRR 99-1624DRR 11-239412/18/9912/18/9912/18/9912/18/99217/072/7/0710/28/0910/28/0910/28/0910/28/0910/28/0910/28/0910/28/0910/28/0910/28/0910/28/0910/28/094/8/083/9/114/8/084/8/084/8/084/8/084/8/084/8/0811/6/1411/6/144/8/0812/18/9912/18/9910/12/0410/12/0412/18/9911/16/11Wolf Creek -Unit 1XRe,%tson 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.7 PLANT SYSTEMSB 3.7.5-2B 3.7.5-3B 3.7.5-4B 3.7.5-5B 3.7.5-6B 3.7.5-7B 3.7.5-8B 3.7.5-9B 3.7.6-1B 3.7.6-2B 3.7.6-3B 3.7.7-1B 3.7.7-2B 3.7.7-3B 3.7.7-4B 3.7.8-1B 3.7.8-2B 3.7.8-3B 3.7.8-4B 3.7.8-5B 3.7.9-1B 3.7.9-2B 3.7.9-3B 3.7.9-4B 3.7.10-1B 3.7.10-2B 3.7.10-3B 3.7.10-4B 3.7.10-5B 3.7.10-6B 3.7.10-7B 3.7.10-8B 3.7.10-9B 3.7.11-1B 3.7.11-2B 3.7.11-3B 3.7.11-4B 3.7.12-1B 3.7.13-1B 3.7.13-2B 3.7.13-3B 3.7.13-4B 3.7.13-5B 3.7.13-6B 3.7.13-7B 3.7.13-8B 3.7.14-1B 3.7.15-1B 3.7.15-2B 3.7.15-3B 3.7.16-1(continued) 54060444432143200000010000033336441414157576441640576363024142575764646400005DRR 11-2394Amend. No. 123DRR 13-2562DRR 09-1744DRR 09-1744DRR 07-0139DRR 03-0102DRR 07-0139Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 99-1624Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 134Amend. No. 134Amend. No. 134Amend. No. 134DRR 14-1822DRR 09-0288DRR 09-0288DRR 09-0288DRR 13-0006DRR 13-0006DRR 14-1822DRR 09-0288DRR 14-1822Amend. No. 123DRR 13-0006DRR 14-1572DRR 14-1572Amend. No. 123DRR 06-0051DRR 99-1624DRR 09-1009DRR 13-0006DRR 13-0006DRR 14-1822DRR 14-1822DRR 14-1822Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 00-142711/16/1112/18/9910/25/1310/28/0910/28/092/7/072/12/032/7/0712/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/997/14/007/14/007/14/007/14/008/28/143/20/093/20/093/20/091/16/131/16/138/28/143/20/098/28/1412/18/991/16/137/1/147/1/1412/18/992/28/0612/18/997/16/091/16/131/16/138/28/148/28/148/28/1412/18/9912/18/9912/18/9912/18/9910/12/00Wolf Creek -Unit IxiRevision66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.7 PLANT SYSTEMS (continued)

B 3.7.16-2 23 DRR 05-1995 9/28/05B 3.7.16-3 5 DRR 00-1427 10/12/00B 3.7.17-1 7 DRR 01-0474 5/1/01B 3.7.17-2 7 DRR 01-0474 5/1/01B 3.7.17-3 5 DRR 00-1427 10/12/00B 3.7.18-1 0 Amend. No. 123 12/18/99B 3.7.18-2 0 Amend. No. 123 12/18/99B 3.7.18-3 0 Amend. No. 123 12/18/99B 3.7.19-1 44 DRR 09-1744 10/28/09B 3.7.19-2 54 DRR 11-2394 11/16/11B 3.7.19-3 54 DRR 11-2394 11/16/11B 3.7.19-4 61 DRR 14-0346 2/27/14B 3.7.19-5 61 DRR 14-0346 2/27/14B 3.7.19-6 54 DRR 11-2394 11/16/11B 3.7.19-7 54 DRR 11-2394 11/16/11TAB -B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.1-1 54B 3.8.1-2 0B 3.8.1-3 47B 3.8.1-4 54B 3.8.1-5 59B 3.8.1-6 25B 3.8.1-7 26B 3.8.1-8 35B 3.8.1-9 42B 3.8.1-10 39B 3.8.1-11 36B 3.8.1-12 47B 3.8.1-13 47B 3.8.1-14 47B 3.8.1-15 47B 3.8.1-16 26B 3.8.1-17 26B 3.8.1-18 59B 3.8.1-19 26B 3.8.1-20 26B 3.8.1-21 33B 3.8.1-22 33B 3.8.1-23 40B 3.8.1-24 33B 3.8.1-25 33B 3.8.1-26 33B 3.8.1-27 59B 3.8.1-28 59B 3.8.1-29 54B 3.8.1-30 33B 3.8.1-31 33B 3.8.1-32 33B 3.8.1-33 39B 3.8.1-34 47DRR 11-2394Amend. No. 123DRR 10-1089DRR 11-2394DRR 13-1524DRR 06-0800DRR 06-1350DRR 07-1553DRR 09-1009DRR 08-1096DRR 08-0255DRR 10-1089DRR 10-1089DRR 10-1089DRR 10-1089DRR 06-1350DRR 06-1350DRR 13-1524DRR 06-1350DRR 06-1350DRR 07-0656DRR 07-0656DRR 08-1846DRR 07-0656DRR 07-0656DRR 07-0656DRR 13-1524DRR 13-1524DRR 11-2394DRR 07-0656DRR 07-0656DRR 07-0656DRR 08-1096DRR 10-108911/16/1112/18/996/16/1011/16/116/26/135/18/067/24/069/28/077/16/098/28/083/11/086/16/106/16/106/16/106/16/107/24/067/24/066/26/137/24/067/24/065/1/075/1/0712/9/085/1/075/1/075/1/076/26/136/26/1311/16/115/1/075/1/075/1/078/28/086/16/10Wolf Creek -Unit 1xiiRevision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.2-1 57B 3.8.2-2 0B 3.8.2-3 0B 3.8.2-4 57B 3.8.2-5 57B 3.8.2-6 57B 3.8.2-7 57B 3.8.3-1 1B 3.8.3-2 0B 3.8.3-3 0B 3.8.3-4 1B 3.8.3-5 0B 3.8.3-6 0B 3.8.3-7 12B 3.8.3-8 1B 3.8.3-9 0B 3.8.4-1 0B 3.8.4-2 0B 3.8.4-3 0B 3.8.4-4 0B 3.8.4-5 50B 3.8.4-6 50B 3.8.4-7 6B 3.8.4-8 0B 3.8.4-9 2B 3.8.5-1 57B 3.8.5-2 0B 3.8.5-3 57B 3.8.5-4 57B 3.8.5-5 57B 3.8.6-1 0B 3.8.6-2 0B 3.8.6-3 0B 3.8.6-4 0B 3.8.6-5 0B 3.8.6-6 0B 3.8.7-1 0B 3.8.7-2 5B 3.8.7-3 0B 3.8.7-4 0B 3.8.8-1 57B 3.8.8-2 0B 3.8.8-3 0B 3.8.8-4 57B 3.8.8-5 57B 3.8.9-1 54B 3.8.9-2 54B 3.8.9-3 54B 3.8.9-4 0B 3.8.9-5 0B 3.8.9-6 0(continued)

DRR 13-0006Amend. No. 123Amend. No. 123DRR 13-0006DRR 13-0006DRR 13-0006DRR 13-0006DRR 99-1624Amend. No. 123Amend. No. 123DRR 99-1624Amend. No. 123Amend. No. 123DRR 02-1062DRR 99-1624Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 11-0449DRR 11-0449DRR 00-1541Amend. No. 123DRR 00-0147DRR 13-0006Amend. No. 123DRR 13-0006DRR 13-0006DRR 13-0006Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 00-1427Amend. No. 123Amend. No. 123DRR 13-0006Amend. No. 123Amend. No. 123DRR 13-0006DRR 13-0006DRR 11-2394DRR 11-2394DRR 11-2394Amend. No. 123Amend. No. 123Amend. No. 1231/16/1312/18/9912/18/991/16/131/16/131/16/131/16/1312/18/9912/18/9912/18/9912/18/9912/18/9912/18/999/26/0212/18/9912/18/9912/18/9912/18/9912/18/9912/18/993/9/113/9/113/13/0112/18/994/24/001/16/1312/18/991/16/131/16/131/16/1312/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9910/12/0012/18/9912/18/991/16/1312/18/9912/18/991/16/131/16/1311/16/1111/16/1111/16/1112/18/9912/18/9912/18/99Wolf Creek -Unit 1xiiiRevLision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.9-7 0 Amend. No. 123 12/18/99B 3.8.9-8 1 DRR 99-1624 12/18/99B 3.8.9-9 0 Amend. No. 123 12/18/99B 3.8.10-1 57 DRR 13-0006 1/16/13B 3.8.10-2 0 Amend. No. 123 12/18/99B 3.8.10-3 0 Amend. No. 123 12/18/99B 3.8.10-4 57 DRR 13-0006 1/16/13B 3.8.10-5 57 DRR 13-0006 1/16/13B 3.8.10-6 57 DRR 13-0006 1/16/13TAB -B 3.9 REFUELING OPERATIONS B 3.9.1-1 0 Amend. No. 123 12/18/99B 3.9.1-2 19 DRR 04-1414 10/12/04B 3.9.1-3 19 DRR 04-1414 10/12/04B 3.9.1-4 19 DRR 04-1414 10/12/04B 3.9.2-1 0 Amend. No. 123 12/18/99B 3.9.2-2 0 Amend. No. 123 12/18/99B 3.9.2-3 0 Amend. No. 123 12/18/99B 3.9.3-1 24 DRR 06-0051 2/28/06B 3.9.3-2 51 DRR 11-0664 3/21/11B 3.9.3-3 51 DRR 11-0664 3/21/11B 3.9.3-4 53 DRR 11-1513 7/18/11B 3.9.4-1 23 DRR 05-1995 9/28/05B 3.9.4-2 13 DRR 02-1458 12/03/02B 3.9.4-3 25 DRR 06-0800 5/18/06B 3.9.4-4 23 DRR 05-1995 9/28/05B 3.9.4-5 33 DRR 07-0656 5/1/07B 3.9.4-6 23 DRR 05-1995 9/28/05B 3.9.5-1 0 Amend. No. 123 12/18/99B 3.9.5-2 32 DRR 07-0139 2/7/07B 3.9.5-3 32 DRR 07-0139 2/7/07B 3.9.5-4 32 DRR 07-0139 2/7/07B 3.9.6-1 0 Amend. No. 123 12/18/99B 3.9.6-2 63 DRR 14-1572 7/16/09B 3.9.6-3 42 DRR 09-1009 7/16/09B 3.9.6-4 42 DRR 09-1009 7/16/09B 3.9.7-1 25 DRR 06-0800 5/18/06B 3.9.7-2 0 Amend. No. 123 12/18/99B 3.9.7-3 0 Amend. No. 123 12/18/99Wolf Creek -Unit 1xiVRevision 66 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASESPAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)Note 1 The page number is listed on the center of the bottom of each page.Note 2 The revision number is listed in the lower right hand corner of each page. The Revisionnumber will be page specific.

Note 3 The change document will be the document requesting the change. Amendment No.123 issued the improved Technical Specifications and associated Bases which affectedeach page. The NRC has indicated that Bases changes will not be issued with LicenseAmendments.

Therefore, the change document should be a DRR number inaccordance with AP 26A-002.Note 4 The date effective or implemented is the date the Bases pages are issued by DocumentControl.Wolf Creek -Unit 1XVRevision 66