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Wolf Creek Generating Station Changes to Technical Specification Bases - Revisions 67 Through 73
	ML16076A357
Person / Time
Site:	Wolf Creek
Issue date:	03/10/2016
From:	Hafenstine C R; Wolf Creek
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA 16-0008
	Download: ML16076A357 (85)
	v • d • e

W0LF CREEK7 NUCLEAR OPERATING CORPORATION Cynthia R. Hafenstine Manager Regulatory AffairsMarch 10, 2016RA 16-0008U. 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 67 through 73Gentlemen:

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 67 through 73) under the provisions to TSSection 5.5.14 and a List of Effective Pages. This submittal reflects changes from January 1,2015 through December 31, 2015.This letter contains no commitments.

If you have any questions concerning this matter, pleasecontact me at (620) 364-4204.

Sincerely, Cynthia R. Hafenstine CRH/rltEnclosure cc: M. L. Dapas (NRC), w/eC. F. Lyon (NRC), w/eN. H. Taylor (NRC), w/e 0Senior Resident Inspector (NRC), w/e -P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831An Equal Opportunity Employer MIFIHC/VET Enclosure to IRA 16-0008Wolf Creek Generating StationChanges to the Technical Specification Bases(44 pages)

FQ(Z) (EQ Methodology)

B 3.2.1BASESSURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS a precise measurement in these regions.

It should be noted that while thetransient FQ(Z) limits are not measured in these axial core regions, theanalytical transient FQ(Z) limits in these axial core regions aredemonstrated to be satisfied during the core reload design process.This Surveillance has been modified by a Note that may require morefrequent surveillances be performed.

When FQc(Z) is measured, anevaluation of the expression below is required to account for any increaseto FQ(Z) that may occur and cause the FQ(Z) limit to be exceeded beforethe next required FQ(Z) evaluation.

If the two most recent F0(Z) evaluations show an increase in theexpression maximum overz [FQ z)it is required to meet the FQ(Z) limit with the last FQw(Z) increased by theappropriate factor specified in the COLR, or to evaluate FQ(Z) morefrequently, each 7 EFPD. These alternative requirements prevent FQ(Z)from exceeding its limit for any significant period of time without detection.

Performing the Surveillance in MODE 1 prior to exceeding 75% RTPensures that the FQ(Z) limit will be met when RTP is achieved, becausepeaking factors are generally decreased as power level is increased.

FQ(Z) is verified at power levels > 10% RTP above the THERMALPOWER of its last verification, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.The Surveillance Frequency of 31 EFPD is adequate to monitor thechange of power distribution with core burnup. The Surveillance may bedone more frequently if required by the results of FQ(Z) evaluations.

The Frequency of 31 EFPD is adequate to monitor the change of powerdistribution because such a change is sufficiently slow, when the plant isoperated in accordance with the TS, to preclude adverse peaking factorsbetween 31 day surveillances.

Wolf Creek -Unit 1 ..- eiin2B 3.2.1-9Revision 29 F0(Z) (F0 Methodology)

B 3.2.1BASESREFERENCES

°.2.3.4.5.6.10 CFR 50.46, 1974.USAR, Section 15.4.8.10 CFR 50, Appendix A, GDC 26.WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel FactorUncertainties,"

June 1988.Performance Improvement Request 2005-3311.

WCAP-1 2472-P-A, "BEACON Core Monitoring and Operations Support System,"

August 1994 (including Addendum 4, September 2012).Wolf Creek.- Unit I B3211 eiin7B 3.2.1-10Revision 70 B 3.2.2BASESACTIONS A.1.2.1 and A.1.2.2 (continued) condition for an extended period of time. The Completion Times of4 hours for Required Actions A.1 .1 and A.1 .2.1 are not additive.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to reset the trip setpoints perRequired Action A.1 .2.2 recognizes that, once power is reduced, thesafety analysis assumptions are satisfied and there is no urgent need toreduce the trip setpoints.

A..22Once the power level has been reduced to < 50% RTP per RequiredAction A.1 .2.1, a power distribution measurement (SR 3.2.2.1 ) must beobtained and the measured value of verified not to exceed theallowed limit at the lower power level. The unit is provided 68 additional hours to perform this task over and above the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed by eitherAction A.1 .1 or Action A.1 .2.1. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months isacceptable because of the increase in the DNB margin, which is obtainedat lower power levels, and the low probability of having a DNB limitingevent within this 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period. Additionally, operating experience hasindicated that this Completion Time is sufficient to obtain the powerdistribution measurement, perform the required calculations, and evaluateI*A.3Verification that is within its specified limits after an out of limitoccurrence ensures that the cause that led to the FNAJH exceeding its limitis identified, to the extent necessary, and corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates thatthe FNAN limit is within the LCO limits prior to exceeding 50% RTP, againprior to exceeding 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMALPOWER is >95% RTP.This Required Action is modified by a Note that states that THERMALPOWER does not have to be reduced prior to performing this Action.B.._IWhen Required Actions A.1.1 through A.3 cannot be completed withintheir required Completion Times, the plant must be placed in a mode inwhich the LCO requirements are not applicable.

This is done by placingthe plant in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Wolf Creek -Unit 1 ..- eiin4B 3.2.2-5Revision 48 B 3.2.2BASESACTIONS 8.1 (continued)

Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in anorderly manner and without challenging plant systems.SURVEILLANCE SR 3.2.2.1REQUIREMENTS SR 3.2.2.1 is modified by a Note. The Note applies during powerascensions following a plant shutdown (leaving MODE 1). The Noteallows for power ascensions if the surveillances are not current.

It statesthat THERMAL POWER may be increased until an equilibrium powerlevel has been achieved at which a power distribution measurement canbe obtained.

Equilibrium conditions are achieved when the core issufficiently stable at the intended operating conditions to perform themeasurement.

The value of FNAH is determined by using either the movable incoredetector system or the Power Distribution Monitoring System to obtain apower distribution measurement.

A calculation determines the maximumvalue of FNAH- from the measured power distribution.

The measured valueof FNAH must be increased by 4% (if using the movable incore detectorsystem) or increased by (if using the Power Distribution Monitoring System, where UAH is determined as described in Reference 4, with aminimum value of 4%) to account for measurement uncertainty beforemaking comparisons to the limitAfter each refueling, FNAN must be determined in MODE I prior toexceeding 75% RTP. This requirement ensures that FNAH~ limits are metat the beginning of each fuel cycle.The 31 EFPD Frequency is acceptable because the power distribution changes relatively slowly over this amount of fuel burnup. Accordingly, this Frequency is short enough that the limit cannot be exceeded forany significant period of operation.

REFERENCES

1. USAR, Section 15.4.8.2. 10 CFR 50, Appendix A, GDC 26.3. 10 CFR 50.46.4. WCAP-1 2472-P-A, "BEACON Core Monitoring and Operations Support System,"

August 1994 (including Addendum 4, September 2012).Wolf Creek -Unit 1B3226Reion7 B 3.2.2-6Revision 70 RCS P/T LimitsB 3.4.3B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.3 RCS Pressure and Temperature (PIT) LimitsBASESBACKGROUND All components of the RCS are designed to withstand effects of cyclicloads due to system pressure and temperature changes.

These loads areintroduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure andtemperature changes during RCS heatup and cooldown, within the designassumptions and the stress limits for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown, inservice leakand hydrostatic (ISLH) testing, and data for the maximum rate of changeof reactor coolant temperature (Ref. 1).Each PIT limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup orcooldown maneuvering, when pressure and temperature indications aremonitored and compared to the applicable curve to determine thatoperation is within the allowable region. Vacuum fill of the RCS isnormally performed in MODE 5 under sub-atmospheric pressure andisothermal RCS conditions.

Vacuum fill is an acceptable condition sincethe resulting pressure/temperature combination is located in the region tothe right and below the operating limits provided in Figures 2.1-1 and2.1-2 of the PTLR.The LCO establishes operating limits that provide a margin to brittle failureof the reactor vessel and piping of the reactor coolant pressure boundary(RCPB). The vessel is the component most subject to brittle failure, andthe LCO limits apply mainly to the vessel. The limits do not apply to thepressurizer, which has different design characteristics and operating functions.

10 CFR 50, Appendix G (Ref. 2), requires the establishment of PIT limitsfor specific material fracture toughness requirements of the RCPBmaterials.

Reference 2 requires an adequate margin to brittle failureduring normal operation, anticipated operational occurrences, and systemhydrostatic tests. It mandates the use of the American Society ofMechanical Engineers (ASME) Code,Section III, Appendix G (Ref. 3).The neutron embrittlement effect on the material toughness is reflected byincreasing the nil ductility reference temperature (RTNDT) as exposure toneutron fluence increases.

The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vesselmaterial specimens, in accordance with ASTM E 185 (Ref. 4) andWolf Creek -Unit IB343-Reion6 B3.4.3-1Revision 67 RCS P/T LimitsB 3.4.3BASESBACKGROUND (continued)

Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limit curves will beadjusted, as necessary, based on the evaluation findings and therecommendations of Regulatory Guide 1.99 (Ref. 6).The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vesseland head that are the most restrictive.

At any specific

pressure, temperature, and temperature rate of change, one location within thereactor vessel will dictate the most restrictive limit. Across the span of theP/T limit curves, different locations are more restrictive, and, thus, thecurves are composites of the most restrictive regions.The heatup curve represents a different set of restrictions than thecooldown curve because the directions of the thermal gradients throughthe vessel wall are reversed.

The thermal gradient reversal alters thelocation of the tensile stress between the outer and inner walls.The criticality limit curve includes the Reference 2 requirement that it be> 40°F above the heatup curve or the cooldown curve, and not less thanthe minimum permissible temperature for ISLH testing.

However, thecriticality curve is not operationally limiting; a more restrictive limit exists inLCO 3.4.2, "RCS Minimum Temperature for Criticality."

The consequence of violating the LCO limits is that the RCS has beenoperated under conditions that can result in brittle failure of the RCPB,possibly leading to a nonisolable leak or loss of coolant accident.

In theevent these limits are exceeded, an evaluation must be performed todetermine the effect on the structural integrity of the RCPB components.

The ASME Code, Section Xl, Appendix E (Ref. 7), provides arecommended methodology for evaluating an operating event that causesan excursion outside the limits.APPLICABLE SAFETY ANALYSESThe P/T limits are not derived from Design Basis Accident (DBA)analyses.

They are prescribed during normal operation to avoidencountering

pressure, temperature, and temperature rate of changeconditions that might cause undetected flaws to propagate and causenonductile failure of the RCPB, an unanalyzed condition.

Reference 1establishes the methodology for determining the P/T limits. Although theP/T limits are not derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Wolf Creek -Unit 1 ..- RvsoB3.4.3-2Revision 0

RCS Loops -MODE 4B 3.4.6B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.6 RCS Loops -MODE 4BASESBACKGROUND In MODE 4, the primary function of the reactor coolant is the removal ofdecay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via theresidual heat removal (RHR) heat exchangers.

The secondary function ofthe reactor coolant is to act as a carrier for soluble neutron poison, boricacid.The reactor coolant is circulated through four RCS loops connected inparallel to the reactor vessel, each loop containing an SG, a reactorcoolant pump (RCP), and appropriate flow, pressure, level, andtemperature instrumentation for control, protection, and indication.

TheRCPs circulate the coolant through the reactor vessel and SGs at asufficient rate to ensure proper heat transfer and to prevent boric acidstratification.

In MODE 4, either RCPs or RHR loops can be used to provide forcedcirculation.

The intent of this LCO is to provide forced flow from at leastone RCP or one RHR loop for decay heat removal and transport.

Theflow provided by one RCP loop or RHR loop is adequate for decay heatremoval.

The other intent of this LCO is to require that two paths beavailable to provide redundancy for decay heat removal.APPLICABLE In MODE 4, RCS circulation is considered in the determination of the timeSAFETY ANALYSES available for mitigation of the accidental boron dilution event.The operation of one RCP in MODES 3, 4, and 5 provides adequate flowto ensure mixing, prevent stratification, and produce gradual reactivity changes during RCS boron concentrationi reductions.

With no reactorcoolant loop in operation in either MODES 3, 4, or 5, dilution sources mustbe isolated or administratively controlled.

The boron dilution analysis inthese MODES take credit for the mixing volume associated with having atleast one reactor coolant loop in operation (Ref. 1 ).RCS Loops- MODE 4 satisfies Criterion 4 of 10 CER 50.36(c)(2)(ii).

Wolf Creek -Unit IB346-Reion5 B3.4.6-1Revision 53 RCS Loops-MODE 4B 3.4.6BASESLCO The purpose of this LCO is to require that at least two loops beOPERABLE in MODE 4 and that one of these loops be in operation.

TheLCO allows the two loops that are required to be OPERABLE to consist ofany combination of RCS loops and RHR loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forcedcirculation.

An additional loop is required to be OPERABLE to provideredundancy for heat removal.Note 1 permits all RCPs or RHR pumps to be removed from operation for_< 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests thatare required to be performed without flow or pump noise. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months timeperiod is adequate to perform the necessary

testing, and operating experience has shown that boron stratification is not a problem during thisshort period with no forced flow.Utilization of Note I is permitted provided the following conditions are metalong with any other conditions imposed by test procedures:

a. No operations are permitted that would dilute the RCS boronconcentration with coolant at boron concentrations less thanrequired to assure the SDM of LCO 3.1.1, thereby maintaining themargin to criticality.

Boron reduction with coolant at boronconcentrations less than required to assure the SDM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; andb. Core outlet temperature is maintained at least 1 0°F belowsaturation temperature, so that no vapor bubble may form andpossibly cause a natural circulation flow obstruction.

Note 2 requires that the secondary side water temperature of each SG be_< 50°F above each of the RCS cold leg temperatures before the start ofan RCP with any RCS cold leg temperature

_< 368°F. This restraint is toprevent a low temperature overpressure event due to a thermal transient when an RCP is started."

An OPERABLE RCS loop is comprised of an OPERABLE RCP and anOPERABLE SG, which has the minimum water level specified inSR 3.4.6.2.Similarly for the RHR System, an OPERABLE RHR loop comprises anOPERABLE RHR pump capable of providing forced flow to anOPERABLE RHR heat exchanger.

RCPs and RHR pumps areOPERABLE if they are capable of being powered and are able to provideforced flow if required.

Management of gas voids is important to RHRSystem Operability.

Wolf Creek -Unit 1 ..- eiin7B3.4.6-2Revision 72 RCS Loops -MODE 4B 3.4.6BASESSURVEILLANCE SR 3.4.6.4REQUIREMENTS (continued)

RHR System piping and components have the potential to develop voidsand pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required RHRloop(s) and may also prevent water hammer, pump cavitation, andpumping of noncondensible gas into the reactor vessel.Selection of RHR System locations susceptible to gas accumulation isbased on a review of system design information, including piping andinstrumentation

drawings, isometric

drawings, plan and elevation

drawings, and calculations.

The design review is supplemented bysystem walk downs to validate the system high points and to confirm thelocation and orientation of important components that can becomesources of gas or could otherwise cause gas to be trapped or difficult toremove during system maintenance or restoration.

Susceptible locations

.................depend on plant and system configuration, such as stand-by versusoperating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.Acceptance criteria are established for the volume of accumulated gas atsusceptible locations.

If accumulated gas is discovered that exceeds theacceptance criteria for the susceptible location (or the volume ofaccumulated gas at one or more susceptible locations exceeds anacceptance criteria for gas volume at the suction or discharge of a pump),the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas(i.e., the system is sufficiently filled with water), the Surveillance may bedeclared met. Accumulated gas should be eliminated or brought within theacceptance criteria limits.RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location.

Susceptible locations in the same system flowpath which are subject to the same gas intrusion mechanisms may beverified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due toradiological or environmental conditions, the plant configuration, orpersonnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location.

Monitoring is not required for susceptible locations where themaximum potential accumulated gas void volume has been evaluated anddetermined to not challenge system OPERABILITY.

The accuracy of themethod used for monitoring the susceptible locations and trending of theresults should be sufficient to assure system OPERABILITY during theSurveillance interval.

Wolf Creek -Unit 1 ..- eiin7B 3.4.6-5Revision 72 RCS Loops -MODE 4B 3.4.6BASESSURVEILLANCE SR 3.4.6.4 (continued)

REQUIREMENTS This SR is modified by a Note that states the SR is not required to beperformed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 4. In a rapid shutdown, there may be insufficient time to verify all susceptible locations prior toentering MODE 4.The 31 day Frequency takes into consideration the gradual nature of gasaccumulation in the RHR System piping and the procedural controlsgoverning system operation.

REFERENCES

1. USAR, Section 15.4.6/Wolf Creek -Unit 1 ..- eiin7B3.4.6-6Revision 72 RCS Loops -MODE 5, Loops FilledB 3.4.7BASESLCO b. Core outlet temperature is maintained at least 10°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

< 368°F.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.

Management of gasvoids is important to RHR System OPERABILITY.

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 1 ..- eiin7B 3.4.7-3Revision 72 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-MODES5, 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 I 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications andalarms available to the operator in the control room to monitor RHR loopperformance.

Wolf Creek -Unit I1 ..- eiin4B 3.4.7-4 RCS Loops -MODE 5, Loops FilledB 3.4.7BASESSURVEILLANCE SR 3.4.7.2REQUIREMENTS (continued)

Verifying that at least two SGs are OPERABLE by ensuring theirsecondary side wide range water levels are >_ 66% ensures an alternate decay heat removal method is available via natural circulation in the eventthat the second RHR loop is not OPERABLE.

If both RHR loops areOPERABLE, this Surveillance is not needed. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency isconsidered adequate in view of other indications available in the controlroom to alert the operator to the loss of SG level.SR 3.4.7.3Verification that a second RHR pump is OPERABLE ensures that anadditional pump can be placed in operation, if needed, to maintain decayheat removal and reactor coolant circulation.

Verification is performed byverifying proper breaker alignment and power available to the RHR pump.If secondary side wide range water level is > 66% in at least two SGs, thisSurveillance is not needed. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has beenshown to be acceptable by operating experience.

SR 3.4.7.4.RHR System piping and components have the potential to develop voidsand pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required RHRloop(s) and may also prevent water hammer, pump cavitation, andpumping of noncondensible gas into the reactor vessel.Selection of RHR System locations susceptible to gas accumulation isbased on a review of system design information, including piping andinstrumentation

drawings, isometric

drawings, plan and elevation

drawings, and calculations.

The design review is supplemented bysystem walk downs to validate the system high points and to confirm thelocation and orientation of important components that can becomesources of gas or could otherwise cause gas to be trapped or difficult toremove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versusoperating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.Acceptance criteria are established for the volume of accumulated gas atsusceptible locations.

If accumulated gas is discovered that exceeds theacceptance criteria for the susceptible location (or the volume ofWolf Creek -Unit 1 ..- eiin7B3.4.7-5Revision 72

....." ...... RCS Loops -MODE 5, Loops FilledB 3.4.7BAS ESSURVEILLANCE SR 3.4.7.4 (continued)

REQUIREMENTS accumulated gas at one or more susceptible locations exceeds anacceptance criteria for gas volume at the suction or discharge of a pump),the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas(i.e., the system is sufficiently filled with water), the Surveillance may bedeclared met. Accumulated gas should be eliminated or brought withinthe acceptance criteria limits.RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location.

Susceptible locations in the same system flowpath which are subject to the same gas intrusion mechanisms may beverified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due toradiological or environmental conditions, the plant configuration, orpersonnel safety. For these locations alternative methods (e.g., operating

....................

parameters, remote-monitoring) may be used to monitor-the susceptible-location.

Monitoring is not required for susceptible locations where themaximum potential accumulated gas void volume has been evaluated anddetermined to not challenge system OPERABILITY.

The accuracy of themethod used for monitoring the susceptible locations and trending of theresults should be sufficient to assure system OPERABILITY during theSurveillance interval.

The 31 day Frequency takes into consideration the gradual nature of gasaccumulation in the RHR System piping and the procedural controlsgoverning system operation.

REFERENCES

1. USAR, Section 15.4.6.2. NRC Information Notice 95-35, "Degraded Ability of SGs to RemoveDecay Heat by Natural Circulation."

Wolf Creek -Unit 1 ..- eiin7B3.4.7-6Revision 72

-RCS Loops -MODE 5, Loops Not FilledB 3.4.8B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.8 RCS Loops -MODE 5, Loops Not FilledBASESBACKGROUND In MODE 5 with the RCS loops not filled, the primary function of thereactor coolant is the removal of decay heat generated in the fuel, and thetransfer of this heat to the component cooling water via the residual heatremoval (RHR) heat exchangers.

The steam generators (SGs) are notavailable as a heat sink when the loops are not filled. The secondary function of the reactor coolant is to act as a carrier for the soluble neutronpoison, boric acid.In MODE 5 with loops not filled, only RHR pumps can be used for coolantcirculation.

The number of pumps in operation can vary to suit theoperational needs. The intent of this LCO is to provide forced flow from atleast one RHR pump for decay heat removal and transport and to requirethat two paths be available to provide redundancy for heat removal.APPLICABLE In MODE 5, RCS circulation is considered in the determination of theSAFETY ANALYSES time available for mitigation of the accidental boron dilution event. Theflow provided by one RHR loop is adequate for decay heat removal.The operation of one RCP in MODES 3, 4, and 5 provides adequate flowto ensure mixing, prevent stratification, and produce gradual reactivity changes during RCS boron concentration reductions.

With no reactorcoolant loop in operation in either MODES 3, 4, or 5, dilution sources mustbe isolated or administratively controlled.

The boron dilution analysis inthese MODES take credit for the mixing volume associated with having atleast one reactor coolant ioop in operation (Ref. 1 ).RCS loops in MODE 5 (loops not filled) satisfies Criterion 4 of 10 CFR50.36(c)(2)(ii).

LCO The purpose of this LCO is to require that at least two RHR loops beOPERABLE and one of these loops be in operation.

An OPERABLE loopis one that has the capability of transferring heat from the reactor coolantat a controlled rate. Heat cannot be removed via the RHR System unlessforced flow is used. A minimum of one running RHR pump meets theLCO requirement for one loop in operation.

An additional RHR loop isrequired to be OPERABLE to meet single failure considerations.

Wolf Creek -Unit 1B348-Reion5 B3.4.8-1Revision 53 RCS Loops -MODE 5, L~oops Not FilledB 3.4.8BASESLCO(continued)

Note 1 permits all RHR pumps to be removed from operation for _< 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .The circumstances for stopping both RHR pumps are to be limited tosituations when the outage time is short and core outlet temperature ismaintained at least 1 0°F below saturation temperature.

The Noteprohibits boron dilution with coolant at boron concentrations less thanrequired to assure the SDM of LCO 3.1.1 is maintained or drainingoperations when RHR forced flow is stopped.

The Note requires reactorvessel water level be above the vessel flange to ensure the operating RHR pump will not be intentionally deenergized during mid-loopoperations.

Note 2 allows one RHR loop to be inoperable for a period of < 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,provided that the other loop is OPERABLE and in operation.

This permitsperiodic surveillance tests to be performed on the inoperable loop duringthe only time when these tests are safe and possible.

An OPERABLE RHR loop is comprised of an OPERABLE RHR pumpcapable of providing forced flow to an OPERABLE RHR heat exchanger.

RHR pumps are OPERABLE if they are capable of being powered andare able to provide flow if required.

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

In MODES 5and 6, one Diesel Generator (DG) is required to be OPERABLE per LCO3.8.2, "AC Sources -Shutdown."

The same ESW train is required to becapable of performing its related support function(s) to support DGOPERABILITY.

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

Management of gas voids is important toRHR OPERABILITY.

APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal andcoolant circulation by the RHR System. One RHR loop provides sufficient capability for this purpose.

However, one additional RHR loop is requiredto be OPERABLE to meet single failure considerations.

Operation in other MODES is covered by:LCO 3.4.4, "RCS Loops -MODES 1 and 2";LCO 3.4.5, "RCS Loops -MODE 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.7, "RCS Loops -MODE 5, Loops 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).Wolf Creek -Unit 1 ..- eiin7B 3.4.8-2Revision 72 RCS Loops -MODE 5, Loops Not FilledB 3.4.8BASESAPPLICABILITY Since LCO 3.4.8 contains Required Actions with immediate Completion (continued)

Times, it is not permitted to enter LCO 3.4.8 from either LCO 3.4.7, IRCSLoops -MODE 5, Loops Filled,"

or from MODE 6, unless therequirements of LCO 3.4.8 are met. This precludes removing the heatremoval path afforded by the steam generators with the RHR System isdegraded.

ACTIONS A._.1If only one IRHIR loop is OPERABLE and in operation, redundancy forIRHIR is lost. Action must be initiated to restore a second loop toOPERABLE status. The immediate Completion Time reflects theimportance of maintaining the availability of two paths for heat removal.B.1 and B.2_~I~f n~o required RHRloops are OPERABLE orin operation, except duringconditions permitted by Note 1, all operations involving introduction intothe RCS, coolant with boron concentration less than required to meet theminimum SDM of LCO 3.1.1 must be suspended and action must beinitiated immediately to restore an IRHR loop to OPERABLE status andoperation.

Boron dilution requires forced circulation from at least oneIRCP for proper mixing so that inadvertent criticality can be prevented.

Suspending the introduction into the IRCS, 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 Time reflects the importance of maintaining operation for heat removal.

The action to restore must continue until oneloop is restored to OPERABLE status and operation.

SURVEILLANCE SIR 3.4.8.1REQUIREMENTS This SIR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one loop is in operation.

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications andalarms available to the operator in the control room to monitor IRHR loopperformance.

Wolf Creek -Unit 1B348-Reion2 B3.4.8-3

.... ..... RCS Loops -MODE 5, Loops Not FilledB 3.4.8BASESSURVEILLANCE SR 3.4.8.2REQUIREMENTS (continued)

Verification that a second RHR pump is OPERABLE ensures that anadditional pump can be placed in operation, if needed, to maintain decayheat removal and reactor coolant circulation.

Verification is performed byverifying proper breaker alignment and power available to the RHR pump.The Frequency of 7 days is considered reasonable in view of otheradministrative controls available and has been shown to be acceptable byoperating experience.

SR 3.4.8.3RHR System piping and components have the potential to develop voidsand pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the RHR loops andmay also prevent water hammer, pump cavitation, and pumping ofnoncondensible gas into the reactor vessel.Selection of RHR System locations susceptible to gas accumulation isbased on a review of system design information, including piping andinstrumentation

drawings, isometric

drawings, plan and elevation

drawings, and calculations.

The design review is supplemented bysystem walk downs to validate the system high points and to confirm thelocation and orientation of important components that can becomesources of gas or could otherwise cause gas to be trapped or difficult toremove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versusoperating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.Acceptance criteria are established for the volume of accumulated gas atsusceptible locations.

If accumulated gas is discovered that exceeds theacceptance criteria for the susceptible location (or the volume ofaccumulated gas at one or more susceptible locations exceeds anacceptance criteria for gas volume at the suction or discharge of a pump), -the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas(i.e., the system is sufficiently filled with water), the Surveillance may bedeclared met. Accumulated gas should be eliminated or brought within theacceptance criteria limits.RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location.

Susceptible locations in the same system flowWolf Creek -Unit 1 ..- eiin7B3.4.8-4Revision 72 RCS Loops -MODE 5, Loops Not FilledB 3.4.8BASESSURVEILLANCE SR 3.4.8.3 (continued)

REQUIREMENTS path which are subject to the same gas intrusion mechanisms may beverified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due toradiological or environmental conditions, the plant configuration, orpersonnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location.

Monitoring is not required for susceptible locations where themaximum potential accumulated gas void volume has been evaluated anddetermined to not challenge system OPERABILITY.

The accuracy of themethod used for monitoring the susceptible locations and trending of theresults should be sufficient to assure system OPERABILITY during theSurveillance interval.

The 31 day Frequency takes into consideration the gradual nature of gasaccumulation in the RHR System piping and the procedural controlsgoverning system operation.

REFERENCES

1. USAR, Section 15.4.6.Wolf Creek -Unit 1 ..- eiin7B3.4.8-5Revision 72 Accumulators B 3.5.1BASESAPPLICABLE SAFETY ANALYSES(continued)

The worst case small break LOCA analyses also assume a time delaybefore pumped flow reaches the core. For the larger range of smallbreaks, the rate of blowdown is such that the increase in fuel cladtemperature is terminated primarily by the accumulators, with pumpedflow then providing continued cooling.

As break size decreases, theaccumulators and ECCS pumps play a part in terminating the rise in cladtemperature.

As break size continues to decrease, the role of theaccumulators continues to decrease until they are not required and thecentrifugal charging pumps become solely responsible for terminating thetemperature increase.

This LCO helps to ensure that the following acceptance criteriaestablished for the ECCS by 10 CFR 50.46 (Ref. 2) will be met following aLOCA:a. Maximum fuel element cladding temperature is < 2200°F;b. Maximum cladding oxidation is _< 0.17 times the total cladding_ thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is< 0.01 times the hypothetical amount that would be generated if allof the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; andd. Core is maintained in a coolable geometry.

Since the accumulators empty themselves by the beginning stages of thereflood phase of a LOCA, they do not contribute to the long term coolingrequirements of 10 CFR 50.46.For the small break LOCA analysis, a nominal contained accumulator water volume is used, while the large break LOCA analysis samples theaccumulator water volume over the specified range of 6122 gallons to6594 gallons to allow for instrument inaccuracy.

The contained watervolume is the same as the available deliverable volume for theaccumulators.

For large breaks, an increase in water volume can beeither a peak clad temperature penalty or benefit, depending ondowncomer filling and subsequent spill through the break during the corereflooding portion of the transient.

The analysis credits the line watervolume from the accumulator to the check valve.Wolf Creek -Unit I B 3.5.1-3 Revision 73B 3.5.1-3Revision 73

........Accumulators B 3.5.1BASESAPPLICABLE The minimum boron concentration limit is used in the post LOCA boronSAFETY ANALYSES concentration calculation.

The calculation is performed to assure reactor(continued) subcriticality in a post LOCA environment.

Of particular interest is thelarge break LOCA, since no credit is taken for control rod assemblyinsertion.

A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sumpboron concentration for post LOCA shutdown and an increase in themaximum sump pH. The maximum boron concentration is used indetermining the cold leg to hot leg recirculation injection switchover timeand minimum sump pH.The small break LOCA analysis is performed at the minimum nitrogencover pressure, since sensitivity analyses have demonstrated that highernitrogen cover pressure results in a computed peak clad temperature benefit.

The maximum nitrogen cover Pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity.

Thelarge break LOCA analysis samples the accumulator pressure over therange of 568.1 psig to 681.9 psig.The effects on containment mass and energy releases from theaccumulators are accounted for in the appropriate analyses (Refs. 1and 3).The accumulators satisfy Criterion 2 and Criterion 3 of 10 CFR50.36 (c)(2)(ii).

LCO The LCO establishes the minimum conditions required to ensure that theaccumulators are available to accomplish their core cooling safetyfunction following a LOCA. Four accumulators are required to ensure that100% of the contents of three of the accumulators will reach the coreduring a LOCA. This is consistent with the assumption that the contentsof one accumulator spill through the break. If less than threeaccumulators are injected during the blowdown phase of a LOCA, theECCS acceptance criteria of 10 CFR 50.46 (Ref. 2) could be violated.

For an accumulator to be considered

OPERABLE, the isolation valvemust be fully open, power removed above 1000 psig, and the limitsestablished in the SRs for contained volume, boron concentration, andnitrogen cover pressure must be met.APPLICABILITY In MODES I and 2, and in MODE 3 with RCS pressure

> 1000 psig, theaccumulator OPERABILITY requirements are based on full poweroperation.

Although cooling requirements decrease as power decreases, Wolf Creek -Unit 1 ..- eiin7B 3.5.1-4Revision 73 Accumulators B 3.5.1BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.5.1.2 and SR 3.5.1.3Every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , borated water volume and nitrogen cover pressure areverified for each accumulator.

The limit on borated water volume isequivalent to >_ 30 % and < 70.3 % level. Only one set of non-safety channels (1 of 2) is required for water level and pressure indication.

The12-hour Frequency is sufficient to ensure adequate injection during aLOCA. Because of the static design of the accumulator, a 12 hourFrequency usually allows the operator to identify changes before limits arereached.

Operating experience has shown this Frequency to beappropriate for early detection and correction of off normal trends.SR 3.5.1.4The boron concentration should be verified to be within required limits foreach accumulator every 31 days since the static design of theaccumulators limits the ways in which the concentration can be changed.The 31 day Frequency is adequate to identify changes that could occurfrom mechanisms such as dilution or inleakage.

Sampling the affectedaccumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a 70 gallon increase (approximately 8%level) will identify whether inleakage has caused a reduction in boronconcentration to below the required limit. It is not necessary to verifyboron concentration if the added water inventory is from the refueling water storage tank (RWST) and the RWST has not been diluted sinceverifying that its boron concentration satisfies SR 3.5.4.3, because thewater contained in the RWST is normally within the accumulator boronconcentration requirements.

This is consistent with the recommendation of NUREG-1 366 (Ref. 4).SR 3.5.1.5Verification every 31 days that power is removed from each accumulator isolation valve operator when the RCS pressure is > 1000 psig ensuresthat an active failure could not result in the undetected closure of anaccumulator motor operated isolation valve. If this were to occur, only twoaccumulators would be available for injection given a single failurecoincident with a LOCA. Since power is removed under administrative

control, the 31 day Frequency will provide adequate assurance that poweris removed.This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is < 1000 psig, thus allowing operational Wolf Creek -Unit 1 ..- eiin7B 3.5.1-7Revision 71 Accumulators B 3.5.1BASESSURVEILLANCE REQUIREMENTS SR 3.5.1.5 (continued) flexibility by avoiding unnecessary delays to manipulate the breakersduring plant startups or shutdowns.

Should closure of a valve occur in spite of the interlock, the SI signalprovided to the valves would open a closed valve in the event of a LOCA.REFERENCES

1. USAR, Chapter 6.2. 10OCFR 50.46.3. USAR, Chapter 15.4. NUREG-1 366, February 1990.5. WCAP-1 5049-A, Rev. 1, April 1999.Wolf Creek -Unit 1 ..- RvsoB 3.5.1-8Revision 1

ECCS -Operating B 3.5.2BASESLCO In MODES 1, 2, and 3, two independent (and redundant)

ECCS trains arerequired to ensure that sufficient ECCS flow is available, assuming asingle failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.

In MODES 1, 2, and 3, an ECCS train consists of a centrifugal chargingsubsystem, an SI subsystem, and an RHR subsystem.

Each trainincludes the piping, instruments, and controls to ensure an OPERABLEflow path capable of taking suction from the RWST upon an SI signal andautomatically transferring suction to the containment sump.During an event requiring ECCS actuation, a flow path is required toprovide an abundant supply of water from the RWST to the RCS via theECCS pumps and their respective supply headers to each of the four coldleg injection nozzles.

In the long term, this flow path may be switched totake its supply from the containment sump and to supply its flow to theRCS hot and cold legs. Management of gas voids is important to ECCSOPERABILITY.

The LCO requires the OPERABILITY of a number of independent subsystems.

Due to the redundancy of trains and the diversity ofsubsystems, the inoperability of one component in a train does not renderthe ECCS incapable of performing its function.

Neither does theinoperability of two different components, each in a different train,necessarily result in a loss of function for the ECCS. Reference 6describes situations in which one component, such as an RHR crossover valve, can disable both ECCS trains.During recirculation operation, the flow path for each train must maintainits designed independence to ensure that no single failure can disableboth ECCS trains.As indicated in Note 1, the SI flow paths may be isolated for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inMODE 3, under controlled conditions, to perform pressure isolation valvetesting per SR 3.4.14.1.

The flow path is readily restorable from thecontrol room, and a single active failure is not assumed coincident withthis testing (Ref. 7). Therefore, the ECCS trains are considered OPERABLE during this isolation.

As indicated in Note 2, operation in MODE 3 with ECCS pumps madeincapable of injecting, pursuant to LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System,"

is necessary for plants with anLTOP arming temperature at or near the MODE 3 boundary temperature of 350°F. LCO 3.4.12 requires that certain pumps be rendered incapable of injecting at and below the LTOP arming temperature.

When thistemperature is at or near the MODE 3 boundary temperature, time isneeded to restore the inoperable pumps to OPERABLE status.Wolf Creek -Unit 1 ..- eiin7B 3.5.2-5Revision 72 ECCS -Operating B 3.5.2BASESLCO(continued)

Either of the CCPs may be considered OPERABLE with its associated discharge to RCP seal throttle valve, BG-HV-8357A or BG-HV-8357B, inoperable.

APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for thelimiting Design Basis Accident, a large break LOCA, are based on fullpower operation.

Although reduced power would not require the samelevel of performance, the accident analysis does not provide for reducedcooling requirements in the lower MODES. The centrifugal chargingpump performance is based on a small break LOCA, which establishes the pump performance curve and has less dependence on power. The SIpump performance requirements are based on a small break LOCA.MODE 2 and MODE 3 requirements are bounded by the MODE 1analysis.

This LCO is only applicable in MODE 3 and above. Below MODE 3, thesystem functional requirements are relaxed as described in LCO 3.5.3,"ECCS -Shutdown."

In MODES 5 and 6, plant conditions are such that the probability of anevent requiring ECCS injection is extremely low. Core coolingrequirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled,"

and LCO 3.4.8, "RCS Loops -MODE 5, LoopsNot Filled."

MODE 6 core cooling requirements are addressed byLCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation

-HighWater Level," and LCO 3.9.6, "Residual Heat Removal (RHR) andCoolant Circulation

-Low Water Level."ACTIONSA.__1With one or more trains inoperable, the inoperable components must bereturned to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on an NRC reliability evaluation (Ref. 5) and is areasonable time for repair of many ECCS components.

An ECCS train is inoperable if it is not capable of delivering design flow tothe RCS. Individual components are inoperable if they are not capable ofperforming their design function or supporting systems are not available.

The LCO requires the OPERABILITY of a number of independent subsystems.

Due to the redundancy of trains and the diversity ofsubsystems, the inoperability of one component in a train does not renderWolf Creek -Unit 1 ..- eiin4B 3.5.2-6Revision 42 ECCS -Operating B 3.5.2BASESACTIONS A.1 (continued) the ECCS incapable of performing its function.

Neither does theinoperability of two different components, each in a different train,necessarily result in a loss of function for the ECCS. This allowsincreased flexibility in plant operations under circumstances whencomponents in opposite trains are inoperable.

An event accompanied by a loss of offsite power and the failure of anEDG can disable one ECCS train until power is restored.

A reliability analysis (Ref. 5) has shown that the impact of having one full ECCS traininoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .B.1 and B.2If the inoperable trains cannot be returned to OPERABLE status within theassociated Completion Time, the plant must be brought to a MODE inwhich the LCO does not apply. To achieve this status, the plant must bebrought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Theallowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full powerconditions in an orderly manner and without challenging plant systems.C.1lCondition A is applicable with one or more trains inoperable.

The allowedCompletion Time is based on the assumption that at least 100% of theECCS flow equivalent to a single OPERABLE ECCS train is available.

With less than 100% of the ECCS flow equivalent to a single OPERABLEECCS train available, the unit is in a condition outside of the accidentanalyses.

Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.5.2.1REQUIREMENTS Verification of proper valve position ensures that the flow path from theECCS pumps to the RCS is maintained.

Misalignment of these valvescould render both ECCS trains inoperable.

Securing these valves in thecorrect position by a power lockout isolation device ensures that theycannot change position as a result of an active failure or be inadvertently misaligned.

These valves are of the type, described in References 7 and8, that can disable the function of both ECCS trains and invalidate theaccident analyses.

A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered reasonable in viewof other administrative controls that will ensure a mispositioned valve isunlikely.

Wolf Creek -Unit IB3.27Reion4 B 3.5.2-7Revision 42 ECCS -Operating B 3.5.2BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.2Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flowpaths will exist for ECCS operation.

This SR does not apply to valves thatare locked, sealed, or otherwise secured in position, since these wereverified to be in the correct position prior to locking,

sealing, or securing.

This SR does not apply to manual vent/drain valves, and to valves thatcannot be inadvertently misaligned such as check valves. A valve thatreceives an actuation signal is allowed to be in a nonaccident positionprovided the valve will automatically reposition within the proper stroketime. This Surveillance does not require any testing or valvemanipulation.

Rather, it involves verification that those valves capable ofbeing mispositioned are in the correct position.

The 31 day Frequency isappropriate because the valves are operated under administrative control,and an improper valve position would only affect a single train. ThisFrequency has been shown to be acceptable through operating experience.

The Surveillance is modified by a Note which exempts system vent flowpaths opened under administrative control.

The administrative controlshould be proceduralized and include stationing a dedicated individual atthe system vent flow path who is in continuous communication with theoperators in the control room. This individual will have a method to rapidlyclose the system vent flow path if directed.

SR 3.5.2.3ECCS piping and components have the potential to develop voids andpockets of entrained gases. Preventing and managing gas intrusion andaccumulation is necessary for proper operation of the EGCS and may alsoprevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.Selection of ECCS locations susceptible to gas accumulation is based ona review of system design information, including piping andinstrumentation

drawings, isometric

drawings, plan and elevation

drawings, and calculations.

The design review is supplemented bysystem walk downs to validate the system high points and to confirm thelocation and orientation of important components that can becomesources of gas or could otherwise cause gas to be trapped or difficult toremove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versusoperating conditions.

Wolf Creek -Unit 1 ..- eiin7B 3.5.2-8Revision 72 ECCS -Operating B 3.5.2BASESSURVEILLANCE SR 3.5.2.3 (continued)

REQUIREMENTS The ECCS is OPERABLE when it is sufficiently filled with water.Acceptance criteria are established for the volume of accumulated gas atsusceptible locations.

In conjunction with or in lieu of venting, Ultrasonic Testing (UT) may be performed to verify the ECCS pumps and associated piping are sufficiently full of water. The design of the centrifugal chargingpump is such that significant noncondensible gases do not collect in thepump. Therefore, it is unnecessary to require periodic pump casingventing to ensure the centrifugal charging pump will remain OPERABLE.

If accumulated gas is discovered that exceeds the acceptance criteria forthe susceptible location (or the volume of accumulated gas at one or moresusceptible locations exceeds an acceptance criteria for gas volume atthe suction or discharge of a pump), the Surveillance is not met. If it isdetermined by subsequent evaluation that the ECCS is not renderedinoperable by the accumulated gas (i.e., the system is sufficiently filledwith water), the Surveillance may be declared met. Accumulated gasshould be eliminated or brought within the acceptance criteria limits.ECCS locations susceptible to gas accumulation are monitored and, if gasis found, the gas volume is compared to the acceptance criteria for thelocation.

Susceptible locations in the same system flow path which aresubject to the same gas intrusion mechanisms may be verified bymonitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety.For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location.

Monitoring is not required for susceptible locations where the maximumpotential accumulated gas void volume has been evaluated anddetermined to not challenge system OPERABILITY.

The accuracy of themethod used for monitoring the susceptible locations and trending of theresults should be sufficient to assure system OPERABILITY during theSurveillance interval.

The 92 day Frequency takes into consideration the plant specific nature ofgas accumulation in the ECCS piping and the procedural controlsgoverning system operation.

Wolf Creek -Unit 1 ..- eiin7B 3.5.2-9 ECCS -Operating B 3.5.2BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.4Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may beaccomplished by measuring the pump developed head at only one pointof the pump characteristic curve. The following ECCS pumps arerequired to develop the indicated differential pressure on recirculation flow:Centrifugal Charging PumpSafety Injection PumpRHR Pump> 2490 psid>_ 1468.9 psid>_ 183.6 psidThis verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that theperformance at the test flow is greater than or equal to the performance assumed in the plant safety analysis.

SRs are specified in the applicable portions of the Inservice Testing Program, which encompasses the ASMECode. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.

SR 3.5.2.5 and SR 3.5.2.6These Surveillances demonstrate that each automatic ECCS valveactuates to the required position on an actual or simulated SI signal andon an actual or simulated RWST Level Low-Low I Automatic Transfersignal coincident with an SI signal and that each ECCS pump starts onreceipt of an actual or simulated SI signal. This Surveillance is notrequired for valves that are locked, sealed, or otherwise secured in therequired position under administrative controls.

The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned planttransients if the Surveillances were performed with the reactor at power.The 18 month Frequency is also acceptable based on consideration of thedesign reliability (and confirming operating experience) of the equipment.

The actuation logic is tested as part of ESF Actuation System testing, andequipment performance is monitored as part of the Inservice TestingProgram.Wolf Creek -Unit 1 ..-0Reiin7B 3.5.2-10 ECCS -Operating B 3.5.2BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.7The position of throttle valves in the flow path is necessary for properECCS performance.

These valves are necessary to restrict flow to aruptured cold leg, ensuring that the other cold legs receive at least therequired minimum flow. The 18 month Frequency is based on the samereasons as those stated in SR 3.5.2.5 and SR 3.5.2.6.

The ECCS throttlevalves are set to ensure proper flow resistance and pressure drop in thepiping to each injection point in the event of a LOCA. Once set, thesethrottle valves are secured with locking devices and mechanical positionstops. These devices help to ensure that the following safety analysesassumptions remain valid: (1) both the maximum and minimum totalsystem resistance; (2) both the maximum and minimum branch injection line resistance; and (3) the maximum and minimum ranges of potential pump performance.

These resistances and pump performance rangesare used to calculate the maximum and minimum ECCS flows assumed inthe LOCA analyses of Reference 3.SR 3.5.2.8This SR requires verification that each ECCS train containment sump inletis not restricted by debris and the suction inlet strainers show no evidenceof structural distress or abnormal corrosion.

A visual inspection of thesuction inlet piping verifies the piping is unrestricted.

A visual inspection of the accessible portion of the containment sump strainer assemblyverifies no evidence of structural distress or abnormal corrosion.

Verification of no evidence of structural distress ensures there are noopenings in excess of the maximum designed strainer opening.

The 18month Frequency has been found to be sufficient to detect abnormaldegradation and is confirmed by operating experience.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 35.2. 10 CFR 50.46.3. USAR, Sections 6.3 and 15.6.4. USAR, Chapter 15, "Accident Analysis."

5. NRC Memorandum to V. Stello, Jr., from R.L. Baer,"Recommended Interim Revisions to LCOs for ECCS Components,"

December 1, 1975.6. IE Information Notice No. 87-01.Wolf Creek -Unit 1 B3521 eiin7B 3.5.2-11 ECCS -Operating B 3.5.2BASESREFERENCES

7. BTP EICSB-18, Application of the Single Failure Criteria to(continued)

Manually-Controlled Electrically-Operated Valves.8. WCAP-9207, "Evaluation of Mispositioned ECCS Valves,"September 1977.Wolf Creek -Unit 1 ..-2Reiin7B 3.5.2-12 ECCS -ShutdownB 3.5.3B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)B 3.5.3 ECCS -ShutdownBASESBACKGROUND The Background section for Bases 3.5.2, "ECCS -Operating,"

isapplicable to these Bases, with the following modifications.

In MODE 4, the required ECCS train consists of two separatesubsystems:

centrifugal charging (high head) and residual heat removal(RHR) (low head).The ECCS flow paths consist of piping, valves, heat exchangers, andpumps such that water from the refueling water storage tank (RWST) canbe injected into the Reactor Coolant System (RCS) following theaccidents described in Bases 3.5.2.APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also appliesSAFETY ANALYSES to this Bases section.Due to the stable conditions associated with operation in MODE 4 and thereduced probability of occurrence of a Design Basis Accident (DBA), theECCS operational requirements are reduced.

It is understood in thesereductions that certain automatic safety injection (SI) actuation is notavailable.

In this MODE, sufficient time exists for manual actuation of therequired ECCS to mitigate the consequences of a DBA.For MODE 3, with the accumulators

blocked, and MODE 4, theparameters assumed in the generic bounding thermal hydraulic analysisfor the limiting DBA (Reference

1) are based on a combination of limitingparameters for MODE 3, with the accumulators

blocked, and parameters for MODE 4. However, assumed ECCS availability is based on MODE 4conditions; the minimum available ECCS flow is calculated assuming onlyone OPERABLE ECCS train.Only one tr'ain-of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE ofoperation.

The ECCS trains satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO In MODE 4, one of the two independent (and redundant)

ECCS trains isrequired to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a DBA.Wolf Creek -Unit 1 ..- eiin5B3.5.3-1Revision 56

.. .." ...' ....EGCS -ShutdownB 3.5.3BASESLCO In MODE 4, an EGGS train consists of a centrifugal charging subsystem (continued) and an RHR subsystem.

Each train includes the piping, instruments, andcontrols to ensure an OPERABLE flow path capable of taking suctionfrom the RWST and transferring suction to the containment sump.During an event requiring ECGS actuation, a flow path is required toprovide an abundant supply of water from the RWST to the RCS via theEGGS pumps and their respective supply headers to two cold leg injection nozzles.

In the long term, this flow path may be switched to take itssupply from the containment sump and to deliver its flow to the RCS hotand cold legs. Management of gas voids is important to ECCSOPERABILITY.

This LCO is modified by a Note that allows an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, ifcapable of being manually realigned (remote or local) to the ECCS modeof operation and not otherwise inoperable.

This allows operation in theRHR mode during MODE 4. Only one RHR train is placed into operation to reduce RGS temperature.

For an RHR train to be considered OPERABLE during shutdown, the train cannot be placed in service untilRCS temperature is less than 225 0F (plant computer)/21 5 0F (maincontrol board). For an RHR train to be considered OPERABLE duringstartup, the train must be isolated from the RCS prior to RCS temperature exceeding 225 0F (plant computer)/215

°F (main control board).APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for EGGS arecovered by LCO 3.5.2.In MODE 4 with RCS temperature below 350°F, one OPERABLE EGGStrain is acceptable without single failure consideration, on the basis of thestable reactivity of the reactor and the limited core cooling requirements.

In MODES 5 and 6, plant conditions are such that the probability of anevent requiring EGGS injection is extremely low. Gore coolingrequirements in MODE 5 are addressed by LGO 3.4.7, "RGS Loops -MODE 5, Loops Filled,"

and LCO 3.4.8, "RGS Loops -MODE 5, LoopsNot Filled."

MODE 6 core cooling requirements are addressed byLGO 3.9.5, "Residual Heat Removal (RHR) and Goolant Girculation

-HighWater Level," and LGO 3.9.6, "Residual Heat Removal (RHR) andGoolant Girculation

-Low Water Level."AGTIONS A Note prohibits the application of LGO 3.0.4b. to an inoperable EGGScentrifugal charging pump subsystem when entering MODE 4. There isan increased risk associated with entering MODE 4 from MODE 5 with anWolf Greek -Unit 1 ..- eiin7B 3.5.3-2Revision 72 Containment Spray and Cooling SystemsB 3.6.6BASESBACKGROUND Containment Coolinq System (continued)

In post accident operation following an actuation signal, the Containment Cooling System fans are designed to start automatically in slow speed ifnot already running.

If running in high (normal) speed, the fansautomatically shift to slow speed. The fans are operated at the lowerspeed during accident conditions to prevent motor overload from thehigher mass atmosphere.

The temperature of the ESW is an important factor in the heat removal capability of the fan units.APPLICABLE The Containment Spray System and Containment Cooling System limitsSAFETY ANALYSES the temperature and pressure that could be experienced following a DBA.The limiting DBAs considered are the loss of coolant accident (LOCA)and the steam line break (SLB). The LOCA and SLB are analyzed usingcomputer codes designed to predict the resultant containment pressureand temperature transients.

No DBAs are assumed to occursimultaneously or consecutively.

The postulated DBAs are analyzed withregards to containment ESF systems, assuming the loss of one ESE bus,which is the worst case single active failure and results in one train of theContainment Spray System and Containment Cooling System beingrendered inoperable.

The analysis and evaluation show that under the worst case scenario, thehighest peak containment pressure is 51.5 psig and the peak containment temperature is 360.0°F (experienced during an SLB). Both results meetthe intent of the design basis. (See the Bases for LCO 3.6.4,"Containment Pressure,"

and LCO 3.6.5 for a detailed discussion.)

Theanalyses and evaluations assume a unit specific power level ranging to102%, one containment spray train and one containment cooling trainoperating, and initial (pre-accident) containment conditions of 120°F and0 psig. The analyses also assume a response time delayed initiation toprovide conservative peak calculated containment pressure andtemperature responses.

For certain aspects of transient accident

analyses, maximizing thecalculated containment pressure is not conservative.

In particular, theeffectiveness of the Emergency Core Cooling System during the corereflood phase of a LOCA analysis increases with increasing containment backpressure.

For these calculations, the containment backpressure iscalculated in a manner designed to conservatively

minimize, rather thanmaximize, the calculated transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 2).The effect of an inadvertent containment spray actuation has beenanalyzed.

An inadvertent spray actuation results in a -2.72 psigcontainment pressure and is associated with the sudden cooling effect inthe interior of the leak tight containment.

Additional discussion isprovided in the Bases for LCO 3.6.4.Wolf Creek -Unit 1B366-Reion7 B 3.6.6-3Revision 37

--Containment SI5ray and Cooling SystemsB 3.6.6BASESAPPLICABLE The modeled Containment Spray System actuation from the containment SAFETY ANALYSES analysis is based on a response time associated with exceeding the(continued) containment High-3 pressure setpoint to achieving full flow through thecontainment spray nozzles.The Containment Spray System total response time includes dieselgenerator (DG) startup (for loss of offsite power), sequenced loading ofequipment, containment spray pump startup, and spray line filling (Ref. 4).Containment cooling .train performance for post accident conditions isgiven in Reference

4. The result of the analysis is that each train canprovide 100% of the required peak cooling capacity during the postaccident condition.

The train post accident cooling capacity under varyingcontainment ambient conditions, required to perform the accidentanalyses, is also shown in Reference 4.The modeled Containment Cooling System actuation from thecontainment analysis is based upon a response time associated withreceipt of an SI signal to achieving full Containment Cooling System airand safety grade cooling water flow. The Containment Cooling Systemtotal response time of 70 seconds, includes signal delay, OG startup (forloss of offsite power), and Essential Service Water pump startup timesand line refill (Ref. 4).The Containment Spray System and the Containment Cooling Systemsatisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO During a DBA, a minimum of one containment cooling train and onecontainment spray train is required to maintain the containment peakpressure and temperature below the design limits (Ref. 3). Additionally, one containment spray train is also required to remove iodine from thecontainment atmosphere and maintain concentrations below thoseassumed in the safety analysis.

With the Spray Additive Systeminoperable, a containment spray train is still available and would removesome iodine from the containment atmosphere in the event of a DBA. Toensure that these requirements are met, two containment spray trains andtwo containment cooling trains must be OPERABLE.

Therefore, in theevent of an accident, at least one train in each system operates, assumingthe worst case single active failure occurs.Each Containment Spray System typically includes a spray pump, sprayheaders,

eductor, nozzles, valves, piping, instruments, and controls toensure an OPERABLE flow path capable of taking suction from theRWST upon an ESF actuation signal and manually transferring to thecontainment sump. Management of gas voids is important toContainment Spray System OPERABILITY.

A containment cooling train typically includes cooling coils, dampers, twofans, instruments, and controls to ensure an OPERABLE flow path.Wolf Creek- Unit 1 ..- eiin7B 3.6.6-4Revision 72 Containment Spray and Cooling SystemsB 3.6.6BASESACTIONS F.1(continued)

With two containment spray trains or any combination of three or morecontainment spray and cooling trains inoperable, the unit is in a condition outside the accident analysis.

Therefore, LCO 3.0.3 must be enteredimmediately.

SURVEILLANCE SR 3.6.6.1REQUIREMENTS Verifying the correct alignment' for manual, power operated, andautomatic valves in the containment spray flow path provides assurance that the proper flow paths will exist for Containment Spray Systemoperation.

The correct alignment for the Containment Cooling Systemvalves is provided in SR 3.7.8.1.

This SR does not apply to manualvent/drain valves and to valves that cannot be advertently misaligned such as check valves. This SR does not apply to valves that are locked,sealed, or otherwise secured in position, since these were verified to be inthe correct position prior to locking,

sealing, or securing.

This SR doesnot require any testing or valve manipulation.

Rather, it involves

.....verification, through a system walkdown (which may include the use oflocal or remote indicators),

that those valves outside containment andcapable of potentially being mispositioned are in the correct position.

The31 day Frequency is based on engineering judgement, is consistent withadministrative controls governing valve operation, and ensures correctvalve positions.

The Surveillance is modified by a Note which exempts system vent flowpaths opened under administrative control.

The administrative controlshould be proceduralized and include stationing a dedicated individual atthe system vent flow path who is in continuous communication with theoperators in the control room. This individual will have a method to rapidlyclose the system vent flow path if directed.

SR 3.6.6.2Operating each containment cooling train fan unit for > 15 minutes -ensures that all fan units are OPERABLE.

It also ensures the abnormalconditions or degradation of the fan unit can be detected for corrective action. The 31 day Frequency was developed considering the knownreliability of the fan units and controls, the two train redundancy available, and the low probability of significant degradation of the containment cooling train occurring between surveillances.

It has also been shown tobe acceptable through operating experience.

SR 3.6.6.3 Not Used.Wolf Creek -Unit IB366-Reion7 B3.6.6-7Revision 72

... Containment Spray and Cooling SystemsB 3.6.6BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.6.6.4Verifying each containment spray pump's developed head at the flow testpoint is greater than or equal to the required developed head ensures thatspray pump performance has not degraded during the cycle. Flow anddifferential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 5). Since the containment spraypumps cannot be tested with flow through the spray headers, they aretested on recirculation flow. This test confirms one point on the pumpdesign curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and detectincipient failures by abnormal performance.

The Frequency of the SR isin accordance with the Inservice Testing Program.This test ensures that each pump develops a differential pressure ofgreater than or equal to 219 psid at a nominal flow of 300 gpm when onrecirculation (Ref. 6).SR 3.6.6.5 and SR 3.6.6.6These SRs require verification that each automatic containment sprayvalve actuates to its correct position and that each containment spraypump starts upon receipt of an actual or simulated actuation of acontainment High-3 pressure signal. This Surveillance is not required forvalves that are locked, sealed, or otherwise secured in the requiredposition under administrative controls.

The 18 month Frequency is basedon the need to perform these Surveillances under the conditions thatapply during a plant outage and the potential for an unplanned transient ifthe Surveillances were performed with the reactor at power. Operating experience has shown that these components usually pass theSurveillances when performed at the 18 month Frequency.

Therefore, theFrequency was concluded to be acceptable from a reliability standpoint.

The surveillance of containment sump isolation valves is also required bySR 3.5.2.5.

A single surveillance may be used to satisfy bothrequirements.

SR 3.6.6.7This SR requires verification that each containment cooling train actuatesupon receipt of an actual or simulated safety injection signal. Uponactuation, each fan in the train starts in slow speed or, if operating, shiftsto slow speed and the Cooling water flow rate increases to _> 2000 gpm toeach cooler train. The 18 month Frequency is based on engineering judgment and has been shown to be acceptable through operating experience.

See SR 3.6.6.5 and SR 3.6.6.6, above, for further discussion of the basis for the 18 month Frequency.

Wolf Creek -Unit I1 ..- eiin7B 3.6.6-8 Containment Spray and Cooling SystemsB 3.6.6BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.6.6.8With the containment spray inlet valves closed and the spray headerdrained of any solution, low pressure air or smoke can be blown throughtest connections.

This SR ensures that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during anaccident is not degraded.

Due to the passive design of the nozzle, aconfirmation of OPERABILITY following maintenance activities that canresult in obstruction of spray nozzle flow is considered adequate to detectobstruction of the nozzles.

Confirmation that the spray nozzles areunobstructed may be obtained by utilizing foreign material exclusion (FME) controls during maintenance, a visual inspection of the affectedportions of the system, or by an air or smoke flow test following maintenance involving opening portions of the system downstream of thecontainment isolation valves or draining of the filled portions of the systeminside containment.

Maintenance that could result in nozzle blockage isgenerally a result of a loss of foreign material control or a flow of boratedwater through a nozzle. Should either of these events occur, asupervisory evaluation will be required to determine whether nozzleblo0ckage is a possible result of the event. For the loss of FME event, aninspection or flush of the affected portions of the system should beadequate to confirm that the spray nozzles are unobstructed since waterflow would be required to transport any debris to the spray nozzles.

An airflow or smoke test may not be appropriate for a loss of FME event butmay be appropriate for the case where borated water inadvertently flowsthrough the nozzles.SR 3.6.6.9Containment Spray System piping and components have the potential todevelop voids and pockets of entrained gases. Preventing and managinggas intrusion and accumulation is necessary for proper operation of thecontainment spray trains and may also prevent water hammer and pumpcavitation.

Selection of Containment Spray System locations susceptible to gasaccumulation is based on a review of system design information, including piping and instrumentation

drawings, isometric

drawings, plan andelevation

drawings, and calculations.

The design review is supplemented by system walk downs to validate the system high points and to confirmthe location and orientation of important components that can becomesources of gas or could otherwise cause gas to be trapped or difficult toremove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versusoperating conditions.

Wolf Creek -Unit I B 3.6.6-9 Revision 72B 3.6.6-9Revision 72

'"; ......

Sprayi and Cooling SystemsB 3.6.6BASESSURVEILLANCE SR 3.6.6.9 (continued)

REQUIREMENTS The Containment Spray System is OPERABLE when it is sufficiently filledwith water. Acceptance criteria are established for the volume ofaccumulated gas at susceptible locations.

If accumulated gas isdiscovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction ordischarge of a pump), the Surveillance is not met. If it is determined bysubsequent evaluation that the Containment Spray System is notrendered inoperable by the accumulated gas (i.e., the system issufficiently filled with water), the Surveillance may be declared met.Accumulated gas should be eliminated or brought within the acceptance criteria limits.Containment Spray System locations susceptible to gas accumulation aremonitored and, if gas is found, the gas volume is compared to theacceptance criteria for the location.

Susceptible locations in the samesYstem flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that areinaccessible due to radiological or environmental conditions, the plantconfiguration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used tomonitor the susceptible location.

Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume hasbeen evaluated and determined to not challenge system OPERABILITY.

The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure systemOPERABILITY during the Surveillance interval.

The 92 day Frequency takes into consideration the plant specific nature ofgas accumulation in the Containment Spray System piping and theprocedural controls governing system operation.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41. GDC42, and GDC 43, and GDC 50.2. 10 CFR 50, Appendix K.3. USAR, Section 6.2.1.4. USAR, Section 6.2.2.5. ASME Code for Operation and Maintenance of Nuclear PowerPlants.6. Performance Improvement Request 2002-0945.

Wolf Creek- Unit 1B 3.6.6-10Revision 72 AC Sources -Operating B 3.8.1BASESAPPLICABLE meeting the design basis of the unit. This results in maintaining at leastSAFETY ANALYSES one train of the onsite or offsite AC sources OPERABLE during Accident(continued) conditions in the event of:a. An assumed loss of all offsite power or all onsite AC power; andb. A worst case single failure.The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two qualified circuits between the offsite transmission network and theonsite Class 1 E Electrical Power System, separate and independent DGsfor each train, and redundant LSELS for each train ensure availability ofthe required power to shut down the reactor and maintain it in a safeshutdown condition after an anticipated operational occurrence (AOO) ora postulated DBA.Each offsite circuit must be capable of maintaining rated frequency andvoltage, and accepting required loads during an accident, while connected to the ESF buses.One offsite circuit consists of the #7 transformer feeding through the13-48 breaker power the ESE transformer XNB01, which, in turn powersthe NB01 bus through its normal feeder breaker.

Transformer XNB01may also be powered from the SL-7 supply through the 13-8 breakerprovided the offsite 69 Ky line is not connected to the 345 kV system.The offsite circuit energizing NB01 is considered inoperable when theEast 345 kV bus is only energized from the transmission network throughthe 345-50 and 345-60 main generator breakers.

For this configuration, switchyard breakers 345-120 and 345-90 OR 345-120 and 345-80 areopen.Another offsite circuit consists of the startup transformer feeding throughbreaker PA201 powering the ESF transformer XNB02, which, in turnpowers the NB02 bus through its normal feeder breaker.Each DG must be capable of starting, accelerating to rated speed andvoltage, and connecting to its respective ESF bus on detection of busundervoltage.

This will be accomplished within 12 seconds.

Each DGmust also be capable of accepting required loads within the assumedloading sequence intervals, and continue to operate until offsite powercan be restored to the ESF buses. These capabilities are required to bemet from a variety of initial conditions such as DG in standby with theengine hot and DG in standby with the engine at ambient conditions.

Additional DG capabilities must be demonstrated to meet requiredSurveillance, e.g., capability of the DG to revert to standby status on anECCS signal while operating in parallel test mode.Wolf Creek -Unit 1 ..- eiin4B 3.8.1-3Revision 47 AC sources -Operating B 3.8.1BASESLCO Upon failure of the DG lube oil keep warm system when the DO is in the(continued) standby condition, the DO remains OPERABLE if lube oil temperature is> 115 0F and engine lubrication (i.e., flow of lube oil to the DO engine) ismaintained.

Upon failure of the DG jacket water keep warm system, theDG remains OPERABLE as long as jacket water temperature is _> 105 °F(Ref. 13).Initiating an EDO start upon a detected undervoltage or degraded voltagecondition, tripping of nonessential loads, and proper sequencing of loads,is a required function of LSELS and required for DO OPERABILtITY.

Inaddition, the LSELS Automatic Test Indicator (ATI) is an installed testingaid and is not required to be OPERABLE to support the sequencer function.

Absence of a functioning ATI does not render LSELSinoperable.

The AC sources in one train must be separate and independent of the ACsources in the other train. For the D~s, separation and independence arecomplete.

For the offsite AC source, separation and independence are tothe extent practical.

-APPLICABILITY The AC sources and LSELS are required to be OPERABLE in MODES 1,2, 3, and 4 to ensure that:a. Acceptable fuel design limits and reactor coolant pressureboundary limits are not exceeded as a result of AOOs or abnormaltransients; andb. Adequate core cooling is provided and containment OPERABILITY and other vital functions are maintained in the eventof a postulated DBA.The AC power requirements for MODES 5 and 6 are covered inLCO 3.8.2, "AC Sources -Shutdown."

ACTIONS A Note prohibits the application of LCO 3.0.4b. to an inoperable DG.There is an increased risk associated with entering a MODE or otherspecified condition in the Applicability with an inoperable DO and theprovisions of LCO 3.0.4b.,

which allow entry into a MODE or otherspecified condition in the Applicability with the LCO not met afterperformance of a risk assessment addressing inoperable systems andcomponents, should not be applied in this circumstance.

Wolf Creek- Unit 1 ..- eiin7B 3.8.1-4Revision 71 AC Sources -Operating B 3.8.1BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.21SR 3.8.1.21 is the performance of an ACTUATION LOGIC TEST usingthe LSELS automatic tester for each load shedder and emergency loadsequencer train except that the continuity check does not have to beperformed, as explained in the Note. This test is performed every 31 dayson a STAGGERED TEST BASIS. The Frequency is adequate based onindustry operating experience, considering instrument reliability andoperating history data.REFERENCES 1.2.3.4.5.6.7.10 CFR 50, Appendix A, GDC 17.USAR, Chapter 8.Regulatory Guide 1.9, Rev. 3.USAR, Chapter 6.USAR, Chapter 15.Regulatory Guide 1.93, Rev. 0, December 1974.Generic Letter 84-15, "Proposed Staff Actions toImprove and Maintain Diesel Generator Reliability,"

July 2, 1984.10 CFR 50, Appendix A, GDC 18.Regulatory Guide 1.108, Rev. 1, August 1977.Regulatory Guide 1.137, Rev. 0, January 1978.ANSI C84.1-1 982.IEEE Standard 308-1978.

Configuration Change Package (CCP) 08052, Revision 1, April 23,1999.8.9.10.11.12.13.14.15.16.17.Amendment No. 161, April 21, 2005.Not used.Amendment No. 163, April 26, 2006.Amendment No. 154, August 4, 2004.Wolf Creek -Unit 1 B3813 eiin7B 3.8.1-33Revision 71 AC Sou~rces

-Operating B 3.8.1BASESREFERENCES (continued)

18. Amendment No. 8, May 29, 1987.19. Condition Report 15727.Woif Creek -Unit 1 ..-4 eiin4B 3.8.1-34Revision 47 Inverters

-Operating B 3.8.7B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.7 Inverters

-Operating BASESBACKGROUND The inverters are the preferred source of power for the AC vital busesbecause of the stability and reliability they achieve.

The function of theinverter is to provide AC electrical power to the vital buses. The inverters are normally powered from the respective 125 VDC bus. An alternate source of power to the AC vital buses is provided from Class 1 E bypassconstant voltage transformers.

The battery bus provides anuninterruptible power source for the instrumentation and controls for theReactor Protection System (RPS) and the Engineered Safety FeatureActuation System (ESFAS).

There are two required inverters per train.Two spare inverters (one per train) are provided for alignment to the 120VAC vital bus when an associated inverter is taken out of service.

If thespare inverter is placed in service, requirements of independence andredundancy between trains are maintained.

Specific details on inverters and their operating characteristics are found in the USAR, Chapter 8(Ref. 1).APPLICABLE SAFETY ANALYSESThe initial conditions of Design Basis Accident (DBA) and transient analyses in the USAR, Chapter 6 (Ref. 2) and Chapter 15 (Ref. 3),assume Engineered Safety Feature systems are OPERABLE.

Theinverters are designed to provide the required

capacity, capability, redundancy, and reliability to ensure the availability of necessary power tothe RPS and ESFAS instrumentation and controls so that the fuel,Reactor Coolant System, and containment design limits are notexceeded.

These limits are discussed in more detail in the Bases forSection 3.2, Power Distribution Limits; Section 3.4, Reactor CoolantSystem (RCS); and Section 3.6, Containment Systems.The OPERABILITY of the inverters is consistent with the initialassumptions of the accident analyses and is based on meeting the designbasis of the unit. This includes maintaining required AC vital busesOPERABLE during accident conditions in the event of:a. An assumed loss of all offsite AC electrical power or all onsite ACelectrical power; andb. A worst case single failure.Inverters satisfy Criterion 3 of the 10 CFR 50.36(c)(2)(ii).

Wolf Creek- Unit 1 ..- eiin6B 3.8.7-1Revision 69 Inverters

-" Operating B 3.8.7BASESLCOThe inverters ensure the availability of AC electrical power for the systemsinstrumentation required to shut down the reactor and maintain it in a safecondition after an anticipated operational occurrence (AQO) or apostulated DBA.Maintaining the required inverters OPERABLE ensures that theredundancy incorporated into the design of the RPS and ESFASinstrumentation and controls is maintained.

The four inverters (two pertrain) ensure an uninterruptible supply of AC electrical power to the ACvital buses even if the 4.16 kV safety buses are de-energized.

OPERABLE inverters require the associated vital bus to be powered bythe inverter with output voltage within tolerances, and power input to theinverter from the 125 VDC battery bus of the same separation group.The required inverters/AC vital buses are associated with the AC loadgroup subsystems (Train A and Train B) as follows:TRAIN A TRAIN BBus NN01 Bus NN03 Bus NN02 Bus NN04energized from energized from energized from energized fromInvert. NN11 Invert. NN13 Invert. NN12 Invert. NN14orNNl15 or NN 15 or NNl16 or NNl16connected to connected to connected to connected toDC bus NK01 DC bus NK03 DC bus NK02 DC bus NK04APPLICABILITY The inverters are required to be OPERABLE in MODES 1, 2, 3, and 4 toensure that:a. Acceptable fuel design limits and reactor coolant pressureboundary limits are not exceeded as a result of AOOs or abnormaltransients; andb. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.Inverter requirements for MODES 5 and 6 are covered in the Bases forLCO 3.8.8, "Inverters

-Shutdown."

Wolf Creek -Unit 1 ..- eiin6B 3.8.7-2Revision 69 Inverters

-Operating B 3.8.7BASESACTIONS A.1With a required inverter inoperable, its associated AC vital bus isinoperable until it is re-energized from its bypass constant voltagetransformer or the bypass constant voltage transformer of the respective spare inverter.

The bypass constant voltage transformers are poweredfrom a Class 1 E bus.For this reason a Note has been included in Condition A requiring theentry into the Conditions and Required Actions of LCO 3.8.9, "Distribution Systems -Operating,"

with any vital bus de-energized.

This ensures thatthe vital bus is re-energized within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .Required Action A.1 allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to fix the inoperable inverter or placethe associated train spare inverter in service.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months limit is basedupon engineering

judgment, taking into consideration the time required torepair an inverter and the additional risk to which the unit is exposedbecause of the inverter inoperability.

This has to be balanced against therisk of an immediate

shutdown, along with the potential challenges tosafety systems such a shutdown might entail. When the AC vital bus ispowered from its bypass constant voltage transformer, it is relying uponinterruptible AC electrical power sources (offsite and onsite).

Theuninterruptible inverter source to the AC vital buses is the preferred source for powering instrumentation trip setpoint devices.B.1 and B.2If the inoperable devices or components cannot be restored toOPERABLE status within the required Completion Time, the unit must bebrought to a MODE in which the LCO does not apply. To achieve thisstatus, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and toMODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions fromfull power conditions in an orderly manner and without challenging plantsystems.SURVEILLANCE SR 3.8.7.1REQUIREMENTS This Surveillance verifies that the inverters are functioning properly withall required circuit breakers closed and AC vital buses energized from theinverter.

The verification of proper voltage output ensures that therequired power is readily available for the instrumentation of the RPS andESFAS connected to the AC vital buses. The 7 day Frequency takes intoaccount the redundant capability of the inverters and other indications available in the control room that alert the operator to invertermalfunctions.

Wolf Creek -Unit 1 ..- eiin6B 3.8.7-3Revision 69 Inverter's

-Operating B 3.8.7BASESREFERENCES

1. USAR, Chapter 8.2. USAR, Chapter 6.3. USAR, Chapter 15.Wolf Creek -Unit 1 B3874Rvso B3.8.7-4Revision 0

Inverters

-ShutdownB 3.8.8BASESAPPLICABLE SAFETY ANALYSES(continued) distribution systems are available and reliable.

When portions of theClass 1 E power or distribution systems are not available (usually as aresult of maintenance or modifications),

other reliable power sources ordistribution are used to provide the needed electrical support.

The plantstaff assesses these alternate power sources and distribution systems toassure that the desired level of minimal risk is maintained (frequently referred to as maintaining a desired defense in depth). The level of detailinvolved in the assessment will vary with the significance of the equipment being supported.

In some cases, prepared guidelines are used whichinclude controls designed to manage risk and retain the desired defensein depth.The inverters satisfy Criterion 3 of the 10 CFR 50.36(c)(2)(ii).

LCOOne train of inverters is required to be OPERABLE to support one train ofthe onsite Class 1 E AC vital bus electrical power distribution subsystems required by LCO 3.8.10, "Distribution Systems -Shutdown."

The requiredtrain of inverters (Train A or Train B) consists of two AC vital busesenergized from the associated inverters with each inverter connected tothe respective DC bus. Each train includes one spare inverter that can bealigned to power either AC vital bus in its associated load group. Eachspare inverter shall be powered from the 125 VDC bus in the separation group to which the spare inverter is connected.

The inverters ensure theavailability of electrical power for the instrumentation for systems requiredto shut down the reactor and maintain it in a safe condition after ananticipated operational occurrence or a postulated DBA. The batterypowered inverters provide uninterruptible supply of AC electrical power tothe AC vital buses even if the 4.16 kV safety buses are de-energized.

OPERABILITY of the inverters requires that the AC vital bus be poweredby the inverter.

This ensures the availability of sufficient inverter powersources to operate the unit in a safe manner and to mitigate theconsequences of postulated events during shutdown (e.g., fuel handlingaccidents).

The required AC vital bus electrical power distribution subsystem issupported by one train of inverters.

When the second (subsystem) of ACvital bus electrical power distribution is needed to support redundant required

systems, equipment and components, the second train may beenergized from any available source. The available source must be Class1 E or another reliable source. The available source must be capable ofsupplying sufficient AC electrical power such that the redundant components are capable of performing their specified safety function(s)

(implicitly required by the definition of OPERABILITY).

Otherwise, thesupported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered.Wolf Creek -Unit 1B388-Reion6 B3.8.8-3Revision 69 Inverters

-ShutdownB 3.8.8BASESAPPLICABILITY The inverters required to be OPERABLE in MODES 5 and 6 provideassurance that:a. Systems to provide adequate coolant inventory makeup areavailable for the irradiated fuel in the core;b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary to mitigate the effects of events that can lead tocore damage during shutdown are available; andd. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

Inverter requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.7.ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, sinceirradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, theACTIONS have been modified by a Note stating that LCO 3.0.3 is notapplicable.

If moving irradiated fuel assemblies while in MODE 5 or 6,LCO 3.0.3 would not specify any action. If moving irradiated fuelassemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations.

Entering LCO 3.0.3, while in MODE 1, 2, 3, or 4would require the unit to be shutdown unnecessarily.

A.1, A.2.1. A.2.2. A.2.3. and A.2.4By the allowance of the option to declare required features inoperable with the associated inverter(s) inoperable, appropriate restrictions will beimplemented in accordance with the affected required features LCOs'Required Actions.

In many instances, this option may involve undesired administrative efforts.

Therefore, the allowance for sufficiently conservative actions is~made-(i.e.,

to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies, and operations involving positivereactivity additions that could result in loss of required SDM (MODE 5) ofLCO 3.1.1 or boron concentration (MODE 6) of LCO 3.9.1). Suspending positive reactivity additions that could result in failure to meet the minimumSDM or boron concentration limit is required to assure continued safeoperation.

Introduction of coolant inventory must be from sources thathave a boron concentration greater than that required in the RCS forminimum SDM or refueling boron concentration.

This may result in anoverall reduction in RCS boron concentration, but provides acceptable Wolf Creek -Unit 1B388-Reion5 B 3.8.8-4Revision 57 Inverters

-ShutdownB 3.8.8BAS ESACTIONSA.1, A.2.1, A.2.2, A.2.3. and A.2.4 (continued) margin to maintaining subcritical operation.

Introduction of temperature

changes, including temperature increases when operating with a positiveMTC, must also be evaluated to ensure they do not result in a loss ofrequired SDM.Suspension of these activities shall not preclude completion of actions toestablish a safe conservative condition.

These actions minimize theprobability of the occurrence of postulated events. It is further required toimmediately initiate action to restore the required inverters and to continuethis action until restoration is accomplished in order to provide thenecessary inverter power to the unit safety systems.The Completion Time of immediately is consistent with the required timesfor actions requiring prompt attention.

The restoration of the requiredinverters should be completed as quickly as possible in order to minimizethe time the unit safety systems may be without power or powered from abypass constant voltage transformer.

SURVEILLANCE SR 3.8.8.1REQUIREMENTS This Surveillance verifies that the inverters are functioning properly withall required circuit breakers closed and AC vital buses energized from theinverter.

The verification of proper voltage output ensures that therequired power is readily available for the instrumentation connected tothe AC vital buses. The 7 day Frequency takes into account theredundant capability of the inverters and other indications available in thecontrol room that alert the operator to inverter malfunctions.

REFERENCES

1. USAR, Chapter 6.2. USAR, Chapter 15.Wolf Creek -Unit 1 ..- eiin6B 3.8.8-5Revision 69 Distribution Systems -Operating B 3.8.9B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.9 Distribution Systems -Operating BASESBACKGROUND The onsite Class 1 E AC, DC, and AC vital bus electrical power distribution systems are divided by train into two redundant and independent AC, DC,and AC vital bus electrical power distribution subsystems as defined inTable B 3.8.9-1.

Train A is associated with AC load group 1 ; Train B, withAC load group 2.The AC electrical power subsystem for each train consists of anEngineered Safety Feature (ESF) 4.16 kV bus and 480 buses and loadcenters.

Each 4.16 kV ESE bus has one separate and independent offsite source of power as well as a dedicated onsite diesel generator (DG) source. Each 4.16 kV ESE bus is normally connected to a preferred offsite source. After a loss of the preferred offsite power source to a4.16 kV ESF bus, the onsite emergency DG supplies power to the bus.Control power for the 4.16 kV breakers is supplied from the Class 1Ebatteries.

Additional description of this system may be found in the Basesfor LCO 3.8.1, "AC Sources -Operating,"

and the Bases for LCO 3.8.4,"DC Sources -Operating."

The 120 VAC vital buses are arranged in two load groups per train andare normally powered through the inverters from the 125 VDC electrical power subsystem.

Refer to Bases B 3.8.7 for further information on the120 VAC vital system.The 125 VDC electrical power distribution system is arranged into twobuses per train. Refer to Bases B 3.8.4 for further information on the 125VDC electrical power subsystem.

The list of all required distribution buses is presented in Table B 3.8.9-1.APPLICABLE SAFETY ANALYSESThe initial conditions of Design Basis Accident (DBA) and transient ainalyses in the-USAR, Chapter 6 (Ref. 1), and in the USAR, Chapter 1 5(Ref. 2), assume ESF systems are OPERABLE.

The AC, DC, and ACvital bus electrical power distribution systems are designed to providesufficient

capacity, capability, redundancy, and reliability to ensure theavailability of necessary power to ESF systems so that the fuel, ReactorCoolant System, and containment design limits are not exceeded.

Theselimits are discussed in more detail in the Bases for Section 3.2, PowerWolf Creek -Unit 1 ..- eiin5B 3.8.9-1Revision 54

.... Distribution Systems -Operating B 3.8.9BASESAPPLICABLE Distribution Limits; Section 3.4, Reactor Coolant System (RCS); andSAFETY ANALYSES Section 3.6, Containment Systems.(continued)

The OPERABILITY of the AC, DC, and AC vital bus electrical powerdistribution systems is consistent with the initial assumptions of theaccident analyses and is based upon meeting the design basis of the unit.This includes maintaining power distribution systems OPERABLE duringaccident conditions in the event of:a. An assumed loss of all offsite power or all onsite AC electrical power; andb. A worst case single failure.The distribution systems satisfy Criterion 3 of the 10 CFR 50.36(c)(2)(ii).

LCO The required power distribution subsystems listed in Table B 3.8.9-1ensure the availability of AC, DC, and AC vital bus electrical power for thesystems required to shut down the reactor and maintain it in a safecondition after an anticipated operational occurrence (AOO) or apostulated DBA. The AC, DC, and AC vital bus electrical powerdistribution subsystems are required to be OPERABLE.

Maintaining the Train A and Train B AC, DC, and AC vital bus electrical power distribution subsystems OPERABLE ensures that the redundancy incorporated into the design of ESF is not defeated.

Therefore, a singlefailure within any system or within the electrical power distribution subsystems will not prevent safe shutdown of the reactor.OPERABLE AC electrical power distribution subsystems require theassociated buses and load centers to be energized to their propervoltages.

OPERABLE DC electrical power distribution subsystems require the associated buses to be energized to their proper voltage fromeither the associated battery or charger.

OPERABLE vital bus electrical power distribution subsystems require the associated buses to beenergized to their proper voltage from the associated inverter via invertedDC voltage, or bypass constant voltage transformer.

In addition, no tie breakers between redundant safety related AC, DC, andAC vital bus power distribution subsystems exist. This prevents anyelectrical malfunction in any power distribution subsystem frompropagating to the redundant subsystem, that could cause the failure of aredundant subsystem and a loss of essential safety function(s).

Wolf Creek- Unit 1 ..- eiin6B3.8.9-2Revision 69 Distribution Systems -Operating B 3.8.9BASESACTIONS C.1 (continued) status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by powering the bus from the associated inverter viainverted DC or bypass constant voltage transformer.

The required ACvital bus may also be restored to OPERABLE status through alignment tothe spare inverter powered from the 125 VDC bus in the same separation group.Condition C represents one AC vital bus without power; potentially boththe DC source and the associated AC source are nonfunctioning.

In thissituation, the unit is significantly more vulnerable to a complete loss of allnoninterruptible power. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for loss ofpower to the remaining vital buses and restoring power to the affectedvital bus.This 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit is more conservative than Completion Times allowed forthe vast majority of components that are without adequate vital AC power.Taking exceptionto LCO 3.0.2 for components without adequate vital ACpower, that would have the Required Action Completion Times shorterthan 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months if declared inoperable, is acceptable because of:a. The potential for decreased safety by requiring a change in unitconditions (i.e., requiring a shutdown) and not allowing stableoperations to continue;

b. The potential for decreased safety by requiring entry into numerousapplicable Conditions and Required Actions for components withoutadequate vital AC power and not providing sufficient time for theoperators to perform the necessary evaluations and actions forrestoring power to the affected train; andc. The potential for an event in conjunction with a single failure of aredundant component.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time takes into account the importance to safetyof restoring the AC vital bus to OPERABLE status, the redundant capability afforded by the other OPERABLE vital buses, and the lowprobability of a DBA occurring during this period.The second Completion Time for Required Action C.1 establishes a limiton the maximum allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence offailing to meet the LCO. If Condition C is entered while, for instance, anAC bus is inoperable and subsequently returned

OPERABLE, the LCOmay already have been not met for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This could lead to atotal of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , since initial failure of the LCO, to restore the vital busdistribution system. At this time, an AC train could again becomeWolf Creek- Unit IB389-Reion6 B 3.8.9-5Revision 69

.......Distribution Systems -Operating B 3.8.9BASESACTIONS C.__I (continued) inoperable, and vital bus distribution restored OPERABLE.

This couldcontinue indefinitely.

This Completion Time allows for an exception to the normal "time zero" forbeginning the allowed outage time "clock."

This will result in establishing the "time zero" at the time the LCO was initially not met, instead of thetime Condition B was entered.

The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is anacceptable limitation on this potential to fail to meet the LCO indefinitely.

0.1_.With DC bus(es) in one train inoperable, the remaining DC electrical power distribution subsystems are capable of supporting the minimumsafety functions necessary to shut down the reactor and maintain it in asafe shutdown condition, assuming no single failure.

The overall reliability is reduced,

however, because a single failure in the remaining DCelectrical power distribution subsystem could result in the minimumrequired ESF functions not being supported.

Therefore, the required DCbuses must be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by poweringthe bus from the associated battery or charger.Condition 0 represents one train without adequate DC power; potentially both with the battery significantly degraded and the associated chargernonfunctioning.

In this situation, the unit is significantly more vulnerable toa complete loss of all DC power. It is, therefore, imperative that theoperator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining trains and restoring power to theaffected train.This 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit is more conservative than Completion Times allowed forthe vast majority of components that would be without power. TakingSexception to LCO 3.0.2 for components without adequate DC power,...which-would have Required Action Completion Times shorter than2 hours, is acceptable because of:a. The potential for decreased safety by requiring a change in unitconditions (i.e., requiring a shutdown) while allowing stableoperations to continue; Wolf Creek -Unit 1 ..- RvsoB3.8.9-6Revision 0

Nuclear Instrumentation B 3.9.3B 3.9 REFUELING OPERATIONS B 3.9.3 Nuclear Instrumentation BASESBACKGROUND The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition.

The installed sourcerange neutron flux monitors are part of the Nuclear Instrumentation System (N IS). These detectors are located external to the reactorvessel and detect neutrons leaking from the core. There are two sets ofsource range neutron flux monitors:

(1) Westinghouse source rangeneutron flux monitors and (2) Gamma-Metrics source range neutron fluxmonitors.

The Westinghouse source range neutron flux monitors (SE-NI-0031 andSE-NI1-0032) are BE3 detectors operating in the proportional region ofthe gas filled detector characteristic curve. The detectors monitor theneutron flux in counts per second. The instrument range covers sixdecades of neutron flux (1 to 1 E+6 cps). The detectors also providecontinuous visual indication in the control room. The NIS is designed inaccordance with the criteria presented in Reference 1.The Gamma-Metrics source range neutron flux monitors (SE-NI-0060A and SE-NIl-0061A) are fission chambers that provide indication over sixdecades of neutron flux (1 E-1 to 1 E+5 cps). The monitors providecontinuous visual indication in the control room to allow operators tomonitor core flux.APPLICABLE Two OPERABLE source range neutron flux monitors are required toSAFETY ANALYSES provide a signal to alert the operator to unexpected changes in corereactivity such as an improperly loaded fuel assembly.

The source range neutron flux monitors satisfy Criterion 3 of 10 CFR50 .36(c)(2)(ii).

LCO This LCO requires that two source range neutron flux monitors beOPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity.

To be OPERABLE, each monitormust provide visual indication in the control room.When any of the safety related busses supplying power to one of thedetectors (SE-NI-31 or 32) associated with the Westinghouse sourcerange neutron flux monitors are taken out of service, the corresponding source range neutron flux monitor may be considered OPERABLE whenits detector is powered from a temporary nonsafety related source ofWolf Creek -Unit 1B393-Reion6 B3.9.3-1Revision 68 Nuclear Instrumentation B 3.9.3BASESLCO(continued) power, provided the detector for the opposite source range neutron fluxmonitor is powered from its normal source. (Ref. 2) This allowance topower a detector from a temporary non-safety related source of power isalso applicable to the Gamma-Metrics source range monitors.

(Ref. 4)The Westinghouse monitors are the normal source range monitors usedduring refueling activities.

The Gamma-Metrics source range monitorsprovide an acceptable equivalent control room visual indication to theWestinghouse monitors in MODE 6, including CORE ALTERATIONS.

(Ref. 4) Either the set of two Westinghouse source range neutron fluxmonitors or the set of two Gamma-Metrics source range monitors maybe used to perform this reactivity-monitoring function.

The use of oneBE3 detector and one Gamma-Metrics detector is not permitted due tothe importance of using detectors on opposing sides of the core toeffectively monitor the core reactivity.

(Ref. 3)APPLICABILITY In MODE 6, the source range neutron flux monitors must beOPERABLE to determine changes in core reactivity.

There are no otherdirect means available to check core reactivity levels. In MODES 2, 3,4, and 5, these same installed source range detectors and circuitry arealso required to be OPERABLE by LCO 3.3.1, "Reactor Trip System(RTS) Instrumentation."

ACTIONSA.1 and A.2With only one source range neutron flux monitor OPERABLE, redundancy has been lost. Since these instruments are the only direct means ofmonitoring core reactivity conditions, CORE ALTERATIONS andintroduction into the RCS, coolant with boron concentration less thanrequired to meet the minimum boron concentration of LCO 3.9.1 must besuspended immediately.

Suspending positive reactivity additions thatcould result in failure to meet the minimum boron concentration limit isrequired to assure continued safe operation.

Introduction of coolantinventory must be from sources that have a boron concentration greater-than that required in the RCS for minimum refueling boron concentration.

This may result in an overall reduction in RCS boron concentration, butprovides acceptable margin to maintaining subcritical operation.

Performance of Required Action A.1 shall not preclude completion ofmovement of a component to a safe position.

Wolf Creek -Unit 1 ..- eiin6B 3.9.3-2Revision 68 Nuclear Instrumentation B 3.9.3BASESACTIONS B.1(continued)

With no source range neutron flux monitor OPERABLE action to restorea monitor to OPERABLE status shall be initiated immediately.

Onceinitiated, action shall be continued until a source range neutron fluxmonitor is restored to OPERABLE status.B..22With no source range n~eutron flux monitor OPERABLE, there are nodirect means of detecting changes in core reactivity.

However, sinceCORE ALTERATIONS and boron concentration changes inconsistent with Required Action A.2 are not to be made, the core reactivity condition is stabilized until the source range neutron flux monitors areOPERABLE.

This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.~The Completion Time of once per-12 hours is sufficient to obtain andanalyze a reactor coolant sample for boron concentration and ensuresthat unplanned changes in boron concentration would be identified.

The12 hour Frequency is reasonable, considering the low probability of achange in core reactivity during this time period.SURVEILLANCE SR 3.9.3.1REQUIREMENTS SR 3.9.3.1 is the performance of a CHANNEL CHECK, which is acomparison of the parameter indicated on one channel to a similarparameter on other channels.

It is based on the assumption that thetwo indication channels should be consistent with core conditions.

Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel shouldbe consistent with its local conditions.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is consistent with the CHANNEL CHECKFrequency specified similarly for the same instruments in LCO 3.3.1.SR 3.9.3.2SR 3.9.3.2 is the performance of a CHANNEL CALIBRATION every18 months. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION.

The source rangeneutron detectors are maintained based on manufacturer's Wolf Creek -Unit 1B393-Reion5 B 3.9.3-3 N uclearlInstrumentation B 3.9.3BASESTECHNICAL SR 3.9.3.2 (continued)

SURVEILLANCE REQUIREMENTS recommendations.

The 18 month Frequency is based on the need toperform this Surveillance under the conditions that apply during a plantoutage. Operating experience has shown these components usuallypass the Surveillance when performed at the 18 month Frequency.

REFERENCES

1. 10 CFR 50, Appendix A, GDC 13, GDC 26, GOC 28, and.GDC 29.2. NRC letter (J. Stone to 0. Maynard) dated October 3, 1997:"Wolf Creek Generating Station -Technical Specification BasesChange, Source Range Nuclear Instruments Power SupplyRequirements."

3. Engineering Disposition for WO 11-339015-002, "Changes to TRM3.3.15,"

March 21, 2011.4. PIR 2004-1625, "Gamma-Metrics Detectors for Core Alterations,"

October 5, 2005.Wolf Creek -Unit I1 ..- eiin6B 3.9.3-4Revision 68

...RHR and Coolant Circulation

-High Water LevelB 3.9.5B 3.9 REFUELING OPERATIONS B 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation

-High 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 exchanger(s),

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 or 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 200°F, 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 wouldeventually challenge the integrity of the fuel cladding, which is a fissionproduct barrier.

One train of the RHR System is required to beoperational in MODE 6, with the water level > 23 ft above the top of thereactor vessel flange, to prevent this challenge.

The LCO does permitde-energizing the RHR pump for short durations, under the condition that the boron concentration is not diluted.

This conditional de-energizing of the RHR pump does not result in a challenge to thefission product barrier.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.

LCOOnly one RHR loop is required for decay heat removal in MODE 6, withthe water level > 23 ft above the top of the reactor vessel flange. Onlyone RHR loop is required to be OPERABLE, because the volume ofwater above the reactor vessel flange provides backup decay heatWolf Creek -Unit 1 ..- RvsoB3.9.5-1Revision 0

R HR and Coolant

-High Water LevelB 3.9.5BASESLCO(continued) removal capability.

At least one RHR loop must be OPERABLEand in operation to provide:a. Removal of decay heat;b. Mixing of borated coolant to minimize the possibility of criticality; andc. Indication of reactor coolant temperature.

An OPERABLE RHR loop includes 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. Management of gas voids is important to RHR System OPERABILITY.

The LCO is modified by a Note that allows the required operating RHRloop to be removed from service for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period,provided no operations are permitted that would dilute the RCS boronconcentration with coolant at boron concentrations less than required tomeet the minimum boron concentration of LCO 3.9.1. Boronconcentration reduction with coolant at boron concentrations less thanrequired to assure the minimum required RCS boron concentration ismaintained is prohibited because uniform concentration distribution cannot be ensured without forced circulation.

This permits operations such as core mapping or alterations in the vicinity of the reactor vesselhot leg nozzles and RCS to RHR isolation valve testing.

During this1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, decay heat is removed by natural convection to the largemass of water in the refueling pool.The acceptability of the LCO and the LCO Note is based on preventing core boiling in the event of the loss of RHR cooling.

An evaluation (Ref. 2) was performed which demonstrated that there is adequate flowcommunication to provide sufficient decay heat removal capability andpreclude core uncovery, thus preventing core damage, in the event of aloss of RHR cooling with the reactor cavity filled and the upper internals installed in the reactor vessel.APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6, withthe water level >_ 23 ft above the top of the reactor vessel flange, toprovide decay heat removal.

The 23 ft water level was selectedbecause it corresponds to the 23 ft requirement established for fuelmovement in LCO 3.9.7, "Refueling Pool Water Level." Requirements for the RHR System in other MODES are covered by LCOs inSection 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 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation

-Low Water Level."Wolf Creek -Unit 1 ..- eiin7B 3.9.5-2Revision 72 RHR and Coolant Circulation

-High Water LevelB 3.9.5BASESACTIONS RHR loop requirements are met by having one RHR loop OPERABLEand in operation, except as permitted in the Note to the LCO.A.1_If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations.

Suspending positive reactivity additions that could result in failure tomeet the minimum boron concentration limit of LCO 3.9.1 is required toassure continued safe operation.

Introduction of coolant inventory mustbe from sources that have a boron concentration greater than thatrequired in the RCS for minimum refueling boron concentration.

Thismay result in an overall reduction in RCS boron concentration, butprovides acceptable margin to maintaining subcritical operation.

A..22If RHR loop requirements are not met, actions shall be takenimmediately to suspend loading of irradiated fuel assemblies in the core.With no forced circulation

cooling, decay heat removal from the coreoccurs by natural convection to the heat sink provided by the waterabove the core. A minimum refueling water level of 23 ft above thereactor vessel flange provides an adequate available heat sink.Suspending any operation that would increase decay heat load, such asloading a fuel assembly, is a prudent action under this condition.

Performance of Required Action A.2 shall not preclude completion ofmovement of a component to a safe condition.

A.3If RHR loop requirements are not met, actions shall be initiated andcontinued in order to satisfy RHR loop requirements.

With the unit inMODE 6 and the refueling water level > 23 ft above the top of thereactor vessel flange, corrective actions shall be initiated immediately.

A.4If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outsideatmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RHR looprequirements not met, the potential exists for the coolant to boil andrelease radioactive gas to the containment atmosphere.

Closingcontainment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.

Wolf Creek -Unit 1 ..- eiin3B 3.9.5-3

........

.. '........RHR and Coolant Circulatiorn-High Water LevelB 3.9.5BASESACTIONS A.4 (continued)

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on the lowprobability of the coolant boiling in that time.SURVEILLANCE SR 3.9.5.1REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation andcirculating reactor coolant.

The flow rate is determined by the flow ratenecessary to provide sufficient decay heat removal capability and toprevent thermal and boron stratification in the core. The Frequency of12 hours is sufficient, considering the flow, temperature, pump control,and alarm indications available to the operator in the control room formonitoring the RHR System.SR 3.9.5.2RHR System piping and components have the potential to develop voidsand pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the RHR loops andmay also prevent water hammer, pump cavitation, and pumping ofnoncondensible gas into the reactor vessel.Selection of RHR System locations susceptible to gas accumulation isbased on a review of system design information, including piping andinstrumentation

drawings, isometric

drawings, plan and elevation

drawings, and calculations.

The design review is supplemented bysystem walk downs to validate the system high points and to confirm thelocation and orientation of important components that can becomesources of gas or could otherwise cause gas to be trapped or difficult toremove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versusoperating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.Acceptance criteria are established for the volume of accumulated gas atsusceptible locations.

If accumulated gas is discovered that exceeds theacceptance criteria for the susceptible location (or the volume ofaccumulated gas at one or more susceptible locations exceeds anacceptance criteria for gas volume at the suction or discharge of a pump),the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas(i.e., the system is sufficiently filled with water), the Surveillance may bedeclared met. Accumulated gas should be eliminated or brought withinthe acceptance criteria limits.Wolf Creek -Unit 1 ..- eiin7B 3.9.5-4Revision 72

..... RHR and Coolant Circulation

-High Water LevelB 3.9.5BASESSURVEILLANCE SR 3.9.5.2 (continued)

REQUIREMENTS RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location.

Susceptible locations in the same system flowpath which are subject to the same gas intrusion mechanisms may beverified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due toradiological or environmental conditions, the plant configuration, orpersonnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location.

Monitoring is not required for susceptible locations where themaximum potential accumulated gas void volume has been evaluated anddetermined to not challenge system OPERABILITY.

The accuracy of themethod used for monitoring the susceptible locations and trending of theresults should be sufficient to assure system OPERABILITY during theSurveillance interval.

The 31 day Frequency takes into consideration the gradual nature of gasaccumulation in the RHR System piping and the procedural controlsgoverning system operation.

REFERENCES

1. USAR, Section 5.4.7.2. SAP-06-1 13, "Loss of RHR Analysis with the Refuel CavityFlooded and Upper Internals Installed,"

November 16, 2006.Wolf Creek -Unit 1 ..- eiin7B 3.9.5-5Revision 72

-~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 GOC 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 200°F, 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 reactorLCOvessel 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 1 ..- RvsoB3.9.6-1Revision 0

...- RHR and Coolant Circulation

-Low Walter 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. Management of gas voids is important to RHRSystem OPERABILITY.

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 1 ..- eiin7B 3.9.6-2Revision 72

......RHR-and Coolant Circulation

-Low Water LevelB 3.9.6BASESACTIONS A.1 and A.2 (continued) vessel flange. When the water level is > 23 ft above the reactor vesselflange, the Applicability changes to that of LCO 3.9.5, and only one RHRloop is required to be OPERABLE and in operation.

An immediate Completion Time is necessary for an operator to initiate corrective actions.B.1If no RHR loop is in operation, there will be no forced circulation toprovide mixing to establish uniform boron concentrations.

Suspending positive reactivity additions that could result in failure to meet theminimum boron concentration limit of LCO 3.9.1 is required to assurecontinued safe operation.

Introduction of coolant inventory must befrom sources that have a boron concentration greater than that requiredin the RCS for minimum refueling boron concentration.

This may resultin an overall reduction in RCS boron concentration, but providesacceptable margin to maintaining subcritical operation.

B.2If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation.

Since the unit isin Conditions A and B concurrently, the restoration of two OPERABLERHR loops and one operating RHR loop should be accomplished expeditiously.

B.3If no RHR loop is in operation, all containment penetrations providing direct access from the containment atmosphere to the outsideatmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RHR looprequirements not met, the potential exists for the coolant to boil andrelease radioactive gas to the containment atmosphere.

Closingcontainment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable at water levels abovereduced inventory, based on the low probability of the coolant boiling inthat time. At reduced inventory conditions, additional actions are takento provide containment closure in a reduced period of time (Reference 2). Reduced inventory is defined as RCS level lower than 3 feet belowthe reactor vessel.Wolf Creek -Unit 1 ..- eiin4B 3.9.6-3

...........

RHRand Coo~lant Circulation

-Lbw Water LevelB 3.9.6BASESSURVEILLANCE SR 3.9.6.1REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation andcirculating reactor coolant.

The flow rate is determined by the flow ratenecessary to provide sufficient decay heat removal capability and toprevent thermal and boron stratification in the core. The Frequency of12 hours is sufficient, considering the flow, temperature, pumpcontrol,and alarm indications available to the operator for monitoring theRHR System in the control room.SR 3.9.6.2Verification that the required pump is OPERABLE ensures that anadditional RHR pump can be placed in operation, if needed, to maintaindecay heat removal and reactor coolant circulation.

Verification isperformed by verifying proper breaker alignment and power available tothe required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown tobe acceptable by operating experience.

SR 3.9.6.3RHR System piping and components have the potential to develop voidsand pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the RHR loops andmay also prevent water hammer, pump cavitation, and pumping ofnoncondensible gas into the reactor vessel.Selection of RHR System locations susceptible to gas accumulation isbased on a review of system design information, including piping andinstrumentation

drawings, isometric

drawings, plan and elevation

drawings, and calculations.

The design review is supplemented bysystem walk downs to validate the system high points and to confirm thelocation and orientation of important components that can becomesources of gas or could otherwise cause gas to be trapped or difficult toremove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versusoperating conditions.

Wolf Creek -Unit 1 ..- eiin7B 3.9.6-4Revision 72

..... ......RHR and Coolant Circulation

-Low Water LevelB 3.9.6BASESSURVEILLANCE SR 3.9.6.3.

(continued)

REQUIREMENTS The RHR System is OPERABLE when it is sufficiently filled with water.Acceptance criteria are established for the volume of accumulated gas atsusceptible locations.

If accumulated gas is discovered that exceeds theacceptance criteria for the susceptible location (or the volume ofaccumulated gas at one or more susceptible locations exceeds anacceptance criteria for gas volume at the suction or discharge of a pump),the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas(i.e., the system is sufficiently filled with water), the Surveillance may bedeclared met. Accumulated gas should be eliminated or brought withinthe acceptance criteria limits.RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location.

Susceptible locations in the same system flowpath which are subject to the same gas intrusion mechanisms may be;-

by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due toradiological or environmental conditions, the plant configuration, orpersonnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location.

Monitoring is not required for susceptible locations where themaximum potential accumulated gas void volume has been evaluated anddetermined to not challenge system OPERABILITY.

The accuracy of themethod used for monitoring the susceptible locations and trending of theresults should be sufficient to assure system OPERABILITY during theSurveillance interval.

The 31 day Frequency takes into consideration the gradual nature of gasaccumulation in the RHR System piping and the procedural controlsgoverning system operation.

1. USAR, Section 5.4.7.2. Generic Letter No. 88-17, "Loss of Decay Heat Removal."

Wolf Creek -Unit 1 ..- eiin7B 3.9.6-5Revision 72 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES-PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -Title Page Technical Specification Cover PageTitle PageTAB -Table of Contentsi34 DRR 07-1 057 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 D RR 03-0102 2/12/03B 2.1.1-3 14 DRRO03-0102 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 DRRO04-1414 10/12/04B 3.0-5 19 DRRO04-1414 10/12/04B 3.0-6 19 DRR 04-1414 10/12/04B 3.0-7 19 DRRO04-1414 10/12/04B 3.0-8 19 DRRO04-1414 10/12/04B 3.0-9 42 DRR 09-1009 7/16/09B 3.0-10 42 DRR 09-1 009 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 DRRO07-1057 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 07-1 057 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 1 eiin7Revision 73

.....LIST OF EFFECTIVE P~AGES -TECHNICAL SPECIFICATION BASES ... ....PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.1 REACTIVITY CONTROL SYSTEMS (continued)

B 3.1.3-5 0 Amend. No. 123 12/18/99B 3.1.3-6 0 Amend. No. 123 12/18/99B 3.1.4-1 0 Amend. No. 123 12/18/99B 3.1.4-2 0 Amend. No. 123 12/18/99B 3.1.4-3 48 DRR 10-3740 12/28/10B 3.1.4-4 0 Amend. No. 123 12/18/99B 3.1.4-5 0 Amend. No. 123 12/18/99B 3.1.4-6 48 DRR 10-3740 12/28/10B 3.1.4-7 0 Amend. No. 123 12/18/99B 3.1.4-8 0 Amend. No. 123 12/18/99B 3.1.4-9 0 Amend. No. 123 12/18/99B 3.1.5-1 0 Amend. No. 123 12/18/99B 3.1.5-2 0 Amend. No. 123 12/18/99B 3.1.5-3 0 Amend. No. 123 12/18/99B 3.1.5-4 0 Amend. No. 123 12/18/99B 3.1.6-1 0 Amend. No. 123 12/18/99B 3.1.6-2 0 Amend. No. 123 12/18/99B 3.1.6-3 0 Amend. No. 123 12/18/99B 3.1.6-4 0 Amend. No. 123 12/18/99B 3.1.6-5 0 Amend. No. 123 12/18/99B 3.1.6-6 0 Amend. No. 123 12/18/99B 3.1.7-1 0 Amend. No. 123 12/18/99B 3.1.7-2 0 Amend. No. 123 12/18/99B 3.1.7-3 48 DRR 10-3740 12/28/10B 3.1.7-4 48 DRR 10-3740 12/28/10B 3.1.7-5 48 DRR 10-3740 12/28/10B 3.1.7-6 0 Amend. No. 123 12/18/99B 3.1.8-1 0 Amend. No. 123 12/18/99B 3.1.8-2 0 Amend. No. 123 12/18/99B 3.1.8-3 15 DRR 03-0860 7/10/038 3.1.8-4 15 DRR 03-0860 7/10/03B 3.1.8-5 0 Amend. No. 123 12/18/998 3.1.8-6 5 DRR 00-1427 10/12/00TAB -B 3.2 POWER DISTRIBUTION LIMITSB 3.2.1-1 48B 3.2.1-2 0B 3.2.1-3 48B 3.2.1-4 48B 3.2.1-5 48B 3.2.1-6 48B 3.2.1-7 488 3.2.1-8 48B 3.2.1-9 29B 3.2.1-10 70B 3.2.2-1 48B 3.2.2-2 0B 3.2.2-3 48B 3.2.2-4 48B 3.2.2-5 48B 3.2.2-6 70DRR 10-3740Amend. No. 123DRR 10-3740DRR 10-3740DRR 10-3740DRR 10-3740DRR 10-3740DRR 10-3740DRR 06-1984DRR 15-0944DRR 10-3740Amend. No. 123DRR 10-3740DRR 10-3740DRR 10-3740DRR 15-094412/28/1012/18/9912/28/1012/28/1012/28/1012/28/1012/28/1012/28/1010/17/064/28/1512/28/1012/18/9912/28/1012/28/1012/28/104/28/15Wolf Creek -Unit 1 iRviin7iiRevision 73 LIST: OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES -...-PAGE (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 5"B 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 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 99-1 624DRR 99-1 624Amend. No. 123DRR 04-1533DRR 10-3740DRR 04-1533DRR 04-1414DRR 04-1533DRR 04-1533DRR 04-1533DRR 06-080012/18/9912/18/9912/18/9912/18/9912/18/9912/18/9910/12/00

-12/18/9912/18/9910/17/0612/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9911/6/1411/6/1412/18/9912/18/992/28/0212/18/9912/18/9912/18/9912/18/994/24/0012/18/9912/18/9912/18/992/16/0512/28/102/16/0510/13/042/16/052/16/052/16/055/18/06Wolf Creek -Unit 1 i eiin7iiiRevision73 LIST OF EFFECTIVE PAGES -. TECHNICAL BASES ..PAGE (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 3.3.2-3 1 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 04-1 533DRR 06-1 984DRR 04-1 533Amend. 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 00-0 147Amend. 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 11-07242/16/052/16/052/16/052/16/052/16/052/16/0512/28/102/16/0512/28/102/16/052/16/054/20/0 52/16/052/16/052/16/055/18/0611/6/142/16/0510/17/062/16/0512/18/9912/18/9912/18/9912/18/9912/18/995/1/10112/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/994/24/0012/18/9912/18/9912/18/9912/18/994/8/084/8/084/8/084/8/084/8/088/28/088/2 8/088/28/084/8/084/8/0812/18/9912/18/994/11/11Wolf Creek -Unit 1 vRviin7ivRevision 73 LIST OF EFFECTIVE PAGES --TECHNICAL SPECIFICATION BASES --.PAGE (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 20-B 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 04-1 533DRR 04-1 533DRR 04-1533DRR 04-1533DRR 06-0800DRR 04-1533Amend. No. 187 (ETS)Amend. No. 187 (ETS)DRR 04-1 533DRR 04-1 533DRR 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 99-16244/11/1112/18/9912/18/992/16/052/16/052/16/052/16/055/18/062/16/053/5/103/5/102/16/052/16/052/16/0511/16/111 9/2/094/8/082/16/052/16/059/2/099/2/099/2/099/2/099/2/099/2/0 99/2/0912/18/9910/12/0012/18/9912/18/9912/18/999/19/014/20/059/19/019/19/0110/12/0410/12/044/20/054/20/059/19/019/19/0112/18/992/28/027/10/0310/12/0412/18/992/28/0212/18/9912/18/9912/18/99Wolf Creek -Unit 1 eiin7VRevision 73 IST OF EFFECTIViEPAGES

-TECHNICAL SPECIFICATION BASES"PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE!

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

B 3.3.5-4 1 DRR 99-1 624 12/18/99B 3.3.5-5 0 Amend. No. 123 12/18/99B 3.3.5-6 22 DRR 05-1 375 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)0101000000067000000029005329" 0Amend. No. 123DRR 02-0411DRR 02-0411Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 15-0116Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 06-1 984Amend. No. 123Amend. No. 123DRR 11-1513DRR 06-1 984Amend. No. 12312/18/994/5/024/5/0212/18/9912/18/9912/18/9912/18/9912/18/9912/18/992/10/1512/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9910/17/0612/18/9912/18/997/18/1110/17/0612/18/99Wolf Creek -Unit I v eiin7viRevision 73 LIST OF EFFECTIVE TECHNICAL SPECIFICATION BASES, ..-...*... PAGE (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 72B 3.4.6-3 12B 3.4.6-4 72B 3.4.6-5 72B 3.4.6-6 72B 3.4.7-1 12B 3.4.7-2 17B 3.4.7-3 72B 3.4.7-4 42B 3.4.7-5 72B 3.4.7-6 72B 3.4.8-1 53B 3.4.8-2 72B 3.4.8-3 42B 3.4.8-4 72B 3.4.8-5 72B 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 32B 3.4.12-13 0B 3.4.12-14 32B 3.4.13-1 0B 3.4.13-2 29B 3.4.13-3 29(continued)

DRR 02-1 062DRR 02-1 062DRR 11-1513DRR 15-1918DRR 02-1062DRR 15-1918DRR 15-1918DRR 15-1918DRR 02-1062DRR 04-0453DRR 15-1918DRR 09-1009DRR 15-1918DRR 15-1918DRR 11-1513DRR 15-1918DRR 09-1009DRR 15-1918DRR 15-1918Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 00-1427DRR 00-1427Amend. No. 123DRR 07-01 39Amend. 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 12-1 792Amend. No. 123DRR 14-0346DRR 07-01 39Amend. No. 123DRR 07-01 39Amend. No. 123DRR 06-1984DRR 06-19849/26/029/26/027/18/1110/26/159/26/0210/26/1510/26/1510/26/159/26/025/26/0410/26/157/16/0910/26/1510/26/157/18/11110/26/157/16/0910/26/1510/26/15

-,12/18/9912/18/9912/18/9912/18/9910/12/0010/12/0012/18/992/7/0712/18/9912/18/9910/12/0412/18/9912/18/9912/18/992/7/072/27/142/27/1412/18/992/27/142/27/1411/7/122/27/1412/18/9911/7/1212/18/992/27/142/7/0712/18/992/7/0712/18/9910/17/0610/17/06Wolf Creek -Unit 1 iReson3viiRevision 73

LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES-PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

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

B 3.4.13-4 35 DRR 07-1553 9/28/07B 3.4.13-5 35 DRR 07-1553 9/28/07B 3.4.13-6 29 DRR 06-1984 10/17/06B 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 DR R 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 5/1/107B 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 DR R 06-2494 12/13/06B 3.4.16-2

31. DR R 06-2494 -- 12/13/06B 3.4.16-3 31 D RR 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 DR RI1-0724 4/11/111B 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 SYSTEMSB 3.5.1-1 0B 3.5.1-2 0B 3.5.1-3 73B 3.5.1-4 73B 3.5.1-5 1B 3.5.1-6 1B 3.5.1-7 71B 3.5.1-8 1B 3.5.2-1 0B 3.5.2-2 0B 3.5.2-3 0B 3.5.2-4 0B 3.5.2-5 72B 3.5.2-6 42B 3.5.2-7 42B 3.5.2-8 72B 3.5.2-9 72B 3.5.2-10 72B 3.5.2-11 72B 3.5.2-12 72(ECCS)Amend. No. 123Amend. No. 123DRR 15-21 35DRR 15-21 35DRR 99-1624DRR 99-1 624DRR 15-1528DRR 99-1624Amend. No. 123Amend. No. 123Amend. No. 123Amend. No. 123DRR 15-1918DRR 09-1009DRR 09-1009DRR 15-1918DRR 15-1918DRR 15-1918DRR 15-1918DRR 15-191812/18/9912/18/9911/17/1511/17/1512/18/9 912/18/997/30/1512/18/9912/18/9912/18/9912/18/9912/18/9910/26/157/16/097/16/0910/26/1510/26/1510/26/1510/26/1510/26/15Wolf Creek -Unit I1iiRviin7 viiiRevision 73

.. .... LIST-OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES ... .PAGE (! REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

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

B 3.5.3-1 56 DRR 12-1792 11/7/12B 3.5.3-2 72 DRR 15-1918 10/26/15B 3.5.3-3 56 DRR 12-1792 11/7/12B 3.5.3-4 56 DRR 12-1792 11/7/12B 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 06-1 350 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 08 3.6.1-2 0B 3.6.1-3 0OB 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 3.6.3-9 368 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 37Amend. 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 08-1 096DRR 08-1096DRR 08-0255DRR 08-0255DRR 08-0255DRR 08-1096Amend. No. 123Amend. No. 123Amend. No. 123DRR 08-050312/18/9912/18/9912/18/995/26/0412/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/991/31/111/31/111/31/113/20/093/11/083/11/089/19/013/11/083/11/083/9/1113/11/088/28/088/28/083/11/083/11/083/11/088/28/0812/18/9912/18/9912/18/994/8/08Wolf Creek -Unit 1 xRviin7ixRevision 73 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES -.......PAGE (1 REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

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

B 3.6.5-3 13 DRR 02-1458 12/03/02B 3.6.5-4 0 Amend. No. 123 12/18/99B 3.6.6-1 42 DRR 09-1 009 7/16/09B 3.6.6-2 63 DRR 14-1572 7/1/114B 3.6.6-3 37 DRR 08-0503 4/8/08B 3.6.6-4 72 DRR 15-1918 10/26/15B 3.6.6-5 0 Amend. No. 123 12/18/99B 3.6.6-6 18 DRR 04-1018 9/1/104B 3.6.6-7 72 DRR 15-1918 10/26/15B 3.6.6-8 72 DRR 15-1918 10/26/15B 3.6.6-9 72 DRR 15-1918 10/26/15B 3.6.6-10 72 DRRI15-1918 10/26/15B 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 06-1 984 10/17/06B 3.6.7-5 42 DRR 09-1 009 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 3.7.4-30 Amend. No. 123 12/18/990 Amend. No. 123 12/18/990 Amend. No. 123 12/18/990 Amend. No. 123 12/18/9932 DRR 07-01 39 2/7/0732 DRR 07-0139 2/7/0744 DRR 09-1744 10/28/0944 DRR 09-1744 10/28/0944 DRR 09-1 744 10/28/0944 DRR 09-1 744 10/28/0944 DRRO09-1744 10/28/0944 DRR 09-1 744 10/28/0944 DRRO09-1744 10/28/0944 DRRO09-1744 10/28/0944 DRR 09-1744 10/28/0944 DRRO09-1744 10/28/0944 DRRO09-1744 10/28/0937 DRR 08-0503 4/8/0850 DRRI11-0449 3/9/11137 DRR 08-0503 4/8/0837 DRR 08-0503 4/8/0837 DRR 08-0503 4/8/0837 DRR 08-0503 4/8/0837 DRR 08-0503 4/8/0837 DRR 08-0503 4/8/0866 DRRI14-2329 11/6/1466 DRRI14-2329 11/6/1437 DRR 08-0503 4/8/081 DRR 99-1624 12/18/991 DRR 99-1624 12/18/9919 DRRO04-1414 10/12/04Wolf Creek -Unit 1 eiin7XRevision 73 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES.- .-.*PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.7 PLANT SYSTEMSB 3.7.4-4B 3.7.4-5B 3.7.5-1B 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 3.7.8-13.7.8-2B 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 3.7.11-2*

B 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 3.7.15-1(continued) 1915454060444432143200000010000033336441414157576441640576363024142575764646400DRR 04-1414DRR 99-1 624DRR 11-2394DRR 11-2394Amend. No. 123DRR 13-2562DRR 09-1 744DRR 09-1744DRR 07-01 39DRR 03-01 02DRR 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 99-1 624DRR 09-1009DRR 13-0006DRR 13-0006DRR 14-1 822DRR 14-1822DRR 14-1822Amend. No. 123Amend. No. 12310/12/0412/18/9911/16/1111/16/1112/18/9910/25/1310/28/0910/28/092/7/072/12/032/7/0712/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/9912/18/997/14/007/14/007/14/007/14/008/28/143/20/093/20/093/20/091/16/131/16/138/28/143/20/098/28/1412/18/991/16/137/1/1147/1/11412/18/992/28/0612/18/997/16/091/16/131/16/138/28/148/28/148/28/1412/18/9912/18/99Wolf Creek -Unit 1 iRviin7xiRevision 73

"::' ...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 3.7.15-2 0 Amend. No. 123 12/18/99B 3.7.15-3 0 Amend. No. 123 12/18/99B 3.7.16-1 5 DRR 00-1427 10/12/00B 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 DRRO01-0474 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 Am end. 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 DRRI11-2394 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 71B 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 33DRR 11-2394Amend. No. 123DRR 10-1089DRR 15-1528DRR 13-1524DRR 06-0800DRR 06-1350DRR 07-1553DRR 09-1 009DRR 08-1 096DRR 08-0255DRR 10-1 089DRR 10-1089DRR 10-1089DRR 10-1089DRR 06-1350.DRR 06-1350DRR 13-1 524DRR 06-1 350DRR 06-1 350DRR 07-0656DRR 07-0656DRR 08-1846DRR 07-0656DRR 07-0656DRR 07-0656DRR 13-1524DRR 13-1524DRR 11-2394DRR 07-0656DRR 07-065611/16/1112/18/996/16/107/30/156/26/135/18/067/24/069/28/077/16/098/28/083/11/086/16/106/16/106/16/106/161107/24/067/24/066/26/137/24/067/24/065/1/075/1/0712/9/085/1/075/1/075/1/076/26/136/26/1311/16/111 5/1/075/1/07Wolf Creek -Unit 1 i eiin7xiiRevision 73 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES,'-,

-- ... -..PAGE (1 REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.8 ELECTRICAL POWER SYSTEMSB 3.8.1-32 33B 3.8.1-33 71B 3.8.1-34 47B 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 69B 3.8.7-2 69B 3.8.7-3 69B 3.8.7-4 0B 3.8.8-1 57B 3.8.8-2 0B 3.8.8-3 69B 3.8.8-4 57B 3.8.8-5 69B 3.8.9-1 54B 3.8.9-2 69B 3.8.9-3 54(continued)

DRR 07-0656DRR 15-1528DRR 10-1 089DRR 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 00-1 541Amend. 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. 123DRR 15-0493DRR 15-0493DRR 15-0493Amend. No. 123DRR 13-0006Amend. No. 123DRR 15-0493DRR 13-0006DRR 15-0493DRR 11-2394DRR 15-0493DRR 11-23945/1/1077/30/156/16/101/16/1312/18/9912/18/991/16/131/16/131/16/131/16/1312/18/9912/18/9912/18/9912/18/9912/18/9912/18/999/26/0212/18/9912/18/9912/18/9912/18/9912/18/9912/18/993/9/113/9/1113/13/0112/18/994/24/001/16/1312/18/991/16/131/16/131/16/1312/18/9912/18/9912/18/9912/18/9912/18/9912/18/993/26/153/26/153/26/1512/18/991/16/1312/18/993/26/151/16/133/26/1511/16/113/26/1511/16/111 Wolf Creek -Unit 1 iiRviin7xiiiRevision 73

...LIST OF EF~FECTIVE PAGES -TECHNICAL SPECIFICATION BASES .. ....PAGE (1) ,REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

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

B 3.8.9-4 0 Amend. No. 123 12/18/99B 3.8.9-5 69 DRR 15-0493 3/26/15B 3.8.9-6 0 Amend. No. 123 12/18/99B 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 DRRO04-1414 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 68 DRR 15-0248 2/26/15B 3.9.3-2 68 DRR 15-0248 2/26/15B 3.9.3-3 51 DRR 11-0664 3/21/11B 3.9.3-4 68 DRR 15-0248 2/26/15B 3.9.4-1 23 DRR 05-1 995 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 5/1/107B 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 72 DRRI15-1918 10/26/15B 3.9.5-3 32 DRR 07-0139 2/7/07B 3.9.5-4 72 DRRI15-1918 10/26/15B 3.9.5-5 72 DRR 15-1918 10/26/15B 3.9.6-1 0 Amend. No. 123 12/18/99B 3.9.6-2 72 DRRI15-1918 10/26/15B 3.9.6-3 42 DRR 09-1009 7/16/09B 3.9.6-4 72 DRR 15-1918 10/26/15B 3.9.6-5 72 DRR 15-1918 10/26/15B 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 1 i eiin7xivRevision 73 LIST OF EFFECTIVE PAGES -TECHNICAL SPECIFICATION BASES .... -PAGE (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 1 vRviin7XVRevision 73 W0LF CREEK7 NUCLEAR OPERATING CORPORATION Cynthia R. Hafenstine Manager Regulatory AffairsMarch 10, 2016RA 16-0008U. S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555