ML20093H720

Forwards marked-up Proof & Review Version of Tech Specs,In Response to 840628 Request.Encl Includes All Changes Issued to Plant in NUREG-1058 That Are Also Applicable to Wolf Creek

ML20093H720
Person / Time
Site:	Wolf Creek
Issue date:	07/20/1984
From:	Koester G KANSAS GAS & ELECTRIC CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
KMLNRC-84-118, NUDOCS 8407250251
Download: ML20093H720 (592)

Text

KANSAS GAS AND ELECTRIC COMPANY THE ELECTAC COMPANY

@LENN L KOESTER vect entssorNT- NucLEam July 20, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 KMLNRC 84-118 Re: Docket No. STN 50-482 Ref: Letter dated 6/28/84 from BJYoungblood, NRC, to GLKoester, KG&E Subj: Technical Specifications

Dear Mr. Denton:

The Referenced letter requested that page changes to the " Proof and Review" version of the Wolf Creek Technical Specifications be submitted for NRC approval.

Transmitted herewith are the Wolf Creek Technical Specifications " marked up" to indicate changes. These " marked up" pages include all changes issued to Callaway in NUREG-1058 that are applicable to Wolf Creek, as well as any plant specific changes and justifications required for Wolf Creek.

A listing and categorization of all changes made in the " marked up" copy of the Wolf Creek Technical Specifications has also been provided.

This information is hereby incorporated into the Wolf Creek Generating Station, U71t No.1, Operating License Application.

Yours very truly, gff, i N GLK bb Attach

\

xc: PO'Connor (2) I HBundy

• B407250251 840720

• PDR ADOCK 05000482 3 j @i l A PDR V 201 N. Market -Wichita, Kansas - Mail Address: RO. Box 208 I Wichita, Kansas 67201 - Telephone: Area Code (316) 261-6451

m-i h

r s

OATH OF AFFIRMATION

! STATE OF KANSAS )

) SS:

COUNTY OF SEDGWICK )

r I, Glenn L. Koester, of lawful age, being duly sworn upon oath, do depose, state and affirm that I am Vice President - Nuclear of Kansas Gas and Electric Cotrpany, Wichita, Kansas, that I have signed the foregoing letter of transmittal, know the contents thereof, and that all statements contained therein are true.

KANSAS GAS AND ELECTRIC COMPANY By ) 4,y) f)

Gle'nn L. Koester Vice President - Nuclear E.D. Prothro, Assistant Secretary

[

STATE OF KANSAS )

L '

) SS:

COUNTY OF SEDGWICK )

BE IT REMEMBERED that on this 20th day of July, 1984 , before me, Evelyn L. Fry, a Notary, personally appeared Glenn L. Koester, Vice President - Nuclear of Kansas Gas and Electric Company, Wichita, Kansas, who is personally known to me and who executed the foregoing instrument, and he duly acknowledged the execution of the same for and on behalf of and as the act and deed of said corporation.

n o "d"*i.JUe,and WITNESS yearWHEREOF, I have hereunto set my hand and affixed my seal the above written.

> + . )- . *

. l4s - ,..

f;- g F

UCV '/ E Yy' n I.. ry, Notary rl llg~.9..e

... r. . . .h,7 Commission expires on August 15, 1984.

t I

L _ _ _ . - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

TECHNICAL SPECIFICATION CHANGES Page Section Reason for Change VIII Table 3.4-1 Typo IX 3/4.5. 5 Consistent with Callaway Technical Specifications (T.S.)

IX 3/4.5.6 Consistent with Callaway T.S.

IX 3/4.6.4 Qansistent with Callaway T.S.

XII 3/4.8.4.2 Consistent with Callaway T.S.

XII Table 3.8-2 Cbnsistent with Chllaway T.S.

XV 3/4. 3.4 Consistent with Callaway T.S.

XVI 3/4. 5. 5 Consistent with Chilaway T.S. ,

XVI 3/4.5.6 Consistent with Callaway T.S.

i (bnsistent with Callaway T.S.

1-2 1.8 1.2 1.11 Typo 1.'4 1.20 Typo 1-6 1.38 Consistent with Callaway T.S. .

2-1 2.1.1 .Cbnsistent with Callaway T.S.

2-2 Figure 2.1-1 Consistent with Callaway T.S.

2-5 Table 2.2-1 Wolf Creek specific values provided.

(itens 1 through 11) 2-6 Table 2.2-1 Wolf Creek specific valtas provided.

(itens 12 through 15) 2-7 Table 2.2-1 Cbnsistent with Chilaway T.S.

2-7 Table 2.2-1 Wolf Creek specific values provided.

(items 18b throt*3h ISf)

TECINICAL SPECIFICATION CHAIES Page Section Reason for Change 2-8 Table 2.2-1, Wolf Creek specific value provided.

Table Notations (K2) 2-9 Table 2.2-1, ensistent with Callaway T.S.

Table Notations 2-9 . Table 2.2-1, W lf Creek specific value provided.

• Table Notations (Note 2) 2-10 Table 2.2-1, Wolf Creek specific value provided.

Table Notations (K4) 2-11 Table 2.2-1, Nblf Creek specific value provided.

Table Notations (Note 4)

B 2-2 2.1.2 (bnsistent with Callaway T.S.

B 2-4 Bower Range Typo Neutron Flux B 2-5 Intennediate & (bnsistent with Callaway T.S.

Source Range B 2-5 Overpower AT (bnsistent with Callaway T.S.

B 2-6 Pressurizer- Consistent with Callaway T.S.

Pressure and Typo

-B 2-6 Pressurizer -Consistent with Callaway T.S.

. Wter. Ievel B 2-8 Reactor Trip (bnsistent with Callaway T.S.

Systen Interlocks 3/4 1-1 3.1.1.1 (* note) Consistent with Callaway T.S.

3/4 1-4 3.1.1.3 (# note) Consistent with Callaway T.S.

3/4 1-6 3.1.1.4 (* note) Consistent with Callaway T.S.

3/4 1-7 3.1.2.1.a & Consistent with Callaway T.S.

3.1.2.1.b

TEGNICAL SPECIFICATICN CHANGES Page Section Reason for Change 3/4 1-8 3.1.2.2 (* note) Consistent with Callaway T.S.

-3/4-1-10 3.1.2.4 (* note) Consistent with Callaway T.S.

3/4 1-10 3.1.2.4 (Action) Consistent with Callaway T.S.

3/4 1-11 3.1.2.5.a.1) & Cbnsistent with Chllaway T.S.

3.1.2.5.b.1) 3/4 1-11 4.1. 2. 5. a . 3) (bnsistent with Callaway T.S.

3/4 1-12 3.1.2.6, Consistent with Callaway T.S.

3.1.2.6.a &

3.1.2.6.b 3/4 1-12 Action a,b,&c (bnsistent with Callaway T.S.

3/4 1-13 4.1.2.6 Consistent with Callaway T.S.

L 3/4 1-14 3.1.3.1 (* note) Consistent with Callaway T.S.

3/4 1-15 Action e Consistent with Callaway T.S.

3/4 1-15 4.1. 3.1. 2 _ Typo 3/4 1-17' 3.1. 3. 2 Consistent with Callaway T.S.

3/4 1-17, 3.1. 3. 2 (bnsistent with Ch11away T.S.

. Action a & b 3/4 1-17 4.1. 3. 2 (bnsistent with Callaway T.S.

3/4 1-18 3.1.3.3 (# note) Consistent with Callaway T.S.

3/4 1-20 3.1. 3.5 (* note) Consistent with Callaway T.S.

3/4 1-21 3.1. 3. 6 Consistent with Callaway T.S.

3/4 2-1 3.2.1 (bnsistent with Callaway T.S.

3/4 2-1 3.2.1 (Action 6) Consistent with Callaway T.S.

3/4 2-1 3.2.1 (* note) Cbnsistent with ballaway T.S.

3/4 2-8 3. 2. 3.c Consistent with Callaway T.S.

'IEGNICAL SPECIFICATION CHANGES Page Section Reason for Change 3/4 2-9 Figure 3.2-3 MakeWolfCreekspecif[$.

3/4 2-10 ' 4. 2. 3. 5 (bnsistent with Callaway T.S. Incitries Wolf Creek specific instrunent.

3/4 2-11 3.2.4 Consistent with Callaway T.S.

3/4 2-15 Table 3.2-1 (bnsistent with Callaway T.S.

3/4 3-2 Table 3.3-1 Consistent with Callaway T.S.

(6b & 6c)

'~

'3/43-3 Table 3. 3-1 (bnsistent with Callaway T.S.

(12a & 12b) 3/43-4 Table 3. 3-1 (19) Consistent with Chilaway T.S.

3/4 3-5 Table 3.3-1, Consistent with Callaway T.S.

Table Notations 3/43-6 Table 3. 3-1 (bnsistent with Callaway T.S.

(Action 5) 3/4 ~ 3-6 Table 3.3-5 Justification provided by M4INRC (Action 5 - 84-034, dated 3/16/84 deletion of Joron Diluticrt Techni-cal Specification) .

3/4 3-7 Table 3.3-2 (1) Consistent with Callaway T.S.

3/43-8 Table 3. 3-2 (bnsistent with Callaway T.S.

(Heading &

-Item 12) 3/43-9 - Table 4. 3-1 (item Justification provided by M4IIRC 6 - deletion of 84-034, dated 3/16/84 note (12))

3/4 3-12 Tabic 4. 3-1, (bnsistent with Callaway T.S.

Table Notations 3/4 3-12 Table 4. 3-1, Table Justification provided by M4INRC Notations (note 84-034, dated 3/16/84 (9) and note (12))

6 i

, - - - .- , ,c-,,.,, r- w,--, ,--,_,

-- ,,-,-,e- r-- .----,,--,,m- --, -gr

n-TEQWICAL SPECIFICATION CHANGES Page Section Reason for Charge 3/4 3-13 3. 3.2 ( Action c) Consistent with Callaway T.S. .

3/4 3-14 Table 3. 3-3 (1, Omsistent with Chilaway T.S.

lb, le, & 2b) 3/4 3-15 Table 3.3.-3 Cbnsistent with Callaway T.S.

( 3.b. 2) , 3.c. 3) ,

& 3c4))

3/4 3-16 Table 3.3-3 (bnsistent with Callaway T.S.

(4.a.1), 4b, .

4d, 4e, Sa, 5b, Sc) 3/4 3-17 Table 3. 3-3 Cbnsistent with Chilaway T.S.

(Heading, 6a, 6b, 6c, 6d, 6e,

& 6f) 3/4 3-18 Table 3.3-3 (bnsistent with Callaway T.S.

(Headirg, 6g, 6h, & 7a) 3/4 3-19 Table 3.3-3 (9a, Cbnsistent with Callaway T.S.

9b, c, 9d, 10

• & 11) 3/4 3-20 Table 3. 3-3, (bnsistent with Callaway T.S.

Table Notations,

(## note &

1. 1. 1. note) .

3/4 3-21 Table 3. 3-3 (Action (bnsistent with Callaway T.S.

24, 25, 26, & 27) 3/4 3-22 Table '3. 34 (1, Ib, Cbnsistent with Os11away T.S.

Id, & le) 3/4 3-22 Table 3. 3-4 (1.c, Walf Creek specific values provided, l.d, & l.e) 3/4 3-23 Table 3. 3-4 (2b, (bnsistent with Callaway T.S.

3s.2), & 3.b.2))

3/4 3-23 Table 3. 3-4 (2.c Wolf Creek specific values provided.

& 3.b.3))

l

TECIMICAL SPECIFICATION CHANGES Page Sectim Reason for Change 3/4 3-24 Table 3.3-4 Consistent with Callamy T.G.

(Heading, 3.c.2), 3.c. 3) ,

3.c.4), 4b, 4d, and 4e) 3/4 F24 Table 3. 3-4 , 21f Creek specific values provided.

(4.c, 4.d, & 4.e) 3/4 3-25 Table 3.} 4 (bnsistent with Callamy T.S.

(Heading, 5a, 5.b, 5.c, 6.b, 6.c, 6.d, 6.d.1

& 6.d.2) 3/4 3-25 Table 3. 3-4 (5.b, 2 1f Creek specific values provided.

6.d.1), & 6.d.2))

3/4 3-26 Table 3.3-4 (7a) Consistent with Callaway T.S.

3/4 3-26 Table 3.3-4 (7b) 21f Creek specific value provided.

3/4 3-27 Table 3. 3-4 Cbnsistent with Callaway T.S.

(Heading, 9b, 9c, 9d, le & 11) 3/4 3-27 Table 3. >4 (11.a) 2 1f Cieek specific value provided.

3/4 3-29 Table 3.3-5 (im, consistent with Callamy T.S.

lo, 2.a, 2.a.2),

2.a. 3), 2.a.4),

2.a.5), 2.a.6),

2.a.7), 2.a.8) ,

2.a.9))

3/4 3-30 Table 3. 3-5 (bnsistent with Callamy T.S.

( 3.a, 3.a.2) ,

3.a. 3), 3.a.4),

3.a.5), 3.a.6) ,

3.a.7), 3.a.8),

& 3.a.9), 4.a, 4.a.1), 4.a.2),

4.a. 3), 4.a.4) ,

4.a.5), 4.a.6), ,

4.a.7), 4.a.8),

& 4 a.9)) ,

=

e TEGNICAL SPECIFICATION CHANTS Page Secticri Reason for Change 3/4 3-31 Table 3.3-5 (5.a, consistent with callaway T.S.

7, & 8b) 3/4 3-32 Table 3. 3-5 Cbnsistent with Callaway T.S.

(14a, 14b, &

15b) 3/4 F33 Table 3. 3-5 Consistent with Callaway T.S.

((2), (3), (4),

(5), (6), (7),

& (8))

3/4 3-34 Table 4. 3-2 (1, Cbnsistent with Callaway T.S.

lb, & 2b) 3/4 3-35 Table 4. 3-2 (bnsistent with Callaway T.S.

(Heading, 3.a . 2 ) , 3.b . 2 ) ,

3.c.2), 3.c. 3) ,

& 3.c.4)) -

3/4 3-36 Table 4. 3-2 Omsistent with Callaway T.S.

(4.b, 4.e, 5.a, 5.c, 6.b, 6.c,

& M) .

3/4 3-37 Table 4. 3-2 (bnsistent with Callaway T.S.

. (Headir;,&

7.a) 3/4 3-38 Table 4. 3-2 (bnsistent with Callaway T.S.

(Hearif rg, 95, 9c, 9d, 10, 11,

& footnotes (2)

& (3))

3/4 3-40 Table 3. 3.6 (la, Cbnsistent with Ch11away T.S.

Ib, ic. Id, 2a, 2b, & 3) 3/4 3-41 Table 3. 3-6 (* Cbnsistent with Callaway T.S.

note, ** note,

1. note, ## note,

1. 1. 1. note, Action 27, 28, & 30) l l

I

TEWNICAL SPECIFICATION CHALTES Page Sectial Reason for Change 3/4 3-42 Table 4.3-3 (la, Consistent with Callamy T.S.

Ib, ic, Id, 2a, 2b, &3) 3/4:3-46 Table 3. .F7 (bnsistent with Callaway T.S.

and includes Wolf Creek setpoints.

3/4 3-46 Table 4.3-4 (2c, Consistent with Callaway T.S.

2d, 2e, 2f, 3,

• note, & **

note) 4 3/4 3-50 3.3.3.5 (IC Os (bnsistent with Callamy T.S.

Actions a, b,

& c)

'3/4 3-50 3.3.3.5 (* note) Justification provided.

3/4 3-50 4. 3. 3. 5.1, Consistent with Callaway T.S.

4. 3. 3. 5. 2, & -

4. 3. 3. 5. 3 1

3/4 3-51 Table 3. 3-9 (5, (bnsistent with Callaway T.S.

9, & 12) 3/4 3-52 tble 4. M (12) Consistent with Callaway T.S.

3/4 3-53 3. 3. 3.6 ( Actions Consistent with Callamy T.S.

a, b, c,.& d) 3/4 .3-54 Table 3. 3-10 (1, Cbnsistent with Callaway T.S. ,

12, 13, 14, 15, 16, 17, 18, 19, 20, & *** note) 3/4 3-55 Table 4. 3-7 (1, Cbnsistent with Callaway T.S.

12, 13, 14, 15, 16, 17, 18, 19, 20, & *** note) 3/4 3-58 Table 3. 3-11 Cbnsistent with Callaway T.S.

(1110 & 1331) i 3/4 3-59 Table 3. 3-11 (bnsistent with Callaway T.S.

(Headirg, 1509,

~

I

& 1500) l

TECHNICAL SPECIFICATION CHANGES Page Section Reason for Change 3/4 3-60 Table 3.3-11 Consistent with Callaway T.S.

(Heading, 3409

& 3601) 3/4 3-61 Table 3. 3-11 Cbnsistent with mlf Creek Design (ESW Planphouse,

& ESW Cboling

'Ibwer) 3/4 3-61 Table 3. 3-11 Cbnsistent with Callaway 'I'.S.

(ESF Transform-ers & note (1))

3/4 3-62 4. 3. 3. 9. b (bnsistent with Chilaway T.S.

3/4 3-64 Table 3. 3-12 Consistent with Callaway T.S.

(la, lb, ic,

& Id) 3/4 3-64 Table 3. 3-12 (2d) Consistent with m lf Creek design 3/4 3-65 Table 3.3-12 Consistent with Callaway T.S.

(Actions 32 & 33) 3/4 3-66 Table 4.3-8 (la, (bnsistent with Callaway T.S.

lb, Ic, id, 2a, 2b & 2c)

• 3/4 3-66 Table 4.3-8 (2d) Consistent with Elf Creek design .

, 3/4 3-67 Table 4.3-8 Consistent with Callaway T.S.

(Notes (1),

(1)a, & (1)b) 3/4 3-67 Table 4. 3-8 Cbnsistent with 21f Creek specific j (Note (2))' program 3/4 3-69 Table 3. 3-13 Cbnsistent with Callaway T.S.

< (la, 2a, 2b, 2c, 3n, 3b, 3c, I M, & h) 3/4 3-69 Table 3. 3-13 Cbnsistent with Elf Creek design (2d)

(

1

-- g v--.. - ,_ .__.,_,.,,,,__,_.,._,,.______--.,m__r _ , , . -_ _.

TEGNICAL SPECIFICATION OIANGES Page Section Reason for Change 3/4 3-70 Table 3.3-13 Consistent with Callaway T.S.

(4a, 4b, 4c,

& 4d) 3/4 3-71 Table 3. 3-13 Cbnaistent with W if Creek design (Action 39) . .

3/4 3-71 Table 3 3.-13 mnsistent with Ch11away T.S.

(Actions 42

& 45) 3/4 3-72 Table 4. 3.9 Cbnsistent with Callaway T.S.

(ld, 2a, 2b, 2c, 3a, k, 3c, 3d, &3e) 3/4 3-72 Table 4. 3-9 Cbnsistent with Wif Creek design (2d) .

3/4 3-73 Table 4. 3-9 Cbnsistent with Callaway T.S.

(4a, 4b, 4c, 4d, & 4e) 3/4 3-74 Table 4. 3-9 (bnsistent with Callamy T.S.

(Notes (1), (1)a, '

(1)b, (6), & (7) 3/4 3-74 Table 4. 3.-9 2 nsistent with m lf Creek specific (Note (3)) program 3/4 3-75 3. 3.4 (* note) Cbnsistent with Ch11away T.S.

3/4 3-75 4. 3.4.2.a. 3) & 4) Consistent with Wolf Creek design 3/4'3-75 4. 3.4.2 (a2), a3), W if Creek specific justification b, c, d, e)) provided.

3/4 4-1 3.4.1.1 (Action Cbnsistent with Callaway T.S.

&

• note) 3/4 4-2 3.4.1.2 & 3.4.1. 2 (bnsistent with Callaway T.S.

(Actions b & c,

& ** note) 3/4 4-2 4.4.1.2.3 @nsistent with Callaway T.S.

e TECHNICAL SPECIFICATICN QWES Page Section Reason for Gange 3/44-3 3.4.1. 3 (* note) Typo 3/4 4-4 4. 4.1. 3. 2 Consistent with Callaway T.S.

3/4 4-5 3.4.1.4.1 Typo

(** note)

, 3/4 4-10 3.4.4 (Actions Cbnsistent with Chilaway T.S.

a,b,c,d,&

e, &

• note) 3/4 4-10 4.4.4.2 (bnsistent with Chllaway T.S.

3/4 4-11 3.4.5 (Action) Consistent with Callaway T.S.

3/4 4-13 4.4 5. 3.c. 3) (bnsistent with Callaway T.S.

& 4) '

3/4 4-14 4.4.5.4.a.6) & 7) Consistent with Callaway T.S.

3/4 4-15 4.4.3.5.c Consistent with Callaway T.S.

3/4 4-18 4,?.6.1.6 (bnsistent with Callaway T.S.

3/4 4-19 3.4.6.2.f (* note) Consistent with Callaway T.S.

3/4 4-20 4.4.6.2.lc, (bnsistent with Chllaway T.S.

4.4.5.2.2.b,

& 4.4.6.2.2.d 3/4 4-21 Table 3.4-1 Cbnsistent with Ch11away T.S.

, 3/44-22 3.4.7 (Action a) Consistent with Callaway T.S.

3.4 4-25 3.4.8 (Actions b (bnsistent with Ch11away T.S.

arx1 d) 3/4 4-26 3.4.8 (Action Cbnsistent with Chllaway T.S.

for Modes 1-5, Actions 2 & 5) 3/4 4-28 Table 4.4-4 Cbnsistent with Ch11away T.S.

(2 & 3) 3/4 4-28 Table 4.4-4 (4.a)) Typo

-l I

l TE31NICAL SPEIFICATION 01ANGES Page Section Reason for Change 3/4 4-29 4.4.9.1.2 Cbnsistent with Callamy T.S.

3/4 4-30 Figure 3.4-2 Consistent with Calla m y T.S.

3/4 4-31 Figure 3.4 ~ (bnsistent with Chilamy T.S. with Wolf Creek specific notch in curve.

3/ 4 4 - 32 Table 4.4-5 Consistent with Wolf Creek analysis 3/4 4-34 3. 4. 9. 3 Cbnsistent with Callamy T.S. and justification provided 3/4 4-34 3. 4. 9. 3 (a , ' b, Consistent with Callaway T.S. and

& c and Actions justification provided.

a, b, & c) 3/4 4-35 4. 4. 9. 3. 2 Consistent with Callaway T.S.

& 4.4.9.3.3 3/4 4-36 Figure 3.4-4 consistent with Calleway T.S.

3/4 5-1 3.5.1.a ard Consistent with Callaway T.S.

Action b 3/45-2 4.5.1.1 (b,c,&d) Cbnsistent with Chilaway T.S.

3/4 5-3 3.5.2 (* note) Consistent with Callaway T.S.

3/45-5 4.5.2.e.1) Cbnsistent with Callaway T.S.

3/4 5-6 4. 5. 2.h . 3) Consistent with callaway T.S.

& 4.5.2.1 3/45-7 3/4. 5. 3 Cbnsistent with Callaway T.S.

(11eading) 3/45-7 3. 5. 3.d and Cbnsistent with Chilaway T.S.

3.5.3 (Action a

& c) ,

3/4 5-8 4. 5. 3. 2 (bnsistent with Callaway T.S.

3/4 5-9 3. 5.4 ' ( Ac hen Consistent with Callaway T.S.

&

• note)

F

~ . .

m i I TBORGCAL SPECIFICATION CHAtGS Page Section Dam m for Change 3/4.5-18 3/4.5.5 Consistent with Callaway T.S.

3/4 5-11 3/4.5.6 Consistent with Chilamy T.S.

(Heading) 3/4 5-11 3.5.6 (c & d) Q)nsistent with on11away T.S.

3/4 6-1 4.6.1.1.a Consistent with Callaway T.S.

3/46-2 3.6.1.2.a.2) Osnsistent with 19C8R, Appendix J.

3/4 6 3.6.1.2.b Consistent with Callawny T.S.

V4 6-2 4.6.1.2.a Typo 1

3/4 6-3 4.6.1.2(c.1), Consistent with Callaway T.S.

c.3), f & g) 3/46-3 4.6.1.2.d M)1f Creek specific justification l Prwided 3/4 6-5 4. 6.1. 3.a Consistent with Callaway T.S.

V4 6-6 3.6.1.4 (bnsistent with on11amy T.S.

3/4 6-8 3.6.1.6 Revise to be consistent with thru thru Callaway T.S.

3/4 6-16 4.6.1.6.2

& 4.6.1.6.3 3/4 6-11 3.6.1.7 (a & b Q)nsistent with On11am y T.S.

and Actions a, b, & c)

E 3/4 6-12 4.6.1.7 (1, 2, 03nsistent with Oillaway T.S.

3, 4, &

• note) 3/4 6-13 3.6.2.1 and Consistent with Oillawny T.S.

Acticn 3/4 6-13 4.6.2.1.c (1) & Osnaistent with 1hble 3. 3-3 2))

3/4 6-14 4.6.2.1.(b, Cbnsistent with Oillamy T.S.

c.1) % e.2)) ,

l s

t

'11!K21NICAL SPECIFICATION CHANGES I

Page Section Reason for Charge 3/4 6-14 4.6.2.2.c (bnaistent with Callamy T.S. and Table 3. 3-3 3/4 015 3. 6. 2. 3 consistent with Callamy T.S.

(Action c) 3/4 6-15 4.6.2.3. (bnsistent with Callaway T.S.

(a.2) & b) 3/4 6-16 3.6. 3 ( Action, Oshsistent with callamy T.S. -

Actions b & c) 3/4 6-16 4. 6. 3.1 (bnsistent with Chilamy T.S.

3/4 6-17 4. 6. 3. 2 Cbnsistent with Callamy T.S.

3/4 6-18 Table 3.6-1 Consistent with callam y'T.S.

(** note) 3/4 6-18

• Table 3.6-1 Typo (P-25) 3/4 6-19 Table 3.6-1 consistent with Ch11away T.S.

(** note) 3/4 6-20 Table 3.6-1 Typo *

(P-101 t P30) 3/4 6-21 Table 3.6-1 Cbnsistent with Cs11aw y T.S.

(** note, first P49, % P-58) 3/4 6-22 Table 3.6-1 (bruistent with callamy T.S.

(** note,

• • • rete, P-16, P-13, P-45, P-65 (both), %

P-67) 3/4 6-22 Table 3.6-1 Typo (P-45) 3/4 6-23 Table 3.6-1 (bnsistent with chilamy T.S.

(*** note & 5) e a G

n -_- - - - - - . . . . - _ _ _ _ _ - - _ _ _ - . -

r

~

TEWNICAL SPECIFICATICH CHAtRS PLge Section Reason for Charrje

. 3/4 6-23 Table 3.6-1 Typo (P-40) 3/4 6-24 Table 3.6.1 Cbncistent with Callam y T.S.

(** note, P-15

& P-14) 3/4 6-25 Tabla 3.6-1 Typo (P-15) 3/4 6-26 Table 3.6-1 Typo (P-14 & P-89) 3/4 6-28 Table 3.6-1 Typo l

(P-38) 3/4 6-29 Table 3.6-1 (bnsistent with Callam y T.S.

(P-98 & 9) .

3/4 6-30 3.6.4.1 (busistent with Callam y T.S..

(Actions a

& b) 3/4 6-30 4.6.4.1 Cbnaistent with Callamy T.S.

3/4 6-31 3.6.42 Consistent with Callamy T.S.

& Action 3/4 6-31 4.6.4.2 (bnsis' tent with Callamy T.S.

(1.b.2) & 2) 3/4 7-5 4. 7.1. 2.1.b . 3) (bnsistent with Callamy T.S.

3/4 1-6 3. 7.1. 3 & Consistent with Callamy T.S.

Action b 3/4 7-11 . 4.7. 3 (b.1) & Cbnsistent with Callamy T.S.

b.2))

3/4 7-13 3.7.5.b (bnsistent with ibit Creek analysis 3/4 7-15 4.7.6 (c.1), (Lnsistent with Callamy T.S.

c.2), d, & m2))

4 TBONICAL SPECIFICATIOti CHANGES Page Section Reason for Charge 3/4 7-16 4.7.,6.e.5) Consistent with Wolf Creek design

& Typo 3/4 7-16 4.7.6.(f & g) Consistent with Callamy T.S.

3/4 7-17 3.7.7 (Action) (bnsistent with Callaway T.S.

3/4 7-17 4.7.7 (b.1) & Consistent with Callamy T.S.

b.2))

3/4 7-18 4.7.7 (c, d2)," Cbnsistent with Callam y T.S.

e, & f) 3/4 7-19 3.7.8 (bnsistent with Callamy T.S.

3/4 7-19 4.7.8 Consistent with Callamy T.S.

3/4 7-19 4.7.8 (* note) Typo

-3/4 7-20 4.7.8.c Consistent with Callamy T.S.

3/4 7-20 4.7.8.e.2) Typo 3/47-21 4.7.8.(e.2) Consistent with Callamy T.S.

& e.3)) & Typo

. 3/47-21 4.7.8.e Typo 3/4 7-22 4.7.8 (f.2) & g) Consistent with Callam y T.S.

3/4 7-27 3.7.10.1.a Wolf Creek specific justification provided 3/4 7-27 3.7.10.1 Wolf Creek specific justification (Action a) provided 3/4 7-28 4.7.10.1.1.f.1) Consistent with Callam y .T.S.

3/4 7-28 4.7.10.1.1.f.2) W31f Creek specific justification . .

prwided

3/4 7-28 4.7.10.1.2.a Wolf Creek specific justification f

.gwMM i

i f

l t.

m. - - --

o I

, TBOSECM SPECIFICATION 09dGES j l

Page . Section Reason for Change r I

1. 4 7-38 3.7.18 2 (a, c, (bnsistent with on11away T.S. ,

&

• note) l

1. 4 7-31 4.7.18.2.d ,

(bnsistent with Onllaway T.S.

3/4 7 - 32 3.7.18.3 (* note) Consistent with Callamy T.S.

3/4 7-32 4.7.18. 3 (a, b, (bnsistent with callamy T.S. I c.1) & c.2)) {

3/4 7-34 Table 3.7-5 (bnsistent with Callaway T.S. r

( h Bldg.) ,

3/4 7-36 3.7.11 & 3.7.11 0)nsistent with On11away T.S.

(Action a) r 3/4 7-36 4.7.11.1, (bnsistent with callaway T.S.

4.7.11.1.b, &

4.7.11.1.c 3/4 7-36 4.7.11.1.c Typo l

1. 4 7-36 4. 7.11. 2.b (bnsistent with Callaway T.S.

3/4 7- 37 3.7.12 consistent with Callaway T.S. ,

!3/48-1 3.8.1.1 (b.1) consistent with Qillaway T.S. *

& Actions a

• a b) i 3/48-2 3.8.1.1 (bnsistent with callaway T.S.

(Action e) ,

3/48-2 4.8.1.1.1 Cbnsistent with Oillaway T.S.

(a & b) 3/4 8-3 4.8.1.1.2 (a.4), consistent with Callaway T.S.

d, e, &

• note)

.3/48-4 4.8.1.1.2 (f. , (bnsistent with Oillaway T.S.

f.2), f.4)b),

f.5))

I 1_ A.

TB39tICAL SPECIFICATION CHAN3!lS Page Section Bananri for Change 3/46-5 4.8.1.1.2 (f.5), Consistent with Callamy T.S.

f.6), f.7), &

• note)

1. 4 8-6 4.8.1.1.2 (g, Cbnsistent with Ch11away T.S.

h.1))

3/48-6 4. 8.1.1. 3 (bnsistent with Callaway T.S.

3/48-7 Table 4.8-1 Car.sistent with Callaway T.S.

(* note) .

3/4 See 3.8.1.2 (Action) Cbnsistent with callaway T.S.

3/4 8-9 3.8.2.1 Consistent with Callaway T.S.

(Actions a & b) 3/48-9 4.8.2.1.a.2 03nsistent with callam y T.S.

3/4 8-19 4.8.2.1. (b.2)', Consistent with Callaway T.S.

c. 3), c.4), d,

& e) 3/4 8-11 Table 4.8-2 Cbnaistent with callaway T.S.

(Float voltage, Specific gravity) 3/4 8-12 3.8.2.2 (Actions Cbnsistent with Otilaway T.S. ,

a & b) 3/4 8-13 3. 8. 3.1 03nsistent with callaway T.S.

(a through j, .

Action c) 3/4 8-15 3.8. 3. 2 (bnaistent with Callaway T.S.

3/4 8-15 3.8. 3.2 (a & b, consistent with Callawny T.S.

and Action) 3/4 8-16 4. 8. 4.1.a .1 Consistent with Os11away T.S.

3/4 8-17 4. 8. 4.1.a . 2 consistent with callaway T.S.

V4 8-18 Table 3.8-1 Cbnsistent with Ch11away T.S.

thru .

3/4 8-24

TEQtiICAL SPECIFICATION CIANGES Page Section Reason for Charv3e 3/4 8-25 3.8.4.2 Consistent with Callaway T.S.

3/4 0-26 Table 3.0-2 Cbnsistent with Callaway T.S.

3/4 9-1 3/4.9 (Heading) Consistent with Callaway T.S.

3/49-1 4. 9.1. 3 Justification provided by ININRC 84-034 dated 3/16/84 ,

3/4 9-6 3.9.6 (a.2)a), Consistent with Callaway T.S.

a.2)b), & a.3) 3/4 9-6 3.9.6.b.2) Justification provided 3/4 9-7 4.9.6.1 & 4.9.6.2 Cbnsistent with Callaway T.S.

3/4 9-10 3.9.8.2 (* note) Consistent with Callaway T.S.

3/4 9-14 3.9.11 & 4.9.11 Cbnsistent with Chilaway T.S.

3/4 9-14 3.9.11 Justification provided (Applicability)

& 4.9.11 3/4 9-15 4.9.12 (bnsistent with Callaway T.S.

3/4 9-17 3.9.13 -

Consistent with Callaway T.S.

Justification provided 3/4 9-17 3.9.13 ,

3/4 9-17 4. 9.13.b.1) Consistent with Callaway T.S.

3/4 9-18 4. 9.13.b . 2 ) Typo 3/4 10-1 3.10.1 Consistent with Callaway T.S.

3/4'10-2 3.10.2 (Action) Consistent with Callaway T.S.

3/4 10-4 3.10.4 (a, b, arvi consistent with Callaway T.S.

Actions a & b)

3/4 10-4 4.10.4.3 (bnsistent with Callaway T.S.

3/4 11-2 Table 4.11-1 Consistent with Callaway T.S.

(1.b & 2) i I

L

i I

I

'1ECHtICAL SPElCIFICATICH CHAtMS i Page Section Reason for 0 1 ye  !

3/4 11-3 Table 4.11-1, Per discussions with NRC reviewer Table Notations (Bob Fell) ,

(Note (2)) l 3/4 11-4 Table 4.11-1, (bnsistent with Callaway T.S. l Table Notations i

(Notes (4) & (6))

3/4 11-6 4.11.1.3.1 (bnsistent with Callaway T.S.

E S

3/4 11-7 4.11.1.4 Consistent with Callaway T.S. I 3/4 11-9 Table 4.11-2 @ nsistent with Callaway T.S.

(Itsas 1, 2 & 3) 3/4 11-9 Table 4.11-2 21f Cree'k specific justification (Items 3b & 4) provided 3/4 11-9 Table 4.11-2 Cbnsistent with Callaway T.S. l (Item 5) .

l 3/4 11-16 Table 4.11-2 Cbnsistent with Callaway T.S.

l Table Notations (Note (1))

3/4 11-11 Table 4.'11-2, consistent with Callaway T.S. ,

Table Notations (Notes (4), (5),  ;

(7)) - '

3/4 11-12 4 11.2.2 Typo

{

3/4 11-14 3.11.2.4 Typo i i

3/4 11-14 3.11.2.4 Typo 1 j, (Action a.2) '

3/4 11 3.11.2.5 m nsistent with Callaway T.S.

i

!- 3/4 11-15 3.11.2.5 Consistent with Callaway T.S.

(Action a)  !

3/4 11-16 3.11.2.6 (bnsistent with callaway T.S.

3/4 11-16 4.11.2.6 Wolf Creek specific justification provided t

I TECHNICPL SPECIFICATION QPNGES -

Page Section Reason for Otange 3/4 11-18 3.11.4 Typo *

(Action a) 3/4 12-1 3.12.1 Per discussions with NRC reviewer (Action c) (Mike Wangler) 3/4 12-2 3.12.1 Per discussions with NRC reviewer

. (Action c cont'd) (Mike Wangler) 3/4 12-3 Table 3.12-1 Per discussions with NRC reviewer (Item 1) (Mike Wangler) 3/4 12-4 Table 3.12-1 Per oiscussions with NRC reviewer (Itans 2 & 3) (Mib Wangler) 3/4 12-5 Table 3.12-1 Per discussions with NRC reviewer (Items 4a & 4b) . (Mike Wangler) 3/4 12-6 Table 3.12-1 Per discussions with NRC reviewer (Mike Wangler)

. (Item 4c) 3/4 12-7 Table 3.12-1 Per discussions with NRC reviewer Table Notations (Mike Wangler)

(Notes (1) & (2))

3/4 12-0 Table 3.12-1 Per discussions with NRC reviewer Table Notations (Mike Wangler)

(Notes (6) & (7))

3/4 12-9 Table 3.12-2 (bnsistent with Callamy T.S.

3/4 12-10 Table 4.12-1 Consistent with Callamy T.S. ,

3/4 12-13 3.12.2 (* note) Per discussions with NRC reviewer (Mike Wangler) 3/4 12-14 3.12.3 Per discussions with NRC reviewor (Mike mngler)

B 3/4 0-1 3/4.0 Consistent with Callamy T.S.

B 3/4 0-2 3/4.0 (bnsistent with Callamy T.S.

B 3/4 0-3 3/4.0 Consistent with Callamy T.S.

TEl0EICAL SPECIFICATION CHANGES e . ,

Page Sectim Reason for Charrye (

B 3/4 1-2 3/4.1.2 Consistent with Callamy T.S.  !

B 3/41-3 3/4.1. 2 (bnaistent with Callamy T.S.

B 3/4 2-4 3/4.2.2 & 3/4.2.3 Consistent with Callamy T.S.

i B 3/4 2-5 3/4.2.2 & 3/4.2. 3 Cbnsistent with callamy T.S.

B 3/4 2-5 3/4 2.4 consistent with Callamy T.S.

B 3/4 2-6 3/4.2.5 Consistent with callam y T.S.

B 3/4 3-2 3/4. 3.1 & 3/4. 3.2 consistent with Callamy T.S.

B 3/4 3-3 3/4. 3. 3.1 (bnsistent with Callamy T.S.

B 3/4 }-3 3/4.3.3.2 Consistent with Callam y T.S. t B 3/4 F4 3/4. 3. 3. 5 Consistent with Callam y T.S.

B 3/4 3-4 3/4. 3. 3.6 Consistent with Callamy T.S.

B 3/4 3-5 3/4. 3. 3.8 consistent with Callamy T.S.

L B 3/4 3-5 3/4. 3. 3.9 Consistent with Callamy T.S.

B 3/4 3-5 3/4.3.3.10 (bnsistent with callamy T.S.

B 3/4 3-6 3/4. 3.4 consistent with callamy T.S. '

B 3/4 4-1 3/44.1 (bnsistent with Callamy T.S.

B 3/4 4-2 3/4.4.2 Consistent with Callamy T.S.

B 3/4 4-2 3/4.4. 3 (bnaistent with Callamy T.S.

B 3/4 4-3 3/4.4.5 Consistent with Callamy T.S.

B 3/4 4-4 3/4.4.5 -

(bnaistent with Callamy T.S.

B 3/4 4-4 3/4.4.6.2 consistent with Callamy T.S.

B 3/4 4-7 3/4.4.9 (bnsistent with Callamy T.S.

B 3/4 4-8 3/4.4.9 Consistent with Callamy T.S.

a 4

2.____._._ - _ . - - - - - -

'1H:!NICAL SPECIFICATION CIWES Page Section Reason for 01anrJo B 3/4 4-9 3/4.4.9 Consistent with Callamy T.S.

B 3/4 4-14 3/4.4.9 0)nsistent with callamy T.S.

B 3/4 5-2 3/4. 5. 2, 3/4. 5. 3, consistent with callam y T.S.

& 3/4.5.4 B 3/4 5-2 3/4. 5. 5 & 3/4. 5.6 consistent with Ostiamy T.S.

B 3/4 6-1 3/4.6.1.2 consistent with 10CER50, Appenlix J B 3/4 6-2 2/4.6.1.4 Qansistent with C111amy T.S.

B 3/4 6-2 3/4.6.1.6 Consistent with C111amy T.S.

B 3/4 6-3 3/4.6.1.7 0)nsistent with C111amy T.S.

B 3/4 6-4 3/4.6.4 Consistent with Callamy T.S.

B 3/4 7-2 3/4.7.1. 2 consistent with Cs11amy T.S.

B 3/4 7-4 3/4.7.5 Wolf Creek specific justification providel B 3/4 7-4 3/4.7.6 Consistent with Cillamy T.S.

B 3/4 7-4 3/4.7.7 cassistent with Cs11amy T.S.

B 3/4 7-5 3/4.'7.8 Consistent with Callamy T.S.

B 3/4 7-6 3/4. 7.8 Wolf Creek specific justification provided B 3/4 7-7 3/4.7.10 Consistent with Callamy T.S.

B 3/4 0-1 3/4.8.1, 3/4.8.2, Consistent with C111am y T.S.

& 3/4.8.3) n 3/4 8-2 3/4.8.1, 3/4.8.2, 0)nsistent with CT11amy T.S.

& 3/4.8. 3 B 3/4 G-3 3/4.8.4 Cbnslatent with C111amy T.S.

B 3/4 9-1 3/4.9.1 Justiflcation providet by RiumC 84-034 datal 3/16/04

TEQNICAL SPECIFICATION CIANCIE l

Psgo Section Posson for Qiange B 3/411-1 3/4.11.1.1 Cbnaistent with C111am y T.S. .

B 3/411-1 3/4.11.1.2 Granmatical Correcticn B 3/411-2 3/4.11.1.2 0)nsistent with callamy T.S.

B 3/4 11-2 3/4.11.1.3 Consistent with callamy T.S.

B 3/411-3 3/4.11.1.3 consistent with Ctllamy T.S.

B 3/4 11-3 3/4.11.2.1 0)nsistent with Calismy T.S.

u V4114 3/4.11.2.1 Consistent with Otilam y T.S.

B 3/4 11-4 3/4.11.2.2 Consistent with Callamy T.S. an1 per discussions with NRC rwiomr (Dob Foll) -

B 3/4 11-5 3/4.11.2.3 Consistent vith C111awy T.S. aryl per discussions with NRC reviemr (Dob Foll)

B 3/4 11-6 3/4.11.2.4 Consistent with Cs11amy T.S.

B 3/411-6 3/4.11.2.5 Cbnsistent with C111am y T.S.

B 3/4 11-6 3/4.11.2.6 Consistent with Callamy T.9.

B 3/412-1 3/4.12.1 Per discusslons with NRC revieme (Mike Warv31er)

B 3/4 12-1 3/4.12.2 Per discussions with NRC rwlomr (Mike Murylor)

B 3/4 12-2 3/4.12.2 Typo 5-1 5.1. 3 Q)nsistent with Cs11am y T.S.

5-3 Figure 5.1-2 Wolf Crmk specific justification provido1 5-4 Figure 5.1-3 Malf Crmk crncific justification i provblol 5-5 Figure 5.1-4 Wolf Crook specific justification provido1

TECNICAL SPECIFICATION CIMR3ES Page Section Reason for 01ango 5-6 5. 3.1 0)nsistant with Cs11am y T.S.

6-1 6.2.2.e Wbif Crook specific justifiestion providel 6-4 Figure 6.2-3 Wolf Crook specific justification

(** noto) 6-6 6. 2. 3.1 Cbnsistent with C111amy T.S. anct EE onJanization .

6-6 6. 2. 3.4 Consistent with GE onJantLstion 6-9 6.5.1.0 Cbnsistent with 0111amy T.S. nrv1 E E onJantLstion 6-10 6.5.2.2 0)nslatent with G E organization 6-13 6.8.1 0)nalatent with C111amy T.S.

6-14 6.0.1 (* noto) Consistent with GE organimition an1 paticios 6-14 6.0.1 9 03nsistent with CT11amy T.S.

6-14 6.9.2.c ~ Osnalatent with DE orjaniLitton an1 polictos 6-15 6.0. 3.c consistent with M E onlantsitLon

• an1 p)ticion ,

6-15 6.0.4.a Consistent with Cs11am y T.S.

6-10 6.9.1.b Qansistent with Ot1 Limy T.S.

6-19 6.9.1.7 8. Por dlu:ussini 'vith tmc review >r 6.9.1.7 (

• n>to) (Miko Wtrv31or) 6-24 6.15.1.a Cbnaistent with ME onJanization thto ' The organizations dopictml in the Ibtf Crook " Proof arvl PnvioV' copy of Tmhnical Spocificalonn in Piquro 6.2-1 (pl. 6-3) arvi 6.22 (p).

6-4) are na timy oxistol (br Walf Crook Operations in 12/03.

Ilwavor, the rn:)nt rocent onjantL1tions are ntown in the W)1f Crmk PSAR nLto nidonhn. In ordor to rahr:o tho ntmtwr of chsrijos noahnt

. to tw m.21o to the Twhnical Femification figuren intw<wn now arit funt Icwl, thono figuros will tw uptatol just prior to fuel, lorv1.

i

> PR0l!F & [3M COPY TECHN! CAL SPECIFICATIONS fS8 WOI.F CREEK - UNIT 1 DOCKET NO, 50 482 S

e

' e i

I

PRODF & REVIEW COPY DEFINITIONS SECTION PAGE

1. 0 DEFINITIONS 1.1 ACTI0N........................................................ 1-1 1.2 A CTUATI O N LO G I C T EST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.3 ANALOG CHANNEL OP ERATIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

1. 4 AXIAL FLUX DIFFERENCE......................................... 1-1 1.5 CHANNEL CALIBRATION........................................... 1-1

1. 6 CHANNEL CHECK................................................. 1-1 1.7' CONTAINMENT INTEGRITY......................................... 1-2

1. 8 CONTROLLED LEAKAGE............................................ 1-2

1. 9 CORE ALTERATION............................................... 1-2 1.10 DOSE EQUIVALENT I-131........................................ 1-2 1.11 5-AVERAGE DISINTEGRATION ENERGY.............................. 1-2 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.13 FREQUENCY N0TATION........................................... 1-3 1.14 IDENTIFIED LEAKAGE........................................... 1-3 1.15 MASTER RELAY TEST ........................................... 1-3 1.16 MEMBER (S) 0F THE PUBLIC...................................... 1-3 1.17 0FFSITE DOSE CALCULATION MANU/L.~............................. 1-4 1.18 OPERABLE - OPERABILITY......................,.... ........... 1-4 1.19 OPERATIONAL MODE - HCDE...............'...... .... ...........

1-4

...... ..~................. : ........

1.20 PHYSICS TESTS........... 1-4 1.21 PRESSURE BOUNDARY LEAKAGE....T............................... 1-4 1.22 PROCESS CONTROL PROGRAM. . . . . . . .

~

. ..'.......................... ~

1-4 1.23 PURGE - PURGING.......: 2 . . ' . . ' . . . . ............................ 1-4 1.24 QUADRANT POWER TI'LT RAT 10 . . . . . .... .. ..... .............

x 1-5 s

1.25 RATED THERMAL POWERr ................... ..... .............. 1-5 1.26 REACTOR TRIP. SYSTEM RESPONSE TIhs.... .......... ............ ~

1-5 1.27 REPORTABL'E EVENT......... ,

9........ .......... 1-5 1.28 SHUTDOWN MARGIN;..... ......... ........ - . . ............... .

1-5 1.29 SITE BOUNCARY....... ..... ........ ............... ......... 1-5 1.30 SLAVE RELAY TLST.. ... . ...........'.. ..... .. ... ........ 1-5 1.31 S O L I D I F I CAT 10 N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '. : . . . . . . . . . . . . . 1-5

./ .

WOLF CREEK - UNIT 1 I a

W4e*

N 7

r ,. , . , . , , , , - - - - - - . ,

PRODF & RBM COPY DEFINITIONS SECTION PAGE DEFINITIONS (Continued) 1.32 SOURCE CHECK................................................. 1-6 1.33 STAGGERED TEST BASIS......................................... 1-6 1.34 TH ERMA L P0W E R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.35 TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-6 1.36 UNIDENTIFIED LEAKAGE......................................... 1-6 1.37 UNRESTRICTED AREA............................................ 1-6 1.38 VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-6 1.39 VENTING...................................................... 1-7 1.40 WASTE GAS HOLDUP SYSTEM...................................... 1-7 TABLE 1.1 FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 OPERATIONAL M0 DES....................................... 1-9 I

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i l WOLF CREEK - UNIT 1 II

, . - - - . - . , . , y. , y

PROD: & EM CD?Y SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 FIGURE 2.1-2 (BLANK)............................................... 2-3

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-4 YABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-5 BASES SECTION .

PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COO LANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-2

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...... ..... .. B 2-3 l .

WOLF CREEK - UNIT 1 III l

PROD:& REVEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg > 200*F........................... 3/4 1-1 Shutdown Margin - T,yg < 200 F........................... 3/4 1-3 Moderator Temperature Ccefficient............. .......... 3/4 1-4 Minimum Temperature for Criticali ty. . . . . . . . . . . . . . . . . . . . . . 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating................................... 3/4 1-8 Charging Pump - Shutdown................................. 3/4 1-9 Charging Pumps - Operating............................... 3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 Barated Water Sources - Operating........................ 3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEM8 LIES ,

G ro up H e i g ht. . . . . . . . . . . . . . . . . . . . . . : . . . . . . . . . . . . . . . . . . . . . . ' 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00......................................... 3/4 1-16 Position Indication Systems - Operating. . . . . . . . . . . . . . . . . . 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Drop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits............................. 3/4 1-21 FIGURE 3.1-1 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER-FOUR LOOP OPERATION.................. 3/4 1-22 FIGURE 3.1-2 ( B LAN K ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-23 WOLF CREEK - UNIT 1 IV

1 R100F & IEEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..............................._.... 3/4 2-1 FIGURE 3'.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER................................ 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R............................. 3/4 2-4 FIGURE 3.2-2 K(Z)-NORMALIZED Fq (Z) AS A FUNCTION OF CORE HEIGHT... 3/4 2-5 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R................................................. 3/4 2-8 FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN OPERATION................... ............. 3/4 2-9 3/4.2.4 QUdDRANTPOWERTILTRATI0................................ 3/4 2-11 3/4.2.5 DNB PARAriTERS........................................... 3/4 2-14 TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-15 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.......... ........... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................................ 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.................'....................... 3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM.

INSTRUMENTATION............................. ....... 3/4 3-14 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION, TRIP SETP0INTS...................... 3/4 3-22 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-29 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l INSTRUMENTATION SURVEILLANCE REQUIREMENTS..... ..... 3/4 3-34 l

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l WOLF CREEK - UNIT 1 V

I L. $ f LIMITING CONDITIONS FOR OPERATION AND. SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued) 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations................ 3/4 3-39 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS................................ 3/4 3-40 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS........................................ 3/4 3-42 Movable Incore Detectors ................................ 3/4 3-43 Seisaic Instrumentation......................... ........ 3/4 3-44 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-45 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-46 Meteorological Instrumentation........................... 3/4 3-47 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............. 3/4 3-48 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-49 Remote Shutdown Instrumentation.......................... 3/4 3-50 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-51 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION

. SURVEILLANCE REQUIREMENTS........................... 3/4 3-52 Accident Monitoring Instrumentation...................... 3/4 3-53 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION. . . . . . . . . . . . . . . . . . 3/4 3-54 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........................... 3/4 3-55 Chlorine Detection Systems............................... 3/4 3-56 Fire Detection Instrumentation........................... 3/4 3-57 TABLE 3.3-11 FIRE DETECTION INSTRUMENTS........................... 3/4 3-58 Loose-Part Detection System................... .... ..... 3/4 3-62 ,

Radioactive Liquid Effluent Monitoring Instrumentation.. . 3/4 3-63 ',

TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION................ ................... 3/4 3-64 WOLF CREEK - UNIT 1 VI r

PR00T & HM COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE s

INSTRUMENTATION (Continued)

TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-66 Radioactive Gaseous Effluent Monitoring Instrumentation. 3/4 3-68 TABLE 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.................................... 3/4 3-69 TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........... 3/4 3-72 3/4.3.4 TURBINE OVERSPEED PROTECTION............................. 3/4 3-75 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation.............................. 3/4 4-1 Hot Standby............................... .............. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold. Shutdown - Loops Filled................... ......... 3/4 4-5 Cold Shutdown - Loops Not Filled......................... 3/4 4-6 3/4.4.2 SAFETY VALVES Shutdown............................................... 3/4 4-7 0perating.............................................. 3/4 4-8 3/4.4.3 PRESSURIZER.......................'....................... 3/4 4-9 3/4.4.4 RELIEF VALVES............................................ 3/4 4-10 3/4.4.5 STEArt G EN ERATO RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-11 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO B. INSPECTED DURING INSERVICE INSPECTION......................... 3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TVEE INSPECTION.......- ........... 3/4 4-17 l 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l Leakage Detection Systems................................ 3/4 4-18 Operational Leakage.. ........................ .......... 3/4 4-19 l

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WOLF CREEK - UNIT 1 VII l __ _

P100F & 13M COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 0 .

TABLE 3.4-1 REACTOR C0fLANT SYSTEM PRESSURE ISOLATION VA LV E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-21 --

3/4.4.7 CHEMISTRY................................................ 3/4 4-22 TABLE 3.4-2 REACTOR COO LANT SYSTEM CHEMISTRY LIMITS. . . . . . . . . . . . . . . 3/4 4-23 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS........................................ 3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >

1 pCi/ GRAM 00SE EQUIVALENT I-131................. 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND A NA LYS I S P R0 G RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-28 3/4.4.9 PRESSURE / TEMP.ERATURE LIMITS Reactor Coolant System................................... 3/4 4-29 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY.................... ...... 3/4 4-30 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 16 EFPY........................... 3/4 4-31 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE -

PROGRAM - WITHDRAWAL SCHEDULE.'...................... 3/4 4-32 Pressurizer.............................................. 3/4 4-33 Overpressure Protection Systems.......................... 3/4 4-34 FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD

'0VERPRESSURE MITIGATION SYSTEM.................~.... 3/4 4-36 3/4.4.10 STRUCTU RA L I NT EG R I TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-37 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.......... .................................. 3/4 5-1 WOLF CREEK - UNIT 1 VIII

t Pi!DSF & IEEW COPY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.5.2 350*F...........................

ECCS SUBSYSTEMS - T,yg > 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350 F........................... 3/4 5-7 3/4.5.4 ECCS SUBSYSTEMS - T avg 200*F........................... 3/4 5-9

-3/4.5.5 00 ROM I"JCCTIO" SYSTEF - SOROM IM3ECTION T""%...

........ 3/' S-10 ic

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3/4.5.E REFUELING WATER STORAGE TANK............................. 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity....................................

3/4 6-1 Containment Leakage......................................

3/4 6-2 Containment Air Locks.................................... 3/4 6-4 I7ternal Pressure........................................ 3/4 6-6 Air Temperature..........................................

3/4 6-7 Containment Vessel Structural Integrity.................. 3/4 6-8 Containment Ventilation System........................... 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.................................. 3/4 6-13 Spray Additive System.........'........................... 3/4 6-14 Containment Cooling System.......,........................ 3/4 6-15 3/4.6.3 CONTAINMENT ISOLATION VALVES............. ............... 3/4 6-16 TABLE 3.6-1 C0tlTAINMENT, ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen":r-c.(7d1.8'.# . . ................... ......... 3/4 6-30 HydrogenControlSysEemsI.............................. 3/4 6-31 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................... ..... ...... 3/4 7-1 WOLF CREEK - UNIT 1 IX

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I P300F & KE COPY l e l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued)

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION.......................................... 3/4 7-2 TABLE 3.7-2 ( B LA N K ) . . . . . . . . . . . . . . : . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-2 TABL'E 3.7-3 STEAM LINE SAFETY VALVES PER L00P..................... 3/4 7-3 Auxiliary Feedwater System............................... 3/4 7-4 Condensate Storage Tank.................................. 3/4 7-6 Specific Activity........................................ 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY S AMPLE AND ANALYSIS PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-8 Main Steam Li ne Isol ation Val ves. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4 7-11 3/4.7.4 ESSENTIAL SERVICE WATER SYSIEM........................... 3/4 7- 12 3/4.7.5 ULTIMATE HEAT SINK.................'...................... 3/4 7-13 3/4.7.6 CCNTROL ROOM EMERGENCY VENTILATION SYSTEM. . . . . . . . . . . . . . . . 3/4 7-14 3/4.7.7 EMERGENCY EXHAUST SYSTEM................................. 3/4 7-17 3/4.7.8 SNUBBERS................................................. 3/4 7-19 FIGURE 4.7-1 SAMPLING PLAN 2) FOR SNUBBER FUNCTIONAL TEST......... 3/4 7-24 3/4.7.9 SEALED SOURCE CONTAMINATION.. ........................... 3/4 7-25 WOLF CREEK - UNIT 1 X

PROOF & REH COPY l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE PLANT SYSTEMS (Continued) 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System............................ 3/4 7-27 Spray and/or Sprinkl er Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-30 Halon Systems............................................ 3/4 7-32 Fire Hose Stations....................................... 3/4 7-33 TABLE 3.7-5 FIRE HOSE STATIONS............. ...................... 3/4 7-34 3/4.7.11 FIRE BARRIER PEN ETRATIONS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-36 3/4.7.12 AREA TEMPERATURE MONITORING.............................. 3/4 7-37 TABLE 3.7-6 AREA TEMPERATURE MONITORING........................... 3/4 7-38 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES -

Operating...................................... ......... 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE. . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-7 Shutdown................................................. 3/4 8-8 3/4.8.2 0.C. SOURCES 0perating................................................ 3/4 8-9 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..../................ 3/4 8-11 Shutdown................................................. 3/4 8-12 3/4.8.3 ONSITE POWER DISTRIBUTION Operating................................................ 3/4 8-13 Shutdown................................................. 3/4 8-15 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES

• Containment Penetration Conductor Overcurrent Protective Devices..................................... 3/4 8-16 TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES................ ..... 3/4 8-18 WOLF CREEK - UNIT 1 XI

PR00c & E!EV6C0'Y LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE ELECTRICAL POWER SYSTEMS (Continued) 4*rta o Meter. Opec &ted Valves Thermal Ovtricad Prot;; tion and Sypa;; 0cvi;;;.................. .................. 2/4 0 2:

TABLE 0.0-2 MOTOR OPERATED VALVES T"ERMAL OVERLCAO PROTE0 TION ?N0/OR SYPASS DEVICES.................... 3/4 8-26 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION...................................... ... 3/4 9-2 3/4.9.3 DECAY TIME.................................... .......... 3/4 9-3 3/4.9.4 CGaTAINMENT BUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 COMMUNICATIONS........................................... 3/4 9-5 3/4.9.6 REFUELING MACHINE........................................ 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY............. . 3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level......................................... 3/4 9-9 Low Water Leve1.......................................... 3/4 9-10 3/4.9.9 CONTAINMENT VENTILATION SYSTEM........................... 3/4 9-11 3/4.9.10 WATER LEVEL - REACTOR VESSEL Fuel Assemblies.................... ..................... 3/4 9-12 Control Rods....................................... .. 6 3/4 9-13 3/4.9.11 WATER LEVEL - STORAGE POOL .............................. 3/4 9-14 3/4.9. 12 S P ENT FU EL ASS EMB LY ST0 RAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-15 e FIGURE 3.9-1 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2........................................ 3/4 9-16 WOLF CREEK - UNIT 1 XII

- PR09? & REV H 09PY LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE REFUELING OPERATIONS (Continued) 3/4.9.13 EMERGENCY EXHAUST SYSTEM.......................... ...... 3/4 9-17 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4.10.3 PHYSICS TESTS............................................ 3/4 10-3 3/4.10.4 R EACTO R CO O LANT L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11 1 TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM................ ............... 3/4 11-2 0ose..................................................... 3/4 11-5 Liquid Radwaste Treatment System......................... - 3/4 11-6 Liquid Holdup Tanks...... . ............................ 3/4 11-7 3/4.11.2 GASEOUS EFFLDENTS Dose Rate................................................ 3/4 11-8 TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PR0 GRAM................................... 3/4 11-9 Dose-Nob.le Gases......................................... 3/4 11-12 Dose-Iodine-131 and 133, Tritium and Radioactive Material in Particulate Form............... 3/4 11-13 Gaseous Radwaste Treatment System........................ 3/4 11-14 Explosive Gas Mixture.................................... 3/4 11-1S Gas Storage Tanks........ ............................. . 3/4 11-16 WOLF CREEK - UNIT 1 XIII e -- -s. ., - - - _ . . - - - - - - - - - , - - , , - - , , , . , - -

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l PROD & REGl COPY i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE RADIOACTIVE EFFLUENTS (Continued) 3/4.11.3 SO LID RAD IOACTIVE WASTES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-17 3/4.11.4 TOTAL 00SE............................................... 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM...... . 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMP LES. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-9 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL -

SAMPLE ANALYSIS.................................... 3/4 12-10 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... 3/4 12-14 WOLF C:.EEK - UNIT 1 XIV

1 PRODi & HEW COPY BASES SECTION PAGE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVAB LE CONTRO L ASS EMB LIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 3/4.2.1 AXIAL FLUX 0IFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/4 2-2 FIGURE B 3/4.2-1 ~ TYPICAL INDICATED AXIAL FLUX OIFFERENCE VERSUS THERMAL P0WER.................................. B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 3/4.2.5 ONB PARAMETERS............................................ B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3

.Y4 . 3. 4 Tuwt ewoete o pccrecnoV ) -

3[47 REACTOR 30 LANT SYSTEM ~~'~~~y 3/4.4.1 REACTOR COOLANT LOOPS +ND COOLANT CIRCULATION............. B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-2 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEFV5LVES............................................. B 3/4 4-3 WOLF CREEK - UNIT 1 XV

BASES SECTION PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.......................... B 3/4 4-10 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL PCWER SERVICE LIFE........................ B 3/4 4-11 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RT NDT FOR REACTOR VESSEL STEELS EXF0 SED TO IRRADIATION AT 550*F........................... B 3/4 4-12 3/4.4.10 STRUCTURAL INTEGRITY..................................... B 3/4 4-14 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 A C C UMU LATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

B 3/4 5-1 3/4.5.2, 3/4.5.3, and 3/4.5.4 ECCS SUBSYSTEMS..................... B 3/4 5-1 2/4.5.5 "C CC N I NJ :T!ON SY ST:". . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . " 2/? 5-2 4

3/4.5.5 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 OEPRESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 3/4.6.4 COMSUSTIBLE GAS CONTR0L................................... B 3/4 6-4 WOLF CREEK - UNIT 1 XVI

PROOF & RPi:EYl COPY BASES SECTION PAGE 3/4.7 PLANT SYSTEMS

. 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM............................ B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK........................................ B 3/4 7-3 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................. B 3/4 7-4 3/4.7.7 EMERGENCY EXHAUST SYSTEM.................................. B 3/4 7-4 3/4.7.8 SNUBBERS.................................................. B 3/4 7-5 3/4.7.9 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS.................................. B 3/4 7-6 3/4.7.11 FIRE BARRIER PENETRATIONS......................... ....... B 3/4 7-7 3/4.7.12 AREA TEMPERATURE MONITORING............................... B 3/4 7-8 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION............................... B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION DEVICES. . . . . . . . . . . . . . . . . . B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION..................... ................. B 3/4 9-1 l

3/4.9.2 INSTRUMENTATION...................................... .... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS. . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 i

3/4.9.5 COMMUNICATIONS. ........................... ..... . .. . B 3/4 9-1 L

l WOLF CREEK - UNIT 1 XVII l

PRODF & REV;EVI COPY ,

BASES SECTION- PAGE REFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING MACHINE......................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY. . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION. . . . . . . . . . . . . B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION SYSTEM............................ B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and S TO RA G E P0 0 L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-3 3/4.9.12 SPENT FUEL ASSEMBLY ST0 RAGE............................... B 3/4 9-3 3/4.,9.13 EMERGENCY EXHAUST SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10. 4 REACTO R COO LANT L00 P S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS......................................... B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS................ ....................... B 3/4 11-3 3/4.11.3 SOLID RACI0 ACTIVE WASTES................................. B 3/4 11-7 3/4.11.4 TOTAL 00SE............................................... B 3/4 11-7 3/4.12 RADI0 ACTIVE ENVIRONMENTAL MONITORING .

3/4.12.1 MONITORING PR0 GRAM....................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS........................'.... ... ........ B 3/4 12-1 3/4.12.3 - INTERLABORATORY COMPARISON PR0 GRAM....................... B 3/4 12-2 WOLF CREEK ~ UNIT 1 XVIII

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DESIGN FEATURES SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1. 2 LOW POPULATION Z0NE......................................... 5-1 5.1. 3 MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS. . . . . . . . . . . . . . 5-1 FIGURE 5.1-1 EXCLUSION AREA.......................................

5-2 FIGURE 5.1-2 LOW PO P U LATIO N 20N E. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 FIGURE 5.1-3 SITE BOUNDAP.Y FOR GASEOUS EFFLUENTS.................. 5-4 FIGURE 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS................... 5-5 5.2 CONTAINMENT 5.2.1 CONFIGURATION............................................... 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES............................................. 5-6 5.3.2 CONTROL ROD ASSEMBLIES...................................... 5-6 5.4 REACTOR COOLANT SYSTEM .

5.4.1 DESIGN PRESSURE AND TEMPERATURE. . . . . . '. . . . . . . . . . . . . . . . . . . 5-6 ....

5.4.2 v0tuME...................................................... 5-6 5.5 METEOROLOGICAL TOWER LOCATION................................. 5-6 5.6 FUEL STORAGE 5.6.1 CRITICALITV................................................. 5-7 5.6.2 DRAINAGE........ ............ ............................ . 5-7 5.6.3 CAPACITY.................................................... 5-7 FIGURE 5.6-1 MINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2........................................ 5-8 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-7 TABLE 5;7-1 COMPONENT CYCLIC OR TRANSIENT LIMIT............... . 5-9 WOLF CREEK - UNIT 1 XIX A

I PR007 & RNEl CD?Y ADMINISTRATIVE CONTROLS  !

i SECTION PAGE 6.1 RESPONSIBILITY................................................ 6-1 6.2 ORGANIZATION -

6.2.1 0FFSITE..................................................... 6-1 6.2.2 UNIT STAFF.................................................. 6-1 FIGURE 6.2-1 0FFSITE ORGANIZATION................................. 6-3 FIGURE 6.2-2 UNIT ORGANIZATION.................................... 6-4

-TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION........................ 6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function.................................................... 6-6 Composition................................................. 6-6 Responsibilities............................................ 6-6 Records..................................................... 6-6 6.2.4 SHIFT TECHNICAL ADVIS0R..................................... 6-6 6.3 UNIT STAFF QUALIFICATIONS..................................... 6-6 6.4 TRAINING...................................................... 6-7 6.5 REVIEW AND AUDIT 6.5.1 PLANT SAFETY REVIEW COMMITTEE (PSRC)

Function.................................................... 6-7 Composition... ............... ................... ......... 6-7 Alternates.................................................. 6-7 Meeting Frequency................................... ....... 6-8 Quorum........................................... .......... 6-8 Responsibilities........................................... 6-8 Records........................ ................... ........ 6-9 WOLF CREEK - UNIT 1 XX

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PR00F & REVIEW COPY ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE (NSRC)

Function.................................................... 6-9 Cosposition................................................. 6-10 Alternates.................................................. 6-10 Consultants................................................. 6-10 Meeting Frequency........................................... 6-10 0uorum...................................................... 6-10 Review......................................................

6-11 Audits.................................................... . 6-11 Records..................................................... 6-12 6.6 REPORTABLE EVENT ACTI0N........'............................... 6-13

6. 7 S A F ETY L IM I T VI O LATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-13 6.8 PROCEDURES AND PR0 GRAMS....................................... 6-13 6.9 REPORTING RE0UIREMENTS........................................ 6-15 6.9.1 ROUTINE REP 0RTS......................... ................... 6-15 Startup Report.............................................. 6-15 Annual Reports....................... .....'................. 6-16 Annual Radiological Environmental Operating Report.......... 6-17 Semiannual Radioactive Effluent Release Report. . . . ......... 6-18 Monthly Operating Report.................................... 6-19 .

Radial Peaking Factor Limit Report........ ................. 6-19 6.9.2 SPECIAL REP 0RTS............................................. 6-19 6.10 RECORD RETENTION............................ ............... 6-20 WOLF CREEK - UNIT 1 XXI -

PROD & RMEW CDPY ADMINISTRATIVE CONTROLS SECTION PAGE REPORTING REQUIREMENTS (Continued) 6.11 RADIATION PROTECTION PROGRAM.... ............................ 6-21 6.12 HIGH RADIATION AREA.......................................... 6-21 6.13 PROCESS CONTROL PROGRAM (PCP)................................ 6-22 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM).................

.... 6-23 6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RA0 WASTE TREATMENT SYSTEMS............ .... ..... 6-23 l

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! WOLF CREEK - UNIT 1 XXII L .

PRODF & lEW CD?Y 9

9 SECTION 1.0 OEFINITIONS l

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PRODF & EB'l CO?Y 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check,.as a minimum, of output devices.

ANALOG CHANNEL OPERATIONAL TEST

1. 3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a, simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy. .

AXIALFLUXDIFF5RENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION -

~ 1. 5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds with the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK . -

1. 6 A CHANNEL CHECK shall be the qualitative assessment of channel benavior during operation by observation. This determination shall include, where

, possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

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WOLF CREEK - UNIT 1 1-1 i

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COM DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when;

a. All penetrations required to be closed during accident conditions.

are either:

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)- Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

&cm

1. 8 CONTROLLED LEAKAGE shall be that seal water flow :appit : Je the reactor coolant pump seals. ___.

CORE ALTERATION

, 1. 9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

DOSE EOUIVALENT I-131 1.10 OOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be tnose listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of

_ each radionuclide'{n the reactor coolant at the time of samoling) of the sum oftheaverage/eteandgammaenergiesperdisintegration(inMeV)forisotopes,"")

b other than iodine $p th half-lives greater than 15 minutes, making up at least, '

95%ofthetots{nyniodineactivityinthecoolant.

WOLF CREEK - UNIT 1 1-2

DEFINITIONS PR0'F & EEW COPY l

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until th'e ESF equipment is capable of performing its safety function (i.e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmo' sphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.15 A MASTER RELAY TEST shall be the energi;:ation of each tilaster relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include i continuity check of each associated slave relay. -

MEMBER (S) 0F THE PUBLIC 1.16 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

WOLF CREEK - UNIT 1 1-3 s - - .

DEFINITIONS PR00!& RNRl COPY OFFSITE DOSE CALCULATION MANUAL -

1.17 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERA 8LE or have OPERABILITY when it is capable of performing its specified function (s'),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device tn perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power levn1, and average reactor coolant temperature soecified in Table 1.2.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR(g}2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE ,

1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM .

1.22 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and dete,rminations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State

, regulations, burial ground requirements, and other requirements governing tne disposal of the radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

WOLF CREEK - UNIT 1 1-4

OEFINITIONS QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRfP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHUTOOWN MARGIN 1.28 SHUTOOWN MARGIN shall be the instantaneous amount of reactivity by which tne reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE 80LNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST 1.30 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERA 8ILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION 1.31 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

4 WOLF CREEK - UNIT 1 1-5

DEFINITIONS Pi!0DF & REVEW COPY SOURCE CHECK -

1.32 A SOURCE CHECK'shall be the qualitative ase,essment of channel response when the channel sensor is exposad to"a source of increased radioactivity.

STAGGERED TEST iASIS -

1.33 A STAGGERED TEST EASIS shall consist of:

a. A test schedula for n sysicms, subsystems, trains, or other designated

'corrponents obtained by' dividing the specified test interval into n equal subintervalc, and '

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.34 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.35 A TRIS ~ACTOATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or' tr.ip functions. The. TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary,'of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

UNIDENTIFIEU LEAKAGE 1.36 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. -

• UNRESTRICTED AREA .

1.37 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of

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individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, com-mercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM qa\ w 1.38 A VENTILATION EXHAUST TREATMENT SYSTEM 4 any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in'effluants by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or partic-ulates from the gadeous exhaust stream prior to the release to the environment.

Such a systern is not considered to have any effect on noble gas effluents.

Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

WOLF CREEK - UNIT 1 1-6

DEFINITIONS 8 iL .

VENTING 1.39 VENTING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

WASTE GAS HOLDUP SYSTEM 1.40 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off-

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gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

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WOLF CREEK"- UNIT 1 1-7

TABLE 1.1 FREQUENCY NOTATION L!ah[ T NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

O At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

WOLF CREEK - UNIT 1 1-8 6

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l-TABLE 1.2 u L h hff OPERATIONAL MODES REACTIVITY  % RATED AVERAGE COOLANT MODE CONDITION, K THERMAL POWER

• TEMPERATURE ff

1. POWER OPERATION > 0.99 > 5% > 350*F

2. STARTUP > 0.99 1 5% > 350*F

3. HOT STANDBY < 0.99 0 > 350 F

4. HOT SHUTDOWN < 0.99 0 350*F > T

> 200*F avg 5 .' COLD SHUTDOWN < 0.99 0 1 200*F

6. REFUELING ** < 0.95 0 1 140*F

• Excluding decay heat.

• • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

e WOLF CREEK - UNIT 1 1-9

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PROD? & REVH COPY  :

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SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS e

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.0, SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for four and three loop operation, respectively.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANDBY within if hourp, and comply with the requirements of Specification 6.7.1. 1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

%0LF CREEK - UNIT 1 2-1

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FIGURE 2.1-1 j REACTOR CORE SAFETY LIri!T .FCUR LOOPS IN OPERATION i

,.,,0 -L -t -u -n--.t .,,, - U,ll

. i 1 2-2 I

k

PR037 & RPE!W COPY I

l Figure 2.1-2 left blank pending NRC

, approval of three-loop oneration.

l 4

WOLF CREEK - UNIT 1 23 9 - , - . -w ,-

w.,- , - - - - - - - , - - - , e -w, - --e- .c,, -

p + - ,,~m---, ,-n,- -wwe,-

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PR005 & REB'l COPY 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As' shown for each channel in Table 3.3-1.

ACTION:

a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.

b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:

1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or

2. Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

i Equation 2.2-1 Z + R + 5 < TA Where: '

l Z= The value from Column Z of Table 2.2-1 for the affected channel, R= The "as measured" value (ir, percent span) of rack error for the affected channel, S= Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowances) of Table 2.2-1 for the affected channel.

I i

WOLF CREEK - UNIT 1 2-4

P c TABLE 2.2-1

] REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS h

X SENSOR 10TAL ERROR FUNCTIONAL UNIT All0WANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE 5

y

1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A. ,i H 2. Power Ran0e. Neutron Flux p z .3

d. Iligh Setpoint 7. 5 4.56 0 1109% of RTP* IL1k2% of RTP*

b. Low Setpoint - 8.3  : t.3 4.56 0 125% of RTP*

12kf% of RTP*

3. Power Range, Neutron Flux, J,<0 2.4 4 63 0.5 0 18% of RTP* with ifwe% of RTP* with liigh Positive Rate a time constant a time constant -

12 seconds 12 seconds

4. 4 t. 3 Power Range, Neutron Flux, J.A- L9 0.5 0 y liigh Negative Rate 15% of RTP* with $fwe% of RTP* with

• a time constant a time constant y

~>2 seconds ~

>2 seconds  :::.3 S. 35,9 O Intermediate Ran0e. 17.0 8.41 Neutron Flux 0 125% of RTP* 131% of RTP* M Is 90

6. Source Range, Neutron. Flux 17.0 10.01 0 1105 cps it,4 x 105 cps 2.t3 L. %

7. Overtemperature AT fvy' b.'8 _m 2<-79 (Idf See Note 1 See Note 2 r r1

8. Overpower AT 4X 5 6 sA3 3<5

a. M 0< f See Note 3 See Note 4 z.c ty 15 a

9. Pressurizer Pressure-Low 6 b7 8 5'n C

,,8 0.71 1<5 11980 psig 11886 psig

10. Pressurizer Pressure-Iligh 2 r) 24co

]. g 3 7.s 0.71 R$ 12385 psig 12396 psig

,% 93.9

11. Pressurizer Water Level-liigh 5dr V.O 2.18 14 192% of instrument SQk8% of instruinent span span

• RTP = RATED TilERMAL POWER

• • Loop design flow = 95,700 gpm 5

_ _ _ ____-_.____m__.____

'e, 75 TABLE 2.2-1 (Continued)

E n

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS .

]!?i '

SENSOR TOTAL ERROR c fullCTIONAL UlilT

- ALLOWANCE (TA) Z (S L TRIP SETPOINT ALLOWABLE VALUE q 12. Reactor Coolant flow-Low 2.5 T.17 ko o.6 o 1.6 190% of loop f

g 189.2% of loop design flow ** design flow"*

2.s.6 :2 t. s t 2.46 2 s.S 21.3

13. Steam Generator Water 30<o 2MB 1A Level Low-Low 23M of narrow 13J A of narrow range instrument range instrument span span 1.5 8.30 0.0 t o S 7 2 Velfs

14. Undervoltage - Peactor Sr0 ** r+ '

A. C .

p3 5S 4 y * *.

Coolant Pumps

>'^*' 's -

L..'.'n -

7.3 o.o o.o / T 1.2.  ; 67.1 He

15. Underfrequency - Reactor kr e +

, Coolant Pumps { >W llz e

16. Turbine Trip

> S84.62 p;

a. Low Fluid Oil Pressure N. A. - N.A. H.A. e y% 534.75p.I

b. Turbine Stop Valve N.A, #N Closure N.A. N.A. # #b

17. Safety Injection Input N.A. " N.A. N.A. N.A.

from ESF N.A. "I2

~3 C'3 C3 M *1 4 i . 52 o

• x3 V13 h

C3 e

C7 C.g"3 c

l 4 i

t

, f

.t MO. ;;'

$' f.. .- Nfl[

g'n .

. c,7 ri

' : :. P,;y(,

. ' Qv e;

t. <

6 M ULE 2.2-1 (Continued) '

E ,

(

n .

o REAC10R

' 1 RIP SYSlEM INSTRUNENIA110N TRIP SETPOINTS '

fv1

• - SfilSOR 10TAL 'O ERROR

• FlitiCTIOilAL UNIT (S) $I y 18. Reactor Trip System ALLOWANCE (TA) Z TRIP SETPOINT ALLOWABLE VALUE ,'

g Interlocks ~".

a. Intermediate Range 3 N.A. N.A. N.A. 11 x 10 10 amps l' Neutron Flux,.P-6 16 x 10 amps

b. Low Power Reac tor Trips '

Block, P-7

1) P-10 input ti. A.

E4N+86'55% ' '

11. A. N.A. po%'ofRIP* glJM,of RTP*

2) P-13 input I- of Krre izA/ grf

• ti. A. fl. A. N.A. 410%Aturbine

, ' <1M ofA turbi ne Impulse press'ure 4 equivalent, Impulse pressu're equivalent. ]

C3

c. Power Range Neutron N.A. H.A.

6. 3 . CC3 Flux, P-8 N.A.  :~<48% of RTP* ~<5Dd% of RTP*

D

.c s L $3.3 -

d.

Power Range Neutron i A. N.A.

x2 N. A.' k .30% of RTP*

Flux, P-9 ([5N2%ofRIP* Q-

> t 7 4, 6 83.d* ret

e. Power Range Neutron H.A. H.A. N.A q

Flux, P 10 ( po% of RIP * [Aof RTP*

C3

f. Turbine Impulse Chamber NiA. N.A. ft. A 2.4 ~TJ Pressure P-13 $10% of RIP

• turbine impulse

$1&2% of RTP* W.t turbine impulse pressure equivalent pressure equivalent

19. Reactor Trip Dreakers N,A. N.A. N.A N.A. H.A:

20. Automatic Trip and Interlock H.A.

Logic N.A. N.A. H.A. N.A.

l

• RIP = RATED 111ERNAL POWER '

l T

IABLE 2.2-1 (Continued) s-p TABLE NOIATIONS n

Q m

NOIE 1: OVERIEMPERATURE AT AT fly , (3f,,3)<AT,(K -K 2 ft [T (1, I

3) - T'] + K3 (P - P') - f (AI))

E U Where: =

' ],

AT Heasured AT by RID Manifold Instrumentation,

= lead-lag compensator on measured AT, f

ri, T2 =

Time constants utilized in lead-lag compensator for AT, 1 =8s, T2 = 3 S.

1

= L g c mpensator on measured AT, l 1+1 35

=

13 Time constant utilized in the lag compensator for AT, r3 = 0 s,

Indicated AT at RATED TilERMAL POWER, AT, K = 1.10, ot W K2

/ {,

'/

m

' =

c The function generated by the lead-lag compensator for T,yg O

, dynamic compensation, 3 C=.r.

t i, 13 =

Time 13=4s, constants utilized in the lead-lag compensator for T'#9, 1 4 = 28 s, L~3 4 T = Average temperature. *F, @M

~

r I

g,7 3

= Lag compensator on measured T av0' C"3 c3

~t3 is =

Time constant utilized in the measured T,yg lag compensator, is = 0 s, 6 O

c E 1ABlE 2.2-1 (Continued) c3 55 TABLE NOTATIONS (Continued) 1

. 92 140TE 1: (Continued) ~

l '

c: T* $ 588.5*F (Nominal T avg at RATED IllERHAL POWER),

2 I ;4i K3 = Ojo0671)

=

P Pressurizer pressure, psig, P' ' =

2235 psig (Nominal RCS operating pressure),

S = Laplace transform operator, s 8, and f (AI) is a function of the indicated difference between top and bottom detectors of tlee i

power range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARIUP tests such that:

t' .

(i) for q - gt between b -35% and + 7%, f (al) = 0, where qtand qbare percent RATED IllLRMAt POWER in the top and bottom halves of the core respectively, and q t+qb IS total IllERMAL POWER in percent of RATED lilERMAl. POWER; (ii) for each percent that the magnitude of q g "b exceeds -35%, the AT Trip Setpoint shall be automatically reduced by 1.26% of its value at RATED TilERMAL POWER; and

.g3 (iii) =t2 for each percent that the magnitude of yt gbexceeds +7%, the AT Trip Setpoint g shall be automatically reduced by 1.05% of its value at RATED IllERHAL POWER. "'1 4

!"2o N01E 2:  :=3

.The channel's f(3<s)% saxista Trip Setpoint shall not exceed its computed Trip Setpoint by more than of al span.

(s -3.6 i

, _O

~:

. CD "13 M

4, ,

1 TABLE 2.2-1 (Continued) s '

q TABLE NOTATIONS (Continued) k NOIE 3: OVERPOWER AT I E gy (1 + T,5) ( 1 ) I $ ) I 1

• -K (1 + T2 5) (1

• IsS) < ATo {K 4 5 1 + r,5 1+1 65 ) T - Ka [T (1 +sr 5) - I"] - f 2(AI)I E

q :4

- Where: =

AT Measured AT by RID manifold instrumentation; y , ,'3 = lead-lag compensator on measured AT;

=

ri, r2 Time constants utilized in lead-lag compensator for AT, r =8s,12 = 3 s; 1

Lag c spensator on measured AT; 1+1 35 y 13-

Time constant utilized in the lag compensator for AT, is = 0 s; o

AT = Indicated AT at RATED TilERMAL POWER; o s K, =  ;

K3 =

0. 2/ F for increasing average temperature and 0 for decreasing ave rage temperature;

=

3 3 The function generated by the rate-lag compensator for T,,g dynamic compensation; y

3 C=3 12 1

=

Time constant utilized in the rate-lag compensator for T,yg, r2 = 10 s;  % 4 3, g

= Lag compensator on measured T Co avu  ::::s is = -

Time constant utilized in the measured T,,g lag compensator, is = 0 s; q

C3 C"3 I 13 M

e l

t .

l g

1AutE 2.2-1 (Continued)

G n 1ABLE N0fATIONS (Continued) i A -

l p NOTE 3: -(Continued) .

=

i z K. 0.00128/*F for T > I" amt Ks = 0 for T $ 1";

! U ,3

=

\ ~

T Average temperature, "F; I

1" = Indicated I avg at RATED lilERMAL POWER (Calibration temperature for AT I

1 .

. Ins trinnentation,1 588.5*F);

S =

Laplace tras. storm operator, s 8; and

=

f2 (at) 0 for all al. ~

, N0lE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than w

a fje11)% of AT span.

w 3,5 i i l .

l 1

N N

O C.3 m 4

. CS er' l

i r.J l

b,

.r-v c-*

M c3

' O l t * "t3 6

8 M

I

• t Pi!D"? & !!BE COPY 4

t i

I e

BASES t

. FOR .

i SECTION 2.0  ;

SAFETY LIMITS AND L LIMITING SAFETY SYSTEM SETTINGS t

a b

i l

t t

I j . 1

+-

f l

I i

t i

I

i. r i

• i I

i l

3 4

i ,

6 i i

w+m' r vW r- w- ee- . _ _ _ __ _ . . _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ __ __

PC00F 8 REVIEW COPY I

l NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.

4

-. , . - . _ - - - ,_-_y, - - - - y .__ - , , . , - . - - - - - - - - --

2.1 SAFETY LIMITS h( h BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DN8 is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation (R-GRID). The W-3 DNB correlation (R-GRID) has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DN8 heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DN8 at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is. limited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that DN8 will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curvee of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average, temperature for whicn the minimum ONBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor,NF , of 1.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F g.at reduced power based on the expression:

N F = 1.55 [l+ 0.2 (1-P)]

Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allnwable control rod insertion assuming the axial power imbalance is within the limits of the f t(a!) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature aT trips will reduce the Setpoints to provide protection consistent with core Safety Limits. ,

WOLF CREEK - UNIT l- B 2-1

PR0 F & RMW COPY BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides CS contained in the reactor coolant from reaching the contain-ment atmospherec-The reactor vessel, pressurizer, and the RCS piping and valves are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.

g,,su tu. e - <y J*

The entire RCS is hydrotested attl107 ?:4;, 125%sof design pressure, to demonstrate integrity prior to initial operation.

( 50cp.3) 4 e

WOLF CREEK - UNIT 1 B 2-2

_.,-,--m%y_. , - - --,_.-.,.

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensu,re that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents. The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured 3etpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints nave been specified in Table 2.2-1. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combin-ation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equa-tion 2.2-1, Z + R + S < TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift' and the accuracy of their measurement. .TA or Total Allowance is the d.ifference, in percent span, between the' Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as measured" devia-tion, in percent span, for the affected channel from the specified Trip Setpoint.

S or' Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified i'n Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE OCCURRENCES.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the charnels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

'a0LF CREEK - UNIT 1 B 2-3

LIMITING SAFETY SYSTEM SETTINGS ht i BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore, providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of th.e Engineered Safety Features Actuation System.

Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.

Power Rance, Neutron Flux In each of the Power Range Neutron Flux channals there are two independent  !

bistables, each with its own trip setting used for a h'igh and [ow /ange trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

~

l The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power.

The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.

WOLF CREEK - UNIT 1 B 2-4 -

LIMITING SAFETY SYSTEM SETTINGS

}

BASES Intermediate and Source Ranoe, Ufk s i ? = Flux [ A us..~ /

The Intermediate and S a , "ur?c r Flux trips provide core protection during reactor TARTUP o mitigate the consequences of an un-controlled rod cluster contro -assembly bank withdrawal from a subcritical condition. These trips pro' vide-redundant protection tc the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 10s counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemoerature AT The Overtemperature aT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping

' transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the Pressurizer High and Low Pressure

.: trips.

The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to ,the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the-notations in Table 2.2-1. .

Overoower AT The Overpower AT Reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat caoacity of water, and (2) rate of change of temoerature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded. The Overpower aT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Sce+k."

( w.o.,

WOLF CREEK - UNIT 1 8 2-5

O

, LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer pressure ,

. es.ms w In each of the> pre eachwithitsown[ rip /ssurechannels,therearetwoindependentbistables, etting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blockad by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurize.r High Water Level trip is provided to prevent water relief through che pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full equivalent); and on increasing power, automatically reinstated by P-7. S.

p i.

Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consecuences of a . loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressurc at approximately 10%

of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 48% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

WOLF CREEK - UNIT 1 8 2-6 A 8

LIMITING SAFETY SYSTEM SETTINGS

{

BASES Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. -The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System.

Undervoltage and Underfrequency - Reactor Coolant Pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against ONB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltcge, the delay is set so that the time required for a signal to reach the Reactor trip break.ers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, che delay is set, so that the time required for a signal tu reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.

On decreasing power the Undervoltage and Underfrequency Reactor Coolint Pumo Bus trips are automatically blocked by P-7 (a power level of approximately 10*.

of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately .

10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of approximately 50% of RATED THERMAL POWER); n a'd on ir. creasing power, reinstated automatically by P-9.

Safety Injection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.

The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

WOLF CREEK - UNIT 1 8 2-7

• O LIMITING SAFETY SYSTEM SFTTINGS

{ { }

BASES Reactor Trio System Inter 1ceks The Reactor Trip System Interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manual block of the Source Range beester 'eip (i.e., prevents premature block of Source Range trip),

provides a backup block for Source Range Neutron Flux doubling, and allows deenergization of the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked.

'P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the single loop Low Flow trip.

P-9 On increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range Rose 6er trip and the Low Setpoint Power Range E:::::: trip; and automatically blocks the Source Range Reseecr trip and de-energizes the Source Range high voltage power. On decreasing power, the Intermediate Range nooetec trip and the Low Setpoint Power Range Russeor trip are automatically reactivated. Provides input to P-7.

P-13 Provides input to P-7.

b WOLF CREEK - UNIT 1 B 2-8 t

6 PR00F & Ram'l COPY t

l

(

/

SECTIONS 3.0 AND 4.0  !

l LIMITING CON 0!TIONS FOR OPERATION I AND  ;

SURVEILLANCE REQUIREMENTS i

i i

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9 9

9

PR007 & RMM COPY 3/4 i,IMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CON 0! TION FOR OPERATION 3.0.1 Compliance with the Limiting conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other '

conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within tne specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, ,

comoletion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within i hour action shall be initiated to place the. unit in a MODE in which the specification does not apply by placing it, as applicable, in:

a. At least HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. At least HOT SHUT 00WN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

c. At least COLD SHUTOOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Where corr'sctive measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to n.eet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. .

This Soecification is not applicable in MODE 5 or 6.

3.0.4 Entry into an OPERATIONAL kOE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reifance on provisions contained in the ACTION requirements. This (

provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifiestions.

WOLF CREEX - UNIT 1 3/4 0-1 1

_A..

^""'^"'""

PR00F & REVH COPY  !

SURVEILLAkCEREOUIREfENTS 4.0.1_ Surveillance Requirements shall be met during the OPERATIONAL MODES or other condit, ions specifico' for individual Limiting Conditions for Operatio'n unless ntherwise stated in an inc'ividual Surveillance Requirement.

4.0.2 Each Surveillanie R'e quirement shall b' e performeo witnin the specified time interval with:

~

a. A maximum allowable extansion not to exceed 25% of the surveillance interval, but

b. The combined time interval'for'any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

4.0.3 Failure to pe'rform-a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MCDE or ~other specified condition shall not be made unless the Surveillance' Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

4.0.5 Surveillance R'e quirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code' Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler

, and Pressure Ves:el Code and applicable Addenda as required by 10 CFR Part 50, Sectica 50.55a(g), except where. specific written relief has been granted by the Commission purs'u ant to 10 CFR Part 50, Section 50.55a(g)(6)(i)i ,

n WOLF CREEK - UNIT 1 3/4 0-2 46 e

^""'^""

PROSF & lHEW COPY SURVEILLANCE REQUIREMENTS (Continued)

b. Surveillance intarvals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME BOILER AND PRESSURE VESSEL REQUIRED FREQUENCIES FOR CODE AND APPLICABLE ADDENDA PERFORMING INSERVICE TERMIN0 LOGY FOR INSERVICE INSPECTION AND TESTING INSPECTION AND TESTING ACTIVITIES ACTIVITIES Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities;

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and

e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

WOLF CREEK - UNIT 1 3/4 0-3 emels

PR00F & HEW COPY 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 80 RATION CONTROL SHUTDOWN MARGIN - T,y >200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% ak/k for four loop operation.

APPLICABILITY: MODES 1, 2*, 3, and 4.

ACTION:

With the SHUTDOWN MARGIN less than 1.3% ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm baron or equivalent until the required SHUTDOWN MARGIN is restored.

d

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% ak/k:

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while,the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c. When in MODE 2 with K,ff less than 1, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Soecification 4.1.1.1.le. below, with the control banks at the maximum insertion limit of Specification -3.1.3.6; and

• SeeSpecialTestExceptionj3.10.1.

pJ'"

WOLF CREEK - UNIT 1 f 3/4 1-1 l

1

r REACTIVITY CONTROL SYSTEMS C C , ;n SURVEILLANCE REQUIREMENTS (Continued)

e. When in MODE 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 1% ak/k at least once per 31 Effective Full Powe- Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.le. above. The predicted reactivity values shall_be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.

S WOLF CREEK - UNIT 1 3/4 1-2

REACTIVITY CONTROL SYSTEMS I

SHUTDOWN MARGIN - T,yg 5,200 F MOU & RMW C0W LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1% ak/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than 1% ak/k, immediately initiate and continue boration at greater than.or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REOUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1% ak/k:

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is ipoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovaale or untrippable control rod (s); and

b. At l' east once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

1) Reactor Coolant System baron concentration,

2) Control rod position, i

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

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l j WOLF CREEK - UNIT 1 3/4 1-3

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT PROF & REMI COPY LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

a. Less positive than 0 ak/k/*F for the all rods withdrawn, begiryning of cycle life (80L), hot zero THERMAL POWER condition, e and l

b. Less negative than -4.1 x 10 4 ak/k/*F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2#* .

Specification 3.1.1.3b. - MODES 1, 2, and 3#.

ACTION:

a. With the MTC more positive than the limit of Specification 3.1.1.3a.

above, operation in MODES 1 and 2 may proceed provided:

~

, 1. Control red withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 ak/k/ F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> oi be in HOT STANDBY wi. thin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies tnat the MTC has been restored to within its limit for the all rods withdrawn condition; and ,

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all. rods withdrawn condition.

b. With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"With K,ff greater than or aqual to 1.

1. See Special Test Exception 10.3.

se _ . . -

WOLF CREEK - UNIT 1 3/4 1-4

REACTIVITY CONTROL SYSTEMS g} g , if SURVEILLANCE REOUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit of Specifica-tion 3.1.1.3a. , above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, and

b. The MTC shall be measured at any THERMAL POWER and compared to

-3.2 x 10 4 ok/k/*F (all rods withdrawn, RATED THERMAL POWER condi-tion) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 ak/k/ F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specification 3.1.1.3b. , at least once per 14 EFPD during the remainder of the fuel cycle.

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r WOLF CREEK - UNIT 1 3/4 1-5 l

I

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY

)hh h k [ hk LIMITING CCNDITION FOR OPERATION 3.1.1.

shall be4 The Reactor greater Coolant thcn or System equal to 551*F.lowest operating loop temperature (Tavg)

APPLICABILITY: MODES 1 and 2#*.

ACTION:

With a Reactor Coolant System operating loop temperature (Tavg) less than 551'F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

SURVEILLANCE REOUIREMENTS

4.1.1. 4

. bs greater The Reactor than Coolant or equal System to 551 F: temperature (T"V9) shall be determined to

a. Within 15 minutes prior to achieving reactor c'riticality, and

b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 561 F with the T,yg-T ref O(siation Alarm not reset. .

1. With K,ff greater than or equal to 1.

• See Special Test Exception 3.10.3.

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WOLF CREEK - UNIT 1 3/4 1-6

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-FLOW PATH - SHUTDOWN -

. .g.;t.#.%.

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. .* , LIMITING CONDITION FOR OPERATION

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' !TQ;3.1.2.1' As a minimum, one of the following boron injection flow paths shall

.s',n::Q:,.LM be

. .. source:

OPERABLE and capable of being powered from an OPERABLE emergency power

, , 9.;;q~ ,

a. A flow path from the Boric /"id Storage System via a boric acid 4.

. , i;N '. Qfy . ' ,, ./ .',-. -

' transfer pump and a centri *ugal charging pump to the Reactor Coolant

. ,. R. y

. System if the Boric Acid Storage System 3- S;'eci ficction 3.1. 2. Scr

, L.i ,. . ' Y

.is OPERABLEr-**- 45 gian h s g, /, o(-u g s,-n in, s m ;4: & % p : n a tion 3. i . 2 . n . +1 e Mcon 5 a~.I G 3..2..h.'4e . A c oe 4 ; e ,.

C ,1.< . b. The flow path from the refueling water storage tank via a centrifugal

". charging pump to the Reactor Coolant System if the ' refueling water ~

=-

3.l.2.gb,stora e tank in Specif-i etien 3.1.2.5b. is OPERABLE. as y ue- in % 6"h'^

de Xccts 5 ^ s<t te or 45 gh .3 in 5'pe. J6mhn 3.t. 2 .4 b . Me MDF4.

APPLICABILITY: MODES 4, 5, and 6.

4 ACTION:

1 J-};" kith none of the above' flow paths OPERABLE or capable of being powered from an L

OPERABLE emergency power soutce, suspend all operations involving CORE

,. ALTERATIONS or positive reactivity changes. .

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-;s SURVEILLANCE REQUI'REMENTS m ..f. ,, . -

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4.1.2.1~ At least one of the above required flow paths shall be demonstrated

, , f[, . , OPERABLE at least once per 31 days by verifying that each valve (manual, ^-

[

rJ , power-operated, or automatic) in the flow path that is not locked,

x. - sealed, or otherwise secured in position, fs in its correct position. . .

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WOLF CREEK - UNIT 1 '

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REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING U 'l t li . i LIMITING CONDITION FOR OPERATICN 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a. The flow path from the Boric Acid Storage System via a boric acid transfer pump and a centrifugal charging pump to the Reactor Coolant System, and

b. Two flow paths from the refueling water storage tank via centrifugal chargir.] pumps to the4K e tor Coolant System.

APPLICABILITY: MODES 1, 2, anc 3.*

ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y and borated to a SHUTDOWN MARGIN equivalent to at least t ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore a+. least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, -

sealed, or otherwise secured in position, is in its correct position; *

b. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and

c. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the Reactor Coolant System.

s y c Lkd d' WOLF CREEK - UNIT 1 3/4 1-8

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1

• The provisions of Specifications 3 0 4 and 4 0 4 are not applicable for entry '

into M00E 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. prior to the temperature of one or more of the RCS cold legs exceeding 375*F.

O I

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REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN r* }

LIMITING CONDITION FOR OPERATION 3.1.2.3 One centrifugal charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 4, 5, and 6.

ACTION:

" With no centrifugel charging pump ODERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REOUIREMENTS 4.1.2.3.1 The above required centrifugal charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that the pump develops a dif-ferential pressure of greater than or equal to 2400 psid when tested pursuant to Specification 4.0.5.

~

4.1.2.3.2 All centrifugal charging pumps, e~cluding x the above required OPERABLE pump, shall be demonstrated inoperable" at least once per 31 days, except when

! the reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.

• An inoperable pump may be energized for testing or for filling accumulators provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

WOLF CREEK - UNIT 1 3/4 1-9 L

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING hh h hh hff 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F within tne next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in GG4G SHUTDOWN within the nextJWhours. "

is SURVEILLANCE REOUIREMENTS l

l 4.1.2.4 At least two centrifugal charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that the pump develops a differential pressure of greater than or equal to 2400 psid when tested pursuant to Saecification 4.0.5. -

i 4 tda mers en .ne &

WOLF CREEK - UNIT 1 3/4 1-10

sY ~~~~~ '

• The pr

's ions of Specifications 3.0.4 and 4.0.4 are not applicable for entry \

into. d 3 for the centrifugal charging pump declared inoperable pursuant to Spect cation 4.1.2.3.2 provided that centrifugal charging pump is restored to operable statur within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more

, of the RCS cold legs exceeding 375' r.. ..

.. ,1.

e

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN i LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a. A Boric Acid Storage System with:

2ug

1) A minimum contained borated water volume of BMG gallons,

2) Between 7000 and 7700 ppm of baron, and

3) A minimum solution temperature of 65 F.

b. The refueling water storage tank (RWST) with:

is, q i >.

1) A minimum contained borated water volume of S&-SM gallons,

2) A minimum boron concentration of 2000 ppm, and

3) A minimum solution temperature of 37*F.

APPLICABILITY: MODES S and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REOUIREMENTS 4.1.2.5 T'he above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Verifying the contained borated water volume, and

~ *

3) Verifying the boric a'cid storage N solution temperature when it is the source of bora,ted water.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 37*F.

WOLF CREEK - UNIT 1 3/4 1-11

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING

{ {'"t r ,

LIMITING CONDITION FOR OPERATION

/

3.1.2.6 As a minimum, the following; borated water soJrce( )') shall be OPERABLE as required by Specification 3.1.2.2: O som i .a3 1 -~. J A ,o*q

, . .a.t w..e u 3 .,u . .sn 9. ca cui .u q .,.a , q q,....a 3 , 2.i c. 4 ;

a. A Boric Acid Storage System with:

n,sst

1) A minimum contained barated water volume of 16 442 gallons,

2) Between 7000 and 7700 ppm of baron, and

3) A minimum solution temperature of 65 F.

b. The refueling water storage tank (RWST) with:

ym , e c c

1) A minimum contained borated watEP Volume of 112,'75 gallons

2) Between 2000 and 2100 ppm of bor on,

3) A minimum solution temperature of 37 F, and

4) A maximum solution temperature of 100*F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION: .a 2' '. b " # 5

a. With the Boric Acid Storage System inoperabl and being used as one of the above required borated water sources', restore the storage system to OPERA 8LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOBY within the next 6 6:urs and borated to a SHUTDOWN MARGIN equivalent to at least 1% uK/k at 200 F; restore the Boric Acid Storage System to OPERABLE status within the nsxt 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

,,, su i, t , . ')

b. With the RWST inoperable (restore tne tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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, W , ,3 y SQ $ MTh .. . Od .T WOLF CREEK - UNIT l 3/4 1-12 l

REACTIVITY CONTROL SYSTEMS PRODF & E31 COPY SURVEILLANCE REQUIREMENTS

, a .- A 4.1.2.6 Each, borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the baron concentration in the water,

2) Verifying tiie contained borated water volume of the water source, and

3) Verifying the Boric Acid Storage System solution temperature when it is the souice of barated water.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when the outside air temperature is either less than 37*F or greater than 100*F.

l l

WOLF CREEK - UNIT 1 3/4 1-13 m

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT -

LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and contrel rods shall be OPERABLE and positioned within demand position.

12 st.eps (indicated position) of their group step counter APPLICABILITY: MODES 1* and 2*.

ACTION:

a. With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Soecification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With more than one full-length rod inoperable or misaligned from the group step counter demand position by more than t 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a. above, or misaligned from its group step counter demand height by more than 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod' is declared inoperable and the remainder of the roos in the group with the inoperable, rod aie aligned to within

, 12 steps of the inoperable rod while maintaining the' rod secuence and insertion limits of Figures 3.1-1 and 3.1-2. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; "See Special Test Exceptionst3 .10.2 and 3.10.3.

4 pa h

WOLF CREEK - UNIT 1 3/4 1-14

I REACTIVITY CONTROL SYSTEMS -

ri

!f g)y LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained frbm the movable incore detectors and Fq (Z) and F H ,are-ve ified to be within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />'; und d) The THERMAL POWER level is reduced to'less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod position deviation monitor is inoperable, then veriff the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. g,

4.1.3.1.2 ach full-length rod not fully inserted in the core shall be determined OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

o WOLF CREEK - UNIT 1 3/4 1-15

l 1

l 1

TABLE 3.1-1 PRODF a Hylny cgpy I ACCIDENT ANALYSES REOUIRING REEVALUATION IN THE EVENT OF AN IN0PERABLE FULL-

, LENGTH R00 Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at Full Power -

Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)

Major Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

I e

l WOLF CREEK - UNIT 1 3/4 1-16 e

a , - - -, - - * ,- - - , . - --

l l

i l i

i REACTIVITY CONTROL SYSTEMS l POSITION INDICATION SYSTEMS-OPERATING e Pi100r a llEVlEW COPY LIMITING CONDITION FOR OPERATION

~

s' 3.1.3.2 The Shu+d -a ead C rtr;' Rod Position Indication System and the Demand Position Indication System shall be OPERABLE and capable of deter- '

mining the control rod positions within i 12 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

6. M

a. With a maximum of one rod position indicator per bank inoperable either:

1. Determine the position of the nonindicating rod (s) indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Redui:e THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With a maximum of one demand position indicator per bank inoperable either: ,M

1. Verifythatall[odpositionindicatorsfortheaffectedbank

.are OPERABLE and that the most withdrawn rod and tne least withdrawn rod of the bank are within a maximum of 12 steps of

. each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or

- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. -

SURVEILLANCE REOUIREMENTS S') f Y 4.1.3.2 Each dod position indicator shall be determined AOP)dABLE by verifying that the Demand Position Ii.aication System and the Digital / Position Indication System agree within 12 steps at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the red position deviation monitor is inoperable, then compare the Demand Position Indication System and the Digital Position Indication System at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

g WOLF CREEK - UNIT 1 3/4 1-17

-p _ _ _ _ . - - _ _ _ _ . _ _ _ _ ,_- - - __

4 REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUTOOWN g

PP.03F & HEW CDP'i LIMITING CONDITION FOR OPERATION '

3.1.3.3 .One digital rod position indicator (excluding demand position indica-tion) shall be OPERABLE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3*#, 4*#, and 5*#.

ACTION:

With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.

SURVEILLANCE REOUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicator agrees with the demand position indicator within 12 steps wnen exercised over the full range of rod travel at least once per 18 months.

I "With the Reactor Trip System breakers in the closed position.

1. See Special Test Exception 3.10.5.

., ,./

. p' h(

O WOLF CREEK - UNIT 1 3/4 1-18 I

REACTIVITY CONTROL SYSTEMS ROD OROP TIME .

PR00F & ?EV1Et'l COPY LIMITING CONDITION FOR OPERATION l

3.1.3.4 The individual full-length shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T,yg greater than or equal to 551*F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With the rod drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and

c. At least once per 18 months.

D WOLF CREEK - UNIT 1 3/4 1-19 9 - ,-- - + . ,- , . - - - , , , . - - _ ,.- -- - - - - - , , - - - - - , - - -

REACTIVITY CONTROL SYSTEMS SHUTDOWN R00 INSERTION LIMIT

{ { [lf{"jl )

LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.

APPLICABILITY: MODES 1* and 2*#.

ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

a SURVEILLANCE REOUIREMENTS 4.1.3.5 Each shutdown rod shall be determined fully withdrawn:

a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or 0 during an approa,ch to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. .

${# gcab'*I

• See Special Test Exceptions . 10.2 and 3.10.3.

1. With K,7f greater than or equal to 1.

WOLF CREEK - UNIT 1 3/4 1-20 p- ,-m - , - - - - - - - ~ e v ,e --. oe -m- e s - --

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS

{]f} {. }))![ {}

LIMITING CONDITION FOR ODERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures 3.1-1 and 3.1-2.

APPLICABILITY: MODES 1* and 2*#. '

ACTION:

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

. b. Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figures, or

c. Be 'in at least HOT STANOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

J SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

S(" ppd

• See Special Test Exceptions 3.10.2 and 3.10.3.

1. Witn K,ff greater than or equal to 1.

WOLF CREEK - UNIT 1 3/4 1-21

1 l

PROD & REV'N COPY  !

(FULLY WITHDRAWN) 228 - =- --

. _. __ =.._.... ..

200

' / - =

f fBANKB ,.

1M ./

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=f f

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f. 160 (0 % 161) [ (100 % 161);

i 3 140

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_- BANK C 2

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100 l 7

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-[ BANK D 8

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( 30.2%

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0) .I-_-. -- .-

0 0 20 40 60 80 100 (FULLY INSERTED)

FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER-FOUR LOOP OPERATION WOLF CREEK - UNIT 1 3/4 1-22

m .

P3007 & REV;EW COPY Figure 3.1-2 left blank pending NRC approval of three-loop operation l

WOLF CREEK - UNIT 1 3/4 1-23

_ _ - _ , , - , - - - - ,,,g + ,----~---~*m~s' '*v~~" ' ' * * ^ " " ' " " " ~ ~ ' ^

3/4.2 POWER DISTRIBUTION LIMITS .

k 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference:

a. t 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU, and

b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTU.

The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indicated AFD is wi. thin the Acceptable Operation Limits of Figure 3.2-1 and the cumulative penalty deviation t,ime does not exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

! c The indicated AFD may deviate outside the above required target band at hee ter than 15% but less than 50% of RATED THERMAL POWER provided the cumulative alty deviation time does not exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A' PPLICA8ILITY: MODE 1 above 15% of RATED THERMAL POWER *.

ACTION:

a. With the indicated AFD outside of the above required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes, either:

1. Restore the indicated AFD to within the above required target band ifmits, or

2. Reduce THERMAL POWER to less than 90% of RATED. THERMAL POWER.

b. With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWERvr eWee;

1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and .

2. The Power, Range Neutron Fluxg - High"Setpoints to less than or equaVto 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

f ,.,, .. @ *

"See Special Test Exception /3.10.2.

1. Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits-of Figure 3.2-1. A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

WOLF CREEK - UNIT 1 3/4 2-1

L., ',

POWER DISTRIBUTION LIMIT 5 ff(( h hh @hy LIMITING CONDITION FOR OPERATION ACTION (Continued)~ ,

c. With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of cumulated penalty deviation time during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of

~ RATED THERMAL POWER until the indicated AFD is within the above required target band.

SURVEILLANCERE0blREMENTS-4.2.1.1 The indi.cated AFO shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 ddys when the AFD Monitor Alarm is OPERABLE,

~

and

2) At least once per hour for the first 24' hours after restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicated AF0 for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafT.er, when the AFD hionitor Alarm is inoperable., The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels me indicating the AFD to be outside the target band. Penalty' deviation outside of the above required target band shall be accumulated. on a time basis of:

a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target bend at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4. 2.1. 3 The target flux difference of each OPERABLE cxcore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4. 2.1. 4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.

WOLF CREEK - UNIT 1 3/4 2-2

Pi100F & ilETdVI COPY

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40 50 l FLUX DIFFERENCE ( Al)%

l FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER o WOLF CREEK - UNIT 1 3/4 2-3 t

-. - - .~ , . . . . _

. - _ . ~ , - --

I l

POWER DISTRIBUTION LIMITS PR07ER?#TlCOPY 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z) .

LIl!! TING CONDITION FOR OPERATION 3.2.2 F q(Z)_shall be limited by the following r.elationships:

FS (Z) $ [2.32] [K(Z)] for P > 0.5, and P

Fq (Z) 1 [4.64] [K(Z)] for P 5 0.5.

Where:

_ THERMAL POWER

- RATED THERMAL POWER , and K(Z) = the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1.

ACTION: '

With Fq (Z) exceeding its limiti

a. Reduce THERMAL POWER at least U for each HqF (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least U for each H F (Z) exceeds the limit; and 0

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. q t

i i

t i WOLF. CREEK - UNIT 1 3/4 2-4

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, .-l-WOLF CREEK - UNIT 1 3/4 2-5 l

~

POWER OISTRIBUTION LIMITS { L. -

SURVEILLANCE REOUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F shall be evaluated to determine ifqF (Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER;

b. Increasing the measured F xy component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties;

c. Comparing the F above to: xy c mputed (F ) obtained in Specification 4.2.2.2b. ,

1) The F xy limits for RATED THERMAL POWERx(FRTP) for the appropriate measured core planes given in Specifications 4.2.2.2e. and f.,

below, and

2) The relationship:

L ,p F

xy P [1+0.2(1-P)] -

Where F ' is the limit for fractional THERMAL POWER operation express d as a function of F RTP and P is the fraction of RATED x

THERMAL POWER at which F xy was. measured.

d. Remeasuring F xy according to the following schedule:

C N

.1) When F xy is greater than the F limit for the appropriate measured core plane but less than the F relationship, additional power distribution maps shall be taken a d F C compared to F x x and F,y either:

a) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which F, was last determined, or t

l b) At least once per 31 EFPD, whichever occurs first.

WOLF CREEK - UNIT 1 3/4 2-6 i

I!

POWER DISTRIBUTION LIMITS

• {} {'l.]

SURVEILLANCE REQUIREMENTS (Continued)

C

2) When the F y is less than or equal to the F limit for the appropriate measured core plane, additional power distribution maps shall be taken and F RTP xy compared to F x and F ' at least once per 31 EFPD.

e. The F RTP xy limits for RATED THERMAL POWER (Fxy ) shall be provided for all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.9;

f. The F limits of Specification 4.2.2.2e., above, are not applicable xy in the following core' planes regions as measured in percent of core height from the bottom of the fuel:

1) Lower core region from 0 to 15%, inclusive',

2) Upper core region from 85 to 100%, inclusive,

?

3) Grid plane regions at 17.8 2%, 32.1 2%, 46.4 2%, 60. 6 :: 2".

and 74.9 2%, inclusive, and

4) Core plane regions within 2% of core height-( 2.88 inches) about the bank demand position of the Bank "0" control rods.

g. With F"xy exceeding FL , the effects of F xy on Fq (Z) shall be evaluated to determine if qF (Z) is within its limits.

4.2.2.3 When qF (Z) is measured for other than F,y determinations, an overall measured F (Z) shall be obtained from a power distribution map and increased 9

by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

f

! WOLF CREEK - UNIT 1 3/4 7 l

PR0"F & REEEW COPY i POWER DISTRIBUTION LIMITS - - -

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown on Figure 3.2-3 for four loop operation.

Where:

N F

a. SH '

R = 1.49 [1.0 + 0.2 (1.0 - P)]

b' THERMAL POWER , and P

= RATED THERMAL POWER

c. F" = Measured values of-F obtained by using the movable incore detectors 'to obtain a power distribution map. The melsured values of F N shall be used to calculate R since Fidure 3 2-3 includes p- " H-- '- measurement uncertainties of 2-G% for flow and 4% for incore measurement of F" .

APPLICABILITY: MODE 1.

ACTICN:

With the combination of RCS total flow rate and R outside the region of -

acceptable operation shown on Figure 3.2-3:

. a. Within 2 hcurs either: .

1. Restore the combination of RCS total flow rate and R to within the above limits, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL PCWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

e WOLF CREEK - UNIT 1 3/4 2-8 O

MEASUREMENT UNCERTAINTIES OF 2.1% FOR FLOW Ah!D 4.0% FOR INCORE MEASUREMENT OF F4 U ARE INCLUDED IN THIS FIGURE 48 ...

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i F l - i 1 34 l l I-0.90 0.95 1.00 1.05 1.10 R = F4 N/1,49[1.0 + 0.2(1.0.P)] FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R FOUR LOOPS IN OPERATION l u J ch e. C^,LLn""Y - UNIT 1 3/4 2-9 -

                                                   -                                                                                                                                                                         i

1 1 POWER DISTRIBUTION LIMITS hh h [h hf LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED

                         ' THERMAL POWER within the next 2 hours; and

c. Identify and correct the cause of the out-of-limit condition prior 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 leve.ls:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED T'HERMAL POWER, and

3. Within 24 hours of attaining greater than or equal to 95% of '

RATED THERMAL POWER. l SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. , 4.2.3.2 The combination of indicated RCS total flow rate and R shall be determined to be within the region of acceptacle operation of Figure 2.2-3:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel 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 operatien of Figure 3.2-3 at least once per 12 hours when the most recently obtained '.alue 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 i CALIBRATION at least once per 18 months. 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. i N- i a.w ,  ;-

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WOLF CREEK - UNIT 1 3/4 2-10

1 POWER DISTRIBUTION LIMITS i l 3/4.2.4 QUADRANT POWER TILT RATIO O LIMITING CONDITION FOR OPERATION 1 l l 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. I APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *. ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but e less than or equal to 1.09:

1. Calculate the QUADRANT 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 THERMAL POWER.

2. Within 2 hours either:

a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT 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 RATI0'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 next .

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 power may proceed provided that the

                       . QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%

or greater RATED THERMAL POWER.

       "See special Test Exceptiong3 .10.2.              ,,

M' at .

 ~

WOLF CREEK - UNIT 1 3/4 2-11

                                                                  ,               -v m  - - -

POWER OISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT 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 THERMAL POWER.

2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each U: of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes;

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 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 j above 50% of RATED THERMAL POWER may proceed provided that the t

QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at SS% or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1. Calculate the QUADRANT 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 THERMAL POWER. WOLF CREEK - UNIT 1 3/4 2-12

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION PR03F & KW COPY ACTION (Continued)

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 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

3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER.

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and

b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUA'DRANT POWER TILT RATIO shall be determined to be within i.he limit when above 75% of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm that the nomalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. WOLF CREEK - UNIT 1 3/4 2-13

l I

                                                                                 )

POWER DISTRIBUTION LIMITS ~~- 3/4.2.5 DNS PARAMETERS fir 00F & 23,3 COPY LIMITING CONDITION FOR OPERATION 3.2.5 The following DNS related parameters shall be maintained within the limits shown on Table 3.2-1: a. Reactor Coolant System T,yg, and

b. Pressurizer Pressure.

APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the'next 4 hours. SURVEILLANCE REQUIREMENTS . 4.2.5 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours. e WOLF C9EEK - UNIT 1 3/4 2-14

T i TABLE 3.2-1 6 i q DN8 PARAMETERS O m I

     )

LlHITS Four Loops in Three Loops in z PARAMETER Operation Operation ! -i 512.5

     "                Reactor Coolant System I             -< fr%F
                                                                                     ^^

avg Pressurizer Pressure > 2220 h

• M.

z T ti; - N

A:3 C3

                                                      .                                                C3 v1 520

2 m

          " limit not applicable during either a IllERMAL POWER ramp in excess of ST of RAIED IllERMAL J POWER per minute or a TilERHAL POWER step in excess of 10% of RAllD TilERHAL POWER.

c3

         ^^lhese values left blank pending NRC approval of three-loop operation.                        N

3/4.3 INSTRUMENTATION PR00F & RDilEYv COPY 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Resctor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SUR'.'EILLANCE REOUIREMENTS 4.3.1.1 Each Reactor Trip Sys' tem instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1. 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total numoer of redundant channels in a specific Reactor trio function as shown in the

   " Total No. of Channels" column of Table 3.3-1.

WOLF CREEK - UNIT 1 3/4 3-1 Q.

TABLE 3.3-I g REACIOR IRIP SYSTEH INSTRUMENIATION G n HINIMUM g 10TAL NO. CllANNELS CllANNELS APPLICABLE p It!NCIl0NAL UNil 0F CilANNELS T0 sR P OPERABLE H0 DES ACTION [ 1. Manual Reactor Trip 2 1 2 1, 2 1 z a 2 1 2 3*, 4*, 5* 10 y 2. Power Range, Neutron Flux .

a. Ili0h Setpoint 4 2 3 1, 2 2#

h. Low Setpoint 4 2 3 l###, 2 2#

3. Power Range, Neutron Flux 4 2 3 1, 2 2#

liioh Positive Rate .

4. Power Range, Neutron Flux, 4 2 3 1, 2 2#

     ,        lii h 0 Negative Rate          -

s [. S. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3

6. Source Range, Neutron Flux

a. Startup 2 1 2 2##, , j 4

b. Shutdown 2 1 2 3*,4*,S* +& 5 e-Shuttlown -e- -G- - 3, 4, 5 7. Overtemperature AT v3

nC3

a. Four Loop Operation 4 2 3 1, 2 6#

b. Three Loop Operation ** ** ** ** **

f_.3

                              ,                                                                             --r1

8. Overpower AT QO m

a. Four Loop Operation 4 2 3 1, 2 6#

b. Three Loop Operation ** ** ** ** **

                                                                                                            %1 4

9. Pressurizer Pressure-Low -

4 2 3 1 6# cs c3

10. Pressurizer Pressure-liigh 4 2 3 1, 2 6# M

TABLE 3.3-1 (Continued) g REACTOR 1 RIP SYSTEM INSTRilHENTATION G MINIMUM 9 TOTAL NO. CilANNELS CilANt!ELS APPLICABLE y fullCTlatlAL UNIT OF CllAliNELS TO TRI'P OPERABL: MODES ACTION

11. Pressurizer Water Level-liigh 3 2 2 1 7#

7 12. Reactor Coolant flow - Low

a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 7#

any oper- each oper-

                                            .                 ating loop  ating loop

b. Two Loops (Above P-7 and 3/ -

{/loopin 2/ loop 1 75 below P-8) wo oper- ~e7eh oper-ating 100> ating loop euw 9..%3

                                         ,                                 to.g O 13. Steam Generator Water                  4/stm. gen.<    2/stm. gen. 3/stm. gen.      1, 2  6#

] Level-tow-Low , any o:7i. - Mioper- eug 9a,3

ating st ating sim. g, p Den. gen.

14. Iladervoltage-Reactor Coolant Pumps 4-2/ bus 2-1/ bus 3 1 (#

15. Underfrequency-Reactor Coolant.

Pumps 4-2/biis 2-1/ bus 3 1 6#

16. Turbine Trip CD

a. Low Fluid Oil Pressure 3 2 2 1 7# CC3

b. Turbine Stop Valve Closure 4 4 1 1 11# "

CO

17. Safety Injection Input m from ESt- 2 1 2 1, 2 9 y t

R d ca i CD N

k e ( 3. TADLE 3.3-1 (C6ntinued) g REAC10R TRIP SYSTEM INSTRtlMENI'ATION G n HINIMUM

o 101AL NO. CilANNELS CllANNELS APPLICABLE h IllflCIl0NAL Utill Of CilANNELS TO IRIP OPERABLE MODES ACTION

[ 18. Reactor Trip System Interlocks

 =                                            -
 --i        a. Intermediate Ranue
 -              Heutron Flux, P-6                      2              1            2       2##          8

b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 or -

P-13 Input 2 I . 2 1 8

c. Power Range Neutron y Flux, P-8 4 2 3 1 8 w

 <;>        d. Power Range Neutron
 +              flux, P-9                              4             2           3         1           8

e. Power Range Neutron flux, P-10 4 2 3 1, 2 ' 8

f. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8 m

a3

19. I?eactor Trip Breakers 2 1 2 C3 1, 2 9 ca 2 1 2 3*, 4*,

S h- m w 20 Automatic Trip and Interlock Logic 2 1 2 1, 2 9 m 2 1 2 3^, 4*, 5* 10 m

• Y cs C"3 T

M

1 1 i PR007 & RE3V COPY l TABLE 3.3-1 (Continued) TABLE NOTATIONS

                                                     ' ~

C.9 J. pps \ 94+th the Reactor Trip System breakers /in the closed position and the Control Rod Drive System capable of rod withdrawal.

 ** Values left blank pending NRC approval of three loop operation.
   #The provisions of Specification 3.0.4 are not applicable.
 ##Below the P-6 (Intermediate Ra.nge Neutron Flux Interlock) Setpoint.

1. 1. 1. 8elow the P-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: O

a. The inoperable channel is placed in the tripped condition within 1 hour;

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; and

c. Either, THERMAL POWER is restricted to less than or equal 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 Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status' prior to increasing THERMAL POWER above the P-6 Setpoint; and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

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. WOLF CREEK - UNIT 1 3/4 3-5

TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes. ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, verify cser. liens ;th the SuUT00W M^MGIF r;quir;;;ntr of Specificatien-3-E.-1cl-or 3.1.1.2, M app!!cabic, dthin-1-hcur and at-ic=t cece per- 12 hours thereafte r ..h a % ;~ p* <M' k na*\ to Cu%Ma" '"*O'" 4' D5 1

                               #   '

• N,

                                            ~
                                                  $_8b N* b-~-if N Y, rr ---S / 'i- 3. I' Y u ACTION 6 - Wlh'ThN.ii6er"of 0PGRABLE cIiainefs eEn'~less"Han'thTTistaf l"                  .'

Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 1 ho,ur; 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, STARYUP 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 condition within 1 hour.

     . ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 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* Minimum Channels OPERABLE requirement, be in at least HOT STANOBY 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 Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or cpen the Reactor trip breakers within the next hour. ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 1 hour. WOLF CREEK - UNIT 1 3/4 3-6

TAlllE 3.3-2 . , . 8 REACIOR TRIP SYSTEM INSTRl1HENIATION RESPONSE TIMES h FilflCIIONAL lJrilT g RESPONSE TIME

   . 1. Hanual Reactor Tri 2N. A.

b 2. Power Range, Neistron Flux-

 -4                                                                                      -< 0.5 second*

3. Power Range, lieutron Flux, liigt: Positive Rate N.A.

4. Power Range, ticutror Flux,

• liigt Negative Rate 5 0.5 second*

5. Intermediate Range, Neutron Flux

• m N.A. :D

6. Source Range,lieutron Flux c:3 N. n.,

Y -r1

7. Overtemperature al GO 5 6.0 seconds" O 11 . Overpower Al

                                                                                        $ 6.0 seconds" r1

9. Pressurizer Pressure-tow

                                                                                        $ 2.0 seconds                          M

10. Pressurizer Pressure-liigh C3 5 2.0 seconds C3

11. Fressurizer Water Level-liigli N.A.

      ^tleutron tietectors are exempt f rom response time testisig. Respouse time of tile neutron flux signal portion of tiie channel sleall lie sneastered f rosis detector otstpist or inipist of first electronic coniposient in cliannel.

T TABLE 3.3-2 (Continued)

( Rf

ACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES G

l} m FUNCTIONAL UNIT RESPONSE TIME

• lb. [ Reactor Coolant flow-Low N--

EE a. Single Loop (Above P-8) < 1.0 second El b. Two Loops (Above P-7 and below P-8) e i1.0second

13. Steam Generator Water Level-Low-Low _< 2.0 seconds

14. Undervoltage-Reactor Coolant Pumps 5 1.5 seconds

15. Underfrequency-Reactor Coolant Pumps 5 0.6 second

16. Turbine, Trip

u. a. Low Fluid Oil Pressure N.A.

3 b. Turbine Stop Valve Closure N.A.

17. Safety Injection Input from ESF N.A.

18. Reactor Trip System Interlocks N.A.

19. Reactor Trip Breakers N.A.

20. Automatic Trip and Interlock Logic N.A. t3 73 C3 C3 m

                                                                                           =cs m

N c C3 T M

TABLE 4.3-1 g Rp CTOR TRIP SYSTEH INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP O ANALOG ACTUATING MODES FOR S

• CllANNEL DEVICE WHICH CilANNEL CllANNEL OPERATIONAL O'PERATIONAL ACTUATION SURVEILLANCE

[ FittiCTIONAL UNIT CllECK CALIHRATION TEST TEST ' LOGIC TEST IS REQUIRED =

1. Hanual Reactor Trip N.A. N.A. N.A. R N.A. 1, 2, 3*, 4*, S*

2. Power Range, Neutron Flux

d. liigh Setpoint S 0(2, 4) H N.A. N.A. 1, 2 H(3, 4) .

Q(4, 6) R(4, 5)

b. Low Setpoint S R(4) H , N.A. N.A. l###, 2

, 3. Power Range, Neutron Flux, N.A. R(4) H N.A. N.A. 1, 2 g liigh Positive Rate 3 4. Power Range, Neutron Flux, it. A. R(4) H N.A. N.A. 1, 2 liigh Negative Rate S. Intermediate Range, S R(4, 5) S/U(1),H N.A. N.A. 1###, 2 Neutron Flux

6. Source Range, Neutron Flux 5 R(4, S, -le3 S/U(1),H(9) N.A. N.A. 2##, 3, 4, S

7. Overtemperature AT S R(13) H N.A. N.A. 1, 2

8. Overpower AT Oc S R ll N.A. N.A. 1, 2 C:s Cs

9. Pressurizer Pressure-Low S R H N.A. N.A. 1

10. Pressurizer Pressure-liigh S R H N.A. N.A. 1, 2 :Cr

11. Pressurizer Water Level--liigh S R H N.A. N.A. I m

12. Reactor Coolant f low--Low 5 R H N.A. N.A. I M

C3

                                                                                                                /n%

N

TABLE 4.3-1 (Continued) g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS G TRIP h ANALOG ACTUATING H00ES FOR g ' CilANNEL DEVICE WHICH

,                                     CHANNEL CilANNEL        OPERATIONAL     OPERATIONAL    ACTUATION    SURVEILLANCE c   FUNCTIONAL UNIT                    CHECK     CALIBRATION   TEST            TEST           LOGIC TEST   IS REQUIRED z

U 13. Steam Generator Water Level- S R H N.A. N.A. 1, 2 - Low-Low

14. Undervoltage - Reactor N.A. R N.A. H H.A. 1 Coolant Pumps .

15. Underfrequency - Reactor N.A. R N.A. H N.A. 1 Coolant Pumps .

16. Turbine Trip y a. Low Fluid Oil Pressure N.A. R N.A. S/U(1, 10) N.A. 1

• b. Turbine Stop Valve N.A. N.A. S/U(1, 10)

R N.A. I <;> Closure

17. Safety Injection Input from N.A. N. A. N.A. R N.A. ' 1, 2 ESF

18. Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6 N.A. R(4) H N.A. N.A. 2##

b. Low Puser Reactor m

                                                                                                                  =t3 Trips Block, P-7             N.A.        R(4)           M(8)           N.A.           N.A.

c. Power Range Neutron 1

T Flux, P-8 N.A. R(4) H(8) N.A. N.A. 1 52 0

d. Power Range Neutron

• Flux, P-9 N.A. R(4) M(8) N.A. H.A. I h C

c3 C3 "T3 W

TABLE 4.3-1

                      ~

g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS G TRIP S ANALOG ACTUATING 2

• CilANNEL DEVICE MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LCGIC TEST IS REQUIRED j 18. Reactor Trip System Interlocks (Continued)

e. Power Range

;              Neutron Flux, P-10        N.A. R(4)          M(8)            N.A.         N.A.       1, 2

f. Turbine Impulse Chamber Pressure, P-13 N.A. R H(8) N.A. N.A. 1

19. Reactor Trip Breaker N.A. N.A. N.A. N(7, 11) N.A. 1, 2, 3 * , 4 * , S* .

R 20. Automatic Trip and . [ Interlock Logic N.A. N.A. N.A N.A. M (7) 1, 2, 3*, 4*, 5* i I i "T3

                                                                                                                ;X3 C3 C3 m

Ro

=3

                                                                                                                .3 l                                                                                                                @

e C"3 C3

                                                                                                                ~13 i

M 2

TABLE 4.3-1 (Continued) TABLE NOTATIONS c..y el ny . . " ' "

   "W4h the Reactor Trip System breakers. closed and the control rod drive system k capable of rod withdrawal.

M Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

   ##8elow P-10 (Low Satpoint Power Range Neutron Flux Interlock) Setpoint.

(1)If not performed in previous 7 days. (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions af Speci-fication 4.0.4 are not applicable for entry into MODE 2 or 1. (3) Single point comparison of incore to excore AXIAL FLUX OIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applic-able for entry into MODE 2 or 1. (4) Neutron detectors may be excluded from CHANNEL CALIBRATION. (5) Detector plateau curves shall be obtained, evaluated and compared to manu-facturer's data. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable.for entry into MODE 2 or 1. (6)Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provi-sions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (7)Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. (8)With power greater than or equal to the interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST s M11 consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window. (9) Monthly surveillance in MODES 3*, 4* and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Monthly-4 swwi'!:nce cha!' 'nclue ver catter of the Scron-Dilutten 4! ara-Setpoint of-less-than-or-equ:1 to an increece of- twice-the-count-rate within a . 10 minute period. (10)Setpoint verification is not required. (11)At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips. (12-)At-4eest-ence-pee-18-months-during-shutdown c -vertfy-that- on-a-simulated Baron-OHutten-DoubHng-test-signai-the-normal-CVCS-discharge-valves will close and--the-centri fuga4-charging- pumps--suctrion-valves--from-the-RWST will open within40--seconds.-- (13) CHANNEL CALIBRATION shall include the RTO bypass loops flow rate. WOLF CREEK - UNIT 1 3/4 3-12

PROD:& R3!M COPY INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 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 Setooints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. APPLICABILITY: As shown in Table 3.3-3. ACTION:

a. With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive 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 Setp'oint consistent with the Trip Setcoint value of Table 3.3-4 and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or.

d

2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3.3 until the' channel is restored to OPERABLE status with its Setpoint aojusted consistent with the Trip Setpoint value.

Equation 2.2-1 I + R + S < TA Where: I = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel,

                                  ~

S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (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 channeli

c. WA u c s rn .,,,, u w m en. ..i ,,, ,J,, I. A ... '.t. W % bin

                                                                             ' p ,m.~

SURVEILLANCE REOUIREMENTS ^ we t ,1 - 3 4.3.2.1 Each ESFAS instrumentation channel and inter 1cck 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. 4.3.2.2 The ENGINEERED SAFETY FEATURES RESDONCE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 19 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once oer N times 18 months wnere N is the total numcer of redunoant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Taole 3.3-3.

 %0LF CREEK - UNIT 1                           3/4 3-13

TABLE 3.3-3 c

  @                                       ENGINEERED SAFETY FEAIURES ACTUATION SY5 TEM INSTRUMENTATION O                                                                              MINIMUM TUTAL NO. CilANNELS         CHANNELS          APPLICABLE l0     FbriCTIONAL t#ill                       0F CilANNELS  TO IRIP

• OPERABLE MODES ACTION rutera ,

a ' ,i*'M"6!:"is3""n:' r- ~ , =, r - > ',.< , G Component Cooling Water /54 t iesen beneratorAf, Containment

     ,8,     Cooling, and Essential Service                         -

Water a. [ef Manual Initiation 2 1 2 1,2,3,4 18

b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation -

Relays (55tQ

 !:*        c. Containment                        3          2
 **                                                                                      2         1,2,3        IS*

Pressure-liigh-1 Z d. Pressurizer 4 2 3 1,2,3 # Pressure - Low 19*

e. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line Low 1, 2, 3 IS*

any steam~ line

2. Containment Spra)

a. Hanual Initiation 2 pair 1 pair ~ts 2 pair 1, 2, 3, 4 18 operated

x:s C3 simul- "

taneously v1

                                                                                                                      $20
       . b. Automatic Actuation                 2          1                   2             1,2,3,4      14    N Logic arul Actuation Relays (wss)                                                                                        ,

m

c. Containuent Pressure- 4 2 3 1, 2, 3 l liigh-3

• 16 a C3 c

l l

s TADLE 3.3-3 (Continued) 6 ENGINEERfD SAIETY FEATURES ACTilAfl0N SYSIEM INSTRUMENTATION G n - MINIMUM

                                                      '/
   $?  5                                                           101Al fl0.      CilANNELS        CHANNELS        APPLICABLE

fi! FilllCf10NAL UNII Of CilANNELS 10 1 RIP OPERABIE H0 DES ACTION C

3. Containment Isolation z '

4 a. Phase "A" Isolation s

1) ilanual Initiation 2 1 2 1,2,3,4 18

2) Autom'atic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays (m)

3) Sately injection See item 1. above for all Sately injection initiating functions and requirements. f Phase "R" Isolation to b.

h to 1) Hanual Initiation ~ 2 pair 1 pair 2 pair 1, 2, 3, 4 18 A operated simul-

                                                                                 - taneously         -

T

o

2) Automatic Actuation 2 1 2 1,2,3,4 14 C3 Logic and Actuation Relays (Sws) M y

                      .1 ) Coutainment                                  4               2               3            1,2,3            16 Pressure-liigh-3 m

rg "5::

c. Containment Purge ,

Isolation c"3 I 1) Manual Initiation 2 1 2 1,2,3,4 17

2) Automatic Actuation 2 1 2 1, 2, 3, 4 17 l Logic and Actuation Relays (%fs) q .7) Phase "A" Isolation See item 3.a. for all Phase "A" Isolation initiating functions and requirements.

~

3) A 4.. l.. Aeluhes ,

t , p. ,, 4 <b t . . F.. . i', f.7 I 2- I , 2 , 3, 4 8 'l (6Cp CWA 9 1 _ _ . _ . _ - __ _ -

Op lAutE.3J3-3 (Continued) g ENGINEERfD sal:EIY IEAlt,itES ACTilATION SYSTEM INSTRllMENTATION G

n. MINIMUM

    ;j                                        TOTAL NO.         CilANNELS              CllANNELS            APPLICABLE
   !y  flHICIl0NAL llNIT                    Of CllANiilits      TO TRIP                OPERAutE                H0 DES          ACTION

4. Steam Line Isolation

d. Manual Initiation I '

                                                                                          <g&"

1) Individual 1/ steam line 1/ steam line 1// steam line 1, 2, 3 23'

2) System 2 2 1 1,2,3 22

h. Automatic Actuation '2 '

2 Logic and Actuation 1 1,2,3 21 Relays (55FL) t Containment Pressure- 3 2 -2 1,2,3 15*

   ,,              liigh-2 s

[ d. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15* 4 Pressure-tow any steam on line

e. Steam Line Pressure- 3/ steam line 2/ steam line 2/ steam line 3# 15*

                 - Negative Rate-liigh                         any steam line

5. lurbine Trip &

feedwater Isolation "U

c

a. Automatic Actuation 2 CD 1 2 1, 2 Jr 27 co togic and Actuation T Relay,(ssrs)

QC

b. Steam Generator -

4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 19" Water Level- in any oper- in eacli oper-liigh-liigh ating sim. gen. al,ing sim. (jell. Q

c. Sately Jiul.ca . sm

                                             &cs,  s n,    t o.J a         f. ,     .o          s.(,6 3 g. i.. ,      ;. w -. )
                                                                                                                                     ,, 8o
                                                                    ..g.<....ts.

E

u TAlllE 3.3-3 (Continued) c ENGillEERfD SAFETY FFAluRES ACTUATION SYSTEM INSTRUMENTATION

                 '                                                                                            minir. #4 HINIHilH                                               C8"""

O IDIAL t10. CllANilELS CilANNELS hPLICA8L5j i l0

                 ,.           IllNCTI0rlAL llHII                            of CllANilEIS 10 TRIP           OiUI4ilfE        MODES
                                                                                                                                     .) ACTION e           6.        Auxiliary feedwater                                              .

E 36m r) I/P "I' 'If"~P

a. Manual Initiation ' 24 Q

2 -.P- ..-2 1,2,3  ;!2' 13 . Automatic Actuation Logic 2 1 2 1,2,3 21 and Actuation Relays mo) c- k$I'1 MU#'5,TJA "^ z i 1 1, t, 5 28

                                 ' 2 re.       Sim. Gen.Wa$er, Level 3 Low-low

1) Start Motor- -

Driven Pumps 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 19*

                ,,,                                                                        in any opera- in each g                                                                          Ling stm. gen. operat.ing sim. sen.

C 2) Start Turbine-Driven Pump 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 19" in any in each 2 operating operating stm. gen. sim. gen.

c. ( Safety Injection - Start . -

Hotor-Driven Pumps See item 1. above for all Sofety injection initiating functions and requirements. ca fy. Loss-of-Of tsite Power - Start lurbine-Driven Pump 2 E3 g 1 2 1,2,3 22 QO

x2 T = "1 C'2

                                                                                                                                                 ~
                                                                                                                                             ,   O CD
                                                                                                                                                 ~C3 N

1AlllE 3.3-3 (Continued) ' s ENGINEERED sal'ElY FFAlllRES ACIUATION SYSTEM INSTRtiMENTATION f

    '"                                                                               stie):Mue4 MlHitillM                                             08 Add W O                          TOTAL NO.             CllANNEl5       CilANNELS       MPLICABLP              $    i l,"   filllCTIONAL llNil                        Of Cl!AllNELS    10 TRIP          0
   ^
   -                                                                                         i            H0 DES  ACTION             .
     . 5.      Auxiliary Feedwater (Continued)

E ~ri f. Irip of All Main G feedwater Ptunps -

   ~                    Start flotor-Driven Pu;nps               4-(2/ptunp )*

• 2-(1/ptunp 3-(2/in.;;;;;)- 1, 2 19 in see AW 5.p ration) i .g. me aei .)

j, y. Auxiliary feed-water Pump Suction Pressure-Low (Transfer to ESW) 3 2 2 1, 2, 3 IS* w 7. Automatic Switchover to h Containment Sump T a. Automatic Actuation 2 1 2 1, 2, 3, 4 14 G Logic asul Actuation Relays 6 4)

b. RWST Level - Low-Low 4 2 3 1, 2, 3, 4 16 Coincident With Safety injection See item 1. above for Safety Injection initiating functions and requirements.

8. Loss of Power u x8

a. 4 kV Bus Undervoltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 19^ C3

                       -Loss of Voltage

h. 4 LV Dus Undervoltage 4/uus 2/Uus 3/ Bus 1, 2, 3, 4 19^ Go

                       -Grid Degraded Voltage rg h

CS O

                                                 -                                                                      I    O M

M

        $k.WYJ                                 . . &.:') -                                                                               L. 0                    .my                       s.g@@D?$h$$hb&4.Y-9WpS 5
                                                                                                                                                                                              ".U$y & &}M L &Ek!h 4.':MM
       %.y:9'3'fMujfdli@.b5'(%                 :6.~
                                                                                    .- C . ,.! .,. . ~~W c,':'. - U
                                                                                                                             '. ?." ' s L:$
                                                                                                                              'i.s( .0%,R$9p%$g.g N.'i $ $

M: .: ..,'.

                                                                                                                                                                             $ jjp37 ?                                                                                                                                                            $Q
                                     '                            ')'.,

TABLE 3.3-3(Continued)4 .. .,

                                                                                                                                                                                                                                                              }/ . p eig{P f                ,                                             .;                     ;

g r.n

                       ,r              .
                                                                'l[        3                                                                                                                                                                                                     g, ..                                                 .7                  ,1
                                                                                    ' ' ENGINEERCO SAFEfY                                                                 FEATURES                                ' ". ..V a. .s. e.ACTUATION                                                                                                                       SY e

S-

                                                                  .v.

e. ,

MINIMUM ' W Y. . , s x.

                                                                                                                                                                                                                                                                                             .      .%.a.g,MM; K d, f
                                                                                                                                                                                                                                                                                                                                     .L a        .
                                                                                                                                                                                                                                                                                                                                                     -+ fM TOTAL NO.               CHANNELS         ,    CHANNELS lN pc              . FUNCTIONAL UNIT. -                                               0F CHANNELS              TO TRIP               OPERA 8LE , I.' .'.?                 M APPi.ICABLE H0 DES'E-{ ACTION.k0.3 # k'T. N                                                        ~.->
                                                                                                                                                                                                                                                                                                                                                                   ,G e

9. Control Room.' Isolation

                                                                                                                                                                 ~,T5      34 p.;t*:im;rg.W                              8. ' AU...ggj;idN5.,               dl 1-g                                  -
                                                                                                                                                        ~:~
                                                                                                                                                                                                                                                                              - g".g.-gg $k/ggSM T

q a. Manual Initia ion 2 , 1 2 .  : b.

                                                                                                                                                                                           -    m,. All ", C,l y @ ) 0 $..f.i,(i p:s. Q. w : $,    .                        ..                                   .

Automatic Actuation Logic '

                                                                                                                                                                                 <                                                                                                             7.A '.' % w. W.M-                                       o.,$,d..
                                            ' . ___and
                                         ' c-                eh Actuation nRelays       Lo34s, 6SP6  2l- - --- 1               ~ ~ ~ ~ ~

2- ' -A4+3.1.3N . ':.14' -24%kM@fMD A.J.& q,. g cm n..p > g a' d p/~ Phase S..<hiaH.hioSee item '3:a.above' for all-Phase-"A"-Isolation-initiating A"'Isolat i 2.

                                                                                                                                                                                                                                                                 *- " th
                                                                                                                                                                                                                                                                                          \ .'. :JM         . V4 W%ca                                        0(ig
                                                                                                 . functions and requirements.

664

                                                                                                                                                                                                                                                                                                   - ,@'M'                           - -q:N.:.
                                                                                                                                                           ~
                                                                                                                                                                                                                                                                                                                 ?. ,*:7,'a.                          C2
                                                                                                                                                                                                                                                                                                                                                            . 6 v .p 4 10' Engineered Safety features                                                                                                             ^
                                                                                                                                                                        -                                                                                                            . ,,              t of,.......                .

a' y. . - i Actuation System Interlocks

                                                                                                                                                                                ~,         T.'           ~
                                                                                                                                                                                                                           .                 J .U..                   :                       y ..'               1.'9.T 5,, ism / %

.f1 R; .;r.FEM. Ep.n'

                                                                                                                                                                                                                          '" " .             .,.p                                           . yi,'i>f.,
                                                                                                                                                                                                                                                                                                   .: ..                                  m
            , {\ ,                             a.       Pressurizer Pressure,                          ;3
                                                                                                                                                                                                  .                                                    v ,.                                                                       ,..~...y,                    .,,

2; #.. 2 P-11 1, 2, 33 M . (20 p:

                                                                                                                                                                          , z.w,.w - a, q.,. %,. --r: . :.: e . ; . . .A.;S.'j.'g;Q w                                                                                                                                                            . r
                                                                                                                                                                                                                                                                                                                           ,.> y
                                                                                                                                                                                                                                                                                        .:.,, .w
                                                                                          ~

e. U b. Reactor Trip, P-4

                                                                                                                                                                                 '. c..
                                                                                                                                                                                                                                          ... .v . . -3 ,                                                ,.c.          >..n:a.            y
                                                                                                                                                                                                                                                                                                                                                 @m , -,
                                                                                                                                                                                                                                                                                                                                                 .: ct .:: g 4-2/ Train                2/ Train              2/ Train                     ' '
                                                                                                                                                                                           '21',-)2[r, f.3. .U.MM.      s
                                                                                                                                                                                                                                                . ~ . . 1(,22;;?.,/.'.       .            : .                      - :

d .1f.,,'i.-;; ' m

                                                                                                                                                                                               ,; .1-c..                                  s       g , y< .. . . 4. y . s.. a t, ,=;.,n,, .,..,

7.g,; ,. p . : l

                                                                                                                                                                                                                                                                                                                                                  ,.,s.,,. 1.e ; q
                                                                                                                                                                                                               . . .y . p s -
                                                                                                                                                                                                                                                                                       , f,y
                                                                                                                                                                                                                                                      **.-r..
                                                                                                                                                                                              ,.-                                                                                                                    3
                                                                                                                                                                                                ;-]8. '2-j b k                              '[. %ns:.*

a ss.* a ,,\ ,. ;g , , 3,, .., . 7, ,j 3D;.

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e l l i TABLE 3.3-3 (Continued) kh h bf TABLE NOTATIONS

                                                                                                                                                                                                                                                                                                                 /
                                                           # Trip function may be blocked in this MODE below the P-11 (Pressurizer                                                                                                                                                                                                                               ,
                     ',                                       Pressure Interlock) Setpoint.
                                                                                                                                                                                                              .a q - - ,
                ,4 .,

i . .. , . .

                                                                                                                                                                                                                                                            , .,                                                                         7;
                                                        ?? Trip f=:ti= :y 5: byp::::d in thi: "00E :b:;; th: P-11 (Prassurizer
                              ' D'                           Pressur: Int:rl::k) S:tp: fat. '
                    '                                                                                                                                                                                                                                              ~
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                                                                                                                                                                            -- ' t ,. .: W ,a                                            _ _ ~ , ; .s;@                                       ' . . .v       . . f?.n
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          ?(. > jd                                         Trip function automatically blocked abovefP-11 and'may be' blocked below                                                                                                                                                                                                               -

l , Jib.~~'* P-11 when Safety Injection,on low steam ifne piiss~urOis=not blocked.'~-

                                                                                                                                                                                                                   ~
                                                                                                                                                                                                                                                                                                                                                       ~
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                                                                                                                                                                                                                        ..,:                     .,y -...z,      ..

s.;.- - Q,.,,  :

                                   ,                        "The provisions of Specification 3.0.4 are not applicable.
                      .. . . . .                                                                                                      ..-                                    , ..                           ,+.......          .                       .         .o

.1;

                                                         ,**0ne in Separation Group 1 and one in Separation Group 4. 3;
                           .-                                                                                                       r. ' <                                    ,.
                                                                                                                                                                                                     . . . fpsm                  .. - 'lw'                                                                                     ,
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           .                              ~                          .

ACTION STATEMENTS ' ' %: .,,J ';, 3 2

                                                                                                                                                                                                         .. .. ? OC ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY                                                                                                                                                                                                      -

within 6 hours arid in COLD SHUTDOWN within. the following 30 hours; however, one channel may be' bypassed for up to 2 hours for surveillance testing per Specification;4.3.2.1, provided ' the other channel is OPERA 8LE. ,.

                                                                                                                                                                                                                                                                              -1               ,          .                           .c A                     .
                                                                                                                                                                                                                                                                                 . ..                   .T ACTION 15 - With the number of OPERABLE channels 'one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the                                                                                                                                                                                                         ,

inoperable channel is placed in the tripped condition within 1 hour. 9 X' '

                                                                                                                                                                                                                                           ~
                                                                                                                                                                                                                          ~                                                                                                      '
                                                                                                                                                                                                            .o.;                                                                                                                                .,L ACTION 16 - With the number of OPERA 8LE channels .one less than'the Total                                                                                    ~

Number of Channels, operation 'may p'roceed provided the ' inoperable. ~ f .

                                                                                   ' channel is placed in the bypassed condition and the Minimum.                                                                                                                                                                                           3;3 .;
                                                                                  . Channels OPERABLE requirement,is met. .Ond additional channel. ,g, i. -"

may Specification be bypassed 4.3.2.1. for up to 2. hours ' .for. surveillance testing per.T.j;'(. M ,

                                                                    . '           =
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9. . . ;

ACTION 17-WithlessthantheMinimumChannelsOPERABLEreqdirement, . operation may continue provided the" containment purge supply and exhaust valves are maintained closed. * ~ M' 'V Wr8.C

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                                                                                     . %g:. 4. ACTION STATEMENTS (Cohtinued)' f.nf/T:l.'#4                                                                                                                                       M.$.p,            -
                                                                                                                                                                                                                                                                          , . O'.        ',.,7 y h$

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                                                                           ' Nti)d%@hW.:.4:.WWm&M::the WS                                                                                                     .OPERA       d:M(;n18LE channels.5,#;.                      ,

• 7 ' number one of less than the'W.

[ C wACTIOEl[',!;*? Minimum- - M

                                             3:f',W                                Ch'annelf0PERABLE ' req'uirement; . restore the inoperable channel '

I J Nik pk8 - t

                                                                              #'o'l0PERABLE"                                status'within~ 48" hours or' be in at least HOT STANDBY
                                                                                                                                                                        ~
                                                                                                                                                                                                                                                                                               ^

dthin'the hext 6'hours'and'in COLD SHUTDOWN within the following 4@/,# *,30'h7dFsagg;2wmas vu u X .. > - l <s.mACTION

                                                            .4 u sg%<xt;d.                 ~ 19..,q -t,y>ee,.. With             th
                                                                                                                                   . n. 4.yora     &W - N -s-             %;J;44wM
                                                                                                                                                                            +. A r e                                     I?,.NY'bs
                                                                                                                                                                                                                                                                  #/a.^. *.i m m 1.<.                        .c                F* .. .. .
  .w                              -                                                                                                                                                                              .

e n' umber of OPERABLE' channels one less,than the Total -

                          ' d,.9 %$#/4 #T*7 Numb'eir of Chann'els', STARTUP and/or POWER OPERATION m wry
                                           .r
                                           =di            b! bW .n . mvided g.oh;% {.e Q %ph.
                                                                                                            ,,              lhe' ..         following conditions are satisfied:                                                           -              -
  .-                            l v.4                 W Re.orM he;s,*. ;,.. 2.s ;.= P. , s ., . .:3. 5 t'%: . . ;.: - a.w; . , ^: ' .                             ..

fg,hdh .$.N Ti J'97 Ql-%Q M s& Eg$$.8M4The dW ' NWWE!T inoperable and. channelW , ' E

                                                                                                                                                                              ' Y isW'plac:ed in -the
                                                                                                                                                                                                          .W'hour, U ' : ?
                                                                                                                                                                                                                                                 ' ^
                                                                                                                                                                                                                                                        '- tripped           - ' conditi n%      iPMj k}                     hk&WTD                 Q .N Ythe&inoperable                                        bp Y.6                                   .within                              1.

g' channel may De bypassed for up to 2 hours W.;N)W' hf d $% iM hM@vM$g..Nd.i?.b.'@,The

                                                           @t$
                                                                                   /;'3.4.3'.2.1.                                                    .

Minimum C (% .A bk0h2b With less than the Minimum Number of Channels OPERABLE, within Ml..;.._ $$jp.d A 1 hour determine by observation of the associated permissive.

          ;, -            V

' y.] # Q', I I. 1 ;i.h & % ' M0;;W annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification

                      $ ** .' , Q,?.

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                                                                                 , 3. 0. 3.;.
                                                                                                                          '                        ~

r. .

                                                                                                                                                                ;                                  . ~ . .
           ...w cre;y. ACTION 21 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY.-
          ' y.b%..j .,j

9f;,.?.$.Q g g M .:; .$ ([.t.f.{2.'i w ithin 6. hours and in at least HOT SHUTDOWN within the following ~

g fy ' . T: .pr;:3.ql*- 6 hours; however, one channel may be bypassed for up to 2 hours. ln. Gd W. A.My , for surveillance testing per Specification 4.3.2.1 provided the

      .c..s.:.%.s.2' 7q :p.             N,i.C        .. other                 channel is OPERABLE.                                                                                                          '

t. .,p .. ... .,r.. f.F', '. ,

          ..*             t w^;

t

                                                                                       , / .H .+;.-.m>-        .                           . .

pUAp e.. h,p,. . . ' 'a. .

                                                                                                                                           .>-                                                                 ~

M.5kh$. . 7,M/ ACTION m's J 22T-Number With the number of OPERA 8LE channels one less than the Totai^ of Channels, restore the inoperable channel to OPERABLE jydiMMEd.!H.dM .'H'L /j status within 48 hours or be in at least HOT STANDBY within'

 'O                                   NPQ .(2                                      6 hours and in at least HOT SHUTDOWN within the fcilowing'
                                                                                                                                                                                                                                                                             ~~

3 g.$Q  % U.(i Q $;;.y$  %'., . , , .

                                                                                                                                                           ,s                              c,.

7

 .$Hih.0%:;:-:W W W %,,. 6.:s:m.                                                        hours.c             9: 7                                     s- '"                                % ..

R :. 3/ia:.o.yM hdM.Ihjf/.@% YACTION 23.- With the number restore of-0PERABLE channels one less than the T

                                                                                                       ~
                                                          %,; T yn Ndmber                              of Channels,'                                            the inoperate channel                                       to OPERABLE gp2 dv.- -                                       !        u. f . .              status within-48 hours or declare the associated valve inoperable W/ C.4" Z '_^.-and y%s.y($(g@#;

take-the a'ction-required by Speciffcation 3.7.1.5. ,

2 ,, 9.e.ss. 'FT..c.y,. .c .. . . . .. ~.~~. 3 . f[.' ACTiolJ n- ISu. cAcked . Mmu M. '

                                                                                                                                                                                                                      -                              2          -

if&$;?,~.. *.. : .Qince%5 ' hk$?hY.I ' %g:rW:16.w- Q.' . a%G.M 'cWW 'l W" ^& M }

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                               ~

ACTION 24 - With the number of 0PERABLE channels one less than the Minimum Channels OPERABLE requirement, declare the affected auxiliary feedwater pump inoperable and take the ACTION required by Specification 3.7.1.2. ACTION

• 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, declare the affected diesel generator and off-site power source inoperable and take the ACTION required by Specification 3.8.1.1.

ACTION 26 -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 initiate and maintain opera-tion of the Control Room Emergency Ventilation System. ACTION 27 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. b S

d; 4 1 3 c 333" n CC-

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                                                                    .        ,8 u

6 . 3' TABLE 3.3-4 (Continuedl s-P ENGINEEREO SAFETY TEA 1URES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ri Q TOTAL SEllSOR TRIP ALLOWABLE lll fUllCI10tlAt UNII . AL,t0WAtlCE (TA) Z SETPOINT x ERROR (S1 VALUE

2. Containment Spray -

E . Q a. Ibnual Initiation N.A. N.A. fl. A. N.A. id N.A. e

b. Automatic Actuation Logic and Actuation Relays (ss >s) N.A. N.A. N.A. N.A. N.A.

c. Containment Pressure- 43 i .it t liigh-3 u.3 .,

3d 0.71 lA . f 27.0 psig 5 2,84 psig

3. Containment Isolation w

   )       a. Phase "A"    Iselation

1) Manual I.nitiation H.A. N.A. H.A. H.A. H.A.

2) Automatic Actuation Logic and Actuation Relays (55f5) N.A. N.A. N.A. N.A. N.A.

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

b. Phase "11" Isolation l t3 M

c.m

1) Manual Initiation N.A. e .2 -4 N.A. N.A. N.A. N.A. m

2) Automatic Actuation C4 Logic and Actuation m t'

Relays (Urs) g N.A. N.A. N.A. N.A. N.A. M,3 r-c1

3) Containment 4.3 f.'js ,g 3 G Pressure-liigh-3 _Iwo 0.71 5 27.0 psig 5 2 M psig a JAf C3

                                                                                                               *%J

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            )

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A A A .l T T S ( aa E .V R 3 E l t C nl

                                                                                                      .e l                N                                                        eb A                A                                                        t a E

F W O ) I w l o e lyd L d Y L e A A. -4.el el A. A. oo T E >A n u N, N9SA I. N N 4 % MIJ

3. 3i

                                                                                                                                                                                    ,k F

i A t nn t n . S n oo h,io o n i i D C o tt - - E R ( i aa "Y a nn e r e e0 h cc)ro t E n a tuu t ii n oo u r ri E o i s n o ii s u ui l l n tt s s s i l i t o t 5 o i aa e s s-Acd5A 'vN."I i G a i n i L u u' r e e l l t . t a t t P r re E l o t a I a i cc AAg P Pt s nl l itana (5s h"A l o L e t n e a _ hA l I eo ms a u mcyae oia s I s I I n cds it an (s e m n eR n w t n uoePhPa i i nI t gl N n a n L v dt t i e a s ihn. iaa mte 1 l e ae M i2 i m t g A L R^ n a mcy a- m j i ta oia th a L n nr A i ou ) ) L n a t Dl uoe ng eW eg 0 t a CP 1 *2 1 )3, m N AtR oi Ci t O SL t e SN 1 n a l e 1 Cl C o c

                                                                                      -/ (4'                t S       a

b. c d

e li i . F 3 4 gG Oty [=U -  ! yu y~

o , e. lABl.E 3.3-4 (Continued) 6 . ENGINEERED SAFE 1Y I[AltlRIS ACIUATION SYST[M IllSIRUMENTATION TRIP SETPOINTS n N' OTAl,)2 filllCTIONAl Illl I ALLOWANCE (TA) Z

                                                                               ->h               t u        >       s m                                                                 -               ERROR (S)       SElPOINT          VALUE

5. Turliine Trip and C

feedwater Isolation -

   =                                                                                                                                                ,1 y              a. Automatic Actuation                                                                                                           ..

r

• 1.ogic and Actuation Relays (5595) N.A. N.A N.A. N.A. N.A.

13 Steam Generatof Water 2.51 19, ; Level-liigh-liigh 5.0 2.18 k'J ' 5 78% of < ZSc6% of r4isrrow ranti _ narrow ran0e

                                                                               !     instrumen         nstrumen             >

span span

C. _ S=[cly L *;esh m g

6. Auxiliary feedwater gy , , t , p , y , (. ,,, g ,y yj;g,

                                                                                                            '/

s. j w a. Mannai Ini t.ia tion H.A. 11. A. N.A. 11. A. N.A.

ud

b. Aiitomatic Acttiationi logic and Actuatiori

                ,    He lays (55P5 )             N.A.               N.A.           H.A.              N.A.              N.A.

c. bru t g';b- g p.s. v.a.

M p - Steam Generdtor dal,te;r ,i. s . v.a. n.i. , Level-Low-Low

1) Start Hotor- 'p,5 :21. r f 1.41 D6 25 3 Driven Pumps .30-tr 42d8 ,LAf 1 33e2% of 2 3&r4% of y rrow ratige- -

narrow range o 4 inst 'nstrument  ? O (una>rumeni.ian C.J m

2) Start Turbine- 27 3 as.sf 24s 'I56 22.3 Driven Pumps 30ett ZLM le'5 iarrow rangi instrument 1 3k2% o f t 3Dfti% of narrow range ]

instrumt. y

                                                                                      ,a n          < >an                                  ]r+:

C CD

                                                                                                                                           ~O M

e f

                                                                                                                      'r, TABLE 3.3-4 (Continued)
      ,                            ENGINEERED SAFETY FEAltlRES ACIUATION SYSTEM IN5fRUMENTATION TRIP SETPOINTS O

m g TOTAL SENSOR TRIP ALLOWADLE

     ,m . FUNCTIONAL UtilT                         ALLOWANCE (TA)   Z                             SETPOINT       VALUE m                                                                                 ERROR (SJ

6. Auxiliary Feedwater (Continued)

E Q d. Safety Injection - I *}

     "             Start Motor-
                 ~ Oriven Pumps                    See Item 1. above for. all Safety Injection Trip setpoints and Allowable Values.

e. Loss-of-Of fsite' Power-Start Turbine-Driven Pump N.A. N.A. N.A. N.A. N.A. *

f. Trip of All Main Feed-water Pumps - Start Motor-0 riven Pumps N.A. N.A. N.A. N.A. N.A.

w g. Auxiliary Feedwater A Pump Suction Pressure- 2 2/.56 psia Low (Transfor to ESW) N.A. N.A. N.A. 2 20 63(Sea. N (tMEE)

7. Automatic Switchover to Containment Sump

a. Automatic Actuation Logic a'nd Actuation

Relays ( 55Ps)  ?

N.A. N.A. N.A. N.A. N.A. b. i RWST. Level-Low-Low Coincident with (' .') .) d<

                                                                                                    ..uwa   t5e 5    %                y4
 -                                                                                                                 n'aurusand Sr"   c3 Safety Injection                See Item 1. above for Safety injection Trip Setpoints and Allowable Values.      ~M Go

a 8

                                                                                                                                    =s.$

C3 T

s.

                                                                                                              't,                  .

I r . TAlllE 3.3-4 (Continued) g ENGINEERED SAFETY FLAlllHES ACIUAll0N SYSILM INSIRUMENTATION TRIP SETPOINTS 6 h gi filNCTIONAL ! TOTAL 4 h(l{ij e ALLOWABL  % ALLOWANCE (lA) Z Lilit0R (S) S NT VALUE [ 8. Loss of Power y; ,

  -4           a. 4 kV Undervoltage                                                283Y (120V Bus)  2 ~47V                   820V B"5)       : #.

s -Loss of Voltage N.A. w/js Alay (IMf44,-0.[s w/i tO.0 delay N.A. N.A. BMS3C)

b. 4 kV Undervoltage l V f g

                   -Grid Degraded Voltage I $7Q,)

w/ils de lay 2 y/Il9 w r I. os ( delay

11. A. H.A. H.A. fi" !?J -tt A M ft)

9. Control Room Isolation

a. Manual Initiation N.A. N.A. N. A. N.A. N.A w

D b. Automatic Actuation to Logic and Actuation A Re1ays ('2'40 N.A. N.A.

 ~                                                                      H.A.            N.A.~         N.A

c. 9-p !ti;c y;y,,,, n.a ru. v. s.

u. a . a. x .

        ; ,( f.

fhase"A"Isol$ tion See Item 3.a. above for all Phase "A" Isolation Trip Setpoints and Allowable Values. 11 Jtf. Engineered Safety features Actuation System Interlocks

a. Pressurizer Pressure, P-11 1979 Q H.A. H.A. N.A. $ 1970 psig < L981 psig Ea] 4 L.3

b. Reacter Trip, P-4 N.A. N.A. N.A. N.A. N.A. ~*'

(2o

                                                                                                                                          ;ts

p. $d.A StJe tc.l 9.A. ti . A . p A. g. A p. A . p]Q

                     %  s ,w Q

c3

                                                                                                                                         ~~O M

TABLE 3.3-4 (Continued) R00F & HEW COPY ! TABLE NOTATIONS

• Time constants utilized in the lead-lag controller for Steam Pressure-Low are it > 50 seconds and T2 1 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.

l t l l i l WOLF CREEK - UNIT 1 3/4 3'23

l i l TABLE 3.3-5 FR007 & REV!EW C ENGINEERED SAFETY FEATURES RESPONSE TIMES 1 l INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual Initiation

a. Safety Injection (ECCS) N.A.

b. Containment Spray N.A.

c. Phase "A" Isolation N.A.

d. Phase "B" Isolation N.A.

e. Containment Purge Isolation N.A.

f. Steam Line Isolation N.A.

g. Feedwater Isolation N.A.

h. Auxiliary Feedwater N.A.

i. Essential Service Water N.A.

j Containment Cooling N.A.

 ;           k. Control Room Isolation                             N.A.

1. Reactor Trip _

N.A.

m. M 'hiesel Generator N.A.

n. Component Cooling Water N.A.

c. Lb.a. Trie y,a,

2. Containment Pressure-Hich-1

a. Safety Injection (ECCS) 1 29(1)/12( )

1) Reactor Trip <2

2) Feedwater Isolation 7 5

3) Phase "A" Isolation i 1.5(#)

4) Oc,r ta i .....cr,t P e r;: !::1:ti~1 W

44) Auxiliary Feeawater < 60

                 .C &)   Essential Service Water                         60(1)

G 7) Containment Cooling ~ 60(1)

            ,    72)     Component Cooling Water'                    N.A. 6
                               ~

3F) M t Diesel Generator i 14(#)

9) 'Tur6in Traf n, A .

WOLF CREEK - UNIT 1 3/4 3-29 _ -,M--

TABLE 3.3-5 (Continued) t

                                                                                                         !. hh

• ENGINEERED SAFETY FEATURES RESPCNSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECCN05 _

3. Pressurizer Pressure-Low -

4

a. Safety Injection (ECCS) I 129(1)/12$)~

 ,                  1)   Reactor Trip                                   <2

2) Feedwater Isolation h7Y ,

3) Phase "A" Isolation - 12 e r ~ 1-----

m.r;e n g ep e,, y 4 S) Auxiliary Feedwater ' 1 SO ffr) Essential Service Water 1 60(1) 4-7) Containment Cooling 1 60(1) , 7 F) Component Cooling Water N.A. G

                 ;F)
                    ')

M77ieselGenerator Tu b.ne 0 Top

                                                                       <18)

[q.A.

4. Steam Line Pressure-Low

a. Safety Injection (ECCS) M

                                                               --     1 ,20' )/12(2)

1) Reactor Trip (frem 5f) 12

2) Feedwater Isolation 17@

3) Phase "A" Isolation 12

2) C c ; u m .. e ni.' N . g a D c h S c- ++

4d Auxiliary Feedwater 1 50 (F) Essential Service Water 1 60(1)

67) Containment Cooling Jans 1 60(1)

78) Comconent Cooling Water N.A. t.

r F) M70iesel Generators < 14( )

                  .r. fu4.= %p                                      7p.

b. Steam Line Isolation 17 l

WOLF CREEK - UNIT 1 3/4 3-30 e

M005 & O!EW COPY TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

5. Containment Pressure--Hich-3 -

a. Containment Spray 5 M 1)/M(2) g

b. Phase "B" Isolation 5 31.5- --

6. Containment Pressure--Hich-2

 ~

Steam Line Isolation 57,

7. Steam Line Pressure-Necative Rate-Hich 4

                                                                    -7 Steam Line Isolation ~                         M a    8. Steam Gererator Water Level--Hich-Hicn

a. Turbine Trip ~

                                                                   < 2.5 -

b. Feedwatei Isolation _7N

9. Steam Generator Water Level - Low-Low

a. Start Motor-Driven Auxiliary Feedwater Pumps 5 60

b. Start Turbine-Oriven Auxiliary Feedwater Pumps 5 60

10. Loss-of-Offsite Power Start Turbine-Oriven Auxiliary Feedwater N.A.

Pumps

11. Trio of All Main Feedwater Pumos Start Motor-Driven N.A Auxiliary Feedwater Pumps WOLF CREEK - UNIT 1 3/4 3-31 i

                                                                                                         >             3 TABLE 3.3-5 (Continued)                                      [ R00F & RDilEW 8 ENGINEERE'D SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION                                                         RESPONSE TIME IN SECON05

12. Auxiliary Feedwater Pumo Suction Pressure-Low Transfer to Essential Service Water N.A.

13. RWST Level-Low-Low Coincident with Safety Injection Automatic Switchover to Containment -

                                                                                                < 60 Sump                                              .

14. Loss of Power I4

a. 4 kV Bus Undervoltage- .

                                                                                               <H Loss of Voltage

c. 4 kV Bus Undervoltage-a -< 144-Grid Degraded Voltage

, 15. Phase "A" Isolation 1

4. Control Room Isolation ~

N.A.

c. O

• h..c,-e- r tk < p % .l N ea ~ ~ g CO WOLF CREEK - UNIT 1 3/4 3-32 l-t

TABLE 3.3-5 (Continued) h b / ll COM TABLE NOTATIONS 1 (1) Diesel generator starting and sequence loading delays included. (2) Diesel generator starting ud ::quencu k; E ,; delay not included. Offsite power available. . . - .

           -t   2. v . ~ . i a a^.
          ~ _ _ . , _ .       -

(s (M Diesel generator start.ing and sequence loading delay included. RHR pumps M included. [4) fM Diesel generator starting and sequence loading delays not included. Offsite power available.* RHR pumps not included. (S) Sequence-dc!cy; ,J -in ' < h *. ( 6)f7/ Does 'not include valve closure time. (O (h) Includes time for diesel to reach full speed. J

                                                          ~

er WOLF CREEK - UNIT 1 3/4 3-33

TABLE 4.3-2 r ( ENGINEERED SAFEIY FEATURES AClHAIION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3 h m TRIP ANALOG ACIUATING MODES 7 CilANNEL DEVICE MASTER SLAVE FOR WHICH-c CllANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION- RELAY RELAY SU'lVEILLANCE

l. = FUNCil0NAL tlNIT CllECK CAtIURATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED w

  -      1. Saf ety Injection (Reactor Trip,hese YLI. I...          '

feedwater Isolation,fCoinpodeh& Lt e 7 p Coog Water,ASErrTIII'esci ^ A..;t.., r;, b. ;,, .sg r., _ b..u. L ,pco...y 3 c(p- Generatiir#, Containment Cooling, . and Essential Service Water g,

a. Manual Initiation N.A. N.A. N'. A.

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

b. Automatic Actuation N.A. N.A. N.A. N.A.

Logic and Actuation M(1) H(1) , Qu') 1, 2, 3, 4 i s y Relays (5srs') { j [ c. Containment Pressure- S R H N.A. N.A. N.A. N.A. 1, 2, 3 4 liigh-1 l

d. Pressurizer Pressure- S R H N.A. H.A. N.A. H.A. 1, 2, 3 Low h

c. Steam Line Pressure- S R H N.A. N.A. N.A. N.A. 1, 2, 3 Low

2. Containment Spray

a. Manual Initlation H.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) ;Q ( s) 1, 2, 3, 4 1ogic and Actuation Relays (55 4)

c. Containment Pressure- S R H N.A. N.A. N.A. N.A. 1,2,3 liigh-3 I

I l$b L

IABIE 4.3-2 (Continued) 6 1 RIP Q; n lX

      ~

ANALOG CllAfillEL ACIUATING DEVICE hJ MASTER SDLVE d FOR WHICH I CilAllNEL CititlNEL OPERATIONAL OPERA 110NAL ACTUATION Q flillCIInflAL UtlII CilLCK CA!!DRATION TES1 IES1 RELAY RELAY SURVEILLANCE

   '                                                                                                 10GIC TEST TEST        TESI       IS REQUIRED c-  3.' Containment Isolation 5
  -4

4. Phase "A" Isolation r 1) Hanual Initiation N.A. fl. A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

2) Automatic Actuation N.A N.A. H.A. N.A. M(1)

Logic and Actuation M(1) ' Q(3) 1, 2, 3, 4 Relays (%Fs) -

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

b. . Phase "B" Isolation

1) Hanual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 1:*

• 2) Automatic Actuation N.A. fl. A. fl. A. fl. A. M(1)

Logic and Actuation M(1) Q 1, 2, 3, 4

 ','                 Relays (%fs)

3) Containment S R H N.A.

Pressure-Iligh-3 N.A. N.A. N.A. 1,2,3 f

c. Containment Purge Isolation

1) Hanual Initiation N.A. N.A. fl. A. H N.A. N.A. N.A. 1,2,3,4

2) Automatic Actuation N.A. fl. A. N.A. fl. A. H(1) togic and Actuation H(1) QG) 1, 2, 3, 4 J) kit. 'h t.' "'"' '^' NI #~#' #'#' #8 * )

           .4 2)    fhase "A"   Isola"la'd't.e-/~4 tion           See llem 3.d. above for all Phase        "A" Isolation Surveillance Requirements.-

6# > D

                                                  .                                                            \

 .                                                                                  a I Alll E 4.3-2 (Continued) t-ENGINEEREI) SAFETY fEAluRES ACTUA110ft SYSTEM INSTRUMENTATION y                                                           SilRVEILLANCE REQlilREllENIS n

o TRIP l2 ANA10G ACillATING

    ^

H0 DES

      '                                                                  CilAtitlEL     DEVICE                    MASTER SLAVE-     FOR WilICH CilANNEL CilANilEL         OPERATIONAL OPERATIONAL ACTUATION        RELAY      RELAY E      fullCII0llAl littlT                                                                                                      SURVEILLANCE CllECK     CAllBRATION TEST               TEST         LOGIC TEST TEST         TEST   IS REQUIRED 7

4. Steam Line Isolation

d. hantlal IniLial. ion N.A. N. A. N.A. R N.A. H.A. N.A. 1, 2, 3 13 Automatic Actuation  !!. A. N.A. N.A. N.A. M(1)

Iogic and Actuation M(1) Q 1, 2, 3 Relays (656)

c. Containment Pressure- S R H H.A. fl. A. fl. A. N.A. 1, 2, 3 liigh-2

   ,,,         d. Steam Iine Pressure-     S           R               H                   it. A. N.A.            N.A. N.A. 1,2,3 h                Low
   ';"         e. Steam Line Pressure-   (S            R              H                    H A. N.A.            N.A.

g flegative Rate-liigh S R H N.A. H.A.

                                                                                                                            ?!. A. 1, 2 N.a. H.A. 3, g S. Turbine Irip and feedwater Isolation

a. Automatic Actuation H.A. N.A. ft. A. fl. A.

logic and Actuation Relay M(1)  !!(1) QU) 1, 2

b. Steam Generator Water S R H N.A. N.A. N.A. it. A. 1, 2 l evel-Iligh-liigh 6.AbxiliihI((l[ater "" ~

                                                                           '"       1
                                                                                                       ~

i

d. Manual IniLiation II. A. fl. A. N.A. R N.A. 11. A. N.A. 1, 2, 3

b. Automatic Actuation it. A. H.A. N.A. N.A. M(1) M(1) Q 1, 2, 3 Logic and Actuation Relays (strs) d r. Steam Generator Water S H  !! II. A. N.A. N.A. H.A. 1, 2 ~ 3 -

LeveI-Low-1ow a ..,n n I C 4Ja.lk hful,o,e leje' M.A. M.A. bl. A . MA. ggg @M i t , t;l: g ~ " v 4 J.

                                                                                                                        /ub                      (

1AllIE 4.3-2 (Continued} P0007 h

   $5

[ G O EllGilitt RLD sal ETY TEAluRES~ AC1tlAIION SYSTEM INSTRilMENTATION SilRVEIll ANCE IqlllHEMENTS h(( m 50 TRIP s ANALOG ACTUATING MODE I CilANNEL DEVICE MASTER SLAV e CllANi;EL CllAllNEL FOR WillCli { lifflCfl0NAL UtilT CilECK OPERATIONAL OPERATIONAL ACTUATION CAllilRATION TEST TEST RELAY LOGIC TEST TEST RELAY TEST SURVEILLANCE IS REQUIRED

6. Auxiliary Feedwater (Continued)

d. Safety Injection See Item 1 above for all Safety Injection Surveillance Requirements

e. Loss-Offsite Power il. A. R N.A. M N.A. N.A. N.A. 1, 2, 3

f. Trip of All Main N.A. fl. A. N.A Feedwater Pumps '

R N.A. N.A. N.A. 1, 2 w g. Auxiliary Feedwater S R M H.A. h Pump Suction Pressure- N.A. H.A. N.A. 1, 2, 3 3, Low w

7. Automatic Switchover to Containment Sump

a. Automatic Actuation il. A. fl. A. fl. A. fl. A. M(1) M(1) 1, 2, 3, 4 logic and Actuation Q(5)

Relays (ssfQ

h. RWST Level - low-Low S R  !! ti. A. H.A. fl. A. N.A. 1, 2, 3, 4 Coincident With ~

Safety injection See Item 1. above for all Safety injection Surveillance Requirements.

8. Loss of Power

a. 4 kV lindervoltage - it. A. fl. A.

R Il 11. A. fl. A. 13. A. 1, 2, 3, 4 loss of Voltage

b. 4 kV UndervoItage - N.A. R N.A. 11 N.A. 11. A. fl. A. 1, 2, 3, 4 Grid Degraded Voltage

(ABIE 4.3-2 (Continueil) 6 EllGillEEltEI) SAFE 1Y FEAltJRES AC1tlATION SYSTfH INSTRUMENTATION G; SilRVEltt ANCE REQtilllEMEtJIS O m

                                                                                                                 TRIP j

ANALOG ACTUATING MODES CHANilEL DLVICE , HASTER SL'AV c- FOR WHICH CilANNEL CilANilEl. OPERAl!0NAL OPERATIONAL ACTUATION RELAY RELAY Fut ClluMAL tit!Il SURVEILLANCE

  .-s CllECK         CAllllRAll0N TEST                           IEST                  LOGIC TEST TEST            TEST          IS REQUIRE 0

9. Control Room isolation

a. Hannal Initiation N.A. N.A. H.A. R N.A. N.A. H.A. All ti . Automatic Actuation N.A. N.A. fl. A. N.A. H(1)

Logic anil Actuation H(1) .QO) -A+1- 4, 2, 3, 'l Relays ('>54

             .A e. Pliase "A" Isolation C*   d*E,h,2'N

• See M-item 3.a. aliove f or all Pliase "A" Isolat. ion Surveillance Requirements.

                                                                       # A-Engineerett Sa'Ded'i%)                                                                                 N A-          f4U)(1)                     Il' A -     ail kJ6. Actuation                    fety Features System Interlocks es. A-                                                       F. A -                        '

(;' a. Pressiirizer Pressiere, N.A. R

 'g H                              N.A.           N.A.            N.A.       N.A.          1, 2, 3 c                   P-ll

13. Reactor Trip, P-4 H.A. N.A. N.A. N.A.

R N.A. N.A. 1, 2, 3 ga <u, A % t, L.3 Se, < . < - N A- N.A-u.O. 9 A. A(OO) g 9.A. ) ,1, 3, $ (1) Eacli train steall lie testeil at 'least every 62 ilays on a SI AGGEREI) lEST DASIS. Co.J.. .q ,1,..L I 6e Ia=I< l fiu- t he- AC 3 A 7 8# L- " ' ' ' '517' t

                                            ..i. 7 n i((n
      'l    3)                                                                                                                                                                   T3

[,c.g} a K ,t,;.a p 1(L o 7 , K t.10, K 6 2 2 , K 61'i , A 6 3 0 , kTM c. 4 k 74 l ' h L. S b 18 be I'd' 'I , . co-b sibiht N

                                               <x<

u.&.s

                                                        ..,    is ..wA

h. ec Lem a.3 -1. d. A

                                                                                              . , <.. . i. 3
                                                                                                     %L . in, c . .. i
                                                                                                                         %.,y           ee                        -

c., h<9 g .73

                                               . u. ,, u m .,

Iss , _1

                                                                                                                                                                              . m..

C .3

                                                                                                                                                                !             CD N

i M

1 INSTRUMENTATION 1-

                                                                         ?i10078 Hell COPY 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operaticns shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within

. the specified limits. APPLICABILITY: As shown in Table 3.3-6. ACTION:

a. With a radiation monitoring channel Alarm / Trip Setooint for plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours or declare the channel inoperable. .

b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown i,n Table 3.3-6.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. '

SURVEILLANCE REOUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MODES and at the fre-quencies shown in Table 4.3-3. WOLF CREEK - UNIT 1 3/4 3-39

1

• a

                                                                                                                                                                               'd.'

g TAtil.E 3. 3-6

                                                                                                                                                                                       ~~

n RADIATION HONITORING INSTRUMENTATION FOR PLANT OPERATIONS . 2'. gr MINIMUM

        ,                                                      CllANNELS           CilANNELS        APPLICABLE FU!!CTIONAL Utill                                                                                   ALARM / TRIP c
       =                                                       TO TRle ALARM      OPERABLE          H0 DES      SETPOINT
       ~~

ACTION .

       -4    1. Containment                                                                                              g            .

d. Containment Atmosphere 64ws R3:licactivity-liigh 1 2

                                                                                                                   -M M$

All (cr-ne-se 4 n) 26 I A- L,;-t;!.

                             .          ;;rP 'ui:                                                                                  '
                                                                                                                                                       ~

d L c..., t h" =0 t-t vity-4 444tt 36 - 2 2- - All . # 20 -

c. Gaseous Radioactivity-4

{ RCS Leakage Detecticn (sr-te -se { n) N.A. I 1, 2, 3, 4 N.A. 29 Y d. Particula te N.A.

       $                                                                          1                1,2,3,4      N.A.            29 i

Radioactivity RCS Leakage Detection (tr - A t - 5 432) s;

2. Fuel Building E f. 5p# (vkyst-S.,f

a. m. . . . . ._.

(uu-a Radioactivity-lii h p 3o (6 -As-%7 4 ag) 0 1 2 ^^ g 'p"/'.

                                                                                                                               -i9-

b. Criticality- $3 h Radiation Level 1 2 *

(so-n -37 i ss) 5 15 mR/h 28 M

3. Control Room

• c3 Air Intake -6*sceus m

Radioactivity-liigh Ro

                                                                                                                   #                            m

( GK- u-M (of) 1 2 All d? "'S 27 (S ...

              ,                                          .                                                                                      CD t3 W

         . e .g . , p.       . .v o:  .~
           .m .

fi . i e 1 TABLE 3.3-6 (Continued) I g D gC {. g{} { TABLE NOTATIONS fc "With fuel in the fuel storage areas pf fuel building. of

                              **With irradiated fuel in the fuel storage areas p                 r fuel buildi n.

g k;sDfbh

                                    .8-tg                I d,s.,.U r.fa F RmACTION ClA MSTATEMENTS ulk N d k l !.     <d4Lh.1" s.A      I' *R%Af k g
                                                                                                  %=% fa [ %U kbi se =m em          A.                        4.e<
                               ,,            . h .t , ,

s ea sas s % ked o n e ru ACTION 26 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue provided the containment purge valves are main-tained closed. ACTION 27 - With the number of OPERABLE channels one less than the Minim Channels OPERABLE requirement, within 1 hour isolate the Control g U "1 RoodeclNVentilation System and initiate operation of the Control Ventilation System in the recirculation mode. ACTION 28 - With less than the' Minimum Channels OPERABLE requirement, opera-

     '         ~

tion may continue for up to 30 days provideo an appropriate e portable continuous monitor with the same Alarm Setpoint is provided in the fuel pool area. Restore the inoperaole monitors to OPERABLE status within 30 days or susoend all operations involving fuel movement in the fuel building. , l ACTION 29 - Must satisfy the ACTION requirements for Specification 3.4.6.1. Aenap 3o . 5,e .dbekJ Acb s 30 L

3 s

WOLF CREEK - UNIT 1 3/4 3-41

           -z                          .

w ,- pd,

             .t                                                                                                           _

4. r?. [ ## Trip Setp6 Tnt c5heeritratlon 8 value (iici 7cii8)is70'ts~eKablish

  ' ### Trip Setpoint concentration value (pCi/cm ) is to be established such that the actual The      submersion                dose rate would not exceed 9 mR/h in the containment fetpoint value may be increased up to the equivalent limits building.

of Specification 3.11.2.1 in accordance with the methodology and parameters in the 00,CM during containment purge or vant provided the Setpoint value.does not_exceedjthe maximum concentration activity in the containitent. determined y the sample analysis performed prict to each release in accordance with Table 4.11-2.

 .p
   ' [' ACTIONChannels           30 - With'              the requirement, OPERABLE                     number within   of OPERABLE           I hour isolate the channels Fuel         on Building Ventilation System and initiate operation of the Emergency Exhaust             S.-, tem to maintain the fuel building at a/
                        ,, negative pressure.                                              :,                                 ,

4 ., , g. w

n. r- ,...

                                                                                                                                                                                   ':. 'a
                                                                                  *                                                                                           .       .M y <
                                                                                                                                                                                     % g{
                                                                                                                                                                            ,f
      ~f r-t n                                       .

TABLE 4.3-3 . O

• RADIATION MONITORING INSTRUMENTATION FOR PLANT y OPERATIONS SURVEILLAllCE REQUIREMENTS j

E ANALOG MODES FOR M CilANNEL WilICil SURVIL-e CllANNEL CilAllNEL * ~t OPERATIONAL LANCE IS - FUNCTIONAL UNIT CllECK CALIBRATION TEST REQ 61 RED 5

1. Containment- ,

i

a. Containment Atmosphere .

S R M Al1 6esas gdi,o,agi(v,1,ty-lii0h S C ; . . i a o ... . . i-i u s u. r-- r .m. i . _, . _. ,... . _._

n. _a . ___ _ . :. ,..-

y liiGit _

 ~*    *

c. Ga'seous Radioactivity- -

     '{"
                             , g ,lgak,aq ,1 tection                                    S                           R                H            1,2,3,4

d. Particulate Radioactivity - l

    '                                                                                  S                           R                H RCS (c,r- ce       Leakag(e - n s a.)Detection                                                                       1, 2, 3, 4 2.

FuelBuildingE""$"

a. nt- "

Gaius Igli,ogcgvig-liigh S R H

b. Criticality-lligh Radiation Leve 5

• c un-seRoom -n R M N

3. Control $sg} l O C

                                                                                                                                                                            -ri Air Intake-c sms                                                                                                                                   C2O             :

Radioactivity- p ' lii 0h (6M-Ae.of f os) S R M All Q M

                     *With fuel in the fuei stora0e areas or fuel building.                                                                                                 c-a
                   **With irradiated fuel in the fuel stora0e areas or fuel building.                                                                                       o c.

INSTRUMENTATION 3i100 & EEW CDl)Y MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

a. At least 75% of the detector thimbles,

b. A minimum of two detector thimules per core quadrant, and

c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the Novable Incore Detection System is used for:

a. Recalibration of the Excore Neutron Flux Detection System,ef

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. Measurement of NF , Fq(Z) and F ACTION: ,

a. With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours by normalizing each detector output when required for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or N

c. Measurement of F , F (Z), and F q .

WOLF CREEK - UNIT 1 3/4 3-43

l l l INSTRUMENTATION i h hh ' SEISMIC INSTRUMENTATION

                                                                                     -~

LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE. APPLICABILITY: At all times. ACTION:

a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction ano the plans for restoring the instrument (s) to OPERABLE status.

b. The provisions ~of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS - 4.3.3.3.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-4. 4.3.3.3.2 Each of the above required seismic monitoring instruments actuated during a seismic event greater than or equal to 0.01 g shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 10 days following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 14 days describing the magnitude, fre-quency spectrum, and resultant effect upon facility features important to safety. WOLF CREEK - UNIT 1 3/4 3-44

                  -8                                                                                                     *,                                                                                         *
                                                                                                      ..e r                               -              -
                                                                                                                                                                                                                      ./

m. y w:.w;@[..

      ,p.,
     . . . . % Q,'.no
               =

4 urm.

                                        >r 9.,.,a . .' d* .*
                                                    .. :.: '. y .w w..<~.'.t.
                                                                                                                                                                                                                                       *    ~.~. m .            .         . . - .

c.4 n... y v.u.w. s . .... . . . t 9 hk h'.g.n.,U.,.b.S, 5 7..,.. ,

                                                                                                                                           .                                                                                                       IJ
      ..f-je.k.                k.,.e.hd.t,                      ,                                             .

c TABLE 3.3-7

   .<             ygf'.".~jj.;."'-                                               .               _
                                                                                                                  ' SEISMIC MONITORING INSTRUMENTATION
     . m . ~.. : g-n,t.y %.a.. ru eu.; .3-                                  ~.              .s,                                                                                                                                -                -

e..  : MINIMUM

                                                                                           . .'u
                                                                                                                                                                                    .r 6.Ei.7$;2 d$q@6 3 /s.e... x" .
     ;'.2Q D @ , h.i.M:e'                                                                        , 2 6., 4 . w. .6 ';                                                                                                                                      ,
                                                                                                               .                                               A ,- %.J;h-c.                                 MEASUREMENT _ -G                                            INSTRUMENTS ~'                                ;
  .$g. $w $. INSTRUMENTS AND SENSOR LOCATIONS
                           .p or. - .                                                                                                                                 . ~. . .. _ .                                      RANGE                           ,7                OPERABLE-                                         .'

y ,, , . r. .m,.. . .'..,.3 . .. ., -

, ,. 2 .;

K $c:

                                                                                                                                                                                                 . ,.                                      3 . . . . . .
   .p,9.               Jap %e n:12 Triaxial Peak Recording Accelerographs~'                                                                                                                                                              9" %~" :.. ~,.q.

4 21 b '

                                                                                                                                                                                                                                                                                                                           . . ,i
                                                                                                                                                                                                                                                                                    .~ , .n. -
  $m@,:MTSS'                                 :# ?;s. .w&                    ,.c       m.

q ,mBase

                                                                                                           ~
                                                                                                           ; m Slab
                                                                                                                                .o.s.,.
                                                                                                                                                         . . w . .. ..,.~..                ..
                                                                                                                                                                                                      . p l.Og'       .. .
                                                                                                                                                                                                                                  -..    .-                  n r

Radwaste ' ~.:.s - 1 1 P '~ bi Control Room O - '

                                                                                                                                                                                                                                                          ~ O                                                         -E
  .k'r @9.f W .? d D. ai.h'$.hk%.$.Qf7cl.                                                                                                                                                                              1 Og $ 6ESW
                                                                                                                                                                                                                                                                                 - l' '.

Pump~Fac

  .i 4.ykf?y$U.df
                                                                                                                    ~

4 .- " +1.0g k D ENsM "1:2.0g SRV~ 4W i d e f,T Fd"7 Ctmt-Structurei- F Suctions@ '2': -  : AuxiliaryJBldg. SI. Pump " 1.Ogi 4 'sMA','M. 1 W M U W'.- I s.7.@bb.h'@M.f M.P W -%. SGB Piping %.m'd %M a6.m ..v. - .n. . .a _.-

                                                                                                                                                                                     .4-                     + 2.0g. M                                           M'                        S
  .$.c.~                 h:,.,w      .V [k,:w,.t.               9    p    SGB       ,
                                                                                             .5upportyg,3-c.:s.                .
                                                                                                                               ;     y,qT.$C                        et               '.         "     '                           r.....

1. Og @*;6.w5Mj? M E..q '  ?;....

l'.o. "s3.' @%;;. :M. w$M.:'. w% .ns . .w.st.) 9 m t'  ? - - a - em 1 '. '; .r *c

                                                                                                                                                                                                                                     - Y. . : +'?' -

fM8dMEM2.9191' axial ~ The History and Response' ^

     !.Dbf[ .591W.E.T m/J          M.                            Ik              Spectrum Recording System, Monitoring (Active)                                                      .

j:.D

                                                                                                                                                                                                                                        .9,;       '    .  "     .      '
                                                                                                                                                                                                                                                                                                              ~

r~

                                                                                                                                                                                                                                                                                                                                   " C.'.T[i s,e.s,,M. wE

o M

                                        .:n.e n v : .    .     'the.,Following           ,

Accelerometers. .. e n n

                                                                                                                                                                                                                                                    .<~

a. Ctmt. ~ Base' Slab'.' '

N 1 Sh

   -M'M5@!?h}['                                                      b.

iWME. Ctmt. Oper. F1cor, ' 1.0g 1.0g '_. #..' .j.->9 .- W+

   @QG'ju..grpb.W>g.ir    e                w .. .c?. Reactor Support .' ~. .                                                                  n".'.-

A W: 1

1. ' M.,W:

                                                                                                                                                                                                                                                                                                                                 '+ -

Aux. Bldg. Base Slab' . . t 1.0g 1. 0g ,.? W.n$ W.c FF A l

                                                                                                                                                                                                                                      -1                                  -
                                                                                                                                                                                                                                                                                                                  ~~          -
                  , ;,!q.,,,fq;t;.                            d.                                                                                                                                                                         9
                   $,9 .

i 1 0g e r m' A. 1; 1

                                  .v "r 6.4%, r. Free F.ield" W.k..U...A..i;iy,R..??

n. yy .~ c.

                       % E dis 9?dFMid N!E N O N . * ^ .Y - @@N..d. @ .! 4 W M M
                                                                                                                        ' elc Aux. Bldg.+ 0.5g Control      W..?.. W..Room              N.., D 1'.

Air, Filter P d ff %.y $ T3$'. Triaxial'Res@ponse-Spectrum Recorder' r lh j $ " g-M M' M e .M /tp$,yyg'.M ffh:9p ' @ W L WW' *'WQ:d% +~ % -

                                                                                                                                                                                                                                   ?-
     ..m.3.g,L Q                            $$m.Ens;(Passive)*%C..

v a.9A- 2: .W >.

• 149l w..

c g.#;* t,tqqw?;g: *:l,. l?. ."4a. .

u3 li * ' 3 -5W M> o. . .::: . L- .a ; , . . - .

                                                                                                                                                                                             ,~
                                                                                                                                                                                                                             + . s , v. .            .                      w        - . . . .

Ctmt.

                                                              . .. wBas
                                                                                                                                                                                                                                                                                                                     . . ;. . .n
   *Q#%pAlkQ;i:?          2                                                   e e Slab " , '. . vW                                       M n ~v..                                                                   1.

,.r 0g..c..%&m-r

                                                                                                                                                                                                                                                      *%g      ~ . -g;n: 1,L                               ' :..:

t ~ ~%hn,%e h .w-=.>+- n~;w.i&

      ..                                                                                       . - ~ ~          +*:,. ^ X o
                                                                                                                                                           .' '.- t w                       ,

v > ~ v W W X ,. . v: ' s ':!: a.L*.'. *

                                                                                                                                                                                                                                                                                                    ' w.W; 'c . . '               .
     %. ?. U.im.:.1,4"(a.,4C,.n.J.riaxial , Sei smic Swi tches                                                                                                 .>J, ..>:,t e

n ,o. ACCELERATIOr.; .' .c e '.+.;; W. W; ?.a ; . x ;y-n;4f s.,-ves k-C,p"w,L -in./,O:,P. ' 9. a.Q.s. M.%W~m'.d, .% 4.Q.,

• M.rms ,a. A:  ::7 -- ' . < , .. # = LEVEL.W 'e. cO'.e.. * ~

                                                                                                                                                                                                                                                                                                                                 .n. .

ec 4.. n e. . te . +: s-

                                                                                                                                . g e . n, *N l ?.
      &,k,e"3$R,th.f"heiS$..Ne$Yf *f                                                                                                      :~

e -

                                                                                                                                                                "&Y;.k                                  l*!t                  ' $'" "';jf'. .T. ,'.: .,..&sD.OBE                  . %+i'Wl f y&,M. .f. d.%

s 78:.$ p a. .. O.06glM

                                                                                                                                                                                                                                        .1 Tf.

W %.9 5, Q (,bi :SSE Ctmt.0per Base Slabf : itf S . e,oa,3 3 ;12 i lt!M ~.8 Qi.g;Q,.c,.y 0BE Ctmt[r0 F12,'- peri. ,lin (".%Qo4M:. F1Wl[M.$W'S'Qy@G ;0.15g

3 @

1 pQ g Mr. pqlW Wlg.pt M@'Mys W '^

   ., i-h 3.2di..SSE.Ctmt.
  $.                                               #e~6LSy                                                                               EQ%hoh.D:.01h                                                              h.Tg e.Mg.&.:4caly[15[n 'O.                                                                  tr h.M          e..h.(h'          . W. mmh: stem Trigger                                                                             d                  l?@dh5G'#hNWsiW@gP.*M.QQEdfg.71                                                $%%$$N.W. $M %&&
  ?.Y:     E   W$lWUh,)&yWh:

d W. o NW.%CdM@n@$8:5$9'N$1;Q$$::N$Mkb$&???i.8(4.i.Whl5 Q&. .$,...

    .7W,%p f..

s

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4 a R . g.cY....y,cE5-&-

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                                                  . @WMNEh.w%%,w$p&,. :M.

u/ w $ .w. .-- r W$s$$. c.$lM

                                                                                                                            ,mi.

ob.W. . n.~ ?.2. 4. -E'5

                                                                                                                                                                                          . .. W...Wf           .W   7,.              p..M~ r.5Y.'.$$.w$NW.t:r.p,           m~.p,. .m >  r :, . P...u.r:.      2./.m.y        7t h.

o. :: -

W'MW WCIW."N. i .m~%*A"fM??iW WW Y.xO . M.::. Q., yp$ Nf--%.e%W.i?!%~M'.W@.w .%:tM.%% htf W bW X+ j'NM

   \R .R % &M& M k2 8 % ?NDt!W W M :
                                                                     %f WidlMi@dkMS&M$3RN                                                                   8,~,OW@$QM4p$s@W.W%.                                    Il ' W!b$%M ^'                                                                                              5 ANT                P.%
  ; nW?fh%?Qf.1%.s%Gl5W%.W5%WGWlNSWW?W9M6WMh%iW.3 W           ASW.#%                                          &j&

h9nb.f&*W$ h? r%... w

                                      %n .M. % . Mn@a,.@.np%.'lM.y.m(.'.IIk'@.QS%'h.~;w@
                                                     /D                      .$                                           &O              f $ &ntk             e                             $u.M.cD                              Eln. N .5v. % W;%'$                                                   $%..n>h.e.s                 FE, %

g. h&4 .. 5 An 5 t.Y,bg4hg 5SN h,.-.u..u..

                                                                                                                          .     ,        o. if w vr 5 .      m,.                    w.:n.c@ym
                                                                                                                                                                                 .m,u

p. t u.. . . . , . .

S... ..i : w

                                                                                                                                                                                                                                                                                                          .'-                       w.. .-

g < w ..%+-=O.k. .w ft:.&..c,$'5 k

                                                                                                                                         .I'(,,.IWllp.:..                                    ve        hnllj/&s.?:i                 *
                                                                                                                                                                                                                                    '.    . Yc..

y, A. p,w,. . q, W ! &.m. 8 R,.y:* & p.w:4.j)n.?gW. wt g .g, ur c 4 g, W M M  : ;p M .t E*'U 3/4.3-45e'g.u.p$.?.?v

                                                                                                                                                                                                                           .h                 5                0             M.                       m.                             N                 ~

f,u$,

  ?g.

s MS$3N . .,w.mp,s.c.My MM

y. u. Y.6M.qplF ppfin 7m.eh.

                                                       .%,a' ,ww w;.m<:                            3         CREEK             . m,                   ww.                         {;IS e ... UNIT     .

wmmwv

av.n.<. :p @b@. m pt?.m.

                                                                                                                                                                                                                                                                                      .m lN.k          , .m./M       %      .n W.a w .W%%,.; x:.qgq.:N '4. 9. u W.. W W 3 M y p W W W . w s.8 4 N W M M We.

M.QM

PRODI& RBH COPY TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS ANALOG - CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST l

1. Triaxial Peak Recording Accelerographs l

f

a. Radwaste Base Slab N. A. R N.A.

b. Control Room -

N.A. R N.A.

c. ESW Pump Facility N.A. R N.A.

d. Ctmt Structure N.A. R N.A.

e. -Auxiliary Bldg. SI Pump Suction N.A. R N.A.

f. SGB Piping N.A. R N.A.

g. SGB Support N.A. R N. A.

2. Triaxial Time History and Response Spectrum Recording System, b.tonitoring the Following Accelerometers (Active)

a. Ctat. Base Slab M R SA .

b. Ctmt. Oper. Floor M R SA

c. Reactor Support M R SA e -

         ,         d. Aux. 81dg. Base Slab                                     M                       R              SA + +

e. Aux. Bldg. Control Raom Filt*ers M R- SA * *

f. Free Field M R SA 4

• L3908-

3. TriaxialsSpectrum Recorder (Passive)

18. +

Ctat. Base Slab N.A. R SP

4. Triaxial Seismic Switc.nes

                                                                            ~

a. OBE Ctat. Base Slao M R SA

b. SSE Ctat. Base Slab M R SA

c. OBE Ctat. Oper. F1. M R SA

                                                                                                                                 ~

d. SSE Ctmt. Oper. Fl. M R SA

e. System Trigger M _

R SA

                  + cmau:.ny             a   4g        gag     (;.J, a 5,uA A nn.na, An fe ,,sg                           cl.c m
                                           %        ( .Jr.>               sksil ve. p hr"cl J                   iu.&      c~s ed(d       ss                           Co om pr        if+         days.

6x ii. b,Mie M nd Lde ua..aa( r b n, -i k ju, Le .,~u u ef 4.,p%nsdpe. A u A.d-uz- 08%h ' MEW M + t_ Ye w). i {

           'a0LF CREEK - UNIT 1                                  3/4 3-46

INSTRUMENTATION

                                                          ,:P.00.: & Uhi VI L)OPY METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4    The m3teorological monitoring instrumentation channels in Table 3.3-8 shall be OPERABLE.

APPLICABILITY: At all times. ACTION:

a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

b. The provisions'of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCEREOUIREMENiS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-5. l l WOLF CREEK - UNIT 1 3/4 3-47 i

TABLE 3.3-8 ' PRODF & REVRI CO METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATI N OPERABLE

1. Wind Speed Nominal Elev. 10m 1 Nominal Elev. 60m 1

2. Wind Direction Nominal Elev. 10m 1 Nominal Elev. 60m 1

3. Air Temperature - AT Nominal Elev. 10m-60m 1 J

                           ~

l l 1 WOLF CREEK - UNIT 1 3/4 3-48

TABLE 4.3-5 l -- MOOF u REV'EW COPY! I METEOROLOGICAL MCNITORING INSTRUMENTATION , SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL

    ' INSTRUMENT                                  CHECK      CALIBRATION

1. Wind Speed

a. Nominal Elev. 10m 0 SA

b. Nominal Elev. 60m D SA

2. Wind Direction

a. Nominal Elev. 10m D SA

b. Nominal Elev. 60m D SA I

3. Air Temperature - AT

a. Nominal Elev.10-60m D SA F

l I

   'a0LF CREEK - UNIT 1                    3/4 3-49 l

L -

INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels given in - Table 3.3-9 11 be OPERABLE with readouts displayed external to the control room. g,_ anA ,t,,;g,, g(t,,,, p.,t (,*sv.) c a6is APPLICABILITY:# MODES 1, 2, and 3. ACTION:

a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore to ta A le a M mW x the inoperable channel (s) to OPERABLE status within 7 days l e 6 ebtem se, w ., + c .t (, k.m .-A, PHOT SHUTDOWN within -

                                                                           .&tl the  =t 12 hours.
                                                                                  .$ , 4

c. M.

                         .         The provisions of Specification 3.0.4 are nnt applicable.

b. 14 A A ALP cc1.b lacpu. Ale ecsbra, A *inep. .ad e. AsA ct.dn.g

4. cres.Ae.4 -)sh.s .. tw o 7 q s -[ cth u is. m , de a x4 leusf kr trAww w. N ., h nes (, x, ,5 Ana in 4 r- w%u 4. A %

Mcwg bl%<s . SURVEILLANCE REOUIREMENTS 4.3.3.5.1Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies given in Table 4.3-6.

w. s M ,d 453.5.2.  % 650 cc4A Atl de. o ftv.Au-rads , h ock u.k f 4 <(

d leut e ,s e. pu lt

                                                                                                *[e M "3 cqcned               &cm         twe.      asp.

4.53.6 3 ef 4 # 'I A '*- "*4 'W '#"'id '-

                                       %            ,y3;. 5                      %:*Iliede

• hche  % '% k st -cktN'^ M. l. Icehu fe' L. A 5 Sw be. a ke c.' dwg v'W88-gp e<

V Do <8!c,i. 4.- h ,.+ c. ; .. . . , / 6, oc cp . ~ 4 /, h b n e_ Ay G o, w s.,

                      .        AsA,     1.       n., /.       . kh        z.      abc.         % A (n b ,.C.Jc L I r       L W w.kT w ep..J .

WOLF CREEK - UNIT 1 3/4 3-50 l

i specification 3.3.3.5 Justification -

                    '1he asterisked note has been added to the applicability requirenents _of this specification to be consistent with the
               . applicability requirements of Table 3.3-1 (item 6).          -

i i a S ( f, i' o D a

TABLE 3.3-9 - g REMOTE SilulDOWN MONITORING INSTRilMENTATION

   .      G                                                                                                   '
       ~                                                                       TOTAL NO.         MINIMUM S                                              READOUT                  OF             CHANNELS E

n INSTRIIMENT LOCAlION CilANNELS OPERABLE ,i [ 1. RCS Pressure-Wide Range ASP

• 2 1 z

U 2. Reactor Coolant Temperature-

         -          Cold Leg                             ASP"                 4                  1

3. Source Range Neutron Flux ASP

• 2* I

4. Reactor Irip Breaker Indication RIS** 1/ trip breaker 1/ trip breaker Reactor C

• perature - ASP
• 2 1

                   -iwerage- Hot ( $                -

u 6. Reacto- Coolant Pimp Breakers ^^^ 1/ptmp 1/ pump D w 7. Pressurizer Pressure ASP ^ l 1

         .A

8. Pressurizer Level ASP

• 2 1

9. Steam Generator Pressure ASP ^ 2/kmpshare gew=/d 1/leep sfram geanh

30. Steam Generator Level ASPA 2/steata generator 1/ steam generator Auxiliary Feedwater Flow Rate II. ASP ^ 4 1 i

                                                                                                                   -r.1 l             31,    Ac.Luy FedA b6 freu.,e              Asta                  3                 -
                ' Auxiliary Shutdown Pauel                                                                         c3
                                                                                                       .           -i
               ^^ Reactor Trip Switchgear                                                                          M
             ^^*13.8 kV Switchgear                                                                                 ::c3 rg h

r-r1 C"3

C3

                                                                                                                    - t:3
                                                                                                                     -4

                                                        .        TABLE 4.3-6                                        .

g REMOTE SHUTDOWN MONITORING INSTRUMENTATION' q SURVEILLANCE REQUIREMENTS g CHANNEL CHANNEL m INSTRtlMENT CllECK CALIBRATION [ 1. RCS Pressure - Wide Range M' R z

2. Reactor Coolant Temperature - Cold Leg M R

3. Source Range, Neutron Flux -

M R

4. Reactor Trip Breaker Indication M N.A.

S. Reactor Coolant Temperature - flot Leg M R

6. Reactor Coolant Pump Breakers N.A. N.A.

7. Pressurizer Pressure M R y 8. Pressurizer Level M R

9. Steam Generator Pressure M R

10. Steam Generator Level M R

11. Auxiliary Feedwater Flow Rate M R .

t2. Am 1.6 y U<h.Je r b f N Ii I* " ' M C- ' m I P :8 c:3 C=3

                                                                                                           -ri CO

x2 m

M c2 C3

                                                                                                            -O u

r 1 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION tO LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLICA8ILITY: MODES 1, 2, and 3. ACTION:

a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3-10, restore fh,e, inoperable channel (Q$o OPERABLE status within 7 days' l r, ij a,tg,eagHjJ_ ,

within the next k2* hours.ted 'a c k.J

b. With the number cf OPERABLE acudent monitoring instrumentation '

channels except the reactcr coo h nt radiation level monitor and the

     ,              unit vent - high range noble gas monitor,less than the Minimum Channels OPERA 8LE requirements of Table 3.3-10, restore the inoper-able channel (s) to OPERABLE status within 48 hours;Wbe in at least HOT S,WWEGWN within the next 12 hours. tad go #cr serve # A *e /*###) d ' ?

s4~ody e d r. The provisions of Specification ;.0.4 are not applicable. cMu k A.i c. SURVEILLANCE REOUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7. WOLF CREEK - UNIT 1 3/4 3-53 i

W ACTION: c. With the number of OPERABLE channels for the containment radiation level monitor or t.ke unit vent - high range noble gas monitor less than the Minimum Channels OPERABLE requirements of Table 3.3-10, initiate the preplanned alternate method of monitoring the appropriate parameter (s) within 72 hours and either restore the inoperable channel to OPERABLE status within 7 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days that provides actions taken, cause of the inoperability and plans and schedule for restoring the channels to OPERABLE status. l 4 I e

             -   -      ,-..-._.--,m- - , .%_,__.,w,,.. - . . , .   ...-y.-r----_   ,_ _ _ _ . , . _ , - - - - - - ~,...-.,<..-,--_c-- - - - , - - , - -.- - . . .

                                                                                    ~

1 Alli E 3. 3-10

,.                                                       ACCID [llI iknill0Ril1G lilSTRI!!1Elif Alloti

? m TOTAL MINIMUM g ' t10. OF CilAfillELS m IliSTI:llMEllT ty - CilAtitlELS OPERABLE a)dcrawl

1. 2 Containment Pressure - 4hta-1 Rang 65 E=f<.d ed (eeg 2

                                                                                                                  ,        I g            2. Reactor Coolant Outlet Temperature - lil0T (Wide Range)                         2                       1 s

,, 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1.

4. Reactor Coolant Pressure - Wide Range 2 1 S. Pressurizer Water level ~

2 1

6. Steam Iine Pressure 2/ steam generator 1/ steam generator

            /. Steam Generator Water Level - fla: Tow Range                                    1/ steam generator      1/ steam generator

8. Steam Generator Water Level - Wide Hasige 1/ steam generator 1/ steam generator

$ 9. Retueling Water Storage Tank Water level 2 1 y 10. Containment llydrogen Concentration level 2 1 $ 11. Anxiliary feedwater flow Rate 1/ steam generator 1/ steam generator 17---Re*etor-Curlon L-Sys tem- Subcoo l i ng 4ta rg i n - Hottittte Y 4- ~

    #1 M.         PORV Position Indicator ^                                                      1/ Valve                1/ Valve s5 14.       PORV lllock Valve Position Indicator ^^                                         1/ Valve                1/ Valve
     ,4 1Y.      Safety Valve' Position Indicator                                                1/ Valve                1/ Valve
    ,5 M.        Containment Water level                                                         2 1

m iG V. Containment Radiation Level (lligh Range) e7 M. Ibermocouple/ Core Cooling Detection System j 24 A- l' $

                                                                                              / *l /c"e pal <*y+-     (t 1/ N* 1.~64.          E],

SL1W--=iteactuctoa%kC?Tiliana,. Lai1 T M \',.  ;-i& , GO iA 20. tinit Vent - liigh Rangef floble Gas Monitor ' i M ae.k p IM. W b_.

             ^llot applicable if the associated block valve is in the closed position.                                                         r~m
           ^^tiot applicable if the block valve is verified in the closed position and power is removed.                                       '
        ^^^ t t-inoperable--Actb. o. i.. S i mc i; ics L ;i., 3. 3. 3. E cha! ' he .:pp! icahte.
                                                                                                                                    ,          Q
        -fr -nu ,)               l _- ~F    '

k r y ~4- U* d'O -'*"M'^j'r " ' b# "' ' Ni%'y - A bnf k p >

                               ,=

3

TABLE 4.3-7  : ACCIDENT HONITORING INSlRilHENTATION SURVEILLANCE REQUIREMENTS G CilANNEL CilANNEL g INSlRilHENT CllECK CALIBRATION r

    ]          1. Containment Pressure - C..a! L..g                                  H-                  R e        2. Reactor Coolant Outlet. Iemperature - Til0T (Wide Range)           H                   R

{

3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) H R

4. Reactor Coolant Pressure - Wide Range H R S. Pressurizer Water Level H R

6. Steam Line Pressure H R

7. Steam Generator Water Level - Narrow Range H R

8. Steam Generator Water Level - Wide Range *H R

9. Refueling Water Storage Tank Water Level H R D 10. Containment flydrogen Concentration Level H R

{ 11. Auxiliary Feedwater flow Rate H R tr. Reacto.- Cuoiosic 5yst . E coe!!ng Mara,1 i n Hunicu m -H--- 4 i 2. M . PORV Position Indicator

• H N.A. -T3

3 15 M'. PORV Block Valve Position Indicator ** H N.A. c.:2 c:3 Aj M. Safety Valve Position Indicator H N.A. m

             )

15 1 . Containment Water Level H R U#

                                                                                                                        ;t.3 sv 14. Containment Radiation Level (High Range)                            H                  R***

rg-8714. Thermocouple / Core Cooling Detection System M R p:i-- J9-- -Reactor-Coolant- Radiation-Leve1 -# - M- 4-- 4

         #1( 20. Unit Vent - liigh Range Noble Gas Monitor                           M                  R O

c3

                                                                                                                         -c 4
                *Not applicable if the associated block valve is in the closed position.
,              ^^tiot applicable if the block valve is verified in the closed position and power is removed.
             ^^*CilANNEL CALIBRAIION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a osie point calibytion check of the detector below 10,R/h with an installed or portable gamma source.
           .u , o,                  n                                                   A N              #"']

• _e ,, m

                                                        ,.7.. I - <* e w J f

(> ,L' -

INSTRUMENTATION CHLORINE DETECTION SYSTEMS 1> 00c~U" H?# COPY LIMITING CONDITION FOR OPERATION 3.3.3.7 Two independent Chlorine Detection Systems, with their Alarm / Trip Setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppm, shall be OPERABLE. l APPLICABILITY:' All MODES. ACTION:

a. With one Chlorine Detection System inoperable, restore the inoperable system to OPERABLE status within 7 days or within the next 6 hours initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation made of operation.

b. With both Chlorine Detection Systems inoperable, within 1 hour initiate and maintain operation of the Control Room Emergency Ventilation System in the recirculation mode of operation.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.3.3.7 Each Chlorine Detection System shall be demonstrated GPERA8LE by performance of a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least 6nce per 18 months. t WOLF CREEK - UNIT 1 3/4 3-56

l l INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE. ACTION:

a. With any, but not more than one-half the total in any fire zone, Function A fire detection instruments shown in Table 3.3-11 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.6.

b. With more than one-half of the Function A fire detection instruments in any fire zone shown in Table 3.3-11 inoperable or with any Function B fire detection instruments shown in Table 3.3-11, inoperable, or with any two or more adjacent fire detaction instruments shown in Table 3.3-11 inoperable, within 1 hour establish a i!re watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the t.strument(s) is located inside the containment, then inspect that containment zone at least once per 8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.6.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

, SURVEILLANCE REOUIREMENTS 4.3.3.8.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERA 8LE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST ducirg each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months. 4.3.3.8.2 The NFPA Standard 720 supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. WOLF CREEK - UNIT 1 3/4 3-57

T^8'e 3 3-11 MOD:& REV!EW COPY  ! 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. Emp. Rm. 8 101 2/0 1108-Safety Inj. Pmp. Rm. B 101 2/0 1109-Res. Ht. Remov. Pmp. Rm. B 101 1/0 1110-Ctmt. Spray Pmp. Rm. /6 101 1/0 1111-Res. Ht. Remov. Pmp. Rm. A 101 1/0 1112-Ctmt. Spray Pmp. Rm. A 101 1/0 1113-Safety Inj. Pmp. Rm. A 101 2/0 1114-Cent. Charg. Pmp. Rm..A 101 2/0 1115-Pos. Disp. Charg. Pmp. Rm. 101 2/0 1115,1117-Boric Acid Tk. Rms. ~ 101 2/0 1116, 1117-Boric Acid Tk. Rms. 101 2/0 1120-Aux. Bldg. 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. & Pmp. Rm. 101 1/0 1127-Stair A-Z 109 1/0 1128- 117 2/0 1130-Aux. Bldg. 1974' N. Corr. 100 0/2 1206-W. Pipe Chase Below AFWP Area 117 2/0

         -Aux. Bldg. Elec. Chase S. 1988'       117                     1/0 1301-Aux. Bldg. 2000' Corridor #1            103                     0/10 1301-Aux. Sidg. 2000' Corridor #1            117                     2/0 1311-Aux. Bldg. Sampling Rm.                117                      2/0 1312-Boron Meter /RC Activity Mon. Rm.      103                      0/1 1314-Aux. Bldg. 2000' Corridor #3           103                      0/3 1314-Aux. Bldg. 2000' Corridor #3           117                      2/0 1315-Ctmt. Spray Add. Tk, Area              103                      0/2 1316-Viv. Rm. 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. Rm. C                     111     /         7/d 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. B1dg. 2026' Corridor #1 IJ4 0/6 1403-MG Set Rm. 105 0/9(3) 1403-l'G Set Rm. 112 0/9(1) WOLF CR'5K - UNIT 1 3/4 3-58

[ABLE3.3-11(Continued) FIRE DETECTION INSTRUMENTS TOTAL NUMBER -- 0F INSTRUMENTS

• INSTRUMENT LOCATION ZONE HEAT FLAME SMOKE (x/y) (x/y) (x/y) 1405-Chemical Stg. Area 118 6/0 1/06-Comp. Cool. Pmp. & Ht. Exch. A 104 0/1 1406-Comp. Cool. Pmp. & Ht. Exch. A lla 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 1) 1409-Elec. Pene. Rm. B 113 0/4((1) 0/4 1410-Elec. Pene. Rm. A 107 0/8(1) 1410-Elec. Pene. Rm. A 114 0/8(1) 1413-Aux. Shutdown Pn1. Rm. 118 4/0 1501-Ctri. Rm. A/C & Filt. Units 8 110 10/0 1504-Ctmt. Purge Exh. & Mech. Equip. 8 108 13/0 1506-Cmt. Purge Sup. AHU Rm. A 109 18/0 1507-Personnel Hatch Area 108 3/0 d!512-Ctr1.Rm.A/C&Filt.UnitsA 110 10/0 1513-Ctri. Bldg. Vent Sup. A/C Unit Rm. 109 3/0 Aux. Bldg. Duct 2047'6" 119 1/0 l Containment ** 201 1/0(2)

Containment ** 202 2/0(2) l Containment ** 203 1/0(2) l Containment ** 204 1/0

    ?

Containment ** 206 Containment ** 215 3/0(2) 1/0 I Containment ** 216 Containment ** 217 1/0((2) Containment ** 218 1/0(2) 1/0 2) Containment ** 219 4/0 Containment ** 220 1/0(2) 3101-Ctri. Bldg. 1974' Pipe Space 300 11/0 3105-Ctri. 81dg. Elec. Chase S. 1974' 300 1/0 { 3106-Ctr1. Bldg. Elec. Chase N. 1974' 300 1/0 ' -Area Above Access Control 301 12/0 3229-Ctri. Bldg. Elec. Chase S. 1984' 300 1/0 3230-Ctr1. Bldg. Elec. Chase N. 1984' 300 1/0 3301-ESF Swgr. Rm. #1 314 0/7(1) 3301-ESF Swgr. Rm. #1 315 0/7(1) 3302-ESF Swgr. Rm. #2 316 0/5(1) 3302-ESF Swgr. Rm. #2 317 0/5(1) 3305-Ctr1. B1dg. Elec. Chase S. 2000' 301 1/0 3306-Ctrl. Bldg. Elec. Chase N. 2000' 301 1/0 3403-Non-V!t. Swgr. & Xfmr. Rm. #1 304 3403-Non-Vit. Swgr. & A fmr. Rm. #1 305 0/1((g) 1) 3404-Switchboarc Rm. 44 321 0/1(1) 0/2 3404-Switchboarc Am. 44 322 0/2 CU 3405-Battery Rm. #4 303 2/0 3407-Battery Rm. #1 303 2/0 3408-Switencoarc Rm. #1 325 0/2( ) (%0LFCREEK-UNIT 3/4 3-59 a ., . ro.a .. . . . 1 . . ,e 4 l l$dI - Aa.= %..a 6 ,We kW1 l4S 9

e e LE 3.3-11 (Continued) ' h[ h ), i COM FIRE DETECTION INSTRUMENTS TOTAL NUMBER - 0F INSTRUMENTS

• INSTRUMENT LOCATION ZONE HEAT FLAME SMOKE (x/y) (x/y) (x/y) 3408-Switchboard Rm. #1 326 0/2(1) 3409-Non-Vit. Swgr. & Xfmr. Rm. #2 323 3409-Non-Vit. Swgr. & Xfmr. Rm. #2 0/1(1) 327 0/

3410-Switchboard Rm. #2 324 1) 3410-Switchocard Rm. #2 328 0/2(1) 0/2 3411-Battery Rm. #2 303 3413-Battery Rm. #3 2/0 303 1/0 y) 3414-Switchboard Rm. #3 318 3414-Switchboard Rm. #3 320 0/2(I) 0/2 C 3415-Acc. Ctr1, & Elec. Equip. A/C 303 4/0 Units #1 3416-Acc. Ctr1. & Elec. Equip. A/C 303 4/0 Units #2

• 3418-Ctr1. 81dg. Elec. Chase S. 2016' 303 1/0 3419-Ctr1. 81dg. Elec. Chase N. 2016' 303 '1/0

                 -Ctr1. Bldg. Elec. Chase N. 2016'        303                       1/0
                 -Ctri. 81dg. Elec. Chase S. 2016'        303                      1/0 3501-Lower Cable Spreading Rm.                306                     0/13 3504-Ctr1. Bldg. Elec. Chase N. 2032'         303                      1/0 3505-Ctr1. 81dg. Elec. Chase S. 2032'         303                      1/0
                 -Ctri. Bldg. Elec. Chase N. 2032'        303                      1/0
                 -Ctr1. 81dg. Elec. Chase S. 2032'        303                     1/0 3601-Control Room                              308 3601-Control Room                              309                     4/0(1) 0/7 3601-Control Room                            g19                       0/7(1)

AN - @) 73602-Pantry 308 "# 71/0 3603-Shift Supv. Office 308 1/0 3605-Equipment Cabinet Area 308 13/0 3606-Emerg. Equip. Storage Rm. 308 1/0 3608-Janitor's Closet 308 1/0 3609-SAS Rm. 308 1/0 3617-Ctr1. 81dg. Elec. Chase S. 2047'6" 308 1/0 3618-Ctr1. Bldg. Elec. Chase N. 2047'6" 308 1/0

                -Ctr1. Bldg. Elec. Chase S. 2047'6" 308                          1/0 3801-Upper Cable Spreading Rm.                307                     0/18 3804-Ctr1. 81dg. Elec. Chase S. 2073'6" 308                           ' 'a
                -Ctr1. Bldg. Elec. Chase S. 2073'6" 308 1/0 5201-W. Diesel Gen. Rm.                       501             4/0 5201-W. Ofesel Gen. Rm.                       502    0/8 5203-E. Olesel Gen. Rm.                       500             4/0 5203-E. Olesel Gen. Rm.                       503    0/8 6102-Fuel 81dg. Rafircad Bay                   600    0/3 6104-Fuel Pool Cool. HX Rm. 8                 601                      6/0 6105-Fuel Pool Cool. HX Rm. A                  601                     6/0 6202-Elec. Equipment Rm.                      601                      3/0 WOLF CREEK - UNIT 1                     3/4 3-60

y- i TABLE 3.3-11 (Continued) ht b. t FIRE DETECTION INSTRUMENTS

                                                                                                                ~

TOTAL NUMBER OF INSTRUMENTS

• INSTRUMENT LOCATION ZONE HEAT FLAME SM0KE (x/y) (x/y) (.x/y) 6203-Air Handling Equip. Rm. 601 3/0 6301-Fuel B1dg. 2047'6" Gen. Fir. 602 2/0 6303-Fuel Bldg. Exh. Filt. Absorb. 601 2/0 Rm. A 6304-Fuel Blag. Exh. Filt. Absorb. 601 2/0 Rm. B

                 -Ner% ESW Pumphouse vrain a                  002                     3/0
                -Seeth ESW Pumphouse hi A   '

001 3/0

                -ECU C::1!n; 7~er              -     '
                                                           ~ ~401                    -tr+
                -ESM C:: ling h er                            462--                   1.J ';
                -tv w.u fc w we i                             o.u   %
                -t w Trmk.vo xch '      ,     TABLENOTATI0 tis
             *(x/y): x is number of Function A (early warning fire cetection ana a               notification only) instruments.

y is number of Function B (actuation of fire suco'ression 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.

(1) Zone is associate [ wl 5 [ Halon protected scace. Each scace has two separate detection circuits (zones). One :ane, in~its entirety, neecs to remain . rao (2) Line-type heat detector, c Y

                                    ~...._._._..-

WOLF CREEX - UNIT 1 3/4 3-61

1 - f INSTRUMENTATION g k( h LOOSE-PART DETECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.9 The Loose-Part Detection System shall be OPtRABLE. APPLICABILITY: MODES 1 and 2. ACITON: _

a. With one or more Loose-Part Detection System channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not appiicable.

J 1 SURVEILLANCE RE0VIREMENTS 4.3.3.9 Each channel of the Loose-Part Detection System shall be demonstrated OPERABLE by performance of. -

a. A CHANNEL CHECK at least once per 24 hours, p,h..J #'

agsf hr -

b. An ANALOG CHANNEL OPERATIONAL TEST 4at least once'per 31 days, and

_ _ -r_

c. A CHANNEL CALIBRATION at least once per 18 months.

WOLF CREEK - UNIT 1 3/4 3-62

INSTRUMENTATION RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend *.ne release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation ' channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.7, why this inoperability was not corrected within the time specified.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i SURVEILLANCE RE0VIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequen-cies shown in Table 4.3-8.

 'a0LF CREEK - UNIT 1                      3/4 3-63

                                                                                                                                        .# n ..                   ,s
                                        . er                                      .:                                  a
                                                                                           .)                                          :.t. q

t. -

                                                                                                                                     . y:e ' ,.
                                                                            *                                           <              .e       i.1 M $

TABLE 3.3-12 .. , a- - , ,

                                                                                                                                                                 .?

RADI0 ACTIVE LIQUID EFFLUENT HONITORIllG INSTRUMENTATION 4 n .'.!

o .

HINIMUM , i?) l!! CilANNELS ,Aj INSTRUMENT OPERABLE s ACTION ' '.!) E 1.- Radioactivity Monitors Providing Alarm and ~ Q Automatic Termination of Release ,

• l'g

a. Liquid Radwaste Discharge Monitor (E-18)

                                                                                                    .            1              31                               .!  .

b. Steam Generator Blowdown Discharge Hoisitor (4tE-52) 1 32

                                                              . Lt.

c. Turbine Building Drain Monitor 6ftE-59) * -

1 32

d. Secondary Liquid Waste System Honitor iE-45) 1 33

2. Flow Rate Heasurement Devices 1:* a. Liquid Radwaste Discharge Line

 .. s.                                >                                                                                                                             -
    ';"                     1) Waste Honitor Tank A Discliarge Line                                             1              34                                -

S .

2) Waste Monitor Tank B Discharge Line 1 34

b. Steam Generator Blowdown Discharge Line 1 34

c. Secondary Liquid Waste System Discharge Line 1 34 c d. C2:!:.au-Tu- -G b i- Se t .

                                                                                                                                        =c                         6 CD                         I
                                                                                                                                                                *f CO,                        -

f?O 4 1

• rm -

                                                                                                          .                              rn g

C3

                                                                                                                                 ;       CD
                     '                                                                                                                   N M

j .

     ;.::-[;                                        -

TABLE 3.3-12 (Continued) 100 & EliM COM ACTION STATEMENTS ACTION 31 - 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 14 days provided that prior to

                       .                   initiating a release.

a. At least two independent samples are analyzed in accordance '

with Specification 4.11.1.1.1, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 32 - With the number of channels .0FERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continuepprovided grab samples are analyzed for p mp. p4 I ist

                                 ..Ai         a--"
                                                             '" Qup to .su cayjDat a lower limit of detection +6 is Td/* 4 il-l .

4s spec,*M

a. At least once per 12 hours when the specific activity of the J secondary coolant is greater than 0.01 microcurie / gram DOSE

    .                                           EQUIVALENT I-131, or

b. At least once per 24 flours when the specific activity of the secondary coolant Is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.

                                                      'ha number of channels OPERABLE less                          wec-tpThe ACTION Minimum        33[)WPh           Ch h west N guire                     e   filient releases via this
                                      .I                                                  .   .,1' % m tLded that, at least
                          $8%

p ^ (/ pathway once per 12 hou maysamolar continue are collecteu for l :" at a lower limit of detection of no more tn 7 ACTION 34 - With the numbai of channels OPERABLE less than required by the - Minimum Channels 0FC?.aBLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. Pump - performance curves generated in place may be used to estimate

                  ,     .   .              flow.

b. WOLF CREEr( - UNIT 1 3/4 3-65

l ACTION 33 - With the number of channels OPERABLE Less than required by the Minimum Channels OPERA 8LE requirement, effluent releases via this pathway may continue for up to 30 days'provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and

b. - At least two technically qualified members of the facility staff independently verify the release rate calculations and

 .                discharge line valving.

Otherwise, suspend release of radioactive affluents via this , c pathway. , '

s O

e b

9 ,

                                              /                                                                                      .                                  .
                                                                                                                                                                      'n..

TABLE 4.3-8 e ' RADIDACTIVE LIQUID EFFLUENT HONITORIllG INSTRUMENTATION SURVEILLANCE REQUIREMENTS - n . o

      ,q                                                                                               .

ANALOG

      ^

CllANNEL

         '                                                                                CilAHilEL      SOURCE           CllANNEL IllslRuMI:llT                                                                                                            OPERATIONAL c      -                                                                             CilECK          Cl!ECK      CALIBRATION i      =                                                                                                                                        TEST y      1.        ,1!adioactivi1.y Honitors Providing g                 Alarm and Autontatic Termination of Release IIG-                                                                               !

a. LiquidRaddasteDischargeMonitordRE-18) D 'P R(2) Q(1)

b. Steam Generator filow'own d Discharge D H . R(2) Q(1)

Honitor (IRE-52) . On* .

                     '-c.   .lurbine Building Drain Monitor
 ,    y                       j!:E-59)                                                        D
   .. 5                    Li~                                                                                H             R(2)         Q(1) 8;>               il. Secontrary liquid Weste System 8;                                                               '

0 A R(2) Q(1) lionitor GRE-46) f

p. ,

2. Flow Rate fleasurement Devices

a. Liquid Radwaste Discharge Line D(3) it. A. R .Q W.A.

b. Steam Generator Blokdown Discharge Line D(3) ti. A. R Jr # #<

c. Secondary Liq' tid Waste System D(3) ll. A.

Discharge Line R &ar,A, 3 c3

                  ; d                                           '                                                                                          c3 C u o '. .. 3 " . _ . L '. .. i ..      ..;
                                                                                           @                  "^   _

D _ n. Ro

                                                                                                                                             ~

M Q m cs

                                                                                                                                   +

l C3

                 ,                                                                                                                                        T

                .n     -

l TABLE 4.3-8 (Continued)

                                                                                                                    ][ h {] ((

TABLE NOTATIONS . as attel';,n (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur (if any of the following conditions exists:

                                                                                                                                                   -          M (h3 ,g , M ,g

a. Instrument indicates measured levels above the Alarm / Trip Setpoint.( or Circuit fai%rg(A<a l

b. or cal ) -

                                                                                                    /                       54f

c. Instrument indicates a downscale failure ( i r _.-7 ), or

d. Instrument controls not set in operate mode (alarm only).

qo or ag..a ..A s.1;A) (2) The initial CHAdEL CALIBRATION shall be performed using one or more of the referencepitandards certified by the National Bureau of Standards (NBS) or using stancards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shell permit calibrating the system over its intended range of energy,aa4 measurement ra@eg For subsecuent CHANNEL CALIBRATION, scur :: thn :::- __r m = .. ;.i. 'nitial c;;;s. nica an;; :e .mm. (3) CHANNEL CHECK shall consist of verifying indication p flow during periods of release. CHANNEL CHECK shall be made at least gnce per 24 hours on [ days en which continuous, periodic, or batch reipa'ses are made. (mi al.htisk mo~6 resps 4 A f*'"I / f l c.h ..s s..n c. / wes %c e.J.le k d edud ( ys, hpid ," 5'U O , vai uid *

                                                                                            ,         er     o.      sam            of N.        7 shua                                        t;A p<ccess w        u b;q       of (ps,M.u:A th.

1, ean is

                                                                                                                      ,    .<     s.pte.

p,.w:ded i pim+gbg.,gv9u.& otb <ds ddu-)-;44 using lI usia3 965 referene. ted ebbi-ed from sylius A oc sbda .

   ". -Y'-               '~*

• y pad, icipafe. ira measurcuni assurance ae%; lie <, sik Mss.

i s

WOLF CREEK - UNIT 1 3/1 3-67 _ , , , - . -s =

Table 4.3-8, Table Notations Justification - ~ The proposed wording currently in the Callaway Technical l Specifications implies that the same geonetry as found in i the monitors will be used for calibrations. It is not - always possible to duplicate geonetry on count roon equipment. ICME's calibration program may also differ fran other plants; therefore, a time frame is not justified.

    - Note (2) has been inodified to resolve these concerns. This modification allows the use of a process stream for calibration. A portion of the process stream can be                           !

analyzed for radioactivity usirg the currently calibrated system. This system is calibrated using NBS reference standards or standards obtained fran suppliers participatirg in measurement asurance activities with the NBS. e g y ,

                       .w--  . --        ----   -    ,_      ,    , _. . - . . ,

INSTRUMENTATION P 00F & B31 COPY RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.11 The radioactive gas'eous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints. set to ensure that the limits of Specifications 3.11.2.1 and 3.11.2.5 are not exceeded. The Alarm / Trip Setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the 00CM. APPLICABILITY: As shown in Table 3.3-13.

     'CTION:

A

a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent

monitoring instrumentation channels OPERABLE take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report, pursuant to Specifica-tion 6.9.1.7, why this inoperability was not corrected within the time specified.

c. -The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.3.3.11 Each radioactive gaseous effluent moritoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequen-cies shown in Table 4.3-9. WOLF CREEK - UNIT 1 3/4 3-68

                         ,                          l.                                           .

i TABLE 3.3-13

                     =c

( RADI0 ACTIVE CASE 005 EFFLUENT HONITORING INSTRUMENTATION O HINIMUM CllANNELS lll INSTRUME_I{T OPERABLE APPLICABILITY s ACTION

1. WASTE GAS Il0 LOUP SYSTEM Explosive Gas

                   @       _    Honitoring System                                              .
                   '^
                   -l                                                                                                                                                 .-

g. a. Ilydrogen Monitor 1/recombiner **

7, 44

b. Oxy 0en Monitor 2/recombiner **

42  !

2. Unit Vent System *

a. Noble Gas Activity Monitor - Providing 1

• 40 Alarm (RE-21) '

sc-Iodine A u .vii.,  ::....t r '": " ' l- Savaf l 4 R b. 1 *

                '~

• 43

              . y            c. Part,-ic,ul$ te ?c . . . . L, ::.;; . tu. '"" 20 b{l**        1

• g 43

d. Flow Rate N 1

• M 46

e. Sampler Flow Rate Monitor 1

• 39

3. Containment Purge System

                                                       ~

l

q. Noble Gas Activity M5st' tor - Providing -

Alarm and Automatic Termination of Release IE-22, itE-33) "E M ."E 22} 1 * - 41 y w

b. LIIn*- Saa'If er 3 "

• 4' O C3

c. Ihrlic=k fe, Sey ler M3

'                                                                                                  3                   y              43                 p J. Fl.a ble                                                       nA.                 ,               qs                  c, i

e, hgl,r Fl.d~%fs H5Ei*f I

                                                                                                                       +               3*l              b r v,
                                                                 .                                                                                      k i

C~3 T r &

I

                                                   ; 'l                                                    ,s

' ~

• d' ',t.;

a' . .o

                                                                                                                                                                                              -i.! .j c
                                                       '                        lABLE 3.3-13 (Continued}                                  ,

Sn - RADI0 ACTIVE GASEOUS EFFLUENT HONITORIllG INSTRUMENTATION O HINIMUM CllANNELS p, INSTRtlMENT OPERAllLE m APPLICABILITY ACTION

4. Radwaste DuildinD Vent System e . ..

Z , q 'a. Noble Gas Activity Honitor - Providing . g Alarm and Automatic Termination of Release ytE-10) 1

• all- 38, 40 t

h. lodine ^.c'.i.i'y 17.. . ' L; .u-; 5yh 4 1 .

43

c. Particulate f.c' hit, L... .. su; l,) .9m[er- 1

• 43

d. Flow Hate "; ,. . -1 N. A . * -

                                                                                                                                               .ag.4g

e. Sampler Flow Rate Honitor 1 *

       ' R.

r 39 - La a'* 23 C3 CD m C'O

• .a s r c1 me:C Ti8

                                                                                                                                    .       g
                    ,                                     .                                                                                            co

• C3

                                                                                                                                                   ,   nJ M

l

       ..                                                            .,                                            z j, l t.s .

t , 7,-

                                          ' ' , . ,, , ' . 4 gi                                                                          ,
                                                                                                                                                                                                       ,4    i ,o l
   @.' .y < ,                        . ;                          ,s
     .my                                   g..

TABLE 3.3-13 (Continued) PRODF & REM COPY C'q.' ' 7 7;.. TABLE, NOTATIONS ,

                                                                                                                                                                                                                                                )

gj q.,,.M. < n ,:

• At all times. .

                                                                                                                                                                                                                       .             ~ ,:

W T/ . 7 ** During' WASTE GAS HOLDUP SYSTEM operation.

                                                                                                                                                                                                                  -                             l

r. m.

Oy;. f. . - >o '

                                                                                                              ~ ACTION STATEMENTS QD%.y...
                                                                                                            '.            .                                    z.-                                         ,
 'M'                      '. r ACTION 38 - With the number of channels OPERABLE less than required by the                                                                                                                      ~
        -W-
                                                                            . Minimum Channels OPERABLE requirement, the contents of the tank (s) yhA?':.                             .,c.,                               -may be released to the environment for up to 14 days provided that                                                                            ,

y,[.r/M t ) /;. , . , pr.ior to 1nitiating the release: -

                                                                                                                                                                                                , u . , ;.y , .

J ,a. ,.At'least two independent samples of the tank's contents are,

 .;.W fyg. hl a.[ ,
                                                                                  .., analyzed, and owd . ;                           .,                                                    ..o-                                       ..

l;L' ,,. ,i'

              .                                                             : b.  At least two technically qualified members of the facility                                                                       ,i
  ;Q"'f
                                                                                       ' staff independently verify the release rate calculations and discharge valve lineup.                         ~
       ,d s.                                    -

a., - 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 releases 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 l A for rcdfoactivity within 24 hours. *

                                   -ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERA 8LE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 42 - 7ttb. e number of channels OPERABLE one less than required b ' 'io Minimum OPERABLE requirement, operation ~ system h may continue provi samples ar and analyzed at least 68 once per 24 hours. With s inoperable, operation may

o. go e continue provid samples are taken lyzed at least b6 # once pe urs during degassing operation and a east once hours during other operations.

ACTION 43 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent 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 than required by the - Minimum Channels OPERABLE requirements, suspended oxygen supply to the recombiner. ~ WOLF CREEK - UNIT 1 3/4 3-71 sd bc ks'al e^ l Ac.Ticu % - Flos ede gws a.ey G Ns k 5(sb use s e o.cJw\ mers a<m~ .

l 4 ACTION 42 - With the Outlet Oxygen Monitor channel inoperable, operation of the system may continue provided grab sampiss are taken and analyzed at least once per 24 hours. With both oxygen channels or both 4db the inlet oxygen and inlet hydrogen monitors inoperable, suspend oxygen supply to the recombiner. Addition of waste gas to the - system may continue provided grab samples are taken and analyzed at least once per 4 hours during degassing operations and at least once per 24 hours during other operations. l l t m

                                                                                                                                            'l                                            .,        .                                                                                                   .
                                                                                                                                                                                                                                                                                        .(  .-

i . c TABLE 4.3-9 ~j - g RADI0 ACTIVE GASEOUS EFFLUENT MONIT0lilNG INSTRUMENTATION SURVEILLANCE REQUIREMENTS r,-, - n ANALOG g CllANNEL H0 DES FOR WHICH y t CilANNEL SOURCE CilANNEL OPERATIONAL SURVEILLANCE

                                                                                                     ,     lilSIRUMEllT                                                     CilECK        CilECK        CAL 10 RATION TEST _          IS REQUIRED                                                 !

1; WASTE GAS Il0LDUP SYSTEM Explosive-H Gas Monitoring System - N

a. Inlet flydrogen Monitor D N.A. Q(4) H **

s

b. Outlet flydrouen Monitor D fl. A. Q(4) H ^^ *

c. Inlet Oxy 0en Honitor D N.A. Q(S) H **

d. Outlet Oxy 0en Monitor 6

D it. A. , Q(p) H **

2. Unit Vent System .

a. floble Gas A'ctiv'ity Honitor- D H. s R(3 ) *- Q(2)

• u Providin(f Alarm (RE-21) -

h S tkr

b. Iodine ^.t!.; , ou,,; w, s':; 23 W N.A. N.A.

• N.A.

c. Particulate Art %

i;:yle<- t; ::.,,,;;u. (nu 2d W N.A.

                                                                                                                                                                                       .                   N.A.              N.A.                         *

d. F l ow Ra te !'_.. . to.- -D- N. A N.A. R(7) -(t-N, A. *

e. Sampler Flow Rate Monitor D. N.A. R

• Q

3. Containment Purge System u -'

x3 c- 3

a. Noble Gas Actikity Honitor -

Providing Alarm and Automatic M Termination of Release g i

gE-22,,RE-33)Pr-21,.RE2)- D P P(3) Q(2)

• p;g' y cp -,

W F "h 1,. L ata. Her W.A. n.a. g. 4. 4-  ::G

                                                                                                             .g. Pu k kle. W <c                          .               W             y A.            u.3,             , ,4 ,      .             , ,                              a
                                                                                                                & fb        A*+e.

N ^- 9.A. PA) Ar.A.

• E3

e. Seger A*8 MMMr D ti A- M A- #

                                                                                                                                                                                                             &                                          4 l

di - .

                                                                                                                                                      ,.4...',
                                    ,                                                                                                                  k.::t TABLE 4.3-9 (Continued)                                                                        .j' 2-                                                                                                                                                       .

o RADI0 ACTIVE GASE0US EFFLUENT 110NIIORING IllSTRUMENTATION SURVEILLANCE REQUIREMENTS O ANALOG lIl CilANNEL . MODES FOR WHICH

• CilAtlNEL SOURCE CllANNEL OPERATIONAL INSTRUMENT SURVEILLANCE

     '                                                              CllECK          CllECK        CALIBRATION      TEST IS REQUIRED E    4.      Itadwaste Guilding Vent System                                 -

g . r a. tioble Gas Activity Mohitor - D, P HP R(3) *

                   .Providing Alarm and Automatic                                                                   Q(1)

TerminationofReleasegtE-10) -

                                           %i

b. Iodine Activity "ep n.ite. (RE IG) .

W N.A.

                                                                                                     !!. A.

• N. A.

lo-SofMc;;itar '

c. Particulate Act!"ity (RE-10} W ll. A. N.A. N.A. *

d. Flow Rate "- . .. c 46 N.A.

• ca
• R(1) .Q-9. A .

 . 2            e. Sampler Flow Rate Monitor                           D
                                                                             ~

N.A. 11 *

                                                                                                                    -Q el.A.

Y .~' m W CD C3

              .                                                                                                                              -u1                  '

W

                         '                                                                                                                   ce g,

i - G C3 i m

                  ,,7                   .c      .                                  .                             -
                 /.   -

.f TABLE 4.3-9 (Continued)

TABLE NOTATIONS f it. ' l

• At all times. -

                         ** During WASTE GAS HOLDUP SYSTEM operation.

(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstra:e tha any of the following conditions exists: gp

a. Instrument indicates measured levels above the Alarm / Trip Setpoint4 or b.

Circuitfailurbfo

c. Instrument indicates a downscale failure (alarm on1y) or

                                                                                                           ~

d. Istrument controls not set"in operate mode (alarm only).

(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that' control room alarm annunciation occ~urs if any of the following conditions exists:

a. ~ Instrument indicates measured levels acove the Alarm Setooint. or J

b. Circuit failure, or

c. Instrument ind'icates a downscale failure, or

d. Instrument entrglsn.gsef cerate mode. "

                                                                                                                           .((f((

(3) The initial CH NEL CALIBR.CION shall be , rmec using one or more of the reference tanoards .rtified une National Sureau of Standards (NBS) or using standards tna ave been obtained from suopliers that participate in measureme ssurance activities with NBS. These standards shall permit calibrat . the system over its intended range of energy acMi measurement range 3 or subsequent CHANNEL CALIBRATION, _.3o

                                                                                                                   . . . ~
                  %. &                                                                                                     \

(4) The CHANNEL CAT.IBRATION shall include the use of standara gas sampled - containing a nc.minal:

a. One volume percent hydrogen, balance nitrogen, and v '

b. Four volume percent hydrogen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall incluce',the use of standard gas samples containing a nominal: .

a. One volume percent oxygen, balance nitrogen, and

b. Four volume percent oxygen, balance nitrogen.

 .                    ;&h            %        .k.~.NE           110tc (Q             '
                                                                                                                                        ~

(7) A dfd 4 7tst. ( 7)

                                                            ~

WOLF CREEK - UNIT 1 3/4 3-74 a e

Insert NBS traceable standard (gas, liquid, or solid) may be used; or a sample of the process stream (gas, liquid, or solid), provided the activity of the stream is determined usirg plant equipnent calibrated usirq NBS reference standards or standards obtained fron suppliers that participate in measurenent assurance activities with NBS. G

                                                     =

1 e e D p _ , _ _

Table 4.3-9, Table Notations Justification -

        'Ihe proposed wording currently in the Chllaway Technical Specifications implies that the same gemetry as found in the monitors will be used for calibrations. It is not always possible to duplicate geonetry on count roon equipnent. 10&E's calibration program may alm differ fron other plants; therefore, a time frame is not justified.

Ibte (3) has been modified to resolve these concerns. e S e

                                         ~ TABLE 4.3-9 (Continued)

_ TABLE NOTATIONS kh5 The CHANNEL CALIBRATION shall include the use of standard gas samples frgi containing a nominal:

.t.

y a. 10 ppm by volume oxygen, balance nitrogen, and

b. 80 ppm by volume oxygen. balance nitrogen.

(7) If flow rate is determined by exhaust fan status and fan performance curves, the following surveillance operations shall be performed at least once per 18 months:

a. The specific vent flows by direct measurement, or i

b. The differenti'al pressure across the exhaust fan and vent flow established by the fan's " flow pP" curve, or

c. The fan motor horsepower measured and vent f?ow established by the fan's " flow-horsepower" curve.

e e I 5 e o I

                                       . _ . _ .             _ _    _ . . _   .o ___ _ _ _ _ _

1 INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION 300F & ElEl COPY LIMITING CONDITION FOR OPERATION l 3.3.4 At least one Turbipe Over M Protection System shall be OPERABLE. 1

                                                                         \

APPLICABILITY: MODES 1, nd . ACTION:

a. With one stop valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoper-able, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in'the affected steam lines or isolate the turbine from the steam supply within the next 6 hours.

b. With the above required Turbine Overspeed Protection System otherwise inoperable,'within 6 hours isolate the turbine from the steam supply.

SURVEILLANCE REOUIREMENTS o . 4.3.4.1 The provisions of Specification 4.0.4 are not applicable. 4.3.4.2 The above required Turbine Overspeed Protection System shall be demonstrated OPERABLE:

a. At least once per 7 days by cycling each of the following valves through at least one complete cycle from the running position:

1) Four high pres:ure turbine stop valves,

0) Icur M gh pr a ra t cbi..e ga.cr. r : M ,

S:. 2.2) Jeer low pressure turbine reheat stop valves, and gu 3 #) -4:s.5eer low pressure turbine reheat intercept valves. w,,t c W. At least once per 31 days by direct observation of the movement of each of the above valves through one complete cycle from the running position; d a. At least once per 18 months by performance of a CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems; and

e. t. At least once per 40 months by disassembling at least cne of each of the above valves and parforming a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or Corrosion.

         +  U +'     a gsle r d k         i .s     .S c t.t    2 .       3    "I"      GD          r. . . a    ; { v.s      I %           s.. '< h-savo       c.uk             Luc c .t t. A            su        glon     in       "n e_        (knA                            u,y     ,, g g c n. .      v . . .. ,      A -.                                                                          I , L , (,,

s% h.  % -k . L - . a .J ,,i , WOLF CREEK - UNIT 1 3/4 3-75 _,p. __ _ - - - . .

i l 1 sasat 4.3.4.2b -At least once per 31 days by cycling each of the four high pressure' Main Turbine Governor Valves through at least one complete cycle from the running position. I e O ( e 4 e e i .

Specification 3 3.4.2b Justification - This revision to the surveillance requirements is based on the vendor recu m a dations. These reuumedations are provided in the attached General Electric Technical Information Intter No. %9, dated 5/22/84. O e 4 4 e*

j. '

{ 4 e 6 e i

v-a GENERAL ELECTRIC '

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         .-       LARGE STEAM TURBINE-GENERATOR DIVISION
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GENF3AL ELECTRIC COMPANY SCHENECTADY, NEW YORK Technical Information Letter No. 969 MAY 22, 1984 Periodle Turbine Steam Valve Test - Nuclear Units n CE recommendations for periodic nuclear turbine steam valve tests as contained in the Turbine Instruction Book call for daily test of the main stop, inter-mediate stop, and intercept valves and weekly test of the control valves. These recommendations are similar to the test frequencies that have been in practice since the late 40's and early 50's on fossil-fueled turbines. The operating experience accumulated on in-service nuclear units during the past 24 years has shown considerably lower ' valve failure rates than those values upon which the recommendations were based. These reduced failure rates are due to many design improvements to the nuclear turbine valves and controls that have been incorporated through Technical Information Letters (TIL's) and Engineering Change Notices (ECN's). In the " Memo Report - Hypothetical Turbine Missiles - Probability of Occur-rence," dated March 14, 1973, the probability of runaway failure and wheel burst of a GE nuclear turbine was given, based on the nuclear experience up to that time. Included in the probability calculations were the recommended valve test intervals; i.e., daily for main stop and intermediate stop & inter-cept valves, weekly for control valves. The Nuclear Wheel Information Letter No. 2, dated November 8, 1982, gave comparative values for the increased over- ' speed probabilities due to increasing test intervals. . I S Based on past in-service experience with nuclear turbine steam valves, turbine steam inlet valve reliability and testing intervals are no longer the major

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contributing factors in determining hypothetical turbine missiles. The over-

   '4
              .-            all probability of a hypothetical missile is therefore increased only a

('] negligible amount by increasing the test interval of the valves. Increasing V test intervals will correspondingly decrease the probability of a system upset during such testing and should therefore increase the nuclear plant avall-ability. Of course any problems detected during any testing should be brought to the attention of your local A&ES ' service engineer. The service engineer may call on the LST-G Dept. if furtler assistance is necessary. l l COPYRIGHT GENERAL ELECTRIC COMPANY, 1984 t l g Thea' laserumauem orga =ana= futulehed is this Technical laformaues IAtter le effered to you by General Electrie la costlaus&Aem of its eagegeg selee and eerstee relationship with your However. stace the operataen of yeer plaat tavelves umany factere met withia our knowledge, and names opermales of the plaat le withtm yearsostret and ulumate ,w DI 43 A roepeaethutry may fee its- eestamulag successful operause roete with you, General Electrie specifleauy disclaine any reopensabuiry for liebtitry based om etalane for damages of type.La diseet. meta al er sportal that enay be allegod se hase booe taeurred ae a reemit er applytag thee tafsemasaom regardloos ef whetherilisennamod that General ., . ,. Elaetwee is eartetty Mahaa to beoech of contract, breach of warrasty. mogugent, or le la ether respecte roepa==dm for Fy atteged tajury or ' 5 "*"'**' by your , , ,,, organaamates ne a result of applytag this taformaanoe.

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I 969 ' b Effective with this Technical Information Letter, the recommended valve test intervals for nuclear turbines are: Main Stop Valves Weekly Intermediate Stop Valves Weekly Intercept Valves Weekly control Valves Monthly Recommended test intervals for other control components remain unchanged. Utilities should revise their Instruction Books to the new test intervals based on this TIL. Revised Instruction Book Articles will not be sent from General Electric Co. Because of the higher temperature and resulting increased oxidation build-up on the stems and bushings of fossil-fueled turbines, the valve test interval recommendations remain unchanged; i.e., daily test of the main stop, combined intercept and rebeat stop valves, and weekly test of the control valves. If your Technical Specification or other documentation upon which your URC operating license is based contains any obligation or commitment to test on a specific schedule, it is suggested that you take appropriate steps to modify that document if you wish to change your test intervals to these new recommendations. .... 9 page 2 [ .I

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                                                                  ? ROD & RSHI CO?Y 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND CCOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION
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3. 4.1.1 All reactor coolant loops shall be in operation.

APPLICABILITY: MODES 1 and 2.*- ACTION: With less than the above required reactor coolant locos in operation, be in at least HOT STANDBY within Y hourg. b J SURVEILLANCE REOUIREMENTS

4. 4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at le'ast once per 12 hours.

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     "See Special Test Exception /3.10.4 WOLF CREEK - UNIT 1                   3/4 4-1

l REACTOR COOLANT SYSTEM

                                                                  %[h          hhh HOT STANOBY                                                                          )

LIMITING CONDITION FOR OPERATION

                         -ovw 3.4.1.2 At least two- f the reactor coolant loops listed below shall be
                                ~

OPERABLE and at least of these reactor coolant loops-shall be in operation:"

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,

b. Reactor Coolant Loop 8 and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant. pump, and

d. Reactor Coolant Lo'op D and its associated steam generator and reactor coolant pump.

APPLICABILITY: MODE 3.**- ACTION:

a. With less than the above required reactor coolant loops OPERABLE,

             , restore the required loops to OPERABLE statu:; within 72 hours or be in HOT SHUTDOWN within the next 12 hours.

b.  % on a Amcu * .

c Jr. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate correc'.ive action to return the l required reactor coolant loop to operation. SURVEILLANCE RE0VIREMENTS 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 b 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 l Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature. 6+  %. qu:.1 % c.c.fh ... 9eg.swa 3 w4. l L WOLF CREEK - UNIT 1 3/4 4-2 i

l 1 ACTION:

b. With only one reactor coolant 1 cop in operation, restore at least two loops to operation within 72 hours or be in HOT SHUTDOWN within the next 12 hours.

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l REACTOR COOLANT SYSTEM ]{ { )) {} HOT SHUTDOWN LIMITING CONDITION FOR OPERATION I

3. 4.1. 3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:* .

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,**

b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,**

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,**

d. Reactor Coolant Loop D and its associated steam generator anc reactor coolant pump,**

e. RHR Loop A, and -

f. RHR Loop B.

APPLICABILITY: MODE 4. ACTION:

a. With less than the above required reactor coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return tne required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours,

b. With no reactor coolant or RHR loop in operation, suspend all operations involving a. reduction in boron concentration of tne Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

             *All reactor coolant pumps and RHR pumps may be deenergized for up to 1 hour t-tow, provided:     (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
           **A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50 F above each of the Reactor Coolant System cold leg temperatures.

WOLF CREEK - UNIT 1 3/4 4-3

PR007 & REVal COPY REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pump, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

                                 ,yA-4.4.1.3.2 The required s am generator (s) shall be determined OPERABLE by verifying secondary side ater level to be greater than or equal to 10% at least once per 12 hours.

4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours. l l c ! WOLF CREEK - UNIT 1 3/4 4-4

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l l REACTOR COOLANT SYSTEM ^ l { {I[ hk COLD SHUTDOWN - LOOPS FILLED

     ' LIMITING CONDITION FOR OPERATION                                                   I l

3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:

a. One additional RHR loop shall-be OPERABLE #, or

b. The secondary side water level of at least two steam generators shall be greater than 10% of the wide range.

APPLICABILITY: MODE 5 with reactor coolant loops filled ##. ACTION:

a. With one of the RHR loops inoperable and with less than tne recuired steam. generator level, immediately initiate corrective action to return the inoperable-RHR loop to OPERABLE status or restore the required steam generator level as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System anc immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REOUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.

       #0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
     ##A reactor coolant pumo shall not be started unless the secondary water tamperature of each steam generator is less than 50 F above each of the R

j eactor Coolant System cold leg temperatures. Q*The RHR pump may be deenergized for up to I hour.provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature. WOLF CREEK - UNIT 1 3/4 a-5

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REACTOR COOLANT SYSTEM DM & win COM _  ! 1 COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE # and at least one RHR loop shall be in operation.* APPLICABILITY: MODE 5 with Reactor Coolant loops not filled. ACTION:

a. With less than the above required RHR loops OPERABLE, immediately initiate torrective action to return the required RHR loops to OPERABLE status as soon as possible.

            .b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REOUIREMENTS 4.4.1.4.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. l

     #0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
     *The RHR pump may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.

WOLF CREEK - UNIT 1 3/4 4-6

REACTOR COOLANT SYSTEM f;t & R win COPY 3/4.4.2 SAFETY VALVES SHUTOOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig i IE.* APPLICABILITY: MODES 4 and 5. ACTION: With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode. 8 SURVEILLANCE REQUIREMENTS l 4.4.2.1 No additiona'l requirements other than those required by ' l Specification 4.0.5. 1 s i

                  *The lift setting pressure shall correspond to ambient conditions of the valve at nominal cperating temperature and pressure.

WOLF CREEK - UNIT 1 3/4 4-7 O

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l REACTOR COOLANT SYSTEM ((hf h h a _ l l OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 'All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig i 1%.* APPLICABILITY: MODES 1, 2, and 3. ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or'be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specification 4.0.5.

   *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

WOLF CREEK - UNIT 1 3/4 4-8

r-. _ REACTOR CCOLANT SYSTEM - h [hh h I 3/4.4.3 PRESSURIZER-LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERACLE with at least two groups of backup pressurizer heaters each having a capacity of at least 150 kW and a water level of less than or equal-to 92% (1657 cubic feet). APPLICABILITY: . MODES 1, 2, and 3. ACTION:

a. With one group of backup pressurizer heaters inoperable, restore at

                                          . least two groups of backup heaters to OPERABLE status within 72 hours aor be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.

b. With the pressurizer otherwise inoperable,~be in at least HOT STANDBY with the Reactor trip breakers open within 6 hourr and in HOT SHUTDOWN within the following 6 hours.

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             ,J SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water level shall be determined to be within its
                              , limit at,least once per 12 hours.
                            ~4.'4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit
                            ' current at'lcast once per 92 days.

4 k, J WOLF CREEK - UNIT 1 3/4 4-9 4

a s l 1- U REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES fhh l LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: , j ,g

a. WithoneormorePORV(s) inoperable [within1houreitherrestore the PORV(s) to OPERABLE status or close the associated block valve (s) E n.v.- gu . .' rw thc M cck =1 4 ; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTCOWN

                ,,_  s,W i.t g g t% f [flowing 30 hours.

c. s.. , u .4 Sct,cu c. -

d e. with one or more block valve (s) inoperable, within i nour e - LD restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s) or close the PORV and remove power from its associated solenoid valve;-s tar,i;c, be i.. - isasi ntn siAnUbi wir.nirrTne-nen. huur:r end-Ju40LO SHUTCCWN w$E.Mos cpp:.k , 6e rc H%.idd A c. unuum W4.) . ed W 'yl) Ac1 nog e. u c. .4.. 3

e. . z". The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.4.4.1 In addition to the requiremerits ,0f Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by performance of a CHANNEL CALIBRATION. 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 unless the block valve is closed with power removed in order to meet the requirements of ACTION e in Specification 3.4.4.

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• WOLF CREEK - UNIT.1 3/4 a-10 L

ACTION:

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b. WithonePORVinoperableduetocausesotherthanexcessiveseai' leakage,within1houreitherrestorethePORVtoOPERABLEstatus,o]r 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 and in COLD SHUTDOWN within the following 30 hours.

c. With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.

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REACTOR COOLANT SYSTEM _ l 3/4.4.5 STEAM GENERATORS 1-h i ___ LIMITING CONDITION FOR OPERATION ~ 3.'4. 5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: e With one or more steam generators inoperable, restore the inoperable nerator(s) to.0PERABLE status prior to increasing T,yg above 200 F. SURVEILLANCE REOUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the reouirements of Specification 4.0.5. 4.4.5.1 Steam Generator Samole Selection and Insoection - Each steam generator shall be determined OPERABLE during shutcown by selecting and inspecting at least the minimum numoer of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Samole Selection and Insoection - The steam generator tuce minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specifIed in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteric of Specification 4.4.5.4. The

   ' tubes selected for'each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

WOLF CREEK - UNIT 1 3/4 4-11

REACTOR COOLANT SYSTEM P Pit Ul & E H COPY l SURVEILLANCE RE0VIREMENTS (Continued)

1) All nonplugged. tubes that previously had detectable wall penetrations (greater than 20%),

2) Tubes in those areas where experience has indicated potential priblems, and

3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall

                   ,be selected and subjected to a tube inspection.

c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and

2) The inspections include those portions of the tubes where imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories: Catecory Insoection Results

     .           C-1                  Less than 5% of the total tubes inspected are
            -                         degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubec, but not more than 1". of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. C-3 More than 20% of the total tubes inspected'are degraded tubes or trore than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. l l WOLF CREEK - UNIT 1 3/4 4-12 l

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I REACTOR COOLANT SYSTEM (( g {j{. Q l SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.3 Inscection Frecuencies - The above required inservice inspections of steam. generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed a.fter 6 Effective Full Power Months but within 24 calendar nanths of initial criticality.

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the pre-service inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that pre-viously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency o

shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

 ,              1)    Reactor-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of
                      <pecification 3.4.6.2, or

2) A seismic occurrence greater than the Operating Basis Earthquake.

or

3) A C;cf E' loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A C;ndi;im :',' main steam line or feedwater li.ne break.

A WOLF CREEK - UNIT 1 3/4 4-13

l l REACTOR COOLANT SYSTEM . g gr SURVEILLANCE REOUIREMENTS (Continued) 4.4.5.4- Acceotance Criteria

a. As used in this specification:

1) Imoerfection means an exception to the dimensions, finish or contotr of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;

2) Deoradation means a servica-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube;

3) Oeoraded Tube means a tube containing imoerfections greater than or equal to 20% of tne nominal wall thickness caused by degradation;

4)  % Deoradation means the percentage of the tube wall tnickness affecteo or removed by degradation;

5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective;

6) Pluccina Limit means the imperfection depth at or beyond which '

the tuce snall be removed from service and is equal to jig 8748% of the nominal tube wall thickness; , . . .

7) Unserviceable describes the condition of a tube if it leaks or contains a cefect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in(4.4.5.3c., above; mya&w

8) Tube Insoection means an inspection of the steam generator tube from tne point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and WOLF CREEK - UNIT 1 3/4 J-la

1 1 REACTOR COOLANT SYSTEM FnDDF & RE\lEW COPY SURVEILLANCE RE0VIREMENTS (Continued)

9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to

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initial POWER OPERATION using the equipment and techniques ex-pected to be used during subsequent inservice inspections,

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes ccataining through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3,and r:quirc pecmpt not;ricetic ei Um CcE 4 pon shall be reported3 pursuant to Specification 6.9.2; prior to resumption of 2

plant opbration. The writtenMowuo-ot This report shall provide a descriktion of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence. 5 .., a g at t- w C -+ " \ q

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                                                                                  ..s i    WOLF CREEK - UNIT 1                     3/4 4-15 i

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Y r w n Nngn ir ei h T O o nU h pr T 3i t it t s c ge c a r s. e n se l inp er po h ie d o e e s s T bs n w n n s n ota O O ai o psh er T . ll e d e ai t mut o ne t c hs c u ois g , s r cv e e r u nepna er cm n p s nt s ion o e ae i t i ce E B F luo d r t t s s . mr ns b e eh pt e l s O N o e e ets u e o l nw T N r l l n h i r rchit t l a iel o O h A O cf s i oi h S I T ge c d s icl o R T C nh e nn n vf O E o it t m ei oc o r l el T P af os e it s a o os o r c t A S w s rh e e inh s N T r p R t et v e s t s s E I a slud ns n r n N E C S ne na or uh t i e i f o E G I N r e t t s o c e itc 1 V R O oh t aem h r iver v ht e p 4 M E I T t a t e v e e s o gs nn 4 A S A rf eieehst t inba i i r E n) o un t T N T E S I O gnNet et r c t s I d e L B G N a o ep i n r dequ A F N E mlp r. a msn f i e O I L ee n ei e ed ht e t c s T R R eb st h nb o t B b pu E U A a gr i s s B D T nimt otd en inc s idn n M D osr ed i e r e ud t U E k nf d o a N T tooiud t a l oo i dn s nh s t C d r ee af eo r r u M m U E P s it n ne t i ct e o o e ib e puc t n af M S N i o imggn yb s r r e e I l ns t nh e mial t N I c e a mn la i in eT I b er h t i g M p s ysf r o t s s oe s. it n a r oe nh m n a n i mfopran et c r t eo U e e ow t s it c nr e ne t u ao c o eaer o ep i r q e v ib rl e ws l r p e s t cms gs e nf o s n n e t n r o o i euroml pn s t a p a gd l r i ed n t a it c t n ine ae m r t mu a o h ir t o r e e e h es a eh s oh p u et n es t i n s q c i is e t s s e e r t c e n e v go e r n h d G i I h eNmmts eo t p s r e b hi f ra s s c u in reae r od m t oce t lo a e i v S et o e c h n e c t s r e & h eh s en na ep h sn c o a c S e ivr f n d T (w aos ll t Ti E s e o i n t t s o .

2. 3.

e o r c 1 i e Pi N F S co5 m h,x ' C h -

mA p5

TABLE 4.4-2 5r-- STEAM GENERATOR TUBE INSPECTION A

                      'N                       IST SAMPLE INSPECTION                          2ND SAMPLE INSPECTION             3RD SAMPLE INSPECTION h             Sample Size        Result         Action Required        Result       Action Required      Result      Action Required
                      ==

U A minimum of ' C-1 None N. A. N. A. N. A. N. A.

                      -            S Tubes per 4                                   S. G.

C-2 Plug defective tubes C-1 None N. A. N. A. and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Plug defective tubes 4S tubes in this S. G. Perform action for C-3 C-3 result of first sample Perform action for a C-3 C-3 result of first N. A. N. A. 2 sample w C-3 Inspect all tubes in All other this S. G., plug de- S. G.s are None N. A. N. A. fective tubes and C-1 inspect 2S tubes in , Some S. G.s Perform action for N. A. each other S. G. N. A. C-2 but no C-2 result of second additional sampic Notification to NRC S. G. are pursuant to 950.72 (b)(2) of 10 CFR C-3 Additional Inspect all tubes in O Part 50 S. G. is C-3 each S. G. and plug defective tubes. M Notification to NRC N. A. N. A. pursuant to $50.72 N (b)(2) of 10 CF R Part50 h g 4 3 N, Where N is the pumber of steam generators in the unit, and n is the number of steam generators inspected C"3 t7 -, durinu an inspection 1

                                                                                    ~

REACTOR COOLANT SYSTEM g{ hh 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4'6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Atmosphere Particulate Ra,dioactivity Monitoring System,

b. The Containment Normal Sump Level Measurement System, and

c. Either the Containment Air Cooler Condensate Flow Rate or the Containment Atmosphere Gaseous Radioactivity Monitoring System.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only two of the above required Leakage Detection Systems OPERABLE, opera-tion may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for gaseous and particulate radioactivity or a gamma isotopic analysis of the containment atmosphere is performed using the Post Accident Sampling System at least once per 24 hours when the required Gaseous or Particulate Radioactivity Monitoring System is inoperable; otherwise, be in at least HOT STANDBf within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. SURVEILLANCE RE0VIREMENTS

 '4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Monitoring System performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table

4. 3-3,-

b. Containment Normal Sump Level Measurement Systed performance of CHANNEL CALIBRATION at least once per 18 month

c. Containment Air Cooler Condensate Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.

WOLF CREEK - UNIT 1 3/4 4-18

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE

                                                                                   '(( hh LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. 1 gpm total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 8 gpm per RC pump CONTROLLED LEAKAGE at a Reactor Coolant System oressure of 2235 20 psig, and

f. 1 gpm leakage at a Reactor Coolant System pressure of 2235 : 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.+

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANOBY within 6 hours and in COLD SHUTDOWN within the following 30 hours,

b. With any Reactor Coolant System leakage greater than any one of the acove limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, reduce the leakage rate to within limits within 4 hours, or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 12 hours with

                .      an RCS pressure of less than 600 psig.
         -y-       5,,   .. H.1.l     gaLi3L.(       n, d.

WOLF CREEK - UNIT 1 3/4 4-19

                    " Test pressures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one-half power.

                                                                                            .40 k

i,

                ,e-.      ---...v.- -. . . - , , . .n---.-,e----.,w,-.- -
                                                                          ,,w,

i _ REACTOR COOLANT SYSTEM NOT & EJE COM - SURVEILLANCE RE0VIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within _ each of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulate radioactivity monitor at least once per 12 hours;

b. Monitoring the containment normal sump inventory and discharge at least once per 12 hours;

c. MeasurementoftheCONTROLLEDLEAKAGE)4ev' af the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 t 20 psig at least once per 31 days,e4th t' 2 2 ; ; i ng- . .. . . . . , m _ ; .m The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of'a Reactor Coolant System water inventory balance at least once per 72 hours; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours.

4.4.6.2.2 Each Reactor Coolant System Pressure' Isolation Valve scecified in Taole 3.4-1 shall be demonstrated CPERABLE by verifying leaxage to oe within its limit:

a. At least once oer 18 months, asp

b. Prior to entering MODE 2 whenever tne 9 44*t has been in COLD SHUTDOWN for 72 hours or more and if leakage testing nas not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and

d. Within 24 hour _s. following valid ~ahtuation due to autcmatiE or manual

                              ~

action oF~ flow througn the valve, eveyt he 4tve> ecVI7c 4/6 c73 ervsm Ms. ,, The provisions- of"Scecification~ 4: 0J4 are not ace'1FdibTe~for entry into M00E 3 or 4. WOLF CREEK - UNIT 1 3/4 J-20

l l TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION _ BBV8948 A, B, C, D M/N/Accum.+fn~jection BBV8949 A, B, C, D SI/RHR Hot Leg Injection BBV001, 022, 040, 059 Bit Bwup@c Cold Leg 'InyA.'

     $  BBPV8702 A, B                   RHR Normal Suction 8841 A, B                    RHR Hot Leg 3 Recirc C      Iso E 8701 A, B                     RHR Normal Suction EMV001, 002, 003, 004 SI Hot Leg Inj Ctmt I !!Me '

EM8815 it Inj. Ctat Isolation EM&E EPV010, 020, 030, 040 04 % S asse Inj Ctmt Iso \ EPV8818, A, B, C, D Gd S 6 Asse Inj Ctmt Iso EPV8956 A, B, C, D Accum Inj Isolation 'a0LF. CREEK - UNIT 1 3/4 A-21

REACTOR COOLANT SYSTEM -- 3/4.4.7 CHEMISTRY g{ { ))i COM LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2. APPLICABILITY: At all times. ACTION: MODES 1, 2, 3, and 4:

a. With any one or more chemistry parameter in excass of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours pend,

b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.

At All Other Times: I With the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4. SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant Syste:n chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3. l WOLF CREEK - UNIT 1 3/4 4-22

TABLE 3.4-2 . REACTOR COOLANT SYSTEM hhh"gf h fg i i b CHEMISTRY LIMITS STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen

• 1 0.10 ppm i 1.00 ppm Chloride 1 0.15 ppm. 1 1.50 ppm Fluoride 1 0.15 ppm i 1.50 ppm
• Limit not applicable with T avg less than or equal to 250*F.

l - l WOLF CREEK - UNIT 1 3/4 4-23

TABLE 4.4-3

                                                                          &       E'sCOM REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS SAMPLE AND PARAMETER                                 ANALYSIS FREQUENCY Dissolved Oxygen

• At least once per 72 hours Chloride At least once per 72 hours Fluoride At least once per 72 hours "Not required with T,yg less than or equal to 250*F WOLF CREEK - UNIT 1 3/4 4-24

 ~_

REACTOR COOLANT SYSTEM - 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and

b. Less than or equal to 100/E microcuries per gram of gross radioactivity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2, and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line).shown on Figure 3.4-1, operation may continue for up to 48 hours provided that the cumula-tive operating time under these circumstances does not exceed 800 hours in any consecutive 12-month period. The provisions of Specification 3.0.4 are not applicable;

b. With the total cumulative operating time at a reactor coolant specific activity greater than 1 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive 6-map % repare and submit a Special Report to the Commission githin 3_0 dayf, pursuant to Specification 6.9.2C, ndicating the numcer of hours above this limit. The provisions of Specification. 3.0.4 are not applicable;

c. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANOBY with T avg less than 500*F within 6 hours; and

d. With the specific activity of the reactor coolant greater than 100/E microceries per gram of gross radioactivity, be in at least HOT STANDBY with T 3yg less than 500*F within 6 hours W s

 *With T,yg greater than or equal to 500*F.

WOLF CREEK - UNIT 1 3/4 4-25

   /

REACTOR COOLANT SYSTEM PTU u:> d il1L P# COPY LIMITING CONDITION FOR OPERATION ACTION (Continued) MODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram of gross radioactivity, perform the sampling and ane. lysis requirements of Item 4.a) of Table _.1,_4-4_untiL.the-specific-activity of the.. reactor-coolant ~f s~ restored to within its limits.4

                           ~
                                                                    -         _ d' . '

iM: '"T:O' :t:t: rt, are and submit _,a_SpeciaLReport~to~ thNommission

     ' pursuant 70 Spectfication-or9.-2 within~30 days with a copy to the Director, Nuclear Reactor Regulation, Attention: Chief, Core' Performance Branch, and Chief, Accident Evaluation Branch, U.S. Nuclear Regulatory Commission, Washing-ton, D.C. , 20555. This report shall contain the results of the specific activity analyses together with the following information:

1. Reactor power history starting _4.8_ hours prior to the first samole in which the limit was exceeded -

d 3, *O 0" f . 2. Results ofA the last i otopic analysis for radioied[ne performed prior to exceeding the limit, while limit was exceeded and/one analysis after tne radiciodine :t' ';-was~reducedtolessthan!thelimit, including for each isotonic analysis, the date and time of samoling and tne raciciocine concentrations,

3. Clean up flow history starting 48 hours prior to the first samole in which the limit was exceeded, -

4. History of degassing cperations, if any, starting 48 hours prior to tne first sample in which the limit was exceeded, and fe u W

5. The time duration when the specific activity of the prs;; coolant exceeded 1 microcurie per gram 00SE EQUIVALENT I-131.

SURVEILLANCE RE0VIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4 ' WOLF C.'iEEK - UNIT 1 3/4 4-26

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   < 100 m                                         .            . . . .          ..                               . . . . . . . .       . _ . . . . . .

m _ .. . . . . . . . . . . . . . . . . . . . m .. . .. . . _ . . . _. . . . _ . . . y I . H ._ . . . . . _ . . . . . . . . . _ . _ _ . . . . . . . . . ._ z ACCEPTAB LE.. _ . . . . _ . . .

    ".]                                           OPERATION                                                       - -- --          - --                                     - - -

f -- - - -

    <      50                                                                                                                        ...               .                    _.              _ _ _ .                              . .

3 3 . ... . . . . . . c - . . _. ._. . . _ _ . . w . . . . . . . . . .. . . . ua .. . m . _ . . . g . . 0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4.1 00SE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY '.IMIT VERSUS PERCENT OF RATED THERMAL PCWER WITH THE REACTOR CCOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131 WOLF CREEK - UNIT 1 3/4 4-27 l

,                                                         IABLE 4.4-4 O

5 REACIOR C00l ANI SPECIFIC ACHVITY SAMPLE g AND ANALYSIS PROGRAM .7 x 1YPE OF HEASUREHENT SAMPLE AND ANALYSIS H0 DES IN WillCll SAMPLE AllD ANALYSIS FREQUENCY AND ANALYSIS REQUIRED k H

1. Gross Radioactivity At least once per 72 hours 1, 2, 3, 4 Determination

2. Isotopic Analysis for DOSE EQUIVA- per 14 days 1 LEHf I-131 Concentration .

1

3. Radiochemical for E Determinatiort Dac.e per 6 months

• 1

4. Isotopic Analysis for Iodine a) Once per 4 hours, 1#, 2#, 3#, 4#, 5#

Including I-131, I-133, and I-135 whenever the specific activity exceeds 1 pCi/ gram DOSE y N cs EQUIVALEN I-131 C.3 [ or 100/E j[ilCi/ gram of gross " A radioactivity, and W b) One sample between 2 and 6 hours following 1,2,3 a TilERMAL POWER change exceeding IST,of the y 4 RAlED TilERMAL POWER o within a 1-hour period. C3

   #Until the specific activity of the Reactor Coolant System is restored within its limits
   ^ Sample to be taken af ter a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.

l l 1 3/4.4.9 PRESSURE / TEMPERATURE LIMITS ROM & EU COM REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100*F in any 1-hour period,

b. A maximum cooldown of 100*F in any 1-hour period, and

c. A maximum temperature change of less than or equal to 10*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILIT,Y_: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STAND 8Y within the next 6 hours and reduce the RCS Tavg and pressure to less than 200 F and 500 psig, respectively, within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2)aw 3.4-3; w i M -4 . WOLF CREEK - UNIT 1 3/4 4-29

MATERIAL PROPERTY BASIS' COPPER CONTENT GON998WWWIEWWEdrY, ASSUMED TO BE 0.10 WT% / pf/) RTuoy INITIAL I40T RTuay AFTER 16 EFPY :1/4T.1!?F 3/4T, 87'F ' CURVE APPLICABLE FOR HEATUP RATES UP TO 60'F/HR AND 100'F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 10'F AND 60 PSIG - FOR POSSIBLE INSTRUMENT ERRORS. 3000 ' l I I i

                                                                                                                                                     /

l l i Il l I l lll 11 l LEAK TEST ll f l/l f llll 1 ' , , ','"4T , , y. j gij 7ll l L ll II'llil IM / I /l l '/I l l lil! l l l l Ill  ! ll ll! ll 7 ll fl /l / l i/IJ'S$#'S'l$' ' 0 I 2000 I IIIll II II Ill II / IIII I I i/MTf ' HEATUP CURVE Il _ ll! I ll lllllil11 l l 11 l I /11/ fillilill lilllilli

    $            lli       i    ll     lillll!Il                lilill/ l /                   /l/!!!!Il                  lllllllll 3    $           ll ll l                     11lllll             l i f f l '/            l/   li l / I l l ! I l          llllllll 5           II il I                     lllll11             ll lll /li l'/ It!Ill                                    llllllll 0           ll l 'l                     ll11111 il l '/l / / if I l lCRmCAUTY UMIT
    !           Il                l         111llll ll l/ l fl/ ' M N$$P UR E '"" T
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    $           ill          ll        lllllllll                ll'/ L/l/Il /lllllllll                                   llllll           l g

lill! 111 !Illilill il/ l /iA l / 11llll111 lllllllll 3gga lllll lll 1111lllll f/l/ill/l !Ililllli ll11lllll llll ll l HEATUPllI /{ l [I l l [I IIIIIIIll lllII l l Ill l l C,uRyE q .j jflly  ;;jjj;l; ;j;;; l l l  !! !/ l l CRITICALITY LIMIT BASED ON L l l l l l l/ Q l l 1NSERVICE HYDROSTATIC TEST W 4, TEMPERATURE I255'FI FOR THE -- _ 60'F/ H R I' I ll SERVICE PERIOD UP TO 16 EFPY 1"?/,HR[ , , ,,;llllll ll l l l lllllllll llll ll l l'! li lllllllll lli l l l' lllllllll Illi lilll ! l ll111llll lilll l l. ll111llll Illi 11ll l I llllillll 11 i l lllll11 l Ill 0 IIIIII II IiIIIIIII II I II IIIIIII  !!I O 100 200 300 400 500 IN01CATED AVERAGE TEMPERATURE (DEG. F) FIGURE 3.4-2 REACTOR CCOLMIT SYSTE!! HEATUP LIMITATIONS APPLICASLE UP TO 16 EFPY WOLF CREEK - UIIT 1 3/4 4-30

- PROT & REM COPY h MATERIAL PROPERTY BASIS ) E 5 COPPER CONTENT : GamegippsusewetV ASSUMED TO BE 0.10 WT% E-RTNOT INITIAL .: 40'F  :

, y..- --

g RT,,7AFTER 18 EFPY: 1/4T,'110'F g 3/4T,87'F  : 3000 E. E_ E CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100'F/HR FOR THE SERVICE' E

PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 10*F and 60 PSIG FOR POSSIBLE :

- INSTRUMENT ERRORS. 2

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100 ? 0 0 100 200 300 400 500 INDICATED AVERAGE TEMPERATURE (DEG. F) FIGURE 3.4-3 REACTOR C00LAtlT SYSTE!1 C00LDOWil LIMITATI0tlS APPL: CABLE UP TO 16 EFPY '

         "0LF CREEK . UilIT 1                                                         3/4 4 31

TABLE 4.4-5 PROD & RWcW . COM REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE CAPSULE VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY) U 58.5* 4.00 1st Refueling Y 241* 3.69 g6

       -  V            61*            3.69       9 X            238.5*         4.00       15 W            121.5*         4.00       Standby Z            301.5*         4.00       Standby 8

O l O r WOLF CREEK - UNIT 1 3/4 4-32

                                                                    ..\                            .

[ . p Table 4.4-5 ' Justification - N; - , Revised to be consistent with WCAP-10015 " Kansas Gas & Electric' Cdnpany, Wolf Creek Generating Station, thit No.1, Reactor. Vessel Radiation Surveillance Program." I s i

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REACTOR COOLANT SYSTEM PP1007 & F iN M M PRES $URIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 100*F in any 1-hour period,

b. A maximum cooldown of 200*F in any 1-hour period, and

c. A maximum spray water temperature differential of 583*F.

APPLICABILITY: At all times. ACTION: With the pressurizer temperatur'e limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. ' SURVEILLANCE REOUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. WOLF CREEK - UNIT 1 3/4 4-33

m

                  ~

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION-FOR" OPERATION 3.'4.9.3 / At least one of the following Overpressure Protection Systems shall beOPERABLE( g , y g ,y j ,,, gj y3, g, ,,, _

                                                                                                                                                ,,   g

e. g. Twocowe%,39.,,y ise g,.3 r_ operated relief valves (PORVs) with Setpoints which do not exceed the limit establ.ished in Figure 3.4-4,.or

c. b'. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2 square. inches.

              ,              APPLICABILITY: MODE 1 when the temperature of any RCS cold leg is less than or equal tc 368*F, MODES 4 and 5, ano: MODE 6 wi                               he' reactor vessel head on.
                                                                                            ,.a w y*/'                                          , .,.o ACTION:
                                =

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                                                                        ' g4p M '#

s t, ic & - ' '- 7

                          ,,"-           a.       'Wi trbne PORV/g.neper:Me, either cestore 'the--inoperanle-+0RV to
                                              ,._ OPERABLE status within 7 days.or depressurize and vent the RCS
          <                                        th-ough at least a 2 souare inch vent           w, within the next 8 hours.
                                                                           . a S , s. ,s..t'

( b. With both PORVs/ inoperable dcpressur'ize and' vent the RCS through at least el2 square inch vent within 8 hours.,s...r u a,, ~. i c%

                              ,                                                                .. m

c. In th6 avent e4hes the PORVs $r the RCS vent (s) are used to mitigate

         ,,-                            ,          an RCS pressure transient, a Special Report shall be prepared and submitted to tn'e Commission pursuant to Specification 6.9.2 within 30 days.        The report shall describe the circumstances initiating the
            ,                                      transient, the effect of the PORV por-RCS vent (s) on the transient, and any corrective ~ action necessarf to prevent recurrence.
                           , _          d.      . The provisions of Specification 3.0.y are not applicable.
       ,-                  ,           /                                                             \
                                                                                                          .g      n,e    ,.A.       i ' d  I " * >

i p is WOLF CREEK - LHIT 1 3/4 4-34 L .; _

Specification 3.4.9.3a Justification - The existing tolerances were chosen based on the tolerances for the pressurizer code safety valves. They should have been established to be consistent with the requirenents of A9E B & PV (bde as lorg as the maximun pressure allowed is less than 530 psig (the lowest pressure setting of the ECRV's) . This specification has therefore been revised. 4 9 e e 1 a , v- - , - -- --- , ,- , .,9-- g r-- .r,. .+ - ,cm-- -, , . . . ..

Specification 3.4.9.3 (Action a) l l Justification - It was the intent of this Specification to give credit for the availability of the RHR suction relief valves for cold overpressure protection (CDP). In the Standard

      ' Technical Specifications, when relying on PORV's for CDP, if
    . one IORV_ becanes inoperable, you have 7 days to restore it or open a 2 square inch vent. 'Ihis same action should be applicable if you are relying on the MIR suction relief valves for CDP and you lose one, regardless of the status of the FORV's. 'Ihe action statement has therefore been revised.

t 4 4 e. e 9 e 4

  ,        +                      .-            e9-      -S.. - - - -- -   , - - -. x     - , . , . - - - . .

REACTOR COOLANT SYSTEM fQQ]} g Q{[j] {}{y SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of a ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE;

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and

c. Verifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.

4.4.9.3.I'TheRCSvent(s)shallbeverifiedtobeopenatleastonceper 12 hours

• when the vent (s) is being used for overpressure protection.

 "Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open positio'n, then verify these valves open at least once per 31 days.

WOLF CREEK - UNIT 1 3/4 2-35

                                ,.9         g      -

O sasse t - 4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a. For RHR suction relief valve 87088:

1) By verifying at least once per 31 days that RHR RCS Suction Isolation Valve (RRSIV) B701B is open with power to the valve operator removed, and ,

, 2). By verifying at least once per 12 hours that RRSIV 8702B is open.

b. For RHR suction relief valve 8708A:

1) By verifying at least once per 31 days that RRSIV 8702A is open with powe,r to the valve operator removed, and

2) By verifying at least once per 12 hours that RRSIV 8701A is open. .

   \ s. c. Testing pursuant to Specification 4.0.5.

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100 200 300 400 MEASURED RTD TEMPERATURE,'F FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM ucos F C Rta. CALLL'AY - UNIT 1 3/4 4-36

c. ..

O

                                                                                     -l REACTOR COOLANT SYSTEM r

3/4.4.10 STRUCTURAL INTEGRITY '"n0"UI i & RM!W COM _ { LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class l', 2 and 3' components shall be maintained in accordance with Specification 4.4.10. APPLICABILITY: All MODES. ACTION:

a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT, considerations.

b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.

c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate

, the affected component (s) from service. l

d. The provisions of Specification 3.0.4 are not apolicable.

SURVEILLANCE REOUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatcry Guide 1.14, Revision 1, August 1975. WOLF CREEK - UNIT 1 3/4 4-37

i

          /

PROOF & REWEW COPY 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS - LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

a. TheisolationvalveopenhW few Wa*" [ ' -

b. A contained borated water volume of between 6122 and 6594 gallons,.

c. A baron concentration of between 1900 and 2100 ppm, and

d. A nitrogen cover pressure of between 602 and 648 psig.

APPLICABILITY: MODES 1, 2, and 3". ACT!ON: e

a. With one accumulator inoperable, except as a result of a closec isolation valve, restore the inoperable accumulator to OPERABLE status within 1-nour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTCOWN within the following 6 hours.

b. With one accumulator inocerable due to the isolation valve Oeing closed, either immediatelv open the isolation valve or be in at least HOT S NDBYwithin}I-housandinHOTSHUTDCWNwithintne following hours. &

SURVEILLANCE REOUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 12 hours by:

1) Verifying, by the absence of alarms, the contained boratea water volume and nitrogen cover pressure in the tanks, and

2) Verifying tnat eacn accumulator isolation valve is open.

              " Pressurizer pressure acove 1000 psig.
             'a0LF CREEK - UNIT 1                        3/4 5-1 b.

EMERGENCY CORE COOLING SYSTEMS PROD! & RIVliW COPY SURVEILLANCEREOUIREMENT5(Continued) 1 l l 1

b. At least once per 31 days and within 6 hours after each solution s s volume increase of greater than or equal to 70 gallons by > .)

verifying the boron concentration of the accumulator solution, and ! lete l

c. At least once per 31 days when the RCS pressure is above 3G00 psig by verifying that the circuit breaker supplying power.%o_the' isolation valve operator is open, y

             ,' ~       At least once per 18 months by verifying that each accum            dr^
           /                  tion valve opens automatically under each of + fo'llowing         '
         ,              condi tto                                    _ .

i 1)

                                   \__

• When an actull'or. ed RCS pressure signal exceeds the-P-11 (Pressuriz essu ck of Safety Injection) Setpoint, and . ....

                ~

pon receipt of a Safety Injection test signa . 4 4. 5.1. 2 Each accumulator ~ water level and pressure channel shal1 be demonstrated OPERABLE at least once per 18 months by the performance of a CHANNEL CALIBRATION. WOLF CREEK - UNIT 1 3/4 5-2

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T

                                                                         )R00F & Ra'iEV1 COPY l avo

• 35C F LIMITING CONDITION FOR OPERATICN 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE Safety Injection pump,

c. One OPERABLE RHR heat exchanger,

d. One OPERA 8LE RHR pump, and

e. An OPEFABLE flow path' capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation a phase of operation.

APPLICABILITY: MODES 1, 2, and 3.* ACTION:

a. With one ECCS suosystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hcurs and in HOT SHUTDOWN within the following 6 hours.

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and sub"mitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

             +       w     iflu:k(         akab.( -nc n F
                   " ~ . .     , , _ , ,

WOLF CREEK - UNIT 1 3/4 5-3

ions of Specifications 3.0.4 and 4.0.4 are not applicable for entry

        *The into MOD 3 for the centrifugal charging pump and the Safety Injection pumps deci        inoperable pursuant to Specification 4 5 3 2 provided the centrifugal charging pump and the Safety Injection pumps are restored to OPERABLE status
        -within 4 hour op prior to the temperature of one or more of the RCS cold legs exceeding 37 *F l

t D g

EMERGENCY CORE COOLING SYSTEMS PRODF & EEW COPY SURVEILLANCE REOUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position BN-HV-8813 Safety Injection to Open RWST Isolation Viv EM-HV-8802A(B) SI Pump Discharge Closed Hot Leg Iso V1vs EM-HV-8835 safety Injection Open Cold Leg Iso Valve EJ-HV-8840 RHR/SI Hot Leg Closed Recirc Iso Valve EJ-HV-8809A RHR to Accum Inj Open loops 1 & 2 Iso V1v EJ-HV-88098 RHR to Accum Inj Open Loops 3 & 4 Iso Viv

b. At least once per 31 days by:

1) Verifying Vat the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and

2) Verifying that each valve (manual, power-operated, or automatic) in the. flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containnient sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1) For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

d. At least once per 18 months by:

1) Verifying automatic isolation and interlock action of the RHR System from the Reactor Coolant System by ensuring that:

 ,                   a)    With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the interlocks prevent the valves from being opened, and b)   With a simulated or actual Reactor Coolant System pressure signal less than or equal to 750 psig the interlocks will cause the valves to automatically close.

i i WOLF CREEK - UNIT 1 3/4 5-4

EMERGENCY CORE COOLING SYSTEMS PR00F & R32 COPY SURVEILLANCE REOUIREMENTS (Continued) _ 2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. At least once per 18 months, during shutdown, by:

1) Verifying that each automai.ic valve'in the flow path actuates to its correct position on a Safety Injection test signal andfy en an RH4 Automatic Switchovery RWSTyjow-Low test signal, and kcc.nc,ae.J A L G.A. ~ 4 s g . V <

2) Verifying that each of the following pumps start automatically % Iv 6.

upon receipt of a Safety Injection actLction test signal: a) Centrifugal charging pump, b) Safety Injection pump, and o c) RHR pump.

f. By verifying that each of the following pumps develops the reouired differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1) Centrifugal charging pump > 2400 psid,

2) Safety Injection pump 1 1445 psid, ana

3) RHR pump 1 165 psid.

g. By verifying the correct position of eacn mechanical position stop for the following ECCS throttle valves:

1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and

2) At least once per 18 months.

HPSI System CVCS System Valve t4umoers Valve riumoers EMV095 EMV109 BGV-198 EMV096 EMV110 BGV-199 EMV097 EMV089 BGV-200 EMV098 EMV090 BGV-201 EMV107 EMV091 BGV-202 EMV108 EMV092 l l t WOLF CREEK - UNIT 1 3/4 5-5

                                                                                                                                      \

J 1 EMERGENCY CORE COOLING SYSTEMS  ! PRODF & REVIEW COPY ' SURVEILLANCE REQUIREMENTS (Continued) h. By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1) For centrifugal charging pump lines, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 346 gpm, and b) The total pump flow rate is less than or equal to 550 gpm.

2) For Safety Injection pump lines, with a single pumo running:

a) The sum of'the injection line flow rates, excluding the highest flow rate, is greater than or eoual to 462 gpm, and b) ._ The . total pump flow rate is-less-than-or equal to 650 gpm.

3) rec R.;d rump ri nm , + M - 9 g'a mm~ runni n g, the n;'f M A l

fnjectieu i;oc Tivo . itsr i 5- gevater -th .n er ^qu2' M.348 ccm. l }, Sw s: N M Y L iN :.

                                ~

9 WOLF CREEK - UNIT 1 3/4 5-5 -

i. .By performing a flow test, durin~g shutdown, following completion of .

modifications'to the RHR subsystems that alter the subsystem flow characteristics and verifying that for RHR pump lines, with a single pump running:

1) The sum of the injection line flow rates is greater than or equal to 3800 gpm, and -

2) The total pump flow rate is less than or equal to 5500 gpm. ,

t

     \
      /

EMERGENCY CORE COOLING SYSTEMS f 3/4.5.3 ECCS SUBSYSTEMS - Tavog350F

                                                                                   .gl g'[

LIMITING CONDITION FOR OPERATI0tl 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump, and

d. An OPERABLE flow path capable of taking suction from the refuding OW we hr sterage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation. ,

APPLICABILITY: MODE 4. $ ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperaoility of either the centrifugal charging pump or the flow path from the Rws refreHng ueter sterage t=k, restore at least one ECCS suosystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours.

b. With no ECCS subsystem OPERABLE because of the inopersoility of either the RHR heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or. maintain the Reactor Coolant System avg less than 350*F by use of alternate heat removal methods.

T

c. In the event the ECCS is actuated and infects water into the Reactor Coolant System, a Soarial ReDort shall ha neonarad and iqttraitted to the Commission pursuanttoSpecification6.9.2Y Ithin90 describing the circum day so [ftheactuationandthetotalaccumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

S WOLF CREEK - UNIT 1 3/4 5-7

l l EMERGENCY CORE COOLING SYSTEMS gf}1 { {.{l1,I @f 3 SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2. 4.5.3.2 All centrifugal charging pumps and Safety Injection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable

• by verifying that-the-motorcircuit. breakers.are-secured-in-the open 'p6sition -' '---' ^~~'

f );. 21 m.,;. toi.tW 4 W dk e*ftriaj AtCbE 4 (rea 14o06 3 b ik ( +, m p,r.,4v of ca . v more cf N M nId I 'p ..e 55 ,

    '-s a u ,. 5;o) be le, 3 Ls* cod       t    les.s+ ce          p<  34 dqs   #<<u     W.
             --%,~                                                                     '

i "An inoperable pumo may be energized for testing or for filling accumulators provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position. h0LF CREEX - UNIT 1 3/4 5-8

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 ECCS SUBSYSTEMS - T 3R00F & 'Em COPY avn < 200 F LIMITING CONDITION FOR OPERATION 3.5.4 All Safety Injection pumps shall be inoperable. APPLICABILITY: MODE 5 and MODE 6 with the reactor vessel head on. ACTION:

                                ~

With a Safety Injection pu$p OPERhBLE, restore all Safety Injection pumps to an inoperable statusul% 4 h=es. SURVEILLANCE REOUIREMENTS 4.5 4 All Safety Injection pumps shall be cemonstrated inoperaole* by a verifying that the motor circuit breakers are secured in the open position at least once per 31 days.

   *An inoperable pumo may be energized for testing par Spe iiicesivo 4. 0.1 or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

WOLF CREEK - UNIT 1 3/4 5-9

I l l I EMERGENCY CORE COOLING SYSTEMS 3/4K5.5 BORON INJECTION SYSTEM BORON INJECTION TANK LIMITING CO ITION FOR OPERATION

                                                                           /

3.5.5 The boron jection tank shall be OPERABLE with:

a. A minimum c tained borated water volume of 9 gallons, and

b. A boron concen ation of between 2000 and 00 ppm.

APPLICABILITY: MODES 1, 2, a 3. ACTION: With the boron injection tank inoper le restore the tank to OPERABLE status within 1 hour or be in HOT STANDBY and orated to a SHUTDOWN MARGIN equivalent to 17. ak/k at 200*F within the next 6 o rs; restore the tank to OPERABLE status within the next 7 days or be n H0 SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.5.5 The boron inj tion tank shall be demonstrated OPE. LE by:

a. Verifyi g the contained borated water volume at lea t once per 7 days, and

b. Ve fying the boron concentration of the water in the tan at least .

ce per 7 days.

                                                                                  \
                                                                                    'N i

l l I l I WOLF CREEK - UNIT 1 3/4 5-10 l-l

EMERGENCY CORE COOLING SYSTEMS (, fi b I 3/4.5.6 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.6 The refueling water storage tank (RWST) shall'be OPERABLE with:

a. A minimum contained borated water volume of 394,000 gallons,

b. A baron concentration of between 2000 and 2100 ppm of boron, y(Jim

c. A minimum we4er- temperature of 37*F, and s whe~

d. A maximum e temperature of 100*F.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: . With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY.within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS ' 4.5.6 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the contained borated water volume in the tank, and

2) Verifying the boron concentration of the water.

b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is either less than 37 F or greater than 100 F.

WOLF CREEK - UNIT 1 3/4 5-11

l PP,05 & HEW COPY 3/4.6 CONTAINMENT SYSTEMS I 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

3. 6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3, ana 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore' CONTAINMENT INTEGRITY within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the follo. wing 30 hours. 3 SURVEILLANCE REOUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations

• not caoable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of 3pecification 4p(mes;T.'

6 3;.

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and .

c. After each closing of each penetration subject to Type-B testing, except the containment air locks, if opened following a Type A or 8 test, by leak rate testing the seal with gas at a pressure not less than Pa , 48 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rata is less than 0.60 L '

a

     *Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured in the closed position.       These penetrations shall be verified closed during each COLD SHUTOOWN excect that suca verification need 'not be performed more often than once per 92 days.                                               -

l WOLF CREEK - UNIT 1 3/4 6-1 e

f CONTAINMENT SYSTEMS MODF & win UM CONTAINMENT LEAKAGE LIMITING CGNDITION FOR OPERATION - 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of: i

1) Less than or equal to La, 0.20% by weight of the containment air per 24 hours at P,, 48 psig, or

2) Less than cr equal to Lt ' ~

U '"O O' contair .cnt air-per-24-hours- at a reduced pressure of P , 24 psig.

b. A combined leakage rate of less than.10.60a L for all penetrati ns and valves subject to Type B and C tests, when pressurized to Pg4$p;$.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With either the measured overall integrated containment leakage rate exceeding 0.75 L, or 0.75 L t, as applicable, or the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L,, restore the overall integrated leakage rate to less than 0.75 L, or less than 0.75 Lt , as applicable, and the comoined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 aL prior to increasing the Reactor Coolant System temperature aoove 200 F. SURVEILLANCE REOUIREMENTS - 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-fled in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:

a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 t 10 month intervals during shutdown at a pressure not less than either P,, 48 psig or at P ' 24 psig, during ea h 10 year service t t 4

eriod. The th rd 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 1

a .-ma -,. S.A_ e= de . K.-4+S-4*A541 a ua 4 s. . , -a *--+6*-* +da -_.ah. a m.4,a 4 A L f k e. t

                                      . Specification                     3.6.1-'2.a 2)  .

t

                                      - Justification -

1 This specification has been revised to be consistent 2 2 with 10GR50, Apperrlix,J. 1 I e

 )

i i p . , u I f-d 9 k - e 3-i a 4-b e wer ewm ~ ,e .cenv-wvm w -wwwwn.?.nm..~m,-+~.woww-m-e-w-~~~e.-+,m v o ew- n - r m e me,,r + m e -me_ .m e,n . . _ _.- -w ww em e-n

CCNTAINMENT SYSTEMS 1RODF & R911T COPY -- SURVEILLANCE REOUIREMENTS (Continued) b. If any periodic Type A test fails to meet either 0.75 L, or 0.75 L t, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet either 0.75 L, or 0.75 Lt, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either 0.75 L, or 0.75 L tat which time the above test schedule may be resumed;

c. The accuracy of each Type A test shall be verified by a supplemental test which:

1) Confirms the accuracy of the Type-A test by vcrifying that the dMferencesbetween supplemental ad Typ: a test data-is-*Mbin 0.

                         ~

25 L,,; er 0. 254-i us It, Le3 m.no h sm c1 A T p 4 a~l

                                .. i, ,,;r ..a 61. , t . ,
                                                            ., e,y ic c .- t e., um,,      m z,, c. c c; 9 ,

2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and

                                      . ,,s    rh d Mk            *

3) Requirestthe.\q m tity of gas [ injected into the containment or bled from the containment during the supplemental test to-be equ i val ent_to-at-leas t-25%%f~ th e- total-mea s u red--lea kag e-at- P p 48 pri gLt, o"d c 75 r .;t ' i2' U N "' "" "'

d. Type 8 and C tests shall be conducted with gas at a pressure not less than P 48 psig, at intervals no greater than 24 months except for testsin,volving:

1) Air locks and

2) Purge supply and exhaust isolation valves with resilient material sealsond rsut

e. Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3;

f. Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.3 or 4.6.1.7.4, as applicable; and

g. The provisions of Specification 4.0.2 are not applicable.

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

Insert

3) Valves, which are in systems that are designed to contain water subsequent to a leakage design basis bss-

                        ~ of-coolant accident shall be pressurized with that fluid to a pressure not less than 1.10 P. , 52.8 psig.

w a .i e J ?- o 4 e O t f

                      .     -_._,.f__*..,-  , - . - -_.--,--S---_. ..,--.-.,m   --,. ,.. _ _ _ . _ . -., _ _ _ _    .- -. _ . _ . .      . ..,.u.,-

Specification 4.6.1.2.d3) Justification - 10CFR50, Appendix J, Section III.C.3 states leakage fran containment isolation valves that, are sealed with fluid fran a seal system may be excitded Wen detennining the conbined leakage rate providal that a) such valves have been demon-strated to have fluid . leakage rates that do not exceed those specified in the technical specifications or associated bases and b) the installed isolation valve sealwater systen fluid inventcry _is sufficient to assure the sealing function for at least 30 days at a pressure of 1.10P. . The essential service water penetrations, P-28, 29, 71 and 73, qualify for this exclusion based on the fact that the water volune contained.in the pipirg inside containnent provides a passive inter seal on the applicable valves. The water inventory inside containnent is sufficient to assure the sealirg function for at least 30 days. This specification has therefore been revised. r l l

s  ; i tf . , l s

                                                                       \x 9

i. )

    ~ ' ~s                                                                                   %
                       ' CONTAINMENT SYSTEMS                                                       F CONTAINMENT AIR LOCKS                                         s 1
                                                                                \_       \

LIMITING CONDITION FOR OPERATION ' 3.6.1.3 Each containment air lock shall be OPERABLE with: ,

a. Both doors closed except when the air lock is being used for normal transit entry and exits through the' containment, then at least one air lock door shall be closed, and U

b. An overall air lock leakage rate 'of. less than or equal to 0.05 L"

( at P,, 48 psig. m APPLICABILii"Y: iMODES 1, 2, 3, and 4.

• ACTION:

a. Wiln one containment air lock door inspenble:

1. Maintain at least the CPERABLE air lock d'oor closed and either s 3 a s\ restore the inoperable afb lock door to OPERABLE status within 24 hours or lock the4DPERABLE air lock dcor closed,

                                                                                   ?x                                -

N 2. Operation may then continue until perfomance of the next l

                          \

required overall air lock leakage test provided that the OPERABLE

                                            >g     Air lock door is verified to be locked closed at least once per
                                                  -31 days,       m
                                                                         ~

3. Otherwise, be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and

          \                           .

4. The p'ovisions r of Specification 3.0.4 are not appli:able.

b. With the containment air lock inoperable, except as the result of an

              ,                            inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STAND 3Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

N - r 's

               !             \
                                 \              t WOLF C3EEK - UNIT 1 '                             3/4 6-4

_p

CONTAINMENT SYSTEMS PROOF & EEll COPY SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. Within 72 hours following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours, by g presserising-the-vol=0 between the door-sea +s4o-at-least-10-pr4g

                   ,          fo     t-4een 30 seconds-and-ver4fy4ng-the-leakage-does not-exceed
                              ---_-_g,
               ,     b.       By conducting overall air lock leakage tests at not less than P,,

f 48 psig, and verifying the overall air lock leakage rate is within l its limit:

1) At least once per 6 months,# and

2) Prior to establishing CONTAINMENT INTEGRITY when maintenance -

has been performed on the air lock that could affect the air lock sealing capability.* I i c. At least once per 6 months by verifying that only one door in each

        !                     air lock can be opened at a time.

s

          \  \
                 \ Y <  .        *
                                          'k h k    'kho          'g._I
                                                                  ,         \ t s. y              .,,

e . .u . .w d

                                                                                            ,3            -4614 ,g   ( C l *) L,,

t. 3. u i .. .:.s &- .,w.n... ..,,, 4 ,3 ,3 ,, % ._, , , (

q ,. g rd lead 2f 5<< . c c b c th Tr ' u h* '"c  % h u <.s -t y g ca c. c.v s ks.W g7. c,u. m d ,- i c p . .) , The provisions of Specification 4.0.2 are not applicable.

      *This represents an exemption to Appendix J of 10 CFR Part 50.

WOLF CREEK - UNIT 1 3/4 6-5 f 9 _ _ _ . . _

                                 .:                                                                                                                     e I

r , ,

                                      ^

m.

                                                                                                                    ;.                   y C0!4TAINMENT SYSTEMS /

e PROD: l & PEEW COPY

j. INTERNAL PRESSUFli

                                                                                                                          ~

,./ ~ .< LIMITING CONDITION FOR OPERATION- ^ e-  ;< -

w , 3. 6.1. 4 Primary containment, internal pressur2 shall be maintained between

                                 /                 +E and -2 psig., /
                                                  + .s                 -d. n                .

/, . APPLICASILITY: MODES 1, 2, 3,' and 4. ,' ' -

                                                                                                 ,                                                        s

. i u

ACTION: . - -

l. ,.y ~

2l/ g With4he containment internal pressure outside of the limits above, restore

                     !                             the' internal pressure to within the limits within 1 hour or be in at least HOT
                                               'STAhtBY vithin the next 6 hours and in col.0 SHUTDOWN within the following 30 hcurs.

<3, 1 - l

                                     -                                                                               e
                                     -.I j                                                                                                       ,

O ,, i[, c ' . ,' l. '

                                                                                                                                                                             ,            ~

a Mu , 4 -

    .Is,-

ll 4  ?

                                                 ,4                   %.

v f .. , y gDRVEILLANCE SE@IREMENTS '/ x, , , i' ,

                             '/
              ,,                                  4. 6. f.'4                  The pririary containment internal pressure shall De deter:nined to be

l. . - within the limits at least once per 12 hours., .

                                             !           s g          ~ . . .

y '

1. r . , , ,

                     -+

t 1 i l , [  ?

                                          )                                                                                        ?

( I L I WOLF CREEK - UNIT 1 3/4 6-6

CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION

3. 6.' 1. 5 Primary containment average air temperature shall not exceed 120'F.

APPLICABILITY: MODES 1, 2, 3, and 4. 1 ACTION: With the containment average air temperature greater than 120*F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS

               <4. 6.1. 5 The primary containment average air temperature shall be the
              . arithmetical average of the temperatures at the following locations and shall E
              - be determined at least once pei 24 hourc:
                                                                                                                                   ~

Location .

a. Containment Cooler Inlet located near NNE wall (El 2068'-8");

b. Containment Cooler Inlet located near West wall (El 2068'-8");

i

c. Containment Cooler Inlet located near NNW wall (El 2068'-8"); and.

d. Containr. ant Cooler Inlet located near East wall (El 2068'-8").

I L i i f CALLAWAY - UNIT 1 3/4 6-7 L -

b CONTAINMENT SYSTEMS

                                   \p /f C'  7 </'                       -

CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION c 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6. APPLICABILITY: MODES 1, 2, 3, and e ACTION:

a. With the structural integrity not conforming to the requirements of Specification 4.6.1.6.1, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

b. With the structural integrity of the containment not conforming at a level consistent with the acceptance criteria of Specification 4.6.1.6.2, restore structural integrity or complete an engineering evaluation that assures structural integrity prior to increasing reactor coolant temperature above 200 F.

J - SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Vessel Tendens. End Anchoraces and Adjacent Concreta Surfaces. The containment vessel tencons structural integrity snall be comon-strated at the end of 1, 3, and 5 years following the initial containment vessel. structural integrity test and at a 5 year intervals thereafter. The tendons' structural integrity shall be demonstrated by:

a. Determining that a random but representative sample of at least 11 tendons (4 inverted U and 7 hoop) each have an observed lift off force within predicted limits for each. For each subsequent inspection one tendor, from each group may be kept unchanged to develop a history and to correlate the observed data. If the unobserved lift off -

force of any one tendon in the original sample ~ population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of this tendon should be checked for their lif t-of.' 'arces. If toth or these adjacent tendons are found to be within c. heir predicted limits, all three tendons should be restored to the required level of integrity. This single deficiency may be considered unique and acceptable. Unless there is abnormal degradation of the containment vessel during the first three inspec-tions, the sample population for subsequent inspections shall include at least 6 tendons (3 inverted U and 3 hoop). If more than one tendon has an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit, or withie one tendon below 90% of the predicted lower limit, it snall be considered as evicence of possible abnormal degradation for the purposes of Specification 4.6.1.6.lg. WOLF CREEK - UNIT 1 3/4 6-8

v i f \# CONTAINMENT SYSTEMS g c< y  %[r} [] hhh ytn SURVEILLANCE REOUIREMENTS (Continued)

b. Performing tendon detensioning, irusections, and material tests on a previ~ously stressed tendon from each group (inverted U and hoop). A randomly selected tendon from each group shall be completely deten-sioned in order to identify broken or damaged wir'es and determining that over the entire length of the removed wired that:

1) The tendon wires are free of unacceptable (pitting of 1/64 inch or deeper and minimum of 1/32 inch in diameter) corrosion, cracks, and damage. The presence of unacceptable corrosion, .______S cracks, or other damage shall be considered evidence of peeemese-abnormal degradation of the containment structure for'tha.c-.- -

purooses of Specification 4.6.1.6.lg.;

2) There are no. changes in the presence or physical appearance of -

the sheathing fi.ller grease Abnormal changes in the presence or physical appearance-of the sheathing filler grease shall be considered evicencecof r-Te abnormal degradation of the containment structure for the purposes of Specification 4.6.1.6.lg; and ' Mm ,' w=- - opsh '

3) A minimum tensile strength of 240jdOO psi (guaranteed ultimate

                   -strength of the tendon material)/for at least three wire er             -

etefi%r samples (one from each end and one at mid-length) cut from each removed wire. Failure of any one of the wire er - sWemt samples to meet the minimum ^ tensile strength test shall mbe -considered as evidence of.amaanm6e abnormal degradation of the containment vessel structure for the purposes of Specifi-cation 4.6.1.6.lg. "

c. Performing tendon retensioning of those tendons detensioned for inspection to the1r observed lift-off force with a tolerance limit of +6%. During retensioning of th.se tendons, the changes in load and elongation should be measured simultaneously at a minimum of three approximately equally spaced levels of force between zero and the seating force. If the elongation corresponding to a specific load differs by more than 5% from that recorded during installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in 'nchorages;

d. Assuring the observed lift off stresses adjuste account for elastic losses exceed the average minimum design ,atue given below:

Inverted U 139 ksi Hoop: Cylinder 147 ksi Dome 134 ksi WOLF CREEK - UNIT 1 3/4 6-9

m l CONTAINMENT SYSTEMS RDE & m3 CM SURVEILLANCE.REOUIREMENTS (Continued)

                                                           %o

e. Verifying the OPERABILITY of the sheathing filler grease by assuring:

1) If the installed quantity of grease exceeds that withdrawn by 5% or more, an investigation shall be conducted to assure that excessive leakage has not occurred in the tendon duct system,

2) Minimum grease coverage exists for the different parts of the anchorage system, and

3) The chemical properties of the filler material are within th' e tolerance limits as specified by the manufacturer.

Failure to satisfy 1pecification 4.6.1.6.le. 2) or 3) above for OPERABILITY of~the sheathing filler grease shall be considered as evidence of p:r S'^ abnormal degradation of the containment structure for the purposes of Specification 4.6.1.6.lg. J

             .       f. Determining througn inspection that no apparent degradation has occurred in the visual appearance of the end anchorage or the concrete surfaces adjacent to the end anchorages. If apparent degracation has occurred in the visual appearance of the end anchorage or the
                                                        ~

concrete surfaces adjacent to'the end anchorages, it shall be con-sidered as evidence of pr- " '9 abnormal degradation of the contain-ment structure for the purposes of Specification 4.6.1.6.lg.; and

g. If evidence of fenwisinde abnormal degradation of the containment structure is detected during the performance and/or evaluation of the.

                ,         results of the above tests, the following actions shall be completed:

1) , Reported to the NRC within 10 days,

2) Perform an engineering evaluation demonstrating the continued ,

ability of the containment structure to perform its design function. If continued containment integrity cannot be assured by engineering analysis within 00 days, ACTION a. required by Specification 3.6.1.6 shall be taken, and

3) Provide a determination of the cause of the apparent degradation and performance of any corrective actions necessary to ensure continued containment integrity.

4.6.1.6.2 Containment Vessel Surfaces. The structuiJ integrity of the exposed accessiDie interior anc exterior surfaces of the containment vessel, including the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of tnese surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in aopearance or other abnormal degradation. WOLF CREEK - UNIT 1 3/4 6-10 4 0

• CONTAINMENT SYSTEMS C0lnAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION

3. 6.1. 6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY: MODES 1, 2,'3, and 4. ACTION:

a. With more than one tendon with an observed lift-off force between

              '                  the predicted lower limit and 90% of the predicted loyer limit or with one tendon below 90% of the predicted lower limit, restore the teidon(s) to the required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

b. With any abnormal degradation of.the structural integrity other than

                     ,           ACTION a. at a level below the acceptance criteria of Specifica-tion 4.6.1.6, restore the containment vessel to the required level of integrity within 72 hours and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 15 days in accordance with Specification 6.9.2 or be in at
            ~                    least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours.

i SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Vessel Tendons. The containment vessel tendons' struc-tural integrity shall be demonstrated at the end of 1, 3, and 5 years following

  '               the initial containment vessel structural i.ntegrity test and at 5 year intervals thereafter.        The tendons' structural integrity shall be demonstrated by:

a. Determining that a random but representative sample of at least 11

 ]                               tendons (4 inverted U and 7 hoop) each have an observed lift-off force within predicted limits for each. For each subsequent inspec-tion one tendon from.each group may be kept unchanged to develop a i

history and to correlate the observed data. If the observed lift-off force of any one tendon in the original sample population lies between the predicted lower limit and 90% of the predicted lower limit, two ten' dons, one on each side of this tendon should be checked for their lift-off forces. If both of these adjacent tendons are found to be within their predicted limits, all three tendons should r be restored to the required level of integrity. This single deficiency 2

 -                               may be considered unique and acceptable.      Unless there is abnormal degradation of the containment vessel during the first three inspec-
 ;                               tions, the sample population for' subsequent inspections shall include
 =                               at least 6 tendons (3 inverted U and 3 hoop);
 -                                                                                                       l i

CALLAWAY - UNIT 1 3/4 6-8

                                                             ~

CONTAiNMENTSYSTEMS . SURVEILLANCE REQUIREMENTS (Continued)

b. Performing tendon detensioning, inspections, and material tests on a

_previously stressed tendon from each group (inverted U and hoop).

                   -A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and deter-mining that over the entire length of the removed wire that:

1) sThe tendon wires are free of corrosion, cracks, and damage,

2) There are no changes in the presence or physical appearance of the sheathing filler grease, and

3) A minimum tensile strength of 240,000 psi (guaranteed ultimate strength of the tendon material) exists for at least three wire samples (one from each end and one at mid-length) cut from each removed wire. Failure cf any one of the wire samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment vessel structure.

I

c. Performing tendon retensioning of those tendons detensioned for inspection to.their observed lift-off force with a tolerance limit of +6%. During retensioning of these tendons, the changes in load

 ,          ,      and elongation should be measured simultaneously at a minimum of three approximately equally. spaced levels of force between zero and the seating force.                   If the elongation corresponding to a specific .

load differs by more than 5% from that recorded during installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages;

d. Assuring the observed lift-off stresses adjusted to account for elastic losses exceed the average minimum design value given below:

Inverted U 139 ksi Hoop: Cylinder 147 ksi Dome 134 ksi

e. Verifying the OPERABILITY of the sheathing filler grease by assuring:

1

1) No voids in excess of 5% of the net duct volume, .

2)- Minimum grease coverage exists for the different parts of the anchorage' system, and

3) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer.

CALLAWAY  ; UNIT 1 .' 3/4 6-9 i

     ~ CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.6.2    End Anchorages and Adjacent Concrete Surfaces. The structural integrity of the end anchorages of all tendons inspected pursuant to Specifi-cation 4.6.1.G.,1 and the adjacent concrete surfaces shall be demonstrated by determining through inspection that no apparent changes have occurred in the visual appearance of the end anchorage or the concrete crack patterns adjacent to. the end anchorages. Inspections of the concrete shall be performed during the Type A containment leakage rate tests (reference Specification 4.6.1.2) while the containment vessel is at its maximum test pressure.

4.6.1.6.3 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate,'shall be d.etermined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. O I e l l e l CALLAWAY - UNIT 1 3/4 6-10 1 I

i CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM PR00F & RElH COPY LIMITING CONDITION FOR OPERATION - 3.6.1.7 Each containment purge supply and. exhaust isolation valves shall be

     . OPERABt.E and:

5

a. Each 36-inch containment shutdoWrrpurge supply and exhaust isolation s

valve shall be closed and g gig g' s and

b. The 18-inch containment mini-purge sup'p'ly and exhaust isolation valve (s) may be open for up to.507 hours during a calendar year.

APPLICABILITY: MODES 1, 2, 3, and 4. p, ..mp, ACTION:

                                                       /

4 fM J' ~h ,

                                                                                         /

a. Witha36-inchcontainmenMurgesupply'and/orexna t isolation valve open or not 52_ ?:'::: ...close and/or e:M '= tnat valve or isolate the penidation(s) within 4 hours,~ otherwise be in

                                             ~

a at least HOT STANDBY within the next 6 hours and in COLD SHUTCOWN

 .                 within the following 30 hours.                       s b.
                                                                   /

Withthe18-inchcontainmentminipurgesufplyand/orexhaust isolation valve (s) open for more than 66& hours curing a calendar year, close the open 18-inch valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANOSY within the next 5 hours, and in COLD SHUTDOWN within the following 30 hours. I 'f.2 1

c. With a containment purge L;;oly and/or e naust isolation valve (s) having a measured leak e rate in exces of the limits of Specifications 4.6.1. and/or 4.6.1.F.4, restore the inoperable valve (s) to OPERABLE status within 24 hours, otherwise be in at least

                . HOT STANDBY within the next 6 hours, and in COLD SHUTCOWN within the fol b ing 30 hours.

WOLF CREEK - UNIT 1 3/4 6-11

          .I CONTAINMENT SYSTEMS P300F 3 EISl COPY      l t                        l SURVEILLANCE REOUIREMENTS                                                                                   l 4.6.1.7.1, he 36-inch containment _ shutdown purge supply and exhaust isolation
                  ?

valv'e(s)*shall_be verified Oc;kd ck;;C cnd closed at least once per 31 days.

                  ! 4.G.I. 7 2-           See a bW- hka-             iL BD,ak    7.u )

j i.4.6.1.7.),The~cumulativetimethatall'18-inchcontainmentminipurgesupply

                     ~and/ee. exhaust isolation valves have been open during a calendar year shall be determined at least once per 7 days.

g= -

s. 6.1. 7. ' least once per 6 months on a STAGGERED T m E T ;, ..m snodard and outboard va v ~ f er)Lmatecert Taals in sealed closed 36-inch containment shutdown _e 50pp-ty st penetrations shall be demonstrated

                                                                                                   '~

OPERABLE at the measured leakage ess than-0.05.L '*"*" p a to P,. 4.6.1.7.4 At least once per 3 months each IS-inch containment mini-purge supply and exhaust isolation valve with resilient material seals shall be a demonstrated OPERABLE by verifying that the measured leakage rate is less than W L, when pressurized to P,. o.05 l

        ,    ^~~._...                     ,

y ~. See <M tulty,L.( x./c

 \

i t

             ~ - -                                                             -

( - . - . . - _ . . . . _ . . - WOLF CREEK - UNIT 1 3/4 6-12 l e

l l l

 ~

l 4.6.1.7.2 Each 36-inch containment shutdown purge supply and exhaust isolation valve and its associated blank flange shall be leak testeo at least once per 24 months and following each reinstallation of the blank flange when pressurized to P valv$s, ar.d 48 psig, and including flanges, verifying that stemwhen the measured leakage, is added to leakage rate forrates the leakagc these determined pursuant to Specification 4.6.1.2d for all other Type B and C pene.trations, the combined leakage rate (s less than 0.60 L,. e l l *Except valves and flanges which are located inside containment. These valves i shall be verified to be closed with their blank flanges installed prior to L entry into MODE 4 following each COLD SHUTOOWN. i I l I

f

                                                                                                                        )
               ,                                                                                                      7 I CONTAINMENT SYSTEMS g { hh bh.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS u

                   , CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Cw**;* = t Spray System capable of taking suction from the RWST and transferring suction

__._..to -the containment sump. APPLICABILITY: MODES 1, 2, 3, and 4.

    ~

g.C ACTION:

            ,      With one Containment Spray System inoperab , restore the inoperaole pray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to GPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REOUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, l sealed, or otherwise secured in position, is in its correct position;

b. By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 265 psig when tested pursuant to Specification 4.0.5;' MO

c. At least once per 18 months during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment W G (CSAS) test signal, and pum p;g.3

2) Verifying that each spray pump starts automatically on a Containment (M-4 (CSAS) test signal.

Pmsw<-WlA-3

d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

1 WOLF CREEK - UNIT 1 3/4 6-13

I CONTAINMENT SYSTEMS 3 g SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION . 3.6.2.2 The Spray Additive System shall be OPERABLE with:

a. A spray additive tank containing a volume of between 4340 and 4540 gallons of between 28 and 31% by weight NaOH solution, and

b. Two spray additive eductors each capable of adding NaOH solution from the chemical additive tank to a Containment Spray System pump flow.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the Spray Additive System inoperable, restore the system to OPERABLE statu within 72 hours or be in-at least HOT STAND 8Y within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

b. At least once per 6 months by:

             ~

1) Verifying the contained solution volume in the tank, and

2) Verifying the concentration of the NaOH solution by chemical analysis.

c. At least once per 18 months during shutdown, by verifying that each automatic valve'in the flow path actuates to its correct position on (CSAS test signal; and aContainment,g+his"<-

d. 6-3 At least once per 5 years by verifying:

1) Each eductor flow rate is greater than or equal to 52 gpm using the RWST as the test source throttled to 17 psig at<ho eductor inlet, and

2) The lines between the spray additive tank and the eductors are not blocked by verifying f!cw.

WOLF CREEX - UNIT 1 3/4 6-14

     ~  /                                                                                        I
       /   CONTAINMENT SYSTEMS                                                                   !

CONTAINMENT CCOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.3 Two independent groups of containment cooling fans shall be OPERABLE with two fan systems to each group. . APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

a. With one group of the above required containment cooling fans inoperable and both Containment Spray Systems OPERABLE, restore the inoperable group of cooling fans to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

b. With two groups of the above required containment cooling fans inoperable and both Containment Spray Systems GPERABLE, restore at least one group of cooling fans to OPERABLE status witnin 72 hours or be in at least HOT STANOBY within the next 6 hours and in COLO SHUTDOWN within the following 30 hours. Restore both acove requireo

 -                     grouos of cooling fans to OPERABLE status within 7 days of initial

' loss or be in at least HOT STANDBY within the next 6 hours anc in COLD SHUTDOWN within the following 30 hours. c. grW ~ With one roup of the above required containment cooling fans inoperable and one Containment Spray System inocerable, restore the inoperable pray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOCWN within the following 30 hours. Restore the inocerable group of containment cooling fans to OPERABLE status within 7 days of initial loss or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.6.2.3 Each group of containment cooling fans shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1) Starting each non-operating fan group from the control room, and verifying that each fan group operates for at least 15 minutes, and

2) Verifying a cooling water flow rate of greater than or equal to 2200 gpm to each cooler

b. (M At least once per 18 months by verifying that on a Safety Injection test signal, the fans start in slow speed or, if operating, shift to slow speed and the cooling water flow rate increases to 4000 gpm to each cooler group.

jg WOLF CREEK - UNIT 1 3/4 6-15

1 CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES PR00F & RBH1 COPY 4 LIMITING CONDITION FOR OPERATION ~ 3.6.3 The containment isolation valves specified in Taole 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

                                 **"'"~

With one or more of the/ isolation valve (s) specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:

a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or ,

b. Isolate each afi'ected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position

               -end the provi;ien; cf Specification 3.0. ! are not appMesole,-or , c-

c. Isolate each affected penetration within 4 h'ours by use of at least one closed manual valve or blind flange :nd the ;1r_azisiens nf

               - Specificatica 3.0.4 are act applicabl% or         3

d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REGUIREMENTS

                     ,, daw 8"Y 4.6.3.1 The olation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuatcc, control or power circuit by performance of a cycling test, and verification of isolation time.

WOLF CREEK - UNIT 1 3/4 6-16

I CONTAINMENT SYSTEMS r)" g gP] (( SURVEILLANCE REOUIREMENTS (Continued) tc*4ganent 4.6.3.2 Each/Isolatica valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at leas.t once per 18 months by: . ( a. Verifying that on a. Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and

c. Verifying that on a Containment Purge Isolation test signal, each purge supply and exhaust isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of.each power operated ~or automatic valve of Table 3.6-1 shall be determined to be within its ifmit when tested pursuanc to Specification 4.0.5. h t I WOLF CREEK - UNIT 1 3/4 6-17

4 TABLE 3.6-1 h.  ! CONTAINMENT ISOLATION VALVES MAXIflUM TYPE LEAK . ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)

1. Phase "A" Isolation (active)

I 44 P-62 BB HV-80 D PRT Nitrogen C 10 Iso Valve P-62 BB HV-802 PRT Nitrogen C 10 Iso Valve P-24 BG HV-8100 Seal Water Return C

• 10 CTMT Iso Valve P-24 BG HV-8112 Seal Water Return C 10 CTMT. Iso Valve P-23 BG HV-8152 Letdown System CTMT C 10 Iso Valve P-23 BG HV-8160 Letdown System CTMT C 10 Iso Valve P-25 BL HV-8 7 Reactor Makeup Water C 10 CTMT Iso Valve P-21 EJ HCV-882 RHR to SI Test Line A 10 Iso Valve P-82 EJ HCV-8890A RHR A to SI Pumps Test A 13 Line Iso Valve P-27 EJ HCV-8890E RHR B to SI Pumps Test A 13 Line Iso Valve ,

P-49 EM HV-8823 [ SI/ Accumulator Injection A 10 Test Line Iso Valve P-48 EM HV-882

                                  ")   Safety Injection Pump B Test Line Iso Valve A         10 n TIw pc.; cans cf S fee.'f;c b c~      3 c 4 "" * * + 'ff 'I' d ' -

WOLF CREEX - UNIT 1 3/4 6-18 i k_

8 TABLE 3.6-1 (Continued) l CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)

1. Phase "A" Isolation (active) - (Continued)

P-88 EM HV-8843 Boron Injection Up- A 10 stream Test Line Iso P-92 EM HV-8871 SI Test Line to RNST C 10 Iso Valve P-87 EM HV-8881 " Safety Injection Pump A 10 Test Line Iso Valve P-92 EM HV-8964'" SI Test Line System C 10 Outside CTMT Iso P-99 GS tiV-3 Hydrogen Analyzer B A,C 5 Inlet Iso P-99 GS HV-4 Hydrogen Analyzer B A,C 5 Inlet Iso P-99 GS HV-5 Hydrogen Analyzer B A,C 5 Inlet Iso P-56 GS HV-8 Hydrogen Analyzer B A,C 5 Disch Iso P-56 GS HV-9 Hydrogen Analyzer B A,C 5 - Disch Iso P-101 GS HV-12 Hydrogen Analyzer A A,C 5 Inlet Iso P-101 GS HV-13 Hydrogen Analyzer A A,C 5 Inlet Iso P-101 GS HV-14 Hydrogen Analyzer A A,C 5 Inlet Iso P-97 GS HV-17 Hydrogen Analyzer A A,C 5 Disch Iso P-97 GS HV-18 Hydrogen Analyzer A A,C 5 Disch Iso

      % (,,;;s: A s    el ym she du 3 C'I u <- "I M f "' '        '

WOLF CREEK - UNIT 1 3/4 6-19 t _

TABLE 3.6-1 (Continued) {h h h, si CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS' VALVE NUM3ER FUNCTION TEST REOUIT<ED (Seconds)

1. Phase "A" Isolation (active) - (Continued)

P-101 GS HV-31 Sample Line to CTMT A,C 5 Atmos Monitor P-101 GS HV-32 Sample Line to CTMT A,C 5 Atmd'{ Monitor P-97 GS HV-33 Hydrogen Sample Return A,C 5 From PASS P-97 GS HV-34 Hydrogen Sample Return A,C 5

                                     , From PASS P-99           GS HV-36            Samole Line to CTMT         A,C             5
    ,       ,                           Atmos Monitor
 ,'   P-99           GS HV-37            Sample Line to CTMT         A,C             5
              ,                         Atmos Monitor-l      P-56           GS HV-38            Sample Return CTMT          A,C             5 Atmos Monitor F-56           GS HV-39           Sample Return CTMT           A,C             5 Atmos Monitor P-44           HB HV-7126         RCOT Vent Inside CTMT        C             10 P-26           HB HV-7136         RCDT Pumps Disch Hdr         C             10 Outside CTMT Iso P-44           HB HV-7150         RCDT Vent Outside     -

C 10 CTMT P-26 HB HV-7176 RCDT Pumps Disch Hrd C 10 Inside CTMT Iso P-30 KA FV-29 Readt3 Bldg Instr Air C 5 Sucpls Outside CTMT Iso Y P-32 LF FV-95 CTMT_ Normal Sumps to C 30 l Floor'Orain Tank Inside CTMT Iso i WOLF CREEK - UNIT 1 3/4 6-20

I i l 1 TABLE 3:6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)

1. Phase "A" Isolation (active) - (Continued)

P-32 LF FV-96 CTMT Normal Sumps to C 4 Floor Drain Tank Outside CTMT Iso P-93 SJ HV-5 PZR/RCS Liquid Sample C 5 Inner CTMT Iso u P-93 SJ HV-6 PZR/RCS Liquid Sample C 5 Outer CTMT Iso

                                  **                        dar P-69              SJ HV-12          PZR Vapor Sample I m r        C          5 CTMT Iso P-69              SJ HV-13          PZR Vapor Sample Inner        C          5 CTMT Iso

. +1 P-95 SJ HV-18 Accumulator Sample C 5 Inner CTMT Iso

                                 'r t P-95              SJ HV-19         Accumulator Sample             C          5 Outer CTMT Iso p-93               SJ HV-127         PZR/RCS Liquid Sample         C          5 Outer CTMT Iso P-64              SJ HV-128 "       PZR/RCS Liquid                A,C        5 Sample Inner CTMT Iso
                                   =+

P-64 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 RCDT 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 . Line Iso Valve

 ..       .'. 8.',"

• b " s

       *[YOLFCREEK-UNIT 1an{ng,rmit}entbjsis$derpgg,nistrativecontrol.

3/4 6-21 f%

TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES { { [ f r$ h hi b MAXIMUM - TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REOUIRED (Seconds)

2. Phase "A" Isolation (passive)" - (Continued)

P-16 EN HV-01 CTMT Recire Sump to CTMT A L A. 30 Sprar Pump A Iso o P-13 EN HV-07 CTMT Recire Sump to CTMT A NA Sc. Spray Pump B Iso P-45 EP HV-8800 CTMT Nitrogen Supply C -M-/n ic Iso Valve r+ P-65 GS HV-20 Hydrogen Purge Inner C h 5 CTMT Iso P-65 GS HV-21 Hydrogen Purge Outer C -NA- 5 CTMT Iso P-67 KC HV-253 Fire Protection System C bb & Hdr Outer CTMT Iso

3. Phase "B" Isolation (active) 0-74 EG HV-58 CCW to RCS Iso C 30 P-75 EG HV-59 CCW Return From C 30 RCS Iso P-75 EG HV-60 CCW Return From -

C 30 RCS Iso P-76 EG HV-61 CCW Return From C 30 RCS Iso P-76 EG HV-62 CCW Return From C 30 RCS Iso

4. Containment Purge Isolation (active)

V-161 GT HZ-4 CTMT Mini-Purge C 3 Supply Outside CTMT Iso V-161 GT HZ-5 " ' CTMT Mini-Purge C 3 Supply Inside CTMT Iso "May be opened on an intermittent basis under acministrative control.

 + + h , gig.g e f    b   .4<-m      3 c.4  n.c  ,t     a               ,
                                                       , gs, <- 6 / *

~ , "* abkl l,$s'.~.? e e$$,, ~ 'ff ""- f'- "I'd '^ c f'"a h . Nu

                                                                                                                   's          i      I
                                  .                        TABLE 3.6-1 (Continued)                    \

CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST RE0VIRED (Seconds)

4. Containment Purge Isolation (active) - (Continued)

V-160 GT HZ-11

• CTMT Mini-Purge C 3 Exh Inside CTMT Iso V-160 GT HZ-12 * CTMT Mini-Purge C 3 Exh Outside CTMT Iso

5. Containment Purge Isolation (passive)

V-161 GT HZ-6 *" CTMT.S/D Purge Supply Outside C b CTMT Iso IJ s V-161 GT HZ-7 *** CTMT S/D Purge C Jkk Supply Inside CTMT Iso

                                                                                                .                    10

V-160 GT HZ-8 "

• CTMT S/0 Purge C -McA~

Exh Inside CTMT Iso 10 V-160 GT HZ-9

• CTMT S/D Purge C S k.

Exh Outside CTMT Iso

6. Remote Manual P-41 BB HV-8351A RCP A Seal Water C N.A.

Supply P-22 BB HV-8351B RCP B Seal Water C N.A. Supply P-39 BB HV-8351C RCP C Seal Water C N.A. Supply r P-40 BB HV 8351b ' RCP O Seal Water C N.A. -

                                            ...         Supply P-79                BB PV-8702A               RCS Hot Leg 1 to RHR            A                         N.A.

Pump A Suction

                                                                                 .s         , tr o.u c                            w
                                        <f     s . , h u. h'. 4   t.t.4    e.. .                           ',,. , , g , A 5+,     p       . . . . :u s "p"

g . . w w. .' " b' r' " "' - " ' WOLF CREEK - UNIT 1 3/4 6-23 e _. . -. . _ _ .

i

                  ~

TABLE 3.5-1 (Continued) PROP!' & RBS COPY - CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REOUIRED (Seconds)

6. Remote Manual - (Continued)

P-52 BB PV-87028 RCS Hot Leg 4 to RHR A N.A. Pump B Suction

                                        '                                                                          5 P-15              EJ HV-23             PASS Sump Sample CTMT                           C           NrA; Iso n                                                                          6 P-15              EJ HV-25             PASS Sump Sample CTMT                           C           N;A.

Iso P-14 EJ HV-24 PASS Sump Sample CTMT C N .

                                                 . Iso
            'P-14              EJ HV-26             PASS Sump Sample CTMT                           C           N. N.

Iso . P-71 EF HV-31 ESW Supply To C N.A. Containment Coolers P-28 EF HV-32 ESW Supply To C N.A. Containment Coolers P-71 EF HV-33 ESW Supply To C N.A. Containment Coolers P-28 EF HV-34 ESW Supply To C N.A. Containment. Coolers P-73 EF HV-45 ESW Return From C N.A. Containment Coolers P-29 EF HV-46 ESW Return From C N.A. Containment Coolers P-73 EF HV-47 ESW Return From C N.A. Containment Coolers a N. ,i 4 a, , e?

                .,v.....,s  .

sg , 4 , ,i..s 3.< . . g . . < ,& . WOLF CREEK - UNIT 1 3/4 6-24

i P;100F & IHiEW COPY TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME 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 P-29 EF HV-50 ESW Return From C N.A. Containment

                                   . Coofers 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-76 EG HV-132* CCW Return From RCP C N.A. Thermal Barriers P-76 EG HV-133* CCW from RCP Thermal C N.A. _ Barrier 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-27 EJ HV-88098 RHR Pump B Cold Leg A N.A. Injection Iso Valve P-15 EJ HV-8811A CTMT Reci K Sump to A N.A. RHR Pump Suc, ion (,

     "These valves were assumed to be closed during the accident analysis, and are normally closed but may be opened on an intermittant basis under adminstrative control.

WOLF CREEK - UNIT 1 3/4 6-25

n. _ _ _ _ _ . - - _ _ __ - - _ _ _ _ _ . - - . _ __

TABLE 3.6-1 (Continued) PPM & WiE CD CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds)

6. Remote Manual - (Continued)

P-14 EJ HV-88118 CTMT Recirc Sump to A N.A. RHR Pump B cioni L )c P-21 EJ HV-8840 RHR Hot Leg A N.A.

 ,                                            Iso Valve P-87             EM HV-8802A*      SI Pump A Disch Hot        A             N.A.

Leg Iso Valve P-48 EM'HV-88028" SI Pump B Disch Hot A N.A. Leg Iso Valve P-49 EM HV-8835 ~ SI Pumps Disch to A N.A. Cold Leg Iso Valve P-89 EN HV-6 CTMT. Spray Pump A A N.A. so Valve / P-66 EN HV-12 Disch{Aff) CTMT Spray Pump 8 A N.A. Discharge Iso Valve -

7. Active for SIS P-80 BG HV-8105 CVCS Charging Line C th*.10 P-88 EM HV-8801A Boron Injection to A N.A.

RCS Cold Legs P-88 EM HV-88018 Boron Injection to A N.A. - RCS Cold Legs

8. Hand-Operated and Check Valves l

[ P-41 88 V-118 RCP A Seal C N.A. l Water Supply P-22 88 V-148 RCP 8 Seal C N. A. Water Supply P-39 88 V-178 RCP C Seal C N.A. Water Supply P-40 BB V-208 RCP 0 Setl C N.A. Water Supply l

           *These valves were assumed to be closed during the accident analysis and are normally closed cut may be opened on an intermittent basis under administrative control.         ,

WOLF CREEK - UNIT 1 3/4 6-26 M

                                                                        05u U!!Emk COPY 11; TABLE 3.6-1 (Continued)

CONTAINMENT ISOLATION VALVES MAXIMUM TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REQUIRED (Seconds) __

8. Hand-Operated and Check Valves - (Continued)

P-24 BG V-135 RCP Seal Water Return C N.A. P-80 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 Orain 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. From 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 P65 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 EG V-204 CCW Supply to RCP C N.A. P-82 EJ 8818A RHR Pump to Cold A N.A. Leg 1 Injection I P-82 EJ 88188 RHR Pump to Cold A N.A. Leg 2 Injection 1 WOLF CREEK - UNIT 1 3/4 6-27 I L

                                ~ s                   4 i    t~          th
                       \'

r

s. s TABLE 3.6-1 (Continued)

                                                                       * \;                   \

hih b t i I

                                                            ' CONTA1HMENT ISOLATION VALVES
                                                          ,          t MAXIMUM -

s TYPE LEAK ISOLATION TIME PENETRATIONS VALVE NUMBER FUNCTION TEST REOUIRED (Seconds)

 -                                                                                                                                                    ~
                                                  >                      x

8. Hand-Cperated'and Check Valves b (Continued)

P-27 EJ 8818C RHR Pump t.o Cold A N.' A. s Leg 3 Inje'ction-P-27 EJ 88180' RHR Pump to Cold A N.A. Leg 4 Injection l'P-21 EJ 8841A RHR Pump Disch to A N.A. RCS Hot Leg 2 P-21 [. EJ 88418 RHR Pump Disch to A N.A. t t 'T4 RCS Hot Leg 3 P-87 <. EM V-001 SI Pump Hot Leg 1 A , N.A.

                                  \                                  Injection P-8'7 - \                    EM V-002                 SI Pump Hot Leg 2                   A                  N.A.

Injection

                              ^
            .P-48                           EM V-003                 SI Pump Hot Leg 3               'A                     N.A.

b( ' Injection (, P-48 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. Leg 2 P-49 EM \-030 SI Pump Disch to Cold A N.A. Leg 3 P-49' ' EM V-040 SI P6mp Disch to Cold A N.A. Leg 4 V-P-88 EMl8815 BIT to RCS Cold L.eg A N.A. Injection P-89 EN V-013 CTMT Spray Pump A A N.A. to CTMT Spray Nozzles l WOLF CREEK - UNIT 1 3/4 6-28 l

m ,... a > g c . .v .y,. o.w. ,.,e. . , . . . .

      .a v ,g .g y.,                  -
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r

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Q,. . ; , . TABLE 3.6-1 (Continued)  !

           %,      ?...

v: . . c: m. ' . . '. CONTAINMENT ISOLATION VALVES ..'

                                                                                                                                                                                                                                                                                                                                                                      '.                      e     c' b.? '

MAXIMUM 't c.

        ".% i /
                                                                                'L TYPE LEAK                                                                                   ISOLATION TIME d l - PENETRATIONS VALVE NUMBER s

2' FUNCTION TEST REQUIRED (Seconds) <N W

        ? 1'f@3          t *' 6 C M .b '                           ~6.                Hand-Operated and C' heck Valves - (Continued) '
                                                                                                                                                                                                                                                      "              ~
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Ql,M.',P-66 ch. EN V-017 CTMT Spray Pump B ~ to CTMT,S,epray Nozzles . . 2

                                                                                                                                                                                                                                                            'A ,

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                                                                                                                                                                                                                                                                                                    .; Mt . c,n N.A.
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                                                                                                                                                                                                                                                                                                                                                                                     . ,s y EP V-046                                                       Accumulator Nitrogen                                                                        ;C                                     "
                                                                                                                                                                                                                                                                                                              ;. . . . N.A.

M, h'o~.m;. . ~ ... :. h. ' . ~, .m.w .

Supply Line ~

[ , ' ' &.$. \ , 4">

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    . 4,;;$$'y.:,, ,                                                                           HD V-016                                         .*            Auxiliary Steam to
                                                                                                                                                                                                                                                                                                                ; .,',N.A.

d ~yf,P-43 ' C ..

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Decon System j%. ,g. s p,fy.

             .t
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vi- P-43 HD V-017 Auxiliary Steam to ' , C N.A. . , Decon System e P-63 KA V-039 Rx Bldg Service Air C ' .N.A. I Supply 7 .4.a . 9 *1. P-63 KA V-118 Rx Bldg Service Air C* N .'A . . . 'l . T' . " C Supply.,

                                                                                                                                                                                                                                                                                                                                                                . l, !~<f. ,,F.,.

KA V-204 Rx Bldg Instrument C

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c4 1 ' .'.# %

                                          )30.

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( 'l P-67 KC V-478 Fire Protection C N.A.:.','. % Supply to RX Bldg '

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i.asert.

9. Other Automatic Valves . . . ,

P-1 AB-HV-11** Mn. Stm. Isol. A 5 P-2 AB-HV-14** Mn. Stm. Isol. A 5 . P-3 AB-HV-17** Mn. ,Stm. Isol. A 5 P-4 AB-HV-20** Mn. Stm. Isol. A 5 P-5 AE-FV-42** Mn. FW Isol. A 5 P-6 AE-FV-39** Mn. FW Isol. A 5 P-7 AE-FV-40** Mn. FW Isol. A 5 P-8 AE-FV-41** Mn. FW Isol. A 5 P-9 BM-HV-4** SG Blowdn. Isol. A 10

    ' P-10           BM-HV-1**       SG.Blowun. Isol.         A     10 P-11           BM-HV-2**                                             .

SG Blowdn. Isol. A 10 P-12

                ~

BM-HV-3** SGBlowdn.ksol. A 10 i l p **The provisions of Specification 3.0.4 are not applicable. . i

  .                                                                   e I

CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: A. With one containment hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours.

b. W* t" M" b]d u y .3 arug tu s i.vju.We ,

r< 4 4 J lemst c .s c

u. tu 4c C Pf. R.Stk i skte wi fh;^

7 2. b. - 6 o ,- b e. ,,, :t

              ;ad       kr       st^eb6I         ed 6.- %     nae       (, wo .. . n J

SURVEILLANCE REOUIREMENTS - c 4.6.4.1 Each containment hydrogen analyzer shall be demonstrated OPERABLE by the performance of a--C::ANNEL C;;ECK Et lesst cr,ce pe. 12 huu,y anANALOGCgANNEL 9 OPERATIONAL TEST _atjeast_once per 31 days a r.d .e t ' -M t ech ge c 02 cys fon a , F STAGGERLD TCST BASIS by performing a CHANNkt--CALIBRAT_ ION using sample gas containing b vel e pe,# hdnpn, bl*6 a*%en. c N

a. Dns u i uaa- p cent nycry.n, alenc: -it~ ;aa and-

                        ~                                                                          '

t. ca_ volume. perc+nt--hydrogeny-balance-nitrogeng_ - L

                                                     ~

WOLF CREEK - UNIT 1 3/4 6-30 l

                                                                   .-                                      . -, .                ~_. - .

CONTAINMENT SYSTEMS gar & E a CO?Y HYOROGEN CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.4.2 A Hydrogen Control System shall be OPERABLE with two independent Hydrogen Recombiner Systems, .. . .., . 3... .......,.....ono wu. independent Hydregen P.eca-biner Sy:te . APPLICABILITY: MODES 1 and 2 ACTION:

      'With one ofy-'+- the two independent Hydrogen Recombiner Systems and the Hydrogen f%n9rs 9t               inoperable, restore the inoperable Hydrogen Recombiner System or tf.; Hydreg;r Purge Sabsystem to OPERABLE status within 30 days or be in at least H'JT STANOBY within the next 6 hours.

SURVEILLA CE REQUIREMENTS 4.6.4.2.1 Each Hydrogen Recombiner System shall be demonstrated OPERABLE: a. At least o'nce per 6 months by verifying, during a Hydrogen Recombiner System functional test, that the heater air temperature increases to ! greater than or equal to 1150*F within 5 hours; and

b. At least once per 18 months by:

                            ~

1) Performing a CHANNEL CALIBRATION of all hydrogen recombiner system instrumentation and control circuits,

2) Verifying through a visual examination that ,there is no, evidence of abnormal conditions within the hy' drogen recombiner . system enclosure (i.e., loose wiring or structural connections, deposits of foreign materials, etc.), and l

3) Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.

d . 6. d . 2. 2 The Hydrogen-Perge-Gubsystem-sha'i be demonstreted CPERASLE by-

     -cycling valve: GS-HV20, GS-HV21, and KA-HV30 at 10:st cacc par 21 days.

l l l l l WOLF CREEK - UNIT 1 3/4 6-31 l l \

PR007 & REV!EW COPY 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-3. APPLICABILITY: MODES 1, 2, and 3. ACTION:

a. With four reactor coolant loops and associated steam generators in

      ,             operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

b. With three reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves associated with an operating loop inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS i-4.7.1.1 No additional requirements other than those required by Specification 4.0.5. . o l WOLF CREEK - UNIT 1 3/4 7-1 O

l i f3h h i il TABLE 3.7-1 MAXIMUM ALLOWABLE P"WER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES OURING FOUR LOOP OPERATION MAXIMUM NUM8ER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (Percent of RATED THERMAL POWER) 1 87 2 65 3 44 TABLE 3.7-2

l. . .

MAXIMUM ALLOWA8LE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH l INOPERABLE STEAM LINE SAFETY VALVES OURING THREE LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWEP RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT CPERATING STEAM GENERATOR * (Per:ent of RATED THERMAL POWER) 1 ==

                                                         ~

g MR 3 **

                       *At least two safety valves shall be OPERABLE on the non-operating steam generator.

These values left blank pending NRC approval of three loop operation. WOLF CREEK - UNIT 1 3/4 7-2 A

6 F * [ TABLE 3.7-3 y STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (11%) ORIFICE SIZE e Loop 1 Loop 2 Loop 3 Loop 4 V055 V065 V075 V045 1185 psig 16.0 sq. in. V056 V066 V076 V046 1197 psig 16.0 sq. in. V057 V067 V077 V047 1210 psig 16.0 sq. in. V058 V068 V078 V048 1222 psig 16.0 sq. in. V059 \069 V079 V049 1234 psig 16.0 sq. in. 7 w t:s

c C"3 C. 3 a mt The lif t setting pressure shall correspond to ambient conr'itions of the m valve at nominal operating temperature and pressure. .

cm. O C.3 "O M

l I PLANT SYSTEMS I AUXILIARY FEEDWATER SYSTEM PRODF & REV!EW COPY l LIMITTNG CONDITION FOR OPERATION 3.7.1.2 At least three irdependent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor-driven auxiliary feedwater pumps, each capable of being powerad from separate emergency busses, and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3. ACTION:

a. With one auxili,ary fe'edwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTCJHN within the following 6 hours.

b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY witin 6 hours and in HOT SHUTDOWN within the following l 6 hours.

c. With three auxiliary feedwater pumps inoperable, immediately iniciate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by:

1

1) Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1535 psig on recirculation

! flow when tested pursuaat to Specification 4.0.5; 23 Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1625 psig at a flow of l greater than or equal to 120 gpm when the secondary steam supply pressure is greater than 900 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3; l . WOLF CREEK - UNIT 1 3/4 7-4

PLANT SYSTEMS PDflr1: nUu, Q nLvi!.Il udi i9 DO,lStf poDV SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and

4) Verifying that each automatic valve in the flow path is in the fully cpen position whenever the Auxiliary Feedwater System is placed in automatic control or whan above 10% RATED THERMAL POWER. -

b. At least once per 18 months during shutdown oy:

1) Verifying that each automatic valve in the ESW supply to the auxiliary feedwater pumps actuates to its full open position upon receipt of an Auxiliary Feedwater Pump Section Pressure-Low test signal,

2) Verifying that each auxiliary feedwater-pump. starts as designed automaticallyuponreceiptofan'Auxiliapf,feedwaterActuation test signal, and g Sy __-

3) Verifying that each au liary feedwater motor-operated discharge valve limits the ficw from the motor-driven pump to less than or equal to 320 gpm. -;

4.7.1.2.2 An auxiliary feedwater flow path shall be demonstrated 0FERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying normal flow to at least two steam generators from one auxiliary. feedwater pump. W0LF CREEK - UNIT 1 3/4 7-5

I l hhh[ 't fhg ff bb CONDENSATE STORAGE TANK . __

                                                                                                                   ~

LIMITING CONDITION FOR OPERATION

3. 7.1. 3 The condensate sto g tank (CST) shall be OPERABLE with a contained water volume of at leastone M allons.

APPLICABILITY: MODES 1, 2, and 3. ACTIONf With the CST inoperable, within 4 hours either:

a. Restore the CST to OPERABLE status or be in at least HOT STANOBY within the next 6 hours and in HOT SHUTDOWN within the following 6

         .                    hours, or e

b. DemonstratetheOPERABILITYoftheEssentialServiceWater(ESW)kas. a backup supply to the auxiliary feedwater pumps and restore the CST to OPERABLE status within 7 days or be in at least HOT STANDSY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REOUIREMENTS 4.7.1.3.1 The CST shall be demonstrated OPERABLE at least once per 12 hours l- cy verifying the contained water volume is within its limits when the tank is ! the supply source for the auxiliary feedwater pumps.

• 4.7.1.3.2 The ESW System shall be demonstrated OPERABLE at least once pe'r .

 .               12 hours by verifying that the ESW System is in operation wt.enever the ESW System is the supply source for the auxiliary feedwater pumps.                              -

WOLF CREEK - UNIT 1 3/4 7-6

                            -                       ,       ,                                                  -+-
                                                                        -w -       -          v---g--

PLANT SYSTEMS

                                                                          }}f h k        g i COM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 ~ The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurfe/ gram DOSE EQUIVALENT I-131.

1 APPLICABILITY: MODES 1, 2, 3, ana 4. ACTION: With the spesific activity of the Secondary Coolant System greater than 0.1 microcurie / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the Secondary Coolant System snall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. 4 l ! WOLF CREEK - UNIT 1 3/4'7-7 e

                                                      -_              ~. _ , - - -

PROM & RUi3<' M . TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FRE00ENCY

1. G Nss Radioactivity Determination At least once per 72 hours.

2. Isotopic Analysis for DOSE a) Once par 31 days, when-EQUIVALENT I-131 Concentration ever the gross radioactivity determination indicates concentrations greater than 10% of the allowable limit for radiofodines.

• b) Once per 6 months, when-ever the gross radioactivity

                                                                                        . determination indicates concentrations less than or equal to 10% of the allowable limit for radiciodines.

S 1 1 l WOLF CREEK - UNIT 1 3/4 7-3 l .

                                                                                                             - - - - - -

• y- --

w .e--,- g .4 q-- ---- - -----y i-- - y

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES PR0"* & W 'M COPY LIMITING CONDITION FOR OPERATION

 '3.7.1.5 Each main steam line isolation valve (MSLIV) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3. ACTION: MODE 1: With one MSLIV inoperable,but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. MODES 2 and 3: With one MSLIV inoperable, subsequent operation in MODE 2 or 3 may preceed provided the isolation valve is maintained closed. Otherwise. be in HOT STANOBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSLIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3. i 1 i I WOLF CREEK - UNIT 1 3/4 7-9 i

     -                 +                                      ,.. . -             ,.--

PLANT SYSTEMS I . [ 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2 The temperatures of both the reactor and secondary coolants in the

     , steam generator shall be greater than 70*F when the pressure of either coolant in the steam generator is greater than 200 psig.

APPLICABILITY: At all times. ACTION: With the requirements of the above specification not satisfied:

a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and

             .b. Perform an engineering evaluation to determine the effect of the over?ressurization on the structural integrity of the steam generator.

Ootermine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F. SURVEILLANCE REOUIREMENTS 4.7.2 The pressure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the temperature of either the reactor or secondary coolant is less than 70*F. t l l WOLF CREEK - UNIT 1 3/4 7-10 i i _ __ _ _

           "'^"' S'S " S PROD: & HD!! COPY 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.7.3 At least two component cooling water loops shall be demonstrated OPERABLE: J a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct positJon. In addition, an ANALOG CHANNEL OPERATIONAL TEST of the surge tank level and flow instrumentation which provide automatic isolation of the non-nuclear safety-related portion of the system shall be performed at least once per 31 days;

b. At least ones. per 18 months during shutdown, by verifying that:

1) Each automatic valve servicing safety-related equipment or isolating the non-nuclear safety-related portion of the system actuates to its correct position on a Loss-of-Power or Safety I W ction :::: si;;=1 and on a simulated High Flow and Low Surge Tank Level test signal, and -

2) Each OPERABLE Component Cooling 'JaDir System pump starts auto-matically on a Safety Injection gloss-of-Power test signal.

c. Atleastonceper18monthsduringshutdbwn,byperformingaCHANNEL CALIBRATION of the surge tank level and flow instrumentation which provide automatic isolation of the non-nuclear safety-related portion of the system.

WOLF CREEK - UNIT 1 3/4 7-11

PLANT SYSTEMS 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent essential service water (ESW) loops shall be OPERA 8LE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only one ESW loop OPERABLE, restore at least two ESW loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. SURVEILLANCE REQUIREMENTS - 4.7.4 At least two ESW loops shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position. In addition, at least once per 31 days, an

ANALOG CHANNEL OPERATIONAL TEST of the differential pressure l instrumentation for automatic isolation of the ESW to the air compressors shall be performed;

b. At least once per 18 months during shutdown, by verifying that:

l 1) Each automatic valve servicing safety-related equipment or l isolating the non-nuclear safety-related portion of the system actuates to its correct position on a Loss-of-Power or Safety Injection test signal and on a simulated High Differential Pressure test signal, and

2) Each Essential Service Water System pump starts automatically on a Safety Injection, Low Suction Pressure (AFW pumps), and Loss-of-Power test signal.

l c. At least once per 18 months during shutdown, by performing a CHANNEL CALIBRATION of the differential pressure instrumentation for automatic isolation of ESW to the air compressors. WOLF CREEK - UNIT 1 3/4 7-12 l

PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK PROSF<!: E ?<? P>0M LIMITING CONDITION FOR OPERATION Mm 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with: a. The crest of the UHS dam without the rip rap cover on top and correspcnding water level at or above elevation 1069.5 F.ean Sea Level, USGS datum, and

b. The lake .;;;t (plant inlet)-water temperature of less than or equal to 90.*F.,___ __.

APPLICABILITY: MODES 1, 2 3, and 4. ACTION: With the requirements of the above specification not satisfied, be in at least l HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS l 4.7.5 The UHS shall be determined OPERA 8LE: a.

   -                 At least once per 24 hours by verifying the above required water temperature and water level to be within their limits, and

b. At least once per 12 months by verifying that the crest of the VHS dam without the rip rap cover on top is above elevation 1069.5 Mean Sea Level, USGS datum.

1 l l i I WOLF CREEK - UNIT 1 3/4 7-13 j 1

a PLANT SYSTEMS

   '3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two independent Control Room Emergency Ventilation Systems shall be OPERABLE.

APPLICA8ILITY: All MODES. ACTION: MODES 1, 2, 3 and 4: With one Control Room Emergency Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT LTANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

                                        ~

MODES S and 6:

a. With one Control Room Emergency Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Emergency Ventilation System in the recirculation mode. .

b. With both Control Room Emergency Ventilation Systems inoperable, or with the OPERABLE Control Room Emergency Ventilation System, required to be in the recirculation mode by ACTION a., not capable of being

} powered by an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS

 ,  4.7.6 'Each Control Room Emergency Ventilation System shall be demonstrated OPERABLE:

a. At least once per 12 hours by verifying that the control room

, air temperature is less than or equal to 84*F; I l b. At least once per 31 days on a STAGGERED TEST BASIS, by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers of both the Filtration and Pressurization Systems and verifying that the Pressurization System operates for at least 10 continuous hours with the heaters operating; WOLF CREEK - UNIT 1 3/4 7-14

            -                                                                         ~

PLANT SYSTEMS 100 & REW COPY SURVEILLANCE REOUIREMENTS (Continued)

c. At lea'st once per 18 months or (1) after any structural maintenance _

on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, orby: v chemical release in ang*zgtilation zone communicating with the system

• ggq

1) Verifying that theed:p,.\=p Eca^

stem satisfies the in place penetration and bypass leakage testing acceptance criteria; of less than 1% and uses the test procedure guidance in Regulatory Positions C.S.a, C.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2000 cfm 1 10% for the Filtration System and 2000 cfm 10% for the Pressurization System with 500 cfm 10% going through the Pressurization System filter adsorber unit;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.bA Regulatory Guide 1.52, Revision 2, March 1978, meets /the laboratory testing criteria of Regulatory Po'sition C.6.afof Reg'ylatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;

and - V

                                                            ~

3) Verifying a system flow rate of 2000 cfm + 10% for the Filtration

   '                       System and 2000 cfm ! 10% for the Pressurization System with 500 cfm       10% going through the Pressurization System filter adsorber unit during system operation when tested in accordance with ANSI N510-1975.

d. After every 720' hours of charcoal adsorber operation by verifying within 31 days after removal, that a laboratory analysis of a represen-P tative 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.Cj.E.a of Regulatory

                  . Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;

e. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and chercoal adscroer banks is 1.ess than 5.4 inches Water Gauge while operating the system at a flow rate of 2000 cfm 210% for the Filtration System and 500 cfm : 10% for the Pressurization System filter adsorber unit, 2)* Verifying 7 hat orrrCdntr5T Room Ventilation Isolation $

                                                        , .he system automatically-switches into
                                                                        ~-~
                          ~5* recirculation ~ mode of~ operation w'ith flow through the HEPA filters and charcoal adsorcer banks, WOLF CREEK - UNIT 1                      3/4 7-15

st ,,.

• _

  ;d; .";           .                                                                                                                                                                  .
  ' . ;..s

$s . . ' .; ,

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) _+

                                                                                                           '           ~

z' . . Verifying that the system maintains the control room at a f -: - 'f 3) .. f,. .

       ~

9,,. :

                                         ,3                                      positive pressure of greater than or. equal to 1/4 inch Water . , .; g -                                                      .
      , j ;- O                                7
                                                            ~,

Gauge relative to the outside atmosphere during system operation," .g.

       ,O,.h.
           *                                   * ,"y'
                                             , .                   ')

4 Verifying that .the Pressurization System filter adsorber unit i ,4. ,'c.. y .. c? Sd/ - .. .

                                                                               . heaters dissipate 15 i 2 kW in the Pressurization Syst.em.when                                                       . ~ 4rm",'
    % ,M'd'N,-d<                                                          D tested in accordance with ANSI H510-1975, and                                                         l-; .....                   T.D
                                                                                                                                                                                      -c.~u                    .6$. 3W   . ..: .
    . .; a. ~, , .x.
                                                                    ?
                                                                                                                  ..w. c. m          .;                 ,
                                                                                                                                                                  -                                      e
                                                                                                                                                                                             .. . . . c..,
                                    .,i? S G S) .                                                                                                                                                                            @

L. ' - .

 , F,M. :. //w?ffkU j/:

Verifying that on a High Chlorine "% test signal," - VG c: g9

                                .-      ?/ IP . '                         j~thesystemautomaticallyswitchesintoarecirculationmode'ccy#l@

of operatio with flow through the HEPA filters'and charcoal ^ .::

 - My
 . p r.p4 v.jg, m... . .3; ;;:.r...-  -

o g% ',s p ,,, p gi:c ;y ..Jadsorber ba within 15 seconds.

                                                                                                                                                  ,         T
                                                                                                                                             .; @c-- _r ' '.m -a r" _, 'qJ                ; w'i--?? ,. c.. N;g
                                                                                                                                                                                        ~ .. . ..
    - ."                         c'Wf. " :After each complete or partial replacement of a,iiEPA filter bank,' by A.W.n.
                                                                 -verifying that the cleanup system satisfies the in place penetra . L l.E Y
           ,t. '~. M q/',
                                                                 ' tion and bypass leakage testing 'P 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 2000 cfm

  ~
                                                        /

10% for the < L

                                           ..                    . Filtration System and 500 cfm i 10% for the Pressurization System
                                           :: y--                filter adsorber unit; and
                                                      .3 t'                             "g.' . . After each complete or partial replacement of a charcoal adsorber                                                                                        Ar
                                               "          ~

bank, by verifying that the cleanup system satisfies'the in place - ./- penetration and bypass leakage testing aseeptonee criteria of less C .j i-than 1% in accordance with ANSI N010-1975 for a halogenated hydrocarbon refrigerant test gas while operating the system at a r f , flow. rate of 2000 cfm + 10% for the Filtration System and 500 cfm *

                                      ~

! , 10% for the Pressurization System filter adsorber unit. '

                                                                                                                                                            '     ~

,_.,._ i y k .

                                                                          . . .(

l 4 WOLF CREEK - UNIT 1 3/4 7-16

                                                                     ?R00F& H M CD?Y PLANT SYSTEMS 3/4.7.7 EMERGENCY EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION                                                     _

3.7.7 Two independent Emergency Exhaust Systems shall be OPERABLE. l APPLICABILITY: MODES 1, 2, 3, and 4. U,1 b ACTION: vb pt6 With one Emergency Exhaust Systen inoperable, restore the inoperabledystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.7.7 Each Emergency Exhaust System shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, l from the control room, flow through the HEPA filters and charcoal adsort ers and verifying that the system operates for at least 10 continuous hours with the heaters operating;

b. At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following

   -             painting, fire,orchemicalreleaseinangventilationzone communicating with the system, by:      6M

1) Verifying that wttr the stem satisfies the in place penetration and bypass leakage ' testing acceptance criteria of less than 1%

and uses the test procedure guidance in Regulatory Positions C.S.a C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 9000 cfm 10%;

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 criterie of Regulatory Position C.6.a of egulatory Guide 1.52, Revision ,

2, March 1978, for a methyl 10 de penetration of less than 1,%; a d WOLF CREEK - UNIT 1 3/4 7-17

PLANT SYSTEMS 3R00F & REMEW COPY SURVEILLANCE REOUIREMENTS (Continued)

3) Verifying a system flow r-te of 9000 cfm 10% during system operation when tested in accordance with ANSI N510-1975. _

c. After every 720 hours of charcoal adsorber operation, by 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 ofblatory Guide 1.52, Revision 2, March 1978, for a methyl i dide enetration of less than L%;

d. At least once per 18 mohths by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks of less than 5.4 inches Water Gauge while operating the system at a fick rate of 9000 cfm i 10%, .

7

2) Verifying that the system maintains the uel Building at a' negative pressure of greater than or equal to % Water Gauge relative to the outside atmosphere during systam operation,

3) Verifying that the system starts on a Safety Injection test signal, and

4) Verifying that the heaters dissipate 37 3 kW when tested in accordance with ANSI N510-1975.

1

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing aeeecAeese 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 10%; and

f. After each comolete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acr-" ;e criteria of less than 1% - .-;.; ,-- - - . in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 9000 cfm 10%.

WOLF CREEK - UNIT 1 3/4 7-18

PLANT SYSTEMS 3/4.7.8 SNU8BERS O { { 1[I Q f l LIMITING CONDITION FOR OPERATION 3.7.8 All hydr = lid =d =haMcal snubbers shall be OPERABLE.h cJ sa. LM c.- s. y.d on n d a.t .a ~.a.my - avw s ,, a n ,, m;, u,e

                                                                          ~
          'Asystems
             'PPLIC3BIITY:

L YD5[1,T, 3, indl. T0dEIS aN 6 hor ,s" nub 57r^s ifoIa't required OPERABLE in those MODES. ACTION: With one or more snubbers i,noperable on any system, within 72 hours replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.8g on the attached component or declare the attached system inopera-ble and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS 4.7.8 Each snubber shall be demonstrated OPERABLEl%rfoAa~ncTof the foMing, augmented inservice inspection progra a fu

                                                                          # 44. rega.a.=d cf IF LA** 4            )

a. Insoection Tvoes As used in this specification, type of snubber shall mean snubbers

   .J 9

of the same design and manufacturer, irrespective of capacity.

b. Visual Insoections Snubbers are categorized as inaccessible or accessible during eactor operation. Each of these groups (inaccessible and accessible) may be inspected independently according to the schedule below. The first inservice visual inspection of each type of snubber shall be performed after 4 months but within 10 months of commencing POWER OPERATON and shall include all hydraulic and mechanical snubbers. If all snubbers of each type on any system are found OPERABLE during the first inservice visual inspection, the second inservice visual inspection of that system shall be performed at the first refueling outage. Otherwise, subsequent visual inspections of a given system shall be performed in accordance with the following schedule:

No. of Inoperable Snubbers of Each Subsequent Visual Tyoe on Any System cer Inscection Period Insoection Period *# 0 18 months 25% 1 12 months 25% 2 6 months : 25% 3,4 124 days : 25% 5,6,7 62 days 25% 8 or more 31 days 25% 0

          *The inspection intery fo'r each type of snubber on a given system shall not be 1

lengthened more than onestep at a time unless a generic problem has been identi- ~ fied and corrected; in tha% event the inspection interval may be lengthened one step the first time and wwo steps thereafter if no inoperable snuboers of that type are found on that system.

          #The provisions of Specification 4.0.2 are not applicable.

WOLF CREEK - UNIT 1 3/4 7-19 _ _ _ m,- ,---,-s - --

               ,LANT       SYSTEMS                                                                                                                    hh      Yhh SURVEILLANCE REQUIREMENTS (Continued)

c. Visual Insoection Acceptance Criteria Visual inspections shall verify that: ( h re are no visible indi-cations of damage or impaired OPERABILI , r(2 attachments to the foundation or supporting structure are s ur , and (3) fasteners for attachment of the snubber to the component and to the snubber anchorage are. secure. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type on that system that may be generically susceptible; or (2) the affected snubber is func-tionally tested in the as-found condition and determined OPERABLE per Specification 4.7.8f. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers. For those snubbers common to more than one system, the OPERABILITY of such snubbers shal' oe considered in assessing the surveillance schedule for each of the r ated systems.

d. Transient Event Insoections An inspection shall be performed of all hydraulic and mechanical snub-bers_ attached to sections of systems that have experienced unexpected,

                              -potentially damaging transients as determined from a review of opera-tional data and a visual inspection of the systems within 6 months following such an event. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following:

(1) manually induced snubber movement; (2) evaluation of in place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.

e. Functional Tests During the first refueling shutdown and at least once per 18 months .

thereafter during shutdown, a representative sample of snubbers of each type shall be tested using one of the following sample plans. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the ' sample plan selected for each snubber type prior to the test period or the sample plan used in the prior test period shall be implemented:

1) At feast 10% of the total of each type of snubber shall be function-ally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.8f. , an additional 10% of that type of snubber shall be functionally tested until no more failures are found or until all snucbers of that" type have been functionally tested; or

2) A representative sample of ach type of snubber shall be func- -

                                         -tionally tested in accordance,with Figure 4.7-1.      ,
                                                                                                                                                      "C" is the WOLF CREEK - UNIT 1                                    3/4 7-20
         - - -       - - .          - . . . - - -   e  u-w- ,    --a-    r - ,,,- , , - - - -e-m--ww,.m-,w,---,,-            -w, e ,e- - - , - - ,,--      ,-   ,-     - , - - - - - - -

PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)

e. Functional Tests (Continued) -

total number of snubbers of a type found not meeting the accep-tance requirements of Specificat 4.7.8f. The cumulative number of snubbers of a type tes s denoted by "N". At the end of each day's, testing, the values of "N" and "C" (pre-vious day's total plus current day's increments) shall be plotted on the Figure 4.7-1. If at any time the poig falls in the " Reject" region, all snubbers of that g44, plotted shall be functionally tested. IfatanytimethepointplottfMallsin the " Accept" region, testing of snubbers of that t,fpb may be terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" regip

                                           " Reject" region, or all the snubbers of that tyye'4hdve                  pr the been tested. Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume-anew at a later time provided all snuceers tested with the. failed' equipment during the day of equipment failure J

_ are retested; or An initial representative sample of 55 snubcers shall be 3) functionally tested. For eacn snubber type wh.ich does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is e ual to the initial sample size multiplied by the factor,1 C/2 where "C" is the number ^ of snubbers found which do not me t the functional test acceptance criteria. The results from this sample plan shall be plotted using an " Accept" line which follows the equation N = 55(1 + C/2). Each .9nubber point should be plotted as soon as the snubber is tested. If the point plotted falls on or below the " Accept" line, testing of that type of snuboer may be terminated. If the point plotted falls above the .

                                          " Accept" line, testing must continue until the point falls in the " Accept" region or all the snubbers of that type have been tested.

The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reviewed before beginning t testing. The review shall ensure, as

                           ,     far as practicable, that t ey are, representative of the various con-figurations, operating env nd.ents, range of size, and capacity of snubbers of each type.          Sn bbept placed in the same location as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the samole plan.         If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional test results shall be reviewed at that time to deter-mine if additional samples should be limited to the type of snuceer which has failed the functional testing.

WOLF CREEK - UNIT 1 3/4 7-21

                                   -w--                              -    . - , . , n .- . . . ,         - - . , . -

PLANT SYSTEMS N.05 & M C0W SURVEILLANCE REQUIREMENTS (Continued)

f. Functional Test Accectance Criteria The snubber functional test shall verify that:

1) Activation (restraining action) is'~ achieved within the specified range in both tension and compression; ,

o< rvicase. re. .An AT 'ed

2) Snubber bleed rate fcr 4 1 a.li r fber: 1s present in both tension and compression, w,ithin the-specified range; and

3) Where required, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel.

Testing methods may.be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods. Serwd LW<- He=%ri5 he

g. fmtt:=1 -Tem o il u. c man

                                             ~

An engineerin'g ev'a'l' u hion shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluatinn shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode. For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are l attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service. If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen-in place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be func-tionally tested. This testing requirement shall be independent of the requirements stated in Specification 4.7.8e. for snubbers not meeting the functional test acceptance criteria.

h. Functional Testino of Reoaired and Reolaced Snubbers

! Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement l WOLF CREEK - UNIT 1 3/4 7-22

                                                                                          ~

PLANT SYSTEMS li i b S'URVEILLANCE REQUIREMENTS (Continued) snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.

1. Snubber Service Life Program The service life of hydraulic and mechanical snubbers shall be moni-tored to ensure that the service life is not exceeded between sur-veillance inspections. The maximum expected service life for varous seals, springs, and other critical parts shall be determined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history.

Critical parts shall be replaced so that the maximum service life will not be exceeded durin'g a period when the snubber is required to be OPERABLE. The parts replacements shall be documented and the docu-mentation shall be retained in accordance with Specification 6.10.2. l - l WOLF CREEK - UNIT 1 3/4 7-23

                                                                               ,S

( , PROOF & RBIEW COPY 10, 9 8 7 -- REJECT 6 QV

                                                                 .                           /          CONTINUE
                                                                                                                                                                            ~

3 l [

                                                                                  "                    TESTING                                                 [

2 , 3 ACCEPT F

                                                                                             }

O 10 20 30 40 50 60 70 80 90 100 N FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST I WOLF CREEK - UNIT 1 3/4 7-24

l

                                                                                       ~

PLANT SYSTEMS 3/4.7.9 SEAL.ED SOURCE CONTAMINATION ha kh([ l LIMITING CONDITION FOR OPERATION 3.7.9 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination. APPLICABILITY: At all times. ACTION:

a. 'With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and either:

1. Decontaminate and repair the sealed source, or

2. Dispose of the sealed source in accordance with Commission Regulations.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS l 4.7.9.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by the Commiscion or an Agreement State. ,

The test method shall have a detection sensitivity of at least 0.005 l microcurie per test sample. 4.7.9.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.

a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:

1) With a half-life greater than.30 days (excluding Hydrogen 3),

and ! 2) In any form other than gas. WOLF CREEK - UNIT 1 3/4 7-25

 =    _ . . .
                                                               - - - , e
                                                                                     -   ,g---v-

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate' indicating the last test date shall be tested prior to being placed into use; and

c. Startup sources and fissicie detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed ire the core and following repair or maintenance to the source.

4.7.9.3 Reports - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the prescace of greater than or equal to 0.005 microcurie of removable contamination. . l WOLF CREEK - UNIT 1 3/4 7-26 w - - w - ee--.s-- ----- -- ,.-a w,. , . - . . ---.,w-v.---,,,-.---yem-- c- -- ---e- - - - -- -- -

(

                                                                                              ~
     }  .

PLANT SYSTEMS hhh f i 3/4.7.10 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.10.1 The Fire Suppression Water System shall be OPERABLE with: l

a. 3eco At least two fire suppression pumps, each with a capacity of _>$660 gpm, with their discharge aligned to the fire tuppression header, and-

b. An OPERABLE flow path capable of taking suction from the Wolf Creek Generating Station cooling lake and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standoipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3.7.10.2 and 3.7.10.4 APPLICABILITY: At all times.

ACTION:

a. With one pump inoperable, restore the inoperable equipment to OPERABLE status within 7 days. The provisions of Specifications 3.0.3 and 3.0.4 are .not applicaole. .

b. With the Fire Suppression Water System otherwise inoperable establish a backup Fire Suppression Water System within 24 hours.

SURVEILLANCE REOUIREMENTS 4.7.10.1.1 The Fire Suppression Water System shall be demonstrated OPERARLE:

a. At least cnce per 7 days by verifying that lake level exceeds 1075 feet,

! b. At least once per 31 days on a STAGGERED TEST BASIS by starting 'th electric motor-driven pump and operating it for at least 15 minutes on recirculation flow,

c. At least once per 31 days by verifying that each valve (manual, power-operated, or autematic) in the flow path is in its correct position, '

WOLF CREEK - UNIT 1 3/4 7-27

 ~'

b .

r. .

a Specification ' 3.7.10.1.a

           - Justification -
                '1he Technical Specf.fication capacity of the fire suppression ptanps has been^ revised to allow for ptrnp de-gradation. .. The revised value is the rated capacity of the
       .   - pimp. 'Ihis rated capacity still exceeds the design requirements for the system.-

m Y O e e e u J w

l Specification- 3.7.10.1 (Action a) Justification - The wrds "and/ce one water supply" have been deleted fran Action a. '1his was done because the Wolf Creek fire suppression eter system has only one mter supply (the Elf Creek cooling lake). 9 l 6 4 4 e O

               , , . . , . , . ,   ,,   ,.  .--.--r r -- ,- - - - - - - - -     v--   - - , ~ -v-- - ~ ~ - ~ - - --- = - . ~ - - - - - - -r'- -- ~ ~ - ' - ~ ~

1 PLANT SYSTEMS SURVEILLANCE REGUIREMENTS (Continued) h ffh hk l I l d. At least once per 6 months by performance of a yard loop and fire hydrant flush,

e. At least once per 12 months by cycling each testable valve in the '

flow path through at leact one complete 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 oper.ating-sequence._.and-l ', Keny'n;  : = ; m u c h ea*=*t i - *

                                                                                 **=-% - po w % w ete s ',
            %'q 1; ;O ::r ::t 7 Q :r.,

w.- coo i Verifying that each oum op .- . . .t Jeast $tt:t) gpm at a system m a = x , rus ~p dev tu , gM Cycling each valve in the flow path that is not testable during plant operation through at least one co=plete cycle of full travel, and b 3 41 Verifying that each fire suppression pumo starts (secuentially) on decreasing pressure in the fire suppression header at a

                  ,;        header pressure greater than or equal to 80 psig.

g. At least once per -3 years by performing a flow test of the system in accordance with Chanter 5, Section 11 of the Fire Protection Hancocok, 14th Edition, published by the National Fire Protection Association.

4.7.10.1.2 Each fire pumo diesel engine shall be demonstrated CPERABLE:

a. At least once per 31 days by verifying: #

zoo

1) The fuel storage tank contains at least 17; gallons of fuel, and

2) The diesel start:s 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 die:a1 fuel from 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

c. At least once per 18 months, during shutdown, by subj2cting the diesel n an inspection in acco* dance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

WOLF CREEK - UNIT 1 3/4 7-28

Specification- 4.7.10.1.1.f.1 Justification - Se Technical Specification capacity and pressure of the fire suppression punps has been revised to allow for punp degradation. he revised values are the rated capacity and pressure of the punps. W e values still exceed the design requirements of the system. e 9 e i; I l - 1 r

    *k
    ]

                            ' Specification '4.7.10.1.2.a 1)
                          - Justification -
                                                                                                                                                             )
                                  'Ihe voltme of fuel in the fuel storage tank has been revised to acmunt for the 5% stnp of tnusable fuel in the
 ~

j

                            . tank.

e 4 O 4 e I f e

  *t     ., .- pr- . <m..g..          r.. . ,   ,_ . _ , r, _ - . . . .._ . . . .r. . , , , , ...-..,-r,,   _ ,...,.-yw,r.,,w,-,r,y ~...mm ..- . ,,r.,. -e.

PLANT SYSTEMS 007 & .EE.W COPY SURVEILLANCE REOUIREMENTS (Continued) 4.7.10.1.3 Each fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1

1) The electrolyte level of each battery is above the plates, and

2) The overall battery voltage is greater than or equal to 24 volts.

b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and

c. At least once per 18 months, by verifying that:

1) The batteries, cell plates, and battery racks show no visual

           ,               indication. of ph'ysical damage or abnormal deterioration, and

2) The battery-to-battery and termir el connections are clean, tight, free of corrosion, and coa.ed with anticorrosion material.

WOLF CREEK - UNIT 1 3/4 7-29

                                ,--  --,-g- ,-, , ,--. ,          a,._.,--- -- ,-       - , - ,     , _, , , - - - - 4

l PLANT SYSTEMS SPRAY AND/OR SPRINKLER SYSTEMS kb b.iI LIMITING CONDITION FOR OPERATION 3.7.10.2 The following Spray and/or Sprinkler Systems shall be OPERABLE:

a. Wet Pioe Sorinkler Systems Building Elevation Area Protected Auxiliary 2000/2016 North Electric Cable Chase Auxiliary 1988/2000/un South Electric Cable Chase Control 1974 - 2073 Vertical Electrical Chases Control Control 1974 1992 Pipe Space and Tank Room Cable Area Above Access Controi 4 l

[../ j

b. 1re-Action Sorinkler Systems

                                                                                                       ]/

Building Elevation Area Protected Auxiliary 1974 Cable Trays 4 Auxiliary 2000 Cable Trays & Auxiliary 2026 . Cable Trays & Control 2032 Lower Cable Spreading Room - Control 2073 Upper Cable Penetration Area Reactor 2026 North Cable Penetration Area Reactor 2026 South Cable Penetration Area Diesel Gen. (E) 2000 East Diesel Generator Room Di'esel Gen. (W) 2000 West Diesel Generator Room

c. Water Sorays Systems Building Elevation Area Protected Auxiliary 2000 Auxiliary Feedwater Pump Turbine IN Y.UUZ Ud7 DI1*! UI.*

APPLICABILITY: Whenever equipment protected by the Spray / Sprinkler System is required to ce OPERABLE. ACTION: ,

a. With one or more of the above required Spray and/or Sprinkler Systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.10.2 Each of the above required Spray and/or Sprinkler Systems shall be demonstrated OPERABLE:

   + Areas         ccdain (dundani- sysfes                  of COP'                          '"
                                                                                                              ~ '

WOLF CREEX - UNIT 1 3/4 7-30 l 1 J

PLANT SYSTEMS fh,'[ h f M yf (Off SURVEILLANCE REQUIREMENTS (Continued)

a. At least once per 31 days, by verifying that each valve (manual, power- l operated, or automatic) in the flow path is in its correct position,

b. At least once per 12 months, by cycling each testable valve in the flow path through at least one complete cycle of full travel,

c. At least once per 18 months:

1) By performing a system functional test which includes simulated automatic actuation of the system, and:

a) Verifyim that the automatic valves in the flow path actuate 50 their correct positions on a Simulated Fire test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

2) By a visual inspection of the dry pipe spray and sprinkler

• neaders to verify their integrity, and

3) By a visual inspection of each nozzle's spray area.to verify the spray pattern is not obstructed. j b'

d. At least once per 3 years by performing an air.[ flow test through each open head spray / sprinkler header and verifying each open head spray /sorinkler nozzle is unobstructed.

WOLF CREEK - UNIT 1 3/4 7-31 I . . . _ .

                       .   - . -    .         .     - _ .         .-. .=          . - .. --

PLANT SYSTEMS (( { g] [] HALON SYSTEMS -

• LIMITING CCNDITION FOR OPERATION -

3.7.10.3 The following Halon Systems shall be OPERABLE: Building Elevation Area Protected Auxiliary 2026 North Electrical Penetration Room Auxiliary 2026 South Electrical Penetration Room Auxiliary 2026 Lead Center and M. G. Sets Room

• Control 2000 ESF Switchgear Rooms * -

Control 2016 Switchgear Rooms - Control 2047 Control Room Cable Trenches and Chases APPLICABILITY: Whenever equ'pmerit protected by the Halon System is required to De OPERABLE. ACTION:

a. With one or more of the acove required Halon systems' inoperable, within 1 hour establish a continuous fire watch with cackup fire

       #                   suppression equipment for those areas in which redundant systems or
     -                     components could be damaged; for other areas, estaolish an hourly fire watch patrol.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not acplicable.

SURVEILLANCE REQUIREMENTS 4.7.10.3 Each of the above requireo Halon Systems shall be demonstrated OPERABLE:

                                                                                                                  /

a. At least once per 31 days, by verifying that each valve (manua ,

power-operated, or automatic) in the flow path is in its correct

                              . : thw  -.                                                               -

A-g At least once per' 6 m'o nths'by verifying Halon storage tank weight

                ~         (or level) to be at least 95% of full charge weignt and pressure to be at least 90% of full charge pressure, and V       At-least once per 18 montns by [ .          .
                                                                                                                      ~'

gerifyingthesystem,includingassociatedVentilationSystem , c.. fire dampers and fire door release mechanisms, actuates manually and automatically, upon receipt of a simulated actuation signal,

                                                                                                          ^~...

C -F:ri ..uancarc P6-flew .eu Larougii fieddics and narrhwo assON F.o P GG KdSt:.- , ,,

                                  '~               - -
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           ' $ h ryg3       {eglgh        reb    A     q5iW         &

Mfh @ 6

                                                                                                       .-~....--_./

WOLF CREEK - UNIT 1 3/4 7-32

PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.10.4 The fire hose stations given in Table 3.7-5 shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERA 8LE. ACTION: a. With one or more of the fire hose stations shown in Table 3.7-5 inoperable, provide equivalent capacity backup hose protection to the unprotected area from the spare hose connection on the adjacent OPERABLE standpipe. If two standpipe hose connections are not available at the adjacent OPERABLE hose station (s), provide gated wye (s) to ensure continued OPERABILITY of the affected hose station. Where it can be demonstrated that the physical routing of the backup hose would result in a recognizable hazard to operating technicians, plant equipment, or the hose itself, or would require the blocking open of a fire door, the hose shall be stored at the point of origin and properly identified as to its intended use. The above action shall be accomplished within 1 hour if the inoperable fire hose is the primary means nf fire suppression; otherwise route the additional hose within 24 hours. *

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.10.4 Each of the fire hose stations given in Table'3.7-5 shall be dc. nstrated OPERABLE:

a. At least once per 31 days, by a visual inspection of the fire hose stations accessible during plant operations to assure all required equipment is at the station,

b. At least once per 18 months, by:

1) Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station,

2) Removing the hose for inspection and reracking, and

3) Inspecting al! gaskets and replacing any degraded gaskets in i the couplings.

c. At least once per 3 years, by: )

1) Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and

2) Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, ,

whichever is greater. l WOLF CREEK - UNIT 1 3/4 7-33 V 9

                                                                                ?RODF & REWEW COPY TABLE 3.7-5 FIRE HOSE STATIONS BUILDING           ELEVATION                              AREA          HOSE RACK Auxil'fary         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 Auxiliary 2000 1329 KC-HR-111

        -Auxiliary          2000                                  1320          KC-HR-048 Auxiliary          2000                                  1320          KC-HR-046#

Auxiliary 2000 - - 1314 KC-HR-030 Auxiliary *2000 1321 KC-HR-029# Auxiliary 2000 1301 KC-HR-035# Auxiliary 2000 1301 KC-HR-039 Auxiliary ~ ~ 2000 1301 KC-Hf'-041# 0 J u, ,ne N 4322 n.', CL Auxiliary --

                          -2026            _ _ . _ . .      _.1408             KC-HR-049 Auxiliary         2026                                   1408         KC-HR-044 Auxiliary         2026'                                  1408         KC-HR-032#

Auxiliary 2026 1408 KC-HR-025# Auxiliary 2026 1401 KC-HR-034 Auxiliary 2026 14C3 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 Auxiliary 2047 1504 KC-HR-033 Auxiliary 2047 1502 KC-HR-027 Auxiliary 2064 1119 KC-HR-028# Control 1974 3101 KC-HR-002# 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 WGLF CREEK - UNIT 1 3/4 7-34

TABLE 3.7-4 (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-129*

Reactor 2026 N.A. KC-HR-121* Reactor 2026 N.A. KC-HR-132*# Reactor 2026, N.A. KC-HR-125* 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 Feel 2026 6203 KC-HR-052 Fuel 2047 6301 KC-HR-055# Fuel 2047 6302 KC-HR-056# I Fuel 2047 6301 KC-HR-053# - ESW 2000 N.A. KC-HR-140 ESW 2000 N.A. KC-HR-141 ! TABLE NOTATIONS

     # Secondary means of fire suppression to Water Sprays / Deluge or Halon System.

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

I CALLAWAY - UNIT 1 3/4 7-35

t. .

PLANT SYSTEMS 3/4.7.11 FIRE BARRIER PENETRATIONS h LIMITING CONDITION FOR OPERATION 3.7.11 All fire barrier pene'trations (i t ; walls, floor / ceilings, cable -.

                                                                                                                 ~

tray enclosures, and other fire barriers) separating safety related fire areas or separating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire-rated assembly penetrations (fire doors, fire windows, fire dampers, cable, piping, and ventilation duct penetra-tion seals) shall be OPERABLE. APPLICABILITY: At all times. . - . . . . . _ ACTION: a.

                                                                          ' Q;y              gh With one or more of the above required fire rated use.m i 6 ond-                          l bg da ico inoperable, within 1 hour establish a continuous fire watch on hgPERABILITY of atleastonesideoftheaffectedpenetration,orverig'(msyandestab-                   .

lish an hourly fire n< watch patrol, on._inct il ; to,c;ry mencMectible -

                 ' .sa= 1 m n + m e a 41      2,  n.a c t an an"=1 " r: rca . nar.ca ca ny co- tne - moper-dl 0 2000"ilj Oiid acalany      ucvsua anu es c :ui i sen da nuuriy Tl re wdw.6 rMrO l".

b. The provisions of Specifications 3.0.3 and 3.0far~e~not applicaole.

SURVEILLANCERE0bIREMENTS rde4 d * *c5 4.7.11.1 At least once per 18 months the above required fire Mmer p'enetu-twns. a visualandinspection penetrationof: sealing devices shall be verified OPERABLE by performing

a. The exposed surfaces of each fire rated assembly, [

b. Each fire window / fire damper and associated hardware [od

c.  ! yawa .A +4 x At least 10% of each type Aof sealed-penetration. If acparent changes _

in appearance or abnormal' degradations are found, a visual inspection of an additional 10% of each type of sMed penetration shall be made. This inspection process shall continue until a 10% samole with no apparent changes in accearance or abnormal degradation is found. Samples shall be selected such that each psnetration seal will be inspected every 15 years. . 4.7.11.2 Each of the above required fire doors shall bc verified OPERABLE by inspecting the automatic hold-open, release and closing mechanism and latches at least once per 6 months, and by verifying:

a. The OPERABILITY of the Fire Door Supervision System for each electrically supervised fire door by performing a TRIP ACTUATING DEVICE CPERATIONAL TEST at least once per 31 days,

b. A > eu.4 9 4.w.a That each locked closed fire door /is closed at least once per 7 days,

c. That doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours and perf orming a functional test at least once per 18 months, and ,

d. That each unlocked fire door without electrical supervision is closed at least once per 24 hours.

WOLF CREEK - UNIT 1 3/4 7-26

PLANT SYSTEMS l am ll AL 3b ' COM 3/4.7.12 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION ~- b & Y gaaelN Gv ews 3.7.12...The~ temperature f.each area _. ts,- tm_ >< 4...a- 4,..s in. Table 3.7-6 shall/be xi..;a; -

                                       ,_ T:es;-:. 7 S.                                                - - - _.

tb F hears or by m.<s N ~5c T. APPLICA8ILITYi'~Whenever the equipment in an"affected area is required to be OPERABLE. ACTION:

a. With one or more areas exceeding the temperature limit (s) shown in Table 3.7-6 for more than 8 hours, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s)

         ,        and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specifications 3.0.3 ano 3.0.4 are not appif cable.

b. With one or more areas exceeding the temperature limit (s) shown in Tabl.e 3.7-6 by more than 30*F, prepare and submit a Special Report as required by ACTION a. above, and within 4 hours either restore the area (s) to within the temperature limit (s) or declare the equipment in the affected area (s) inoperable.

SURVEILLANCE REOUIREMENTS 4.7.12 The temperature in eacn of the areas shown in Table 3.7-6 shall be determined to be within its limit at least once per 12 hours. C e WOLF CREEK - UNIT 1 3/4 7-37

l l l I PR00F & REIN CD)Y TABLE 3.7-6 AREA TEMPERATURE MONITORING i MAXIMUM TEMPERATURE AREA LIMIT (*F)

1. ESW Pump Room A 119

2. ESW Pump Room B 119

3. Auxiliary Feedwater Pump Room A 119

4. Auxiliary Feedwater Pump Room B 119

5. Turbine Driven Auxiliary Feedwater Pump Room 147

6. ESF Switchgear Room I 87

7. ESF Switchgear Room II 87

8. RHR Pump Room A 119

9. RHR Pump Room B , 119 l

10. CTMT Spray Pump Room A 119

11. CTMT Spray Pump Room B 119

12. Safety Injection Pump Room A 119

13. Safety Injection Pump Room B 119

14. Centrifugal Charging Fump Room A 119

15. Centrifugal Charging Pump Room B 119

16. Electrical Penetration Room A 101

17. Electrical Penetration Room B 101

18. Component Cooling Water Room A 119

19. Component Cooling Water Rocq B 119

20. Diesel Generator Room A 119

21. Diesel Generator Room B 119

22. Control Room 84 WOLF CREEK - UNIT 1 3/4 7-38

l l 3/4.8 ELECTRICAL POWER SYSTEMS { Qf 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the Onsite C1 ass 1E Distribution System, and b.

Twoseparateandindegegdentdieselgenerators,eachwith:

1) A separate day /'containing a minimum volume of 390 gallons of fuel, _.

2) A separate Fuel 011 Storage System containing a minimum volume of 85,300 gallons of fuel, and

3) A separate fuel transfer pump.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

a. With either an offsito circuit or diesel generator of the above required A.C. elc-trical power sources inoperable, demonstrate the OPERABILITY f the remaining A.C. sources by performing Specificati'..is 4.8.1. 14. nd 4.8.1.1.2a.4) within 1 hour and at least once per 8:1our the after; restore at least two offsite circuits and two diesel 9 tors to OPERA 8LE status within 72 hours or be in a'.

least HOTSTANDBY within the next 6 hours and in COLD SHUTCOWN within the following 30 hours,

b. With one offsite circuit and one diesel generator of the above require C. electrical power sources inoperable, descastrate the OPERAB LITY of the remaining A.C. sources by performing Specifications 4.8.1. 14. and 4.8.1.1.2a.4) within 1 hour and at least once per 8 hour t 'eafter; restore at least one of the inoperable sources to OPER status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN wittin the following 30 hours. Restore at least two offsite circn. _, and two diesel generators to OPERABLE status within 72 hauts from the time of initial loss or be in at least HOT STANDBY u. thin the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

c. With one diesel generator inoperable in addition to ACTION a. or h.

above, verify thatt

1. All required systems, subsystems, trair,5,~ components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and 8

WOLF CREEK - UNIT 1 3/4 8-1 4

r ELECTR @ L POWER SYSTEMS PROOF & EH COPY LIMITING CONDITION FOR OPERATION ACTION / Continued)

2. When in MODE 1, 2, or 3, the steam-driven auxiliary feedwater pump is OPERA 8LE.

If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Specification 4.8.1.1.2a.4) within 1 hour and at least once per 8 hours thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sour:es to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. Wi'h t only one offsite source restored, restore at least two offsite cirguits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

e. With two of the above required diesel gencrators inoperable, demonstrate the OPERABILI f two offsite A.C. circuits by perform-ing Specification 4.8.1.1 within 1 hour and at least once per 8 hours thereafter; resto t least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HJT STANDBY witnin the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two diesel generators to OPERABLE status within 72 hours from time of initial loss or be in least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REOUIREMENTS 4.8.1.1.1 Each of the above required indepen(ent circuits between the offsite transmission network and the Onsite Class 1E Distribution System shall be(' a$' v D'etermined OPERABLE at least once per 7 days by verifying correct breaker alignmentsifindicated power availability,-end

        -b. Camcastrated CPECABLE at least caca pw-48-montas-durfogshutd'own-by transferrirrg-manually unit pcwar supply frs; the normal :ircuit to-the alternate circei4r.

4.d.1.1.2 Each diesel generator shall be demonstrated OPERABLE: WOLF CREEK - UNIT 1 3/4 3-2 a

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continuedi

a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:

1) Verifying the fuel level in the day tank, I

2) Verifying the fuel level in the fuel storage tank,

3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day tank, nj

                                                                                        ,y                                                          no

4) Verifyingthedipelstartsfromambientco itionan[ccelerates to at least WHrpm in less than or equal o 12 secop'ds

• he generator voltage and frequeppshall bek4MG + 409Fvolts and 60 + 1.2 riz 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 syn:hronized, loaded to gr ater than O
         /                                                   or equal to 6201 kW in less than or equal to 60 seco ds? 'cderates with a load greater than or equal to 6201 kW for at east 60 minutes, and 4

6) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

l b. At least once per 31 days and after each operation of the diesel

               ~
                   '_                                vNre the period of operation was great 3r than or equal to 1 hour by c.hecking for and removing accumulated water from the day tanks;
                                       'n..
                                         ' c.        At' least once per 31 days by checking for and removing accumulated n              . --          ..         : water from the fuel oil storage tanks;
                    'am              -                     .i-         s~                      . , , .      -

east once per 92 dcys and from new fuel oil prior to its#, addit o he storage tanks by verifying that a sample'oEtained V- ' in accordance w TM-0270-1975 meets thejf 14cning minimum requirements in accorda ith the tesis-specified in ASTM-0975-1977:

                                      .         b

1) A water and sediment c tent of than or eiual l to 0.05

g. , g*

• volume percent;

2) A kin

• c visocity at 40*C of greater thatn or equal to

                                                                 - centistokes, but less than or equal to 4.1 centisEBkes; S%                ins k 41ls, e                      p                            .

WOLF CREEK - UNIT 1 3/4 8-3 b q 9

                                                                          .3                ,          - . . _ . _         _    , . . , _ _  -

e I

            , INSERT
       ~
         /

t

           *These diesel generator starts from ambient conditions shall be perfon starts for the purpose of this surveillance testing 2 an engine prelube period and/or oth
          .the manufacturer so that engine is minimized. the mechanical stress and wear on the dieseler w 4

e 0 e e 9 h

inwt A 4.3.1.1.2.d. By sampling new fuel oil in accordance with ASTM D4057 prior to addition to storage tanks and: (1) By verifying in accordance with the tests specified in ASTM 0975-81 prior to addition to the storage tanks that the sample has: (a) An API Gravity of within 0.3 degrees at 60% or a specific gravityofwithin0.0016at60/60%,whencomparedtothe , supplier's certificate or an absolute specific gravity at 60/60% 6f 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 degreds but less than or' equal to 39 degrees. (b) A kinematic viscosity at 40'C of greater than or equal to 1.9 centistokes, but less than or equal to-4.1 centistokes. if gravity was not determined by comparison with the supplier's certification. (c)Aflashpoint'equaltoorgreaterthan125%.and . (d) A clear and bright appearance with proper color when tested in accordance with ASTM 04176-82. (2) By verifyingwithin 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 D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 l or ASTM D2622-82. l i

e. At least once every 31 days by obtaining a le of fuel oil in accord-ance with ASTM D2276-78, andverifyingthItparticulatecontamina' tion is less than 10 mg/ liter when checked,in accordance with ASTM D2276-78; W A.

V

             ^

y . ELECTRICAL POWER SYSTEMS j[h'{Mhhh SURVEILLANCE REOUIREMENTS (Continued) A specific gravity as specified by the manufacturer at 60 60 F ~ reater than or equal to 0.83 but less than or e to 0.8 an API gravity at 60'F of greater than equal to 27 degree at less than or equal to 39 de s; -

4) An impurity level less than 2 of insolubles per 100 ml when tested in accorda wit .-02274-70; analysis shall be completed within 7 days er obtaining the sample but may be performed after the ad ion o fuel oil; and

5) The other proper es specified in Table ASTM-0975- 1977 and Regulato guide 1.137,' Revision 1, 1979, Position .a., when tested in accordanc/,0ctob with A -0975-1977; analy ' shall be completed within 14 days after obt ing the sa e but may be performed after the addition of new f 0el-11.

I.r f. At least once pe' r 18 months, during shutdown, by: a 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service, 2)

                                                     /

Verifying the 4enerator capability to reject a load of greater than or'eq_ual to kW (ESW pump) while maintaining voltage at 4MB + Mfr volts and frequency at 60 + 14 Hz, 4ce@ 32+'

                                                        ~

3)

                                                                                      ~ft Verifying't'he diesel generator capability to reject a load of 6201 kW without tripping. The generator. voltage shall not exceed 4784 volts during and following.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 wnile its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency ~of the emergency busses shall be maintained at 4t90 + 499 volts and 60 + 1.2 Hz during this test.. % ~- 324

5) Verifying that on a Safety Injection test . signal without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to -

5 minutes r The generator voltage and frequency shall be N ; 6.a % c h tt . . . . . .% a ,u . 3. m %e < h " M WOLF CREEK - UNIT 1 3'/4"8'- T' l._

F-1 ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) N 30 volts and 60 1 1.2 Hz within 12 seconds after the auto-start signal; the generator steady-state generator voltage and frequency shall be maintained within these limits during this test;

6) Simulating a loss-of-offsite power in conjunct;on with a Safety 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 througn the LOCA sequencer ano Y A5 m epitsmgeneratortcr: for greater than or equal ,to 5 minutes whiles is loaded with emergenity loads. After energization, the steady-state voltage and frequency-ofs the emergency busses shall be maintained at 6 + 4e6 volts and 60 1 1.2 Hz during this test; and           & ~ 3%
  ,                           c)     Verifying that all automatic diesel generator trips, except 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 b.us. concurrent with a Safety Injection Actuation signal.

Y3 T SW '

7) Verifying the diesel generator oper .es for a least 24 hours.

During the first 2 hours of this st, the d'esel generator shall be loaded to greater than equal to 6821.kW and during the remaining 22 hours of this st, the d se1' generator shall be loaded to greater than or' eq al to 620 kW. The gener tor N voltage and frequency shall be 1 *f0 volts and 60 1. 2 We,-3 6.] _, wi. thin 12 seconds after the start signal; the steady s Etem N generatorvoltage and frequency shall be maintained within 4ccc h28 4 n= t nHa C :: 't: during this test. ~ Within 5 minutes a 'ter como7Et:Wg- ~~~

           ~               -this-24-hotir test, perform Specification 4.8.1.1.2ii.6)b)*;

8) Verifying that the auto-connected loads to each diesel generator do not exceed 6635 kW;

9) Verifying the diesel generator's capability to:

a) Synchronize with the offsite power scurce while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, "If Specification 4.8.1.1.2k.6)b) is not satisfa:torily completed, it is not necessary to repeat *the preceding 24-hour test. Instead, tne diesel generator may be operated at 6201 kW for 1 nour or until operating temperature has stabilized. WOLF CREEK - UNIT 1 3/4 8-5

L . ELECTRICAL POWER SYSTEMS b SURVEILLANCE REOUIREMENTS (Continued) b) Transfer its loads to the offsite power source, and _ c) Be restored to its standby status.

10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal overrides the test mod? by: (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offsite power;

11) Verifying that the fuel transfer pump transfers fuel from each fual storage tank to the day tank of each diesel via the installed cross-connection lines; and

12) Verifying that the automatic LOCA and Shutdown sequence timer is OPERABLE with the interval between eacn load block within 10*.' of its design interval.

                                                                   ^
           ~                           g T.      At least once per 10 years or after any modifications wnica could
                                                . affect diesel generator interdependence by starting botn diesal 3

generators simultaneously, during shutdown, and verifying that coth diesel generators accelerate to at least Gi-rpm in less than or equal to JF seconds; and 6l* IF v ( g. At least once per 10 years by:

1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, and g Q< *- ,

2) Performing a p(ressure test of those portions of the diesel fuel oil system designed to Section III, subsection N0 of the ASME Code at a test _p,ressure equal to 110% of the system desig (

                                                     / PIessure
                                                         - ;               g,                           . g,% 3g    3' 4.8.1.1.3       Re
                                                            - All diesel generator failures, valid or g n valid, shall' be reported Ao tne Commiss.f on. pursuant to Specification 6.9.2 Repo rts-o f diesel gener3 toc failaris shall include the information recommended ~in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If tne number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be suoplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory luide 1.108, Revision 1, August 1977.
                                   .JLF CREEK - UNIT 1                        3/4 8-6

f h i TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN LAST 100 VALID TESTS

• TEST FREOUENCY 11 At least once per 31 days 2 At least once per 14 days 3 At least once per 7 days

                                     >4                                  At least once per 3 days

• Criteria for determining r. umber of failures and numoer of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the last 100 tests are determined on a per nuclear unit basis. For 3he purposes-of-this-test-schedule,-or.ly... valid tests-condt.md IftTFttie CL ...r.: d_t: -':t-- _:ruee .e G.E ' w

     ' c :- r L . . w. . v .7 .. i e   e a s i. Avu vaiiu Les ea.       ".i.y
                                                                        .  -          ...v .. i a -sa+   a wi Me s M l '. w d matN a i - i.iic .a A u ay -w s.1 '-^ T '^ ^ ^y g gr*[le b M cf O&                 (4 f/A NC"N iest m *rementJ                                                                                                  .

corngd b d the. ef " %J% Gide, b+ t- loo %d k in f A.) 4 he,

           ~ . . _            __.

i.scgfs,"3 hat fes ,

                                                                             \
                                                                              ,n..s._ t, y m ,'.

WOLF CREEK - UNIT 1 3/4 8-7 n---- n - .--v - , - , a.,. -

ELECTRICAL POWER SYSTEMS N00f Pa FiP# COM A.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION

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

a. One circuit between the offsite transmission network and the Onsite 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 gal.lons o~f fuel, and

3) A fuel transfer pump.

APPLICABILITY: MODES 5 and 6. ACTION: 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, and within a hcur:, deprc: urire and vent the Reactor-Occlent Systeir, thicugh at least a 2 squarc inch--vente 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 correctiv'e action to restore the required sources to OPERABLE status as soon as poss.ble. SURVEILLANCE REOUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPE' TABLE 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 8-8

l 1 ,. ELECTRICAL POWER SYSTEMS i

'l 3/4.8.2 0.C. SOURCES .

11 gy OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:

a. 125-Volt Battery Bank NK11 and NK13, and its associated Full Capacity i Chargers NK21 and NK23, eranA

b. 125-Volt Battery Bank NK12 and NK14, and its associated Full Capacity Chargers NK22 and NK24.

APPLICABILITY: MODES 1, 2, 3, and 4. $d i ACTION: r ch

a. With one of the required ery banks inoperable, restore the inoperable bat.t'ery banks o OPERABLE status witnin 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

b. one of the required full capacity chargers ino ,-des ~nstrate the OP fts associated bat +=r, y performing Specification 4.8.2.

• 1 hour, and at least once per 8 hours thereaft any Category A limi 4.8-2 is not met, d e the battery inoperable.

     .         SURVEILLANCE REQUIREMEdTS 4.8.2.1 Each 125-volt battery bank and chargtr shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1) The parameters in Table 4.8-2 meet the Category A limits, and p 2) The total battery terminal voltage is greater than or equal to t

4de-volts on float charge. - I so. L l i e WOLF CREEK - UNIT 1 3/4 8-9

ELECTRICAL POWER SYSTEMS OF & Ell:W COM . SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:

1) The parameters in Table 4.8-2 meet the Category B limits, p.us - wu 2)

Therei[connno the visible corrosion at either terminals or connectors, tion resistance of these items is less than 50 x 10 s ohm, and

3) The average electrolyte temperature of at least every sixth cell is above 60*F.

c. At least once per 18 months by verifying that:

1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal ceterioration,

2) The cell-to-cell and terminal connections are clean, tight, and

        ,                    coated with anti-corrosion material, 3). The resistance of each pell-to-celJ and terminal connection is less than or equal to g150 x 10 8) ohm, and j                      4)    .The battery charger will supply at least 300 amperes at i

' MH!b-volts for at least 1 hour.

d. At le % ,1 ast once per 18 months, during shutdown, by verifying that the g g battery capacity is adequate to supply and maintain in OPERABLE /" d

 ,         g M ail cf the rtu:! emergency loads for 200 minute W4 hen the g           battery is subject to a battery service test;                    ,,3 ,,u ,.4 e.

duringverifying that the At least battery onceis at capacity per 6080% least months, of the manu shutdownb[cturer's rat subjectedtoaperformancedischargetest.(Thisperformancedischarge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d. ; and f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battary that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating. i l WOLF CREEK - UNIT 1 3/4 8-10

PRODF & REEW COPY TABLE 4.8-2 BATTERY SURVEILLANCE REOUIREMENTS CATEGORY A(1) CATEGORY B(2) PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) DE"IGNATED PILOT CONNECTED CELL VALUE FOR EACH L2LL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates., and < h" above and < " above and not maximum level maximum level overflowing ind,1 cation mark indication mark L3 2.it 2' 7 Float Voltage > -a=*J vol ts > e-M volts (6) > 1r{4 volts

                                               ~
                                                                                                        -Not more than o.ow 4-99tr below the
      '                                                                                                  average of all

_ > 1.195 - connected cells 240 i Specifi Gravity {4) 3 W (5) Average of all Average of all connected cells connecta

                                                                 > 1.205                                 >1.195'g) cells l

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 allowable values, and provided all Category A and B parameter (s) are restored to within limits - within the next 6 days. (2) For any Category 8 parameter (s) outside che 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 8 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

                                                          , , ,  ,p---,     - - , _ -

ELECTRICAL POWER SYSTEMS (1 g {iging g ] D.C. SOURCES SHUTOOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, the following 0.C. electrical sources shall be OPERABLE:

a. 125-Volt Battery Bank NK11 and NK13, and its associated full capaci+y Chargers NK21' and NK23, or

b. 125-Volt Battery Bank NK12 and NK14, and its associated full capacity Chargers NK22 and NK24.

APPLICABILITY: MODES 5 and 6. ACTION:

• A' #

16 With t.he required battery ba noperable, immediately susoend all operations involving CORE A ERATIONS, positive reactivity changes or movement of irradiated uel; initiate corrective action to restore the required battery bank to OPERABLE status as soon as possible,

nd within C h;ur:, depre;;urize and vent th: R c;t:r CC icnt Syster thr; ugh t I:::t : 2 :qu:re 'nch 'fent.

l b_ With the reouired full-capacity char e, demonstrate the OPERABILITY of its associ '

                                                      , bank b performing Specification 4.8.2.la.1) wit         our, and at least once pe            thereafter.

If any ory A limit in Table 4.8-2 is not met, declare . ttery_ rable. SURVEILLANCE REOUIREMENTS 4.8.2.2 The above required 125-volt battery, banks and associated chargers shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1. WOLF CREEK - UNIT 1 3/4 8-12

i i, ELECTRICAL POWER SYSTEMS It 3/4.8.3 ONSITE POWER DISTRIBUTION [ OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the specified manner with tie breakers open (trtretry between redundant busses within the unit ("d b;ti;;.. _r.i t: .t t' = "- at i c r.h L a. . Division #1 A.C.. Emergency Busses consistin f 1 4160-Volt Emergency Bus #N801, and s

                             ! t

2) N

                                                       -Volt Emergency Bus #NG01, NG03 and NG0c".
                          .         b.      Division #2        c. Emergency Busses consisti      of:                     (

hh 1) 4160-Volt Eme cy Bus #N802, a p .

2) 480-Volt Emergenc i <

c. us#N)02 . 04.and NG06E.

to A I 120-Volt A.C. Vital Busses ' I and NK03 energized from their h<*D associated inverter conne ed o C. Busses # hWd and -Me./,

                                                                                           #       M cs         uros 3

d. 120-Volt A.C. Vital sses # 02 and i energized from their associated inver* connected to 0.C. Buss e.

                                                                                                 #.g   and g ,

125-Volt D . Busses NK01 and NK03 energized fro, tteries NK11

       ,                                   and NK      , and N

f. -Volt 0.C. Busses NK02 and NK04 energized from Batterie ' 12 s ~and~NK14.

o - APPLICABILITY: M00ES 172;-3, and-4. ~ ~ ' ACTION:

a. With one of the required divisions of A.C. emergency busses not fully energized, reenergize the division within 8 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN witnin the following 30 hours.

b. With one A.C. vital bus eithar not energized from its associated inverter, or with the inverter not connected to its associated 0.C.

bus: (1) reenergize the A.C. vital bus within 2 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and (2) reenergize the A.C. vital bus from its associated inverter connected to its associated 0.C. bus within 24 hours or be in at least HOT STAND 8Y within the next s 6 hours and in COLD SHUTDOWN within the following 30 hours, g*g'

c. With one O.C. bus not energized from its associated battery bank [ y '*J" reenergize the O.C. bus from its associated battery bank 4itkin 2 hours ~

or be in at least HOT STANC8Y within the next 6 hours and in COLD SHUT 00WN within the followinq 30 hours. i i I o' WOLF CREEK - UNIT 1 3/4 S-13 i - c

i

        -acDivision #1'A.C.~ miiir~~gency k        80sses consisting'of:
                                                                     ^N      '

1) 4160-Volt Emergency Bus #N801, and '

2) 480-Volt Emergency Busses #NG01, NG03 and NG05E. ,

b. Division #2 A.C. Emergency Busses consisting of:

(

1) 4160-Volt Emergency Bus #N802, and i'

2) 430-Volt Emergency Busses #NG02, NG04 and NG06E.

c. 120-Volta.C.VitalBus#NN01energizedfromItsassociatedinverter connected to D.C. Bus #NK01,

d. 120-Volt A.C. Vital Bus #NN02 energized from its associated inverter connected to D.C. Bus #NK02, l

l

e. 120-Volt A.C. Vital Bus #NNO3 energized from its associated inverter connected to 0.C. Bus #NK03,

f. 120-Volt A.C. Vital Bus #NN04 energized from its associated inverter connected to D.C. Bus #NK04,

g. 125-Volt 0.C. Sus #NK01 energized from Battery #NK11M Cy

• Nl,

h. 125-Volt 0.C. Bus #NK0x energized from Battery #NK12 M C % d ur.t, 1.

125-Volt D.C. Bus #NK03 energized from Battery #NK13. and cup f *ti j.

      ,       125-Volt D.C. Bus #NK04 energized from Battery #NK14Ai Cy '#

l O a 4

ELECTRICAL POWER SYSTEMS RODF S RDA.:W CFI SURVEILLANCE REQUIREMENTS 4.3.3.1 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker ~ alignment and indicated voltage on the busses. I WOLF CREEK - UNIT 1 3/4 8-14

ELECTRICAL POWER SYSTEMS (OT & E !.Y1C0F ONSITE POWER OISTRIBUTION

   .                      SHUTOOWN LIMITING CONDITION FOR OPERATION jn                p.o or 3.8.3.2 As a minimum,/f.he following e ectrical busses shall be energized
                , . in the specified manner t j

[ a.

                                                                                                                         \

t e division of A.C. emergency busses consisting of j l an -/olt A.C. emergency bus, s h[h ,

b. Two 120-volt A.C. vi energized from their associated

                   ,o.                 inverters con                to their res       'a 0.C. busses,.and a.4 '        c.             5-volt D.C. busses energized from their as             '

ted battery

                  ,                    banks.
                     'w-APPLICABILITY: MODES 5 and 6.                               ..__._~~~----              -
                                                                          , ,3 ACTIONI'-                            "
                                                                       ,i#
     .,c
       ~                 Withfany[of'the above required electrical busses no.t energized in the required manner,-immediately suspend all operations involving CORE ALTERATIONS, positive l '    .[*"{p"[
        ,                reactivity changes, or movement of-irradiated fuel; initiate corrective action
                    .,_ to_ energize)the--required ch!TEMelrl busses in the specified manner.as 500n 25.

p.. . . , Q ;. .d i t. ; .. G oww. a ucp. meow.-.;m ...u ,e..; ..e 2CS un w3u a6 . c a a t.

                       ;. e: ;&. .        ...a . . . . . . -                                                          --

SURVEILLANCE REOUIREMENTS 4.8.3.2 The specified busses she.i be determined energized in the required. ' manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses. l l l l l WOLF CREEK - UNIT 1 3/4 8-15 1 I l

l l

a. Division 1, consisting of: ,

1) 4160-volt , emergency bus #NB01 and

2) 480-volt emergency busses #NG01, NG03, and NG05E, and ,

on 03 \'

3) 120-volt A.C. vital busses #NNai and NN29 energized from their associated inverter connected to D.C. busses #NK01 and NK03, and

4) 125-volt D.C. busses #NK01 and NK03 energized from batteries

                    #NK11 and NK13, as ecA 4 9 W Kag a onzs,,ec

b. Division 2, consisting of:

J

1) 4160-volt emergency bus #NB02, and

2) 480-volt emergency busses #NG02, NG04 and NG06E, and 3)~ 120-volt A.C. vital busses #NNh and NNN energized from their associated inverter connected to D.C. busses #NK02 and NK04, and g 4) 125-volt D.C. busses #NK02 and NK04 energized from batteries

                         #NK12 and NK14x2d An #N/M eM 444 l

i I

ELECTRICAL POWER SYSTEMS 100;~ & EN CDR 3/4.8.4 ELECTRICAL EOUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES L'IMITING CONDITION FOR OPERATION 3.8.4.1 All containrant penetration conductor overcurrent protective devices given in Table 3.8 'A shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one or more of the above required containment penetration conductor overcurrent protective device (s) inoperable:

a. Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated backup circuit breaker or racking out 'or removing the inoperable circuit breaker within 72 hours, declare the affected system or component inoperable, and verify the backup circuit breaker to be tripped or the inoperable l

circuit breaker racked out, or removed, at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, their inoperable circuit breakers racked out, or removed, or

b. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REQUIREMENTS 4.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERABLE:

a. At least once per 18 months:

1) By verifying that the 13.8 kV circuit breakers are OPERABLE by

selecting, on a rotating basis, at least 10*.' of the circuit breakers ef # ^ ' S;

L;;!, and performing the following:

a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated , automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits l ! function as desi.gned and as specified in Table 3.8-1, and WOLF CREEK - UNIT 1 3/4 8-16

ELECT'tICAL POWER SYSTEMS t b 'i SURVEILLANCE REQUIREMENTS (Continued) c) ror each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. 11$~f

2) Byselectingandfunctiona.lytest[ngarepresentativesample of at least 10% of each type of/"C0 .;14 circuit breakers.

Circuit breakers selected for functional testing shall be selected on a rotating basisj .... ... m icael test shall b w.t e ::n:::t :r in;;c.ing : :;rr;nt input at th ;;;;ified h tpcint te each selected circuit brc k;r and ;;rifying th:t :::b circuk creaker functi;n; ;; designed ;nd th; r;;;;n;; tim: i: Ic:: p~~ .... m ..... ...... Circuit breakers found inoperable during. functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be func-tionally tested.until no more failures are found or all circuit breakers of that type have been functionally tested; and

3) By selecting and functionally ~ testing a representative sample of each type of fuse on a rotating basis. Each representative sample of fuses shall include at least 10% af all fuses of that type. The functional test shall consist of a nondestructive

' resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria. Fuses found inoper-able during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation. For each fuse found inoperable during these functional tests, an additional representative sample of at least 10% of all fuses of that type shall be functionally tested unti1 no more failures are found or all fuses of that type have been functionally tested. ,

b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

WOLF CREEK - UNIT 1 3/4 8-17 l .

Insert-Testing of these circuit breakers shall consist of injectirg a current in excess of the breakers noninal setpoint and measuring the response time. %e measured response time will be ecmpared to the manufacturer's data to

 ,     ensure that it is less than or equal to a value specified by the manufacturer.

e e 4 8

e Use revised Table 3.8-1 fran Gilaway specs. attachei behird Wolf Creek Tabl.e 3.9-1 rather than Wolf Creek table. 4 9 y. e 9 e 0 h

5 9 X 6 TABLE 3.8-1 g CONIAINNENT PENETRATION CONDUCIOR m Q DVERCURRENT PROTECTIVE DEVICES

     .                      x e                         \ s 5
    ~*

BREAKER IRIP RESPONSE TIME AT PROTECTIVE DEVICE SEIPOINT MAX. SHORT CIRCUIT POWERED BRSEER AIO LOCATION (Amperes) (Sec/ Cycles) EQUIP 9ENT

                                                                  /

13.8-kV Switchgear j/ .

                                                             /

Primary (P) 252PA0107 3600 (50)/ / 0.1 Reactor coolant pump 372 (51) & 840-(51) . DP8801A

                                                      /    .

P-252PA0108 0.1 Reactor coolant pump 5:' 3163(50)/372 (51) \

                                         & 840 (51) i a                                                            N                  DP88018
    ?             P-252PA0205            3163'(50)/372(51)               O.1        Reactor coolant pump i    g                                    1.840 (51)                                 DPP801C j                 P-252PA02G4            3163 (50/372 (51)               0.1        Reactor coolant pump
                                         & 840 (51)                                 DP8801D 480-Volt Load Center           .

P-52NG0304 1200 (Inst.) 0.05 Hydr n recombiner i

!                 8-52NG0301             4320 (5.T.)                     0.5        SGS01A                   rJ

23 P-52NG404 8-52NC0401 -

1200 (Inst.) 4320 (5:1.) 0.05 0.5 flydrogen recombiner SG5018 b m Ro P-52PG2102 375 (Inst.) 0.025 Pressurizer backup p Ihrough 52PG2112 heater rr l B-350 A Fuse N.A. 15,) i M

                                                                                                         . O t

TJ 4

                                                           .         s b

8 IABLE 3.8-1 (Continued) G . g CONTAlleqENT PENETRATION CONDUCTOR h OVERCURRENT PROTECTIVE DEVICES c '

                                                                          /

35 BREAKER

 "              \            TRIP                     RESPONSE TINE AT/

PROTECTIVE DEVICE SE1 POINT MAX. Sil0RI CIJtCUIT POWERED NUfeiR AND LOCAIION (Amperes) (Sec/ Cycles)' EQUIPMENT P-52PG2202 375 (Inst.) 0.025 Pressurizer Backup 1hrough 52PG2212 Heaters B-350 A Fuse N.A.

                           . 1cco                                       -

P-52NG0lTAFI 4464 (Inst.) 0.016 4 Containment Cooler B-52EDIO8 2400 (5.1.) 0.19 DSGN01A T P-52E03fAF B-52E0305 ( N (Inst. 0.016 Contalnment Cooler w 2400 (S.T 0.19 DSGN01C P-52NG02TAFI ( N nst.) 'x0.016 Containment Cooler B-52NG0208 24 (5.1.) ' Ogl9 DSGN01B o . N P-52NG041AFI ( +456-(Inst.) 0.016 Containment Cor,1er . B-52NG0405 2400 (S.T.) 0.19 DSGN010

                                    ~
                       /

, 480 V-Etor Control Center P-52E0180F3 75 (Inst.) 0.016 RilR Loop Inlet Iso B-40A Fuse N.A. Viv EJilV8701B m

                                                                                                      ~J P-52E02BilR2            10 (Inst.)

B-15A Fuse 0.016 it. A. ESW from Ctat Air Coolers Iso Viv EFilV46 Q-.n P-52tlG02BDF2 45 (Inst.) 0.016 CCW to Ctat Iso Viv B-40A Fuse N.A. EGilV60

2 4

                                                                                                         .~.

C"3 l C3 i

P 6 TABLE 3.8-1 (Continued) ,/ n ' CONTAIMtENT PEMETRATION CONDUCTOR N ty OVERCURRENT PROTECTIVE DEVICES E n

• BREAKER

            .N PROTECTIVDDEVICE TRIP                        RESPONSE. TIME AT SEIPOINT                    MAX. Sil0RT CIRCulT  POWERED NIHlER AND LdCATION       (Amperes)                   IS d/ Cycles)        EQUIPMENT P-52NG02BDF3             29 (Inst.)                             0.016      Ctat Iso Viv Ret from B-ISA Fuse               N,                                     H.A.       Thra Barrier Cooling N                                     '

Coil EGHV62 P-52NG02BEF2 170 ( s t.' ) - 0.016 Sump to RHR Pump B-60A Fuse ' N.A. Vlv EJHV88118 z* ~

                                                        ~

j m P-52NG02BEF3 29 (Inst.) -s 0.016 Ctat Recirc Sump B-ISA Fuse . N N.A. Iso Vive ENilV7

                                                               \

P-52NG02BGF3 [ $sS De&-(Inst.) \0.016 Accumulator Iso V1v B-60A Fuse- .. 3 N: . EPilV8808B 1 isf P-52NG02BilF2 (, 33&-(Inst.) 0.016 Accumulator Iso V1v B-60A Fuse -

                                                                     - N.A.      EPilV88080 P-52NG02BFF3             10 (Inst.)                             0.016     I   control System      '

c, B-ISA Fuse N.A. Mak up Air Valve KAHV30 J7 P-52NG01BBF3 29 (Inst.) 0.016 RC Pump Seal Water 7; B-15A Fuse H.A. Iso V1v BGliV8112 ,go P-52HG018FF3 170 (Inst.) 0.016 Sump to RilR Pump p, B-60A Fuse N.A. V1v EJilV8811A t-$ P-52HG01BEF3 29 (Inst.) 0.016 C.1 Ctat Recirc Sump -i; B-15A Fuse N.A. Iso V1v ENilV1 ca

                                   .                .                                                      O T

M l I ~

TABLE 3.8-1 (Continued) g CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 8 j E - / " BREAKER " TRIP RESPONSE TIME AT PROTECTIV VICE SETPOINT MAX. Sil0RT CIRCUIJ POWERED HI)MBER AND LO T10N (Amperes) (Sec/ Cycles) / EQUIPMENT P-52NG01BGF3 / (Inst.) .016- Acc ator Iso Viv B-60A Fuse . 1 H.A. EPilV8808A . P-52NG01BGF2 / w t. ) 0.016 - Accumulator Isa Viv B-60A Fuse (~~ ~ , N.A. EPilV8808C ti l0 R P-52NG01BFF2 If'JJ-(Inst. )' 0.016 Ctat Air to Aux Bldg B-15A Fuse - N.A. ESF Filter Iso Vlv GSilV20 0 P-52NG01BBR2 29 ( st.) 0.016 React Bldg Discharge B-ISA Fuse 'g H.A. Iso Vlve LFFV95 P-52NG02BBF3 'v;~42 -(Inst.) 016 RilR Loop Inlet B-40A Fuse N.

                                                             '                                         Iso V1v BBPV87028 75        t                     .

P-52NG02BCF2 f 44 (Inst.) 0.016 RilR Loop Inlet B-40A Fuse N.A. Iso Vlv BBPV8702A P-52NG02BilF3 10 (Inst.) 0.016 -v.3 E y to Ctat Air ;J B-15A Fuse N.A. Coolers Iso Viv EFilV34

                                                                                                                                  @-n P-52NG01BCF2               10 (Inst.)             ,                 0.016      ESW to Ctat Air B-15A Fuse                                                          N.A.                                   7,.o Coolers Iso Viv EFilV33 P-52tlG01BDF2                                                                             -} I 10 (InsL.)                               O.016      E5W from;Cptet Air          $y .-s B-ISA Fuse                                                         N.A.        Coolers Iso.Viv EfifV45     rn 21:

C"3 CD t3 4

                                                                                       /

6 TABLE 3.8-1 (Continued) G r., CONTAINMENT PENETRATION CONDUCTOR NI g OVERCURRENT PROTECTIVE DEVICES BREAKER

 -i                                                                                                        .

TRIP H RESPONSE TJ E AT PROTECTIVE DEVICE SETPolNT HAX. Siloff CIRCUIT POWERED NUMBER AND LOCATION (Amperes) (Sec/Cyftes) EQUIPMENT P-52NG01BEF2 75 ( .) 0.016- RilR Loop Inlet Iso Viv B-40A Fuse N.A. EJHV8701A P-52NG03CDF4 29 (Inst.) 0.016 . RCP Thermal Barrier B-15A Fuse . N.A. CCW Iso Valve BBHV13 y P-52NG03CliF1 29 (Inst.) 0.016 RCP Thermal Barrier

• B-15A Fuse .

CCW Iso Viv BBHV14

 'I' y    P-52PG19NAF4              400     nst.)                    0.016      Reactor Cavity Cooling H-100A Fuse                                                N.A.       Fan PCGN02A P-52PG19NCF3              260 (Inst.)                      0.016      C t Atmospheric Control B-60A Fuse                        i N.A.       Sys     Fan DCGR01A L16 P-52PGl9NGF2 B-40 Fuse

( M (Inst.)

• 0.017 RCP A Spa Heater e 1 N.A.

4 g3 P-52PGl9NGF3 [ (Inst.) 0.017 RCP B Space Heat'er B-40 fuse N.A. $

c. 3

                                                                                                    '1 P-52PGl9NEF1              170 (Inst.)                      0.016      RCP A Oil Lift Pump B-40A Fuse                                                                                    Q .1.

N.A. m ssj P-52PGl9NGR3 170 (Inst.) 0.016 RCP B Dil Lift Pump "= B-40A Fuse N.A. P-52PGl9NFF1 22.(Inst.) 0.016

                                                                                                    ]

Ctat Normal Sump C3 B-15A Fuse N.A. Pump DPLF05A

                                                                                                    @- 4:

   =?

p TABLE 3.8-1 (Continue <t) n n CONIAIMIENT PENETRATION CONOUCTOR - lE A OVERCHRRENT PROTE..TIVE DEVICES C

5. BREAKER 1 RIP RESPONSE TIME PROTECTIVE DEVICE SEIPOINT MAX. Sil0RT C UIT POWERED NilHBER AND LOCATION Amperes) (Sec/Cyc1 EQUIPMENT P-52PGl9NFF2 2 Inst.) 0.016- Ctat Normal Sump 8-15A Fuse N.A. Pump DPLFOSC i

P-52PGl9NAF2 84 (Inst.) 0.016 - Instr 7 1 t/ Tunnel 8-40A Fuse N.A. Sump'P DPLF07A t' P-52NG03CBF4

• 29 (Inst.) 0.016 RCP Thermal Barrier CCW B-ISA Fuse N. A.

  'P                                                                                  Iso Viv 8811V15 O   P-52NG03COF2                  29     nst.)                         0.016        RCP Thermal Barrier B-ISA Fuse                                      *

h. . CCW Iso Viv 88HV16 P-52PG20NOFS 320 (Inst.) 0.016 Reac!or Cavity Cooling B-100A Fuse N.A. Fan DLGN028

                                                                  ~

P-52PG20NFF4 260 (Inst.) 0.016 tat Atmospheric Control B-60A Fuse ,

                                              ,                          N.A.         Sy em Fan DCGR018 L1S                                                                            '"CE P-52PG20N0F1           (. INN 6(InsL.)                             0.017        RCP C    ace Heater          N B-40A Fuse           ' ,' l'    ,. I                            N.A.

P-52PG20NCF1 wf. \- - 0.017 RCP O Space lleater cs B-40A fuse f. -3fe (Inst. )

                              ,                                          M.A.

m P' P-52PG20NFF3 170 (Inst.) 0.016 RCP C 011 Lift Pump ' 5]: B-40A Fuse N.A. Q C"3 C".3

                                                                                                                    "'L"3 4

r- , 6 1ABLE 3.8-1 (Continued) G / n / CONIAlfMENT PENETRATION CONIMACIOR / E .

   'A                                         DVERCURRENT PROTECTIVE DEVICES C

5-i x BREAKER

  ~

TRIP RESPONSE TIME AT PROTECTIVE DEVICE SEIPOINT MAX. 5110RT CIRCUIT RED NUMBER AND LOCATION (Amperes) (Sec/ Cycles) _ EQUIPMENT P-52PG20NFF2 170 (Inst.) .016 RCP D 011 Lift Pump B-40A Fuse N.A. P-52PG20 NEE 2 22 (Ins 0.016 . Ctat Normal Sump B-15A Fuse , N.A. Pump DPLF058 R5- P-52PG20NGF4 22 (Inst 0.016 Ctat Normal Sump B-15A Fuse N.A. Pump DPLF050

  ?

y P-52PG20NDR2 (Inst.) 0.016 Instrument Tunnel B-40A Fuse - N.A. Sump DPLF07B VN?$$ { CRIM Control Ro

                                " NO                                     os s

it w ga d 4 og'- Drive Powe ' P-10A Fuse - -- N.A. Gripper Coils (106 fused B-30A Fuse - -- N.A. circuits) , P-50A Fuse B-150A Fuse N.A. N.A. Li Coils (53 fused circ ' s) C3

                                                                                          ,                       3 c_ n (50) - Protective Relay Instantaneous Unit (51) - Protective Relay Inverse Time unit                                             .

Inst. - Instantaneous Protection f.J, S.I. - Short lime Protection oc E-cs C.~3 "YJ

1 TA81.E 3.8-1 CONTAINMENT PENETRATION CON 00CTOR i 0VERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT .  ; i l , 13.8-kV $witchaear I. P-252PA0107 Reactor Coolant Pump - 8-252PA0110/252PA0101 OP8801A P-252PA0108 Reactor Coolant Pump 8-252PA0110/252PA0101 OP88018 l l P-252PA0205 Reactor Coolant Pump 8-252PA0211/252PA0202 DP8801C P-252PA0204 Reactor Coolant Pump (. g 8-252PA0211/252PA0202 DP88010  ; L 480-V Load Center P-125A Fuse Hydrogen Recombiner 8-52NG0304 -

                                                                                                         '5G501A                                                       .

t P-125A Fuse Hydrogen Recombiner 8-52NG0404 $G5018  : P-52PG2102 . Pressurizer Backup  ! Through 52PG2111 Heaters 0-250 A Fuse . n' . i P-52PG2202 Pressurizer Backup I Through 52PG2211 Heaters 8-250 A Fuse 7 t P-52NG01TAF1 Containment Cooler 0-52NG0108 0$GN01A  ; P-52NG03TAF1 Containment Cooler ' 8-51NG0305 0$GN01C P-52NG02TAF1 Containment Cooler I 8-52NG0208 0$GN018 P-52NG04TAF1 Containment Cooler , 8-52NG0405 OSGN010 , P-52PG2007 PG20P MCC  ! 8-52PG2001 j CALLAWAY - UNIT 1 3/4 8-18 _ m v e-n m r?e m, v

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVEPCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT 480-V Load Center (Continued) P-52PG2402 B-500A Fuse Pressurizer Heater P-52PG2403 B-500A Fuse Pressurizer Heater P-52PC2404 B-500A Fuse Pressurizer Heater P-52PG2405 B-500A Fuse Pressurizer Heater P-52PG2406 B-500A Fuse Pressurizer Heater P-52PG2a07 B-500A Fuse Pressurizer Heater 480-V Motor Control Center P-52NG01BOF3 RHR Loop Inlet Iso B-40A Fuse , Viv EJHV8701B P-52NG02BHR2 B-15A Fuse ESW from Ctmt Air Coolers Iso VIv EFHV46 , P-52NG0280F2 B-30A Fuse CCW to Ctmt Iso Viv EGHV60 P-52NG01BHF3 B-40A Fuse CCW Containment Isolation Valve EGHV132 P-52NG0180F1 B-15A Fuse PRT to Containment Valve BBHV8037A P-52NG01BBR3 B-30A Fuse Pressurizer Relief Valve BBHV8000A P-52PG19NEF5 B-100A Fuse Reactor Coolant Orain Tank Pump DPHB02A P-52PG19HCR3 Lighting Transformer B-150A Fuse XQA26 CALLAWAY - UNIT 1 3/4 8-19

I TA8LE 3.8-1 (Continued) t CONTAINMENT PENETRATION CONOUCTOR  ! OVERCURRENT PROTECTIVE DEVICES I PROTECTIVE DEVICE POWERED - NUMBER AND LOCATION EQUIPMENT j 400-V Motor Control Center (Continued) f P-52PG19NCRS . Machine Rm Exhaust Fan 0-15A Fuse OCGN04 P-5A Fuse Flux Mapping Motor Starters  ! S-52PG19GDF3 SR06A P-5A Fuse Flux Papping Motor Starters B-52PG19GDF6 SR068 , P-52NG03CLF2 RCP thermal barrier j 8-15A Fuse- return isolation V1v 88HV16 ^ P-52NG018HF4 CCW Containment Isolation Valve i B-40A Fuse EGHV130 P-52NG0288F4 Hydrogen Mixing Fan t-250A Fuse DCGNO38 i P-52NG02SCF3 . Hydrogen Mixing Fan 8-250A Fuse OCGN030 P-52NG0188F4 Hydrogen Mixing Fan 5-250A Fuse OCGNO3A  ! P-52NG018CF3 Hydrogen Mixing Fan B-250A Fuse DCGNO3C ' P-52NG029JF5 CROM Cooling Fan B-150A Fuse DCGN018

.           P-52PG20 GAR 2                   ,                                                   CROM Cooling Fan B-150A Fuse                                                                          DCGN01A P-52PG20NFR3 B-15A Fuse Pressurizer Cooling Fan DCGN05 it P-52PG20NEF5                                                                        Reactor Coolant O-100A Fuse                                                                        Orain Tank Pump DPH8028 P-52PG20NEF1                                                                        Lighting Transformer 8-150A Fuse                                                                        XQA28 i

CALLAWAY - UNIT 1 3/4 8-20

                                                                                                                               -._-___-.h_-.__                         L__.,.

i TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CON 00CTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED

• NUMBER AND LOCATION EQUIPMENT '

400-V Motor Control Center (Continued) P-52NG028JF1, . 5tandby Lighting L-40A Fuse P-52NG028JF2 Standby Lighting 8-40A Fuse P-52NG028HR3 PRT to Contctnment V1v 8-15A Fuse 88HV80378 P-52NG0280F1 Pressurizer Relief Valve 8-30A Fuse 88HV80008 P-52NG018JF5 CROH Cooling Fan 8-150A Fuse DCGN010 P-52PG19G8R2 CROM Cooling Fan 8-150A Fuse ' OCGN01C P-52NG0280F3 Ctat Iso V1v Ret from 8-15A Fuse Thre Barrier Cooling

    .                                                   coil EGHV62 P-52NG02SEF2                                  Sump to RHR Pump

• 8-60A Fuse Viv EJHV88118 P-52NG028EF3 Ctat Recire Sump 8-15A Fuse Iso Viv ENHV7

       , P-52NG028GF3                                  Accumulator Iso Viv 8-60A Fuse                                   EPHV88088 P-52NG028HF2                                 Accumulator Iso Viv 8-60A Fuse                                   EPNV88080 P-52NG028FF3                                 H Control System 8-15A Fuse                                   Nake-up Air Viv KAHV30 P-52NG0188F3                                 RC Pump Seal Water 8-15A Fuse                                   Iso V1v BGHV8112' P-52NG018FF3                                 Sump to RHR Pump 8-60A Fuse                                   Vlv EJHV8811A CALLAWAY - UNIT 1                  3/4 8-21

l i TA8LE 3.8-1 (Continued) CONTAINMENT PENETRATION CON 00CTOR  ! OVERCURRENT PROTECTIVE DEVICES r PROTECTIVE DEVICE POWERED i NUMBER AND LOCATION EQU!PMENT 480-V Motor Control Center (Continued) t P-52NG018EF3 Ctat Rectre Sump t-15A Fuse . Iso Viv ENHV1 P-52NG018GF3 B-60A Fuse Accumulator Iso Viv EPHV8808A P-52NG018GF2 B-60A Fuse Accumulator Iso Viv ' EPHV8808C P-52NG018FF2 B-15A Fuse Ctat Air to Aux 81dg ESF Filter Iso Viv GSHV20 P-52NG0188R2 5-15A Fuse React b1dg 0ischarge Iso Viv LFFV95 - P-52NG0288F3 RHR Loop Inlet B-40A Fuse l Iso V1v 88PV87028 *

                                                                                   +

P-52NG028CF2 RHR Loop Inlat 5-40A Fuse . Iso V1v 88PV8702A P-52NG028HF3 ' B-15A Fuse ESW to Ctat Air Coolers Iso V1v EFHV34 . P-52NG018CF2 S-ISA Fuse

• E5W to Ctat Air Coolers Iso Viv EFHV33 P-52NG0180F2 3-15A Fuse ESW from Ctat Air Coolers Iso Viv EFHV45 P-52NG018EF2 RHR Loop Inlet Iso V1v B-40A Fuse EJHve701A P

P-52NG03C0F4 RCP Thermal Barrier 8-15A Fuse , CCW Iso Valve 88HV13  ; P-52NG03CHF1 RCP Thermal 8arrier B-15A Fuse CCW !so V1v 88HV14 P 52PG19NAF4 Reactor Cavity Cooling 8 100A Fuse Fan DCGN02A P-52PG19NCF3 Ctmt Atmospheric Control B-60; Fuse System Fan DCGR01A CALLAWAY - UNIT 1 3/4 8 22 '

                                            .I

o 7 , _ TABLE 3.8-1 (Continued) _ CONTAINMENT PENETRATION CONOUCTOR _0VERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED-NUM8ER AND LOCATION EQUIPMENT 480-V Motor Control Center (Continued) P-52PG19NGF2 RCP A Space Heater 8-40 Fuse P-52PG19NGF3 8-40 Fuse RCP 8 Space Heater P-52PG19NEF1 8-40A Fuse RCP A 011 Lift Pump P-52PG19NGR3 8-40A Fuse RCP B 011 Lif t Putnp P-52PG19NFF1 8-15A Fuse Ctmt Normal Sump Pump OPLF05A P-52PG19NFF2 8-15A Fuse Ctmt Normal Sump Pump OPLF05C P-52PG19NAF2 8-25A Fuse Instrument Tunnel Sump Pump OPLF07A P-52NG03CBF4 B-15A Fuse RCP Thermal Barrier CCW Iso Viv 88HV15 . P-52NG03CLF2 8-15A Fuse ' RCP Thermal Barrier , CCW !$o Viv BDHV16 P-52PG20NBF5 8-100A Fuse Reactor Cavity Cooling Fan DCGN028 P-52PG20HFF4 8 60A Fuse Ctmt Atmospheric Control System Fan DCCR018 P-52PG20NBF1 RCP C Space Heater 8 40A Fuse P-52PG20NCF1 RCP O Space Heater 8-40A Fuse P-52PG20NFF3 8 40A Fuse R"P C Oil Lift Purrp 1EPR08C P 3A Fuse RP139 8 3A Fuse Accumulator Tank A ! sol Vlv CPHV8800A F CALLAWAY - UNIT 1 3/4 8 23

u - - - - - - - - - - - - - - - - f

• I

                                                                                                                     . t

TA8LE 3,8-1 (Continued)

CONTAINMENT PENETRATION CONOUCTOR PROTECTIVE OtVICE OVERCURRENT PROTECTIVE DEVICES _ NUM8ER AND LOCATION ' POWERED EQUIPMENT *! l 480-V Motor Control Center (Continued) _ t itPR000 P-3A Fuse , RP139 8-3A Fuse Accumulator Tank C ! sol Viv EPHV8804C i 4tPR04A P-3A Fuse RP140 8-3A Fuse Accumulator EPNV88088 Tank 8 ! sol Viv 1 4tPR088 P-3A Fuse  ; RP140 3-3A Fuse i Accumulator EPNV88080 Tank 0 Isol Viv IEPK098 P 3A Fuse - RL014 0 3A Fuse l Accumulator Tank 8 Vent V1v EPNV89508 i IEPK090 P-3A Fuse RL018 8-3A Fuse Accumulator Tank C Vent V1v

                                                                *EPNV89500                                                -

ltPK09F P-3A Fuse RL014 0 3A Fuse {

    .                                                            Accumulator Tank 0 Vent Viv EPH/8950F 4EPK09A P 3A Fuse                                                                                             :

RL014 6-3A Fuse > Accumulator Tank A Vent V1v EPHV8950A 4 TPK 09C P 3A Fuse ( RL018 8 3A Fuse

                                                         .      Accumulator Tank 8 Vent Viv EPNV8950C        .

4tPN09E P-3A Fuse RL01p 8 3A Fuse Accumulator Tank,C Vent Viv IPNV8950E c 447K038 P-3A Fuse j RLO20 5 3A Fuse Cont. Mini Purge Isol V1v GTHZ11 44TK030 P-3A Fuse RLO20 0-3A Fuse Cont, rurge Isol V1v GTH28 588A01A P ISA Fuse PA0107 Salla Fuse RCP Breaker Control 588A018 P-15A Fuse

• PA0108 8 15A Fuse RCP areaker Control I 888A01C P 15A Fuse PA0205 8 15A Fuse RCP Sreaker '

Control 888A010 P 15A Fuse i i PA0204 6 15A Fuse RCP Breaktr

                                              ,              Control                                                      I l

CALLAWAY UNIT 1 3/4 8 24

        -_._____.____.______.__.I._____._.___

u - TABLE 3.8-1 (Continued) CGNTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE . POWERED NUMBER AND LOCATION EQUIPMENT 480-V Motor Control Center (Continued) ,

      '5BBG02A P-2A Fuse                                                                                           dCP Oil Lift PG19NEF1 8-1A Fuse                                                                                         Pump Control P-52PG20NFF2                                                                                               RCP D Oil Lift Pump B-40A Fuse P-52PG20NER2                                                                                               Ctmt Norr.al Sump B-15A Fuse                                                                                                 Pump DPLF05B P-52PG20NGF4                                                                                                Ctmt Normal Sump B-15A Fuse                                                                                                  Pump DPLF050 P-52PG20NOR2                                                                                                Instrument Tunnel B-25A Fuse                                                -

Sump Pump DPLF07B P-52PG1904 Polar Crane

     -B-600A Fuse                                                                                                HKE13 1

CROM Control Rod Drive Power P-10A Fuse Gripper Coils (106 fused B-30A Fuse . circuits) P-50A Fuse

                                                                                                 -  Lift Coils (52 fused B-150A Fuse                                                                             '
                                                              ~

c circuits) , ' Low Voltage Power and Control. ' 6HBK04B P-3A Fuse 4 RCOT Heat I.xchanger return to RCDT Viv HB115 B-3A Fuse - HBHV7144. 6HBK05A P-3A Fuse -

                                                                        .                                      FICDT'Dischar'geVlv HB115    B-3A Fuse                       ~,       'V,,           s                                       HBHV7143 1HBK19A P-3A Fuse                          ,             s   ',                                          ContGinment Isclation Viv RLO21                                     e                      'N B-3A Fuse ~~'.     ~                                            v                  . s.H8W T176            ~
                                                   ~

o'T , 4KAG04A P-2A Fus,e s Hydrogsn Purge Makeup Air NG02BFF3 B-1A Fuse-x

                                                                                  ."' N.Supoly Viv                                  .
                                                                          ,                                .KAHV30' SK Q155 P-3A Fuse '                                   ;                  s                      "- F' Ire Protection Discharge V1v                                i KC274A B-3A Fuse      s'                         x-                    ,
                                                                                                      ' KCXV261                         '
                                           ,,                   3      ;      -

g "h CALLAWAY - UNIT,1

3/4 8-25

                                        .         s -

4

                   . e*,*'   ^

i TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CON 90CTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT - Low Voltaae Power and Control (Continued) 6KCQ155 P-3A Fuse KC274A 8-3A Fuse Fire Protection Discharge Viv KCXV262 5KCQ19X P-3A Fuse KC274A 8-3A Fuse Fire Protection Detector KCHPS261-002 5KCQ19Y P-3A Fuse KC274A 8-3A Fuse Fire Protection Detector KCHPS262-002 6KES01A P-5A Fuse KE124 8-15A Fuse Fuel Transfer Panel KE124 6KES01A P-5A Fuse KE125 B-15A-Fuse Fuel Transfer Panel KE125 SLFG06A P-2A Fuse PG19NFF1 8-1A fuse Containment Normal Sump Pump A DPLFOSA SLFG06C P-2A Fuse PG19NFF2 8-1A Fuse Containment Normal Sump Pump C DPLF05C

     '6LFG068-P-2A Fuse L. PG20NER2 8-1A fuse                            Containment Normal Sump Pump B DPLF05B l      6LFG06D P-2A Fuse PG20NGF4 B-1A Fuse Containment Normal Sump Pump D DPLF050 5GNG03A P-5A Fuse NG01BJF5 B-3A Fuse                             CRDM Cooling Fan D Discharge Isolation Damper GNHZ44 SGNG03C P-5A Fuse CRDM Cooling Fan C PG19GBR2 8-3A Fuse Discharge Isolation Damper GNHZ43 IEMK048 P-3A Fuse RLO17     8-3A Fuse                            SIS Test Line Viv EMHV8824 1EMK04D P-3A Fuse-RLO17     B-3A Fuse SIS Test Line Viv EMHV8881 4EMK04C P-3A Fuse RLO17     8-3A Fuse                            BIT Test Line Viv EMHV8843 CALLAWAY - UNIT 1                    3/4 8-26

_ _ _ , , . ,., ,we= - - v"'~ ""

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR _0VERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE NUMBER AND LOCATION POWERED EQUIPMENT ' ~. _ Low Voltage Power and Control (Continued) 4EMK04E P-3A Fuse RLO17 B-3A Fuse SI Test Line Viv EMHV8871 5EMK05A P-3A Fuse RLO17 B-3A Fuse SI Test Line Viv EMHV8889A SEMK05C P-3A Fuse RL017 B-3A Fuse SI Test Line Viv EMHV8889C SEMKOSE P-3A Fuse RLO18 B-3A Fuse BIT Test Line Viv EMHV8882 6EMK05B P-3A Fuse. RLO17 B-3A Fuse SI Test Line Viv EMHV88890 6EMK05D P-3A Fuse RLO17 B-3A Fuse SI Test Line Viv EMHV88898 1ENG02A P-2A Fuse l NG01BEF3 8-1A Fuse Containment Spray Sump Isol Viv ' ENHV1 4ENG028 P-2A Fuse NG02BEF3 B-1A Fuse Containment Spray Sump Isol Viv ENHV7 1EPG02A P-2A Fuse . NG01BGF3 B-1A Fuse Accumulator EPHV8808A Tank Isolation Valve 1EPG02B P-2A Fuse NG01BGF2 B-1A Fuse t Accumulator EPHV8808C Tank Isolation Valve i 1EPK02C P-3A Fuse RLO18 B-3A Fuse Accumulator Tank Isolation Valve EPHV88088 Indication 1EMK04A P-3A Fuse L RLO17 i B-3A Fuse SI Test Line Viv EMHV8823 r 4EJG048 P-2A Fuse NG02AFR3 B-1A Fute RHR to charging /SI pump suctions EJHV88048 1EJG05A P-2A Fuse NG01BEF2 B-1A Fuse RHR Shutdown Suction Line Isol Valve , EJHV8701A CALLAWAY - UNIT 1 3/4 8-27

TABLE 3.8-1 (Continued) _ CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE-NUMBER AND LOCATION POWERED-EQUIPMENT Low Voltage Power and Control (Continued) 1EJG05B P-2A Fuse NG018DF3 B-1A Fuse RHR Valve Shutdown Suction Line Isolation EJHV87018 IEJG06A P-2A Fuse NG01BFF3 B-1A Fuse Cont Recirc Sump Isolatjon Valve EJHV8811A 4EJG068 P-2A Fuse NG02BEF2 B-1A Fuse Cont Recirc Sump Isolation Valve EJHV88118 1EJK07A P-3A Fuse RLO17 B-3A Fuse Test Line Isol Viv Hot Leg Inj Line Solenoid EJHCV8825

             '1EJK07C'P-3A Fuse JRL017     B-3A Fuse                           'RHR Test Line Viv

! EJHCV8890A 4EJK078 P-3A Fuse ' RLO17 B-3A Fuse RHR Test Line V1v i EJHCV88908 P-1EJY13A 3A Fuse RLO11 Ctat Sump Sample Isolation Viv B-1RLYO1G 15A Breaker EJHV21 . NG01ACR119 . P-4EJY138 3A Fust RLO11 , Ctmt Sump Sample Isolation Viv 8-4RLYO1G 15A Breaker EJHV22 L NG02ACR140 ! P-48 MYOID 3A Fuse RLO24 L S.G.C Cnt to Nuc Sample Sys Viv 8-4RLYO1H 15A Breaker BMHV22 NG02ACR127- ! P-48MYO2A 3A Fuse

         'RLO24                                                                                       .

S.G.A Tuoe Sheet Sample Viv B-4RLYO1H 15A Breaker BMHV35

s. NG02ACR127 .

I P-4BMY028 3A Fuse RLO24 S.G.B Tube Sheet Sample Viv B-4RLYO1H ISA Breaker BMHV36 t NG02ACR127 CALLAWAY - UNIT 1 3/4 S-28 L.

I TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR

              ,                 OVERCURRENT PROTECTIVE DEVICES                           i PROTECTIVE DEVICE                              POWERED NUMBER AND LOCATION                            EQUIPMENT Low Voltage Power and Control (Continued)

P-4BMYO2C 3A Fuse - S.G.C Tube Sheet Sample Viv RLO24 BMHV37 B-4RLY0lH ISA Breaker er NG02ACR127 P-4BMY020 3A Fuse S.G.D Tube Shee't Sample Viv RLO24 BMHV38 B-4RLYO1H ISA Breaker NG02ACR127 1BNG03A P-2A Fuse RHR Pump RWST Suction Valve NG01ACR2 B-1A Fuse *- BNHV8812A 4BNG038 P-2A Fuse RHR Pump RWST Suction Valve NG02AFF4 B-1A Fuse BNHV88128 1EFG09A P-2A Fuse Cont Cooler Isolution Valve NG01BCF2 B-1A Fuse EFHV33 1EFG09C P-2A Fuse Cont Cooler Isolation valve NG01BDF2 B-1A Fuse EFHV45 4EFG098 P-2A Fuse Cont Cooler Isolation Valve NG02BHF3 B-1A Fuse EFHV34 4EFG090 P-2A Fuse Cont Cooler Isolation Valve NG02BHR2 B-1A Fuse - EFHV46 4 EGG 06A P-2A Fuse RC Pump CCW Return Cont Isol Viv NG02BDF2 B-1A Fuse EGHV60 4 EGG 10A P-2A Fuse Cont Isol Viv CCW Return from NG02BDF3 B-1A Fuse RC pump Ther Barr V1v l EGHV62 1 EGG 17A P-2A Fuse EGHV-60 Bypass Valve Cont Isol Viv NG01BHF4 B-1A Fuse EGHV130 1 EGG 17B P-2A Fuse EGHV-62 Bypass Valve Cont Isol Viv . NG01BHF3 B-1A fuse EGHV132 l l 1EJG04A P-2A Fuse RHR to Charging /SI Pump Suction Viv l NG03CMF4 B-1A Fuse EJHV8804A l l CALLAWAY - UNIT 1 3/4 8-29 I -

Y TABLE 3.'8-1 (Continued)_ CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE NUMBER AND LOCATION POWERED EQUIPMENT , Low Voltage Power and Control (Continued) 1BBK35A P-3A Fuse RL001 8-3A Fuse

Excess Letdown Path to PRT Isol Viv BBHV8157A 4BBK35B P-3A Fuse RLOO1 B-3A Fuse Excess Letdown BBHV81578 Path to PRT Isol Viv 1BBK37A P-3A Fuse RLO21 8-3A Fuse Pressurizer Safety Viv BBHV8010A IBBK37B P-3A Fuse RLO21 B-3A Fuse Pressurizer Safety Viv

,; BBHV8010B 4BBK37C P-3A Fuse RLO21 B-3A Fuse Pressurizer Safety Viv BEHV8010C 1BBG39A P-2A Fuse l NG01BBR3 8-1A Fuse Pressurizer BBHV8000A PORV Isolation Valve 5GRK02A P-3A Fuse RLO20 B-3A Fuse Filtration Unit Oamper GRP0Z5 6GRK028 P-3A Fuse RLO20 B-3A Fuse Filtration Unit Damper GRPDZ15 1GTK03A P-3A Fuse

        -RLO20      B-3A Fuse Ctmt Minipurge Isol Valve GTHIS
        - P-1GSYOID 3A Fuse RLO11 i

Hydrogen Analyzer Ctat Isol Vlv B-1RLYO1G 15A breaker GSHV13 NG01ACR119 P-1GSY01E 3A Fuse RLO11 Hydrdgen Analyzer Ctmt Isol Vlv

      - B-1RLYO1G 15A Breaker                               GSHV14 NG01ACR119 P-1GSY10A 3A Fuse RLO20

Ctmt Atmosphere Monitor Isol Viv B-1RLYO1A 15A Breaker GSHV31 NG01ACR123 a

CALLAWAY - UNIT 1 3/4 8-30 w -, *- ,--e,- -

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)

           ~

P-1GSY108 3'A Fuse RLO20 Containment Atm Monitor Isol Viv GSHV34 B-1RLY01A 15A Breaker NG01ACR123 P-65QYO2B 15A Breaker PG20GBR238 Loose Parts Simulator B-65QY028 30A Fuse PG20GBR1 P-6BBY16A 15A Breaker PG20GBR239 - Reactor Coolant Sys L'evel Alarm B-6BBY16A 30A Fuse PG20GBR1 P-6GNYO9F 15A Breaker PG20NBR225 CRDM Cooling Fan Space Heater B-6GNYO9F 30A Fuse PG20NBR1 P-6GNYO9H 15A Breaker PG20NBR226 CRDM Cooling Fan Space Heater B-6GNYO9H 30A Fuse ~ PG20NBR1 P-6GNYO9K 15A Breaker PG20NBR223 Reactor Cavity Cooling Fan Space Heater B-6GNYO9K 30A Fuse PG20N8R1 ' P-4EPYO2C 15A Breaker NG028AR114 Accumulator Isolation Valve Space Heater B-4EPY02C 30A Fuse NG02BGR4

l. P-4EPY02D ISA Breaker

! NG02BAR115 Accumulator Isolation Valve Space Heater B-4EPYO2D 30A Fuse NG02BGR4 i P-4GNYO98 15A Breaker NG02BAR110 Containment Cooler Fan Space Heater

    'B-4GNYO9B 30A Fuse NG02BGR4 CALLAWAY - UNIT 1                    3/4 8-31

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued)

         'P-4GNYO9D 15A Breaker                                                     Containment Cooler Fan Space NG02BAR111                                                                 Heater B-4GNYO90 30A Fuse NG02BGR4 P-4GNYO9M 15A Breaker             ,                                       Hydrogen Mixing Fan Space NG028AR120                                                                  Heater B-4GNYO9M 30A Fuse NG02BGR4 P-4GNYO9P 15A Breaker                                                      Hydrogen Mixing Fan Space NG02BAR121                                                                 Heater B-4GNYO9P 30A Fuse NG02BGR4 P-1GNYO9A 15A Breaker                                                      Containment Cooler Fan Space

! NG01BAR119 Heater i B-1GNYO9A 30A Fuse. NG01BER3 P-1GNYO9C 15A Breaker Containment Cooler Fan Space - NG01BAR120 Heater B-1GNYO9C 30A Fuse NG01BER3 P-1GNYO9L 15A Breaker Hydrogen Mixing Fan Space . NG01BAR122 Heater B-1GNY09L 30A Fuse NG018ER3 P-1GNYO9N 15A Breaker Hydrogen Mixing Fan Space NG018AR123 Heater i B-1GNYO9N 30A Fuse ! NG01BER3 P-5GNYO9E 15A Breaker - CRDM Cooling Fan Space Heater PG19NHF228 B-5GNYO9E 30A Fuse PG19NHF1

                                                           ~

P-1EPYO2A 15A Breaker Accumulator Iso NG01BAR116 Valve Space Heater B-1EPYO2A 30A Fuse

NG01BER3 CALLAWAY - UNIT 1 3/4 B-32

                                                                                 -y,..,,...,,-,---,-+,,-w- .,- , . , - - . - ., , - . - - , - - - - - - - - ~ ,. ---,s-,

e.- 5 - -- -.. - . +-_ .r_---.r,,s.3- v- - - - <

TABLE 3.8-1 (C'ontinued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Con;rol (Continued) ' P-1EPY028 15A Breaker Accumulator ISO NG01BAR117 Valve Space Heater 8-1EPY028 30A Fuse NG01BER3 P-4GSY10C 3A Fuse Ctmt Atm Monitor Isol Viv RLO20 GSHV36 B-4RLY01A 15A Breaker NG02ACR130 P-4GSY100 3A Fuse Ctmt Atm Monitor Isol Viv RLO20 GSHV39 - B-4RLYO1A 15A Breaker NG02ACR130 l P-5GNYO9G 15A Breaker CRDM Cooling Fan Space Heater PG19NHF229 B-5GNYO9G 30A Fuse PG19NHF1 . P-5GNYO9J 15A Breaker Reactor Cavity Cooling Fan Space l PG19NHF225 Heater B-5GNYO9J 30A Fuse PG19NHF1 SSFY11AA P-30A Fuse Rod Position Panel SF109

     . PN0711     B-60A Fuse SSFY11AB P-30A Fuse                                                                       Rod Position Panel SF1098 l         PN0710      B-60A Fuse ILFG08A P-2A Fuse                                                                         Normal sump Ctmt Isol Viv

, NG01BBR2 B-1A Fuse LFFV95 l SLFG15A P-2A Fuse Instrument Tunnel Sump Pump l PG19NAF2 B-1A Fuse DPLF07A l

   ~ ~ ~

6LFG158 P-2A Fuse Instrument Tunnel Sump Pump PG20NDR2 B-1A Fuse DPLF078 1GSG03A P-2A Fuse Hydrogen Purge Ctat Isolation Viv NG018FF2 B-1A Fuse GSHV20 l 4BBG398 -P-2A Fuse Pressurizer PORV Isol Viv NG02BDF1 B-1A Fuse BBHV80008 CALLAWAY - UNIT 1 3/4 8-33

                 -    ,,    ,   y,. _--. , , , ,        -- , - - . . , _ _ . _ . , . , , . _ _ _ .              .,,,.,y  . ,,,,-. ---. -+-.---.---

_ TABLE 3.8-1 (Continued) l CONTAINMENT PENETRATION CONDUCTOR [ , OVERCURRENT PROTECTIVE DEVICES .

                                                                                        )

PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Pewer and Control (Continued) IBBK40A P-30A Fuse PZR PORV NK5108 B-30A Fuse BBPCV455A 4BBK04B P-30A Fuse PZR PORV NK4421 B-30A Fuse BBPCV456A 5BGK04B P-3A Fuse Alternate Charging Path Isol Valv RLOO1 B-3A Fuse BGHV8147 6BGK04A P-3A Fuse Normal Charging Path Isol Valv RL001 B-3A Fuse BGHV8146 P-5LFY10A 3A Fuse Containment Cooler Drain Valve RLO23 LFLV97 B-SRLYO1H 15A Breaker PG19GCR217 P-5LFY10C 3A Fuse - Containment Cooler Drain valve ! RLO23 LFLV99

 '   B-5RLYO1H ISA Breaker L

PG19GCR217 P-6LFY108 3A Fuse Containment Cooler Drain Valve RLO23 LFLV98 B-6RLYO1G 15A Breaker PG20GBR217 P-6LFY10D 3A Fuse Containment Cooler. Drain Valve RLO23 LFLV100 B-6DLYO1G 15A Breaker PG20GBR217 , P-6LFY17A 3A Fuse Refueling Pool Stand Pipe RLO23 Discharge Valve B-6RLYO1G 15A Breaker LFLV122 PG20GBR217 P-5LFY20A 15A Breaker Instrument Tunnel Sump Moisture PG19NHF224 Sensor B-5LFY20A 30A Fuse TLVF01 PG19NHF1 CALLAWAY - UNIT 1 3/4 8-34 9

TABLE 3.8-1 (Continued) 4 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES . ( . PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) P-6LFY208 15A Breaker Instrument Tunnel Sump Moisture PG20NBR216 .. ' Sensor B-6LFY20B 30A Fuse TVLF02 PG20NBR1 P-SSDYO6C 15A Breaker Local Radiation Monitor Power PG19NHF215 Supplies B-SSDYO6C 30A Fuse SPRIA39-42 PG19NHF1 P-ISJYOID 3A Fuse Press. Ctat Isol Viv RLO11 SJHV128 B-1RLYO1G 15A Breaker NG01ACR119 P-4SJY01A 3A Fuse Press. Liq /HL 1&3 Sample Clr Viv RLO11 SJHV5 B-4RLYO1G ISA Breaker . NG02ACR140 - IGTK03C P-3A Fuse Ctmt Purge Isol Viv RLO20 B-3A Fuse GTHZ7

P-1GSY01E 3A Fuse Hydrogen Analyzer Ctmt Sample Viv l

RLO11 GSHV14 B-IRLYO1G 15A Breaker NG01ACR119 P-1GSYO1F 3A Fuse Hydrogen Anal Samp. Return to Ctmt Viv RLO11 GSHV18 ' B-IRLYO1G 15A Breaker NG01ACR119 P-4GSYOIA 3A Fuse Hydrogen Anal Ctmt Sample Viv RLO11 GSHV4 i B-4RLYO1G NG02ACR140 15A Breaker P-4GSYO1B 3A Fuse Hydrogen Anal Ctmt Sample Viv RLO11 GSHV5 B-4RLYO1G 15A Breaker NG02ACR140 t . CALLAWAY - UNIT 1 3/4 8-35 g 5

                                                     ,-._--_____,_y_,.         m---,_..   . _ , . _ . . _. , .,, , , .     ,,,-,%. . , , ,,__, m__,   , ,_, , ,___m,

l TABLE 3.8-1 (Continued) _ CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE. POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) P-4GSYO1C 3A Fuse RLO11 Hydrogen and Samp Return to Ctmt Viv GSHV9 B-4RLYO1G 15A Breaker NG02ACR140 5BBG028 P-2A Fuse RCP Oil Lift PG19NGR3 B-1A Fuse Pump Control 6BBG02C P-2A Fuse RCP Oil Lift PG20NFF3 B-1A Fuse Pump Control 6BBG020 P-2A Fuse RCP Oil Lift

      - PG20NFF2 B-1A Fuse                            Pump Control 183G03A P-2A Fuse NG03C0F4 B 1A Fuse                             RCP Thermal Barrier Cooler Isol Viv BBHV13 1BBG038 P-2A Fuse NG03CHF1 B-1A Fuse                            .RCP Thermal Barrier Cooler Isol Viv BBHV14 1BBG03C P-2A Fuse NG03CBF4 B-1A Fuse                            RCP Thermal Barrier Cooler Isol Viv BBHV15 1BBG03D P-2A Fuse NG03CLF2 B-1A Fuse                            RCP Thermal Barrier Cooler Isol Viv BBHV16 5BBK05A P-3A Fuse RLOO1 B-3A Fuse RCP Seal Water Return Viv BBHV8141A 5BBK05B P-3A Fuse RLOO1 B-3A Fuse RCP Seal Water Return Viv BBHV81418 6BBK05C P-3A Fuse RCP Seal Water Return Viv RL001 B-3A Fuse                                BBHV8141C 6BBK050 P-3A Fuse
    -                                                RCP Seal Water Return Viv RLOO1 8-3A Fuse                               BBHV81410 6BBK07A P-3A Fuse                             PRT Discharge to RCOT Viv RLO21 .B-3A Fuse                              BBHV8031 6BBK07B P-3A Fuse                             Reactor Makeup Water to PRT Viv RLO21 B-3A Fuse                              BBHV8045 CALLAWAY - UNIT 1                  3/4 8-36
 =_

_ TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE NUMBER AND LOCATION POWERED EQUIPMENT Low Voltace Power and Control (Continued) 6BBK08A P-3A Fuse H8115 B-3A Fuse RCDT Ht. Exch. to PRT Viv BBHV7141 6BBK09A P-3A Fuse . RLO21 B-3A Fuse RV Flange Leakoff Line to RCDT Viv BBHV8032 1BBK30C P-3A Fuse RLO21 B-3A Fuse Reactor Vessel Head Vent Viv BBHV8002A 483K30D P-3A Fuse . RLO21 B-3A Fuse Rx Vessel Head Vent Viv BBHV8002B 1BBG31A P-2A Fuse

   -NG01BDF1 8-1A Fuse                           PRT Emergency Drain Line Viv BBHV8037A      .

4BBG318 P-2A Fuse NG02BHR3 B-1A Fuse PRT Emergency Drain Line Viv BBHV8037B IEPK020'P-3A Fuse i RL018 B-3A Fuse Accumulator Tank D Isol Viv EPHV88080 Indication 1EPK02E P-3A Fuse RL018 B-3A Fuse Accumulator Tank A Isol Viv EPHV8808A Ind'- ,.on IEPK02F P-3A Fuse RL018 B-3A Fuse Accumulator Tank C Isol Viv EPHV8808C Indication 4EPG02C P-2A Fuse - NG02BGF3 B-1A Fuse Accumulator Tank B Isol Viv EPHV8808B 4EPG02D P-2A Fuse NG02BHF2 B-1A Fuse Accumulator Tank D Isol Viv EPHV880bD 4EPK02A P-3A Fuse RLO18 B-3A Fuse Accumulator Tank A Isol Viv ..,. ; ;_. . . ' . EPHV8808A Indication - 4EPK028 P-3A Fuse - RLO18 B-3A Fuse ,, Accumulator Tank C Isol Viv EPHV8808C Indication 4EPK02G P-3A Fuse RLO18 B-3A Fuse Accumulator Tank B Isol Viv EPHV8808B Indication 4EPK02H P-3A Fuse RLO18 B-3A Fuse Accumulator Tank D Isol Viv EPHV8808D Indicattun CALLAWAY - UNIT 1 3/4 8-37

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltace Power and Control (Continued) 5EPK03A P-3A Fuse RLO18 B-3A Fuse Accumulator Nitrogen Supply Viv EPHV8875A SEPK038 P-3A Fuse RL018 B-3A Fuse Accumulator Nitrogen Supply Viv EPHV8875C 5EPK03C P-3A Fuse RL018 B-3A Fuse Accumulator Test Line Viv EPNV8877A 4BBK30B P-3A Fuse RLO21 B-3A Fuse Reactor Vessel Head Vent Viv

• BBHV8001B SEPK030 P-3A Fuse RLO18 B-3A Fuse A'ccumulator. Test line Viv EPHV8877C SEPK03E P-3A Fuse RL018 B-3A Fuse Accumulator Water Fill Line Viv EPHV8878A-SEPK04A P-3A Fuse RL018 B-3A Fuse Accumulator Water Fill Line Viv EPNV8878C 5EPK048 P-3A Fuse RL018 B-3A Fuse Accumulator Test Line Viv EPHV8879A 5EPK04C P-3A Fuse RLO18 B-3A Fuse Accumulator Test Line Viv

                                                      'EPHV8870C 6EPK04D P-3A Fuse RLO18 E-3A Fuse Accumulator Test Line Viv EPHV88798 6EPK04E P-3A Fuse RLO18 B-34 Fuse                               Accumulator Test Line Viv EPHV88790 6EPK05A P-3A Fuse RLO18 B-3A Fuse Accumulator Nitrogen Supply Viv EPHV88758 6EPK058 P-3A Fuse
    - c;NQ318 B-3A. Fuse                               Accumulator Nitrogen Supply Viv EPNV8875D 6EPK05C P-3A Fuse Accumulator Test Line Viv RLO18 B-3A Fuse                                EPHV8877B

', 6EPK050 P-3A Fuse ii :umulator Test Line Viv . RLO18 B-3A Fuse EPHV88770

       'CALLAWAY - UNIT 1                    3/4 8-38

i

                                                                                                              .                  l TABLE 3.8-1.(Continued)

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltaae Power and Control (Continued) ' 6EPKOSE P-3A Fuse RL018 8-3A Fuse Accumulator Water Fill Viv , EPHV88788 6EPK05F P-3A Fuse Accumulator Water Fill Viv RL018 8-3A Fuse EPHV88780 P-4SJY018 3A Fuse Press. Vapor. Cont. Iso. Space Viv. RLO11 SJHV12 8-4RLY01G 15A Breaker , NG02ACR140 P-4SJY01C 3A Fuse - Accums Sample Cont Isol Viv RLO11 SJHV18 8-4RLYO1G 15A Breaker ' NG02ACR140 P-5SJYO38 3A Fuse Accumulator Sample Line Viv RP211 SJHV16 8-5RPYO90 15A Breaker PG19NHF236 , P-5SJYO3C 3A Fuse Accumulator Sample Line Viv RP211 SJHV17 8-5RPYO90 15A Breaker PG19NHF236 P-5SJYO48'3A Fuse Accumulator Sample Line Viv RP211 _ SJHV14 8-5RPYO90 15A Breaker PG19NHF236 ' P-5SJ0Y4C 3A Fuse Accumulator Sample Line Viv RP211 SJHV15 8-5RPYO90 15A Breaker PG19NHF236 . _ P-1SJYO68 3A Fuse HL Sample 3 V1v RP332 SJHV4 8-1RPYO9F 15A Breaker NG018AR140-P-4SJYO6A 3A Fuse HL Sample 1 Viv - RP333 SJHV3 8-4RPYO9F 15A Breaker NG028AR140 l CALLAWAY.- UNIT 1 3/4 8-39 emw-~-,ww, ~w,~mmw, .-m,c.,-,--~,-,-,~ .em,+-,c-c.--

TABLE 3.8-1 (Continued) CL(TAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE NUMBER AND LOCATION POWERED EQUIPMENT _ Low Voltace Power and Control (Continued)

            .P-5SJYO6C 3A Fuse RP211         -

Press Liquid Space Samp Isol Viv 8-5RPYO90 ISA Breaker SJHV20 PG19NHF236 P-48MY01A 3A Fuse RLO24 S.G. A Out to Nuc Sample Sys Viv 8-4RLYO1H 15A Breaker BMHV19 NG02ACR127 P-48MY018 3A Fuse RLO24 S.G. 8 Out to Nuc Sample Sys Viv 8-4RLY01H 15A Breaker BMHV20 - NG2ACR127 i P-48MY01C 3A Fuse RLO24 . S.G. C Out to Nuc Sample Sys Viv 8-4RLYO1H 15A Breaker BMHV21 NG02ACR127 P-5GNYO8A 3A Fuse . RLO20 CROM Cooling Discharge Damper 8-5RLYO1L 15A Breaker GNHZ71 i PG19GCR230 P-5GNYO8C 1A Fuse RLO20 CRDM Cooling Discharge Damper - 8-5RLYO1L 15A Breaker GNHZ73 PG19GCR230 P-6GNY08A 3A Fuse RLO20 CROM Cooling Discharge Damper 8-6RLY01J ISA Breaker GNHZ72 PG20GBR222 , P-6GNY08C 3A Fuse RLO20 CROM Cooling Discharge Damper ' GNHZ74 B-6RLY01J 15A Breaker PG20G8R222 58GK10A P-3A Fuse

     .RL001 8-3A Fuse                             Normal Letdown Isolation Viv BGLCV459 E

CALLAWAY - UNIT 1 3/4 8-40 _~

_ TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONOUCTOR _0VERCURRENT PROTECTIVE DEVICES . PROTECTIVE DEVICE NUMBER AND LOCATION POWERED EQUIPMENT Low Voltace Power and Control (Contin'ed) u SBGK108 P-3A Fuse ' RLOO1 B-3A Fuse Normal Letdown Isolation Viv BGLCV460 SBGK19A P-3A Fuse RLOO1 B-3A Fuse Pressurizer Spray Viv BGHV8145 6BGK20A P-3A Fuse RLOO1 B-3A Fuse Excess Letdown Line Isolation Viv BGHV8143 5BGK35A P-3A Fuse RLOO1 B-3A Fuse Letdown Orifice Isolation Viv BGHV8149A SBGK358 P-3A Fuse RL001 8-3A Fuse, Letdown Orifice Isolation Viv BGHV81498 5BGK35C P-3A Fuse RL001 B-3A Fuse Letdown Orifice Isolation Viv BGHV8149C 1BGK36A P-3A Fuse RL001 8-3A Fuse Letdown Containment Isolation Viv BGHV8160 IBGG38A P-2A Fuse NG01BBF3 8-1A Fuse. Seal Water Ctmt Isolation Viv BGHV8112 1BGK48C P-3A Fuse RLOO1 B-3A Fuse Excess Letdown /RCS Isolation Viv BGHV8153A IBGK480 P-3A Fuse RLOO1 B-3A Fuse Excess Letdown /RCS Isolation Viv BGHV8154A - 4BGK48A P-3A Fuse RLOO1 8-3A Fuse Excess Letdown /RCS Isolation Vlv BGHV81538 48GK488 P-3A Fuse RLOO1 8-3A Fuse Excess Letdown /RCS Isolation Viv BGHV8154B 4BBG12A P-2A Fuse NG02BCF2 B-1A Fuse RHR Pump Suction Isolation Valve BBPV8702A 4BBG128 P-2A Fuse NG02BBF3 8-1A Fuse RHR Loop 2 Inlet Isolation Valve BBPV8702B 1BBK13A P-3A Fuse i RLO21 B-3A Fuse Ctmt Isol Nitrogen Supply to PRT Viv BBHV3026 CALLAWAY - UNIT 1 3/4 8-41

l TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low' Voltage-Power and Control (Continued) 5BBK14C P-3A Fuse RCP Standpipe Makeup Viv RK021 B-3A Fuse BBLCV180 5BBK14D P-3A Fuse RCP Standpipe Makeup Viv l RK021 B-3A Fuse BBLCV181 6BBK14A P-3A Fuse RCP Standpipe Makeup Viv

                                     ^

RLO21 B-3A Fuse + BBLCV178 6BBK14B P-3A Fuse RCP Standpipe Makeup Viv RLO21 B-3A Fuse BBLCV179 , 5BBK158 P-3A Fuse Reactor Coolant Loops RLO21 B-3A Fuse Instrumentation 5BBK15C P-3A Fuse Reactor. coolant Loops RLO21 B-3A Fuse Instrumentation 6BBK15D P-3A Fuse Reactor Coolant Loops RLO21 B-3A Fuse Instrumentation

           -6BBK15E P-3A Fuse-                               Reactor Coolant Loops RLO21 B-3A Fuse                                 . Instrumentation SBBK19A P-3A Fuse                                Pressurizer Spray Valve
           .RLOO2 B-3A Fuse                                  BBPCV455B
           ~5BBK198 P-3A Fuse                                Pressurizer Spray Valve RLOO2 B-3A Fuse                                  BBPCV455C

IBBK30A_P-3A Fuse Reactor Vessel Head Vent Viv RLO21 B-3A Fuse ,

BBHV8001A , 6GNG038 P-5A Fuse CRDM Cooling Fan B iz . NG02BJF5 B-3A Fuse Discharge Isolation Damper l GNHZ42 , . ^ 6GNG03D P-5A Fuse CRDM Cooling Fan A PG20 GAR 2 B-3A Fuse Discharge Isolation Damper j GNHZ41

          '5GNG04A P-6A Fuse                                Cavity Cooling Fan Discharge Damper i

PG19NAF4 B-4A Fuse GNHZ47

          'CALLAWAY - UNIT 1                      3/4 8-42 i

b.

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES *

 .I                                                                                                                       ,

PROTECTIVE DEVICE POWERED NUMBER AND LOCATION EQUIPMENT Low Voltage Power and Control (Continued) 6GNG04B P-EA Fuse PG20NBF5 B-4A Fuse Cavity Cooling Fan Discharge Damper GNHZ4B 1HBK03A P-3A Fuse RCDT Vapor Space CTMT Isol Viv RLO21 B-3A Fuse HBHV7126 6HBK04A P-3A Fuse RCDT Vapor Space CTMT Isol Viv HB115 B-3A Fuse HBHV7127 SEPYO7B P-3A Fuse RPO43 B-15A CB-1 Accumulator Tank Discharge Valve Position Switch EPHV8808DA EPHV8808BA - 6EPYO7A P-3A Fuse Accumulator Tank Discharge RPO44 B-15A CB-1 Valve Position Switch EPHV8808AA

                                                                       . EPHV8808CA 6GTY12A P-15A Breaker                                              CTMT Minipurge Exhaust.

PG20GBR134 B-20A Fuse Isolation Damper GTHZ41 6GTY12A P-15A Breaker CTMT Minipurge Exhaust PG20GBR134 B-20A Fuse Isolation Damper i GTHZ42 P-55RYO9A SA Fuse In-Core Neutron Monitoring Drive SR057 Unit Heater ! B-55RYO9A 20A Breaker SRO1A, B l Dc!cGEc6 P-55RYO9A SA Fuse In-Core Neutron Monitoring Drive SR057 Unit Heater B-55RYO9A 20A Breaker SRO1C, D PG19GEF6 1 . CALLAWAY - UNIT 1 3/4 8-43 i ...  : w

               -..m-   y 4-+-,,   y.--. - -..,,,, .. . ,      _                                                 . . - __,

h

                                                                ,                          \_                         l
         ~

x / t' m ELECTRICAL POWER SYSTEMS 0

                                                         -y                                   100'5 3 *OOE-1l
                                                                                              '_                      C MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES LIMIT     CONDITION FOR OPERATION 3.8.4.2 The ermal overload protection and bypass devices, integy motor starter o .each valve listed in Table 3.8-2, shall be OPE. LE. with the APPLICABILIT_Y: When. er the motor-operated valve is required o be OPERABLE.

ACTION:

                                        \

With one or more of the tihe inoperable, declare the affect 1 overload protection a, /or bypass devices valve (s) inoperable and apply the appropriate ACTION statement (s) for the affe ted valve (s). SURVEILLANCE REOUIREMENTS 4.8.4.2 Theaboverequiredthermalover1Ia rotection and bypass devices shall be demonstrated OPERABLE:

a. At least once per 18 month, , by the pe formance of a TRIP ACTUATING DEVICE OPERATIONAL TEST of the bypass ci cuitry for those thermal overload devices which re either:

1) Continucusly bypassed and temporarily p ced in force only when the valve motors are undergoing periodic maintenance testing, or

2) Normally i force during plant operation and passed under accident donditions.

b.

                                      /

At least once ,er 18 months by the performance of a CHA NEL CALIBRATION of a represent tive sample of at least 25% of: y s

1) A1,T thermal overload devices which are not bypasse'd, s ch that each non-bypassed device is calibrated at least once-pe 6 years, and 2

All thermal overload devices which are continuously bypass d

                  '          and temporarily placed in force only when the valve motors re undergoing periodic or maintenance testing, and thermal over load devices normally in force and bypassed under accident conditions such that each thcamal overload is calibrated and each valve is cycled through at least one complete cycle of full travel with the motor-operator when the thermal overload is OPERABLE and not bypassed, at least once per 6 years.

WOLF CREEK - UNIT 1 3/4 8-25

                         ,.r                 -.y.-                ,e. , , .    ..-,, -                          - - -

l N E N, l

                                          ~,  -
                                                ",, /       TABLE 3.8-2
                                                                                    ~
                                                                                       ,3  35 fj ,I l'/

j g 1[ h MOTOR OPERATED VALVES THERMAL OVERLOAD 3 PROTECTION AND/OR BYPASS DEVICES

                                                                                 /:

l' A /,

                 *#s., \      VALVE NUM8ER              FUNCTION BYPASS DEVICE
                                                                            /(YES/NO) ty65 Q .Th i

RM ' x . J ' f l l l

                    -[ ,
                     /
                 /
                                                                                                 ~

i l WOLF CREEK - UNIT 1 3/4 8-26 L

1

 /                                                                                              !

3/4.9REF(t/NGOPERATIONS 3/4.9.1 BORON CONCENTRATION

                                                                                      ~

LIMITING CONDITION FOR OPERATION 3.9.1 The baron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conoitions is met:

a. A K,ff of 0.95 or less, or

b. A baron concentration of greater than or equal to 2000 ppm.

APPLICABILITY: MODE 6*. ACTION: . With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to'30 gpm of a solution containing greater than.or equal to 7000 ppm boron or its equiv-alent until K,ff is reduced to less than or equal to 0.95 or the baron concentration is restored to greater than or equal to 2000 ppm, whichever is the more restrictive. SURVEILLANCE REOUIREMENTS l

4. 9.1.1 The more restrictive of the above two rea:tivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full-length control red in excess of 3 feet from '

       ,           its. fully inserted position within the reactor vessel.

4. 9.1. 2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours.

4. 0.1. 3 Valves CC-V170 and OC V601 :h !' 50 verified locked-eiosed-and-secured-in position at least caca per 31 daysr--

      *The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

WOLF CREEK - UNIT 1 3/4 9-1 ~

REFUELING OPERATIONS t a' t 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE each with continuous visual indication in the control room and one with audible indication in the containment and control room. APPLICABILITY: MODE 6. ACTION:

a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. -

b. With both of the above required monitors inoperable or not operating, determine the b.oron concentration of the Reactor Coolant System at least once per 12 hours.

l SURVEILLANCE REOUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK at least once per 12 hours,

b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and

c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

l i l WOLF CREEK - UNIT 1 3/4 9-2 l l

REFUELING OPERATIONS 300: & MH COM 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least 100 hours. APPLICABILITY: During movement of irradiated fuel in the reactor vessel. ACTION: With the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel. i WOLF CREEK - UNIT 1 3/4 9-3 l i l

REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be 'in the following status:

a. The equipment door closed and held in place by a minimum of four bolts,

b. A minimum of one door in each airlock is closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1) Closed by an isolation valve, blind flange, or manual valve, or

2) 8e capable of being closed by an OPERABLE automatic containment purge isolation valve.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. SURVEILLANCE REOUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated ccndition or capable of being. closed by an OPERABLE automatic containment purge isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:

a. Verifying the penetrations are in their closed / isolated condition, i or j b. Testing the containment purge isolation valves per the applicable portions of Specification 4.6.3.2.

WOLF CREEK - UNIT 1 3/4 9-4

REFUELING OPERATIONS { h i 3/4.9.5 COMMUNICATIONS ' LIMITING CONDITION FOR OPERATION 3.9.5. Direct communications shall be maintained between the control room and personnel at the refueling station. APPLICABILITY: During CORE ALTERATIONS. ACTION: When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. SURVEILLANCE REOUIREMENTS 4.9.5 Ofrect communications between the control room and personnel at the refueling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS.

                                                                                ?'

WOLF CREEK - UNIT 1 3/4 9-5

e. REFUELING OPERATIONS " g {l]{ 3/4.9.6 REFUELING MACHINE -- LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine sfiall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:

a. The refueling machine used for movement of fuel assemblies having:

1). A minimum capacity of 4800 pounds,

2) Automatic overload cutoffs with the following Setpoints:

                                                        !,,,. % -      .14 a)      Primary -                      bove td

• indicated suspended veight for wet conditions an 1 W pounds above the indicated suspended weight% for dry.-%.. 4. conditions, and

                                                                 %.o b)     Secondary -

O pounds above the Primary overload,_ cut. _off.~.

3) An automatic load reduction trip with a Setpoint of 50 pounds below the suspended weight for wet conditions and Z 0 pewnda g_m ,,,, .. m m

                                                                        ,_vdry conditions.

b. The auxiliary hoist used for latching and unlatching drive rods and thimble plug handling operations having:

1) Agnjmumcapacityof3000 pounds,and

2) 1000 pound load indicator which shall be used to monitor lifting loads for these operation.

APPLICABILITY: Ouring movement of drive rods or fuel assemblies within the - reactor vessel. ACTION: With the requirements for refueling machine and/or auxiliary hoist OPERABILITY not satisified, suspend use of any inoperable refueling machine crane and/or auxiliaiy hoist from operations involving the movement of drive rods and fuel assemblies within the reactor' vessel. SURVEILLANCE REQUIREMENTS 4.9.6.1 The refueling machine used for movement of fuel assemolies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior r 4 b a

t 8 Specification 3.9.6.b.2) Justification -

         'Ihis specification was revised to be consistent with the lost test requirements in 4.9.6.2.

9 e 9 4 9 0 4 D

                                  . .g               ,,
                                                   ,    N, '

=

REFUELING OPERATIONS . 0 ,9 I hh SURVEILLANCE REQUIREMENTS-[(Continued) a,.mtcF % <td.4tw\ U to .: atst of sud-oper-ati n- by - perforniing a load test' of at least 125% of the secondary automatic overload cutoff and . demonstrating an. automatic load cutoff when the ref uelitig' machine . load exceeds the Setpoints of Specification 3.9.6a.2) a l S e .. kb, <, %- s Lac ro.A . akd6 k r u- -$ + l -a <

                          .                                                                  L dc -

ce ,as % Sep. tl .d. Sp..(u h.t. 3A L. Q. 4.9.6.2 Each auxiliary hoist and associated load ir.dicator used for movement

                                                                           ~

of drive rods within the reactor f assel shall be demonstrated OPERABLE within 100 hours. prior to si.ori cf oat, spa. etitres-by performing a load test of at least 1250 pounds. ree. I J E" "" E*I s s t e i V

                                                             - - ~ .

e

                                                                         =

n WOLF CREEK - UNIT 1 3/4 9-7

                                    - _ .                                ^

REFUELING OPERATIONS ' 1 y 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2250 pounds shall be prohibited from travel over fuel assemblies in the spent fuel storage facility. APPLICABILITY: With fuel assemblies in the spent fuel storage facility. ACTION:

a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS l 4.9.7 Crane interlocks and physical stops which prevent crane trave.1 with loads in excess of 2250 pounds over fuel assemblies shall be demonstrated OPERA 8LE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. l l \ l i WOLF CREEK - UNIT 1 3/4 9-8

REFUELING OPERATIONS n nne o fi UUf Ctn'vtrdu Itli t ongV

                                                                                                             .            bun 1 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation."

APPLICA8ILITY: MODE 6 when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no RHR loop OPERABLE and in operation, suspend all operation ing an increase in the reactor decay heat load or a reduction in boron s a-tion of the Reactor Coolant System and immediately initiate correcti to. return the required RHR loop to OPERABLE and operating status as a possible. Close all containment penetrations providing ofrect access fron. e containment atmosphere to the outside atmosphere within 4 hours. l SURVEILLANCE RE0VIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at least once per 12 hours. -

              *The RHR loop may be removed from operation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor l                vessel hot legs.

l l WOLF CREEK - UNIT 1 3/4 9-9

                                         ^

                                                                                         -l PROD & EW COPY   --

REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.* APPLICABILITY: MODE 6 when the water level above the top of the reactor vessel flange is less than 23 feet. ACTION:

a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.

b. With no RHR lodp in operation, suspend all operations involving a reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours.

SURVEILLANCE REOUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 2800 gpm at least once per 12 hours.

g. h.-

• Prior to initial ['riticality, the RHR loop may be removed frcm operation for up to 1 hour per._-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.

WOLF CREEK - UNIT 1 3/4 9-10

7_ _ REFUELING OPERATIONS 3/4.9.9 CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Ventilation System shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION:

a. With the Containment Ventilation System inoperable, close each of the purge valves providing direct access from the containment atmosphere to the outside atmosphere.

b. The provisions of Specifications 3.0.3 and 3.0 4 are not applicable.

SURVEILLANCE REO.UIREMENTS 4.9.9 The Containment Ventilation System shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment purge isolation occurs on manual initiation and on a High Radiation test signal from each of the containment radiation monitoring instrumentation channels. WOLF CREEK - UNIT 1 3/4 9-11

REFUELING OPERATIONS MDF & ElEty COM 3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEM8 LIES LIMITING CONDITION FOR OPERATION 3.9.10.1 At least 23 feet of water shall be maintained over the top of the reactor vessel flange. t APPLICABILITY: Ouring movement of fuel assemblies within the containment when the fuel assemblies being moved are irradiated. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies within the reactor vessel. SURVEILLANCE REOUIREMENTS 4.9.10.1 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies. WOLF CREEK - UNIT 1 3/4 9-12 t -

I REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL t .1

                                                                                                                                )

CONTROL RODS , LIMITING CONDITION FOR OPERATION 3.9.10.2 At least 23 feet of water shall be maintained over the top of the irradiated fuel assemblies within the reactor pressure vessel. APPLICABILITY: During movement of control rods within the reactor pressure vessel while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of control rods within the pressure vessel. LURVEILLANCE REQUIREMENTS 4.9.10.2 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of control rods within the reactor vessel. l l l WOLF CREEK - UNIT 1 3/4 9-13

 \   REFUELING OPERATIONS

(({f, 3 EU C0W 3/4.9.11 WATER LEVEL-STORAGE POOL LIMITING CONDITION FOR OPERATION ~ 3.9.11 At least 23 feet, of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. 3 d APPLICABILITY: Whenever irradiated fuel assemblies are in the/:ter:; pool. ACTION:

a. With the requirsments.of the above specification not satisfied, suspend all movemertt of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit with,in 4 hours. ,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not acplicable.

SURVEILLANCE REOUIREMENTS yp r 4.9.11 The water level in the/E+fatl r:ge pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel stge pool. S(d l l W3LF CREEK - UNIT 1 3/4 9-14

l

• l

           - Specification 3.9.11
                     .                                                                    l
           - Justification -

This specification was revised to be consistent with Wolf Creek ranenclature. t 9 4 t f .' r 4 e e 4 4 9 e Y f l-e i 1 [ - I r f I i a--en--.,_,,.

REFUELING OPERATIONS

                                                                                     }        . hh 3/4.9.12 SPENT FUEL ASSEMBLY STORAGE LIMITING CONDITION FOR OPERATION 3.9.12 Spent fuel assemblies stored in Region 2 shall ba subject to the following conditions:

a. The combination of initial enrichment and cumulative exposure shall be within the acceptable domain of Figure 3.9-1, and

b. No spent fuel assemblies shall be placed in Region 2, nor shall any storage location be changed in designation from being in Region 1 to being in Region 2, while refueling operations are in progress.

APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool. ACTION: - a. With the requirements of the above specification not satisfied,

      ,               suspend all other movement of fuel assemblies and crane operations with loads in the fuel storage areas and move the non-complying fuel assemblies to Region 1. Until these requirements of the above specification are satisfied boron concentration of the spent fuel pool shall be verified to be greater than or equal to 2000 ppm at least once per 8 hours.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS l m ,,,4 , q .s a< .s tig e6^ 4.9.12 Theburn3ofeachspentfuelassemblystored[nRegion2shallbe ascertained by W analysis of its burnup history /, prior to storage in ,

Region 2. A complete record of such analysis shall be kept for the time period l

that the spent fuel assembly remains in Region 2 of the spent fuel pool. I l I

                                                 ~

WOLF CREEK - UNIT 1 3/4 9-15

i i I

                                                 ..!       :. ..      ...1._...                          ..;
                                                                                                                                        .p
   .        #,000                                     '                       '                              '                            '

I e . j j; - ..:. ...., . . i. .. ..L l g 38,000 -- -! . o ACbEPTABLE FOR  ! 3  ! REGION 2 STORAGE ' 3 EM , .

 .      4                                                          i                         l                             I z                                                                                    '                              i m                                .i l

m 28,000 -- -

                                   -+----------------.--                                                   '

C g, x ' w . w 24,000 -

        >                               '                         r A

5 3 20, 000 - . . - - - . - E - s . r u *

        > 18,000 a                               ,

e  : i NOT ACCEPTABLE FOR 3 us 12, 000 REGION 2 STORAGE us 4  : J

       $     8,000          - - . - -              - - - - - -                          . . -                         --                   --

4,000 - - - - - - - - -- - - - - - - - - - - - - - - - - . . . . . . - . . 0 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 l FUEL ASSEMBLY INITIAL ENRICHMENT. w/o U.235 FIGURE 3.9-1 HINIMUM REQUIRED FUEL ASSEMBLY EXPOSURE AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2 I l WOLF CREEK - UNIT 1 3/4 9-16 , I L .

1 REFUELING OPERATIONS gf ] 3/4.9.13 EMERGENCY EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION - 3.9.13 Two independent Emergency Exhaust Systems sp l{ be OPERABLE. - APPLICABILITY: Whenever irradiated fuel is in the/!ter;;; pool.

                                                             ~~
   , ACTION:

a. With one Emergency Exhaust System inoperable, fuel movement within the fuel storage areas or crane operation with loads over the fuel storage areas may proceed provided the OPERABLE Emergency Exhaust System is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers,

b. With no Emergency Exhaust System OPERABLE, suspend all ocerations

         ,        involving movem,ent of fuel within the fuel storage areas or crane operation with loads over the fuel storage areas until at east one' Emergency Exhaust System is restored to OPERABLE status,

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.9.13 The above required Emergency Exhaust Systems shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the syste.m operates for at least 10 continuous hours with the heaters operating;

b. At least once per 18 months, or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone ccmmunicating with the system, by t, yuyst)

1) Verifying that the(c!: ~ > /ystem satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regulatory o

ositions C.S.a. C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 9000 cfm 10%; f WOLF CREEK - UNIT 1 3/4 9-17 t s

                                                               -     m , ,  _ _ ._    h. ..--m   __      ,

Specification 3.9.13 Justification - This specification e s revised to be consistent with Wbif Creek ncraenclature. - 9 6 5 4. l .g f a i i t t t 8

REFUELING OPERATIONS t . SURVEILLANCE REQUIREMENTS (Continued) t

2) Verifying, within 31 days afer 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 Regulatory Guide 1.52, ' Revisison 2, March 1978, for a methyl iodide penetration of . less than 1%; and

3) Verifying a system flow rate of 9000 cfm 10% during system operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours of charcoal adsorber operation, by verifying, 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 tes. ting criteria of Regulatory Position C.6.a 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 2 10%.

2) Verifying that on a Spent Fuel Pool High Radioactivity test signal, the system automatica.11y starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorcer banks and isolates the normal fuel building exhaust flow to the auxiliary / fuel building exhaust fan;

3) Verifying that the system maintains the Faal Building at a negative pressure of greater than or equsi to 1/4 inches Water Gauge relative to the outside atmosphers during system operation; and

4) Verifying that the heaters dissipate .i7 i 3 kW when tested in accordance with ANSI N510-1975.

e. Af ter each complete or partial replacement of a HEPA filter 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 00P test aerosol while operating the system at a flow rate of 9000 cfm 2 10%; and

f. After each comolet'e or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place

             . penetration dnd 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 flow rate of 9000 cfm 210*..

WOLF CREEK - UNIT 1 3/4 9-18 A

l PROD & ES COPY , 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTOOWN MARGIN LIMITING CONDITION FOR OPERATION alcepM6% 3.10.1 The SHUT 00WN MARGIN requirement of Specificab L ' Amay be . ' suspended for measurement of control rod worth and ghutdcwn maroim provid..d reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s). L. s s ,, ' ',, APPLICA8ILITY: MODE 2. ACTION:

a. With any full-length control rod not fully inserted and with less tnan the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm a

boron or its equivalent until the SHUT 00WN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue baration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm baron or its equivalent until the SHUTOOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REOUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full-length control rod not fully inserted shall be demons; rated capable of full insertion when tripoed from at least the 50?. withdrawn position within 24 hours prior to reducing the SHUTOOWN MARGIN to less than the limits of Specification 3.1.1.1. e WOLF CREEK - UNIT 1 3/4 10-1

                                                                  ?Qil.f).I a .. .      .. ?ht.T.UU'!

.t e fUI'l SPECIAL TEST EXCEPTIONS l -

3/4.10.2 GROUP HEIGHT, INSdRTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS' TESTS provided:

a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and

b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2, below.

APPLICABILITY: MODE 1. ACTION: .

                                                                         ~'

With any of the limits of Specifications 3.2.2 or 3.2.3 being [xceeded while the requirements of Specifications 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, ;.1. s. 7, 3. 2.1, and 3.2.4 are suspended, either:

a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or

b. Be in HOT STANOBY within 6 hours.

SURVEILLANCE REOUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.2.2 The requirements of the below listed specifications shall be performed at least once per 12 hours during PHYSICS TESTS:

a. Specifications 4.2.2.2 and 4.2.2.3, and

b. Specification 4.2.3.2.

e o WOLF CREEK - UNIT 1 3/4 10-2 A

SPECIAL TEST EXCEPTIONS  :

                                        '                                                                                 }5]                              {

3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6, may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The Reactor Trip Setpoints'on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and c.

The Reactor is greater Coolant than System or equal to 541*F. lowest' operating loop temperature (Tavg) APPLICABILITY: MODE 2. . ACTION:

a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.

b. With a Reactor Coolant System operating loop temperature (T,yg) less than 541*F, restore T,yg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes. SURVEILLANCE REOUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 541*F at least once per 30 minutes during PHYSICS TESTS. O e WOLF (SEEK - UNIT 1 3/4 10-3

                         ~-

SPECIAL TEST EXCEPTIONS

                                                                                                                                                                                                                                           %                     h              hh 3/4.10.4 REACTOR COOLANT LOOPS o -
                                                      ' LIMITING CONDITION FOR OPERATION

3. $.c TT$e.' liddiens l Jef he., -follogg gre.wed q be. Syedah e'q ' 4. 3,Wy4. M; ' o l ; ;f Specification 3 4. .1. 1 ;ay b; =pe-td [uring th e w.;;;s,;

4; performance' of STARTUP.,and PHYSICS TESTS provided: c , . . . .

                                  .V 'y~
                                                                                  ~. ..                                   ...
                                                                                                                                                                                                     . .~

3..k. :. a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and

                       ..7,                                . g p). .The Reactor Trip Setpoints on the OPERABLE Intermediate and Power
                       ""6                              ,. y.T. ':                           Range channels .are set less than or equal to 25% of RATED THERMAL
              " .' _ 9 ;-.:, .p~y                                                                  WER.

g 4

                                                                                                           %.':wu.                                    ,).[J[wbelow   . U       ru+             h g w asthe               o 1.sP-7         U .. y Interlock                                                     Setp%%

s  %. mmL

                                                , AP  . EL dAffILIiY:. During operation                                                                                                                                                                 ointet. per Wa""

w: nse es by~ ^'. .k me<wsms.h , pa *E

                                                     ' ACTION:                                                                                  ~
                                                                                                                                                          ~

dy NM fE5f' '

                       -, #4WiththeTHERMALPOWERgreaterthantheP-7.InterlockSetpoint[immediately open the Reactor trip breakers.

i*? N'O'

NiW 'leS &;.' ha.  :. As.. f-rn;M reuS coolsY w.. - .gerkcpuce" entu+ toch ' m- cb yeaer  ;,,,ag % .gm a ine s ema,Q i,4,.A r~hI.g jiue Y^h@M to,,

M: ^ l

,',.. -SURVEILLANCE REQUIREMENTS ,.

p. .

                   '. 'a%,G. ',..4.10.4.1J l                                                           ,e:
                                                                . c : . ?The                   . ..           n .r . -                         -+                       7            -
                                                                                                                                                                                                                         .a                                                                                                 - "

THERMAL POWER shall1be deterefe,d to be less than P-7 Interlock'  :- ..

                                  - ~'lSetpoint'atleast.onceperhourduringSTARTUPandPHYSICSTESTS.

r ,

                                                                                                                                                                                                                                                                                            ,                    .          7','
                   .N ,*
                                                                ;g . V- R.a.     .,                  jf@.iH ; ;,' y( ' ;;j,p #'7 . .,, . , _., ;, n '
                                                   .4.10.4.2' Each Intermediate'and Power Range channel, and P-7 Interlock shall                                                                                                                                                                                        <           L
               + pO;;p: "be subjected to sp ANALOG CHANNEL OPERATIONAL TEST.within.12 hours                                                                                                                                                                                                                                   'l            - prio
           < . .g.4.   .. . .K, . initiating..STARTUB and PHYSICS TESTS.,

4m

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\,N -Q 3/4.h'10 '

                                                                                                                                                                                                           , .c
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l l i SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN (1 { h ' LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the

      - performance of individual full-length shutdown and control rod drop time measurements provided only one shutdown or control bank is withdrawn from the              '

fully inserted position at a time. APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements and during surveillance of digital rod position indicators for OPERABILITY. ACTION: With the Position Indication System inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers. SURVEILLANCE REQUIREMENTS 4.10.5 The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least or.ce per 24 hours thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:

a. Within 12 steps when the rods are stationary, and '

b. Within 24 steps during rod motion. '

G I e I WOLF CREEK - UNIT 1 3/4 10-5

                                                                                           .e o k-

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-4) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microcurie /mi total activity. APPLICABILITY: At all times. ACTION: With the concentration 'of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concen-tration to within the above limits. SURVEILLANCE REOUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table' 4.11-1. 4.11.1.1.2 The results of the radioactivity' analysis shall be used in accordance with the methodology and parameters in the 00CM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. EP' ,* *..=

• D O

y WOLF CREEK - UNIT 1 3/4 11-1

                                ,,      _           ._                                  __.-..-_m.       __ _   _

F'. e.e e *

                  , j , s .* ads. .. - * . eme , , >               *

• PRODF & EliW COPY TABLE 4.11-1

                                                                                                                                                 ~

RADI0 ACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM

                       ~

LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING._ ANALYSIS TYPE OF ACTIVITY TYPE FREQUENCY FREQUENCY ANALYSIS (LLD)(1) (uci/ml)

1. Batch Waste P P Release Each Batch Each Batch Principal Gamma 5x10

                                                                                                                               ~7 Tanks (2)                                                            Emitters (3)                 -

1-ul 1x10

a. Waste Monitor Tank  ? M Dissolved and 1x10

                                                                                                                              ~5
                                                                             ~

One Batch /M Entrained Gases (Gamma Emitters)

b. Secondary a Liquid Waste Moni- P M H-3 1x10

                                                                                                                              ~5 tor Tank          .Each Batch             Composite (4)
                             -                                                                                                .j Gross Alpha             1x10 P                       Q     g)        Sr-89, Sr-90         6/x10'd            -

Each Batch Ccmposite Fe-55 1x10

                                                                                                                              -6

2. Continuous -

                                                                                'W      k        Principal Gamma        5x10
                                                                                                                              ~7 Releases (5)                     ,,       (6) CompetiteN  I         Emitters (3)

M %k . I-131 -6 1x10 Steam Generator Blowdown M M Dissolved and -5

    ".         C^

1x10 Grab Sample. Entrained Gases (Gamma Emitters) N H-3 h10 5 bb r n(6) Composite (

                                                                                                                             ~7

__..Gmb bd4 Gross Alpha 1x10 Sr-89, Sr-90 -8

     ~

h kr u(6) composite

                                                   . r.=                        Q.                                      5x10               -

(,a4546 Fe-55 .1x10

                                                                                                                             -6 WOLF CREEK - UNIT 1                                  3/4 11-2
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TABLE NOTATIONS

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