ML040780754

Changes to Technical Specifications Bases - Revisions 14-16

ML040780754
Person / Time
Site:	Wolf Creek
Issue date:	03/11/2004
From:	Moles K Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 04-0029
Download: ML040780754 (44)

Text

ALF CREEK W 'NUCLEAR OPERATING CORPORATION Kevin J. Moles Manager Regulatory Affairs MAR 1 12004 RA 04-0029 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Changes to Technical Specification Bases - Revisions 14 through 16 Gentlemen:

The Wolf Creek Generating Station (WCGS) Unit 1 Technical Specifications (TS), Section 5.5.14, "Technical Specifications (TS) Bases Control Program," provide-the means for making changes to the Bases without prior NRC approval. In addition, TS Section 5.5.14 requires that changes made without NRC approval be provided to the NRC on a frequency consistent with 10 CFR 50.71(e). The Enclosure provides those changes made to the WCGS TS Bases (Revisions 14 through 16) under the provisions of TS Section 5.5.14 and a List of Effective Pages. This submittal reflects changes from January 1, 2003 through December 31, 2003.

There are no commitments contained in this submittal.

If you have any questions concerning this submittal, please contact me at (620) 364-4126 or Ms. Jennifer Yunk at (620) 364-4272.

n J. Mo KJM/rIg Enclosure cc: J. N. Donohew (NRC), w/e D. N. Graves (NRC), w/e B. S. Mallett (NRC), wle Senior Resident Inspector (NRC), wle RO. Box 411 Burlington, KS 66839 Phone: (620) 364-8831 An Equal Opportunity Employer MIFIHCNVET

I r j Enclosure to RA 04-0029 Wolf Creek Generating Station Changes to the Technical Specification Bases

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs

- ,-- :i... - . "l BASES BACKGROUND: GDC 10 (Ref. 1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and Anticipated Operational Occurrences (AOOs). This is

-accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant.i Overheating of the fuel is prevented by maintaining the steady state peak Linear Heat Rate (LHR) below the level at which fuel centerlirie melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface'temperature is slightly above the coolant saturation temperature.

FFuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon

- centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

' Operation above'the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam

- -:- -- film, high cladding temperatures are reached, and a cladding water -

-(zirconium water) reaction may take place. This chemical reaction results in oxidation' of thejfuel cladding to a'structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant. *,

The prope functioning of the' Reactor Protection System (RPS) and steam generator safety valves prevents violation of the reactor core SLs.

Wolf Creek - Unit 1 B 2.1.1-1 - Revision 0

Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as 'atrsult of normal!.

SAFETYANALYSES operation and AOOs. The rbactor core SLs are' established to preclude violation of the following fuel design criteria:

• ' ' t , i - *-

'

• a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the h6t fuel rod in the core does not experience DNB; and

- t'h l b.' -The hot fuel pellet iii theico're rmust not experience centerline fuel melting.

The ReactorTrip System'Alrow-able Valu'es, in'Table 3.3.1-1 in combination with all the L=ss're designed to *prevent any anticipated J combination of transient icn'dit1`6h9 for Rea~ctor Coolant System (RCS)

";'-femperature,, pressure; RCS hiW, Al, and'THERMAL POWER level that Jr 1 would result in-a Departur'e fromfi Nucleate Boiling1Ratio (DNBR) of less

-'than the DNBR lirit and preclude the 'xistenrce of ffow instabilities.

. -',( . *  !' n

' I Pr6tection for thesereractor c6re SLs is pr vided' by the appropriate

':' ' . ' ' '  ;" ' operatiori oflhe RPS and'tlib'ste rrageneratb# safety valves.

The limitation that tha average'6ith'lp'yh the hot lg be less than or

'equal to the enthalpy of S firatid 1iiuld also 6esuires that the AT crpoer iscoepoe measured by instrumentation, used in the RPS design as a measure of core power, is proportion' tocore power. -

- The SLa 'represent a desin'requiremen'tfor establishing the RPS

- Allowable Values identified previously LCO' 34.1', "RCS Pressure, Temperature; and Flow Departure from Nucleate' Boiling (DNB) Limits,"

and the assumed initial conditions of the safety analyses (as-indicatedjin.

the USAR, Ref. 2) provide more restrictive limits to ensure that the SLs

• ' . '4are notexceeded.'- .,! ,-'> ' ..

SAFETY LIMITS shows the loci of points of THERMAL The figure provided in'the' COLRi~re'"ndbelotwhihthe POWER,'pressu d aver[ge' raturebeo which

.- calculated DNBR is,not less than the design DNBRvalue, that the

. average enthalpy in the hotlegis less than or equal to the enthalpy of saturated liquid,,or that the exit quality? is within' the limits defined by the DNBR corr~latiorii. -

.  ; !A 1 ' ,t  ;'>;;¢ *Z4j § o Wolf Creek - Unit 1 B 2.1.1-2 Revision 14

Reactor Core SLs

- B 2.1.1 BASES SAFETY LIMITS - Jhe reactor:core SLs are established to preclude~the violation of the-(continued), , following fuel~design,critera: ,r80..6 2C-

a. There must be at least a95% confidence level (the 95% confidence

Jj; ,level (the.95/95,DNB criterion) that thevhot fuel rod in the core does J.) f <-, *, .. [ -.not experience DNB; and 4 , -

b. There must be at least a 95% probability at a 95% confidence level

,-,.that the hot fuel pellet in the core does not experience centerline fuel melt.: - '

The reactor

-+% , 1 ,

. .4 core....*;SLs

". are used to

. 1... ...- '...

define the various RPS functions such c ,.,, thatthe above criteria are-satisfied during steadystate operation, normal operational transient~s,.and anticipated operational occurrences (AOOs).

,* .,_ ' , Tensure that the RPS lprec'luoeshe violation of the above criteria, 3 *, ,additional critiria ,are applied toethe Overtemperature and Overpower AT i :ri; r..eactortrip functions That is, it must be demonstrated that the average enthalpjini the hot legs is less than or equal to the saturation enthalpy and mono - Ye

• athat the, core exit quality is within the lirrlits defined by the DNBR

• Ld .>V Br' -- ,, .or~r~e~laetion.ppropriate functioningof theRPS ensures that for variations in the THERMAL POWER, RCS pressure, RCS average temperature,

- . R.- flow rate, and Al that the reactor core SLs will be satisfied during

, ,.~ ~ :,teys state opratiop, -normal operational transients, and A0Os.

.. "*.*- .'*.I *. S t..;.'1

, ;tF

' . D* D e. S .. 3 It,),lj:

, 1-:

Reference 3-discusses-theuqse of.4700 'F as the thermal overpower limit to preclude fuel centerline melting, accommodating thermal evaluation

!gi gre1 5,0X1f Reference 2 depicts the protection

• ¶>l >- .-- ,5prqovided bhythe Overpowe :AT reactor trip function against fuel centerline

!-' a *.;, . ' (I-- , -

APPLICABILITY SL 2.1.1 only applies in MODES 1Sand 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are

quired to be OPERABLE duririg'MODES1 anrd'2 to ensure operation

', , .,  ;.s,,vithinthe reactor:coreSLs.:The steam generator safety valves or

. J, <se:.-Automatic protection actions serve to prevent RCS heatup to the reactor

.; ,- core,SL conditions orto initiate a reactor~trip function, which forces the f.-itunit-into MODE 3.

Allowable Values for the reactor trip functions are

,,- E. e.-.,rspecifiedin LCO,33.1,3, ReactorTrip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

Wolf Creek - Unit 1 ,B2.1.1-3 .. I Revision,14

- 2

, I Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT 2.2.1 . . .  : .. , ,, :

VIOLATIONS .- ' -; -'.'

If SL 2.1.1; is violated; the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

-The allowed Completion .nimeof 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of ;- ' .

. '.-:: .; bringing the unit to a MODE of operation where this SL is not applicable, r .:-;; - . and reduces the probabilityof fuel damage.'-

REFERENCES . :1. 10 CFR 50, AppendixA; GDC 10.

2. USAR, Chapter 15. i.

t 1 w-3

USARj,Section 4.4.1;2': ..  : .

I

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• .-. tt I *Ic - '.. *. Ii 4 .I 77- , :.- 1, .-, : " :' ' - ,-,.. : - '. I  ::. , I;- .1 - , - I ;1 I I :i

, " I '.'. . " - - , "-,-- _- 11! C Wolf Creek - Unit 1 B 2.1.1-4 * - . Revision 14'

't _. 4", ,

. .- A11 PHYSICS TEST Exceptions - MODE 2 II 7 B 3.1.8

.,tA, ,1' BASES APPLICABLE criteria for inclusion in the Technical Specifications, since the components SAFETY ANALYSES and process variable LCOs suspended during PHYSICS TESTS meet (continued) *: ; *Criteria 1,2, and 3 ofh10.CFR 50.36(c)(2)(ii).

.1~ ; . ,. , - *- A LCO T.: tz. This LCO allows the reactorrparameters of MTC and minimum

,; l temperature for criticality.to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. One Power Range Neutron Fluixchnniel miiiaybe bypassed, reducing the number of required channels from 4 to 3. Operation:beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met.n -. : -

The requirements of LCO 3.1:3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6,

• ,and LCO 3.4.2 may be suspended and the number of required channels for LCO 3.3:1, bRTS Thdtru4-i6ntatfoh` Fuinctions 2,3, 6, and.1 8.e' 'maybe reduced to 3 required channels during the performance of PHYSICS TESTS provided:

a. RCS lowest operating loop average temperature is 2 541 'F;

b. SDM is within the limits provided in the COLR; and

c. THERMAL POWER is

• 5% RTP.

APPLICABILITY This LCO is applicable in MODE 2 when performing low power PHYSICS TESTS. The applicable PHYSICS TESTS are performed in MODE 2 at HZP. . .

ACTIONS A.1 and A.2

' If the SDM requirement is not met, boration must be initiated promptly. A Completion Time of 15 minutes is'adequate for an operator to correctly

-align and start the required systems and components. The operator should begin boration with the best source available for the plant

' ' conditions.:- Boration will be continued until SDM is within limit.

' Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOsito within specification.

• Wolf Creek - Unit 1 B -3.1 .8-3 a .. Revision-15

I

• ph PHYSICS TEST Exceptions - MODE 2 B 3.1.8 BASES ACTIONS  ;.B1 t ' - : ^ [ ial Bi- ,1 ',)t tal '

(continued) .~.  ; r .. - .

s; - u s  ; -..  :;.,

' rv When THERMAL POWER is > 5% RTP; the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor

- '. beyond its design limits. ' Immediately opening the RTBs will shut down the reactor and prevent operation of the'reactor outside of its design Iimits.

C.1 t -- 7iy c When the RCS lowest operating Ioops Tas is c 541OF, the appropriate action is to restor&Ta, to W9ithin its spkcified liiiiitiLThe allowed

,' -Completion -Time of 15:minrbtes provides time. for restoring Tang to within llimits without allo'w'ng the' plant to remain in aniuna;"ceptable condition for

• f  %' - Kan extended'period of tme: Operation withthe'reactor critical and with an operating loop's temperature below 541OF'couid violate the assumptions for accidents analyzed in the safety analyses.

• , X , ] -*  : *' r f, - -  ; *_

.. 6 D.1l - . .

If the Required Actions cannot be completed within the associated Completion Time,'the plant must be'broight to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MCDE 3 within rin dditi6nal 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, fo. rea'clihg MODE 3 in an oiderly manner and without challenging plant systems.

t5.,  : A -; . ..

SURVEILLANCE SR 3.1.8.1.  :..

REQUIREMENTS The required power rang'dand .1nterrmediate range neutron detectors must be verified to be OPERABLEilA MODE 2 by LCO 3.3.1, Reactor Trip System (RTS) Instrumefitatio;." P A CHANNEL OPERATIONAL TEST is performed on each OPERABLE power range and intermediate range J. . I. .

channels prior to initiationlof the PHYSICSTESTS'. VThis will ensure that the RTS is properly aligned to'provide the required degree of core protection during'the performafte of the'PHYSICS' TESTS. The SR

• 3.3.1.8 Frequency is sufficient to 6nsure'th'at the-instrumentation is OPERABLE before initiating PHYSICSTESTS:4;.

Wolf Creek - Unit 1 B 3.1 .8-4 Revision 15 1

.  ; .' -Remote Shutdown System

' B 3.3.4 BASES APPLICABLE The Remote Shutdown System is required to provide equipment -

SAFETYANALYSES at appropriate locations outside the control room with a capability to :-

- :;., ¢, promptly shut down'and maintain the unit in a safe condition in MODE 3.

.::: .; ,The criteria'governing the design 'and specific system requirements of the

,-;:. ' *. ~.Remote Shutdown System arelocated in 10,CFR 50, AppendixA, GDC19(Ref.1).'--

The Remote Shutdown System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO .. ::.w ^,The Remote Shutdown-System LCO provides the OPERABILITY

;; °' i, '  :'requirements ofhthe functions'and ASP controls necessary to place and i' iO:ji '.'S' c~y maintain the unit imnMODE,3 from..a location other than the control room.

Y;- - i -,nThe 1:.  ; functions required -are,listed in-Tabtle 3.3.4-1 in the accompanying

-: ,! : .ft.LCO.-  :.z -, '; '  : L '. .

The required ASP controls are listed above and described in USAR Table

-7.4-1.1. The'remote'shutdown panel controls not located at the ASP are described in USAR Table 7.4-1.2 and are excluded from the requirements of this LCO.

¢. r; ,. rThe controis

iinstrumentation, and transfer,switches are required for:

! ':,,Core reactivity control (initial and long term);

s'.* -  :. y*r-rRCS-pressure.control; . -i

~~~4.,,,,-. - .ea. - _. .f

• Decay heat removal via the SGs; and

• RCS inventory control..

-

• A Function of a Remote Shutdown System is OPERABLE if the required

, . ri F;:~number of charnuils needped-to support the Remote:Shutdown System

. .- ,;,oj-

,,Functionidentifiedin TabJe 3.3.4-1 .are OPERABLE~

Fn.

- 2.  ; ** S :i ' - .

-. *.. 1 y,':,* SThe remoteshutdowun instruments and required ASP controls covered by

-.,*  ;., -,this.LCO~do-not need.to be epergized to he'considered OPERABLE. This dr,' LCO is intended-to ensure the instruments and controls will be

• , c-

  ~.. ~. .yPERABLE-if unit conditions.require that jhe Remote Shutdown System be placed in operation: t,, ,.

p . . ' " -

Wolf Creek'- Unit 1 B -3.3.4-3 - - 1, -Revision .15

Remote Shutdown System B 3.3.4 BASES APPLICABILITY The Remote Shutdown System ICC is applicable in MODES 1, 2, and 3'.'

This is required so that the unit can be placed and maintained in'MODE-3'

... ,A ,-. extended J;'iforan :_. pe?1bod'f time'.fiorn' a location otherthan the control room.

This LCO is not applicable in'MODE 4, 5, or 6. In these MODES, the

' .:  ; facility is' already subcritical 'nd in a condition of reduced RCS energy.

Urnder theseconditions,'`considerable time is~available to restore the remote shutdown instrumernts nhd required ASP controls if control room

• W,. instrumentsor controls become unavailable. 1 ACTIONS..X .. .~ Note 1 is included which excludes the, MODE change restriction of LCO 3.0.4.:(This exception:;a'lows entry into an applicable MODE while

. . , C relying on the ACTIONS-even though the ACTIONS, may eventually

. . require a unit shutdown.-, This exception is; acceptable due to the low probability of an event requiring. the Remote Shutdown System and 4'-. 1,.

because the equipment can generally be repaired during operation without significantrisk of spurious trip...: . ,

• -:'. -- Ncte 2 has been added to the ACTIONS to clarify the application of

. i *Cormpletion.Timerules. Separate Condition entry isallowed for each Function listed on Tabla 3.3.4-1 'and for each required ASP control. The Completion:Time(s) of the ino6irable channel(s)/train(s) of a Function will be tracked separately for each Fun'tion'i starting frorm; the time the Condition was entered for that Function.

? '* - "' When the Required Channels inTaIle 3.3.4-1 are specified on a per trip breaker, pedrSG' orperpu'rn basis,-the Condition may be entered

• -. ' ' separately for' each trip breaker; SG,6or pump, as appropriate.

,.,,, . ., . .......... ... . . - 4.

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' A. 1 l!. .  : ,. l C- ' Condition A addresses the situation where' one or more required Functions of the Remote Shutdbwn.Syster in Table 3.3.4-1, or one or more required ASP controls are inoperable..

The Required Action is to restore the required Function and ASP control to OPERABLE status within 30 days; The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.

I Wolf Creek - Unit 1 B 3.3.4>4 Revision 1

.'.~*~ :jRCS PIV Leakage B 3.4.14 BASES SURVEILLANCE-r .,SR 3.4.141 (cpntiiiu~ed4) . ,. ,,

REQUIREMENTS o'g~pr rs ft5p peinih'&nomhinal valve .diamheter upto5pmaxumpleso 0 . gp eahvalve. Leakage testing requires a sf~ble pressure condition.

Fo!he tv~oPlVs- in'se'ribsilhe leakage requirement applies to each valve

• '~.JI 'Iindi)iu - o 616eI~i&mobined le'A1agb ckross both valves. If the PIVs

~ " ~iare not individally leai(ahe'teied o'ne-~~alve maj have failed completely

• 1~~2(S I14) '~'and nbt be ~detecteddif the o~ther ii~lve in series meets the leakage

'reqrmet.in tfiis 'it "ation ~the" protection provided by redundant valves

- -- - _would be lost.'_ -

fl ','~:.:~. ,."Testing si d be perfdrmed e'very~'18 months, atypic'al refueling cycl'e,if the

.c~:~~Uplant doei hnot g~into'MODE:5 for at least 7.days'. The 18 month

.'~ "C'Frqu~bj'is within'the'frequ~fic'yallowed by the American Society of

.%'~ .:L.'Mechanical Engineers (ASME) Code,-SUction Xi (Ref. 6). and is based on n~>:'K i Vihelne ed to perform:such-surveillances under the conditions that apply

~-.-

pr~.. during an outage and.the potential for aniunplainne~d -transient if the

• - . Surveillance were performed with the reactor at, power..

Tkest pressures less jthan 2235 psig but greater than 150 psig are allowed 0- b1w, i fo-leswee.higher pressures could tend to diminish leakage channel i~:: .ppningObser yed leakage shall be adjusted for actual pressure to 2235

~ "psig assuming. the leakage.to. be directly'proportional to pressure differential

.1 to the.on~e halfpower.,....' .~ ..

In addition, testing must be performed once after the check valve has been nc ~,:.~ pened by~flow,pr.exerc~ised to ensure tight reieatirig. PIVs disturbed in the

..performnafce of this Syrve-`1Iahce sitio'61d 61so-156e'sted unless

• ~Ldocumentation 4shows thtan Ifnt tting loop cannotpatclyb a'~6ided. T~sGin-mu'St b6 pe~rformed within 24"hours after the check valve has been reseated. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is'a reasonable and practical time limit for performing this test after opening or reseating a check valve.

• The leakage limit is to be met at the RCS pressure associated with MODES I af2.'This pemis leakage testing at high differential pressures w'~'ith stabl6 c-ohditions nbf possible in the MODES with lower pressures.

Wolf Creek - Unit 1 B 3.4.1 4-6 WolfCree

-nit

..B.4:1-5 Revision 16

RCS PIV Leakage B 3.4.14 BASES . . ' '-.' . .

, , -- . . . I  :, .I' I' : J. " !' '

SURVEILLANCE SR 3.4.14.1 (continued ' ; ;! J - - ';-

REQUIREMENTS....'

'Entry into MOD.ES'3and 4 is'al edto establish' the necessary

.~' differential'pressures and stable' criditions to allow for performance of th Surveillance.' The Noi6.that allows this'provision is complementary to

,, the Frequency of prior to entry1into MODE 2 whenever the unit has been

'iriMODE 5 for 7 days orem&, if leakage testing has not been performed in the previous,9 months.*I In addition, this Surveillance is not required to

,be performed on the RHR-System'when the RHR System is aligned to the RCS in the shutdown cooling.mode of operation. PlVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established. . '. a-

• : l SR 3.4.14.22 i . 'MF??Ir -.

• The RHR suction isolati6n vakvd'ihterlock setpofnr that prevents the valves from birig'opendd is seftlth6 actu6lRCS'pressure must be

< 425 psid t6 open the vali '.This'setp6int ensures the RHR design pressure will'not be6xceeded and the RHR relief valves will not lift. The 18 month Frequency'is basedor thenee to perform the Surveillance under'coniditiions that apply'durrng' alant outage. The 18 month Frequencys isalso acceptable' led brVconsideration of the design reliabilIft (and confirmiing oprating experience) of the equipment. This

. ., ,SR,is not required to,b6 p1ome-d eAnsthe RHR suction isolation valves are open to satisfy LCO a.4.12 r>' -:'*".

1.: FR

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REFERENCES 1. 10 CFR 50.2. . t I

- .. .I..

• 2. 10CFR50.55a(c).- ,';-:.t. , i. .

3.:- l10 CFR 50, AppendixA,,Section.V; GDC 55.'

,. - ;, ,, *)

.: I

4. ' WASH-1400 (NUREG-75/014), Appendix V,';October 1975.

5: 'NUREG-0677, MaVi1980.'  ;

6. ASME, Boiler and Pressure Vessel Code, Section Xl.

i'- * ,  ; Z;. Peru I

Wolf Creek - Unit 1 B 3.4.14-6 .' :, -.- Revision 16

• ;, . . e , '<': Accumulators

. lB 3.5.1 BASES,,.', ,--:

SURVEILLANCE SR 3.5.1.2andSR 3.5.1.3 , Is - A:  :

REQUIREMENTS i , _,,_t _ ,,, ,'

(continued),,, . ..Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, borathedt-waterhvolume and nitrogen-pover pressure are

'erified

" for bo'race waater volume an togtenwvater volume is equivalent td 22i;4ilO'anh q 77'8%' level. Only one 'set of non-safety

'd i channels(1 of 2) is required fdr water level and pressure indication. The I2-hour Frequency issufficient to ensure adequate injection during a lLOCA. 'Becuse of the' ttic'des'ign 'of the accum'uiator, a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

-' Frequency usually 'll'ows the operator to identify changes before limits are

'- *-';* -> reached:n' Opefatiri-g-6xperIen6 e'has showh this Frequency to be, i - ' "'"appropriate for early detection and correction of off normal trends. -

SR 3.5.1.4 i@,.I-;,

; r. -:

i f The boron concentration should be verified to be within required limits for each accumulator every 31 days since'the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day

-'

• ;t *;!Frequencyis adequate to identifychanges that could occur from 4 t r ;rech m anismns such as dilution orin!eakage. KSampling the affected

' t ^ .-accumulatorwwithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.after a 70 gallon increase (approximately 8%

- . ,evel)

I wilJ identify whether.iideakage has caused a reduction in boron

; _ ., concentration to belc;6th6, required limit. It isjiqt necessary to verify boron

.'conen'tratio~niffthe added water rbventory is from the refueling water storagetank RWST) andthe.BWST has not beer diluted since verifying

.. 'ithat its boron cocriioatisfies SR 3.5.4.3, because the water

'i, ;c a.' r~cnntainedd

~

~'

in69 VRWST s'Ynorr60ally within the' dccumulator boron r: '!

i.c to iconcentrati6n r'equirer-e-ts, iThis' is 6onsistent with the recommendation of NUREG-1366 (Ref. 4). - -' 'a' S R 3,5.1.5 ,, ,,;-;

Verification every'31 days thatpower is removed from each accumulator isolation valve operator when the RCS pressure is > 1000 psig ensures that an active failuretcould. not~result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two it,ar;,accumulators would-be available for injection given a single failure coincident with a LOCA. Since power is removed under administrative control, the 31 day Frequency will provide.adequate assurance that power is removed.

'Thi SR allows 'power c be sup~plied to the motor operated isolation valves when RCS pressure is 5 1000 psig, thus allowing operational Revision 16 B3.5.i.7.

Wolf Creek Creek - Unit - Unit I1 -B 3.5.1-'7. 1:; - -1 Revision 16

Accumulators B 3.5.1 BASES '.-. ... Y.,.- .'

SURVEILLANCE SR 3.5.1.5 (continued) r'>,

REQUIREMENTS flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. .

• Should, closure of a valve occur in spite of the interlock, the Si signal',

provided to the valves would open a closedvalve in.the event of a LOCA.

REFERENCES .1. ,USAR, Chapter 6. ,- . ,. .,-

2. 10 CFR 50.46.

3. ARiChapter.15.:,:

3-. .U

4. NUREG-1366, Februaryr990.r,  ;-

- ~ 5.- WCAP-15049-A, Rev. 1April 1999. '

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Wolf Creek - Unit 1 B 3.5.1-8 Revision 1

, .T),t- -,- ..: .a.,!. . ECCS - Shutdown i .--. -_z - f.; B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

. I ! ,

B 3.5.3 ECCS - Shutdown BASES BACKGROUND .- The Background section for Bases;3.5.2, tECCS -Operating," is

.t. ' ": *.> ' applicable to these Bases, with the following modifications.

In MODEO4, th6e reqiuir6d ECCS train consists of two separate subsystems: centrifugal charging (high head) and residual heat removal (RHR) (low head). .

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the refueling water storage tank (RWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases'3.52.-.

APPLICABLE .-, -. The Applicable Safety Analyses section of Bases .3.5.2 also applies SAFETY ANALYSES to this Bases section.

4* . - , - , I . . -' . I;.

I Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic safety injection (SI) actuation is not available.' In this MODE, sufficient time'exists for manual actuation of the required'ECCS to mitigate the consequences'of a DBA.

For' MODE 3, with the accumulators blocked, and MODE 4, the parameters assumed in the generic bounding thermal hydraulic analysis for the limiting DBA (Reference 1) are based on a combination of limiting parameters for MODE 3, with the accumulators blocked, and parameters for MODE 4. However, assumed ECCS availability is based on MODE 4.

conditions; the minimum available ECCS flow is calculated assuming only one OPERABLE ECCS train. .;

Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. 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 is required to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a DBA.

Wolf Creek - Unit 1 B 3.5.3-1  :., ..- Revision 16

ECCS - Shutdown B 3.5.3 BASES -

LCO In MODE 4, an ECCS train consists of a centrifugal charging subsystem a (continued) and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure anOPERABLE flow path capableof taking suction from e; . - t RWST and.transferring suction tothe the t containmdnt sump.

Z During an event requiring ECQS actuation, a-flow path is required to provide

' ' an abundant supply of water, from the RWST to the RCS via the ECCS

';' In th andtheir respective spply headers to two cold leg injection nozzles.

In the long term, this floW path nay be switched to take its supply from the

' 'containment Su and tdeliver its flow to the RCS hot and cold legs.

t( This LCO'is'mnodified bya Note that aIlos an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, if

' ' capable c' of being manually fealignred (renmotd or local) to the ECCS mode of operation andnototheraise inoperable. 'This allows operation in the RHR rmodedurinrg MODE 4: 2. .- .*i.:  % .. -5ri

~~~ ,. - ', ..  ! . .... V' . , ........

-.,, _C ,a,.... j.

APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 4 with RCS temperature below 350 0F, one OPERABLE ECCS train

.. is acceptable withoutsinglefailure consideration, on the basis of the stable

. .eactivity of the react6r and.thellimitedcore cooling requirements.

.In MODES5. and 6, plant conditions are sch.'thatthe probability of an event requiring ECCS inj66t6n is extremely`QW. Core6,866ing requirements in MODE'5 are addressedbj Ld 53.4"CS Lo0ps MODE 5, Loops Filled,"

'and LC0324.8, "RCS L'ops 6MObE'Loops Not Filled." MODE 6 core co6ling requiremenis arezaddresse-d ty LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level."

ACTIONS- A. I

- i I - _- '.* C p! 1, ,! 1; C *1 , 8; 1!8{ I With no ECCS RHR subsystem OPERABLE, the plant is not prepared to

.respond to a loss of coolaht accident or to' continue a cooldown using the

• -.- ,1e!

. ;, .Z :-i ;;  !; ;.-*, ;t Wolf Creek - Unit 1 B,3.5.3-2  : l Revision 16

ECCS - Shutdown

. ;.:B 3.5.3 BASES ACTIONS,'e ' ' AJ.i (cb'ntinued)!' ; '

,~ ~: i- 1 '-4, " f. , ', ., .' 'a t; rmF RHR'un'is a hatexchangers. The Completion Time of immediately to initiate a'ctions that would restoreat least one ECCS RHR subsystem to

., ,~ , -. OPERABLE status ensures that prompt action is taken to restore the required cooling capacity., Norrnally, in "MODE 4, reactor decay heat is

.j , - *S .,

' emoved from the"RCS by an RHR loop. If no RHR loop is OPERABLE for this function,',reactor'decay heat must be remoVed by some alternate 4I.method, such as use of the steam generators. The alternate means of heat removal must contin-ue until the inoperable RHR loop components

, cjan be restored to operation so that decay heat removal is continuous.

- i; i With bothRHR pumps and heat exchangersinoperable, it would be

• . ,,~_ ,;. ¢.,*. ,unwise to require the ,plant to go.to MODE 5, where the only available heat removal system is the RHR;' Therefore,-the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status.

B.1

.2>; .'1 I' J ,. .l. A% I

' iLA$ * 'Witho'no l ECCS-bigh-head iubsy'stem OPERABLE, due to the inoperability

.o i' i of the centeifuig'al dharging'gtum-'ou r flow"p'ath from the RWST, the plant is

,, not prepared to provide high pressure response to Design Basis Events

• 5-1 1requirirSIT TirEnto restore at least one ECCS

, nt i'd.* , '.high'head subss'st'e1to' OPERABLEstatus ensures that prompt action is

', taken td prvidetis requiredd&6liri' capacity o'tinitiate actions to place

, ,,.,; the plant in rMOD,"5, where" h ECCS train is~not required.

C.A When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown should be initiated.

. ,; ,- >; -Twenty-four hors:isa reasonable time; based on operating experience,

.:;;; '-, l: :'Miu':. oto reach MODE 5 in-an orderly manner and without challenging plant systems or operators.

SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply.

Wolf Creek - Unit 1 B 3.5.3-3 i - Revision 0

i.. ECCS - Shutdown B 3.5.3 BASES REFERENCES The applicable references from Bases 3.5.2 apply.,.-'

I

• ,***I - WCAP-12476, Revision, 1,. "Evaluation of LOCA During Mode 3 and A'.; Mode 4 Operation for,Westinghouse NSSS,' November 2000. I I I ,_-. o .

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Wolf Creek - Unit 1 B 3.5.3-4' Revision 16

I . ' .

Containment Spray and Cooling Systems B 3.6.6 BASES _ - .... . _ - ... ._ . . ..

ACTIONS F.1 : -..i I -' .

(continued) - I t

,j - - -' * - With two contairnient spray.trains or ariy combination of three or more

• confairiment spray and cooling trains inoperable, the unit is in a condition

-. outside the accident analysis. Therefore,.LCO.3.0.3 must be entered immediately. -

SURVEILLANCE -SR -.3.6.6.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the containment spray flow path provides assurance that the proper flow paths will exist for Containment Spray System operation. The correct alignment for the Containment Cooling System valves is provided in SR 3.7.8.1. This SR does not apply to manual vent/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 in the correct position prior to locking, sealing, or securing. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown (which may include the use of local or remote indicators), that those valves outside containment and

- " capable of potentially being mispositioned are in the correct position. The 31 day Frequency is based on engineering judgement, is consistent with administrative controls governing valve operation, and ensures correct valve positions. . -

SR 3.6.6.2 Operating each containment cooling train fan unit for 2 15 minutes ensures that all fan units are OPERABLE. It also ensures the abnormal conditions or degradation of the fan unit can be detected for corrective action. The 31 day Frequency was developed considering the known reliability 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 to

'be acceptable through operating experience.'

SR -3.6.6.3 Not Used. -

--- ~- - vSR 3..6.4 . : .>..

Verifying each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance BASES Wolf Creek - Unit 1 B 3.6.6-7  : - Revision 0

Containment Spray and Cooling Systems B 3.6.6 BASES -I SURVEILLANCE SR 3.6.6.4 (continued). l 14, .

REQUIREMENTS required by Section XI of.the ASME Code, (Ref. 5)., Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.

. .' ", .  :;, ¢. Such.inservice tests confirm component OPERABILITY, trend performance, anid'detect incipient failures by abnormal performance. The Frequency of the SR is in accordance with the lnservice Testing Program.

This test ensures that'each pump develops a differential pressure of greater thiri or equal to 219 psid at a nominl fldw'6f 300 gpm when on I f  !:.._.,...t recirculation'(Ref.' 6)f:- ' .

J .. '_. ....... _'..... ._.. '

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SR 3.6.6.5 and SR 3.6.6.6 ' . : ,

• ' These SRs'require'verification that' each automatic containment spray valve actuates to'its-correct position and that each containment spray pump starts upon receipt of an actual or simulated actuation of a

- ** - containment High-3 pressure si3rial. ';CThis Surveillance is not required for

-valves that are locked, sealed, or otherwise secured in the required position under administrative conirols...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 an unplanned transient if the Surveillances were performed with the reactor-at power. Operating experience has shown that these components usually pass the Surveillances when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

The surveillance of containrment sump isolation valves is also required by'

- 'k-

• SR 3.5.2.5.-'A singie suirveillanice may be uised to satisfy both requirements' ,;

• . . -a  :  : .f;. -: .,. ..  :

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• SR '3.6.6; 7 -:  !~ '  :.  :  :............ .

-I:.- This SR requires verification th6t each containment cooling train actuates

.;."upon receipt of an actual or simulated safety injection signal; Upon actuation, each fan in the train 'starts in slow speed-or, if operating, shifts to slow speed and the cooling water flow rate increases to Ž 2000 gpm to each 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 anddSR 3.6.6.6, above, forfurtherdiscussion of the basis for the 18 rnonth,,Frequency. ,  !

Wolf Creek - Unit 1 B 3.6.6-8 Revision 14

d. i t } l  % X}AFW System B 3.7.5 BASES ..

APPLICABLE,: a.; Feedwater Line Break (FWLB);,' ,-;' .

SAFETY ANALYSES' i' : '.: -

(continued) - b. ' -Main Steam Line Brdak;,ard "

iLoss f MFW.-"'e-

,; . ,In addition, the minimun avallable AFW flow and system characteristics

. " *8,_ are considerations in'the analysis of a small break loss of coolant accident (LOCA). The AFW ,Systerndesign i' 'such that it can perform its function j'..,,Jollowing an FWLB between the MFW isolation-valves and containment,

- . mp.

be'%-aer.' .. 'i t *trI -- .a o 1 . bl'

, ,Combined with a loss of offsite power.following turbine trip, and a single active failure of one 6t8r driven'AVV pump. ,This"results in minimum assumed flow to the intact steam generators. One motor driven AFW pump would deliver to the broken MFW header at a flow rate throttled by

-the motor operated "smart! discharge valve until the problem was detected, and flow terminated by'the operator.'; Sufficient flow would be

, i i-ci delivered to the intact Steam generator by the residual flow from the

- s:'n .;.;affected.pump plus the turbine~driven AFW purmp.

.; tu ,jir.!..l h tf'!" ;? Pi I: ',-r--g k'iF- - * . Je.*-

' '! ' The E3P'ESFAS automatically actuates the AFW turbine driven pump

" ,'! iii when required to ensure an adequate feedwater supply to the steam

" . ';.'~ .:.;;generatorsduiring 1o'ssof power;' DC powerioperated valves are provided

,*for each'AFW~line to-cbntrol' the'AFW'flow to each steam generator.

.! Ot.-.i . @ I;'2 fq-~'ijwl9C do..9. ,-l..

V,;"¢ C 1C '-**le..

.C{,i;il:, '

- Thi6AFW Swsfm satgsfies thb'requhremrnts' of Criterion 3 of 10 CFR

-,~k F - 50.36(y)(2y(1rflc.oi 'Er*,f.'.Ad2; '

tg*; . !- (')

_ h - *~2. * ._ -

LCO This LCO provides assurance that the AFW System will perform its design

d. ,',,.KSafety furcior toitigte theconsequences of.accidents that could result overpresuraion o'f the ~reactor coolant pressure boundary. Three

,,,in independent AFW pumps in three diverse trains 'are required to be OPERABLE to ensure the availability of decay heat removal capability for

- all events accompanied by a loss of offsite power and a single failure:

This is accomplished by powering two of the pumps from independent f I,* ,emergency J buses: jThe third AFW pump is powered by a different means,

.-a steam driven turbine supplied with steam from a source that is not

. isolated by'closure of the.,MSIVs. ;,:

' i, an The AFW.System is connfigured into three trains': The AFW System is

't f ~'

• .'l~Uc'orsidered OPERABLE'when the components' and flow paths required to

• provide redundant'AFW6l&W tothe steam genbrators are OPERABLE.

This requires that the twvomrrfot6r driikn AFW pumps be OPERABLE in two diverse paths, each capable of automatically transferring the suction from Wolf Creek - Unit 1 B 3.7.5-3 I- I Revision 0

AFW System B 3.7.5 BASES LCO 'the CST to an ESW supply and supplying AFW to two steam generators: I (continued) The turbine driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIVs,

' and shall be capable 6f autom tically transfe'#ririn the suction from the CST to an ESW supply and supplying AFW to any of the steam generators. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE. .The inoperability of a single supply line or a single'suction isolation valve from an'ESW train to the turbine driven AFW pump causes'a loss of redundancy in ESW supply to the pump but does not render the turbine driven AFW train inoperable. The supply line begins at the point whe're the ESW'piping branches into two lines, one supplying the motor driven AFW pump and one supplying the turbine driven AFW pump, and ends at the suction of th6 turbine driven AFW pump (Ref. 3).

' '. ;; ' Therefdre, with one ESW train inoperable; thb associated motor driven AFW train is considered inoperable; and one turbine driven AFW pump supply line is considered inoperable. However, the turbine driven AFW train is

. OPERABLE based on the remairing:OPERABLE ESW supply line.

In order for the turbine driven AFW pump and motor driven AFW pumps to

. be OPERABLE while the AFW` System1is in automati6 control or above 10%

RTP, the discharge flow cohtror valves, shall be in the full open position.

. The turbine an' motor-driven AFW pumps remain OPERABLE with the discharge flow control valves throttled to maintain steam generator levels

• .during plant heatuo, coo'dovin, ofif~started du6 to ah Auxiliary Feedwater Actuation Signal (AFAS) or'mahiually start6d in anticipation of an AFAS.

.  ; The nitrogen accurnulator tanks supplying the turbine driven AFW pump

• '% control valves and the steamrgentratbrfrtmospheric relief valves ensure an eight hour supply for the pumpland vafves. -

Although the AFW System may be used in MODE 4 to remove decay heat, the LCO does not'require'the AFW System to be OPERABLE in MODE 4 since the RHR System is available for decay heat removal.

APPLICABILITY In MODES 1; 2, and 3, the AFW-System' is required to be OPERABLE in the

!. ' *;event that it is called upon to function'when'the MFW is lost. In addition, the

-. AFW System is required to supp!y enough makeup water to replace the

• ;.  ; _-steam generator secondary. inventoryilost as.the unit cools to MODE 4 conditions.

In MODE 4'the AFW System may be used for heat redmoval via the steam generators but is not required since the RHR System'is available and required to be OPERABLE in this MODE. .

Wolf Creek - Unit 1 B 3.7.5-4 Revision 16

AFW System B 3.7.5 BASES APPLICABILITY InMODE 5 or°6, the steam generators are not norma1ly used for heat (continued) .  ; removal, and the AFW System is not required.

ACTIONS A1;

If one of the two steam-suppliesto the turbine driven AFW train is

~*,tt .' ' inoperable, action must be taken to'restore OPERABLE status within 7 days.

*, The 7 day Completion Time is reasonable,'based on the following reasons

,ii. 8: . .tr, ,,-, !,a.., ,3-, The redundant OPERABLE steam supply to the turbine driven AFW

• • ),- . ,-,

-The  : availability of redundantOPERABLE motor driven AFW pumps;

,,. . il:. and!, i,.

... 9i. r 7-'-

• Oo  ;. c..,  :.-The low probability of an event occurring that requires the inoperable steam supplyto the turbine driven AFW pump.

! mi, ,-Thesecond CompletionTime for~Required Action A.1 establishes a limit on

"a.

!*ic r,.-. ,il;Y the maximum-time allowed for~anylcombination of Conditions to be

- .. -inoperable,during any continuous failure to meet this LCO.

J J ', The:10 day Completion Time provides a limitation time allowed in this

,specified Condition after discovery of failure to meet'the LCO. This limit is considered reasonable for situations in which multiple Conditions are

,. entered concurrently. The AND-connector between,7 days and 10 days

• , ,?x 'dictates that bothComnpletion-Times apply simultane'ously, and the more restrictive mustabem 't., ier - '

. , .1  : . ., .. , .. ,, , . .. .r *6. .. -

With one of the required AFW trains (pump or flow path) inoperable for

-- .reasons-other'than'Condition A;' action must be taken to restore OPERABLE w*-

,r,:': ,*,status within 72,hours.'.;This Condition includes the loss of two steam supply

,(,.:t-: lines to the turbine driven AF.W pump. .Th6 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is a:.. : . ,:m *,.m. reasonable;.basead on. redundant capabilities afforded by the AFW System,

os ;.: time needed for repairs, and the low proba ity o a DBA occurring during this time period. - h l

;,i*.xor'The second Completion Time for Required Action B.1 establishes a limit on

,2r.,&.I,'r?.+S-,  ;.;the maximum time allowed for any~combination of Conditions to'be inoperable during any, continuous failure to meet this LCO.

Revision 16 I B 3.7.5-5 Wolf Creek -- Unit Wolf Creek Unit I1 B 3.7.5-5 . , ; - Revision 16 1

AFW System B 3.7.5 BASES

.. I ,  ; . - . J5 * .- . ! T ACTIONS B.1 (continued),

. - The 10 day Com'pletion Time provides a limitation'time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The AND connector betwveen 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 10 days dictates that both Completion Times apply simultaneously, and the more i1 restrictive must be met. - r  ;.

I , ,  :* # I. ,,  ;- ; .. .-'

CA and C.2 a a. - ..

When Required Action A.1 or B.1 cannot be completed within the required

• . iCommpletion Time or if two AFW trains' are inoperable, the unit must be

- placed'inta MODE'in'which' th LCO'does'not apply. To achieve this status, the fiit'miiust be placed intat 1east MODE 3within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The allowed Completion Times are reasonable-, based on operating

,experiences, to reach therequired unit conditions from full power

• conditionshan orderly manner and without challernging unit systems.

• '- ' . -. *-D

, D.1 i. j1 - -, l * > . ;-;- -.

• A S> S.. ...I. .......

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A If all three AFW trains are inoperable, the unit is in a seriously degraded condition.with no safety related mearis for conducting a cooldown, and only limited means for conducting a cooldown with nonsafety related equipment. In such a condition,ithe unit should not be perturbed by any

• . - action; including a poWer'change, that might result: in a trip. The

• seriousness of this condition requires that action be started immediately to

- restore one AFW train to OPERABLE status-:

Required Action D.1 is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AFW train is restored to OPERABLE status; -In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW Wolf Creek - Unit 1 B 3.7.5-6 Revision 0

AFW System B 3.7.5 3 ,.: .. . -8*

p;c BASES v  :

SURVEILLANCE SR 3,7.5.1 (continued) , .;. --

REQUIREMENTS operation. This SR does not apply to valves that are locked,' sealed, or i;therwise secured in position, 'since they are verified to be in the correct positi6n prior't6oldckinb;' sealing,'6r'securing. This SR also does not apply

• I: t::-i>t ~ to'manual vent/drain valves,-and to valves'that cannot be inadvertently

.*'<r,* ' *-"' 'misaligned, such as check valves. 'This Surveillance does not require any

.- "i~'v. .'! testing or valve manipulation; rather,,it involves verification that those valves capable of being mispositioned are in the correct position.

The 31 day Frequency, based on engineering judgment, is consistent with procedural controls governing valve operation, and ensures correct valve positions. ;i

• 'i'; This.SR is.nodified by a.Note indicating that the SRis not required to be X. lperformedfor,theAFW

.v:~ ,-:- flow control valves until the AFW System is placed in

..standby porTHERMA.L PNOWER is above 10O%.RTP,I SR 3.7.5: F. ,

iVerIfytg that each 'AFW purp's developed head at the flow test point is

• ' n -greater than or equaitothe required developed head ensures that AFW

-pump-performance has not degraded during the cycle. Flow and differential

'head are normal tests of centrifugal pump performance required by Section Xl of the-ASME Code (Ref. 2). Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is

. n performed on're'ir'culation flow;' 'This t6st confirms one point on the pump

' A,  :';'; .: t design curve and is'indicative of overall performance. Such inservice tests F s-l confirm component.OPERABllITY;,trend performance, and detect incipient X ^ *¢ ,; ;._failures by.indicating abnormal performance. Performance of inservice 5 .

s-ri *j w' ' testing discussed in the ASME Code, .Section'XI (Ref. 2) (only required at 3 v:W,* r': E ,jm CF-month intervals).satisfies this.requirement. The test Frequency in accordance,with the Inservice 1 Testing Program results in testing each pump once every 3 months, as required by Reference 2.

L. I. -

a Fi il 1 -  ;;ji fto;

• 6, ?!S*  ; 3**.'*<

Ii-' . : '* o - .

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.. i . .

Wolf Creek -_Unit I B 3.7.5-7, WU .7 . Revision 16

AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.2 (continued). i:. . ' . ., ,. v 9.

REQUIREMENTS When on recirculation, the required differential pressure for the AFW pumps (Ref.,4) when tested in accordance with the Inservice Testing Program is:

Motor Driven Pumps'" 1514 psid at a'nominal flow of 110 gpm Turbine Driven Pump 21616.4 psid at a nominal flow of 130 gpm i '- This SR isrnodifhed by a N6te indicating thatthe SR should be deferred until suit'able test conditions are established.. This' deferral is required

' " because there is irisufficidnt steam pressure to p6rform the test.

" ,- .. '  ; 1.; ' 'd v . ' . . i i'-  : i*

SR 3.7.5.3 '-'. ..-.

• " This SR verifies that AFW clan be delivered to the appropriate steam generator intthe of airi 6cid6ht cvent or transie'nit that generates an a"' ESFAS, by'deronstratingjthat eachi automrtic'valve in the flow path

• . actuates td its cdrrect positfiV'on an actual or simulated actuation signal.

This Surveillance is not required for valVes'that are'locked, sealed, or otherwise secured in the required position under administrative controls.

- ; The 18 mohth Frequency is basted orn the' ne' d to 'perform this

- Surveillance under the dcdditiohs'th'at apply during a unit outage and the

- potential for an unplanned trarfsibnt'iftheSurveillance were performed

- .; * - -with the reactcr at power. 'TheA 8 mohth Frequency is acceptable based

' on operating experiencrdand'thd design r'eliability of the equipment.

-'This SR includes the requirement to verify that each AFW motor-operated

- -: discharge valve linits the flowv from tHe'riotor'driven AFW pump to each

• i- ' 'steam'ge'neratorto <320'gpmrand ttat'valves in the ESW suction

• ' fowpath-actuate to the'full'obpe'rlpd'ition upon receipt of an Auxiliary Feedwater Pump Su6tion Pressure-Low signal.---:

SR 3.7.5.4 .

.:' -;

- ,, i.. ., ,,, ,... ., ;, 'C This SR verifies that the AFW pumps will start in the event of any accident

' . T "or transient that generates an AFAS by demonstrating that each AFW

- pump starts automatically on an actual or simulated:actuation signal. The 18 month Frequency is basedon the, need to perform this Surveillance under the conditions that'apply during a unit outage.-and the potential for an unplanned transient if the.Surveillance were, performed with the reactor at power.

Wolf Creek - Unit I B 3.7.5-8 Revision 14

CREVS

• B 3.7.10 B 3.7 PLANT SYSTEMS - -'

B 3.7.10 Control Room EmergencyVentilation System (CREVS)- I;.1^ X .

- i * -z.- 4i:; f. -: .~' . - :fl  : . '. i Lw:

BASES - - -. . -i * ' ,. i-BACKGROUND The CREVS provides a'protected,' controlled temperature environment t, a,  :. ' from which operators can control the unit following an uncontrolled release of radioactivity. - ga u nl

,f ..  ;.  ; . i,,-,  ; .-. 4., -v wd

' The CREVS consists of two independent, redundant trains that

• recirculate, cool, pressurize, and filter the control room air. Each CREVS

• '.1 train &onsists off a recircu'.rtin,and ation sys;e'm a pressurization system

,train.Th6eairbconditioniria each frain consists of a fan, a self-

'otionof contained'refrigeration system, and'a prefilter."The filtration portion of each system'consists of a high efficiency particulate air (HEPA) filter, an activated charcoal absorber section for removal of gaseous activity (principally iodines), and a second HEPA-folows-the absorber section to

.$8: t2 ,-  :-. collect9ca~rbon finesi Each pressurization system train consists of

--, ductwork to bring airfrom outside the building, a moisture separator, an

< *.,,' , e!ectric heater, a HEPA; an activated charcoal adsorber, and a second

'1-; I  ; . 1 .HEPA. Ductwork, valves,or dampers', and instrumentation also form part

-- m; ;e-,'S! ,

'( of the systern^.\. . , .; .

• -*a;/

tl-t*-iti-1l Y It;-, r;x  ; t~lir0;>-'

. -ab.n b ,.< The t CREVS is-an emergency system which may also operate during

.-.,  ;- ,normal 1 unit-operations;: Upon receipt of the actuating signal, normal air

^, -. , .--, ;:,! supply andee ust othecontrol room is isolated, and a portion of the

' *. . ventilation air, is recirculated -through the:filtration system train(s), and the pressurizatiop.systemlis,started. The filtration system prefilters remove any large particles in the air, and the pressurization system moisture

-~'rctt -. z . v'R,'; separator removes any entrained waterdroplets present, to prevent tii '+ i.':lir 4. excessive loading of the HEPA filters and charcoalxadsorbers.

.,;a;,

i.-3 Continuous,operationof each pressurization trainforat least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per e;, -, month, with theeaterfunctioning, reduces moisture buildup on the HEPA filters and adsorbers.:-,The heaters are important to the effectiveness of the charcoal adsorbers:'

Actuation of the CREVS by a Control Room Ventilation Isolation Signal (CRVIS), places the system in the emergency m'ode of operation.

e .  :... Actuation ofithe system to the emergency mode of operation closes the

• - . . -. ;Wnfiltered outside.air~intake' and unfiltered exhaust dampers, and aligns L '.;. ithe

.. system for'recirculation.I A portion'of the recirculation control room air l-r ,.flow through the redundant filtration system trains of HEPA and the

,,;r charcoal adsorbers.: The CRVIS also initiates pressurization and filtered

,C:Jw*, f: ;} j';':, join ventilation of the air st'pply'to the control room.'-

Wolf Creek'- Unit I B 3.7.10-1 :Revision 0

. .* CREVS B 3.7.10 BASES .

BACKGROUND Outside air is filtered, diluted with air from the electrical-equipment and' (continued) cable spreading rooms, and added to the air being recirculated from the control room. Pressurization of the control room prevents infiltration of

= ._ unfiltered air from the surrounding areas of the building...,

The air entering the contrbl buildinrg during normal operation is '-l- t'C

- continuously monitored by radiation and smoke detectors. A high radiation signal initiates the CRVIS; the smoke detectors provide an alarm in the control room. A CRVIS is initiated by the radiation monitors

'- (GKRE0004 and GKRE0005), fuel building ventilation isolation signal,

- containment isolation phase Ai containment atmosphere radiation

monitors (GTRE0031 arid GTRE0032), containment purge exhaust

. -. - a. tradiation monitors'(GTREO22 and GTRE0033),-or manually.

.4 . ..

A single train vill 'pressurize tha control room to 2 0:25 inches water gaug-e. -The CREVS operation'in `mainiaining the control room habitable is

-:discussed'in the USAR, Section 6:4 (Ref; 1)-l Either 6f the pressurization and recirculaiicn trains provide the required filtration and pressurization to the control room. Normally open isolation

' damperara, e arranged in series pairs 'so that the6failure of one damper to shut will not result in a breach of isolationri-The CREVS is designed in

accordance with Seismic Category I'requirements. "

.:: . *- . . -: * .'.  :' . v.:.!

The CREVS is designed to ima~htain the control room environment for 30 days' of continuous occupancy affer'a Design Basis Accident (DBA) without exceeding a 5 rem whole body dose or its equivalent to any part of

.^- , . !,r!ache body (Ref.'2)-; - ' "e  ;

APPLICABLE..' The CREVS components are'arranged in redundant, safety related SAFETY-ANALYSES ventilation t6ains: The location of comaponents and'ducting within the control room envelope ensur-es' an adequate supply of filtered air to all

areas requiring access! The CREVS'jrovides airborne radiological

'protection for the control rocboperators, as'demonstrated by the control room accideht'dose analyses for the' most limiting design basis loss of

• -~*I . coolant accident, fission product, releasepresented in the USAR, I \ - Chapte'r,15,'Appendix 15A'(RefP,2),.' i . : t'j

-:.-. fa....

i'>*f -  :.; 9; i: 9. .-

The worst-case'single active failure of a component of the CREVS, assuming a loss of offsite power, does not impair thie ability of the system to perform its design function.

Tithe CREVS satisfies Critericrr3 of 10 CFR 50;36(c)(2)(ii).

Wolf Creek - Unit I B 3.7.10-2 Revision 15

, Containment Penetrations

, .. B 3.9.4 B 3.9 REFUELING OPERATIONS -

B 3.9.4 Containment Penetrations ATr . ' -  :' ' B; a:; * ', ,

.  !:-?e -'ir ^, -: 4~~'

!;'tr ,; . .: - :

• +~.§Ah ,e sr:fwj t;. a.4. ' " 'C

.tC9 :t. '.- I,. e i .-

BASES -. -

BACKGROUND . . :During CORE ALTERATIONS or movement of irradiated fuel

n. assemblies within .containment, a release of fission product radioactivity

, within containment will be restricted from escaping to the environment i:.f- r. ! i:-owhen the LCO requirements are met. In MODES I, 2, 3, and 4, this is

1.,~-.I r,-,,;,.. 4 ,raccomplished-bymaintaining containment OPERABLE as described in

- -::so; LCO :3.6.1 ,,"Containment.".,In MODE .6,the potential for containment

,,- ,,, r .. : .pressurization-as a result of'an accident is not likely; therefore,

- Al",u,,irequirements

-. to isolate theicontainment from the outside atmosphere can be less stringent. The LCO requirements are referred to as

-t. ,,. : .  ;."containment penetration.closure", ratherthan 7containment

,:l S i aiOPERABILITY.". Containment penetration closure'means that all potential escppe paths are closed or capable of being closed. Since there is no potential for containment pressurization, the 10 CFR 50,

-., . ,-; ,9,>,;-.ppendix-J,leakage criteria and tests are not required.

-, ' aj,..The .. containment seryes to contain. fission.product radioactivitythat may

, ,be releasedjrom the;reactorlcore following an accident, such that offsite radiation exposures are maintained well-within the requirements of 10 CFR 100. Additionally, the containment provides radiation shielding

,; ,) ,Vff.%^ from the fission pyroductsjhat may.be present in the containment

,/,4, i, (

ii,,.,;

atmosphere following accident conditions> ;,. v The containment equipment hatch,-which is part of the containment

- . pressure boundary, provides a means for moving large equipment and

- ---- - '-~----~- cornponernts-lnto-and out of containment. If closed, the equipment'hatch-

!, ,,,j,>s.Ar,, .must be held-ioplace.byat least fourbolts:.Good engineering practice

-,r.1 ,infv,c;,dictates that the boltsyrequir.pd by this LCO, be'approximately equally

., . ,-,spaced.-%Th epquipment.hatch may be open during CORE a . ALTERATIONS ormovementof irradiated-fuel assemblies within

.r ) # era ;ulA it. t containment;.prpovided itcan-be installed with arninimum of four bolts

';:-.~. ,.- is ~' n,;,,i, 'holding it in place: During Shutdown conditions, adequate missile iv',r; e<. nprotection for sazfety4 related equipment in containment is provided with the equipment hatch~beld in place with 6 bolts. Administrative controls

- ensure'the equipment hatch is in place during the threat of severe i.:7',  :.n. ,.weather that could result in the generation of tornado driven missiles.

The containment air locks, which are also part of the containment

,- ;4,pressure boundary, provide a means for.personnel!~access during J . . - ,,..- .  :  :,,, .; .. . . ...

Wolf Creek - Unit 1 B 3.9.4-1  ; Revision -14'

. .. r , , " ' 'I H- L- '  : Containment Pe 'etrations

... .. ;.. ..... . .1, .. i .I .I . .B.3.9.4 I !, i ? - ,;

BASES BACKGROUND MODES 1', 2,'3, and'4'tunit operation in acc6rdanc6With LCO'3.6.2, (continued) - "Containment Air Locks." Each-air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown when cntainnirit penetration closure id not required, the 's door interlock-mechanism may be disabled, allowing both-doors- of arr-air lock-to remain open for extended periods when frequent containment:

entry is necessary.' During CORE ALTERATIONS or movement of  ;;.

irradiated fuel assemblies within containment, containment penetration clbsuredis required; howevr, the do interlock mechanism may remain

'disable'd provided o&e peirsonnel'air locKdoor is'capable'6f being-closed and one emergency air lock door is closed. In the case of the  :

emergencyair lock door, a temporary closure device is an acceptable'

; ., replacement for the air lock door (Ref. 1).-. -

. .: ...-. i}

The requirements for containment penetration closure ensure that a-release of-fission !product radioactivity within containment will be - -

restricted from'escaping to the environment.--The closure restrictions -

are sufficient to restrict fission-pIfoduc't radioactivity release from' -

containment due to a fuel handling accident during refueling.

s.  :

The Containrndnint Purge System includes!two subsystems. The <.

shutdowln purge subsystem includes a 36, inch'supply penetration and a

.L :

36 inch exhaust penetration. The second subsystem, a mini-purge . '

• L system, includes'an 18 inch supply penetration and an 18 inch exhaust r

, . , L.1- penetration. iDuring MODES 1, 2, 3, and 4, the two valves in each of the shutdown purge supply and exhaust penetrations are secured in the,;

,. N closed positiori or blind flange installed. JThe two valves in each of the two minipurge penetrations can be opened intermittently, but are 616ed-l automatically by the Engineered Safety Features Actuation System -

A\ (ESFAS)j Neither of the subsystems is suibject to a Specification in

.v. . .

MODEr5 oir MODE 6 excluding CORE ALTERATIONS or movement of irradiated fuel in containment.;--; a.-. -

. . In MODE 6; large6air exchanges are necessary to conduct refueling .

operations.- The normal 36 inch purge system is used for this purpose,

..- . and all fourvalves may be closed by the ESFAS in accordance with.

.. : LCO 3.3.6, "Containment Purge Isolation Instrumentation," during -

CORE ALTERATIONS or movement of irradiated fuel in containment. -

When the minipurge system is not used in MODE 6, all four 18 inch '

valves are closed.

4> ....  : . ,.,;

The othercontainrnent penetrations that provide direct access from containment almosphere to outside atmosphere must be isolated on at.

I Wolf Creek - Unit 1 B 3.9.4-2 Revision 13.: 1

LIST OF EFFECTIVE PAGES - TECHNICALSPECIFICATION BASES PAGE(') ,, REVISION NO. (2) CHANG.DO UMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)

TAB-Title PageTe~chnicaT'Specificatioh '::  ;  ;;

CoverPage

TitlePage; * ' ' . .- 5,

• _

TAB -Table of ContAmend. No. -123 12118/99 i l'i i ' J 0 Amend. No. 123 12/18/99 iii  ; i": .' 2 , - i' DRR 00-0147 4/24/00 TAB-B2.0'SAFETY'LIMITS'(SLs)'

B2.1.1-1 ' .0 G . s  ; ' ;Amend:No.-123 ' 12/18/99 B 2.1.1-2 1 1 i'3-'-;' ' ' -.DRR03-0102 - 2/12/03.

B 2.1:1-3 :: JO ~ ' C;j.': . 14 ' 3 DRR-03-0102 2/12/03 B 2.1.1-4 14 ':.'. DRR 03-0102; 2/12/03 B 2.1.2-1 0 Amend. No. 123 12/18/99 B 2.1.2-92., .: 7 r '- DRR;02-1062* 9/26/02 4

B 2.1.2-3 Ii;';.i,,- tr .. ,.. - . ,

-m ,, .. .Amend. -°.,:-... No. 123 12/18/99 TAB - B 3.0 LIMITING CONDITIONFOR OPERATION (LCO) APPLICABILTY.

B 3.00-1 , ,:)i;! - ",' ll i;I:i' P'..

. uri::.'y -'q!, f ,A,,mend..No .123 ,1V/18/99 B 3.0-2, - ., Amend. No. 123 12/18/99 B 3.0-3 .. 0 .n Ame'nd. No6 123- 12118/99 B' 3'0'3 0 Amend. No. 1237, 12/18/99 B3.045 '~~~. . 4 ~* 'Ii!... A~c~623 12181899:

B 3.0-6" Ur im . . i 0nc. 1 I7'r Afnend.'No. 123 12/18/99 B 3.0-7. t* 2r ' ( ' 7t,;  ;'. 'Amend. No. 123 i 12/18/99' B3.0-.8K";- ' i I3- 0 -J0.1E .>.-E ,i 1i J' ,Amend.No. 123 o. 12/18199 B 3.0-9S! r; b S

.*_ .  ! ISi- '  :.Amend. No. 123 12/18/99 2

B3.0--10 * -:,r1 Ar;'~: ;j,: L. K DRR 02-1062 ,, Ir' 9/26/02.

B 3.0-?1-. y ',,;i e ^ )J n l ;DRR0205142 5/2/02 B 3 13r-2 .-- _e- *-. *.,j 'I ;1& t'_t2 1DRR 02-0514  :. 5/2/02 311

.b , 1'. .C. _

.0

. _ ..-__ ~..C.

• DRR020514 ' 5/2/02 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS -

B 3.1.1-1 2 , 0 - . Amend. No. 123 12/18/99 B 3..1~2~ ' e,~\lL/..'.Amend. No. 123' 12/18/99 B 3'f. 4:'" jIC'  ; t ' Arnend. N6.)123' 12/18/99 B31-4' 1A> ; 't- i' '133.1 \"4 DRR02-'1458 .: 12/03/02 B 3.1.1-5'1 J ; :. '1 " 0 *1 ^ Amend. No.A123; '- 12/18/99 B 3.1;2-1 ~*l.t¢^  :^,0

'4 ,:' *Z",-

->" '. t5f " 'Amend. No:'1237:' 12/18/99 B 3.1.2-2 -0 - Amend. No.123 12/18/99 B 3.1.2-3t * , 0 .- -* .: Amend. No. 123 ,' 12/18/99 B 3.1.2-4 0 - Amend. No. 123 :. 12/18/99 B 3.1.2-5  ; .0 . Amend. No. 123 12/18/99 B 3.1.3-1. ,. - 4b -  : ,Amend. ,Ar0p No. 123. 12/18/99 B 3.1.3  ; ^. . 0 ,,t,, fJ i. {,Amend. No., 123 12/18/99 B 3.1.3-3 0: - - Amiend. No.123 12/18/99 B 3.1.3-4 0- . Amend. No. 123 12/18/99' B 3.1.3-5 , 0 Amend. No. 123 12/18/99-B 3.1.3-6 0 Amend. No. 123 12/18/99 Wolf Creek - Unit I I. j -I ,.-Revision 16

LIST OF EFFECTIVE PAGES --TECHNICAL SPECIFICATION BASES -' '" Y-:

PAGE(1)'. REVISION NO;(2) CHANGE DOCUMENT(3) DATE EFFECTIVEI

,'IMPLEMENTED .- (4)

TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued).* -2;, ,--

B 3.1.4-1 ' 0 - Amend. No. 123 12/18/99 B 3.1.4-2 0 - Amend. No. 123 12/18/99' B 3.1.4-3 0. Amend. No. 123 12/18/99.'

B 3.1.4-4 :: 0 Amend. No. 123 ,12/18199 -'

B 3.1.4-5' .00 . Amend. No. 123 12/18/99 ' '

B 3.1.4-6' 0. ' . Amend. No. 123 12/18/99 B 3.1.4-7:. O..:... Amend. No. 123 12/18/99. i B 3.1.4:8 ;' 0; . ' Amend. No. 123 12/18/99 .

B 3.1.4-9 ,0 - Amend. No. 123 12/18/99 B 3.1.5-1 0 - Amend. No..123'.2Th:j' '. :.';.12118/99,'

B 3.1.5-2 ., .0 ,:-.  ; Amend. No. 123 12i18I99,'

B 3.1.5-3 .... 0. . - Amend. No. 123 12/18/99, B 3.1.5.4A' 0 Amend. No. 123 12/18/99, B 3.1.6-1 1, 0! j - Amend. No. 123 12/18/99 -2 B 3.1.6-2' . -' 0 - Amend. No. 123 12/18/99 -,(

B 3.1.6-3 , - ,e"i Amend. No. 123! 12/18/99'-

B 3.1.64 '0" Amend. No. 123 12/18/99 ;

B 3.1.6-5 ^ 0 .. ., . Amend. No. 123 12/18/99, B 3.1.6-6, ' 0 -. , Amend. No. 123 12/18/99 B 3.1.7-1 > .. 0  ;- '. Amend. No. 123 12118/99 B 3.1.7-2 -- 0 ., Amend. No. 123 12118/99 -

B 3.1.7-3: 0-- Amend. No. 123 12/18/99 ,-

B3.1.74 ' n; Amend. No. 123 12/18/99 B 3.1.7-5-.:; 0 .- e -,; Amend. No. 123 12/18/99, ;

B 3.1.7-6 ,:, . - 0: A. Amend. No. 123 12/18/99 B 3.1.8-1: '-0 Amend. No. 123 121/18/99 ,

B 3.1.8-2 ! ' 0 ' Amend. No. 123 1218/99 B3.1.8-3 . .15 . DRR03-0860' 7/10/03'.-

B 3.1.8-4'8.  ; ,15 ,, . DRR 03-0860 7/10103-B 3.1.8-5.;. .0 Amend. No. 123 12/18/99 B3.1.8-6 , . ,'-. ., DRR00-1427 10/1?/00 TAB - B 3.2 POWER DISTRIBUTION LIMITS _ _ t B 3.2.1-1 '.. 0, - Amend. No. 123 12/18/99 .-.

B 3.2.1-2' O' Amend. No. 123 12/1¶8/99 C.

B 3.2.1-3..' 0 Amend. No. 123 12/18/99 i.

B 3.2.145.T '.. Amend. No. 123 12/18/99 B3.2.1-5-7 0i DRR 99-1624 12/18/99 B3.2.1-6 0 DRR 02:1062 9/26/02 '.

B 3.2.1-7 . O Amend. No. 123 12/18/99; ?

B 3.2.1-8 0' Amend. No. 123 12/18/99 .

B 3.2.1-9 40 DRR 00-1365 9128100 B 3.2.2-1 0 Amend. No. 123 12/18/99 B 3.2.2 - 0 Amend. No. 123 12/1899 B 3.2.2-3. 0 Amend. No. 123 12/1i899' B 3.2.24 ' ° r' Amend. No. 123 12/18/99 Amend. No. 123 12/18/99 B 3.2.2-6 ' '0 ' Amend. No. 123 12/18/99 B 3.2.31 ' 0 Amend. No. 123 12/18/99 ;

B3.2.3-2 0,' Amend. No. 123 12/18/99 Wolf Creek - UnIt 1 ii Revision 16

LIST OF EFFECTIVE PAGES '.-,TECHNICAL SPECIFICATION'BASES4.n :-., .--.=- -;t.::^. - ,

- PAGE01 ) y,'),t " REtIISION NO,.,(r,'-% CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)

TAB - B 3.2 POWER DISTRIBUTION LIMITS (continued B 3.2.3-3 0 '-.1 Amend. No. 123 12/1 8/99

• Amend. No. 123 12/18/99 B 3.2.4-2 '0 Amend. No. 123 12118/99 -

B 3.2.4-3: 0 .- Amend. No. 123 12/18/99 B 3.2.44,-. - 0 , Amend. No. 123 12118199.

B 3.2A-5 ., , -- Amend. No. 123 12118/99 B 3.2.4-6 .: ,0 Amend. No. 123 12/18/99 12/18/99:

B 3.2.4-7  ; 0, . Amend. No. 123

• ?;?;'*.~., tt,.f -. a 12118/99:-

TAB'-'B 3.3INSTRUMENTATION .:*

B3.3.1-1'-J . 0o. Amend. No. 123 12118/99 B 3.3.1-2 IL ;0. Amend. No. 123 12118199-B 3.3.1-3' " 0 ' Amend. No. 123 Amend. No. 123 12118/99 12/18/99 B 3.3.1-S 0OU B 3.38.1 -5 :0 Amend. No. 123 12/18/99 B3.3:1-4 'r: ;.....:>;

Amend. No. 123 12118/99 B 3.3.1 -9.

B 3.3:1-17 i . i' ;A DRR No.1427 Amend. No. 123 12/18/99 B 3.3.1-9.0 Amend. No. 123 12/18/99 Amend. No. 123 12/18199 B 3.3;-13  :;8 *:.o~'; Amend. No. 123 12118/99 B 3.3.1-r B 3.3.1-123  ;'09 Amend. No. 123 12/18/99 B 3.3.1-23'0 Amend. No. 123 12/18/99 B 3.3.1-14: 0o Amend. No. 123 12/18/99 B 3.3.-18 4 c  ; Amend. No. 123 12/18/99 B 3.3.1-2IC Amend. No. 123 12118/99 B 3.3.1-30. - ' 0i Amend. No. 123 12118/99 B 3.3.1-318; . 0 , Amend. No. 123 12/18/99 B 3.3.139'2 '  ! Amend. No. 123 12/18/99 B 3.3.1-203 07 Amend. No. 123 12118/99 Amend. No. 123 12118/99 B ,33.1-722, ., 12 Amend. No. 123 12/18/99 B 3.3.1-23'_' 9 AeDRR 02No0123 B 3..42 -K.,L*gO ,>{j Amend. No. 123 12118/99 B 3.3.1-'25?', t o ,0' .w-; Amend. No. 123 12/18/99 Amend. No. 123 12/18/99 B .. 0;.-r Amend. No. 123 12/18/99 B 3.3::1-27' 3.3-14210 ~., -tl2.;;z 4/24/00 B 3.3.'1-28; DRR 00-0147 B3.3.'1-2168 " 12 Rl';

-1'no; 12118/99 B 3.3.1-29 :' DRR 99-1624 DRR 99-1624 12/18/99 Amend. No. 123 12/18/99 12/18199 Amend. No. 123 12/18/99-12118/99.

12/18199 Amend. No. 123 B 3.3.1-39: 0 Amend. No. 123 DRR 02-1 062 9/26/02' B 3.3.14C2 : ;0 --. DRR 02-1062 9/26/02 DRR 02-1062 9/26/02-

• 3.31 .12 DRR 02-,1062 9/26/02 B 3.3.1-30 . 0 Amend. No. 123 12/18/99 Amend. No. 123 12/18/99 Wolf Creek - Unit 1  ; iii :I, I--..-..Revision 16

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TAB - B 3.3 INSTRUMENTATION (continued)  : - . i . . : . -" .. - -w - .1 , .  : --.  :

B 3.3.1-41 '0 Amend. No. 123 12/18/99 B 3.3.1-42 13 DRR 02-1458 12/03/02.

B 3.3.1-43 13 DRR 02-1458 12/03/02 B 3.3.1-44 '13 DRR 02-1458 12/03/02 B 3.3.1-45 13 DRR 02-'1458 12/03/02 B 3.3.1-46. 13. DRR 02-1458 12/03/02 B3.3.1-47. ! 13 DRR 02-1458 12/03/02, B 3.3.1-48 i' -13; DRR 02-1458 12/03/02 B 3.3.1149- 13 DRR 02-1458 12/03/02 B 3.3.1-50 13 DRR 02-1458 12/03/02 B 3.3.1 13 DRR 02-1458 12/03/02 B 3.3.1-52 13' DRR 02-1458 12/03/02 B 3.3.1-53 13' DRR 02-1458 12103/02 B 3.3.1 .13 DRR 02-1458 12/03/02 B 3.3.1-55 13' DRR 02-1458 12/03/02 B3.3.1-56 13 :-- DRR 02-1458 12103/02 B 3.3.2-1' *0 - Amend. No. 123 12/18/99 B 3.3.2 0 ' Amend. No. 123 12/18/99 B 3.3.2-3' ' Amend. No. 123 12/18/99 B 3.3.2-4 0

• Amend. No. 123 121.18/99 B 3.3.2-5' 0 ' Amend. No. 123 12/18/99 B 3.3.2 7 DRR 01-0474 5/1/01:>'

B 3.3.2-7'- 0 - - Amend. No. 123 12/18/99 B 3.3.2-8 0 :f. Amend. No. 123 12/18/99 B 3.3.2-9 0'O% - Amend. No. 123 12/18/99.

B 3.3.2-10 Amend. No. 123 12/18/99 B 3.3.2-11. '0 -; Amend. No. 123 12/18/99 B 3.3.2-12  :' I Amend. No. 123 12/18/99 B 3.3.2-13'  :0 Amend. No. 123 12/18/991 B 3.3.2-14 2'--: DRR 00-0147 4/24/00 .

B 3.3.2-15.' - . 0' Amend. No. 123 12118/99 '

B 3.3.2-16'.. 0' Amend. N'o. 123 12l18/99 B 3.3.2-17. .0 OK Amend. No. 123 12/18/99 B 3.3'.2-18 0'; Amend. No. 123 12/18/99 B 3.3.2-19 0 Amend. No. 123 12/18/99 B 3.3.2-20 0 . ' Amend. No. 123 12/18/99' B3.3.2-21 'O! Amend. No. 123 12/18/99 B 3.3.2-22' 0 Amend. No. 123 12/18/99 B 3.3.2-23'4- 0 Amend. No. 123 12/18/99 B 3.3.2-24 !' O..' Amend. No. 123 12/18/99 I B 3.3.2-25' ' 0 Amend. No. 123 12/18/99 B 3.3.2 0  ;*. Amend. No. 123 12/18/99 B 3.3.2-27: 0i Amend. No. 123 12/18/99 K B 3.3.2-28 7 DRR 01-0474 5/1/01 B3.3.2-29 ,0' Amend. No. 123 12/18/99 B 3.3.2-30 '.' 0 Amend. No. 123 12/18/99 B 3.3.2-31 ' -. 0 Amend. No. 123 12/18/99 B 3.3.2-32 0 Amend. No. 123 12/18/99 B 3.3.2-33' 0 Amend. No. 123 12/18/99 B 3.3.2-34' 0 Amend. No. 123 12/18/99 B 3.3.2-35 0 Amend. No. 123 12/18/99 Wolf Creek - UnIt 1 iv Revision 16

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'IPAGE) -'! . . ;RE-VISION NO.:(2) IK, CHANGE DOCUMENT(3) DATE EFFECTIVE/

' .' 4IMPLEMENTED

. (4)

TAB - B 3.3 INSTRUMENTATION (continued) Amend. .. 12 . J. I.'I -

B 3.3.2-36  :..0, s Amend. No. 123 12/18/99 B 3.3.2-37 .io *-, Amend. No. 123 12/18/99.

B 3.3.2-38 7+i0 Amend. No. 123 12118/99 B 3.3.2-39 ' 0' Amend. No. 123 12/18/99 B 3.3.240.- .>;O Amend. No. 123 12/18/99 B 3.3.2-41'.' .12 DRR 02-1062 9/26/02 B 3.3.2-42 10- - Amend. No. 123 12/18/99 B3.3.2-43:., .12 rl DRR 02-.1 062 9/26/02 B 3.3.2-44.' ': Amend. No. 123 121.18199 B 3.3.2-45st r 10 > Amend. No. 123 12/18/99:

B3.3.2-46i 10 Amend. No. 123 12/18/99 B 3.3.2-47 r6 DRR 00-1541 3/13/01 B 3.3.2-4 8: i6? .! DRR 00-1541 3/13/01 B 3.3.2-49. Amend. No. 123 12118/99 B 3.3.2-50.~ 't2Q ,- DRR 00-0147 4/24/00 B 3.3.2-51 0 DRR 99-1624 12/18/99 B 3.3.2-5 2 0, - Amend. No. 123 12/18/99 B 3.3.2-53.1 .0 t; - Amend. No. 123 12/18199 B 3.3.2-54 r ! 6. DRR 00-1541 3/13101 B 3.1.2-55 1, .'. 6,>n..* DRR 00-1541 3/13/01 B 3.3.3-1" .0, Amend. No. 123 12118199 B3.3.3-25 , DRR 00-1427 1i0i12100 B 3.3.3-3 ;.0 -. -' Amend. No. 123 12118/99 B 3.3.3-4.o!'.  :< O Amend. No. 123 12118/99 B 3.3.3-5 j2 8t .' Amend. No. 123 12/18/99 8

B 3.3.3-613- '.'-- r DRR 01;1235 9/1j9101 B 3.3.3-74$ .*>- '8. DRR 01-1235 9/19/01 -

B 3.;3-:38fr;. .8 !." DRR 01-1235 9/19/01 B 3.3.3-2*9, DRR 01--1235 9/19/01.

B 3.3.3-6 .8? 9; D DRR 01-1235 9119101 B 3.3.3-1 1'; DRR 01-1235 9/1 9/01 B 3.3.3-3d 2" I- :8. DRR 01-1235 9/19/01:,.

B3.3.3-61 3.C. O " DRR 01-1235 9/19/01.

B 3.3&3-14!S, r8 .jlt{ DRR 01-1235 9/19101 B 3.133-15(:  ;-.S.§ i4 DRR 01-1 235 9/19/01 B 3.3.461 , Amend. No. 123 12118/99 B 3.3.64-22 9 0* DRR 02-1023 2128/02 B 3.3.'-3 ; 1 DRR 03-0860 7/10/03 B 3.3.4-4x1,,;,:, DRR 99-1624 12118/99 B 3.3.4-'5, t;. <1..r DRR 99-1624 12118/99 B 3.3.4-6,\.; .9- DRR 02-0123 2/28/02 B 3.3.5-1 , ' 0! :1 Amend. No. 123 12/18/99 B 3.3.5-2 '.  : ,1 . -- DRR 99-1624 12/18199 B 3.3.5-3 :;:. ,- ' DRR 99-1624 12/18/99 B 3.3.5-4 , G '.( .il DRR 99-:;1624 12/18199 B 3.3.5-5, k'. I! 0  : - Amend. No. 123 12118199 B 3.3.5-6 ':. ' 0 ), Amend. No. 123 12/18/992 B 3.3.5-7 'xs 0-,: . Amend. No. 123 12118/99 B 3.3.6-1 ~-. 0 ,x - Amend. No. 123 12118/99 B 3.3.6-2 .0 Amend. No. 123 12118/99 Wolf Creek - Unit 1 .1 v - , Revisbnl16

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TAB - B 3.3 INSTRUMENTATIONI (continued)z -  : -

B 3.3.6-3 0 Amend. No. 123 12/18/99 B 3.3.6-430 Amend. No. 123 12/18/99 B 3.3.6-5 0' Amend. No. 123 12/18/99 B 3.3.6-6 0 Amend. No. 123 12/18/99

B 3.3.6-7. '- 0 Amend. No. 123 12/18/399':

B 3.3.7-1 0 Amend. No. 123 12/18/99 B 3.3.7-26. 0: -t Amend. No. 123 12/18/99 B 3.3.7'3i 0- Amend. No. 123 12/18/99.

B 3.3.7-4 L 0 *-' Amend. No. 123 12/18/99 B 3.3.7-5 -' 0 " - Amend. No. 123 12/18/99 B 3.3.7-6 0- .`-- Amend. No. 123 12/18/99 .-

B3.3.7-7 - 0"t "' Amend. No. 123 12/18(99 D~

B 3.3.7 0 Amend. No. 123 12/18/99.

B 3.3.8-1' ' oi '-'4* Amend. No. 123 12/18/99

B 3.3.8-2 0O!. '- Amend. No. 123 12/18(99 B 3.3.8-3 0' Amend. No. 123 12/18/99 B3.3.8-4 0 Amend. No. 123 12/18199 B 3.3.8 5 0' C Amend. No. 123 12/18/99(i B 3.3.8-6' 0; Amend. No. 123 12/18/99

• B 43.i B 3.3.8-7 - 0 *A*;e-ndt. Amend. No.

No. 12 123 2/9 12/18/99 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) .. ! .  :

B3.4.1-1 Amend. No. 123 12/18/99 B 3.4.1-2 ; 10 DRR 02-'0411 415102 '

B 3.4'.1-3 10 DRR 02-0411 4/5/02 '

• B 3.4.1*4 0 Amend. 196. 123 12/18/99 B3.4.1-5 0 - Amend. N6. 123 12/18/99 B 3.4.1-6 ' 0 ,' Amend. No. 123 12i18/99 B 3.4.2-1 '* 0 4 Amend. No. 123 12118/99 B 3.4.2-2 0 Amend. No. 123 12/18/99 B 3.4.2-3 '- 0; Amend. No. 123 12I16/99 B 3.4 3' 1 'O-!<! Amend. No. 123 12/-8/99 B3.4.3-2' O: Amend. No. 123 12/18/99 B 3.4.3-3 0 Amend. Ro,. 123 1218/99 -

B 3.4.3-4 ' 0 Amend. No. 123 12J18/9 9 B 3.4.3-5*: I ' Amend. No. 123 12/1i8b9 B 3.4.3-6 0 O* Amend. No. 123 12/18/99 B 3.4.3-7 0 Amend. No. 123 12/18/99 B 3.4.4-1 0 Amend. No. 123 12/18/99 B 3.4.4-2 0 i " Amend. No. 123 12/18/99 B 3.4.4-3 0 Amend. No. 123 12/18/99 B 3.4.5-1 0 Amend. No. 123 12/18/99 B3.4.5-2 0 Amend. No. 123 12/18/99 ***

B 3.4.5-3 12"' DRR 02-1062 9/26r/02 B 3.4.5-4 0 Amend. No. 123 12118/99 .

B 3.4.5-5' 12 DRR 02-1062 9/26!02,:

B 3.4.5-6 12' DRR 02-1062 9/26W'02' B 3.4.6-1 0 Amend. No. 123 12t18/96 B3.4.6-2' 12 DRR 02-1062 9/26/02 Wolf Creek - Unit I vi Revision 16

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (co ntinued) - 0 ,I.'

B 3.4.6-3 ., ' 12 DRR 02-1062 9/26/02 B3.4.6-4 U 12 DRR 02-,1062 9/26/02 -

B 3.4.6-5. ;12 DRR 02-1 062 9/26/02 B 3.4.7 12 DRR 02-1062 9/26/02 B 3.4.7-2 ,,.. 12; DRR 02-1 062 9/26/02 B 3.4.7-3 :0 Amend. No. 123 12/18/99 B 3.4.7-4 : . 12 ., DRR 02-1062 9/26/02 B 3.4.7-5 .: 12 DRR 02-1062 9/26/02 -

B 3.4.8-1'Žai '0 - Amend. No. 123 12118199-B3.4.8 .- ,,13 DRR 02-.1458 12/03/02:s B 3.4.8-3 j ;;2, , DRR 02-1062 9/26/02 -

B 3.4.8-4'~. P 12 ,:;""¶'t. DRR 02-1062 9/26/02 B3.4.9-3i:&r Amend. No. 123 12/1BI99 B 3.4.9-2 t -. 0 r Amend. No. 123 12/18199,,

B3.4.9-3-;3 -> c* ,0 , Amend. No. 123 12/18/99..

B3.4.9-4V& - 0 Amend. No. 123 12/18199 B 3.4 O , ¶, -. - , 5 .0.;s , DRR 00-1427; 10/12/00:

B 3.4.1D-2 z- 5 DRR 00-1427 - 10/12/00 B 3.4 .10-3 Z . 0  ! ,! 1A Amend. No. 123 12/18/99, B 3.41 004x , *5 ,,,, DRR 00-1427 10/12100 B 3.4.11-1 0 Amend. No. 123 12/18199 ¶

-.DRR 99-1624.. '12J18/99

-B 3.4.11'-3 '--- DRR 99-1624 12118/99 B 3.4.1 .  ; *I , Amend. No. 123 12/18/99 B 3.4.11-4k 5 5,. - DRR 99-1624 12/18/99.

B 3.4,11-6;, - 0 , Amend. No. 123 12/18/99 B3.4.1.1-,7,, .m..- Amend. No. 123- 12/1i8/99 B 3.4.12-12> '1 r DRR 99-1624 12/18/99 ,

B 3.4.12-23c . - . DRR 99-1624 12/18/99.

B 3.4.12-,3, .O Amend. No. 123.

DRR 99-1624 12/18/99 DRR 99;1624 12/18/99.

B 3.4.412-6;: .; ,.'. DRR 99-1624 121218/99 B 3.4.12-75. ,0 Amend. No. 123 12/18/99, DRR 99:1624 12/18/99 B3.4.12-6: 0

° A Amend. No. 123. 12/18/99 B 3.4.12-12 0 ..

Amend. No. 123 12/18/99 B3.4.12-11,  ; 01 . Amend. No. 123 12/18/99 B 3.4.12-1 0 Amend. No. 123 12/18/99 B 3.4.12-131 ! 0-' Amend. No. 123 12/18/99 B3.4.14-2z: -r 0- Amend. No. 123 12118/99 B 3.4.13-2.i 3.4.13;-l'4 0 v, . Amend. No. 123 12/18/99 Amend. No. 123 12/18/99 Amend. No. 123 Amend. No. 123 12/18199 B3.4.13-3,' 0.°::-".

DRR 02-1062 B 3.4.13-1 . 0 9/26/02 .

Amend. No. 123 12/18/99 -

B 3.4.14-21 0 Amend. No. 123 12/1i8/99 Amend. No. 123 12/18/99 B 3.4.14-3 0 Amend. No. 123 12/18/99 B 3.4.14-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 vii .t '.a .. .. - '.'Revision 16

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TAB - B 3.4 REACTQR COOLANT SYSTEM (RCS) .(continued) -  :,.-

B3.4.14-5 16 DRR03-1497 11/4/03 B 3.4.14-6 16 DRR 03-1497 11/4/03 B3.415-1 *2 A - .. DRR00-0147 4/24/00 B 3.4.15-2 0 Amend. No. 123 12118/00 B 3.4.15-3 - 9 ' DRR 02-0123 2128102' &

B 3.4.15-4: 9  ! DRR 02-1023 2/28/02 B 3.4.15-5: -9 ' DRR 02-1023 2/28/02 '

B 3.4.15-6, 0! Amend. No. 123 12/18/99 B 3.4.15-7' " ' 00) Amend. No.123 12/18/99 '

B 3.4:16'1 0 Amend. No. 123 12/18/99 2 B 3.4.-i6-2' DRR 99-1624 1211'8/99, B 3.4.16-3 0 '

• Amend. No. 123 12/18l99 .

B 3.4.16:4 t t.' 0~ Amend. No. 123 12/18/99 .

B3.4.16-5;t 0 Amend. No. 123 12118/99 .t B3.4.16.:6,' 0' Amend. No. 123 12/16/99 '3 TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) -' -

B 3.5.1,-1 0o - Amend. No. 123 12/18/99 - '

B 3.5.1-2 0'  :' ' Amend. Nb. 123 12/1'B99 B 3.5.1-3 0' Amend. No. 123 12/18/99' B 3.5.1-4

• O Amend. No. 123 128/999-'

B 3.5.1-5' DRR 99-1'624' 12/18/99 B 3.5.1-6 DRR 99-i624 B 3.5.1-7 .16. t DRR03'1497 B 3.5.1-8 1 DRR 99:1624 12/18/99 -'

B 3.5.2-1 Amend. No. 123 12i18/99 ' l B 3.5.2-2 -0 Amend. No. 123 12/18/99 -

1211899;'-

12118i99';'

B 3.5.2-3j 0' Amend. No.123 B 3.5.2-4' 0 i' Amend. No.123 12/18/99 ' '

B3.5.2-6 0 ,Amend. No. 123 B 3.5.2-6 0'" - Amend. No. 123 12118i99 -

B 3.5.2-7 r0 Amend. No.123 B 3.5.2-8 01 Amend. No.123 12/18/99 B 3.5.2-9 12~ . DRR 02!1062 9/26/02 ^'

B 3.5.2-10 0' Amend. No. 123 12/18/99 B 3.5.3-1 16' DRR 03-1497 11/4/O3 -

B 3.5.3-2 16 DRR 03-1497 11/4/03' :

B 3.5.3-3 o ' Amend. No. 123 12/18/99'Y B 3.5.3-4 16 DRR 03.1497 11/4/03 B 3.5.4-1 , 0, > Amend. No.123 12/18/99 --

B 3.5.4-2 '0 Amend. No. 123 12/18/99 B 3.5.4-3 0 Amend. No. 123 12/18/99 B 3.5.4-4 0' -' Amend. No. 123 12/18/99 B 3.5.4-5 0' Amend. No. 123 12/18/99 '

B 3.5.4-6 '0 Amend. No. 123 12/1'8199

B 3.5.5-1 2 Amend. No. 132 4/24/00' '

B 3.5.5-2 2 Amend. No. 132 4/24/00 1 ,

B 3.5.5-3 2 Amend. No.132 4/24/00 B 3.5.5-4 2 '- Amend. No. 132 4/24/00

.. . 1 Wolf Creek - UnIt 1 viii Revision 16

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TAB - B 3.6 CONTAINMENT SYSTEMS,.-' ' a . ' . r:: v .

• i -

B3.6.1-1 0, - Amend. No. 123 1Z/16_IY - -

B3.6.1-2', 0- Amend. No. 123 12118/99 B3.6.1-3'. 0 Amend. N o.123 12118/99 B 3.6.1-4 ' ' 8 ', DRR 01-1235 9/19/01 B 3.6.2 .0, Amend. No. 123 12/18/99 B3.6.2-2': 0 Amend. No. 123 12/18/99 B 3.62-3. ,;.. 0 At Amend. No. 123 12/18/99 B 3.6.24- ,~,

O"t;,, Amend. No. 123 12/1 8/99 B 3.6.2-5\j 0 Amend. No. 123 121.18/99 B 3.6.2-6  :,, 0 , Amend. No. 123 12/18/99 .

B 3.6.2-7;,l 0 ,,:z . Amend. No. 123 12/18/99, B 3.6.3-1 *. 0 ,0 sr-0 Amend. No. 123 12118/99 B3.6.3-2'tot 0. Amend. No. 123 12118/99 B 3.6.3-,3 .: ., (0 s Amend. No. 123 12/18/99.

B 3.6.3-4 !c 0. r,r _o Amend. No. 123 12/18/99 B 3.6.3-5 8 DRR 01-1235 9/19/01,

,-.-.  :'DRR01-1235;-'-,'- ;J:919/01 DRR 0171235 9/19/01 8

B3.6.3-8. ci j :.' DRR 01-,1235 9/19/01, B 3.6.3-9 ;r, 8 t DRR 01-1235 911910i~

DRR 01-1235 9/19/01.

DRR 02-0123 2/28/02 B3.6.3-.12, 9 -i DRR 02-0123 2/28/02' B 3.6.3-134.c- DRR 02-0123 2/28/02 DRR 02-0123 2128/02' B 3.6.4,:1 ;...- 29 , DRR 00-,0147 4/24/00 B 3.6.4-2 3- r . '14 Amend. No. 123 12/18/99 B 3.6A7-,jvI 0l,',,....

B 3.6.4-3, t. . 0 Amend. No. 123 12/18/99 Amend. No. 123 12/1,8/99 B 3.6.5 ;0 ° . .' Amend. No. 123 1211.8/99 B 3.6.,5,-3 1;;13 4i DRR 02-1458 12/03/02 B3.6.53;, 13 Amend. No. 123 12/18199 B3.6.s6-1F',.,1o' *- Amend. No. 123 12/18/99 B3.6:6.2c-o i 0 Amend. No. 123 12/18/99 B 3.6.6-3 .. 0 DRR 99-1624 121,18199 B 3.6,..6,4:,,; .° ii Amend. No. 123 12118/99 B 3.6.6-5 Amend. No. 123 12/18/99 B 3.6.6.6 >. 0 iea1.i Amend. No. 123 12/18/99.

B 3.6.6-7,j; a, , Amend. No. 123 12/18/99 B3.6.6-8:,:..' . 14 ... DRR 03-0102 2/12/03 B 3.6.6-9 13 . DRR 02-1458 12/03/02 B3.6..-1 1 .. 0,: Amend. No. 123 12/18/99 B 3.6.7-2:,.;& 0!' (.,.i Amend. No. 123 12/18/99 B 3.6.7-3. t&

• 0' Amend. No. 123 12118/99 B 3.6174 f I, 2'.' .. ' DRR 00-0147 4/24/00 B 3.6.7-5 .  ;- 0 , Amend. No. 123 12/18/99 B3.6.8-11 _.' 0 Amend. No. 123 12/18/99i Amend. No. 123 12/18/99 B 3.6.8-3.. 0 . Amend. No. 123 12/18/99 B 3.6.8-4 0 Amend. No. 123 12/18/99 B 3.6.8-5 '0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 B. ix .. 1: ,Revision 16

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TAB - B 3.7 PLANT SYSTE_MS . I , -;

B 3.7.1-1 Amend. No. 123 12118/99 B 3.7.1-2 0

Amend. No. 123 12/18/99.

B 3.7.1-3 *00 .

Amend. No. 123 12/18/99 >

O z B 3.7.1-4 Amend. No. 123 12/18/99' 8 B 3.7.1-5 0 Amend. No. 123 12/18/99 B 3.7.1-6' o0' .. "I Amend. No. 123 12/18/99 "

B 3.7.2-1 Amend. No. 123 12/18/99 i '

B 3.7.2-Z2' 0' Amend. No. 123 12/18/99 B 3.7:2-3 , O. . Amend. No. 123 12118/99.

B 3.7-24' Amend. No. 123 12118/99:

B 3.7.2-5 0. Amend. No. 123 12118/99,.,

B 3.7.2-6 0 ;v.'. . Amend. No. 123 12118/99 6 B 3.7.3-1 .1 O .'K' -

Amend. No. 123 12118/99 a B 3.7.3-2 Amend. No. 123 12118/99 B 3.7.3-3 *.1 o .. Amend. No. 123 12/18/99 B 3.7.34 i*>o z-'. Amend. No. 123 12/18/99 £ B 3.7.3-5': Amend. No. 123 12/18/99 :.

B 3.7.4-1 DRR 99-1624 12/18/99 ..

B 3.7.4-2 DRR 99-1624 12118/99 "

B 3.7.4-3 DRR 99-1624 B 3.7.44  ! 16' -iw DRR 99-1624 12118/99.'

B 3.7.4-5 DRR 99'1624 12/18/99 :

B 3.7.5-1 0: Amend. No. 123 12118/99 '

' 1:.

0 ....

B 3.7.5-2 Amend. No. 123 12118/99 -

B 3.7.5-3 Amend. No. 123 12/18/99.':

B 3.7.5-4 DRR 03-1497 11/4/03 B 3.7.5-5' B 3.7.5-6 DRR 03-1497 Amend. No. 123 11/4103;.

12118/99 I

B 3.7.5-7 (' 0i' DRR 03-1497 11/4/03:

B 3.7.5-8 DRR 03-0102 2/12/03 B 3.7.5-9 016- :. 'RR02-1458' -- 2703j02<- .

B 3.7.6-1 , 'Amnrid;No. 123 1211'8199,'.7 B 3.7.6-2 Amend. No. 123 B 3.7.6-3 16 ,..\ Amend. No. 123 B 3.7.7-1 14 Amend. No. 123 12/18/99 B 3.7.7-2 Amend. No. 123 12/18/99

.. u , ..

B 3.7.7-3 Amend. No. 123 12118/99 -t B 3.7.7-4 , '0' DRR 99-1624 12/18/99 -

B 3.7.8:1 Amend. No. 123 12/18/99 B 3.7.8-2 Amend. No. 123 12118/99 B 3.7.8-3 O .3 Amend. No. 123 3 v%

B 3.7.8-4 0.;O.. Amend. No. 123 12/18/99 B 3.7.8-5 Amend. No. 123 12/18/99

15. *.

B 3.7.9-1 Amend. No. 134 7/i4/00:'

B 3.7.9-2 *0O > Amend. No. 134 7/14/00- -.

B 3.7.9-3 0 Amend. No. 134 7/14/00 .-

B 3.7.94. Amend. No. 134 7/14/00 B 3.7.10 Amend. No. 123 12i18/99

B 3.7.10-2: DRR 03-0860 7/10/03..

B 3.7.10 Amend. No. 123 12/18/99 B 3.7.10-4 Amend. No. 123 12118/99 Wolf Creek - UnIt 1 x Revisbon 16

3 I - I II LIST OF EFFECTIVE PAGES'-:TECHNICAL'SPECIFICATION BASES '  :

'--PAGE-T 3 REVISION NO.:(2 ?).y.- CHANGE-DOCUMENT (3) DATE EFFECTIVEI

'( :-, . ,'i;.g  ; . -IMPLEMENTED (4)

TAB - B 3.7 PLANT SYSTEMS (continued), - -; ..

B 3.7.10-5 r. 0 .'. ,*- Amend. No. 123 12/18/99 B 3.7.10-6  :'~ 0 *'...,; l Amend. No. 123 12/18/99:

B 3.7.10-7,-  ! ' Amend. No. 123 12118/99 B 3.7.11-t"k si 01 ' r- Amend. No. 123 12/18/99..

B3.7.11-2.  : 0 . ,-: Amend. No. 123 12/18/99 B3.7:11 "; 0' .i ., Amend. No. 123 12118/99 B 3.7.11-4:2i; 0o. ;i

, Amend. No. 123 12/18/99 B 3.7.12-1, t. 0 o ., Amend. No. 123 12/18/99.

B 3.7.13-i:i *: ,rl, ,. DRR 99-1624 12/18/99 B 3.7.A13-2&3 I .- DRR 99-1624 12/18/99' B 3.7.13-3,> E, .0 i ,c : DRR 99-1624 12/18/99!

B3.7A3-4" '-2 I r;-n DRR99-1624 12/18/99 B 3.7.13-5,. , - ", r < DRR 99-1624 12/18/99 B 3.7.13-6"1 1224r; u DRR 02-1062 9/26/02 B 3.7.13-7..:r  :;1 .. . . DRR 99-1624 12/18/99 B 3.7..13-8^t i .R 'i - DRR 99-1624 12/18/99 B 3.7A.4-1,-I'l 8 0' '*- Amend. No. 123 12/18/99 B 3.7.15-1S! I, O !1 Amend. No. 123 12/18/99 B 3.7.15-2 St , e0 2 Amend. No. 123 12/18/99, B 3.7.15-3 St ni- Amend. No. 123 12/18/99 B 3.7A1.: r ,:' 5 TL DRR 00-1427 10/12/00 B 3.7.16-2Sd  ;. n .- 1J DRR 99-,1624 12/18/99 B 3.7:16-3' , .5 .b, ,T.A DRR 0041427 10/12/00 -

B 3.7.177-21 q .' . .:r

• .A DRR 01-0474 5/1/01 I

B 3.7.17-2 ' ' .'7A. !aA DRR 01'0474 5/11/01 B 3.7.17-3 ! ' 5^'. DRR001427 10/12/00.

B 3.7.18-1' ' 'O-':-i Amend. No. 123 12/18/99 B 3.7.18-2^; f'. .b .So Amend. No. 123 12/18199 B 3.7A18-3' , ' Amend. No. 123 12/18/99 TAB'-' B 3.8 ELECTRICAL POWER-SYSTEMS  ; . -.

B 3.8.11'"- ' .' 0 -'V' , Amend. No. 123 12/18/99 B 3.63A1-3'2 -_

Amend. No. 123 12/18/99 DRR 001541 3/13/01 B 3.8.'1-3;i 61 o DRR 00:1 541 3/13/01 B3.8.16 'J Amend. No. 123 12/18/99 Amend. No. 123 12/18/99 B 3.8.1;i '- .0 t, Amend. No. 123 12/18/99 B 3.8. 11-lo B 3.8.1-8 '..: 0 Amend. No. 123 12/18/99 B3.8.*9'3 0 Amend. No. 123 12/18/99 Amend. No. 123 12118/99 B 3.8.1-814 Amend. No. 123 12/18/99 B 'i0-;-'- ;ov i-- '"

B 3.8.1 3.8.1-1'3' B0  : 0 Amend. No. 123 12118/99 B 3.8'. i-114 '- 0;' 12118199 Amend. No. 123 Amend. No. 123 12/18/99 Amend. No. 123 12118/99 B 3.8.'i-i3 ' t- : - ';

DRR 02-0123 2/28102 B 3.8.1-18 ' DRR 01-0474 511/01 -

Amend. No. 123 12/18/99 B 3.8.1-19,' 0 Amend. No. 123 12/ 8199 B 3.8.1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 .. xi . , , -, - -,,Revision 16

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

IMPLEMENTED (4)

TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued). - '  : ~.. 1.

B 3.8.1-21 0 Amend. No. 123 12/18/99.

B 3.8.1-22 8 DRR 01-1235 9/19/01' B 3.8.1-23 7 DRR 01-0474 5/1/01 £'

B 3.8.1-24 0 Amend. No. 123 12118/99 '

B 3.8.1-25 0 Amend. No. 123 12/18/99 -

B 3.8.1 0 Amend. No. 123 12118/99 B 3.8.1-27 6' - DRR 00-1541 3131301 ' .?

B 3.8.2-1' !. Amend. No. 123 12/18199' B3.82-2 0' Amend. No. 123 12/18/99 B 3.8:2-4 'O Amend. No. 123 12/18/99 B 3.8'2-4 3. Amend. No. 123 12/18/99 B 3.8.2-5 12 '$ ' DRR 02-1062 9/261029 1 B3.8.2-61 12' DRR 02.1062 9/26/02'

B 3.8.2i-7' 12; DRR 02-1062 9/26/02,--

B 3.8.3-1' 1:' DRR 99-1624 12/18/99 B 3.8.3-2 0 Amend. No. 123 12/18/99 B 3.8.3-3 .. 0 .Amend..No. .123 B 3.8.34 . .1 DRR 99-41624 .12118199 .

B3.8.3-5  ;' 0 h Amend. No. 123 B 3.8.3-6 0.' Amend. No. 123 12/18/99 -

B 3.8.3-7 '12 DRR 02-1062 9/26/02-..:

B 3.8.3-8 .: '.1. . DRR 99-1624 12/18/99.:'

B 3.8.3-9 - J 0. Amend. No. 123 12/18/99 .

B 3.8.4-1'. ... ' 0 Amend. No. 123 12118/99 B3.8.4-2'" 0 Amend. No. 123 12/18/99 B3.8.4-3 . 0t Amend. No. 123 12/18/99 B 3.8.4-45* ' .. ' 0 ' Amend. No. 123 12/18/99:

B 3.8.4-5 ' .0,,,.,S,. Amend. Nd. 123 12/18/99 :

B 3.8.4-6'  : 0 Amend. No. 123 12/18/99, E B 3.8.4-7" 6 DRR 00-1541 3/13I01- '

B 3.8.4-8 ' 0 Amend. No. 123; 12/18/99!..;

B 3.8.4-9 " z" 2 : DRR 00:0147 - 4/24/00C5.,,'-

B 3.8.521 ' 0. Amend. No. 123 12/18/99!:

B3.8.5-2' '. Amend. No. 123 12/18/99, i B 3.8.5-3 . 0 , Amend. No. 123 12/18/-99-,. f B 3.8.5-4 *k' ' '2, ' . DRR 02^1062 9/26/02 :.- '-.

B 3.8.5-5 12 DRR 02-1062 9/26/02. J B 3.8.6-1 -.. 0 Amend. No. 123 12/18/99.

B 3.8.6-2 $.* 0 Amend. No. 123 12118/99 B 3.8.6-3 ' 0  : Amend. No. 123 12/18/99, B 3.8.6-1 . 0 Amend. No. 123 12/18/99.

B 3.8.6-52 '. 0 Amend. No. 123 12/18/99.

B 3.8.6-6 0 Amend. No. 123 12/18/99.

B 3.8.7-1l.' ' 0 - ' Amend. No. 123 12/18/99 B 3.8.7-2 5 DRR 00-1427 10/12/00.

B 3.8.7-30 Amend. No. 123 12/18/99 B 3.8.i-4 ~ '0 '- ' ' Amend. No. 123 r 2/118/99 .

B 3.8.8-1 0 Amend. No. 123 12/18/99 B 3.8.8-2 0 Amend. No. 123 12/18/99 B 3.8.8-3 0 Amend. No. 123 12/18/99 B 3.8.8-4 12 DRR 02-1062 9/26/02 Wolf Creek - Unit 1 *xii .. ? I Revision 16

C ,A.; [ 1i, I it.;.7 LIST OF EFFECTIVE PAGES ---TEGHNICAL SPECIFICATION BASES - *.

-,PAGE4 1~-.j,~ -; :REVISIONNO._,2)s,4, CHANGE DOCU.MENT(3) DATEEFFECTIVE/

--- ,,,;, . IMPLEMENTED (4)

TAB - B 3:8 ELECTRICAL POWER SYSTEMS (continued) t ~- 7 . -

B 3.8.8-5 , 7 12 ,, - '. -', DRR021062 9/26102.

B 3.8.9-1, .0f Amend. No. 123 12/18/99 B 3.8.9-2 . . 0 .. ..  ;, Amend. No. 123 12/18/99,.

B 3.8.9-3  ;.* 01 Amend. No.123 . 12/18/99 B 3.8.9-4-d 0.., . Amend. No. 123 12/18/99 B 3.8.9-5'f  ; 0° . Amend. No. 123 12/18/99 B 3.8.9-6 ,Q 0 -- Amend. No.123 12118/99 B3.8.9-.7 ' 0 ,.- ' - Amend. No. 123 12/18.199 B 3.8.9-8 -1 - , ,- DRR99-1624 - 12/18199 B3.8.9-9 -0 . . Amend. No. 123 121198/99 B 3.8.10-,1. e, ., -, . Amend. No. 123 1218/99 B 3.8.10-2 .,0 r Amend. No. 123 12118/99 B 3.8.10;,3w~ . £<.;-- .- - 'Amend. No. 123 - 12/18/99 B3.8.10-4n,i. 0 Si. Amend. No. 123 12/18/99 B3.8.1P,~5& 12 DRR02-1062 9/26/02 B 3.8.&10-6,g ,- 12,;,r-r. - DRR 02-1062 9/26102 TAB -.B 3.9 REFUELING OPERATIONS. - *.

B3.9.1-1';I s .. 0, t--. o Amend. No. 123 12/18/99 B 3.9;1-2 .r 13M. .133 c. DRR 02-1458 12103/02 B 3.9.1-3 -'Z D.r,13 :,; ._ DRR 02-1458 12/03/02 B 3.9.1-4':' t`Q( 'r! Amend. No. 123 12118/99 8 3.9.2-1.0 ' Amend. No. 123 12/18/99 B 3.9.2-2 : . 0 V Amend. No. 123 12/18/99 B 3.9.2-3'%' .> 01.I::.' Amend. No. 123 12/18/99 B 3.9.3-1'1 2. DRR02-1062 9/26/02 B3.9.3-2V.. .i2, .iJ A - DRR 02:1062 9/26/02 B 3.93-3","-.  :. 12 7r;'#W DRR 02-1062 9/26/02 B3.9.4-1;i 8 14'.- i.' DRR03-,0102 2112103 B3.9:4-2 r', t:-13:;, i] DRR 02:1458 12/03/02 B3.9.4-3';' ' 13 wn. DRR 02-1458 12103/02 B3.9:44 *'; :13.0'-i; DRR 02:1458 12103/02 B 3.9.'4-5": 4' .13 . DRR 02-1458 12/03/02 B 3.9'4-:K.iA ' .13 DRR 02-1458 12/03/02 B 3.9.54i'2 1!. , ,o . , Amend. No. 123 12/18/99 B 3.9.5-2 P?. 12-2 1'0 l DRR 02-1062 9/26/02 B 3.9:5.3' -,IC12. DRR 02-1062 9/26/02 B 3.9.5-41' '  :.? 12; !x;c;;A DRR 02-1062 9/26/02 B 3.9.6-1! . .0~ ; ;, . Amend. No. 123 12118/99 B 3.9.6-2 " . .13' .t- _1-iA DRR 02-1458 12/03/02 B 3.9.6-3 c F . 121 , . DRR 02-1062 9/26/02 B 3.9.64' .1 i DRR 02-J062 9/26/02 B 3.9.7-1,  : 0 '; .^ Amend. No. 123 12/18/99 B 3.9.7-2  ; £. .0 3mj: Amend. No. 123 12/18/99 B3.9.7-3 C ' 0 .- i Amend. No. 123 12/18/99

~ .. ..,;.

Wolf Creek - Unit 1  ;/X xiii .... - .Revision 16

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

IMPLEMENTED (4)

Note I 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 Revision number 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 affected each page. The NRC has indicated that Bases changes will not be issued with License Amendments. Therefore, the change document should be a DRR number in accordance with AP 26A-002.

Note 4 The date effective or implemented is the date the Bases pages are issued by Document Control.

Wolf Creek - UnIt 1 xiv Revision 16