Text

Forwards Interim Comments on Final Draft Tech Specs.Review Will Be Completed by 841210
	ML20100F105
Person / Time
Site:	Wolf Creek
Issue date:	12/03/1984
From:	Koester G; KANSAS GAS & ELECTRIC CO.
To:	Harold Denton; Office of Nuclear Reactor Regulation
References
	KMLNRC-84-211, NUDOCS 8412060459
	Download: ML20100F105 (104)
	v • d • e

{{#Wiki_filter:" 1 KANSAS GAS AND ELECTRIC COMPANY THE ELECTFhC COMPANY CLENN L. ROESTER VCs peessosst. Muctans ? December 3, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission r Washington, D.C. 20555 KMLNRC 84-211 Re: Doc).et No. STN 50-482 Ref Letter dated 11/07/84 from DGEisenhut, NRC, to GLKoester, KG&E Subj: Technical Specifications

Dear Mr. Denton:

The Reference provided the " final draft" of the Wolf Creek Technical Specifications for KG&E review and affirmation. x KG&E's review of the Technical Specifications is not yet complete. However, to aid the NRC provided herewith are interim comments on the draft. KG&E's review will be complete with the results for-warded to the NRC on or before December 10, 1984. Yours very truly, GLK:bb Attach j xc:PO'Connor (2) HBundy QO k0 e 0412060459 041203 PDR ADOCK 05000402 g PDR 201 N. Market - WIchts, Kansas ~ Med Address: PO. Box 208 i WcNts, Kansas 67201 - Telephone: Area Code (316) 2616451 L i

KANSAS GAS AND ELECTRIC COMPANY e F-G L C N M U R O E S T E le wuts passeoem? nuctaan 5 A December 3, 1984 x Mr. Harold R.' Centon, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 KMLNRC 84-211 Re Docket No. STN 50-482' Ref:. Letter dated 11/07/84 from DGEisenhut, NRC, .to GLKoester, KG&E Subjs Technical Specifications

Dear Mr.-Denton:

The Reference provided the " final draf t" of the Wolf Creek Technical Specifications for KG&E review and affirmation. KG&E's review of the Technical Specifications is not yet complete. However, to aid the NRC provided herewith are interim comments on the. draft. KG&E's review will be complete with the results for-warded to the likC on or before December ~10, 1984. Yours:very truly, GIE:bb Attach xc:PO'Connor (2) IDlundy; s. f 201 N. Market -WIchte, Kansas - Med Address: PO. Ikn 208 i %4chsta, Kansas 67201 - Telephone: Area Code (316) 2616451

-r c c., w ~ TECHNICAL SPECIFICATION CHANGES s .( 9 Page Section Reason-for Change.- o XX115 ' NA - Rs orting requiremnts-was completed on previous p Je and not continued on this page.. <2-4' Table 2.2 Cnange not made with'other changes on this page. '3/4 10 Footnote-Changer required to achieve correct meaning for footnote.

3/4J1-23'

.NA To be consistent with balance of Tech spec (All other references to three-loop operation have y been deleted.) "3/4 2-10 4.2.3.6 Specification not required since NCGS will utilize the leading edge flow meter to determine l degradation of the feedwater venturi. 3/4 2-11 . Action a.4 To be consistent with Tech Spec format.

( 3; 3/4-3-2&S Table 3.3-1 '

Justification with change in attachment. 3/4 3-6 Action 5 Change time frame to be consistent with Spec. 3.1.1.2. Valve lineups are controlled by locked valve lists and surveillance not necessary. 3/4 13 Action b.2 Typo. 3/4 '3-27 ~ Table 3.3-4' Reevaluation of design criteria. Time delay ~~& 3-28 8.a & b. and-changes were made to be compatiable with , Table notation Table 4.3 2. 3/4 3-30 4.a.6 Consistency with rest of table. 3/4 3-35 Table 4.3-2 Consistency with rest of table. 3/4 3 Table 3.3-6 Correction not picked up. To correctly identify 2.b instrumentation. 3/4.3-54 Table 3.3 To correctly identify instrument. Item 18' 3/4 !3-55 Table 4.3.7 Typo. g -*** Footnote 3/4 3-58, Table 3.3-11 Provide additional information. 59,60,& 61 3/4.3-71 Action 42 Typo.. -Action 44 Typo. 1 m ~ .~. ,-._-n-.__,

7, r-s i ~ 3 j

i TECHNICAL SPECIFICATION CHANCES (Continued) 1 Page.

Section Reason for Change 3/4 74 Table 4.3-9 Typos. Notations-(1)d.' & (2)- 3/4 '4-2

- Footnote.

Correctly identify Specification. . i, c 3/4.4-10 '3.4.4 Justification with change in attachment. Action a., b.~, & c. 3/4 4-34 Action a. Consistent with wording in Callaway Tech' Spec. 3/4 6-2 NA-Info to be provided after Performance'of Test. 3/4 6-11 3.6.1.7 Justification provided with change in attachment. Action b.- '3/4 21 Table 3.6-1 Valve functions misidentified.

1. P-69 3/4 6 -Table 3.6-1 Correctly identify. function.

6. P-76

Footnote Typo.

3/4 6-27 Table 3.6-1 Incorrect Nomenclature. & 28 3/4 6-29'& = Table 3.6-1 These valves are no6 considered as containnent. 30 3. isolation valves in the FSAR. See justific1 tion provided with change in attachment. 3/4 7-28 4.7.10.1.lf.- a. Systen does not have automatic valves in flow path b.- Per FSAR, 80 psig is correct value. c. Change required to clarify requirement and to correspond with design. 3/4. 7-34 Table 3.7-3 Consistency with rest of table. 3/4 7-35 Table 3.7-3 Provide necessary information concerning equipment. 3/4 8-3, 4, 4.8.1.1.2a.4), Changed to reflect design and for compatibility &5 f.2)', f.4) b), with Specifications 3.8.3.1 and 3.8.3.2. f.5), f.6) b), f.7). 2

(

n f~ TBCHNICAL' SPECIFICATION CHANGES (Continued) ~ t Page Section Reason for Change 3/4 8 -Action' State-To be consistent with 'est of Tech. Spec. r ment ~

3/4 8-9 c' 3.8. 2.la. _

Both are required,.not one'"or" the other. 3/4 8 Table 4.8-2' Clarification'. 3/4 8-39 Table 3.8 Typos. [ 3/4 8-40 ' Table 3.8-1 Typo. '3/4 9-18 4.9.13b.2) Typo. -3/4 10 4.10.4.3' Typo. '3/4 11-3 Table 4.11-1 Justification with change in attachment. Notation ~(2) 3/4 11-9 Table 4.11 Justification provided with change in 3/4 11-11 attachment. 3/4 11-16 3.11.2.6 Justification provided with change in-attachment. -3/4. 11-18 3/4.11.4 This page missing from Final Draft. 3/4 '.12-1 13.12.1 Justification provided with change in &2 Action C-attachment. s3/4-12-4: Table 3.12-1, Justification provided with change in 3.a. ' attachment. 3/4' 12-8 Table 3.12 Justification' provided with change in Notation'(7) attachment. 3/4 12-14 3/4.12.3-This page missing fro:n Final Draft. . 'B 3/4 = 2 3/4.2.2 & Clarify reference. j'? -s 3/4.2.3: B_3/4 2-5 3/4.2.2 and Just.ification with change in attachment. ~" 3/4.2.3 ,s v 5 k 3 ...,. -.. -. -.,,..... -.... -.,-. ~.. - - --

7... - u. ~ ~ ~ TBCHNICAL~ SPECIFICATION CHANGES (Continued) F I Page' Section Reason.for' Change B 3/4 3 -4 3/4.3.3.5 Consistent with desig'n and rest of Tech Spec. B 3/4 3-5 , 3/4. 3.3.10.- Typo. ' B 3/4 3-6

3/4.3.3.11 Typo.:

' B 3/4 4-9' 3/4.4.9 Typo.- ' B 3/4 '6 3/4.6.1.2 Revised to be consistent with 10CFR50, Appendix J. B 3/4 6-2 3/4.6.1.6 R.G. 1.35 is not a proposed R.G. B'3/4 6-3-- 3/4.6.1.7 Refer to justification for changes R ...7. B 3/4 7-6 3/4.7.8' Justification with change in attachment. . B 3/4 8-2-~ 3/4.8.1, Typo.- - 3/4.8.2 & -. 3/4.8.3 ' B 3/4 11-3 3/4 11.2.1 Typos. 5-2. ~ Fig. 5.1-1; Clarification of Figure. ...5-9 Table 5.7 Typos. 6-3 Fig. 6.2-1 Update organization chart.

6-4.

- Fig. 6.2-2' Update organization chart, 6-23 6.12.2~ Clarification. J t 4 ~ A k 2 M er yJ 1 + 4' s t 4 3-,.~,,,-....,1..-_._+.w,,_,,w,,_ ,,,,..--,,4 -...y..,,,,,

FRA!.MAq . ADMINISTRATIVE CONTROLS SECTION PAGE o,. om o,.,. .m..-_.._..__ .._m o-nwa v na asvu n byU A RL4'1Lll 3 J ( b b'55bscaws u f 6.11 RADIATION PROTECTION PR0 GRAM................................. 6-22 6.12 HIGH RADIATION AREA.......................................... 6-22 6.13 PROCESS CONTROL PROGRAM (PCP)................................ 6-24 0FFSITEDOSEdALCULATIONMANUAL(0DCM)....................... 6.14 6-24 6.15 MAJOR CHANGES 10 LIQUID, GASEOUS, AND SOLID RA0 WASTE TREATMENT SYSTEMS....................... 6-24 WOLF CREEK - UNIT 1 XXII

4 / 'E _ TABLE 2.2-1 q E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E SENSOR 7 -FUNCTIONAL UNIT TOTAL-ERROR-M LOWANCE (TA) Z_ __ (S) TRIP SETPOINT g 1. Manual Reactor Trip k.A. ALLOWABLE VALUE N.A. N.A. N.A. 2. Power Range, Neutron Flux N. A. a. High Setpoint 7.5 i 4.56_ .0 $109% of RTP* 5112.3% of RTP* b. Low Setpoint 8.3 i i 4.56. 0 $25% of RTP* $2d.3% of RTP* 3. Power Range, Neutron Flux,

2. 4 -

High Positive Rate 0.5 0 <4% of RTP* with <6.3% of RTP* with a time constant a time constant _2 seconds 4. _2 seconds Power Range, Neutron Flux, 2.4 i y High Negative Rate 0.5 0 $4% of RTP* with 56.3% of RTP* with + a time constant a time constant i 5. Intermediate Range, 12 seconds 12 seconds i Neutron Flux 8.41 0 52F% of RTP* $35.3% of RTP" 17.0 } 6. Source Range, Neutron Flux 17.0 i i 10.01 0 $105 cps $1.6 x 105 cps 7. W'1 1 Overtemperature AT

6. 9 w

2.83 2.26 See Note 1-l, 8. Overpower AT See Note 2 T "y R@ 5.5 M 1.43 1.35 See Not 3 See Note 4 k" 9. Pressurizer Pressure-Low 3.7-i875 0.71 2.4 1196tf psii 11866 psig d-j 10. Pressurizer Pressure-High 7.5 0.71 _2.49 psig $2400 psig '5id ) 11. Pressurizer Water Level-High 8.0 2.18 1.96 <92% of instrument <93.9% of instrument

if"I ipan ipan
RTP = RATED THERMAL POWER d

t

- Loop design flow = 95,700 gpm I

i ^

FEA!. MAFT . REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4' At least two centrifugal charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3.* ACTION: With only one centrifugal charging pump OPERABLE, restore at least two cen-trifugal charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours. SURVEILLANCE REOUIREMENTS 4.1.2.4 At least two centrifugal charging pumps shall be demonstrat$~d OPERABLE by verifying, on recirculation flow, that the pump develops a differential pressure of greater than or equal to 2400 psid when tested pursuant to Specification 4.0.5.

The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provi he centrifugal charging pump is restored to OPERABLE status within 4 ours rio to the temperature of one or more of the RCS cold legs exceeding,375 F.

OT ' WOLF CREEK - UNIT.1 3/4 1-10.

FINAL D3;R T - +: ob *. O Figure 3.1-2 lef lank pending NRC approval of three-loop o eration .1 - c 4 4 I 1 . WOLF _ CREEK - UNIT 1 3/4 1-23 p

- gg9w+4'***"*

'POWERDISTRIBUTIP'lLIMITS = u LIMITING CONDITION FOR.0PERATION ) ACTION (Continued)- Identify and correct the cause of the out-of-limit condition prior c. to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b., above; subsequent POWER OPERATION .may proceed provided that the combination-of R and indicated RCS-total flow rate _ are demonstrated, through incore flux mapping and-. .RCS total. flow rate comparison, to be within the: region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following ' THERMAL' POWER levels: - 1. A nominal 50% of RATED THERMAL POWER, 2. A nominal 75% of RATED THERMAL POWER, and 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS The provisions of Specification,4.0.4 are not applicable. 4.2.3.1 4.2.3.2 The combination of indicated RCS total flow rate and R shall be determined.to be within the region of accoptable operation of Figure 3.2-3: Prior to operation above 75% of RATED THERMAL POWER after each fuel a.' loading, and b. At least once per 31 Effective Full Power Days. 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the . region of acceptable operation of Figure 3.2-3 at least once per 12 hours when the.most recently obtained value of R obtained per Specification 4.2.3.2, is assumed to exist. 4.2.3.4 ~ The RCS total flow rate indicators shall be subjected to a CHANNEL . CALIBRATION at-least once per 18 months. ~ -4.2.3.5 The RCS total flow rate shall be det+.rmined by precision heat balance measurement at least once per 18 months. Within 7 days prior to performing the precision heat balance, the instrumentation used for determination of steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated. 4.2.3.6 The feedwater venturi shall be inspected for fouling and cleaned as necessary at least once per 18 months. \\ de.le.h WOLF CREEK - UNIT 1-3/4 2-10

s

r..

d Justification' for specification 4.2.3.6,-'pg. 3/4 2-10: s TheLWCGS feedwater system.uses a venturi asithe primary ~ umethod of flow measurement. - WCGS also has:the ability to measure. flow using-the Westinghousejleading edge flow meter (LEFM). iThe'LEFM'has a: higher ~ accuracy l~evel than'thea venturi and its' design-is unaffected by crud buildup. The ' LEFM c'an'therefore be-used to detect crud buildup on the ~ venturi and venturi: cleaning can:then;be scheduled. 'This ~ is a -standard operating.use at other LEFM installations. t b" s y Y Y 2 4 -s<- 6 g

POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO 1 s LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1,-above 50% of RATED THERMAL POWER *. ACTION: With the QUADRANT POWER TILY RATIO determined to exceed 1.02 a. less than or equal to 1.09: 1. Calculate the QUADRA.,T POWER TILT RATIO at least once per hour until either: a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THER POWER. 2. Within 2 hours either: a) Reduce the QUADRANT POWER TILT RATIO to within its e limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL PO for each 1% of indicated QUADRANT POWER TILi RATIO in ~ excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours. 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the n 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less-than or equal to 55% of RATED THERMAL POWER within the next 4 hours, and 4. Identify and correct the cause of the out of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL owe may proceed provided that the .QUADRX;T POWER TILT RATIO is erified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER. git caps ^See Special Test Exception Specification 3.10.2. WOLF CREEK - UFIT 1 3/4 2-11 ~

=, e TABLE 3.3-1 g REACTOR TRT ' SYSTEM INSTRUMENTATION 4 4 E MINIMUM i TOTAL NO. CHANNELS CHANNELS . APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP- ~0PERABLE MODES ' ACTION [ 1. Manual Reactor Trip 2 1 2 1, 2 1 ~ z 2-1 2 3*,4*,5*. 10' --a w 2. Power Range, Neutron Flux a. High Set' point-4 2 3 1, 2 b. Low Setpoint 4 2 3 1###, 2. 2#- i 2# 3. Power Range, Neutron Flux 4 2 3 ' 1, 2 2# j High Positive Rate 4. Powcr Range, Neutron Flux, 4 2 3 1, 2 2# y High Negative Rate i j y 5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 to j 6. Source Range, Neutron Flux a. Startup 2 1 2 4 b. Shutdown 2 1 2-3,4,5 5 7. Overtemperature AT N Four Loop Operation 4 2 3 1, 2 6# ~ j 8. Overpower AT D j Four Loop Operation 4 2 3 1, 2 6# 9. Pressurize Pressure-Low 4 2 3 1 6# 6 10. Pressurizer Pressure-High 4 2 3 1, 2 6# DM 4 i i

RR DRM7 ~ TABLE 3.3-1 (Continued) TABLE NOTATIONS $0nly if the Reactor Trip System breakers happen to be in the closed position and the~Faatrel Daa n,.49. quetam A capable of rod withdrawal. oron dilution flux doubling signal may be ing reactor startupe in accordam.e. wi+h noemt oeua+in3 ecouard. a

e

. i:f em ef Sa-Pir-=tiaa ?P are n:t oppi h.ao e.

1. Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
  1. Below the F-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: The inoperable channel is placed in the tripped condition a. within 1 hour; b. The Minimum Channels OPERABLE requirement is met; nowever, the inoperable channel may be bypassed for up to 2 hours-for surveillance testing of other channels per Specification 4.3.1.1; and Either, THERMAL POWER is restricted to less than or equal c. to 75% of RATED THERMAL ~ POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2. ACTION 3 -.With the number of channels OPERABLE one less than the Minimum c l Channels OPERABLE 7quirement and with the THERMAL POWER level: l Below the P-6 (Intermediate Range Neutron Flux Interlock) a. Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; or b. Above the P r (Intermediate Range Neutron Flux Interlock) l Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER. ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels CPERABLE requirement suspend all operations involving positive reactivity changes. WOLF CREEK - UNIT 1 3/4 3-5

5(- IF::-.. r 1

' V : ;Q h

,.; z, s ,' r p:,: ~ ~. n c. m. s g.; ;-6 s. T.24 % i ;;.,, ' q. 1, ~. ' [, ~- s,

- 0-( t

~ f y '/ !' Justification [forLTable3.N1,. pg.:3/4'3-2.and'3/4.3-5: y- ,~>. 4: . This' change allows startup which'is-othersise prevented .with,isignal?not blocked. I t I a g ?'. g s 4 -s, W T r' a L + F s ' s, O ,s +, 4 h / 4.

~. b ^ F!NA!. DRAFT TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) Re.elace. wig Ls, err. ACTION 5 the number of OPERABLE channels one less than.the Mini um Channe RABLE requirement,_ restore the inoperabl nnel to.0PERABLE s s within 48 hours or open th or trip breakers,- suspend operations _invo positive reactivity -changes and verify Valv G-V d BG-V601 are closed and secured.in position wit ext hour. With-no_ channels OPERABLE. verify ance with t UTDOWN MARGIN requirements of Specific n 3.1.1.1 or 3. 2.1.2, as cable, and_take the actio ated above within 1 hour and verify ance at least: .pm .per 12 hours thereafter. ACTION 6 - With the number of OPERABLE ch&nnels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed' provided the following conditions are satisfied: The inoperable channel is placed in the tripped condition a. within 1 hour; and b. The-Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.1.1. ACTION 7 ~- With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided'the inoperable channel is placed in the tripped

e condition within 1 hour.

ACTION 8 - With less than the Mnimum Number of Channels OPERABLE, within i: 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state-for the existing plant condition, or apply Specification 3.0.3. ACTION 9 - With the number of OPERABLE channels one less than the Minimu Channels OPERABLE requirement, be in at least HOT STANOBY i within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 10 - With the number of OPERABLE channels one less than the M Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor trip breakers within the next hour. ACTION 11 - With the number of OPERABLE channels less than the Total Nu ~ of CLannels, operation may contin'se provided the inoperable i channeis are placed in the tripped condition within 1 hour. d i.0LF CAEEK - UNIT 1 3/4 3-6 i 1_ _-._._._._ _ __ _ _.._._,___ __. _ ____ __ _._ _ _ __ _ _ _

+ -m-n;- ~ -2, an

g $,

t c ~ ' t.( . h> s. ,A ,i lie Insert for ' Action 5,.: pg j 3/4,3--6 :. p yy r - -'a 'Withithe' number of OPERABLE channels one _ n - li . Action?5- - a.- iless than-the Minimum: Channels OPERABLEi ~ ? 1 requirement,lrestoreEthe inoperable channel 'to OPERABLE: status within~48 hour's or open s' -the Reactor Trip Breakers,. suspend.allL operationsfinvolving-positive reactivity-changesLand verify. valves BG-V178 and .BG-V601 are' closed'and secured in position within:the next hour.

&. W

- b. !With no: channels OPERABLE, open.the Reac-stor-Trip Breakers,Jsuspend-all operations Jinvolving-positive. reactivity changes and u

verify compliance.with the_ SHUTDOWN MARGIN f'

W' requirements of-Specification 3.1'.l.1 orJ 3.1.1.2,.as applicable, within.1 hour and' - 1every 12. hours thereafter,.and verify

valves BG-V178 and BG-V601'are closed and

.securedfin position within.4 hours and~ l cverified to be. closed and secured in posi-9 tion every'14 days. w ', ~ 1 J P. i ,d'_,J_-m-9. g ~ s. + 3 v. 1 L. 2. ^ k' yN>-fN'V"e 'v y 't w'-f9v- @W Pta'-T"rT 'fN W'T I 198'M*4 #7TW %' WWT'D@*f"P'Wf'W-*T*F-M#II'T'T'*-T-PAW-'TNyY9'-'**W#*W W*T-YW T'M9 "f 7~ TWYsw fW W W W TfPTV 'F b 9 T'T *'9PT'DYCf'"'NW**"-9'TT '" M

i,, 4_ -= t r. ,--, r Y ' Justification ~for Table-3.3-1,; Action-5, pg. 3/4 :3-6 : - i. -- - Verification time frames have been changed to allow rea-sonable; time..to' perform actions and to reflect the admin- '-istrative controls-provided by-the locked' valve list. J 0 1 b I s .h c y o- + f 1 L I i 4 L s i -'WW ev'T* 9V 4WMw N re w-**-7e si e w "1s4'F' ty- - * *wp -pQ--- 9 g v w-NN'y 7'm38+-Tu--==*$9W* ggv e ew We wg 9%er'T T" M 9e4 w*$ e'Mr h9g=1s W-pg-y W W *p mgtwwg y rg F W'W

.g INSTRUMENTATION 3/4.3.2' ENGINEERED SAFETY FEATURES' ACTUATION SYSTEM INSTRUME LIMITING CONDITION FOR OPERATION ~ '3.3.2~ The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks'shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-S. APPLICABILITY: 'As shown in Table 3.3-3. ' ACTION: With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva--

a.

Ltive than the value shown in the Trip Setpoint column but more conser-vative than the value.shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value. b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conser-vative than the value shown in the Allowable Values column of Table 3.3-4, either:- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or 3,3 - 3 2. Declare the channel inoperable and a p he applicable ACTION statement requirements of Table,.3. until the channel is restored to OPERA 8LE status with its Setpoint adjusted consistent with the Trip Setpoint value. Equation 2.2-1 ~ Z + R + S < TA Where: I = The value from Column Z of Table 3.3-4 for the affected channe!. R = The "as measurec" value (in percent span) of rack error for the affected channel, = Either the' "as measured" value-(in percent span) of the S sensor error, or the value from Column.5 (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column-TA (Total Allowance) of Table 3.3-4 for the affected channel. With an ESFAS instrumentation channel or interlock inoperable, take -c. -the ACTION shown in Table 3.3-3. SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic -actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2. L 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS fun shall be demonstrated to be within the limit at least once per 18 months. Each test _shall include at least one train such that both trains are tested at 1 east once per 36 months and one channel per function such that all channels L are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of L Channels" Column of Table 3.3-3. ~ WOLF -CREEK UNIT 1 3/ 4 3 13

S -TABLE 3.3-4 (Continued) E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPO ng TOTAL SENSOR TRIP. ALLOWABLE p FUNCTIONAL UNIT ALLOWANCE (TA) Z ERROR (S)- SETPOINT VALUE

8. Loss of Power a.

4 kV Undervoltage 3t3V 7N # 5W H -Loss of Voltage N.A. N.A. N.A. (120V (120V Bus) us) w/1.+ 0.2, -0.5s delay b. 4 kV Undervoltage delay i.o t ea s -Grid Degraded sos,.9v s o 9. 3V Voltage H.A. N.A. N.A. > (10f'57) ' > (104.59(120V Bus) T120V Bus), w/11s t-11.6s delay *

9. Control Room Isolation w/119s delay w/ s s.6s delay we y

W/8 t o.ss Jelay s Manu,31 Initiation H.A. N.A. N.A. N.A. N.A a. U b. Automatic Actuation Logip and Actuation Relays (SSPS) N.A. N.A. N.A. N.A. N.A c. Autogatic Actuation M Logjlc'andActuation wy Relays (80P ESFAS) N.A. N.A. N.A. N. A. N.A. g d. Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and E Allowable Values. ~D

10. Solid-State Load Sequencer N.A.

N.A. N.A. N.A. .N.A. M

11. Engineered Safety Features Actuation System Interlocks M

N a. Pressurizer Pressure, P-11 N.A. N.A. N.A. $ 1970 psig 5 1979 psig b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.

FINAL DRAFT TABLE 3.3-4 (Continued) TABLE NOTATIONS

Time constants utilized in the lead-lag controller for Steam Pressure-Low are It >_ 50 seconds and T2 < 5 seconds.

CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values. ~ **The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value. Apphcable. .4o c.kanne.1 cal locaAion only.

A pbcable-4o 4tip acAul;ng device. epera4tona l +es+ only.

p I WOLF CREEK %JJ 3/4 3-28

a 4. TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES R_ESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 3. Pressurizer Pressure-Low , Safety Injection (ECCS) 1 29(1)/12(4) a. 1)- Reactor Trip i2 2) Feedwater Isolation 17- %)~ < Phase "A" Isolation. 5 2(5) 4) Auxiliary Feedwater 5 60 5) Essential Service Water 1 60(1) 6) Containment Cooling 1 60(1) 7) ' Component Cooling Water N.A. 8) Emergency Diesel Generators i 14(6) 9) Turbine Trip N.A. 4. Steam Line Pressure-Low Safety Injection (ECCS) 1 24(3)/12(4) a. 1) Reactor Trip 52 2) Feedwater Isolation 17 3) Phase "A" Isolation 1 2(5) 4) Auxiliary Feedwater 5 60 5) Essential Service W 1 60(1) 6) Containment Cool ng reis i 60(1) S 7)' Component Cooling Water N.A. 8) Emergency Diesel Generators i 14(6) 9) Turbine Trip N. A. b. Steam Line Isolation 17 WOLF' CREEK - UNIT.1 3/4 3-30

u e TABLE 4.3-2-(Continued) g uc,megees suwre wrx,ouc wruariou sv1rs" Wh"'** * "*# ,g sue ueiunoer Du mua****f1R I P.. ' r-awasnn .iCTUATiim MODES [ CilANNEL-DEVICE. MASTER-SLAVE ~ FOR WHICil-J CilANNEL CilAN14EL OPERATIONAL OPERATIONAL ACTUATION. RELAY' RELAY 1 SURVEILLANCt 1 p FUNCTIONAL UNIT CilECK CALIBRATIONL TEST TEST-LOGIC TEST-TEST TEST IS REQUIRED

3. Containment Isolation-C$

a. Phase "A" Isolation , ['

1) -Manual Initiation N.A.

N.A. N.A. R N.A. N.A. N. 'A. 1, 2,.3, 4. 2)' Automatic Actuation N.A N.A. N.A. N. A.- M(1) M(1). Q(3) 1, 2, 3, 4 Logic and Actuation Relays (SSPS) 4

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements, b.

Phase "B" Isolation 1)~ Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4.

2) Automatic Actuation N.A.

'N.A. N.A. N.A. M(1) M(1)' Q 1, 2, 3, 4 i Logic and Actuation Relays (SSPS) l

3) Containment S

.R M N.A. N.A. N.A. N.A. 1,.2, 3-i Pressure-itigh-3 4 l c. . Containment Purge Isolation }

1) Manual Initiation N.A.

N.A. N.A. R N.A. N.A. N.A. 1,2,3,4

2) Automatic Actuation N.A.

N.A. N.A. N.A. M(1) M(1) Q(3) 1, 2, 3,o 4 i Logic and Actuation } Relays (SSPS) j

3) Automatic Actuation N.A.

N.A. N.A. N.A. M(1)(2) N.A. ~N.A. 1, 2, 3, 4 ] Logic and Actuation l Relays (80P ESFAS) j

4) Phase "A" Isolation See Item 3.a. above for all Phase "A" Isolation Surveillance Requirements.

4

m . TABLE 3.3-6 ~ g RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 9 S. MINIMUM CHANNELS . CHANNELS-APPLICABLE-ALARM / TRIP h. FUNCTIONAL UNIT-TO TRIP / ALARM' ' OPERABLE MODES SETPOINT-ACTION [. 1. Containment E -4 a. Containment Atmosphere-e Gaseous' Radioactivity-High (GT-RE-31 &.32) 1 2 All 26 b. Gaseous Radioactivity-RCS leakage Detection (GT-RE-31 & 32) N.A. 1 1,2,3,4 'N.A. 29 c. ' Particulate Radioactivity- [M RCS Leakage Detection (GT-RE-31 & 32) N.A. I 1,2,3,4 N.A. 29 b 2. Fuel Building a. Fuel Building Exhaust- % "J Gaseous Radioactivity-by High (GG-RE-2 28) 1 2 ^^ m 30 b. Criticall y-N;3h gm Radiation

1) Spent Fuel Pool 1

1 g (SD-RE-37 or 38) 5 15 mR/h 28 N

2) New Fuel Pool 1

1 (SD-RE-35 or 36) 1 15 mR/h 28 $~2 "f1 3. Control Room Air Intake-Gaseous i Radioactivity-High i l (GK-RE-04 & 05) 1 2 All 27 i

)

-TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION g n n TOTAL MINIMUM-y INSTRUNENT NO..OF CHANNELS CHANNELS-OPERABLE 7 1. Containment Pressure y a) Normal Range 2 1-M b) Extended Range 2 1 2. Reactor Coolant Outlet Temperature - Til0T (Wide Range) 2 e 1-3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1 4. Reactor Coolant Pressure - Wide Range 2 1 5. Pressurizer Water Level 2 1 6. Steam Line Pressure 2/ steam. generator 1/ steam generator -{ 7. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator y 8. Steam Generator Water Level - Wide Range 1/ steam generator 1/ steam generator 9. Refueling Water Storage Tank Water Level 2 1 10. Containment Hydrogen Concentration Level 2 1 11. Auxiliary Feedwater Flow Rate 1/ steam generator 1/ steam generator 12. PORV Positica Indicator

1/ Valve 1/ Valve 13.

PORV Block Valve Position Indicator ** 1/ Valve 1/ Valve 14. Safety Valve Position Indicator 1/ Valve 1/ Valve 15. Containment Water Level qug 2 1 16. Containment Radiation Level (iiigh Range) N.A. 1 17. Thermocouple / Core C in election System 4/ core quadrant 2/ core quadrant p 18. Unit Vent - liigh R nge/ Nob e Gas Monitor N.A. 1 U.13

Not applicable if the associated block valve is in the closed position.
- Not applicable if the block valve is verified in thp closed position and power is removed.

h-N 6 i3 =l

y TABLE 4.3-7 ACCIDENT MONITORING INSlRUMENTATION SURVEILLANCE REQUIREMENTS G Q INSTRUMENT CilANNEL CHANNEL CilECK CALIBRATION. h 1. Containment Pressure M R ~ s 2. Reactor Coolant Outlet Temperature - THol (Wide Range) M R h 3. Reactor Coolant Inlet Temperature - ICOLD (Wde Range) M R 4. Reactor Coolant Pressure - Wide Range M R 5. Pressurizer Water Level M R 6. Steam Line Pressure M R 7. Steam Generator Water Level - Narrow Range M R. 8. Steam Generator Water Level - Wide Range M R 9. Refueling Water Storage Tank Water Level M R A 10. Containment flydrogen Concentration Level M R y 11. Auxiliary Feedwater Flow Rate H R 12. PORV Position Indicator

M N.A.

13. PORV Block Valve Position Indicator ** M N.A. 14. Safety Valve Position Indicator M N.A. 15. Containment Water Level M R 16. Containment Radiation Level (iiigh Range) M R*** 17. Thermocouple / Core Cooling Detection System q M R 18. Unit Vent - High Range Noble Gas Monitor E38 M-R F&

Not applicable if the associated block valve is in the closed position.
- Not applicable if the block valve is verified in the closed position and power is removed.
  - CilANNEL CALIBRATION may consist of an electro '

~ n of the channel, not including the detector, mj ai for range decades above 10 R/h and a one p 'nt calibuties c 5 ck of the detector below 10 R/h with an s:-c3 installed or portable gamma source. Callrat;en gg M cz4

~. ; r j TABLE 3.3-11 FIRE DETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS

INSTRUMENT LOCATION ZONE HEAT FLAME SM0KE (x/y)

(x/y) (x/y) 1101-Aux. Bldg. 1974' Gen. Fir. #1 100 0/11 ~ 1102-Chiller & Surge-Tks._ Area 100 0/4 1102-Chiller & Surge Tks. Area 101 2/0 1107-Cent. Charg. Pmp. Rm. B 101 2/0 1108-Safety Inj. Pmp..Rm. B 101 2/0 .1109-Res. Ht. Remov. Pop. Rs. 8 101 1/0 1110-Ctat. Spray Pap. Rm. B 101 1/0 .1111-Res. Ht. Remov. Pop. Rs. A 101 1/0 1112-Ctat. Spray Pmp. Rs. A 101 1/0 1113-Safety'Inj. Pop. Rs. A 101 2/0 1114-Cent. Charg. Pap. Rs. A 101 2/0 1115-Pos. Disp. Charg. Pmp. Rm. 101 2/0 -1116,.1117-Boric Acid Tk. Res. 101 2/0 1116, 1117-Boric Acid Tk. Res. 101 2/0 1120-Aux. 81dg. 1974' Gen. Fir. #2 101 4/0 ~ 1122-Aux. Bldg. 1974' Gen. Fir. #3 100 0/3 1122-Aux. Bldg. 1974' Gen. Fir. #3 101 5/0 1126-Boron Inj. Tk. & Pop. Rm. 101 1/0 1127-Sta h ?c! 109 1/0 112(Aus Wh< be #" 8"" 117 2/0 1130-Aux.. ts i og. 1974

n. t.orr.

100 0/2 ipe Chase Below AFWP Area 117 2/0 saeJ-Aux. Bldg. Elec. Chase S. 1988' 117 1/0 1301- . Bldg. 2000' Corridor #1 103 0/10 1-Aux. Bldg. 2000' Corridor #1 117 2/0 1311-Aux.- Bldg. Sampling Rm. 117 2/0 1312-Boron Meter /RC Activity Mon. Rs. 103 0/1 1314-Aux. Bldg. 2000' Corridor #3 103 0/3 1314-Aux. Bldg. 2000' Corridor #3 117 2/0 1315-Ctat. Spray Add. Tk. Area 103. 0/2 1316-Viv. Rs. by Seal Wtr. Ht. Exch. 103 0/1 1320-Aux. Bldg. 2000' Corridor #4 103 0/3 1321-Aux. Bldg. 2000' S. Exit Vest. 103 0/1 1322-Pipe Pene. Rm. B 117 5/0 1323-Pipe Pene. Rm. A 117 6/0 1325-Aux. FW Pmp. Rm. B 117 2/0 1326-Aux. FW Pmp. Rm. A 117 2/0 1331-Aux. FW Pmp. Rs. C 111 2/0 1331-Aux. FW Pmp. Rm. C 117 1/0 '1335-Aux. Bldg. Elec. Chase N. 2000' 117 1/0 1336-Aux. Bldg. Elec. Chase S. 2000' 117 1/0 1401-Comp. Cool. Pmp. & Ht. Exch. B 118 5/0 1402-Aux. Bldg. 2026' Corridor #1 104 0/6(7) 1403-MG Set Rs. 105 0/9(1) 1403-MG Set Rm. 112 0/9 WOLF CREEK - UNIT 1 3/4 3-58

p t ,MI l ' TABLE 3.3-11 (Continued) ~ FIRE DETECTION INSTRUMENTS 0 T ENTS* INSTRUMENT LOCATION ZONE HEAT FLAME-SMOKE (x/y) (x/y) (x/y) 1405-Chemical Stg. Area 118-6/0 1406-Comp. Cool. Prap. & Ht. Exch. A 104. 0/1 1406-Comp. Cool. Pop. & Ht. Exch. A 118 2/0 _1408-Aux. Bldg. 2026' Corridor #2 104 0/9 1408-Aux. Bldg.-2026' Corridor #2 118 5/0 1409-Elec. Pene. Rm. B 106 0/4(1) 1409-Elec Pene. Rm. B 113 0/4(1) 1410-Elec.. Pene. Rm. A 107 0/8(1) 1410-Elec..Pene. Rm. A 114-0/8(1) 1413-Aux. Shutdown Pnl. Rm. 118 4/0 1501-Ctr1. Rm. A/C & Filt. Units 8 110 10/0 1504-Ctat. Purge Exh. & Mech. Equip. B 108 18/0 1506-Cat. Purge Sup. AHU Rm. A 109 18/0 1507-Personnel Hatch Area 108 3/0 -1508-Main Steam Iso. Valve Rm #1 115 1/0 1509-Main Steam Iso.~ Valve Rm #2 115 1/0 2-Ctr1. Rm. A/C & Filt. Units A 110 10/0 151 Ctrl. Bldg. Vent Sup. A/C Unit Rm. 109 3/0 isa-Aux. Bldg. Duct 2047'6" 119 1/0 = . NA - Containment ** 201 1/0(2) WA.Dantainment** 202 2/0((2) 2) ua - (ontainment**, 203 1/0 1/0(2) 2) un-( ontainment** 204 3/0((2) NA - Containment ** 206 .NA-Containment ** 215 1/0 -1/0(2) 2)

4-Containment **

216 1/0((2) NA-C ontainment** 217 N4-ontainment** 218 1/0 44 - ontainment** 219 4/0 d+ - ontainment** 220 1/0(2) 31 -Ctrl. Bldg. 1974' Pipe Space 300 11/0 + 3105-Ctrl. Bldg. Elec. Chase S. 1974' 300 1/0

1. Bldg. Elec. Chase N. 1974' 300 1/0 N4 ea Above Access Control 301 12/0 trl. Bldg. Elec. Chase S. 1984' 300 1/0 3230-Ctrl. Blog. Elec. Chase N. 1984' 300 1/0 0/7(7) 1)

3301-ESF Swgr. Rm. #1 314 0/7(1) 3301-ESF Swgr. Rm. #1 315 0/5( 3302-ESF Swgr. Rm. #2 316 3302-ESF Swgr. Rm. #2 317 0/5(1) 3305-Ctr1. Bldg. Elec. Chase S. 2000' 301 1/0 3306-Ctr1. Bldg. Elec. Chase N. 2000' 301 1/0 3403-Non-Vit. Swgr. & Xfmr. Rm. #1 304 0/1(1) 3403-Non-Vit. Swgr. & Xfmr. Rm. #1 305 0/1(1) 3404-Switchboard Rm. #4 321 0/2(1) WOLF CREEK - UNIT 1 3/4 3-59 h

e y j',.g . 40 [ ,D' -TABLE 3.3-11 (Continued) \\ N FIRE DETECTION INSTRUMENTS ,2" TOTAL NUMBER .g

0F INSTRUMENTS

INSTRUMENT' LOCATION ZONE HEAT FLAME SMOKE

-e 3 (x/y)- (x/y) (x/y) ,,s 3404-Switchboard Rs. #4 L ' 322 0/2(1) ~ 3405'-Battery Rs.' #4 3407-Battery Rs.'.-#1. ' 303 2/0

3408-Switchboard Rs. #1.

303 2/0 +. .325: 0/2(1) 3408-SwitchboardeRm. #1 326 0/2(1) L 3409-Non-Vit. Swgr. & Xfar. Rs.t#2' -323-0/1(1) ip =3409-Non-Vit.1Swgr.&Xfar.-Rm.72 327 0/1((1) 1) 3410-Switchboard Rm. #2-324 W$f 1 3410. Switchboard Rm. #2 328-0/2(1) ? 0/2 3411-Battery Rs. #2- '1 303-2/0 + f3413-Battery Rs. #3-. 303 1/0 F j4 3414-Switchboard Rm. '#'.f " 318 0/2 .r s f3414-Switchboard Rm.^#3 320 0/2 3.l G " 13415-Acc.-Ctr1. & Elec. Equip. A/C 303 4/0 . Units #1 3416-Acc.-Ctr1..& Elec. Equip. A/C 303 4/0 ) Units #2 34 -Ctr1. Bldg..Elec. Chase S. 2016' 303 1/0 419-rl. Bldg. Elec. Chase N. 2016' 303 ^ ~ 1/0 ' JM -C cl. Bldg. Elec. Cha::e N. - 2016' 303 1/0 2Wo tr1.. Bldg. Elec. Chase >S. 2015' i 303 1/0 1-Lower Cable' Spreading Rm.' " 306 0/13 ~ y 3504-Ctr1. Bldg. Elect Chase N. 2032' 303 1/0 Q'2 ' tr1. Bldg. Elec. Chase S. 2032'. 303 1/0 p ,stol-C rl. Bldg.'Elec. Chase N. 2032' i 303 1/0 '3f88 - trl. Bldg. Elec.. Chase S. 2032' 4 03 1/0 m -Control Room a 308 4/0(1) Y 3601-Control' Room 309 ~ 0/7(1) 0/7 13601-Control Room 319 + 3601-Control Room 329 ~20/0 '3602-Pantry 308-1/0 3603-Shift Supv. Office 308 1/0 7 3605-Equipment cabinet Area i 308 15/0 3606-Emerg. Equip. Storage Rs. 308 1/0

3608-Janitor's Closet 308 1/0

3609-SAS Rs. 308 1/0 3617-Ctr1. 81dg. Elec. Chase'S. 2047'G" 308 1/0 tr1. 81dg. Elec. Chase N.,2047'6" 308 1/0 3686 tr1.~ Bldg. Elec. Chase S.-2047'6" 308 1/0 -U per Cable. Spreading Rs. 307 0/18 rl. Bldg. Elec. Chase S. 2073'6" 308 1/0 38er tr1. 81dg. Elec. Chase S. 2073'6" 308 1/0 1-W. Diesel Gen. Rs. 501 4/0 5201-W. Diesel Gen. Rm. 502 0/8 5203-E. Diesel Gen.'Rm. 500 4/0 = WOLF CREEK - UNIT 1 3/4 3-60

FIR. PRM TABLE 3.3-11 (Continued) -v -FIRE DETECTION INSTRUMENTS TOTAL NUMBER OF INSTRUMENTS

INSTRUMENT LOCATION

. ZONE HEAT FLAME SMOKE (x/y) (x/y) (x/y) 5203-E.' Diesel Gen. Rs. 503 0/8 6102-Fuel Bldg.' Railroad Bay 600 0/8 6104-Fuel Pool Cool. HX Rm. 8 601 6/0 6105-Fuel Pool Cool. HX Rs. A 601-6/0 6202-Elec.-Equipment Rm. 601 3/0 6203-Air Handling Equip. Rm. 601 3/0 6301-Fuel Bldg. 2047'6" Gen. Fir. 602 2/0 6303-Fuel ~81dg. Exh. Filt. Absorb. 601 2/0 Rm. A-4-Fuel Bldg. Exh. Filt. Absorb. 601 2/0~ .B-N4 -E9d Pumphouse Train B 002 3/0 NA -ESW Pumphouse Train A 001 3/0 NA ES: Transformer XNB01 016 0/6 NA -E F Transformer XNB02 017 0/6 TABLE NOTATIONS ..*(x/y): x is number of Function A (early warning fire detection and notification only) instruments. y is number of Function B (actuation of fire suppression sys-tems and early warning and notification) instruments.

- The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A containment leakage rate tests.

i (1) Zone is associated with a Halon protected. space. Each space has two l~ separate detection circuits (zones). One zone, in its entirety, needs to remain OPERABLE. i (2) Line-type heat detector. t P 9 L WOLF CREEK - UNIT 1 3/4 3-61

s l .U b .1 'o -TABLE 3.3-13 (Continued)- 9 TABLE ~ NOTATIONS

?At all times.

i** During' WASTE GAS HOLDUP' SYSTEM operation. ACTION STATEMENTS ' ACTION 38 - With the number of channels OPERABLE less than required Minimum Channels OPERABLE requirement, the contents of the tank (s) -may be released to the environment for up to 14 days provided that , prior to initiating the release: At.least two independent samples of the tank's contents are a. ' analyzed, and b. At' least two technically qualified members of the facility staff independently verify the release r' ate calculations and discharge valve lineup. Otherwise, suspend' release of radioactive effluents via this pathway. ACTION 39 --With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releasu via this pathway may continue for up to 30 days provided the flow rate is estimated at least 'once per 4 hours. ACTION 40 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples.are taken at.least once per.12 hours.and these samples are analyzed for radioactivity within 24 hours. ACTION ~41 - With the number of channels OPERABLE less thcn required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway. c ACTION 42 --With the Outlet Oxygen Monitor channel inoperable, operation of the system may continue provided grab samples are taken and analyzed at least once per 24 hours. With both oxygen channels or both the inlet oxygen and inlet hydrogen channels inoperable, -suspend oxygen supply to the recombiner. Addition of waste gas to the system may continue provided grab samples are taken analyzi.d at least once per 4 hours during degassing opera ions n d at least once per 24 hours during other operations. ACTION 43 - With the number of channels OPERABLE less than required by the ~ Minimum Channels OPERABLE requirement, efrluent releases via the affected pathway may continue for up to 30 days provided samples are continuously. collected with auxiliary sample equipment as required in Table-4.11-2. ' ACTION 44 - With the number of channels OPERABLE one less a required by the Minimum Channels OPERABLE requirements, susp nded a ygen supply to the recombiner. ACTION 45 - Flow rate for this system shall be based on fan status and operating curves or actual measurements. -WOLF' CREEK - UNIT 1 3/4 3-71 e-

7 n ? ~ 4 -m 'Q p 9 Sn@h} Ebn [w.,.f; q_ TABLE-4.3-9~(Continued)' TABLE NOTATIONS

AtLalltimes.

e r L** During WASTE GAS: HOLDUP SYSTEM operation. (1): The. ANALOG' CHANNEL OPERATIONAL TEST shall' a 4 isolation of thisipathway-and control, room alarm annunciation as. a C

occur. if. any of the 1following conditions exists::

j[ (isolation and' alarm),lor-Instrument indicates measured level .a. 1 .P b.- 'Circuitifailure (alarm only), or - ~ c. trument indicates a downscale failure (alarm only) or n 1 d. Ijgtumentcontrolsnot. set:inoperatemode(alarmonly). [. J(2)n The ANALOGTCHANNEL OPERATIONAL TEST shall als room alarm annunciation occurs if any of the following conditions e ist/: f t - a. . Instrument indicates measured levels above the Alarm Setpoint, or b.~

Circuit ' failure, or Instrument indicates-a downscale failure, or c.

d. Instrument' controls not set-in-operate mode. .(3). The initial CHANNEL CALIBRATION,shall be performed using one or mo -ine reference (gas.or liquid and solid) standards certified by the National'~ Bureau of Standards =(NBS) or using standards that have been ~ btained-from suppliers that participate in measurement assurance activi-o . ties ~with NBS. These' standards shall permit calibrating the system over i ~

itsfintended range _ of-ener~gy, measurement range, and establish monitor i

response to a: solid ~ calibration source. NBS traceable standard (gas, liquid,:or solid) may be used; or a . liquid,_~or' solid source that has been calibrated by relating it to equip- , ment.that was previously (within 30 days) calibrated by the same geometry and type of source traceable to NBS. (4) The CHANNEL CALIBRATION shall include the use of standard gas samp containing'a nominal: One volume percent hydrogen, balance nitrogen, and -: a. b. -Four volume percent h'drogen, balance nitrogen. y WOLF CREEK - UNIT 1-3/4 3-74

l REACTOR COOLANT SYSTEM i HOT STAN0BY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least three of the reactor coolant loops listed below shall be OPERABLE and at least two of these. reactor coolant loops shall be in operation:* Reactor Coolant Loop A and its associated steam generator and a. reactor coolant pump, b. Reactor Coolan.t Loop B and its associated steam generator and reactor coolant-pump, Reactor Coolant Loop C and its associated steam generator and -c. reactor coolant pump, and d. -Reactor Coolant Loop D and its associated steam generator and reactor coolant pump. APPLICABILITY: MODE'3.** ACTION: With less than the above required reactor coolant loops OPERABLE, a. restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours, b. With only one reactor coolant loop in operation, restore at least two loops to operation within 72 hours or within 1 hour open the Reactor Trip System breakers. With no reactor coolant loop in operation, suspend all operations c. involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation. SURVEILLANCE REQUIREMENTS ~ 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side wide range water level to be greater than or equal to 10% at least once per 12 hours. 4.4.1.2.3 At least two reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. "All reactor coolant pumps may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant Sys oron concentration, and (2) core outlet temperature is maintain at lea 10*F below saturation temperature.

- See Spe al Exceptio Specification 3.10.4.

p WOLF CREET JNIT 1 3/4 4-2 "Tesi }

.f. 0,.Q R h b i}s,'[l$h h, m ,~mm REACTOR COOLANT SYSTEM fJf 3/4.4.4 RELIEF VALVES LIMITING-CONDITION FOR OPERATION' L 3.4.4 - All power-operated relief valves (PORVs) and their associat d bl ~ valves shall be OPERABLE. e ock APPLICABILITY: MODES 1, 2, and 3.* ACTION: With one or more PORV(s) inoperable because of excessi e a. within 1 hour either restore the PORV(s) to OPERABLE s e age, within the next 6 hours and in COLD SHUTD close 30 hours. b. With one PORV inoperable due to causes other than excessi e s # 1eakage, within 1 hour either restore the PORV to OPERABL or. close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours SHUTDOWN within the following 30 hours. With both PORV(s) inoperable due to causes other than excessi e seat'-- c. leakage, within 1 hour either restore each of the PORV(s) to 0 the block valve (s) and be in HOT STANDBY with COLD SHUTDOWN within the following 30 hours. d. With one or more block valve (s) inoperable, within 1 hour: the block valve (s) to OPERA 8LE status (1) restore remove power from the block valve (s),,or close the.PORV to remove power from its associated solenoid valve; and (2) apply ACTION b. or

c. above, as appropriate, for the isolated PORV(s).

The provisions of Specification 3.0,4 are not applicable. e. SURVEILLANCE REOUIREMENTS In addition to the requirements of Specification 4.0.5, each PORV 4.4.4.1 ' CHANNEL CALIBRI.1 ION.shall be demonstrated OPERABLE at least once p 4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unle the block valve is closed with power removed in order to meet the reoufrement of ACTION b. or c. in Specification 3.4.4.

With all RCS cold leg temperatures above 368'F.

WOLF CREEK - UNIT 1 3/4 4-10 ~

fi:l ti _ ) s,,, 4 g ~ Justification ' for, specification 3.4.4, Actions a., b., 'and c.,- .pg.;3/4_4-10: Other;: type's ~ of leakages' develop -in these valves (such as body-to-bonnet) which are just as :isoable by' the block valves 'as seat leakage.- The intent of-thisfspecification 'is:to' allow? continual opera' ion.~if_-the--PORV is operable iexcept for'relatively1small-leakages. Deletion of'the adverb " seat" allows _this intent-to be realized.. I .r + a "wr-1y-q9v-77 t,-w--9%y www 9'eP

e.. '. = 5 y= =. 3 J Irv J I REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS . LIMITING CONDITION FOR OPERATION 3.4.9.3 'At least one of the following Overpressure Protection Systems shall .be OPERABLE: Two residual heat removal (RHR) suction relief valves each w a. Setpoint of 450 psig i 3%, or b. Two power operated relief valves (PORVs) with Setpoints which do not exceed the limit established in Figure 3.4-4, or The Reactor Coolant System (RCS) depressurized with an RCS vent of c. greater than or equal to 2 square inches. APPLICABILITY: MODE 3 when the temperature of any RCS cold leg is less than or equal to 368 F, MODES 4 and 5, and MODE 6 with the reactor vessel head on. ACTION: oNE owe

noreembit-With h= th= tweg ORV or 4we RHR suction relief valvey ^.^: ^m,

a. P 3 either restore two PORVs or two RHR suction relief valves to O status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours. b. With both 90RVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through at least a 2 square inch vent within 8 hours. In the event the PORVs, or the RHR suction relief valves, or the RCS c. vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or the RHR suction relief valves any corrective action necessar,y to prevent recurrence.or RCS vent (s) d. The provisions of Specification 3.0.4 are not applicable. WOLF CREEK - UNIT 1 3/4 4-34

CONTAINMENT SYSTEMS -CONTAINMENT

LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 -Containment = leakage rates shall be limited to:

u..

a. - An overall integrated leakage rate of: O M >-- z 1) Less than or equal to L, 0.20% by weight of the containment ct 4 a IIi o air per 24 hours'at P, 48 si, or og a $I 2) -Less than or equal t L. % by weight of the.

ainment 0

t 7 air per 24 hours at P ' @ l-- - t 9' t-b. A combined leakage rate of less than 0.60 L, for all penetrations and W valves subject to Type B and C tests, when pressurized to P,,48 psig. et I C.i > APPLICABILITY: MODES 1, 2, 3, and 4. c', Z ' h ACTION: With either the measured overall integrated containment leakage rate exceeding g@ 0.75 L r 0.75 L, as applicable, or the measured combined leakage rate for a t {g_allpenetrationsandvalvessubjecttoTypesBandCtestsexceeding0.60L' ' " restore the overall integrated leakage rate to less than 0.75 L a r less than a L, as applicable, and the combined leakage rate for all penetrations subject t to Type B and C tests to less than 0.60 L prior to increasing the Reactor Coolant System temperature above 200'F. a SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972: Three Type A tests (Overall Integrated Containment Leakage Rate) shall a. be conducted at 40 10 month intervals during shutdown at a pressure not less than either P,, 48 psig, or P, 24 psig, during each 10 year t service period. The~ third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspection; WOLF CREEK - UNIT 1 3/4 6-2 ~- -

n, g "). ? .g ^ y w Justification. for specification :3.6.1.2a.2), pg. 3/4 6-2: J This informatiloniwill'be provided 'af ter testing -has been ~ accomplished.. (ILRT is:' currently._ scheduled for Dec. 15. . Data available-by'Dec. 2 5 ~. ) L- / 4

k ~ CONTAINMENT-SYSTEMS CONTAINMENT VENTILATION SYSTEM-LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valves shall be OPERABLE and: ~ ._Each'36-inchcontainmentshutdownpurgesupplyandexhaustisolAtion - a. valve shall be closed and blank flanged, and b. The 18-inch containment mini urge upply and exhaust isolation valve (s) may be open for up 0400 ours during a calendar year. 2000 APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With a' 36-inch containment purge supply and/or exhaust isolation a. valve open or:not blank flanged, close and/or blank flange that valve or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN " within the following 30 hourg. 2000 b. With the 18-inch containment mini pu sup and/or exhaust isolation valve (s) open for more tha curs during a calendar . year, close the open 18-inch valve (s) r isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours, and in COLD SHUT 00WN within the following 30 hours. With a containment purge supply and/or exhaust isolation valvc(s)

c..

having a measured leakage rate in. excess of the limits of Specifications 4.6.1.7.2 and/or 4.6.1.7.4, restore the inoperaole -valve (s) to OPERABLE status within 24 hours, otherwise be in at least HOT STANDBY within the next 6 hours, and in COLD SHUTOOWN within the following 30 hours. I WOLF CREEK - UNIT 1 3/4 6-11

Justification!for Specification.3.6.1.7b. and Action b.: ~KGEE. feels 2000 hours'of-allowedLpurge: time in MODES l Jthrough 4 versus.-the Standard Technical Specification of 500 thours :is justified. based ~ on safety, operational and ' maintenance' considerations. SAFETY. The containment isolation valves in the system were selected, itest'ed, and= located in accordance with the' requirements of 10 .CFR 50~,-Appendix A, General Design. Criteria 54 and 56, and 10 -CFR150,1 Appendix J,1 Type C testing. The actuation! circuitry-for Containment Purge Isolation is Safety. Grade,. Class IE. Isolation is initiated by any Safety Injection Signal, High Containment Airborne Activity, or High y 1 Containment. Purge Exhaust' Activity. OPERATION AND' MAINTENANCE'

KG&E expects that-during. the first year of commercial operation.
the number of pre-planned entries and the number-of unplanned.

-entries into containmentLwould'be higher.than in subsequent ~ years. Good. engineering practice-dictates more-frequent obser- ~vationiand surveillance activities for a new plant. .In addition, the Standard Technical, Specifications mandate special surveil- -lance-of. snubbers'during the first fuel cycle. The " shakedown" phase of the plant will'normally result in more forced outages requiring unplanned entries into containment. The increasedLpurge time would reduce radiation levels within containment and therefore:be consistent with an ALARA philosophy. -CONCLUSION ~The-change poses no' increased threat to the health and safety .of the public and will. allow KG&E to operate the Wolf Creek . Plant within the philosophies of good engineering. practices and ALARA. 1:d

=; A M y .e ^1 qv. TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM PENETRATIONS'_ VALVE NUMBER TYPE LEAK ISOLATION ' FUNCTION TEST REQUIRED (Seconds) -1. Phase "A" Isolation (active) - (Continued) P-32 LF FV-96 CTMT Normal Sumps to' C 4 Floor Drain Tank Out-side.CTMT Iso P-93 SJ HV-5** PZR/RCS Liquid Sample C 5 Inner CTMT Iso P-93 .SJ HV-6** PZR/RCS Liquid Sample C 5 Outer CTMT Iso P-69 SJ HV-12** PZR Vapor Sam le-Cuter C 5 knw CTMT Iso ouw P-69 SJ HV-13** PZR Vapor Samp e-Inner C 5 CTMT Iso 'P-95 SJ HV-18** Accumulator Sample L 5 Inner CTMT Iso P-95 SJ HV-19** Accumulator Sample C 5 Outer CTMT Iso p-93 SJ HV-127** .PZR/RCS Liquid Sampie C 5-Outer CTMT Iso P-64 SJ HV-128** PZR/RCS Liquid A,C 5 Sample Inner CTMT Iso P SJ HV-129** PZR/RCS Liquid A,C 5 Sample Outer CTMT Iso .P-64 SJ HV-130** PZR/RCS Liquid A,C 5 Sample Outer CTMT Iso Valve P-57 SJ HV-131** PASS Discharge to A,C 5 RCOT P-57 SJ HV-132** PASS Discharge to A,C 5 RCDT 2. Phase "A" Isolation (passive)* P-58 EM HV-8888** Accumulator Tank Fill C 5 Line Iso Valve

May be opened on an intermittent basis under administrative control.
- The provisions of Specification 3.0.4 are not applicable.

-WOLF CREEK - UNIT 1 3/4 6-21

REAL M MT ~ TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES c MAXIMUM TYPE LEAK ' ISOLATION i PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds) 6. Remote? Manual - (Continued) P-29 EF HV-48 ESW Return From C N.A. Containment Coolers P-73 EF HV-49 ESW Return From C N.A. Containment Coolers u 'P-29 EF HV-50 ESW Return From C H.A. Containment Coolers .P-74 EG HV-127* CCW Supply to RCP C N.A. P-75 EG HV-130* CCW Return from RCP C N.A. P-75 EG HV-131* CCW Return From RCP C N.A. P EG HV-132* CCW Return From RCP C N.A. Th arriers gben P-76 EG HV-133* CCW from P Thermal C N.A. 3 P-79 EJ HV-8701A RCS Hot Leg 1 to RHR A N.A. Pump A Suction P-52 EJ HV-87018 RCS Hot Leg 4 to RHR A N.A. Pump B Suction P-82 EJ HV-8809A RHR Pump A Cold Leg A N.A. Injection Iso Valve P EJ HV-88098 RHR Pump B Cold Leg A N.A. Injection iso Valve P-15 EJ HV-8811A CTMT Recirc Sump to A N.A. RHR Pump A Suction

These valves.were assumed to be closed during the accident analysis an are normally closed but may be opened on an intermittant basis under a migft tive control.

s WOLF CREEK - UNIT 1 3/4 6-25 -- '~~

FINAL DEFT 4 TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION iI. PENETRATIONS VALVE NUMBER FUNCTION E T REQUIRED (Seconds) 8. Hand-Operated and Check Valves - (Continued) P BG V-135 RCP Seal Water Return C N.A. P BG 8381~ CVCS Charging Line C N.A. P-25 BL 8046 Reactor Makeup Water C N.A. Supply P-78 BM V-045 Steam Generator Drain C N.A. Line Iso Valve P-78 BM V-046 -Steam Generator Drain C N. A'. Line Iso Valve P-53 EC V-083 Refueling Pool Supply C N.A. Frem Fuel Pool Cleanup P-53 EC V-084 Refueling Pool Supply C N.A. From Fuel Pool Cleanup P-54 EC V-087 Refueling Pool C N.A. Return to Fuel Pool Cooling .P-54 EC V-088 Refueling Pool C N.A. Return to Fuel Pool Cooling P-55 EC V-095 Refueling Pool C N.A. Skimmers To Fuel Pool Cooling Loop P-55 EC V-096 Refueling Pool C N.A. Skimmers To Fuel Pool Cooling loop P-74 V-204 CCW Supply to RCP C N.A. P-82 Gfg8818A RHR Pump to Cold A N.A. ) Leg 1 Injection P-82 E F [d 188 RHR Pump to Cold A N.A. Leg 2 Injection WOLF CREEK - UNIT 1 3/4 6-27

FINAL DRAFT TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION 7Ic PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds) 8. Hand-0 arete Land Check Valves - (Continued) x P-27 Ef f f 881'8C RHR Pump to Cold A N.A. ) Leg 3 Injection P-27 Ef ff88180 RHR Pump to Cold A N.A. Leg 4 Injection P-21 8841A RHR Pump Disch to A N.A. RCS Hot Leg 2 P-21 EJ 8841B RHR Pump Disch to A N.A. RCS Hot Leg 3 P-87 EM V-001 SI Pump Hot Leg 1 A N.A. Injection P-87 EM V-002 SI Pump Hot Leg 2 A N.A. Injection l P-48 EM V-003 SI Pump Hot Leg 3 A N.A. Injection P-4" EM V-004 SI Pump Hot Leg 4 A N.A. Injection P-58 EM V-006 Accumulator Fill Line C N.A. From SI Pumps P-49 EM V-010 SI Pump Disch to Cold A N.A. Leg 1 P-49 EM V-020 SI Pump Disch to Cold A N.A. l Leg 2 P-49 EM V-030 SI Pump Disch to Cold A N.A. l Leg 3 P-49 -EM V-040 SI Pump Disch to Cold A N.A. Leg 4 i P-88 EM V-8815 BIT to RCS Cold Leg A N.A. Injection P-89 EN V-013 CTMT Spray Pump A A N.A. l to CTMT Spray Nozzles WOLF CREEK - UNIT 1 3/4 6-28

p i RHi.DB]T .- s TABLE 3.'6-1 (Continued) . CONTAINMENT ISOLATION VALVES L-MAXIMUM TYPE LEAK ISOLATION '! ~ PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)

8.

Hand-Operated and Check Valves - (Continued) 'P-66' 'EN V-017 CTMT Spray Pump 8 A N.A. to CTMT Spray Nozzles P EP.V-046 Accumulator Nitrogen 'C N.A. Supply Line P-43 HD V-016 Auxiliary Steam to C N.A. Decon System P HD V-017 Auxiliary Steam.to C N.A. Decon System P-63 KA V-039 Rx Bldg Service Air C N.A. Supply P-63 KA V-118 Rx B1dg Service Air C N.A. Supply 3 P-98 K3 /-001 Breathing Air Supply C N.A. to RX Bldg P-98 K3 V-002 Breathing Air Supply C' N.A. to RX Bldg P-30 KA V-204 Rx Bldg Instrument C N.A. Air Supply P-67 KC V-478 Fire Protection C N.A. Supply to RX Bldg P-57 SJ V-111 Liquid Simple from A,C N.A. PASS to RCOT 9. Other Automatic Valves P1 A5 nv-il" Fin. 5Lm. Isvi. A 5 J2 A6 nv-14** -t... S tm. Iavi. t a 4 -? '3-P'!- 17 *

l1.,. S im. I2ul.

6

- The provisions of Specification 3.0.4 are not cpplicable.

WOLF CREEK - UNIT 1 3/4 6-29

=- TABLE 3.6-1 (Continued) . CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION T. PENETRATIONS VALVE NUMBER FUNCTION TEST REOUIRED (Seconds) 9. 'Other Automatic Valves (Continued) -r 4 AS "Y-20** Mn. 5 Lm. Isoi. A 5 ^5 AC P! "2** "r.. IJ Isul. A I 'C A: P! ;^** Hn. F4 Isvi. A a-7 AC P! '0** lin. IJ Isvi. A 5 f-S A: Il 41** .in. I,1 Bel A 5 P-9 BM-HV-4** SG Blowdn. Isol. A 10 P-10 BM-HV-1** SG Blowdn. Isol. A 10 P-11 SM-HV-2** SG Blowdn. Isol. A 10 P-12 BM-HV-3** SG Blowdn. Isol. A 10

- The provisions of Specification 3.0.4 are not applicable.

WOLF CREEK - UNIT 1 3/4 6-30

-c .s +- '^ o '~ e g Justification $'or Table 3.6-1.9', cpg.'3/4-6-29 and.3/4'6-30:- f These-.. valves are not. considered containment isolation valves' in the FSAR, pg. 6.2.4-6 and'Section 15.6.3.2. The response ' time-testing-for these valves is contain~ed-in Specification -Table 3.3-5 (steam line ~ isolation testing. and feedwater - isolation. testing). 5 -+ l 4

~ PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d. At least once per 6 months by performance of a yard loop and fire hydrant flush, At least once per 12 months by cycling each testable valve in the e. flow path through at least one ccmplete cycle of full travel, f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and: -1) Verifying thOt C&CIe &UtGiiietk valve in Lite iluw pedi aciuotes tc its correct pcsitim /g) Verifying that each pump .velops at least 3300 gpm' at a system pressure of M-80 p533 2/) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and J/) Verifying th:t :ch fiie suppre>> iva pump sie L5 (acquentially) en dccroesing pie auce in the fir; suppr:::icn header at e- 'headei pie 55 Lie s. eat:r th2." Or OQ"?! t0 M ?N At least once per 3 years by performing a flow test of the system in g. accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association. 4.7.10.1.2 Each fire pump diesel engine shall be demonstrated OPERABLE: At least once per 31 days by verifying: a. 1) The fuel storage tank contains at least 200 gallons of' fuel, and 2) The diesel starts from ambient conditions and operates for at least 30 minutes on recirculation flow. b. At least once per 92 days by verifying that a sample of diesel fuel free the fuel storage tank, obtained in accordance with ASTM-0270-1975, is within the acceptable limits specified in Table 1 of ASTM 0975-1977 when checked for viscosity, water, and sediment; and At least once per 18 months, during shutdown, by subjecting the c. diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service. WOLF CREEK - UNIT 1 3/4 7-28

j; Mffy' ~ f .r P s e 4 4,_. . Insert ' for ; specification.-4.7.10.1.11.3), pg. 3/4 28 : + Verifying thatithe; electric ^ driven fire pump-starts on a start = signal initiated on decreasing header pressure of

75 psig'and-the diesel driven fire' pump starts on a start signal'on. decreasing' header pressure ~of:70 psig after 10 second ; time-delay to avoid simultaneous start of both

~' . pumps.- + i\\ m . e

TABLE 3.7-3 Q]MMgg h. s FIRE HOSE STATIONS d BUILOING ELEVATION AREA HOSE RACK Auxiliary 1974 1122 KC-HR-051 Auxiliary 1974 1122 KC-HR-047 Auxiliary 1974 1120 KC-HR-031 Auxiliary 1974 1120 KC-HR-025# Auxiliary 1974 1101 KC-HR-023# Auxiliary 1974 1101 KC-HR-040 Auxiliary 1974 1101 KC-HR-042 Auxiliary 1988 1201 KC-HR-024 Auxi1iary 2000 1329 KC-HR-111 Auxiliary 2000 1320 KC-HR-048 Auxiliary 2000 1320 KC-HR-046# Auxiliary 2000 1314 KC-HR-030 Auxi1iary 2000 1321 KC-HR-029# Auxiliary 2000 1301 KC-HR-035# Auxiliary 2000 1301 KC-HR-039 Auxi1iary 2000 -1301 KC-HR-041# Auxiliary 2026 1408 KC-HR-049 Auxi1iary 2026 1408 KC-HR-044 Auxiliary 2026 1408 KC-HR-032# Auxiliary 2026 1408 KC-HR-026# Auxiliary 2026 1401 KC-HR-034 Auxiliary 2026 1403 KC-HR-037# Auxiliary 2047 1506 KC-HR-050 Auxiliary 2047 1513 KC-HR-043 Auxiliary 2047 1506 KC-HR-045 Auxiliary 2047 1501 KC-HR-038 Auxi1iary 2047 1504 KC-HR-033 Auxi1iary 2047 1502 KC-HR-027 Auxiliary 2064 1119 KC-HR-028# Control 1974 3101 QCHR-002#b RC-M-00M Control 1974 3101 KC-HR-014# Control 1984 3204 KC-HR-015# Control 1984 3221 KC-HR-001# Control 2000 3301 KC-HR-004# Control 2000 3301 KC-HR-017# Control 2000 3302 KC-HR-016# Control 2016 3401 KC-HR-005 Control 2016 3401 KC-HR-019 Control 2016 3401 KC-HR-018 WOLF CREEK - UNIT 1 3/4 7-34

m ' TABLE 3.7-3'(Continued) FIRE HOSE STATIONS' BUILDING ELEVATION AREA HOSE RACK Control'

2032 3501 KC-HR-006#

Control 2032 3501 KC-HR-020# Control 2047 3604 KC-HR-007 -Control 2047' 3616 KC-HR-021 Control 2073 3801 KC-HR-008# Control.- 2073 3801 KC-HR-022# -Reactor. 2000 2201 KC-HR-120* Reactor 2000 2201 KC-HR-131* Reactor 2000 2201 KC-HR-124* Reactor 2000 2201 KC-HR-Reactor 2026 N.A. KC-HR 21* Reactor 2026 N.A. KC-HR 132* 44-Reactor 2026 N.A. KC-HR-5* Reactor 2026 N.A. KC-HR-130 Reactor 2047 N.A. KC-HR-128* Reactor 2047 N.A. KC-HR-122* -Reactor 2047 N.A. KC-HR-126* Reactor-2068 N.A. KC-HR-123* Reactor 2068 N.A. KC-HR-127* ~ . Fuel-2000 6102 KC-HR-142# Fuel 2000 6102 KC-HR-054# Fuel 2000 6102 KC-HR-143 Fuel 2000 6104 KC-HR-057 Fuel 2026 6201 KC-HR-133 Fuel 2026 6203 KC-HR-052 Fuel 2047 6301 KC-HR-055# Fuel 2047 6302 KC-HR-056# Fuel 2047 6301 KC-HR-053# ESW 2000 N. A. KC-HR-140 ESW 2000-N.A. KC-HR-141 TABLF NOTATIONS

Secondary means of fire suppression to Water Sprays /0eluge or Halon Systems.

Fire hose for station to be stored ' external to Reactor Building.

l WOLF CREEK - UNIT 1 3/4 7-35

e ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank, AO(,o e r6o - Neo vel +.5 4) Verifying the diesel starts from ambient bondition and accelerates to at least 514 rpm in less than or equal to 12 seconds.* The generator voltage and frequency shall b(1000

S and 60 + 1.2 Hz within 12 seconds" after the-start signal.

The diesel generator shall be started for this test by using one of the following signals: a) Manual, or ~ b) Simulated loss-of offsite power by itself, or c) Safety Injection test signal. 5) Verifying the generator is synchronized, loaded to greater than or equal to 6201 kW in less than or eoual to 60 seconds," operates with a load greater than or equal to 6201 kW for at least 60 minutes, and 6) Verifying the diesei gene-ator is alignec te p evice stanc:y - qcws to the associatec emergency cusses. 4 b. At laast enca per 31 days anc after eacn coeration of :ne ciesei wnere the perio'c of operation was greater tnan or ecual :o 1 nour oy checking-for and removing accumulated water from One cay tanks; At least once per 31 days by checking for and removing accumulated c. water from the fuel oil storage tanks; By sampling new fuel oil in accordance with ASTM 04057 prior to d. addition to storage tanks ar.d: (1) By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has: (a) An API Gravity of within 0.3 degrees at 60 F or a specific gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate or an absolute specific gravity at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees;

These diesel generator starts from ambient conditions shall be performed only once per 184 days in these surveillance tests and all other engine starts for the purpose of this surveillance testing shall be preceded by an engine prelube peri.nd and/or other_ warmup procedures recommended'by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.

WOLF CREEK - UNIT 1 3/4 8-3 4 ,._.--m., - - -~ --m,

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) (b) A kinematic viscosity at 40*C of greate'r than or equal to 1.9 centistokes, but less than or equal to 4.1 centistores, if gravity was not determined by comparison with the sup-plier's certification; (c) A flash point equal to or greater than 125 F; and (d) A clear and bright appearance with proper color when tested in accordance with ASTM 04176-82. (2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM 0975-81 are met when tested in accordance with ASTM 3975-81 except that the analysis for sulfur may be performed in accordarce with ASTM 01552-79 or ASTM 02622-82. At least once every 31 days by obtair'ing a sample of fuel oil in e. accordance with ASTM 02276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM 02276-78, Method A. f. At least once per 18 months, during shutdown, by: 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of stardby service, 2) Verifying the diesel generator capability to reject a load of greater t e, cwal t= 352 kW (ESW pump) while maintaining volta. at 1000

20 volt and frequency at 60 + 5.4 Hz,

~ Niko 4-t6o.64 2,0 ~ 3) Verify 1 e diesel ca _ or capability to reject a load of 6201 kW without tripping. The generator voltage shall not exceed 4784 volts during and fol. lowing the load rejection, 4) Simulating a loss-of-offsite power by itself, and: a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected shutdown loads through the shutdown sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and equen he emergency busses shall be maintained t 4000 320 ts and 60 + 1.2 Hz during this te *. q q,,h, 4 WOLF CREEK - UNIT 1 3/4 8-4

ELEt1RICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 5) Verifying that on a Safety Injection test signal without loss-of-offsite power, tha diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes; and the offsite power source ener uto-connected emergency (accident) load through The generator voltage and frequency shall b ',000he LOCA seque r. 320 volt and 60 1 1.2 Hz within 12 seconds after the I auto-art signal; the generator steady-state generator voltag be maintained within these limits during thi and.requency shal te t; u.o+ mo-va - q 6) Simulating a loss-of-offsite power in conjunction Saf Injection test signal, and a) Verifying deenergization of the emergency busses and load shedding from the emergency busses; b) Verifying the diesel starts on the auto start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected emergency (accident) loads through the LOCA sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with emergency loads. After energization, the steady-state voltage and quency % the emergency busses shall be maintained and 60 1 1.2 Hz during this test; and t 1000 1 220 v lts % m.o -uso c) Verifying that all automatic diesel generator trips, except i high jacket coolant temperature, engine overspeed, low lube oil pressure, high crankcase pressure, start failure relay, and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a Safety Injection Actuation signal. 7) Verifying the diesel generator operates for at least 24 hours. During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 6821 kW and during the remaining 22 hours of this tes diesel generator shall be loaded to greater than or e al to 62 kW. The generator voltage and frequency shall beH000 _ 220. Its and 60 + 1.2 Hz, -3Hzwithin12secondsafter'the[startsinal;thesteady-dat; ;;^ or voltage and fre quency shall e maintained within C 4st.o m,o-vs. -1000 324 vo ts and 60 2 1.2 F z during this t st. Within

N tc; si er completing thit tion 4.8.1.1.2f.6)b)*;

24ghourtest,prformSpecifica-g, g o y,, J

If Specification 4.8.1.1.2f.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test.

Instead, the diesel generator may be operated at 6201 kW for 1 hour or until operating temperature has stabilized. WOLF CREEK - UNIT 1 3/4 8-5

' L Y. _ o b k ' Justification for specification 4.8.1.1.2.a.-4), f. 2), ' f. ' 4)l b), f. 5),; f. 6) b), and f. 7). pg. 3/4:8-3,4, and 5:. Shifted center' point but maintained span ~at same points. Change needed to ensure.compatability.with specifications .3.8.3.1 and;3.8.3.2..Also the. generator is a 4160 volt. - machine and the regulator is set to regulate at 4160 volts. m

I l FIML DNFT ELECTRICAL POWER SYSTEMS A.C. SOURCES' SHUTOOWN LIMITING CONDITION FOR OPERATION

3. 8.1. 2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

.One circuit between the offsite transmission network and the Onsite a. Class 1E Distribution System, and b. One diesel generator with: 1) A day tank containing a minimum volume of 390 gallons of fuel, 2) A fuel storage system containing a minimum volume of 85,300 gallons of fuel, and 3) A fuel transfer pump. APPLIC.",3ILITY: MODES 5 and 6. ACTION: e ~ With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the spent fuel pool. =d ::f th'- 9 5:=, epr;;;urize and vant the hnter Ces'e.d 0.1;te.a thingh ;t k r t : 2 ;que.i i..cr..ent. In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible. SURVEILLANCE REGUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.5)), and 4.8.1.1.3. WOLF CREEK - UNIT 1 3/4 8-R

.a; 4 5 4: 4 t hic '. t y.. l3. u.. ' Justification for.-specification-~3.8.1.2' Action, pg.7.3/4 8-8: ,I' 1 x I LMi =.2 This should have.been deleted:.whenLthe-specitications were- ' modified fo'r use of'RHR cold overpressure protection

f

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P

k 4 g 4 i l[* 3 b g: S-f z; c t N 4 I'. $ ,1- . s 9 n 8 p n. .n.]- l h, '. j

g. '

" ^ .,+.,,.a,-

ELECTRICAL POWER SYSTEMS 3/4.8.2- 0.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION e 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE: a. 125-Volt Battery Bank Chargers NK21 and NK2, g e cj 13, and its associated Full Capacity an b. 125-Volt Battery Bank NK12 and NK14, and its associated Full Capacity Chargers NK22 and NK24. APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: With one of the required battery banks and/or ful' capacity chargers inoperable, restore the inoperable battery bank and/or full capacity charger to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ~ a SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLI: At least once per 7 days by verifying that: a. 1) The parameters in Table 4.8-2 meet the Category A limits, and 2) The total battery terminal voltage is greater than or equal to 130.2 volts on float charge. WOLF CREEK - UNIT 1 3/4 8-9

TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A(1) CATEGORY 8(2) PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) DESIGNATED PILOY CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL l Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark,

plates, and < " above and < %" above and not maximum level maximum level overflowing indication mark indication mark Float Voltage

> 2.13 volts > 2.13 volts (6) > 2.07 volts Not more than 0.020 below the j average of all > 1.195 connected cells Specifi Gravity {4) 1 1.200 Average or all Average of all connected cells -> 1.195' ) cells connected > 1.205 _1 TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category 8 measurements are taken and found to be within their allow ele values, and prov!ded all Category A and 8 parameter (s) are restored to within limits within the next 6 days. (2) Foe any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B parameters are within their allowable values and provided the Category 8 parameter (s) are restored to within limits within 7 days. (3) Any Category B parameter not within its allowable value indicates an inoperable battery. (4) Corrected for electrolyte temperature and level. (5) Or battery charging current is less than 2 amps when on charge. (6) Corrected for average electrolyte temperature. WOLF CREEK - UNIT 1 3/4 8-11

RNAL DRAFT TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EOUIPMENT Low Voltage Power and Control (Continued) 6EPK05E P-3A Fuse-Accumulator W ter f@l1 Viv RLO18 B-3A Fuse-EPHV88789 6EPK05F P-3A Fuse Accumulator Water Fill V1v RLO18 B-3A Fuse EPHV88789 P-4SJY018-3A Fase RLO11 Press. Vapor. Cont. Iso. Space Viv. SJHV12 B-4RLYO1G 15A Breaker NG02ACR140 P-4SJY01C 3A Fuse Accums Sample Cont Isol Viv RLC11 SJHV18 B-4RLY01G 15A Breaker NG02TCR140 ?-53JYO3B 3A Fuse Accumulator Sample Line.Viv RP211 SJHV16 B-5RPYO90 15A Breaker PG19NHF236 P-5SJYO3C 3A Fuse Accumulator Sample Line Viv RP211 SJHV17 B-5RPY09D 15A Breaker PG19NHF236 P-5SJYO4B 3A Fuse Accumulator Sample Line Viv -RP211 SJHV14 B-5 90 15A Breaker 19NHF2 P-5SJ[4C3AFuse Accumulator Samole Line Viv P211 SJHV15 B snIY000 d Breaker PG19NHF236-P-ISJYO6B 3A Fuse HL Sample 3 Viv RP332 SJHV4 B-1RPYO9F 15A Breaker NG018AR140 P-4SJYO6A 3A Fuse HL Samole 1 Viv RP333 SJHV3 B-4RPYO9F 15A Breaker NG028AR140 WOLF CREEK - UNIT 1 3/4 8-39

FINE DEFT TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE NUMBER AND LOCATION POWERED EQUIPMENT ., Low Voltage Power and Control (Continued) P-SSJYOGC 3A Fuse RP211 Press Liquid Space Samp Isol Viv B-5RPYO90 15A Breaker SJHV20 PG19NHF236 P-4BMY01A 3A Fuse RLO24 S.G. A Out to Nuc Sample Sys Viv B-4RLYO1H ISA Breaker BMHV19 NG02ACR127 P-4BMY01B 3A Fuse 4 S.G. B Out to Nuc Sample Sys Viv BMHV20 B-4R 1H 15A Breaker N ACR1 P-4BM C 3A Fuse S.G. C Out to Nuc Sample Sys Viv B-4RLYO1H 15A Breaker BMHV21 NG02ACR127 P-5GNY08A 3A Fuse RLO20 CRDM Cooling Discharge Daaper B-5RLYO1L 15A Breaker GNHZ71 PG19GCR2:0 P-5GNY08C 3A Fuse RLO20 CRDM Cooling Discharge Damper B-5RLY01L ISA Breaker GNHZ73 PG19GCR230 P-6GNY08A 3A Fuse RL020 CRDM Cooling Discharge Damper t GMK 72 B-6RLYO1J 15A Breaker PG20GBR222 P-6GNY08C 3A Fuse RLO20 CRDM Cooling Discharge Damper GNHZ74 B-6RLY01J ISA Breaker PG20GBR222 5BGK10A P-3A Fuse RL001 B-3A Fuse Normal Letdown Isolation V1v BGLCV459 WOLF CREEK - UNIT 1 3/4 8-40

FINAL DRAFT REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continu_ed) 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regula ide 1.52, Revisison 2, March 1978, for a methyl odij(epenetrationof less than 1%; and d 3) Verifying a system flow rate of 9000 cfm 10% during system operation when tested in accordance with ANSI N510-1975. After every 720 hours of charcoal adsorber operation, by verifying, c. within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a l of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%; d. At least once per 18 months by: 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5.4 inches Water Gauge while operating the system at a flow rate of 9000 cfm 10%. 2) Verifying that on a Spent Fuel Pool Gaseous Radioactivity-High test signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks and isolates the normal fuel building exhaust flow to the auxiliary / fuel building exhaust fan; ~ 3) Verifying that the system maintains the Fuel Building at a negative pressure of greater than or equal to 1/4 inches Water Gauge relative to the outside atmosphere during system operation; and 4) Verifying that the heaters dissipate 37 3 kW when tested in accordance with ANSI N510-1975. After each complete or partial replacement of a HEPA filter bank, by e. verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1975 for a 00P test aerosol while operating the system at a flow rate of 9000 cfm i 10%; and f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant test gas while operating the system at a i flow rate of 9000 cfm i 10L WOLF CREEK - UNIT 1 3/4 9-18 _ _ ~

D FINA!. MAFT SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 'The limitations of the following requirements may be suspended:

Specification 3.4.1.1 - During the performance of startup and a. PHYSICS TESTS in MODE 1 or 2 provided: 1) The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and 2) The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER. . b. Specification 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least three reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE. APPLIp N LITY: During operation below the P-7 Interlock Setpoint or performance of not roo drop time measurements, ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint during a. the performance.of startup and PHYSICS TESTS, immediately open the Reactor trip breakers. b. With less than the above required reactor coolant loops OPERABLE during performance of hot rod drop time measurements, immediately place two reactor coolant loops in operation. SURVEILLANCE REOUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlo Setpoint at least once per hour during startup and PHYSICS TESTS. 4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating startup and PHYSICS TESTS. 4.10.4.3 At least the above required reactor coolant loops shall be determined OPERABLE within 4 hours initiation of the hot, rod drop time measure-ior ments and at least onc per 4 ho s during the hot rod drop time measurements by verifying correct b eakerjlalig ments and indicated power availability. . WOLF CREEK - UNIT 1 3/4 10-4

FINA!. DRAFT TABLE 4.11-1 (Continued) TABLE NOTATIONS (1)The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system, which may include radiochemical separation: LLD = b E V 2.22 x 105 Y exp (-Aat) Where: LLD = the "a priori" lower limit of detection (microcuries per unit mass or volume), b = the standard deviation of the background counting rate or of s tne counting rate of a~blan.k sample as appropriate (counts per ~ O E = the counting efficiency (counts per disintegration), V = the sample size (units of mass or volume), 2.22 x 108 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (s 1), and at = the elapsed time between the midpoint of sample collection and the time of counting (s). I Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. (2)A batch release is the discharge of liquid wastes of a discrete volume. i Prior to sampling for analyses, each batch be i Ks and then thoroughly mixed by a method describe in t.t CCCM to assure representative sampiing. gp g,Le 3 WOLF CREEK - UNIT 1 3/4 11-3

p' ' ( - 1 I ' Justificationfor: Table 4.11-1 Footnote (2), pg. 3/4: 11-3 : ' P'lant. procedures for the radioactive-release of liquids have been written.- These procedures-include the methods- ~ ~ for mixing:to.' assure. representative sampling.. The mixing; methodology is not in the ODCM. m s g. h e +-w-w,...r -w.-+.mree-n, , c w,-- m 3.. ers w w., =--,y,- ,,---m-, ,mpwa,,w ,-w,y.,,,--,-,rw--eey-,w--,.,9,-e----wt--*w-1--*-- ww-----

TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM S i h MINIMUM LOWER LIMIT OF SAMPLING ANALYSIS TYPE OF DETECTION (LLD)(3) 4 c GASEOUS, RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pC1/ml) 6 P P [ 1. Waste Gas Decay Each Tank Each Tank Principal Gamma Emitters (2) 1x10~4 Tank Grab Sample 2. Containment Purge Each PURGE (3) Each PURGE (3) Principal Gamma Emitters (2) 1x10~4 or Vent Grab Sample O M H-3 (oxide) 1x10 -6 3. Unit Vent

3) I I3 M

Principal Gamma Emitters (2) 1x10'4 Sample MD I -6 H-3 (oxide) 1x10 g ((SpetFu ) [M[/((,Mrin[pa1[mma[ftte/(2)p Bui di M / 4 1 [ / /g(S)/ / [ / n-[(odde) [ [ [ dx10~I / / adwaste Building M ~4 1 Grab Samp Principal Gamma Emitters (2) 1x10 M [ All R h s (in 1. Types Continuois(6)% (7) g 1-131 -12 as listed 1x10 4 p Charcoal Sample I-133 1x10 10 g 2., 3., 4. an Continu(us(6)f8I (7)

5. above Principal Gamma Emitters (2) 1x10 l mens

-11 O I> articulate I Y /*~ Sample M Continu'us(6)% ~Il g% Gross Alpha 1x10 Composi e 4 / Particulate Sample Continuou Q Sr-89, Sr-90 1x10-II Composite Particulate Sample

a TABLE 4.11-2 (Continued) TABLE NOTATIONS (Continued) (2)The principal gamma emitters for which the LLO specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.7, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. (3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within 1 hour period. (4) Tritium grab samples shall be taken and analyzed at least once per 24 hours when the refueling canal is flooded. (5) Tritium grab samples shall be taken and analyzed at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool. Grab samples need to be taken only when spent fuel is in the spent fuel pool. (6)The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3. (7) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing, or after removal from sampler. For unit vent, sampling shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period, and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analyzed, the corresponding LL0s may be increased by a factor of 10.- This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT.I-131 concentration in the reactor coolant has not increased more than a factor of 3, and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3. (8)Continous sampling of the spent fuel building exhaust needs to be performed only when spent fuel is in the spent fuel pool. L WOLF CREEK - UNIT 1 3/4 11-11

] p} ~ ) e / mJustification.fo'r Table.4.11 < c -The Spentcfuel' Building vent is not an effluent releas'e -point.- The, Spent Fuel Building vent'; exhausts into the'

Auxiliary Building Ventilation System which exhausts in: turn into the. Plant Unit-vent.

The-site. release point there-

fore-is the Unit vent.- The Unit vent is-~therefore-the-point

'where:the methodology of the ODCM.is applied to< assure.that" 'the dose. limits-for-Technical Specification 3.11.2.1 are met.

Since Table 4.ll-2Jis designed to aid meeting the criteria z

of. Technical.. Specification 3.~11.~2.1, the Spent Fuel' Building. i eve't:should be deleted. <The tritium' requirement note for n the Spent Fuel ~ vent can'therefore be moved to the unit vent.. + m w + f x e,,, e-- c.,;--, ---.e,--w

.e,

, ---w-wg ,*.yv-, ,,,v im,- g+~,- -y,-ry----v-q, ,-.y,9-m.--,,y - - - - - - +, -, -, y-,i-,.- - -,,,y

RADI0 ACTIVE EFFLUENTS GAS STORAGErTANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal t 2 x 105 Curies of noble gases. (con-sidered as Xe-133 equivalent).

3.3 APPLICABILITY

At all times. ACT10N: With the quantity of radioactive material in any gas storage tank a. exceeding the above limit, immediately suspend all additions of radioactive material to the tank and, within 48 hours, reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent F.elease Report, pursuant to Specificatien 6. 9.1. 7. The pr:.isic.s of Spacificaticcs 3.0.3 and 3.5.4 are not aoplicable. b. t SURVEILLAtiCE REOUIREMENTS 4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 7 days when radioactive materials are being added and within 7 days following any addition of radioactive material to the tank. WOLF CREEK - UNIT 1 3/4 11-16

~ gv= t l e l r 4 w a -.' Justification ~for specification 3.11.2.6, pg. 3/4 11-16:. The metieorologicalJconditions for Wolf Creek. allow this -!va ue to be -increased while-still meeting boundary require-l ments.: h. m. e c =_( s C ""'41 a.s Ww y ywpg g

s,.... 1 f c,. '_RADI0 ACTIVE EFFLUENTS hh 3/4.11.4 TOTAL DOSE _ LIMITING CONDITION FOR OPERATION 13.11.4 ~ The annual (calendar year) dose or dose commitment to any MEMBER OF \\ -.THE'PUBLIC due to releases of radioactivity and to radiation from uranium fuel { cycle socrces shall be-limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less eaan or-equal to.75 mrems. APPLICABILITY: At all times. . ACTION: -With the calculated doses from the release of radioactive materials a. -in liquid or gaseous effluents exceeding twice the limits of Speci fication 3.11.1.2a., 3.11.1. 2b., 3.11. 2.2a., 3.11. 2. 2b., 3.11.2.3a., or 3.11.2.3b., calculations should be made including direct radiation contributions from the units and.from outside storage tanks to determine wh' ether the above limits of Specification 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, 'a Special Report that ^, defines the corrective action to be taken to reduce subsequent releases to prevent recurreate of exceeding the above limits and includes the LC.

chedule for achieving conformance with.the above limits.

This Special . Report, as defined in 10 CFR 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radicactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not aireacy been corrected, the Special Report shall' include a requ6est for a variance in accordance f with the provisions of 40 CFR Part 190. M Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ~ 5 RVEILLANCE REGUIREMENTS ~ 4.11.4.1-Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the 00CM. 4.11.4.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the 00CM. This requirement is applicable only .{ uncer conditions set forth in ACTION a. of Specification 3.11.4. '.~ WOLF CREEK - UNIT 1 3/4 11-18 I ?..

r-- + ' Justification for specification 3*11 4< Pg. 3/4 11-18: This page was missing. i

3/4.12 ' RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The Radiological Environmental Monitoring Program shall be conducted as specified in Table 3.12-1. APPLICABILITY: At all times. ACTION: With the Radiological Environmental Monitoring Program not being a. conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.6, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose

to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3.

When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if: concentration (1) concentration (2) ... > 1. 0 reporting level (1) reporting level (2) When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose

to A MEMBER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 or 3.11.2.3.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.6. With th; eveileb-;1ity of e;1k er frc:P leafy vegetoble semples frem c. cac sc mer; cf th c=pla lacations required by T;ble 3.12-1 net-Ileelw M peac4.4 cable e possib!:, idcatify specific locations for obtaining T9 sch peptatemeni :,amplos and add ther with" 30 days te the Rediclegic&1 Envirensentsi N uiiuc;ng Prcgrc given in the OGE&

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

WOLF CREEK - UNIT 1 3/4 12-1

FiNA! DWA N RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION ACTION (Continued) l D^ epa'4fic l^c2ti0n~, fr0;..hich 022",le3^6i6 Un&v6ilaule suay Lii4G bG-- ) deleted frat. iiia inuiii iu r i ng pi eg e.T.. I'Ursu;nt te Speci'_ icatica 0.14, S,_,_u m.. -o n_,_ ___m ._. in une neAL aemlannuas nausuauwave cansusemw nuicuac nspvi w dOOu;ehleLiun iur a c.nange isi Lisu OCLM M ludi"g 2 r0'tiOCd figurC(3}

m..a_.. u s.._

aus wuc vuun re s i cs w a i.3

u.. _ _ _.. 3..,.

4.. o. r,, s,

. 4.. u. auppurwsuQ w ..s- .s. irf0N.etien identify-ing Lisc can3G Of th0 2"I'!2il3Di'ity Of 30;pl00-und 'u:ti#ying the selectica of the n locatien(:) for obtaining 00;pl65. d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the taole and figure (s) in the ODCM, and shall be analy:ed cursuant to the recuirerents of Table 3.,12-1 and the cetection e:pacilities recuired by Table 4.12-1. i i l l l WOLF CREEK - UNIT 1 3/4 12-2

y-X A 4 g InsertIfor' specification 3.12.1, Action c.,s pg. 3/4 12-1,2: - With1 milk or~ fresh. leafy vegetable. samples temporarily.-una-L vailable from:a ' routine' sampling location, a : sample from an

alternative location..(identified in the ODCM) will be substi-

.tuted,Lnoting the reason-for'the unavailability in the, Annual ' Radiological. Environmental Operating Report. When: changes in ' sampling? ocations are permanent, the sampling schedule in. l ithe'ODCM willebe updated to reflect the.new routine and alter-native: sampling locations,.and.this revision will be^ described-c 'in;th'e--Annual Radiological Environmental Operating Report. r L I g e w $s' g r mem$ e w,os -.,w-ysq ..y_im,-g9 M y9 9 yg i --.g----m g--w

p r .s ,'y[,

t'.

3ustification for' Specification-3.12.1 Action'(c), pg.. 3/41.12-1,2: The change -in ' Action' (c) is' requested to make it' clear that-temporary sample unavailability at a. location (which preoperational experience' hasishown - to, occur) does not require revision' of the program and the associated paperwork..which would. occur for actually deleting: and/or - adding a sampling ~ location. f a 1 4 u w

'M mamme TABLE 3.12-1 (Continued) E 6 q; RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM f k NUMBER OF p REPRESENTATIVE g EXPOSURE PAlllWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/0R SAMPLE SAMPLE LOCATIONS (1) COLLECTION FREQUENCY OF ANALYSIS M

j

2. Airborne M

H Radiolodine and Samples from five locations Continuous sampler Radiofodine Cannister: Particulates operation with sample I-131 analysis weekly. Three samples from close to collection weekly, or the three SITE BOUNDARY locations, more frequently if in different sectors, of the required by dust Particulate Sampler: highest calculated annual average loading. Gross beta radioactivity ground level D/Q. analysis following filter change;(4) and R One sample from the vicinity gamma isotopic analysis (5) of a community having the highest of composite (by g calculated annual average ground-location) quarterly. 4 level 0/Q. One sample from a control loca-tion, as for example 15 to 30 km (10 to 20 mile) distant and in g least prevalent wind direction.

3. Waterbor

% sqk

a. Surfa e One sample upstream. (O Gratr-samp!e ev--

Gamma isotopic analysis and sample downstream. I mv.iu pa icd-monthly. Composite for tritium analysis quarterly.

b. Ground Samples from one or two source Quarterly.

Gamma isotopic (5) and tritium onlyiflikelytobeaffected.[8) analysis quarterly. I

c. Drinking One sample of each of one to Composite sample I-131 analysis on each g

three of the nearest water over 2-week period composite when the dose supplies that could be when I-131 analysis calculated for the consump-affected by its discharge. is performed, monthly tion of the water is composite otherwise. than 1 mrem per year.gater Com-One sample from a control positeforgrossbetaag) location. gamma isotopic analyses monthly. Composite for ,, i t 4,,,,, n, n., 4, ...,.+.a..

1 -.4 Justification,for Table 3.12-1, 3.a.; pg. 3/4 12-4: TheLchange in wording and removal of note "7" is requested

because the volume of.the downstream' sampling point, Wolf

Creek cooling lake (3.6 x E +10 gallons), is-large'enough that concentrations of released effluents would fluctuate

-very: gradually with time, making short term or-composite-sampling unnecessary for.this location. f

FINAL DRAF TABLE 3.12-1 (Continued) TABLE NOTATIONS (Continued) (5) Gamma isotopic analysis means the identification and quantification of ga emitting radionuclides that may be attri'autable to the effluents from the facility. (6)The " upstream sample" shall be taken at a distance beyond significant influen of the discharge. but near the mixing zone.The " downstream" sample shall be taken in an area be g (7)A composite sample is one in which the quantity (aliquot) of is constant over the sampling period and in which the met od of saiquidspled employed results in a specimen that is representative of t ing +n k p reta. id concen-In this program composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample. i (8) Groundwater samples shall be taken when this source is tapped for drinki or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination. (9)The dose shall be calculated for the maximum organ and age group, using t methodology and parameters in the 00CM. (10)If harvest occurs more than once a year, sampling shall be cerformed durin each diccrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products. WOLF CREEK - UNIT 1 3/4 12-8

P.-- 1 4 b: a u; up - Justification 'for Table '3.12-1, Table notation _ (7), ' pg. .3/4 12-8:- ^ = Change is._requestedito-correct-a typographical error.and-make.the wording more_. accurate. = a

j 7 L ?- y ,q , RADIOLOGICAL ENVIRONMENTAL MONITORING Yd i 1 J ~ 4 3 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION -3.12.3 Analyses shall-be performed on radioactive materials supp d as part of an Interlaboratory--Comparison-Program that-has-been-approved _bjf the ~ Commission /{h+ ccernp M +s wp cs re uir.d i g by %W. 3.12. APPLICABILITY: At all times. ACTION: With analyses not being performed as required above, report the a. corrective actions taken to prevent.a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. b. The provisions of Spe'cifications 3.0.3 and 3.0.4 are not applicable. .r SURVEILLANCE P.EOUIREMENTS C '4.12.3 The Interlaboratory Comparison Program shall be described in the 00CM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. i e i _ _ ~. i ' f.~. e - '~ O WOLF CREEK - UNIT 1 3/4 12-14 ~

r I c. l 4 Justification for Specification 3/4.12.3, pg. 3/4 12-14: This page was missing from the final draft. i 4 2

e i FINAL DRAFT + POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE. HOT CHANNEL FACTOR (Continued)- l_ The control rod insertion limits of Specification 3.1.3.6 are i c. j maintained, and l d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. F will be maintained within its limits provided Conditions a. through H

d. above are maintained.

As noted on Figure 3.2-3, RCS flow rate and F l. may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F is also low) to ensure that the calcu g lated DNBR will not be below the design DNBR value. The relaxation of F as N i a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insert ~ ion limits. l spes$ ch ^ R as calc lated in 3.2.3 and use in Figure 3.2-3, accounts for FN less I g than or equal to. O. Thi is used in the various accident analyses whereFhinfluencesparametersotherthanDNBR,e.g.,peakcladtemperature, and thus is the maximum "as measured" value allowed. Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margins, totaling i- -l 9.1% DNBR, completely offset any rod bow penalties. This margin includes the i following: t-1) Design limit DNBR of 1.30 vs. 1.28, 2) Grid spacing (K,) of 0.046 vs. 0.059, 3) Thermal Diffusion Coefficent of 0.038 vs. 0.059, { 4) DNBR Multiplier of 0.86 vs. 0.88, and i 5) Pitch Reduction. _l The applicable values of rod bow penalties are referenced in the FSAR. When an F measurement is taken, an allowance for both experimental error q and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. WOLF CREEK - UNIT 1 B 3/4 2-4

FINAL DRAFT POWER OISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR'ENTHA HOT CHANNEL FACTOR.(Continued)- The Radial-Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (z), remains within its limit. RTP q The F limit for RATED THERMAL POWER (Fxy ) as provided in the Radial Peaking xy Factor Limit Report per Specification 6.9.1.9 was determined from expected power control manuevers over the full range of burnup conditions-in the core. WhenRCS'flowrateandFharemeasured,noadditionalallowancesare necessary prior to comparison with the limits of Figure 3.2-3. Measurement errors of 2.1% for RCS total flow rate and 4% for F have been allowed for in H determination of the design DNBR value. measureme ror for al flow r based up p forming pr isio heat b ance and usin the r sult to calibt.te the CS flo rate M 6 i icator s. Po ntial fouling of the eedwate venture which might n t be d tected bould ias the result from the precis'on heat balanc in a n n-c nservative m nner. Tterefo e, an ins ectio is perf rmed f the fee ate enture e ch efueling ta .The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3. 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The limit of 1.02, at which corrective ACTION is required, provides ONB and linear heat generation rate protection with x y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. The 2-hour time allowance for operation with a ti.lt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod. In the event such ACTION does not correct the tilt, the margin for uncertainty on F is reinstated by reducing q the maximum allowed power by 3% for each percent of tilt in excess of 1. WOLF CREEK - UNIT 1 B 3/4 2-5

~ O <@W W Justification for specification B 3/4.2.2 and B 3/4.2.3 Specification'4.2.3.6 has been requested to be deleted. A 4 --mm-m

FINAL. DRAFT INSTRUMENTATION BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumeatation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974. 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and cafety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onrite Meteorological Programs," February 1972. 3/4.3.7.5 REMOTE SHUTDOWN INSTRUMENTATION STHUD8V The OPERABILITY of the Remote Shutdown Systern asures that suffic g ent capability is available to permit shutdown and maintenance of H C::UTLCM of the facility from locations outside of the control room and that a fisc wi not preclude achieving safe shutdown. The Remote Shutdown System transfer witches, pcwer circuits, and control circuits are independent of areas where a fire could damage systems normally used to shut down the reactor. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 3 and 19 and Appendix R of 10 CFR Part 50. 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 2. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection System ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant control Room Operators Against an Accidental Chlorine Release," Revision 1, January 1977. WOLF CREEK - UNIT 1 B 3/4 3-4

'e. FINAL DRAFT . INSTRUMENTATION ~ BASES -3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that both ade-quate warning capability is available for the prompt detection of fires and that Fire Suppression Systems,' that are actuated by fire detectors, will~ discharge extinguishing agents in a timely manner. Prompt detection and suppression of ~ fires will. reduce the potential for damage to safety-related equipment and is an integral element in the overall facility. fire protection program. Fire detectors that are used to actuate fire suppression systems represent a more critically important com;onent of a plant's fire protection program than detectors that are installed solely for early fire warning and notification. Consequently, the minimum-number of. operable fire detectors must be greater. The loss of detection capability for Fire Suppression Systems, actuated oy fire detectors represents. a significant degradation of fire protection for any As a result, the establishment of a fire watch patrol must be initiated area. at an earlier stage than would be warranted for the loss of detectors that provide only early fire warning. The establishment of frequent fue patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.3.9 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose part detection instrumentation ensures that sufficient capability is available'to detect loose metallic parts in the ' Reactor Coolant System and avoid or mitigate damage to Reactor Coolant System components. The allowable out-of service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981. 3/4.3.3.10 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION The' radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / -Trip Setpoints for these instruments shall be calculated and adusted in -accordance wtth the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The .0PERABILITY and use of this instrumentat' n ' consistent with the requirements of General Design ~ Criteria 60, 63, and 4gA endix A to 10 CFR Part 50. oC WOLF CREEK - UNIT 1 8 3/4 3-5 . ~.

INSTRUMENTATION g BASES 3/4.3.3.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be adjusted to values calculated in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTE GA$ HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appencix A to 10 CFR Part 50. The sensitivity of any noble gas activity m i tor used to show compli-ance with the gaseous effluent release requirement of Sp cification 3.11.2.2 shallbesuchthatconcentrationsaslowas1x1-6/Ci/caremeasurable. 3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is providad to ensure that the turbine overspeed pro-tec icn instrumentation anc i:'e tu-oine speed control valves are OPERABLE and will crotect tne turbine from excessive overspeed. Although the orientation .of the turcine is such that the number of potentially damaging missiles which could impact and damage safety-related components, equipment, or structures is minimal, protection from excessive turbine overspeed is required. WOLF CREEK - UNIT 1 8 3/4 3-6 Y ~

FINAL DRAFT REACTOR COOLANT SYSTEM BASES .FRESSURE/ TEMPERATURE LIMITS (Continued) Where: KIR is the reference stress intensity factor as a function of the metal temperature :T and the metal nil ductility reference temperature RT NOT.

Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

CK73 + kit 1KIR (2) Where: K73 = the stress intensity factor caused by membrane (pressure)

stress, KIt ress intensity factor caused by the thermal gradients,

= K7g = unction of temperatura relative to the RT f the material provided by the Code, 'NDT C = 2.0 for level A and 'B service -limits, and C = 1.5 for inservice hydros'tatic and leak test coerations. , At any time during the heatup or cooldown ' transient,'Kg is ceterEined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and *. hen the corresponding thermal stress intensity factor, kit, f r the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated. COOLDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the.inside of the wall because the thermal gradients proo:uce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temper ature relations are generated for both steady-state and finite cooldown rate ituations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use'of the composite curve in the cooldown analysis is,necessary because control.of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent tc *.5e vessel 10. This condition, of course, is not true for the steaoy-state si*;Uation. It follows that at any given reactor coolant temperature, the AT developed WOLF CREEK - UNIT 1 8 3/4 4-9

N 3/4.6 CONTAINMENT SYSTEMS s BASES _3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. restriction, in conjunction with the leakage rate limitation, will limit the This SITE B0UPDARY radiation doses to within-the dose guideline values of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L r 0.75 L, as applicable, during performance of the periodic a t test to account for possible degradation of the containment leakage barriers between leakage tests. $ustry 7 The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air lock are required to meet the restrictions ;n CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals pro / ides assurance that during t'le intervals between air lock leakage tests.the overall air WOLF CREEK - UNIT 1 B 3/4 6-1 I mit m - _______,.._._,_ _ ___...____m

[,[f 7, 2 :., q ^ I s ..s. T h-( l 1 i i ' Y

Insert for' specification l 3/4.621.2,!pg. B~;3/4 6-1:

V t ' For r' educed ; pressure '. tests, the; leakage-characteristics

yielded by measurements.L m and Lam shall establish the t

i t . max mumcallowable test leakage rate.L ' Of. not.more than - La.(L m/ Lam) '. In c.the : event L m/ Lam is greater than 0.7, - ~ t t (P /P,)1/2, -Lt shall -be. specified asi equal to' La t "a i.L, -g. 3, b' N i s k ? -e h 4 g .2 +. h, J A 6 d e 1 1h + 4 + v w 7 m ~ Ra O ( .5+., .,4 -,---..-,.,-,.,-reg ,,.,---y-+ -,,-,w,,my-r- - = = ~ - im -r- --n v.--r- ,w s--- er.e -. +

p a _.,. m q..... s s s.- i FINAL DRAFT . CONTAINMENT: SYSTEMS ... BASES' 3/4.6.1.'4 INTERNAL PRESSURE' The limitations on containment internal pressure ensure that: (1) the ~ containment. structure is' prevented from exceeding its design negative pressure ' differential'with respect to the outside atmosphere of 3.0 psig, and (2) the

containment peak pressure does' not exceed the design pressure of-60 psig

~during steam line break conditions. The' maximum peak pressure expected to be obtained from a steam l'ine break eventais 48 psig. The limit of 1.5 psig for initial positive containment pressure will limit tha. total pressure to 43.5 psig, which is less than design pressure-and is consistent with the safety analyses. 3/4.6.1.5 ' AIR TEMPERATURE The' limitations on containment average air temperature ensure that the overall containment-averace air temoerature does not exceed tne initial .tamcerature condition assiced'in the safety analysis for a shm line ~ 4 creas accicant. Measurements shal1J e mace at all listec locations, whether by fixed or portaMe instruments, pFior to cetermining the average air. --temperature. 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY .This limitation ensures that the structural. integrity of.the containment

will be maintained in accordance with safety analysis requirements for the life of the facility.

Structural integrity is required to ensure that the contain-ment will withstand the maximum pressure of 50 psig in the event of a steam line break accident. The measurement of containment tendon lift off force, the tensile' tests of the tendon wires or. strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the contain-ment, and the Type A leakage test are sufficient to demonstrate this capability. The Surveillance Requirements for demonstrating the containe structural integrity are in compliance.with the reccmmendations f f r-~ :d t l Regulatory Guide.1.35, " Inservice Surveillance of Ungrouted Tend us in Fre-stressed Concrete Containment Structures," April 1979, and prop ~osed Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of Prestressed . Concrete Containments," April 1979. lThe requir?d Special Reports from any engine 9 ring evaluation of containment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure, the tolerance 'ct, cracking, the results of the engineering evaluation and the -corrective actions taken. WOLF CREEK - UNIT 1 8 3/4 6-2

3 e FINAL DmumW CONTAINMENT SYSTEMS 8ASES 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 36-inch containment purge supply and exhaust isolation valves are required to be closed and blank flanged during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves closed and blank flanged during plant operation ensures that excessive quantities of radioactive material will not be released via the Containment Purge System. To provide assurance that the 36-inch containment valves-cannot be inadvertently opened, the valves are blank flanged. The use of the containment mini purge lines is restricted to the 18-inch purge supply and exhaust isolation valves since, unlike the 36-inch valves, the 18-inch valves are capable of closing during a LOCA or steam line break accident. There-fore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not be 2000 exceeded in the event of accident during containment purging operation. Opera-tion will be limited to 500 hours during a calendar year. The total time the Containment Purge (vent) System isolation valves may be cpen during MODES 1, 2, 3, and 4 in a calendar yt.tr is a function of anticipated need and operating experience. Only safety related reasons, e.g., containment pressure control or the r 'en vi ai rburne rad ived.; a

W = W [ ' " f C_;3 ; ^ ' ^' ~

+.
~ v s L f'. u InsertLfor Basesc3/4.7.8:-~ --& ;g;' . _ ~ For mechanical.' snubbers the' force required t'olinit'iate or ~
3. i
. maintain motion of:the snubber is not great enough.to;over-
. stress.the' attached piping or component'during-thermal
-' movement or':to;-. indicate : impending ' failure of, thei snubber. .x l'- A
e s
k 4 A k r-3 x J k - .J h A 'j,k *A Ll s s bi.- s m h-- ->A w
~
.s JustificationLfor' Bases _3/4.7.8,'pg. 3/4 7-6:
The' insert.'was added to the bases to clarify;that drag force below a specified-range is not required to determine'a snubbers-ability to: reduce the effects..of seismic, events. The snubbers need only _ allow-the' component pipe to move without overstressing -the component / pipe.- 1 7
FINAL DRAFl ELECTRIC POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION (Continued) The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1.129, " Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, ud Replacement of Large Lead Storage Batteries for Generating Stations and Substations." Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the cattery capacity at that time with the rated capacity. 0.085 Table 4.8-2 specifies the norm li. ts for eac designated pilot cell and each connected cell for electr lyte evel, flo voltage and specific gravity. The limits for the desicnated pilot c s float voltage and specific gravity, greater than 2.13 volts e d b w the manufacturer's full charge specific gravity or a battery charg rent that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all tiie connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Operation with a battery cell's parameter outside the normal limit but within the d lowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that *he overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function. Wolf Creek - unit 1 B 3/4 8-2
I O ^ FINAL l)RAF~ RADI0 ACTIVE EFFLUENTS = BASES DOSE RATE (Continued) assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations . exceeding the limits specified in Appendix B Table II of 10 CFR Part 20 (10 CFR 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be suf ficiently low to compensate for any increase in the atmospheric diffusi factor above that for the SITE BOUNDARY. The specified release rate l' 3000 s restrict, at all times, the corresponding gamma and beta dose rates ove ack-ground to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to ess han or equal to 500 mrems/ year to the whole body or to less than or equal o s/ year to the skin. These release rate limits also restrict, at all times the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. gaseous. The required detection capabilities for radioactive materials in li W d waste samples are tabulated in terms of the lower limits of detection (Leus). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.2.2 00SE - NOBLE GASES This specification is provided to implement the requirements of Sections II.B, III. A and IV. A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive materials in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man frca Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimatira Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled WOLF CREEK - UNIT 1 B 3/4 11-3
a-1 ' U':,O EN PE 3 ACCESS ROAD 1 Ti i 9 E DUC ATION ' CENTER NM 2 en , (' m 3 4 F AS 193 5 'i Cz l H E X CLUSION F3 WOLF CREEK AREA-12OO METERS Note:- COOLING LAKE F* 2 1. The exclusion-restricted' 4 area is a 1200 meter radius $^E"I' r circle centered around Unit 1 / containment. I DIK E*8* yp.. SADDLE DAM ll k m I 8 SADDLE DAM Ill \\ \\ t T FIGURE 5.1-1 / EXCLUSION AREA l WOLF CREEK COOLING LAKE s FAS to [ 0UR4 INGTON \\ N F AS 10 Y I
t en t
6q TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS E CYCLIC OR DESIGN CYCLE E COMPONENT TRANSIENT LIMIT OR TRANSIENT 5 Reactor Coolant System 200 heatup cycles t 1 1 0 F/h l s cycle - T,y9 fr m.<_ 2 0 F and 200 cooldown c at lLo 2. 5 0 F. < 100 F/h. I nc e-T avg from >5
F o < 2 0 F.
essurizer couldown cycles Pressurizer cool cycle at < 2 0 F/h. >iperatures f m.>_ 6 0 F to ~ 4- / }2 F. 80 loss of load cycles, without m of RATED TilERMAL POWER to > s a inanediate Turbine or Reactor trip. f RATED TilERMAL POWER. 40 cycles of loss of-of fsite Loss-of-of fsite A.C. electrical ,j A.C. electrical power. ESF Electrical System. i 80 cycles of loss of flow in one Loss of only one. reactor reactor coolant loop. coolant pump. 400 Reactor trip cycles. 100% to 0% of RATED TilERMAL POWER. 10 auxiliary spray aculation Spray water temperature differential M cycles. > 320 F. 50 leak tests. Pressurized o,> 24 5 psf - M 0 P i 5 hydrostatic pressure tests. Pressurized to > 3106 psig. O Secondary Coolant System 1 large steam line break. Break in a > 6-inch steam line. M S hydrostatic pressure tests. Pressurized to > 1350 psig. M ',.l
!? ~ n - Ills ~ il 11ll 18' li 811 - lii e ai -e i 2l i! p - ill - ll ll qi i il r i i !gl m - !!i lil ili - Ill WOLP CREEK UNIT 1 6-3
. = - p -y Plant Manager _ by -Supt of Maint -Supt of Operations -Supt of Tech Supp -Supt of Plant Supp- -Supt'of~ Regulatory, g Quality and Admin 0 -Maint Sves Supvr -Oos Coord - Ops -Reactor Eng Supvr -Fire Prot Spec *- Services. pbintEngr. fShiftSupervisor fReactorEngineer fTrainingSpec M -Mat. Cont Supvr -Supervising Opr -Engineering Spec
a
-Mat Coord Supvr -Reactor Operator -I&C Supervisor -Chief of Security
e
-Storeroom Supvr -Station Operator -Engineer / Spec -Security Ops Supv -Safety Specialist ~ ~ ,2 FWarehouseAtt -Utility Helper -I&C Coordinator- -Lieutenant E -Uti'lity Supvr LI&C Specialist -Sergeant.. -Nuclear Medical Spec. -Utility Mech -Ops Coord - Plan- -I&C Technician -Officer g -Bldg Svcman ning and Projects -Utility Helper -Site Emerg Planning -Engineer / Spec -Lead Computer Eng -Data Mgmt Supvr Administrator- -Maint Supp Supvr .-Surv Coord -Comp Eng/ Spec -Data Base Analyst -Mech Supvr fSystemAnalyst -App Programner -Asst Supv Comp Ops -Document Cont Supvr- -Welder I -Technician -Document Cont Clerk ~ -Machinist -Site Chemist -Computer Opr -Analysis & Film Opr -Mechanic femistrySupvr -Asst Supy Sys.Sftwr' i -Utility Hlpr -HP/ Chem Tech -Data Comu Engr Spec -Administrative Supvr i -Electrical Supv -Utility Helper -Comp Sys Tech flerk l -Electrician -Clerk -Utility Hlpr -Site Health Phys
fHPSupervisor
-Results Eng Supvr -HP/ Chem Tech -Results Eng (M) -Utility Helper -Results Eng (E) -Startup Manager fStartupOrganization e .n -
For technical matters of an immediate nature the respective individual reports-directly_to_the, Plant Manager.
FIGURE 6.2-2 g This position requires an SRO UNIT ORGANIZATION License. t L
o-ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) -E Technician) or_ personnel continuously _ escorted i-exempt from the RWP issuance requirement during the performance of their ci assigned duties in high radiation areas with exposure rates equal to or less . tion procedures for entry into such high radiation areas.th Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following: j A radiation monitoring device which continuously indicates a. the radiation dose rate in the area, or i b. A radiatio'n monitoring device which continuously integrates the radiation (ose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them, or An individual qualified in radiation protection c. procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at j t the frequency specified by the Site Health Physicist 1 in the RWP. '4 [ ore"9 c SJ 6.12.2 In addition to the requirements of Specification 6.12.1, to personnel with radiation levels greater than 1000 mR/h at 45 rea accessible m( in.) fro, the radiation source or from any surface which the radiation pe etr t provided with locked doors to prevent unauthorized entry, and es shall e maintained under the administrative control of the Shift Super isor on d e k ys shal e and/or health physics supervision. Doors shall remain locked e ring periods of access by personnel under an approved RWP which shall specify the do rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, be made by personnel qualified in radiation protection positive exposure control over the activities being performed within the area. For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, such as PWR con-tainment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device. WOLF CREEK - E.IT 1 6-23 .-}}

ML20100F105

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3.3 APPLICABILITY

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ML20100F105

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Dear Mr. Denton:

Dear Mr.-Denton:

3.3 APPLICABILITY

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