ML17334B515

Forwards 1994 FSAR Update for DC Cook Nuclear Plant, Instructions for Incorporating Update Included W/Copy

ML17334B515
Person / Time
Site:	Cook
Issue date:	07/19/1994
From:	Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17334B516	List: ML17334B515
References
AEP:NRC:0509Q, AEP:NRC:509Q, NUDOCS 9407260242
Download: ML17334B515 (388)

Text

RIG RITY PCCELERATED 1 RIDS PROCESSING)

REGULATORY XNFORMATXON DISTRIBUTION SYSTEM (BIDS)

ACCESSION NBR:9407260242 DOC.DATE: 94/07/19 NOTARIZED: NO DOCKET FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME AUTHOR AFFXLIATXON P FITZPATRICK,E. Xndiana Michigan Power Co. (formerly Xndiana & Michigan Ele RECXP.NAME RECIPIENT AFFILXATION RUSSELL,W.T. Document Control Branch (Document Control Desk)

Forwards "1994 FSAR Update for DC Cook Nuclear Plant," j~

P7m

/'UBJECT:

Instructions for incorpoarting update included w/copy.

DISTRIBUTION CODE: A053D COPIES RECEIVED:LTR TITLE: OR Submittal: Updated FSAR (50.71) and Amendments J ENCL +0 SIZE: ~4 R+S R

NOTES RECIPXENT COPIES RECIPXENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 PD 1 0 HICKMA'N,J 1 1 XNTERNAL: AEO IRB 1 1 NRR/PDLR 1 0 ILE 01 1 1 RGN3 1 . 1 EXTERNAL: IHS 1 1 NRC PDR 1 1 NSIC 1 1 SAIC ATEFX,B. 1 1 D

C U

NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM PI-37 (EXT. 504-2083 ) TO ELIMINATEYOUR NAME FROM DISTRIBUTIONLISTS FOR DOCUMENTS YOU DON'T NEEDl TOTAL NUMBER OF COPIES REQUIRED: LTTR 10 ENCL i ~o

A s ~

4

~,

w h

~ <<I S '

indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 AEP NRC 0509$

Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 1994 FINAL SAFETY ANALYSIS REPORT UPDATE U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Attn: Mr. W. T. Russell July 19, 1994 Dear Mr. Russell Attached are ten copies of the changed pages for the 1994 update to the Cook Nuclear Plant Final Safety Analysis Report. These pages are being transmitted to you according to the provisions of 10 CFR 50.71(e). Instructions for incorporating the update are included with eqch copy.

Changed pages have been dated "July, 1994" in the lower right corner.

in order to identify changed pages in addition to vertically barring the specific change. Vertical change bars next to the .July 1994 date in, the lower right corner indicate that the information has only shifted pages.

We hereby certify that the information contained in this update to the FSAR, to our knowledge, accurately presents changes made between January 22, 1993, and January 22, 1994.

Sincerely, QE~y ~j Vice President af Attachment

<<9407260242 '740719

'PDR ADGCK 05000315, K PDR

Mr. W. T. Russell Page 2 AEP:NRC:0509Q cc: w/o attachment A. A. Blind - Bridgman G. Charnoff J. B. Martin J. R. Padgett NRC Resident Inspector - Bridgman NFEM Section Chief

VOLUME I Chapter 1 Introduction and Summar Pacae Date

1. 0-1 1988 1.0-2 1989 1.0- 1989 1.1 1982

l. 2 1982

l. -3 1982

.1-4 1982 Fig. -1 ORIG

1. 2-1 1993 1.2-2 1993 1.2-3 1993 1.2-4 1991

1. 2-5. 1982 Table 1.2-1 (6 pages) 1989 1.3-1 1982 1.3-2

• 1982 1.3-3 1982 1.3-4 1984 1.3-5 1982 1.3-6 1982 1.3-7 1982 1.3-8 1984 1.3-9 1982 Fig. 1.3-1 1993 Fig. 1.3-2 1990 Fig. 1.3-3 1990 Fig. 1.3-4 1990

, Fig. 1.3-5 1990 Fig. 1.3-6 1990 Fig. 1.3-7 1990 ig. 1.3-8 1993 F g, 1.3-9 1993 Fi . 1.3-10 1993 Fig 1.3-11 1982 1.4-1 1991

. 4-2 1991 1987 1987 1991 1982 1.4-7 1991 1 4-8 1991 1.4-9 1982 1.4-10 1987 1.4-11 1991

I

/

VOLUME I Chapter 1 Introduction and Summar Pacae Date 1.4-12 1991 1.4-13 1991 1.4-14 1991 1.4-15 1991 1.4-16 1993 1 ~ 4-17 1987 1.4-18 1987 1.4-19 1992 1.4-20 1992 1.4-21 1991 1.4-22 1991 Table 1 4-1 (pa 1) 1991 (pe 2) 1991 (pe 3) 1991 (pe 4) 1991 1.5-1 1982 1.6-0 1984 1.6-1 1983 1.6-2 1983 1.6-3 1982 1.6-4 1982 1.6-5 1983 1.6-6 1982 1.6-7 1983 1.6-8 1982 1.6-9 1982 1.6-10 1983 1.6-11 1982 1.6-12 1983 1.6-13 1982 1.6-14 1985 1.6-15 1985 1.6-16 1985 1.6-17 1985 1.6-18 1985 1.6-19 1992 1.6-20 1985 1.6-21 1985

l. 6-22 1985 1.6-23 1985 1.6-24 1985 1.6-25 1992 le 1.6-1 (pg 1) 1989 (pe 2) 1989 (pa 3) 1989 (pe 4) 1989 (pe 5) 1989 (pe 6) 1989 (pe 7) 1989 (pe 8) 1989 (pe 9) 1989 (pe 10) 1989

VOLUME I Chapter 1 Introduction and Summar Pacae Date Table 1.6<<1 (pg 11) 1989 (pg 12) 1989 (pg>>) 1989 (pg 14) 1989 (pg>>) 1989 (pg 16) 1990 (pg 17) 1989 (pg>>) 1989 (pg>>) 1989 (pg 20) 1989 (pg>>) 1989 (pg 22) 1991 (pg 23) 1993 1.7-1 to 1.7-119 1992 1.8-1 1989 1.9-1 1982

VOLUME I Chapter 2 Site and Environment Pacae* Date

2. 1-1 1991 2 '-2

'-3 1993 2 1993 2 '-4 2.1-5 1982 1993 2.1-6 1993 2 '-7 1993 F 1-8 1993 2.1-9 1993 2.1-10 1982 2.1-11 1993 2.1-12 1993 Table 2% 1 1 1989 Table 2 ~ 1-2 1989 Table 2.1-3 1989 Table 2 '-4 2.1-5 1993 Table 1993 Table 2 '-6 1993 Table 2~1 7 1993 Table 2.1-8 1993 Table 2.1-9 1993 Table 2.1-10 1993 Table 2.1-11 1993 Table 2.1-12 1993 Fige 2.1-1 1982 Fig. 2.1-2 1982 Fig. 2.1-3 1993 Fig. 2.1-4 1982 Fig. 2.1-4a 1982 Fig. 2.1-4b 1993 Fig. 2 '-5 1993 Fig. 2 '-6 1982 Fig. 2.1-7 1993 Fig. 2 '-8 1993 Fig. 2.1-9 1993 Fig. 2.1-10 1993 2.2-1 1993 2~2 2 1993 2~2 3 1993 2.2-4 1993 2 '-5 2.2-6 1988 1993 2~2 7 1993

VOLUME I Chapter 2 Site and Environment Pacae Date 2.2-8 1993 2.2-9 1993 2.2-10 1993 Table 2 2 1

~ 1993 Table 2.2-2 1993 Table 2~2 3 1993 Table 2 '-4 2.2-5 1993 Table 1993 Table 2.2-6 1993 Table 2 '-7 1993 Table 2.2-8 1993 Table 2.2-9 1993 Table 2.2-10 Deleted Table 2.2-11 Deleted Table 2.2-12 Deleted Table 2.2-13 Deleted Table 2.2-14 Deleted Table 2.2-15 Deleted Table 2.2-16 Deleted Table 2.2-17 Deleted Table 2.2-18 Deleted Table 2.2-19 Deleted Table 2 '-20 Deleted Table 2.2-21 Deleted Table 2.2-22 Deleted Table 202-23 Deleted Table 2.2-24 Deleted Table 2 '-25 Deleted Table 2.2-26 Deleted Table 2 2 27

~ Deleted Table 2.2-28 Deleted Table 2 '-29 Deleted Table 2.2-30 Deleted Table 2.2-31 Deleted Table 2.2-32 Deleted Table 2 ~ 2 33 Deleted Fig. 2 2-1

~ 1982 Fig. 2 2 2

~ 1993 Fig. 2 2 3 1993 Fig. 2 '-4

~

1993 Fig. 2.2-5 1993 Fig. 2.2-6 1993

VOLUME I Chapter 2 Site and Environment Pacae Date Fig. 2~2 7 1993 Fig. 2.2-8 1993 Fig. 2.2-9 1993 Fig. 2.2-10 1993 Fig. 2.2-11 1993 Fig. 2.2-12 1993 Fig. 2.2-13 1982 Fig. 2.2-14 1982 Fig. 2.2-15 1982 Fig. 2.2-16 1982 Fig. 2.2-17 1982 Fig. 2.2-18 1982 Fig. 2.2-19 1982 Fig. 2.2-20 1982 Fig. 2.2-21 1982 Fig. 2 '-22 1982 Fig. 2.2-23 1992 2.3-1 1982 2~3 2 1982 2~3 3 1982 2.3-4 1982 2.3-5 1982 Fig. 2.3-1 1982 Fig. 2 3 2

~ 1982

2. 4-1 1982 2.4-2 1984 2.4-3 1984 2.4-4 1982 2.4-5 1988 2.4-6 1982 2.5-1 1982 2.5-2 1982 2.5-3 1982 2.5-4 1982 2.5-5 1982 2.5-6 1982 2.5-7 1982 Table 2.5>>1 (2pp) 1989 Fig. 2.5-1 1982 Fig. 2.5-1a 1982 Fig. 2.5-2 1982 Fig. 2.5-3 1982 Fig. 2.5-3a 1982 Fig. 2.5-3b 1982 Fig. 2.5-3c 1982 Fig. 2.5-3d 1982 Fig. 2.5-3e 1982 Fig. 2.5-3f 1982 Fig. 2.5-3g 1982

VOLUME I Chapter 2 Site and Environment Pacae Date Fig. 2.5-3h 1982 Fig. 2.5-3i 1982 Fig. 2.5-3) 1982 2.6-1 1993 2 '-2 2.'6-3 1992 1992 2 6-4

~ 1992 2.6-5 1992 2.6-6 1993 2.6-7 1992 2.6-8 1992 2.6-9 1992

2. 6-10 1992 2.6-11 1993 2 '-12 1993 2.6-13 1993 2.6-14 1992 2.6-15 1992 2.6-16 1992 2.6-17 1993 2.6-18 1993 2.6-19 1993 2.6-20 1993 2 '-21 1993 2.6-22 1993 2.6-23 1992 2.6-24 1992 2.6-25 1992 2.6-26 1993 2.6-27 1992 2.6-28 1992 2.6-29 1992 2.6-30 1993 2.6-31 1993 2.6-32 1992 2.6-33 1993 2.6-34 1993 2.6-35 1992 2.6-36 1992 2 '-37 1992 2.6-38 1992 2 '-39 1992 2.6-40 1992 2 '-41 1992 2.6-42 1992 2.6-43 1993 2 '-44 1993 2.6-44a 1993 Table 2.6-1 1992 Table 2.6-2 1992 Table 2.6-3 1992 Table 2.6-4 1992 Table 2.6-5 1992 Table 2.6-6 1992 Fig. 2.6-1 1992 Fig. 2.6-2 1992 Fig. 2.6-3 1992 Fig. 2.6-4 1992 Fig. 2.6-5 1992 Fig. 2.6-6 1992

VOLUME I Chapter 2 Site and Environment Pacae Date Fig. 2.6-7 1992 Fig. 2.6-8 1992 Fig. 2.6-9 1992 Fig. 2.6-10 1992 Fig. 2 '-11 1992 Fig. 2.6-12 1992 2 '~1 1993 2~7 2 1993 2~7 3 1992

2. 7-4 1992 2 '-5

'-6 1989 2 1988 2~7 7 1993 2.7-8 1989 Table 2.7-1 1992 Table 2 7 2

~ 1993 Table 2+73 1991 Table 2 7-4 1992 Table 2.7<<5 (pg 1) 1993 (pg 2) 1993 (pg 3) 1993 (pg 4) 1990 Fig. 2 '-1 1992 Fig. 2 7 2

~ 1989 Fig. 2 7 3

~ 1992 Fig. 2.7-4 1992 2.8-1 1989 2.8-2 1982 2.9-1 1982 2.9-2 1982 2.9-3 1982 2.9-4 1992 2.9-5 1990 2.9-6 1983 2.9-7 1991 2.9-8 1982 2 '-9 1982 2.9-10 1993

2. 9-11 1982 2.9-12 1982 2.9-13 1982 2 '-14 1990 2.9-15 1990 2 '-16 1990 Table 2.9-1 (3pp) 1989 Table 2.9-2 (5pp) 1989 2.10-1 1985 2.10-2 1982

VOLUME II Chapter 3 Reactor Unit 1 Pacae Date 3 '-1 1993 3~1 2 1992

3. 1-3 1993

3. 1-4 1990 F 1-5 1992 3.1-6 1990 3 1-7 1990 3.1-8 1990 3.1-9 1990 3.1-10 1990 3.1-11 1990 3.1-12 1990 3.1-13 1990 3.1-14 1990 3.1-15 1990 3~2 1 1993 3~2 2 1982 3~2 3 1982 3.2-4 1982 3.2-5 1986 3.2-6 1982 3\27 1982 3.2-8 1982 3.2-9 1982 3.2-10 1982 3.2-11 1990 3.2-12 1982 3 ~ 2 13 1984 3.2-14 1982 3.2-15 1982 3.2-16 1982 3.2-17 1982 3.2-18 1983 3.2-19 1982 3 '-20 1982 3.2-21 1982 3 ~ 2 22 1982 3 ~ 2 23 1982 3.2-24 1982 3.2-25 1982 3.2-26 1982 3 ~ 2 27 1982 3.2-28 1982 3 '-29 1982 3.2-30 1982 302-31 1990

VOLUME EI Chapter 3 Reactor Unit 1 Pacae Date 3 ~ 2 32 1982 3 ~ 2 33 1982 3.2-34 1983 3.2-35 1982 3.2-36 1982 3 ~ 2 37 1982 3.2-38 1982 3.2-39 1982 3.2-40 1982 3.2-41 1982 3.2-42 1982 3.2-43 1982 3.2-44 1983 3.2-45 1982 3.2-46 1987 3.2-47 1987 3.2-48 1987 3.2-49 1987 3.2-50 1982 Table 3.2.1-1 (3pp) 1989 Fig. 3.2.1-1 1982 Fig. 3.2.1-2 1982 Fig. 3.2.1-3 1982 Fig. 3.2.1-4 1982 Fig. 3.2 '-5 1982 Fig. 3.2.1-6 1982 Fig. 3.2. 1-7 1982 Fig. 3.2.1-8 1982 Fig. 3.2.1-9 1982 Fig. 3.2 '-10 1982 Fig. 3.2.1-11 1982 Fig. 3.2. 1-12 1982 Fig. 3.2.1-13 1982 Fig. 3.2.1-14 1982 3.3-1 1990 3~3 2 1992 3~3 3 1992 3.3-4 1982 3.3-5 1992 3.3-6 1992 3~3 7 1983 3.. 3-8 1983 3.3-9 1984 3.3-10 1990 3.3-11 1987 3.3-12 1982

VOLUME II Chapter 3 Reactor Unit 1 Pacae Date 3 3-13 1982 3.3-14 1992, 3 3-15 1992 3.3-16 1992 3 ~ 3 17 1992 3.3-18 1992 3.3-19 1992 3.3-20 1992 3.3-21 1992 3 ~ 3 22 1992 3 ~ 3 23 1992 3 '-24 1992 3.3-25 1993 3.3-26 1989 3 ~ 3 27 1990 Table 3.3.1-1 (pg 1) 1992 (pg 2) 1990 (pg 3) 1993 Table 3.3.1-2 1990 Table 3.3.1-3 1993 Table 3.3.1-3a 1990 Fig. 3.3.1-1 1982 Fig. 3.3.1-2 1984 Fig. 3.3.1-3 1984 Fig. 3.3.1-4 1984 Fig. 3.3.1-5 DELETED Fig. 3.3.1-6 DELETED Fig. 3~3~1 7 DELETED Fig. 3.3.1-8 DELETED Fig. 3.3.1-9 DELETED Fig. 3 '.1-10 DELETED Fig. 3.3.1-11 1984 Fig. 3 '.1-12 1984 Fig. 3~3 ~ 1-13 1984 Fig. 3.3.1-14 1984 Fig. 3.3.1-15 1984 Fig. 3.3.1-16 1984 Fig. 3.3.1-17 1992 3.4-1 1990 3 '-2 1993 F 4.3 1993 3.4.4 1982 3.4.5 1982 3.4-6 1982 3.4-7 1993

VOLUME II Chapter 3 Reactor Unit 1 Pacae Date 3.4-8 1993 3.4-9 1993 3.4-10 1982 3.4-11 1993 3.4-12 1987 3.4-13 1993 3.4-14 1993 3.4-15 1983 3.4-16 1982 3.4-17 1982 3 '-18 1982 3.4-19 1982 3.4-20 1982 3.4-21 1982 Table 3.4.1-1 (2pp) 1989 Table 3.4.1-2 1989 Table 3.4.1-3 1989 Fig. 3.4.1-1 1982 Fig. 3 '.1-2 1982 Fig. 3.4.1-3 1982 Fig. 3.4.1-4 1982 Fig. 3 '.1-4a 1982 Fig. 3.4.1-5 1982 Fig. 3. 4. 1-6 1982 Fig. 3.4.1-7 1982 Fig. 3.4.1-8 1982 Fig. 3.4.1-9 1982 3.5-1 1993 3.5.1-1 1993 3.5.1-2 1993 3.5.1-3 1993 3.5.1-4 1993 3.5.1-5 1993 3.5.1-6 1993 3.5.1-7 1993

3. 5. 1-8 1993 3.5.1-9 1993 3 '.1-10 1993 3.5 '-11 1993 3 '.1-12 1993 3.5.1-13 1993 3.5.1-14 1993 3.5.1-15 1993 3.5.1-16 1993 3.5.1-17 1990 3.5 '-18 1993

VOLUME II Chapter 3 Reactor Unit 1 Pacae Date Table 3.5.1-1 1992 Table 3.5.1-2 Deleted Table 3.5.1-3 Deleted Fig. 3.5.1-1 1990 Fig. 3.5.1-2 1990 Fig. 3.5.1-3 1990 Fig. 3.5.1-4 1990 Fig. 3.5.1-5 1990 Fig. 3.5.1-5a 1992 Fig. 3.5;1-6 1992 Fig. 3.5.1-7 1992 Fig. 3.5.1-8 Deleted FIg. 3.5.1-9 Deleted 3.5.2-1 1992 3.5.2-2 1992 3.5 2-3 1993 3.5.2-4 1993 3.5.2-5 1992 3.5.2-6, 1993 Table 3.5.2-1 1993 Table 3.5.2-2 1991 Table 3.5.2-3 1993 Fig. 3.5.2-1 1993 Fig. 3.5 '-2 1992 3.5.3-1 1993 3.5 '-2 1992 3.5.3-3 1990 3.5.3-4 1990 3.5.3-5 1992 3 '.3-6 1992 3.5.3-7 1993 3.5.3-8 1990 3 ' '-9 1992 3.5.3-10 1993 3.5.3-11 1990 3.5.3-12 1990 3.5.3-13 1992 3.5.3-14 1991 Table 3.5.3-1 (pg 1) 1992 Table 3.5.3<<1 (pg 2) 1992 Table 3.5.3-1 (pg 3) 1990 Table 3.5.3-1 (pg 4) 1992 Fig. 3.5.3-1 1990 Fig. 3 ' '-2 1990 Fig. 3.5.3-3 1990

0 VOLUME IZ Chapter 3 Reactor Unit 2 Pacae Date 3.1-1 1991 3.1-2 1991

3. 1-3 1991 3.1-4 1991 3.1-5 1991 Table 3-1-1 (5pp) 1991 (Notes) 1991 Table 3 ~ 1 2 (p91) 1991 Table 3-1-2 (p92) 1991 Table 3.1-2 (P93) 1991 Table 3-1-2 (pe4) 1989 Table 3 1 3

~ 1991 3.2-1 1982 3~2 2 1991 3~2 3 1982 3.2-4 1991 3.2-5 1991 3.2-6 1982 3 ~2 7 1991 3.2<<8 1991 3 2-9 1991 3.2-10 1991 3 2-11 1991 3.2-12 1991 3.2.13 1991 3 '-14 1991 3.2-15 1991 3.2-16 1991 3 ~ 2 17 1991 3.2-18 1991 3.2-19 1991 3.2-20 1991

3. 2-21 1991 3 ~ 2 22 1991 3 ~ 2 23 1991 3.2-24 1991 3.2-25 1991 3.2-26 1991 3 ~ 2 27 1991 3.2-28 1991 3.2-29 1991 3.2-30 1991 3.2-31 1991 3 ~ 2 32 1991 3 ~ 2 33 1991 3.2-34 1991 3.2-35 1991

VOLUME II Chapter 3 Reactor Unit 2 Pacae Date 3.2-36 1991 3 ~ 2 37 1991 3.2-38 1991 3.2-39 1991 3.2-40 1991 3.2-41 1991 3.2-42 , 1991 3.2-43 1991 3.2-44 1991 3.2-45 1991 3.2-46 1991 3.2-47 1991 3.2-48 1991 3.2-49 1991 3.2-50 1991 3.2-51 1991 3.2-52 1991 3 '-53 1991 3 '-54 1991 3.2-55 1991 3.2-56 1991 3.2-57 1991 3.2-58 1991 3 '-59 1991 3.2-60 1982 3.2-61 1982 3.2-62 1982 3.2-63 1982 3 '-64 1982 3.2-65 1982 3 '-66 1982 3.2-67 1991 3.2-68 1982 3.2-69 1982 3 '-70 1982 3.2-71 1982 3 ~ 2 72 1982 3 ~ 2 73 1991 3.2-74 1982 3.2-75 1982 3.2-76 1991 3 ~ 2 77 1982 3.2-78 1982 3.2-79 1982 3.2-80 1982 3.2-81 1982 3.2-82 1991 3.2-83 1991 3.2-84 1991 3.2-85 1982

VOLUME II Chapter 3 Reactor Unit 2 .Pacae Date 3.2-86 1982 3.2-87 1982 3.2-88 1991 3.2-89 1991 3.2-90 1991 Table 3.2-1 1991 Fig. 3.2-1 1991 Fig. 3~2 2 1991 Fig. 3~2 3 1991 Fig. 3.2-4 1991 Fig. 3.2-5 1991 Fig. 3.2-5a 1991 Fig. 3.2-6 1991 Fig. 3~2 7 1991 Fig. 3.2-8 1982 Fig. 3.2-9 1982 Fig. 3.2-10 1982 Fig. 3.2-11 1982 Fig. 3 ~ 2 12 1982 Fig. 3.2-13 1982 Fig. 3.2-14 DELETED Fig.'ig.

3 '-15 1982 3.2-16 1982 Fig. 3 ~ 2 17 1982 Pig. 3.2-18 1990 Fig. 3.2-19 1982 Fig. 3.2-20 1982 Fig. 3.2-21 1982 Fig. 3 ~ 2 22 1982 Pig. 3 ~ 2 23 1982 Fig. 3.2-24 1982

VOLUME III Chapter 3 Reactor Unit 2 Pacae Date 3.3-1 1991 3~3 2 1991 3~3 3 1991 3.3-4 1992 3.3-5 1991 3.3-6 1991 3~3 7 1991 3.3-8 1991 3.3-9 1993 3.3-10 1991 3.3-11 1991 3.3-12 1991 3.3-13 1991 3.3-14 1991 3.3-15 1991 3.3-16 1991 3.3-17 1991 3.3-18 1991 3'.3-19 1991 3.3-20 1991 3.3-21 1991 3 ~ 3 22 1991 3 ~ 3 23 1992 3.3-24 19.91 3.3-25 1991 3.3-26 1991 3 ~ 3 27 1991 3.3-28 1991 3 '-29 1991 3.3-30 1991 3.3-31 1991 3 ~ 3 32 1991 3 ~ 3 33 1991 3.3-34 1991 3.3-35 1993 3.3-36 1991 3 ~ 3 37 1991 3.3-38 1991 3.3-39 1991 3.3-40 1991 3.3-41 1991 3.3-42 1991 3.3-43 1991 3.3-44 1991 3.3-45 1991 3.3-46 1991 3.3-47 1991

VOLUME III Chapter 3 Reactor Unit 2 Pacae Date 3.3-48 1991 3.3-49 1991 3.3-50 1991 3.3-51 1991 3.3-52 1991 3 '-53 1991 3.3-54 1991 3.3-55 1991 3.3-56 1991 3.3-57 1991 3.3-58 1991 3.3-59 1991 3.3-60 1991 3.3-61 1991 Table F 3 1 (3pp) 1991 Table 3.3-2 (2pp) 1991 Table 3~3 3 1991 Table 3.3>>4 1991 Table 3.3-5 1991 Table 3.3-6 1991 Table 3~3 7 1989 Fig. 3.3-1 1991 Fig. 3~3 2 1991 Fig. 3~3 3 1991 Fig. 3.3-4 1991 Fig. 3.3-5 1991 Fig. 3.3-6 1991 Fig. 3~3 7 1991 Fig. 3.3-8 1991 Fig. 3.3-9 1991 Fig. 3. 3-10 1991

VOLUME III Chapter 3 Reactor Unit 2 Pacae Date Fig. 3.3-11 1991 Fig. 3.3-12 1991 Fig. 3.3-13 1991 Fig. 3.3-14 1991 Fig. 3.3-15 1991 Fig. 3.3-16 1991 Fig. 3 ~ 3 17 1991 Fig. 3.3-18 1991 Fig. 3.3-19 1991 Fig. 3.3-20 1991 Fig. 3.3-21 1991 Fig. 3 ~ 3 22 1991 Fig. 3 ~ 3 23 1991 Fig. 3.3-24 1991 Fig. 3 3-25 1991 Fig. 3.3-26 1991 Fig. 3\3 27 1991 Fig. 3 '-28 ~

1991 Fig. 3 '-29 1991 Fig. 3.3-30 1991 Fig. 3.3-31 1991 Fig. 303-32 1991 Fig. 3 ~ 3 33 1991 Fige 3.3-34 1991 Fig. 3.3-35 1991 Fig. 3.3-36 1991 Fig. 3 3 37

~ 1991 Fig. 3.3-38 1991 3.4-1 1982 3.4-2 1991 3.4-3 1991 3.4-4 1991 3.4-5 1993 3.4-6 1991 3.4-7 1991 3 '-8 1991 3.4-9 1991 3.4-10 1992 3.4-11 1991

VOLUME III Chapter 3 Reactor Unit 2 Pacae.. Date 3.4-12 1991 3.4-13 1991 3.4-14 1991 3.4-15 1993 3.4-16 1991 3.4-17 1991 3.4-18 1991 3.4-19 1991 3.4-20 1991 3.4-21 1991 3.4-22 1991 3.4>>23 1993 3.4-24 1991 3.4-25 1993 3.4-26 1991 3.4-27 1991 3 '-28 1991 3.4-29 1991 3.4-30 1993 3.4-31 1991 3.4-32 1993 3.4-33 1991 3.4-34 1991 3.4<<35 1991 3.4-36 1991 3.4-37 1991 3.4-38 1991 3.4-39 1991 3.4-40 1993 3.4-41 1991 3.4-42 1991 3 '-43 1992 3.4-44 1991 3.4-45 1991 3.4-46 1993 3.4-47 1991 3.4-48 1991 3.4-49 1991 3.4-50 1991 3.4-51 1992 3.4-52 1991 3.4-53 1991 3.4-54 1991 3.4-55 1991 3.4-56 1991 3.4-57 1991 3 '-58 1991 3.4-59 1991 3.4-60 1991

VOLUME III Chapter 3 Reactor Unit 2 Pacae Date

3. 4-61 1991 3.4-62 1991 3.4-63 1991 3.4-64 1991 3.4-65 1991 Table 3.4-1 (pg '1) 1992 (pg 2) 1992 (pg 3) 1991 Table 3 '-2 1991 Table 3.4-3 (2pp) 1991 Table 3.4-4 1991 Fig. 3 4-1 1991 Fig. 3.4-2 1991 Fig. 3.4-3 1991 Fig. 3.4-4 1991 Fig. 3 '-5 1991 Fig. 3.4-6 1991 Fig. 3.4-7 1991 Fig. 3.4-8 1991 Fig. 3.4-9 1991 Fig. 3.4-10 1991 Fig. 3.4-11 1991 Fig. 3. 4-12 1991 Fig. 3.4-13 1991 Fig. 3.4-14 1991 3.5-1 1991 3.5-2 1987 3.5-3 1991 3.5-4 1991 3.5-5 1991 3.5-6 1991 3.5-7 1991 3.5-8 1991 3.5-9 1991 3.5-10 1993 3.5-11 1991

VOLUME III Chapter 3 Reactor Unit 2 Pacae Date 3.5-12 1991 3.5-13 1991

3. 5-14 1991 3.5-15 1991 3.5-16 1991 3.5-17 1991 3.5-18 1991 3.5-19 1987 3.5-20 1990 Table 3.5.1-1 (2pp) 1991 Table 3.5.1-2 (2pp) 1991 Table 3.5 '-3 1991 Fig. 3.5.1-1 1983 Fig. 3.5.1-2 ORZG Fig. 3.5.1-3 1991 Fig. 3.5.1-4 1983 3.5-26 1991 3;5-27 1991 3.5-28 1991 Table 3.5 '-1 1991

VOLUME III Chapter 4 Reactor Coolant S stem Pacae Date 4.1-1 1991 4.1-2 1991 4.1-3 1982 4.1-4 1982 4.1-5 1982 4.1-6 1982 4.1-7 1982 4.1-8 1982 4.1-9 1982 4.1-10 1982 4.1-11 1982 4.1-12 1990 4.1-13 1990 4.1-14 1990 4.1-15 1990 4.1-16 1991 4.1-17 1982 4.1-18 1982 4.1-19 1989 4.1-20 1982 4.1-21 1982 4.1.22 1982 4.1-23 1982 4 '-24 1982 Table 4.1-1 1990 Table 4.1-2 1993 Table 4.1-3 1990 Table 4.1-4 1991 Table 4.1-5 ( pe 1) 1991

( pe 2) 1989 Table 4.1-6 1989 Table 4.1-7 1989 Table 4.1-8 1989 Table 4.1-9 1989 Table 4.1-10 (pe 1) 1990 (pe 2) 1989 Table 4.1-11 (3pp) 1989 Table 4.1-12 (2pp) 1991 4.2-1 1982 4.2-2 1982 4.2-3 1982 4.2-4 1989 4.2-5 1982 4.2-6 1982 4.2-7 1982 4 2-8 1982 4.2-9 1982 4.2-10 1982 4.2-11 1989 4.2-12 1989

VOLUME III Chapter 4 Reactor Coolant 8 stem Pacae Date 4.2-13 1989 4.2-14 1982 4.2-15 1982 4.2-16 1982 4.2-17 1990 4.2-18 1983 4.2-19 1991 4.2-20 1991 4.2-21 1991 4.2-22 1993 4.2-23 1982 4.2-24 1982 4.2-25 1982 4.2-26 1986 4.2-27 1982 4.2-28 1986 4.2-29 1982 4.2-30 1982 4.2-31 1982 4.2-32 1982 4.2-33 1982 4.2-34 1987 4.2-35 1987 4.2-36 1987 Table 4.2-1 (3pp) 1989 Table 4.2-2 1992 Table 4.2-3 1993 Fig. 4.2-1 1984 Fig. 4.2-1A 1988 Fig. 4.2-2 1982 Fig. 4.2-2A 1982 Fig. 4.2-3 1982 Fig. 4.2-4 1982 Fig. 4.2-4A 1989 Fig. 4 2-5 1982 Fig. 4.2-6 1982 Fig. 4.2-7 1982 Fig. 4.2-8 1982 Fig. 4.2-9 1982 Fig. 4.2-9 Ref. (4pp) 1982 4.3-1 1982 4.3-2 1982 4.3-3 1982 4.3-4 1982 4.3-5 1982'982 4.3-6 4.3-7 1990 4.3-8 1982 4.3-9 1982

VOLUME III Chapter 4 Reactor Coolant S stem Pacae Date 4.3-10 1982 4.3-11 1989 4.3-12 1989 4.3-13 1989 4.3-14 1989 4.3-15 1989 4.3-16 1989 4.3-17 1989 4.3-18 1989 4.3-19 1989 4.3-20 1989 4.3-21 1989 4.3-22 1989 4.3-23 1989 4.3-24 1989 4.3-25 1989 4.3-26 1989 Table 4.3-1 1990 Table 4.3-2 1990 Table 4.3-3 1989 Table 4.3-4 1989 Table 4 3-5 (2pp) 1990 Table 4.3-6 1990 Table 4.3-7 1990 Table 4.3-8 1989 Fig. 4.3-1 '982 Fig. 4.3-2 1982 Fig. 4.3-3 1982 Fig. 4.3-4 1982 Fig. 4.3-5 1982 Fig. 4.3-6 1990 Fig. 4.3-7 1990 4.4-1 1986 4.4-2 1988 4.4-3 1986 4.5-1 1982 4.5-2 1988 4.5-3 1982 4.5-4 1982 4.5-5 1993 4.5-6 1993 4.5-7 1982 4.5-8 1982 4.5-9 1982 4.5-10 1982

4. 5-11 1982 4.5-12 1982

VOLUME III Chapter 4 Reactor Coolant S stem Pacae Date 4.5-13 1982 4.5<<14 1983 4.5-15 1982 4.5-16 1982 4.5-17 1982 4.5-18 1982 4.5-19 1982 4.5-20 1988 4.5-21 1988 4.5-22 1988 4.5-23 1988 4.5-24 1988 4.5-25 1988 Table 4.5-1 (pg 1) 1989 (pg 2) 1989 (pg 3) 1989 (pg 4) 1990 Fig. 4.5-1 1982 Fig. 4.5-.2 1982 Fig. 4.5-3 1982

VOLUME IV Chapter 5 Containment S stem Pacae Date 5.0-1 1989 5.1-1 1987 5.1-2 1982 5.1-3 1982 5.1-4 1982 Table 5.1-1 (2pp) 1989 5.2-1 1989 5.2-2 1982 5.2-3 1982 5.2-4 1982 5.2-5 1982 5 '-6 1986 5.2-7 1982 5 '-8 1990 5 '-9 1986 5 '-10 1982 5 '-11 1982 5 '-12 1982 5.2-13 1982 5.2-14 1982 5 '-15 1982 5.2-16 1982 5 '-17 1982 5 '-18 1982 5.2-19 1982 5.2-20 1982 5.2-21 1982 5 '-22 1982 5.2-23 1982 5.2-24 1982 5 '-25 1982 5.2-26 1990 5.2-27 1982 5.2-28 1987 5.2-29 1982 5.2-30 1982 5.2-31 1987 5 2-32 1987 5.2-33 1987 5.2-34 1982 5.2<<35 1987 5.2-36 1987 5.2-37 1988 5.2-38 1987 5.2-39 1987 5 '-40 1987 5 '-41 1987 5.2-42 1982 5.2>>43 1987 5 '-44 1990 5 '-45 1989

VOLUME XV Chapter 5 Containment S stem Pacae Date 5.2-46 1989 5.2-47 1989 5.2-48 1982 5.2-49 1988 5.2-50 1989 5.2-51 1989 5.2-52 1987 5.2-53 1988 5.2-54 1988 5.2-55 1989 5.2-56 1990 5.2-57 1987 5.2-58 1987 5.2-59 1991 5.2-60 1989 5.2-61 1988 5.2-62 1987 5.2-63 1987 5.2-64 1987 5.2-65 1987 5.2-66 1987 5.2-67 1987 5.2-68 1988 5.2-69 1988 5.2-70 1987 5.2-71 1988 5.2-72 1987 5.2-73 1987 5.2-74 1987 5.2-75 1987 5.2-76 1987 5.2-77 1988 5.2-78 1988 5.2-79 1990 5.2-80 1991 5.2-81 1990 5.2-82 1990 5.2-83 1990 5.2-84 1990 5.2-85 1990 5.2-86 1990 5.2-87 1990 5.2-88 1990 5 '-89 1993 5.2-90 1990 5.2-91 1990 5.2-92 1990 5.2-93 1990 5.2-94 1990 5.2-95 1990

VOLVME IV Chapter 5 Containment S stem Pacae Date 5.2-96 1990 5.2-97 1990 5 2-98 1990 5.2-99 1990 5.2-100 1990 5.2-101 1990 5.2-102 1990 5.2-103 1990 5.2-104 1990 5.2-105 1990 5.2-3.06 1990 5.2-107 1990 5.2-108 1990 5.2-109 1990 5.2-110 1990 5.2-111 1990 5.2-112 1990 5.2-113 1991 5.2-114 1990 Table 5.2-1 1990 Table 5.2-2 (2pp) 1990 Table 5.2-3 1990 Table 5.2-4 1990 Table 5.2-5 1990 Table 5.2-6 1990 Table 5.2-7 1990 Fig. 5. 2-3. 1982 Fige 5.2-2 1982 Fig. 5.2-3 ORIG Fige 5.2-4 1982 Fig. 5.2-5 1988 Fig. 5.2 '-1 1982 Fig. 5.2.2-1A 1982 Fig. 5.2.2-2 1982 Fig. 5.2.2-2A 1982 Fig. 5.2.2-3 1982 Fig. 5.2.2-4 1982 Fig. 5.'2 2-4A 1982 Fig. 5.2.2-4B 1982 Fig. 5.2 '-5 1982 Fig. 5.2.2-6 1982 Fig. 5.2.2-6A 1982 Fig. 5.2.2-6B 1982 Fig. 5.2.2-6C 1982 Fig. 5.2.2-6D 1982 Fig. 5.2.2-7 1982 Fig. 5.2 '-8 1982 Fig. 5.2.2-9 1982 Fig. 5.2.2-10 1982 Fig. 5.2.2-10A 1982

VOLUME IV Chapter 5 Containment S stem Pacae Date Fig. 5.2.2-11 1982 Fig. 5 2.2-11A 1982 Fig. 5.2.2-12 1982 Fig. 5.2.2-12A 1982 Fig. 5.2.2-13 1982 Fig. 5.2.2-14 1982 Fig. 5.2.2-15 1982 Fig. 5.2 '-16 1982 Fig. 5.2.2-17 1982 Fig. 5.2.2-18 1982 Fig. 5.2.2-19 1982 Fig. 5.2.2-20 1982 Fig. 5.2.2-21 1982 Fig. 5.2.2-22 1982 Fige 5.2.2-23 1982 Fig. 5.2.2-24 1982 Fig. 5.2.2-25 1982

'ig. 5.2.2-26 1982 Fig. 5.2.2-27 1982 Fig. 5.2.2-28 1982 Fig. 5.2.2>>29 1982 Fig. 5.2.2-30 1982 Fig. 5.2.2-31 1982 Fig. 5.2.2-32 1982 Fig. 5.2.2-33 1982 Fig. 5.2.2-34 1982 Fig. 5.2.2-35 1982 Fig. 5.2.2-36 1982 Fig. 5.2.2-37 1982 Fig. 5.2.2-38 1982 Fig. 5.2.2-39 1982 Fig. 5 '.2-40 1982 Fig. 5.2.2-41 1982 Fig. 5.2.2-42 1982 Fig. 5.2.2-43 1982 Fig. 5.2.2-44 1982 Fig. 5.2.2-45 1982 Fig. 5.2.2-46 1982 Fig. 5.2.2-47 1982 Fig. 5.2.2-48 1982 Fig. 5.2.2-49 1982 Fig. 5.2.2-50 1982 Fig. 5.2.2-51 1982 Fig. 5. 2. 2-51A 1982 Fig. 5.2.2-51B 1982 Fig. 5.2.2-51C 1982 Fig. 5 '.2-51D 1982 Fig. 5.2.2-51E 1982 Fig. 5 '.2-52 1982 Fig. 5.2.2-52A 1982 Fig. 5 '.2-53 1982

VOLUME 1V Chapter 5 Containment S stem Pacae Date Fig. 5.2.2-54 1982 Fig. 5.2 '-54A 1982 Fig. 5.2.2-54B 1982 Fig. 5.2.2-55 1982 Fig. 5.2.2-55A 1982 Fig. 5 2.2-56 1982 Fig. 5 2.2-56A 1982 Fig. 5.2.2-57 1982 Fig. 5.2.2-57A 1982 Fig. 5.2 '-58 1982 Fig. 5.2.2-58A 1982 Fig. 5.2.2-59 1982 Fig. 5.2-2-59A 1982 Fig. 5.2 2-59B 1982 Fig. 5.2.2-59C 1982 Fig. 5.2.2-59D 1982 Fig. 5.2.2-59E 1982 Fig. 5.2.2-60 1991 Fige 5 '.2-60A 1991 Fig. 5.2.2-60B 1982 Fig. 5.2.2-60C 1982 Fig. 5.2.2-61 1982 Fig. 5 '.2-62 1982 Fig. 5.2.2-63 1982 Fig. 5.2 '-64 1982 Fig. 5.2.2-65 1982 Fig. 5.2.2-65A 1982 5.3-1 1984 5.3-2 1984 5.3-3 1982 5.3-4 1982 5.3-5 1982 5.3-6 1982 5.3-7 1982 5.3-8 1982 5.3-9 1982 5.3-10 1988 5.3-11 1982 5.3-12 1993 5.3-13 1984 S.3-14 1993 5.3-15 1982 5.3-16 1984 5.3-17 1984 5.3-18 1982 5.3-19 1982 5.3-20 1990 5.3-21 1982 5.3-22 1982 5.3-23 1982

VOLUME IV Chapter 5 Containment S stem Pacae Date 5.3-24 1982 5.3-25 1982 5.3-26 1986 5.3-27 1986 5.3-28 1982 5.3-29 1988 5.3-30 1988 5.3-31 1988 5.3-32 1982 5.3-33 1988 5.3-34 1989 Table 5.3-1 (pg 1) 1993 (pg 2) 1989 Fig. 5.3-1 1982 Fig. 5.3-2 1986 Fig. 5 '-2A 1986 Fig. 5.3-3 1982 Fig. 5.3-4 1982 5.4-1 1992 5 4-2 1992 5.4-3 1992 5.4-4 1992 5 '-5 1992 5.4-6 1992 5.4-7 1992 5.4-8 1992 Table 5-4-1 (pg 1) 1993 (pg 2) 1993 (pg 3) 1993 (pg 4) 1993 (pg 5) 1993 (pg 6) 1993 (pg 7) 1993 (pg 8) 1993 (pg 9) 1993 (pg 10) 1993 (pg>>) 1993 (pg>>) 1993 Fig. 5.4-1 1982 5 '-1 1982 5 '-2 1982 5.5-3 1987 5 '-4 1993 5.5-5 1993 5.5-6 1993 5.5-7 1987 5.5-8 1993 5.5-9 1992 5.5-10 1992 5.5-11 1992 5 '-12 1987 5 '-13 1987

VOLUME EV Chapter 5 Containment S stem Pacae Date 5.5-14 1987 5.5-15 1987 5.5-16 1993 5.5-17 1993 Fig. 5.5-1 1985 Fig. 5.5-2 1985 Fig. 5.5-3 1982 5.6-1 1993 5.6-2 1993 5.6-3 1992 5.6-4 1986 Fig. 5.6-1 1993 5 '-1 1982 F 7-2 1982 5.7-3 1982 5.7-4 1982 5.7-5 1982 5.7-6 1992 5.7-7 1982 5 7-8 ORIG Fig. 5.7-1 1982 Fig. 5.7-2 1982

VOLUME IV Chapter 6 En ineered Safet Features Pacae Date 6.1-1 1989 6.1-2 1989 6.1-3 1989 6.1-4 1989 6.1-5 1989 6.1-6 1989 6.1-7 1989 6.1-8 1989 6.1-9 1989

6. 1-10 1989 6.1-11 1989 Table 6.1-1 1989 6.2-1 1982 6.2-2 1982 6.2-3 1982 6.2-4 1982 6.2-5 1993 6.2-6 1993 6.2-7 1993 6.2-8 1993 6.2-9 1988 6.2-10 1991 6.2-11 1992 6.2-12 1992 6.2-13 1992 6.2-14 1993 6.2-15 1982 6.2-16 1993 6.2-17 1992 6.2-18 1992 6.2-19 1982 6.2-20 1982 6.2-21 1990 6.2-22 1990 6.2-23 1990 6.2-24 1990 6.2-25 1982 6.2-26 1982 6.2-27 1982 6.2-28 1982 6.2-29 1982 6.2-30 1993 6.2-31 1993 6.2-32 1982 6.2-33 1993 6.2-34 1982 6.2-35 1982 6.2-36 1982 6.2-37 1993

VOLUME IV Chapter 6 En ineered Safet Features Pacae Date

6. 2-38 1982

6. 2-39 1982 Table 6.2-1 1990 Table 6.2-2 1989 Table 6.2-3 1993 Table 6.2-4 1989 Table 6.2-5 1991 Table 6.2-6 (pg 1) 1993 (pg 2) 1989 (pg 3) 1989 Table 6.2-7 (2pp) 1991 Table 6.2-8 1989 Table 6.2-9 1989 Table 6.2-10 (2pp) 1989 Fig. 6.2-1 1993 Fig. 6.2-1A 1992 Fig. 6.2-2 1982 Fig. 6.2-3 1982 Fig. 6.2-4 1991 Fig. 6.2-5 1982 6.3-1 1982 6.3-2 1982 6.3-3 1982 6.3-4 1982 6.3-5 1989 6.3-6 1982 6.3-7 1982 6.3-8 1982 6.3-9 1986 6.3-10 1991 6.3-11 1982 6.3-12 1982 6.3-13 1982 Table 6.3-1 1989 Table 6.3-2 1991 Table 6.3-3 1991 Table 6.3-4 (pg 1) 1992 (pg 2) 1989 Fig. 6.3-1 1982

VOLUME V Chapter 7 Instrumentation and control Pacae Date 7.1-1 1993 7 '-2 1982 7.2-1 1982 7~2 2 1990 7~2 3 1982 7.2-4 1982 7.2-5 1982 7.2-6 1982 7~2 7 1982 7.2-8 1982 7 '-9 1982 7.2-10 1982 7.2-11 1982 7.2-12 1982 7.2-13 1990

7. 2-14 1982 7.2-15 1982 7.2-16 1982 7 '-17 1982 7.2-18 1982 7 '-19 1982 7.2-20 1982 7 '-21 1990 7.2-22 1988 7 ~ 2 23 1982 7.2-24 1982 7.2-25 1986 7.2-26 1987 7 ~ 2 27 1982 7.2-28 1982 7.2-29 1992 7.2-30 1987 7.2-31 1992 7 ~ 2 32 1987 7 ~ 2 33 1987 7.2-34 1987 7.2-35 1987 7.2-36 1992 7~2 37 1987 7.2-38 1990 7.2-39 1987 7 '-40 1987 7.2-41 1987 7.2-42 1990 7.2-43 1987 7.2-44 1987 7+2-45 1987 7.2-46 1990

VOLUME V Chapter 7 Instrumentation and Control Pacae Date 7.2-47 1987 7.2-48 1987 7.2-49 1987 7.2-50 1987 7.2-51 1987 7.2-52 1987 7.2-53 1987'989 Table 7.2-1 (pg 1)

(pg 2) 1992 (pg 3) 1989 (pg 4) 1990 (pg 5) 1990 Table 7.2-2 (pg 1) 1991 (pg 2) 1992 (pg 3) 1991 (pg 4) 1991 (pg 5) 1991 (pg 6) 1991 Table 7 2 3

~ 1991 Table 7.2-4 1991 Table 7.2-5 (2pp) 1991 Fig. 7 2-la 1990 Fig. 7.2-1b 1990 Fig. 7.2-1c 1990 Fig. 7.2-1d 1990 Fig. 7~2 2 1982 Fig. 7~2 3 1982 Fig. 7.2-4 1982 Fig. 7.2-5 1982 Fig. 7.2-6 1982 Fig. 7%2 7 1982 Fig. 7.2-8 1982 Fig. 7.2-9 1982

7. 3-1 1987 7~3 2 1990 7~3 3 1982 7.3-4 1982 7.3-5 1982 7.3-6 1992 7~3 7 1983

7. 3-8 1982 7.3-9 1982 7.3-10 1990 7.3-11 1982 7.3-12 1982 7 ~ 3 13 1982 Fig. 7.3-1 1982 7 '-1 1982 7.4-2 1991

VOLUME V Chapter 7 Instrumentation and Control Pacae Date 7 '-1 1982 7.5-2 1982 7.5-3 1982 7.5-4 1982 7.5-5 1982 7.5-6 1991 7 '-7 1987 7.5-8 1987 7.5-9 1989 7.5-10 1987 7.5-11 1982 7.5-12 1982 7.5-13 1991 7.5-14 1982 7.5-15 1982 7 5-16 1982 7.5-17 1982 7.5-18 1991 7.5-19 1982 7.5-20 1989 Table 7.5-1 1989 Table 7.5-2 (2pp) 1989 Fig. 7.5-1 1982 Fig. 7.5-2 1982 Fige 7.5-3 1982 7.6-1 1991 7.6-2 1991 7.6-3 1991 7.6-4 1991 7.6-5 1991 Fig. 7.6-1 1982 Fig. 7.6-2 1982 Fig. 7.6-3 1982 7.7-1 1982 7~7 2 1982 7~7 3 1982 7.7-4 1982 7.7-5 1982 7.7<<6 1993 7~7 7 1991

7. 7-8 1991 7.7-9 1982 7 '-10 1986 7.7-11 1982 7.8-1 1992 7.8-2 1992

VOLUME V Chapter 7 Instrumentation and Control Pacae Date Table 7.8-1 (2 pgs) 1992 Table 7.8-2 (2 pgs) 1992 Table 7.8-3 (2 pgs) 1992 Table 7.8-4 (pg 1) 1992 (pg 2) 1993 (pg 3) 1992 (pg 4) .

1992 (pg 5) 1992 Table 7.8-5 (3 pgs) 1992

VOLUME V Chapter 8 Electrical S stems Pacae Date 8.1-1 1990 F 1-2 1982 8.1-3 1990 8.1-4 1990 8.1-5 1991 8.1-6 1991 8.1-7 1991 8.1-8 1987 8.1-9 1982 Fig. 8.1-1 1986 Fig. 8 1-1A 1990 Fig. 8.1-1B 1990 Fig. 8.1-2 1993 8.2-1 1990 Fig. 8 2-1 1990 8.3-1 1990 8.3-2 1991 8.3-3 1990 8.3-4 1990 8 '-5 1990 8.3-6 1990 8 '-7 1990 8.3-8 1990 8.3-9 1990 8.3-10 1990 8.3-11 1990 Fig. 8.3-1 1990 Fig. 8.3-2 1990 Fige 8.3-3 1990 8 4-1 1990 8.4-2 1990 8.4-3 1990 Fig. 8.4-1 1992 8.5-1 1991 9.5-2 1982 8.6-1 1986 8.6-2 1982

VOLUME V Chapter 9 Auxili.ar and Emer enc S stems Pacae Date 9 '-1 1992 9.1-2 1982 9.1-3 1990 9 '-4 1982 9.2-1 1982 9 '-2 1982 9 '-3 1991 9.2-4 1986 9.2-5 1982 9.2-6 1982 9.2-7 1988 9 '-8 1987 9 '-9 1982 9.2-10 1982 9.2-11 1982 9.2-12 1982 9.2-13 1982 9 '-14 1982 9.2-15 1982 9.2-16 1990 9.2-17 1992 9.2-18 1992 9.2-19 1982 9.2-20 1982 9 '-21 1982 9.2-22 1982 9.2-23 1993 9.2-24 1988 9.2-25 1982 9.2-26 1982 9.2-27 1982 9.2-28 1982 9.2-29 1993 9.2-30 1993 9.2-31 1982 9.2-32 1982 9.2-33 1993 9.2-34 1991 9.2-35 1982 9 '-36 1982 9.2-37 1982 9.2-38 1982 Table 9.2-1 1990 Table 9.2-2 1989 Table 9.2-3 (pe 1) 1989 (pe 2) 1990 (pS 3) 1989 (pe 4) 1992 (pe 5) 1989 (pe 6) 1990

VOLUME V Chapter 9 Auxiliar and Emer enc S stems Pacae Date Table 9.2-3 (pg 7) 1989 (pg 8) 1989 (pg 9) 1989 (pg 10) 1990 (pg 11) 1989 (pg 12) 1989 (pg 13) 1990 (pg 14) 1990 (pg 15) 1989 (pg 16) 1989 (pg 1I) 1989 Table 9.2-4 (2pp) 1989 Fig. 9.2-1 1992 Fig. 9.2-2 1982 Fig. 9.2-3 1993 Fig. 9.2-4 1984 Fig. 9.2-5 ORIG Fig. 9.2-6 1982 9.3-1 1982 9.3-2 1982 9.3-3 1990 9.3-4 1991 9.3-5 1982 9.3-6 1982 9.3-7 1988 9.3-8 1988 9.3-9 1987 9.3-10 1986 9.3-11 1993 9.3-12 1990 9.3-13 1990 Table 9.3-1 1990 Table (lpp) 1990 Table 9.3-2 (2pp) 1991 Table 9.3-3 (3pp) 1990 Fig. 9.3-1 1991 9.4-1 1982 9.4-2 1982 9.4-3 1989 9.4-4 1983 9.4-5 1983 9.4-6 1982 9.4-7 1986 Table 9.4-1 1990 Table 9.4-2 (4pp) 1989 Table 9.4<<3 1989 Fig. 9.4-1 1990 9.5-1 1985 9.5-2 1993 9.5-3 1992

VOLUME V Chapter 9 Auxiliar and Emer enc S stems Pacae Date 9.5-4 1982 9.5-5 1982 9.5-6 1982 9.5-7 1982 9.5-8 1982 9 '-9 1982 Table 9.5-1 1990 Table 9.5-2 1993 Table 9.5-3 1993 Table 9.5-4 1993 Fig. 9.5-1 1993 9.6-1 1991 9.6-2 1983 9.6-3 1989 9 '-4 1987 9.6-5 1987 9.6-6 1988 9 '-7 1983 9.6-8 1983 9.6-9 1982 Fig. 9.6-1 1992 Fig. 9.6-2 1987 9.7-1 1982 9.7-2 1984 9.7-3 1982 9 '-4 1982 9 '-5 1987 9.7-6 1991 9 '-7 1991 9.7-8 1992 9.7-9 1992 9.7-10 1991 9 '-11 1991 9.7-12 1991 9.7-13 1991 9.7-14 1991 9.7-15 1992 9.7-16 1991 9 '-17 1991 9.7-18 1991 9.7-19 1991 9.7-20 1991 9 7-21 1991 9.7-22 1991 9.7-23 1991 9.7-24 1991 9.7-25 1991 Fig. 9.7-1 1982 Fig. 9.7-2 1982 Fig. 9.7-3 1991

VOLUME V Chapter 9 Auxiliar and Emer enc S stems Pacae Date 9.8-1 1993 9.8-2 1993 9.8-3 1993 9.8-4 1993 9 '-5 1993 9.8-6 1993 9.8-7 1993 9.8-8 1993 9 '-9 1993 9.8-10 1993 9 '-11 1993 9 '-12 1993 9 '-13 1993 9 '-14 1993 9 ~ 8-15 1993 9.8-16 1993 9.8-17 1993 9.8-18 1993 9.8-19 1993 9.8-20 1993 9 '-21 1993 9.8-22 1993 9.8-23 1993 9.8-24 1993 9.8-25 1987 9.8-26 1987 9.8-27 1987 9.8-28 1987 9.8-29 1993 9.8-30 1987 Table 9.8-1 1993 Table 9.8-2 (1pp) 1989 Table 9 8-2 (2pp) 1991 Table 9 ' 2 (3pp) 1991 Table 9 '-3 1989 Table 9.8-4 (3pp) 1989 Table 9 '-5 1993 Table 9.8-6 1989 Fig. 9.8-1 1993 Fig. 9.8-2 1982 Fig. 9.8-3 1991 Fig. 9.8-4 1993 Fig. 9.8-5 1993 Fig. 9.8-6 1985 Fig. 9 '-7 1982 9.9-1 1992 9.9-2 1990 9.9-3 1986 9.9-4 1990 9.9-5 1982

VOLUME V Chapter 9 Auxiliar and Emer enc S stems Pacae Date 9 '-6 1991 9.9-7 1991 9.9-8 1986 9 9-9 1985 Fig. 9.9-1 1992 Fig. 9.9-2 1989 9.10-1 1992 9.10-2 1987 9.10-3 1986 9.10-4 1987 Fig. 9.10-1 1986

VOLUME VI Steam and Power Conversion Chapter 10 S stem Pacae Date 10.1-1 1987 10.1-2 1982 10.2-1 1982 10.2-2 1983 10.2-3 1985 10.2-4 1991 10.2-5 1985 10.2-6 1988 Fig. 10.2-1 1984 Fig. 10. 2-1A 1991 Fig. 10.2-1B 1982 Fig. 10.2-1C 1992

10. 3-1 1986 10.3-2 1982 10.3-3 1986 10.3-4 1984 10 3-5 1992 10.3-6 1984 Fig. 10.3-1 1984 Fig. 10.3-1A 1984 10.4-1 ~ 1991 10.4-2 1985 10.4-3 1991 10.5-1 1986 10.5-2 1991 10.5-3 1982 10.5-4 1987 10.5-5 1990 10.5>>6 1989 10 5-7 1993 Table 10.5-1 1991 Fig. 10.5-1 1993 Fig. 10.5-2 1982 Fig. 10.5-2A 1982 Fig. 10.5-3 1982 Fig. 10.5-3A 1982 Fig. 10.5-4 1991 Fig. 10.5-4A 1991 Fig. 10.5-5 1990 Fig. 10.5-5A 1982 10.6-1 1982 10.6-2 1982 10.6-3 1982 10.6-4 1991 Table 10.6-1 1989 Fig. 10.6-1 1982

VOLUME VI Steam and Power Converse.on Chapter 10 S stem Pacae Date 10.7-1 1982 10.7-2 1982 10.8-1 1982 10.9-1 1988 10.10-1 1986 10.11-1 1983 10.11-2 1988 10.11-3 1983

VOLUME VI Waste Disposal and Chapter 11 Radiation Protection S stem Pacae Date 11.1-1 1985 11.1-2 1992 11.1-3 1982 11.1-4 1983 11.1-5 1993 11.1-6 1985 11.1-7 1992 11.1-8 1983 11.1-9 1992 11.1-10 1982

11. 1-11 1982 11.1-12 1993 11.1-13 1985 11.1-14 1985 11.1-15 1985 11.1-16 1982 11.1-17 1982 11.1-18 1992 Table 11.1-1 1989 Table 11.1-2 1990 Table 11.1-3 (pg 1) 1990 (pg 2) 1993 Table 11 1-4 1989 Table 11. 1-5 1989 Table 11.1-6 1989 Fig. 11.1-1 1988 Fig. 11.1-2 1990 Fig. 11 1-2A 1982 Fig. 11 1-2B 1985 Fig. 11.1-3 1982 Fig. 11.1-4 1988 11.2-1 1983 11 2-2 1983 11.2-3 1982 11.2-4 1982 11.2-5 1982 11.2-6 1982 11.2-7 1982 11.2-8 1982 11.2-9 1983 Table Table ll 2-1 11.2-2 1989 1989 Table 11.2-3 1989 Table 11.2-4 1989 Table 11.2-5 1989 Table 11.2-6 1989 Table 11.2-7 1989 Table 11.2-8 1989

VOLUME VI Waste Disposal and.

Chapter 11 Radiation Protection S stem Pacae Date Fig. 11.2-1 1982 11.3-1 1990 11.3-2 1986 11.3-3 1988 11.3-4 1990 11.3-5 1992 11.3-6 1992 11.3-7 1992 11.3-8 1990

11. 3-9 1990 11 3-10 1993 11.3-11 1990 11.3-12 1992 11.3-13 1990 11.3-14 1990 11.3-15 1992 11.3-16 1992 11.3-17 1993 11.3-18 1993 Table 11.3-1 (lpp) 1990 Table 11.3-1 (2pp) 1991 Table 11.3-1 (3pp) 1991 Table 11.3-1 (4pp) 1991 Table 11.3-1 (Spp) 1991 Table 11.3-2 1990 11.4-1 1993 11.4-2 1993 11.4-3 1993 11.4-4 1993 11.4-5 1993 ll. 4-6 11.4-7 1993 1993 11.4-8 1993 11.4-9 1993 11.4-10 1993 11.4-11 1993 11.4-12 1991 11.4-13 1991 Fig. 11 4-1 1993 Fig. 11.4-2 1993 Fig. 11 4-3 1992 Fig. 11.4-4 1992

11. 5-1 1987 11.5-2 1990 11.5-3 1983 Table 11.5-1 1989 Table 11.5-2 1989 Fig. 11.5-1 1983

VOLUME VI Waste Disposal and Chapter ll Radiation Protection S stem Pacae Date 11.6-1 1988 11 6-2 1988 11.6-3 1990 11.6-4 1993 Fig. 11.6-1 1990 Fig. 11.6-2a 1986 Fig. 11.6-2b 1986 Fig. 11.6-2c 1990

VOLUME VI Chapter 12 Conduct of 0 erations Pacae Date 12 ~ 1-1 1991 12.1-2 1982 12.2-1 1991 12 2-2 1988 12 '-3 1988 12.2-4 1988 12.2-5 1988 12.2-6 1988 12.2-7 1988 12 '-8 1988 12.3-1 1988 12.4-1 1983 12.5-1 1982 12 6-1 1991 12.6-2 1991 12.6-3 ORIG 12 '-1 1991

VOLUME VI Chapter 13 Initial Tests and 0 eration Pacae Date

13. 1-1 1991 13 1-2 1991 13i1-3 1991

13. 1-4 1982 Table 13.1-1 (1pp) 1989 Table 13.1-1 (2pp) 1989 Table 13.1-1 (3pp) 1991 Table 13.1-1 (4pp) 1989 Table 13.1-1 (5pp) 1991 Table 13.1-1 (6pp) 1989 Table 13.1-1 (7pp) 1991 Table 13.1-1 (8pp) 1989 Table 13. 1-1 (9pp) 1989 13.2-1 1982 13.2-2 1991 13 '-3 1991 13.2-4 1991 13.2-5 1991 13.2-6 1991 Table 13.2-1 (lpp) 1991 Table 13.2-1 (2pp) 1991 Table 13.2-1 (3pp) 1989 Table 13.2-1 (4pp) 1991 Table 13.2-1 (5pp) 1991 Table 13.2-1 (6pp) 1991 Table 13.2-1 (7pp) 1991 13.3-1 1982 13 3-2 1991 13 '-3 1982 13.3-4 1991 13.3-5 1982 Table 13.3-1 (1pp) 1989 Table 13.3-1 (2pp) 1991 Table 13.3-1 (3pp) 1991 13.4-1 1983

VOLUME VII Chapter 14 Safet Anal sis Unit 1 Pacae Date 14.0-1 1993 14.0-2 1993

14. 0-3 1993

14. 1-1 1993 14.1-2 1993 14.1-3 1993 14.1-4 1993 14.1-5 1993 14.1-6 1993 14.1-7 1993 14.1-8 1993 14.1-9 1993 14.3,-10 1993 14.1-11 1993 14.1-12 1991 14.1-13 1991 14.1-14 1991 Table 14.1-1 1993 Table 14.1-2 1993 Table 14 1-3 (2pp)

F 1993 Table 14 1-4 1993 Fig. 14.1-1 1990 Fig. 14 '-2 1990 Fig. 14.1-3 1990 Fig. 14. 1-4 1990 Fig. 14.1-5 1990 Fig. 14.1-6 1992 14.1.1-1 1993 14.1.1-2 1990 14.1.1-3 1990 14.1.1-4 1990 14.1 1-5 1990 Pig. 14 1.1-1 1990 Fig. 14. 1. 1-2 1990 14.1.2-1 1990 14.1.2-2 1990 14.1.2-3 1990 14.1.2-4 1990 14.1.2-5 1990 14 1 2-6 1990 Pig. 14 1 2-1 1990 Fig. 14.1 2-2 1990 Fig. 14.1.2-3 1990 Fig. 14.1.2-4 1990 Fig. 14.1.2-5 1990 Fig. 14 1.2-6 1990 Fig. 14 1.2-7 1990 Fig. 14.1.2-8 1990 Fig. 14.1.2-9 1990

VOLUME VII Chapter 14 Safet Anal sis Unit Pacae Date 14.1.3-1 1993 14.1.3-2 1992 14.1.3-3 1992 14.1.3-4 1992 14.1.3-5 1990

14. 1. 3-6 1990 14.1.3-7 1992 Fig. 14.1.3-1 1990 Fig. 14. 1. 3-2 1990

~ 14.1.4-1 1990 14.1.5-1 1990 14.1.5-2 1992 14.1.5-3 1992 14.1.5-4 1990 14.1.6-1 1990 14.1.6-2 1990 14.1.6-3 1990 14 '.6-4 1990 14.1.6-5 1990 14.1.6-6 1990 14.1.6-7 1990 14.1.6-8 1990 Fig. 14 1 6-1 1990 Fig. 14.1 6-2 1990 Fig. 14.1.6-3 1990 Fig. 14.1.6-4 1990 Fig. 14.1.6-5 1990 Fig. 14.1.6-6 1990 Fig. 14.1.6-7 1990 Fig. 14.1;6-8 1990 Fig. 14.1.6-9 1990 Fig. 14.1.6-10 1990 Fig. 14.1.6-11 1990 Fig. 14.1.6-12 1990 14.1.7-1 1990 14.1.7-2 1990 14.1.7-3 1990 Fig. 14.1.7-1 1982 Fig. 14.1.7-2 1982 14.1.8-1 1990 14.1.8-2 1990 14.1.8-3 1990 14.1.8-4 1990 14.1.8-5 1990 Fig. 14.1.8-1 1990 Fig. 14.1.8-2 1990 Fig. 14.1.8-3 1990 Fig. 14.1.8-4 1990 Fig. 14.1.8-5 1990

VOLUME VII Chapter 14 Safet Anal sis Unit 1 Pacae Date Fig. 14 1.8-6 1990 Fig. 14.1.8-7 1990 Fig. 14.1.8-8 1990 14.1;9-1 1990 14.1.9-2 1990 14.1.9-3 1990 Fig. 14.1.9-1 1990 Fig. 14.1.9-2 1990 14.1.10-1 1993 14.1.10-2 1993

14. 1. 10-3 1993 14.1.10-4 1993 14.1.10-5 1993 Table 14.1.10-1 1993 Table 14 1.10-2 1993 Table 14.1.10-3 1993 Table 14 1.10-4 1993 Fig. 14.1.10-1 1993 Fig. 14.1 10-2 1993

, Fig. 14.1.10-3 1993 Fig. 14.1.10-4 1993 Fig. 14.1.10-5 1993 Fig. 14 1.10-6 1993 Fig. 14.1.10-7 1993 Fig. 14.1.10-8 1993 14.1.11-1 1990 14 1.11-2 1993 14.l. 11-3 1990 14.1.11-4 1990 Fig. 14.1.11-1 1990 Fig. 14.1.11-2 1990 Fig. 14.1.11-3 1990 Fig. 14.1.11-4 1990 Fig. 14.1.11-5 1990 Fig. 14 '.11-6 1990 Fig. 14.1.11-7 1990 Fig. 14. 1. 11-8 1990 14.1.12-1 1990 14.1.12-2 1990 14.1.12-3 1990 14.1.12-4 1990 Fig. 14.1.12-1 1990 Fig. 14. 1. 12-2 1990 14.1.13-1 1991

14. 1. 13-2 1991 14.1.13-3 1989

14. l. 13-4 1982 14.1.13-5 1991 14.1.13-6 1989

VOLUME VII Chapter 14 Safet Anal sis Unit 1 Pacae Date 14.1.13-7 1990 14.1.13-8 1991 14.1.13-9 1990 14.1.13-10 1990 14 1.13-11 1990 14 1.13-12 1990 14.1.13-13 1982 14.1.13-14 1990 14 1.13-15 1991 14 1.13-16 1990 Table 14.1.13-1 1990 Fig. 14.1.13-1 1982 Fig. 14.1.13-2 1982 Fig. 14.1.13>>3 1982 Fig. 14.1.13-4 1982 Fig. 14.1.13-5 1982 Fig. 14.1 13-6 1982 14.2.1-1 1990 14.2.1-2 1990 14.2. 1-3 1990 14.2.1-4 1990 14.2.1-5 1990 14.2.1-6 1993 14.2.1-7 1993 14.2.1-8 1993 14.2.1-9 1993 14 '.1-10 1993 14.2.1-11 1993 14.2.1-12 1993 14'2.1-13 1993 14.2 1-14 1993 Table 14.2.1-1 1993 Table 14 2 1-2 1993 Table 14.2 '-3 1993 Table 14.2.1-4 1993 Table 14 2.1-5 1993 14.2.2-1 1993 14.2.2-2 1993 14.2.2-3 1993 14.2.2-4 1990 14 ' '-1 1990 14.2.3-2 1990 Table 14.2.3-1 1990 Table 14.2.3-2 1990 14.2.4-1 1990 14.2.4-2 1990 14.2.4-3 1990 14.2.4-5 1992 14.2.4-6 1990 14.2.4-7 1990 14.2.4-8 1992 14.2.4-9 1992

VOLUME VII Chapter 14 Safet Anal sis Unit 1 Pacae Date Fig. 14.2.4-1 1982 14.2 '-1 1990 14.2.5-2 1990 14.2 5-3 1990 14.2.5-4 1990 14.2.5-5 1990 14.2.5-6 1990 14.2 5-7 1990 14.2.5-8 1990 14.2.5-9 1990 Table 14.2 '-1 1990 Fig. 14.2 '-1 1990 Fig. 14.2.5-2 1990 Fig. 14.2.5-3 1990 Fig. 14.2.5-4 1990 Fig. 14.2.5-5 1990 Fig. 14.2.5-6 1990 14.2.6-1 1990 14.2.6-2 1990 14.2.6-3 1990 14.2 6-4 1990 14.2.6-5 1990 14.2.6-6 1990 14 2.6-7 1990 14.2 '-8 1990 14.2.6-9 1990 14.2.6-10 1990 14.2.6-11 1990 14.2.6-12 1990 14 2.6-13 1990 14.2.6-14 1990 14.2.6-15 1990 Table 14 2 6-1 1990 Fig. 14.2.6-1 1990 Fig. 14.2.6-2 1990 Fig. 14.2 '-3 1990 Fig. 14.2.6-4 1990 14.2.7-1 1990 14.2 '-2 1990 14.2.7-3 1990 14.2.7-4 1990 14.2.7-5 1990 14 2 7-6 1990 Table 14.2 7-1 1990 Table 14.2.7-2 1990 Table 14.2 '-3 1992 Fig. 14.2 '-1 1982 Fig. 14.2.7-2 1982 Fig. 14.2.7-3 1987

VOLUME VII 5

Chapter 14 Safet Anal sis Unit 1 Pacae Date Fig. 14.2.7-4 1987 Fig. 14.2.7-5 '1982 Fig. 14.2.7-6 1982 Fig. 14.2.7-7 1982 Fig. 14.2.7-8 1982 Fig. 14.2.7-9 1990 Fig. 14.2.7-10 1990 Fig. 14.2.7-11 1990 Fig. 14.2.7-12 1990 14.2.8-1 1990 14.2.8-2 1990 14.2 '-3 1990 14.2 '-4 1990 14.2.8-5 1990 14.2 '-6 1990 14 2.8-7 1990 Table 14.2 8-1 1990 Fig. 14.2. 8-1 1990 Fig. 14.2.8-2 1990 Fig. 14.2.8-3 1990 Fig. 14.2.8-4 1990 Fig. 14.2.8-5 1990 Fig. 14.2.8-6 1990 Fig. 14.2.8-7 1990 14.3.1-1 1993 14.3.1-2 1993 14.3.1-3 1993 14.3.1-4 1993 14.3.1-5 1993 14.3.1-6 1993 14 3.1-7 1993 14.3. 1-8 1993 14 3.1-9 1993 14.3.1-10 1990 14 3.1-11 1992 14.3.1-11a 1993 14 3.1-12 1993 14.3.1-13 1993 Table 14.3.1-1 (pg 1) 1993 Table 14.3.1-1 (pg 2) 1992 Table 14.3.1-1 (pg 3) 1990 Table 14.3.1-2 1990 Table 14.3.1-3 (2pp) 1990 Table 14.3.1-4 1990 Table 14.3.1-5 1990 Table 14.3.1-6 1990 Fig. 14.3.1-1a 1990 Fig. 14.3.1-1b 1990 Fig. 14.3.1-1c 1990

VOLUME V Chapter 14 Safet Anal ois Unit 1 Pacae Date Fig. 14 3 1-ld 1990 Fig. 14.3.l-le 1990 Fig. 14.3.l-lf 1990 Fig. 14.3.1-1g 1990 Fig. 14.3.1-2a 1990 Fig. 14.3.1-2b 1990 Fig. 14.3.1-2c 1990 Fig. 14.3.1-2d 1990 Fig. 14.3.1-2e 1990 Fig. 14.3 1-2f 1990 Fig. 14.3.1-2g 1990 Fig. 14.3.1-3a 1990 Pig. 14.3.1-3b 1990 Fig. 14.3.1-3c 1990 Fig. 14 3.1-3d 1990 Fig. 14.3.).-3e 1990 Fig. 14.3.1-3f 1990 Fig. 14.3.1-3g 1990 Fig. 14.3. 1-4a 1990 Fig. 14.3 1-4b 1990 Fig. 14.3.1-4c 1990 Fig. 14.3.1-4d 1990 Pig. 14.3.1-4e 1990 Fig. 14.3.1-4f 1990 Fig. 14.3.1-4g 1990 Fig. 14.3.1-5a 1990 Fig. 14.3.1-5b 1990 Fig. 14.3.1-5c 1990 Fig. 14 ~ 3 ~ 1-5d 1990 Fig. 14.3.1-5e 1990 Fig. 14 3.1-5f 1990 Fig. 14.3.1-5g 1990 Fig. 14.3.1-6a 1990 Fig. 14.3.1-6b 1990 Fig. 14.3.1-6c 1990 Fig. 14.3. 1-6d 1990 Fig. 14.3.1-6e 1990 Pig. 14.3.1-6f 1990 Fig. 14.3.1-6g ,1990 Pig. 14.3.1-7a 1990 Fig. 14.3.1-7b 1990 Fig. 14.3.1-7c 1990 Fig. 14.3.1-7d 1990 Fig. 14.3.1-7e 1990 Fig. 14.3.1-7f 1990 Fig. 14.3.1-7g 1990 Fig. 14.3.1-8a 1990 Fig. 14.3.1-8b 1990 Pig. 14.3.1-8c 1990

VOLUME VII Chapter 14 Safet Anal sis Unit 1 Pacae Date Fig. 14.3.1-8d 1990 Fig. 14. 3. 1-Be 1990 Fig. 14.3 1>>8f 1990 Fig. 14.3.1-89 1990 Fig. 14.3.1-9a 1990 Fig. 14.3.1-9b 1990 Fig. 14.3.1-9c 1990 Fig. 14.3.1-9d 1990 Fig. 14.3. 1-9e 1990 Fig. 14.3 1-9f 1990 Fig. 14.3.1-9g 1990 Fig. 14.3.1-10a 1990 Fig. 14.3.1-10b 1990 Fig. 14.3.1-10c 1990 Fig. 14.3.1-10d 1990 Fig. 14.3.1-10e 1990 Fig. '14.3 1-lof 1990 Fig. 14 3.1-10g 1990 Fig. 14.3.l-lla 1990 Fig. 14.3.1-11b 1990 Fig. 14.3.l-llc 1990 Fig. 14 '.l-lid 14.3.1-11e 1990 Fig. 1990 Fig. 14.3.l-llf 1990 Fig. 14.3.1-11g 1990 Fig. 14.3.1-12a 1990 Fig. 14.3.1-12b 1990 Fig. 14.3.1-12c 1990 Fig. 14.3-1-12d 1990 Fig. 14.3.1-12e 1990 Fig. 14. 3. 1-12 f 1990 Fig. 14.3.1-12g 1990 Fig. 14.3.1-13a 1990 Fig. 14.3.1-13b 1990 Fig. 14.3.1-13c 1990 Fig. 14.3.1-13d 1990 Fig. 14.3.1-13e 1990 Fig. 14.3.1-13f 1990 Fig. 14.3.1-13g 1990 Fig. 14.3 1-14 1990 Fig. 14.3.1-15 1990 Fig. 14 '.1-16 1990 Fig. 14.3.1-17 1990 Fig. 14.3.1-18 1990 Fig. 14.3 1-19 1990 Fig. 14.3.1-20 1990 Fig. 14.3.1-21 1990 Fi.g. 14.3.1-22 1990 Fig. 14..3 1-23 1990

VOLUME VII Chapter 14 Safet Anal sis Unit 1 Pacae Date 14.3.2-1 1990 14.3.2-2 1990 14.3.2<<3 1990 14 3.2-4 1990 14 3.2-5 1990 14.3.2-6 1993 14.3 2-7 1993 Table 14.3. 2-1 1990 Table 14 3.2-2 1990 Table 14 3.2-3 1993 Table 14.3.2-4 '1990 Table 14.3.2-5 1990 Table 14 3 '-6 1990 Table 14.3.2-7 1993 Fig. 14.3.2-1 1990 Fige 14.3.2-2 1990 Fig. 14.3.2-3 1990 Fig. 14.3.2-4 1990 Fig. 14.3.2-5 1990 Fig. 14.3.2-6 1990 Fig. 14 3.2-7 1990 Fig. 14.3.2-8 1990 Fig. 14.3.2-9 1990 Fig. 14.3 2-10 1990 Fig. 14.3.2-11 1990 Fig. 14.3.2-12 1990 Fig. 14.3.2-13 1990 Fig. 14.3.2-14 1990 Fig. 14.3.2-15 1990 Fig. 14.3 2-16 1990 Fig. 14.3.2-17 1990 Fig. 14.3.2-18 1990 Fig. 14.3.2-19 1990 Fig. 14.3.2-20 1990 Fig. 14.3. 2-21 1990 Fig. 14.3.2-22 1990 Fig. 14.3.2-23 1990 Fig. 14.3.2-24 1990 Fig. 14.3.2-25 1990 Fig. 14.3.2-26 1990 Fig. 14.3.2-27 1990 Fig. 14.3.2-28 1990 Fig. 14 ' '-29 1990 Fig. 14.3.2-30 1990 Fig. 14.3. 2-31 1990 Fig. 14.3.2-32 1990 Fig. 14.3.2-33 1990 Fig. 14.3.2-34 1990 Fig. 14.3.2-35 1990

VOLUME V II Chapter 14 Safet Anal sis Unit 1 Pacae Date Fig. 14.3.2-36 1990 Fig. 14.3.2-37 1990 Fig. 14.3.2-38 1990 Fig. 14.3.2-39 1990 Fig. 14.3.2-40 1990 14.3.3-1 1989 14.3.3-2 1988 14.3 4-1 1992 14.3.4-2 1992 14.3.4-3 1992 14.3.4-4 1992 14.3 '-5 1992 14.3.4-6 1992 14.3 '-7 1992 14.3.4-8 1992 14.3 4-9 1992 14 3.4-10 1992 14.3 4-11 1992 14.3.4-12 1992 14.3.4-13 1992 14.3.4-14 1992 14 3.4-15 1992 14 3.4-16 1992 14.3.4-17 1992 14.3.4-18 1992 14.3.4-19 1992 14 3.4-20 1992 14.3 4-21 1992 14.3.4-22 1992 14.3.4-23 1992 14.3.4-24 1992 14.3.4-25 1992 14.3.4-26 1992 14.3.4-27 1992 14.3.4<<28 1992 14.3.4-29 1992 14.3.4-30 1992 14.3.4-31 1992 14.3.4-32 1992 14.3.4-33 1992 14.3 '-34 1992 14.3.4-35 1992 14.3 '-36 1992 14.3.4-37 1992 14.3.4-38 1992 14.3.4-39 1992 14.3.4-40 1992 14.3.4-41 1992 14.3.4-42 1992

VOLUME VIII chapter 14 Sachet Anal sis Unit 1 Pacae Date 14 '.4-43 1992 14.3.4-44 1992 14.3.4-45 1992 14.3.4-46 1992 14.3.4-47 1992 14.3.4-48 1992 14.3.4-49 1992 14.3.4-50 1992 14.3.4-51 1992 14.3.4-52 1992 14.3.4-53 1992 14 '.4-54 1992 14.3.4-55 1992 14.3.4-56 1992 14.3.4-57 1992 14.3.4-58 1992 14.3.4-59 1992 14.3.4-60 1992 14.3.4-61 1992 14.3.4-62 1992 14."3.4-63 1992 14.3.4-64 1992

.14.3'.4-65 1992 14.3.4-66 1992 14.3.4-67 1992 14 3.4-68 1992 14.3.4-69 1992 14.3.4-70 1992 14.3.4-71 1992 14.3.4-72 1992 14.3.4-73 1992 14.3.4-74 1992 14.3.4-75 1992 14.3.4-76 1992 14.3.4<<77 1992 14.3.4-78 1992 14.3.4-79 1992 14.3.4-80 1992 14.3.4-81 1992 14.3.4-82 1992 14.3.4-83 1992 14.3.4-84 1992 14.3.4-85 1992 Table 14.3.4-1 1992 Table 14.3.4-2 1992 Table 14.3 '-3 1992 Table 14.3.4-4 (2pp) 1992 Table 14.3.4-5 (2pp) 1992

VOLUME VIII Chapter 14 Safet Anal sis Unit 1 Pacae Date Table 14.3.4-6 1992 Table 14.3.4-7 (2pp) 1992 Table 14.3.2-8 (2pp) 1992 Table 14.3.4-9 1992 Table 14.3.4-10 1992 Table 14 3 4-11 1992 Table 14.3.4-12 1992 Table 14.3.4-13 1992 Table 14.3.4-14 1992 Table 14.'3.4-15 (2 pp) 1992 Table 14.3.4-16 (2 pp) 1992 Table 14.3.4-17 1992

. Table 14.3-4-18 1992 Table 14.3.4-19 1992 Table 14.3.4-20 1992 Table 14.3.4-21 (2pp) 1992 Table 14.3.4-22 1992 Table 14.3.4-23 1992 Table 14 ~ 3 ~ 4-24 1992 Table 14 3 4-25 (3pp) 1992 Table 14.3.4-26 1992 Table 14.3.4-27 (10pp) 1992 Table 14.3.4-28 1992 Table 14.3.4-29 (2pp) 1992 Table 14.3.4-30 1992 Table 14. 3. 4-31 (2pp) 1992 Table 14.3.4-32 (2pp) 1992 Table 14.3.4-33 (2pp) 1992 Table 14.3.4-34 (2pp) 1992 Table 14.3.4-35 (2pp) 1992 Table 14.3 '-36 (2pp) 1992 Table 14.3.4-37 1992 Table 14.3 '-38 1992 Table 14.3.4-39 (2pp) 1992 Table 14.3.4-40 (2pp) 1992 Table 14.3.4-41 1992 Table 14.3.4-42 1992 Table 14.3.4-43 1992 Table 14.3.4-44 1992 Table 14.3.4-45 1992 Table 14.3.4-46 (2pp)

Fig. 14.3. 4-1 1992'992 Fig. 14.3.4-2 Fig. 14.3.4-3 1992 Fig. 14.3.4-4 1992 Fig. 14.3 '-5 1992 Fig. 14.3.4-6 1992 Fig. 14.3.4-7 1992

VOLUME VIII Chapter 14 Safet Anal sis U it 1 Pacae Date Fig. 14.3.4-8 1992 Fig. 14.3.4-9 1992 Fig. 14.3.4-10 1992 Fig. 14.3.4-11 1992 Fig. 14.3 4-12 1992 Fig. 14.3.4-13 1992 Fig. 14 3.4-14 1992 Fig. 14.3.4-15 1992 Fig. 14.3.4-16 1992 Fig. 14 '.4-17 1992 Fig. 14.3.4-18 1992 Fig. 14.3.4-19 1992 Fig. 14.3.4-20 1992 Fig. 14.3.4-21 1992 Fig. 14.3.4-22 1992 Fig. 14.3.4-23 1992 Fig. 14.3.4-24 1992 Fig. 14.3.4-25 1992 Fig. 14.3.4-26 1992 Fig. 14.3.4-27 1992 Fig. 14.3.4-28 1992 Fig. 14.3.4-29 1992 Fig. 14.3.4-30 1992 Fig. 14.3.4-31 1992 Fig. 14.3.4-32 1992

'Fig. 14.3.4-33 1992 Fig. 14.3.4-34 1992 Fig. 14.3.4-35 1992 Fig. 14.3.4-36 1992 Fig. 14.3.4-37 1992 Fig. 14.3.4-38 1992 Fig. 14.3.4-39 1992 Fig. 14.3.4-40 1992 Fig. 14.3.4-41 1992 Fig. 14 ' 4-42 1992 Fig. 14.3.4-43 1992 Fig. 14.3.4-44 1992 Fig. 14.3.4.45 1992 Fig. 14.3.4-46 1992 Fig. 14.3.4-47 1992 Fig. 14.3.4-48 1992 Fig. 14.3.4-49 1992 Fig. 14.3.4-50 1992 Fig. 14.3 4-51 1992 Fig. 14.3.4-52 1992 Fig. 14.3.4-53 1992 Fig. 14 3.4-54 1992 Fig. 14.3.4-55 1992 Fig. 14.3.4-56 1992 Fig. 14.3.4-57 1992

VOLUME VIII Chapter 14 Safet Anal sis Unit 1 Pacae Date Fig. 14.3 '-58 1992 Fig. 14.3.4-59 1992 Fig. 14.3.4-60 1992 Fig. 14.3.4-61 1992 Fig. 14.3.4-62 1992 Fig. 14.3.4-63 1992 Fig. 14.3.4-64 1992 Fig. 14.3.4-65 1992 Fig. 14 3.4-66 1992 Fig. 14.3.4-67 1992 Fig. 14.3.4-68 1992 Fig. 14.3.4-69 1992 Fig. 14.3.4-70 1992 Fig. 14.3. 4-71 1992 Fig. 14.3.4-72 1992 Fig. 14.3.4-73 1992 Fig. 14.3.4-74 1992 Fig. 14.3.4-75 1992 Fig. 14.3.4-76 1992 Fig. 14.3.4-77 1992 Fig. 14 '.4-78 1992 Fig. 14.3.4-79 1990 Fig. 14.3.4-80 1992 Fig. 14.3 '-81 ~

1992 Fig. 14.3.4-82 1992 Fig. 14.3.4-83 1992 Fig. 14.3.4-84 1992 Fig. 14.3.4-85 1992 Fig. 14.3.4-86 1992 Fig. 14.3.4-87 1992 Fig. 14 3.4-88 1992 Fig. 14.3.4-89 1992 Fig. 14 3 4-90 1992 Fig. 14.3.4-91 1992 Fig. 14.3.4-92 1992 Fig. 14.3.4-93 1992

VOLUME TX Chapter 14 Safet Anal sis Unit 1 Pacae Date 14.3.5-1 1986 14.3 '-2 1982 14.3.5-3 1986

'14.3.5-4 1982 14 ' '-5 1982 14.3.5-6 1982 14.3 5-7 1982 14.3.5-8 1982 14.3.5-9 1983 14.3.5-10 1982 14.3.5-11 1983 14.3.5-12 1986 14.3.5-13 1982 14.3.5-14 1986 14.3.5-15 1982 14.3.5-16 1982 14.3.5-17 1982 14.3.5-18 1983 14.3.5-19 1982 14 3 5-20 1982 14.3.5-21 1982 14.3.5-22 1982 14.3 5-23 1982 14.3.5-24 1982 14.3.5-25 1982 14.3.5-26 1987 14.3.5-27 1987 14.3.5-28 1987 14.3.5-29 1987 14.3.5-30 1987 14.3.5-31 1987 14.3.5-32 1987 Table 14.3.5-1 1990 Table 14.3.5-2 1990 Table 14.3.5-3 1990 Table 14.3.5-4 1990 Table 14.3.5-5 1990 Table 14.3.5-6 1990 Table 14.3.5-7 1990 Table 14.3.5-8 1990

VOLUME IX Chapter 14 Safet Anal sis Unit 1 Pacae Date Table 14.3.5-9 pg 1 1990 pg 2 1990 pg 3 1990 pg 4 1990 pg 5 1990 pg 6 1990 Fig. 14.3.5-1 1982 Fig. 14.3.5-2 1982 Fig. 14.3.5-3 1982 Fig. 14.3.5-4 1982 Fig. 14.3.5>>5 1987 Fig. 14.3.5-6 1987 14.3.6-1 1989 14.3.6-2 1982 14.3.6-3 1991 14.3.6-4 1989 14.3.6-5 1982 14.3.6-6 1982 14.3.6-7 1982 14.3.6-8 1982 14.3.6>>9 1982 14.3.6-10 1982 14.3.6-11 1990 14.3 6-12 1990 14 ~ 3 ~ 6-13 1990 14.3.6-14 1990 14.3.6-15 1990 14.3.6-16 1990 14.3.6-17 1990 14 3.6-18 1990 14.3.6-19 1990 14 3 6-20 1990 14.3.6-21 1990 14 '.6-22 1990 14.3.6-23 1990 14.3.6-24 1990 14.3.6-25 1990 14.3.6-26 1990 14.3.6-27 1991 14.3.6-28 1990 14.3.6-29 1991 Table 14 '.6-1 1990 Table 14.3.6-2 1990 Table 14 '.6-3 1990 Table 14.3.6-4 1990 Table 14.3.6-5 1990 Table 14.3.6-6 1990 Table 14.3.6-7 1990 Table 14.3.6-8 1990

VOLUME I Chapter 14 Saiet Anal sis Unit 1 Pacae Date Table 14.3.6-9 1990 Table 14.3.6-10 1990 Table 14.3.6-11 1990 Table 14.3.6-12 1990 Table 14.3.6-13 1990 Table 14.3.6-14 1990 Table 14.3.6-15 1990 Table 14.3.6-16 1990 Table 14.3.6-17 (2pp) 1990 Fig. 14.3.6-1 1982 Fig. 14.3.6-2 1982 Fig. 14.3.6-3 DELETED Fig. 14.3.6-4 DELETED Fig. 14.3.6-5 DELETED Fig. 14 3.6-6 1990 Fig. 14.3.6-7 1990 Fig. 14.3.6-8 1990 Fig. 14.3.6-9 1990 Fig. 14 3 6-10 1990 Fig. 14.3.6-11 1990 Fig. 14.3.6-12 1990 Fig. 14.3 6-13 1990 Fig. 14.3.6-14 1990 Fig. 14. 3. 6-14A 1990 Fig. 14.3 6-15 1990 Fig. 14.3.6-16 1990 Fig. 14.3.6-17 1990 Fig. 14.3.6-18 1990 Fig. 14.3.6-19 1990 Fig. 14.3.6-20 1990 Fig. 14.3.6-21 1990 Fig. 14 3.6-22 1990 14 3 7-1 1993 14.3. 8-1 1993 14.4.1-1 1990 14.4.2-1 1987 14.4.2-2 1982 14.4.2-3 1982 14.4 2-4 1982 14.4.2>>'5 1982 14.4.2-6 1982 14.4.2-7 1982 14.4.2-8 1982 14.4.2-9 1982 14.4.2-10 1982 14.4.2-11 1982 14.4.2-12 1982 14.4.2-13 1987 14.4.2-14 1982

VOLUME IX Chapter 14 Safet Anal sis Unit 1 Pacae Date 14.4.2-15 1982 14.4.2-16 1982 14.4. 2-17 1982 14.4.2-18 1982 14.4.2-19 1982 14.4.2-20 1982 14 4 2-21 1982 14.4.2-22 1990 14.4.2-23 1982 14.4.2-24 1982 14.4.2-25 1982 14.4.2-26 1982 14.4.2-27 1982 14.4.2-28 1990 Table 14.4.2-1 (8pp) 1990 Table 14.4.2-2 1990 Table 14.4 2-3 1990 Table 14.4.2-4 1990 Table 14.4.2-5 (4pp) '990 Fig. 14.4.2-1 1982 Fig. 14.4.2-2 1982 pig. 14.4.2-3 1982 Fig. 14 4.2-4 1982 Fig. 14.4.2-5 1982 Fig. 14.4.2-6 1982 pig. 14.4.2-7 1982 Fig. 14 4.2-8 1982 Fig. 14.4.2-9 1982 Fig. 14.4.2-10 1982 Fig. 14.4 2-11 1982 Fig. 14.4.2-12 1982 Fig. 14 '.2-13 1982 pig. 14.4.2-14 1982 Fig. 14.4.2-15 1982 Fig. 14.4.2-16 1982 Fig. 14.4.2-17 1982 Fig. 14.4.2-18 1982 Fig. 14.4.2-19 1982 Fig. 14.4.2-20 1982 Fig. 14 '.2-21 1982 14 4.3-1 1990 14.4.3-2 1990 14.4.3-3 1993 14.4.3-4 1990 14.4.3<<5 1990 14.4.4-1 1990 Table 14 ' 4-1 1990

~ Table 14.4.4-2 1990

VOLUME IX Chapter 14 Safet Anal sis Unit 1 Pacae Date Table 14.4.4-3 1990 Table 14.4.4-4 (2pp) 1990 Table 14.4.4-5 (3pp) 1990 Table 14.4.4-6 1990 14.4.5-1 1990 14.4.5-2 1990 14.4.6-1 1990 14.4.6-2 1987 14.4.6-3 1987 14.4.6-4 1990 14.4.6-5 1993 14.4.6-6 1990 14.4.6-7 1993 Table 14.4.6-1 1990 Table 14.4.6-2 1990 Table 14.4.6-3 1993 Table 14.4 '-3a 1993 Table 14.4.6-4 1993 Table 14.4.6-4a 1993 Table 14.4.6-5 1993 Table 14.4+6-5a ,1993 Table 14.4 6-6 1990 Table 14.4.6-7 1990 Table 14.4.6-8 1990 Table 14.4.6-9 1990 Table 14.4.6-10 1990 Table 14.4. 6-11 19.90 Table 14.4.6-12 1990 Table 14.4 6-13 1990 Table 14.4.6-14 1990 Table 14.4.6-15 1990 Table 14.4.6-15a 1990 Table 14.4.6-16 1990 Table 14.4.6-16a 1990 Table 14.4.6-17 1990 Table 14.4.6-18 (5pp) 1990 Table 14.4.6-19 1990 Table 14.4.6-20 1990 Fig. 14.4.6-1 1982 Fig. 14.4.6-2 1982 Fig. 14.4.6-3 1982 Fig. 14.4.6-4 1982 Fig. 14.4.6-5 1993 Fig. 14 '.6-6 1993 Fig. 14.4 '-7 1987 Fig. 14.4.6-8 1987 Fig. 14.4.6-9 1993 Fig. 14.4.6-9a 1993 Fig. 14.4.6-9b 1993

VOLUME IX Chapter 14 Safet Anal sis Unit 1 Pacae Date Fig. 14.4.6-10 1993 Fig. 14.4.6-10a 1993 Fig. 14.4 6-11 1993 Fig. 14.4.6-11a 1992 14.4.7-1 1990 14.4.7-2 1990 14 4.8-1 1990 14.4 9-1 1990

'14.4.9-2 1990 14.4.9-3 1990 Fig. 14.4.9-1 1982 Fig. 14.4.9-2 1982 14.4.10-1 1990 14.4 10-2 1990 14.4.10-3 1990 14.4.10-4 1990 Table 14.4.10-1 1990 14 4 11-1 1990 14.4.11-2 1990 14.4.11-3 1990 14.4.11-4 1990 14.4.11-5 1990 14.4.11-6 1990 14.4. 11-7 1990 Table 14.4.11-1 (pg 1) 1993 (pg 2) 1993 (pg 3) 1990 (pg 4) 1993 (pg 5) 1993 (pg 6) 1992 (pg 7) 1993 (p'g 8) 1990 (pg 9) 1993 (pg 10) 1993 (pg>>) 1993 Table 14 4.11-2 1990 Table 14.4.11-3 1990 Table 14.4 11-4 1990 Table 14.4 11-5 1990 Table 14.4.11-6 1990 Table 14.4.11-7 1990 Table 14 4.11-8 1990 Table 14 4 11-9 1992 14A-1 1992 14A-2 1982 14A-3 1982 14A-4 1982 14A-5 1982 14A-6 1982 14A-7 1982 14A-8 1982 14A-9 1982 14A-10 1982

VOLUME IX Chapter 14 Safet Anal sis Unit 1 Pacae Date 14A-11 1982 14A-12 1992 14A-13 1992 14A-14 1992 14A-15 1992 14A-16 1993 14A-17 1993 14A-18 1992 14A-19 1992 14A-20 1992 14A-21 1992 14A-22 1992 14A-23 1992 14A-24 1992 14A-25 1992 14A-26 1992 14.G-1 1987 14.G-2 1987 14.G-3 1987 14.G-4 1987 14 G-5 1987 14.G-6 1987 14.G-7 1987 14 G-8 1987 14.G-9 1987 14.G-10 1987 14.G-ll 1987 14 G-12 1988 Table 14.G-1 1987 Table 14 G-2 1987 Table 14.G-3 1987 Fig. 14. G-1 (3pp) 1988 Fig. 14.G-1 Notes 1988 Fig. 14.G-2 1987

VOLUgg~

Chapter 14 Safet Anal sis Unit 2 Pacaa Date 14 0-1 1993 14.0-2 1993 14.0-3 1993 14.0-4 1993 14.0-5 1993 14.1-1 1993 14.1-2 1993 14.1-3 1993 14.1-4 1993 14 '-5 1993 14.1-6 1993 14.1-7 1993

14. 1-8 1993 14.1-9 1993 14.1-10 1993

14. 1-11 1993 14.1-12 1993 14.1-13 1993 14.1-14 1991 Table 14.1.0-1 (4 pgs) 1993 Table 14.1.0-2 (3 pgs) 1993 Table 14.1.0-3 1993 Table 14.1.0-4 1993 Table 14.1.0-5 (2 pgs) 1993 Table 14.1.0-6 (3 pgs) 1993 Fig 14.1.0-1 1993 Fig. 14.1.,0-2 1993 Fig. 14.1.0-3 1993 Fig. 14 '.0-4 1993 Fig. 14.1.0-5 1993 Fig 14.1. 0-6 1993 Fig. 14. 1- 0-7 1993 14.1.1-1 1993 14.1.1-2 1991 14.1.1-3 1991 14.1.1-4 1991 14.1.1-5 1991 14.1.1-6 1991 14.1 References 1991 Table 14.1.1<<1 1991 Fig. 14.1 l-l 1992

~VO Ug~

Chapter 14 Safet Anal sis Unit 2 Pacae Date Fig. 14 1 1-2 1992 14.1.2A-1 1992 14.1.2A-2 1992 14.1.2A-3 1991

14. 1. 2A-4 1991 14 1.2A-5 1991 14 1 ~ 2A-6 1991 14 1 2A-7 1991 14.1.2A-S 1991 Fig. 14 1 2A-1 1991 Fig. 14 1.2A-2 1991 Fig. 14.1.2A-3 1991 Fig. 14 1 'A-4 1991 Fig. 14.1.2A-5 1991 Fig. 14. 1. 2A-6 1991 Fig. 14 1.2A-7 1991 Fig. 14.1 2A-8 1991 Fig. 14 1.2A-9 1991 14.1.2B-1 1991 14 1.2B-2 1993 14.1.2B-3 1991 14 1 2B-4 1991 14 1.2B-5 1991 14 ' 'B-6 1991 14.1 2B-7

~ 1991 Fig. 14.1.2B-1 1991 Fig. 14.1.2B-2 1991 Fig. 14.1.2B-3 1991 Fig. 14.1.2B-4 1991 Fig. 14.1.2B-5 1991 Fig. 14 1.2B-6 1991 Fig. 14.1.2B-7 1991 Fig. 14.1.2B-S 1991 Fig. 14.1.2B-9 1991 14.1.3-1 1993 14.1.3-2 1991 14.1 '-3 1991 14.1.3-4 1993 14.1.3-5 1991 14.1 3-6 1991 14.1.3-7 1991 14.1.3.4 Refs. 1991 Fig. 14 '.3-1 1991 Fig. 14.1.3-2 1991 14.1.4-1 1991 14.1.5-1 1991 14 1.5-2 1991 14 1.5-3 1991

14. 1. 5-4 1991 14.1.5-5 1991

VOLUME X Chapter 14 Safet Anal sis Unit 2 Pacae Date 14.1.5-6 1991 14.1.5-7 1991 14 1.6-1 1991 14.1.6-2 1991 14.1.6-3 1991 14.1.6-4 1991 14.1.6-5 1991 14.1.6-6 1991 14.1.6-7 1991 14.1.6-8 1991 14 1 6-9 1991 14.1.6-10 1991 14 1.6-11 1991 Table 14 1.6-1 1991 Table 14.1.6-2 1991 Fig. 14.1.6-1 1991 Fig. 14.1.6-2 1991 Fig. 14.1.6-3 1991 Fig. 14.1.6-4 1991 Fig. 14. 1. 6-5 1991 Pig. 14 '.6-6 1991 Fig. 14 1.6-7 1991 Fig. 14.1.6-8 1991 Fig. 14.1.6-9 1991 Fig. 14 1.6-10 F 1991 Fig. 14.1.6-11 1991 Fig. 14.1.6-12 1991 14.1 7-1 1991 14.1.7-2 1991 14.1.7-3 1991 14.1.7-4 1991 Table 14 1.7-5 1991 Fig. 14.1.7-1 1992 14 1 SA-1 1991 14.1.8A-2 1991 14 1.8A-3 1991 14.1.8A-4 1991 14.1.8A-S 1991 14.1.8A-6 1991 Table 14.1.8A-1 (2pp) 1991 Fig. 14.1.8A-1 1991 Fig. 14 1 SA-2 1991 Pig. 14.1.8A-3 1991 Fig. 14.1.8A-4 1991 Fig. 14.1.8A-5 1991 Fig. 14. 1. 8A-6 1991 Fig. 14.1 SA-7 1991 Fig. 14. 1. SA-S 1991 Fig. 14.1.8A-9 1991

VOLUME X Chapter 14 Safet Anal sis Unit 2 Pacae Date Fig. 14 1 8A-10 1991 Fig. 14 1. SA-11

~ 1991 Fig. 14.1.8A-12 1991 14 1 'B-1 1991 14.1 SB-2 1991 14 1.8B-3 1991 14 1 8B-4 1991

14. 1. 8B<<5 1993 14.1. 8B-6 1991 14 1 8B-7 1991 14.1.8B-7 1991 14 F 1 8B-8 1991 Fig. 14.1.8B-1 1991 Fig. 14.1.8B-2 1991 Fig. 14.1.8B-3 1991 Fig. 14.1.8B-4 1991 Fig. 14 1.8B-5 1991.

Fig. 14.1.8B-6 1991 Fig. 14. 1. 8B-7 1991 Fig. 14.1.8B-S 1991 Pig. '4.1.88-9 1991 Fig. 14 1 SB-10 1991 Pig. 14.1.8B-11 1991 Fig. 14.1 8B-12 1991 14.1.9-1 1991 14.1.9-2 1991 14.1.9-3 1991 14.1.9-4 1991 14.1.9-5 1991 14.1.9-6 1991 Table 14.1 9-1 1991 Fig. 14.1.9-1 1992 Fig. 14.1.9-2 1992 Pig. 14 1.9-3 1992 14.1.10A-1 1991

14. 1. 10A-2 1991 14.1.10A-3 1993 14 1 10A-4 1993 14.1 10A-5 1993 14.1 10A-6 1993 14 ~ 1 ~ 10A-7 1993 14.1.10A-S 1993 14.1.10A-9 1991 Table 14.1.10A-1 1993 Table 14.1.10A-2 1993 Table 14.1.10A-3 1993 Table 14. 1. 10A-4 1993 Pig. 14.1.10A-1 1993 Fig. 14.1.10A-2 1993

VOLUME X Chapter 14 Sachet Anal eis Un t 2 Pacae Date Fig. 14.1.10A-3 1993 Fig. 14 1.10A-4 1993 Fig. 14. 1. 10A-5 1993 Fig. 14 1.10A-6 1993 Fig. 14.1.10A-7 1993 Pig. 14 1 10A-8 1993 14 1.10B-1 1991 14.1.10B-2 1991

14. 1. 10B-3 1993 14.1. 10B-4 1993

14. 1. 10B-5 1991 14 1.10B-6 1993 14.1.10B-7 1993 14 1 10B-8 1991 14.1 10B-9 1991 Table 14.1.10B-1 1993 Table 14.1.10B-2 1993 Table 14.1.10B-3 1993 Table 14. 1. 10B-4 1993 Fig. 14.1 10B-1 1993 Fig. 14 '.10B-2 1993 Pig. 14.1.10B-3 1993 Fig. 14. l. 10B-4 1993 Fig. 14 1.10B-S 1993 Fig. 14. l. 10B-6 1993 Fig. 14.1 108-7 1993 Pig. 14 1 ~ 108-8 1993 14 1.11A-1 1991 14.1.11A-2 1993 14.1.11A-3 1991 14.1.11A-4 1993 14.1.11A-5 1991 14.1 11A-6 1991 Table 14.1 llA-1 1991 Fig. 14.1.11A-1 1991 Fig. 14 1 11A-2 1991 Fig. 14 1.1IA-3 1991 Fig. 14.1.11A-4 1991 Fig. 14 1.11A-5 1991 Fig. 14.1 llA-6 1991 Fig. 14.1.11A-7 1991 Fig. 14 1 llA-8 1991

14. 1 11B-1 1991 14 l. 11B-2 1993 l.

14. 11B-3 1991

14. 1. 11B-4 1991

14. 1. 11B-5 1991 Table 14. 1 11B-1 1991 Fig. 14.1.11B-1 1991 Fig. 14.1.11B-2 1991

VOLUME X Chapter 14 Safet Anal sis Unit 2 Pacae Fig. 14.1.11B-3 1991 Fig. 14.1.11B-4 1991 Fig. 14. 1. 11B-5 1991 Fig. 14 1.11B-6 1991 Fig. 14 ' llB-7 1991 Fig. 14.1.11B-8 1991 14 1 12-1 1991

14. l. 12-2 1991 14 1.12-3 1991 14.1.12-4 1991 14 1 12-5 1991 Table 14.1 12-1 1991 Fig. 14. 1. 12-1 1992 Fig. 14.1.12-2 1992 14.1.13-1 1991 14.2-1 1993 14 2.1-1 1993 14.2.2-1 1993 14.2.2-2 1993 14 2.2-3 1993 "14.2 2-4 1991 14.2.2-5 1991 Table 14.2 2-1 1991 Table 14.2.2-2 1993 Table 14.2 2-3 1993 14.2.3-1 1991 14.2.4-1 1991 14.2.4-2 1991 14.2.4-3 1991 14 2 4-4 1991 14.2.4-5 1991 14.2.4-6 1991 14.2.4-7 1991 14.2 4-8 1991 14 2.4-9 1991 Table 14.2.4-1 1991 14.2.5-1'ate 14.'2.5-1 14.2 5-2 1991 1991 14.2.5-3 1991 14 2.5-4 1991 14 2.5-5 1991 14.2.5-6 1991 14.2.5-7 1991 14.2.5-8 1991 14 '.5-9 1991 14 2 5-10 1991 14 2 5-11 1991 Table 14 2 5-1 1991 Table 14.2.5-2 (2pp) 1991 Fig. 1991

VOLUME X Chapter 14 Safet Anal sis Unit 2 Pacae Date Fig. 14.2.5-2 1991 Fig. 14.2 '-3 1992 Fig. 14.2.5-4 1992 Fig. 14.2.5-5 1991 Fig.- 14.2.5-6 1991 14.2. 6-1 1991 14 2.6-2 1991 14 2.6-3 1991 14.2.6-4 1991 14.2.6-5 1991 14.2.6-6 1991 14.2.6-7 1991 14.2 '-8 1991 14.2.6-9 1991 14.2;6-10 1991 14 2 ~ 6-11 1991 14.2.6-12 1991 14.2.6-13 1991 14.2 6-14 1991 14.2.6-15 1991 14 2 6-16 1991 14.2.6-17 1991 Table 14.2 '-1 1991 Fig. 14.2.6-1 1991 Fig. 14.2.6-2 1991 14.2.7-1 1993 14 2.8-1 1991 14.2.8-2 1991 14.2.8-3 1991 14.2.8-4 1991 14.2.8-5 1991 14.2 '-6 1991 Table 14.2.8-1 (3pp) 1991 Fig. 14.2.8-1 1991 Fig. 14.2.8-2 1991 Fig. 14.2.8-3 1991 Fig. 14.2.8-4 1991 Fig. 14.2.8-5 1991 Fig. 14.2.8-6 1991 Fig. 14.2 8-7 1991 Fig. 14.2 '-8 1991

VOLTE~

chapter 14 Safet Anal sis Unit 2 Pacae Date 14.3.1-1 1992 14.3.1-2 1991 14.F 1-3 1991 14.3.1-4 1991 14.3 1-5 1991

14. 3 ~ 1-6 1991 14.3 1-7 1991 14.3 '-8 1991 14.3.1-9 1993 14.3.1-10 1993 14.3.1-11 1993 14.3.1-12 1992 14.3. 1-13 1992 14.3.1-14 1992 14.3.1-14a 1993 14.3.1-15 1991 14.3.1-16 1991 14.3.1-17 1993 Table 14.3.1-1 (2pp) 1991 Table 14.3.1-2 1991 Table 14.3 1-3 1991 Table 14 ~ 3 ~ 1-4 (2pp) 1991 Table 14.F 1-5 1992 Table 14.3 1-6 1993 Table 14.F 1-7 1992 Table 14.3 1-8 1992 Table 14.3.1-9 1992 Table 14.3.1-10 1992 Fig. 14.3 l-l 1991 Fig. 14.3 '-2 1991 Fig. 14 '.1-3 1991 Fig. 14.3 1-4 1991 Fig. 14.3 1-5 1991 Fig. 14.3 '-6 1991 Fig. 14 '.1-7 1991 Fig. 14.3.1-8 1991 Fig. 14.3.1-9 1991 Fig. 14.3.1-10 1991 Fig. 14.3.1-11 1991 Fig. 14.3.1-12 1991 Fig. 14.3.1-13 1991 Fig. 14.3.1-14 1991 Fig. 14.3. 1-15 1991 Fig. 14.3.1-16 1991 Fig. 14.3.1<<17 1991 Fig. 14.3.1-18 1991 "-

Fig. 14.3.1-19 1991 Fig. 14.3.1-20 1991 Fig. 14. 3. 1-21 1991

~VOLUME X Chapter 14 Safet A al sis Unit 2 Pacae Date Fig. 14.3.1-22 1991 Fig. 14.3.1-23 1991 Fig. 14.3.1-24 1991 Fig. 14.3. 1-25 1991 Fig. 14 3 1-26 1991 Fig. 14.3 1-27 1991 Pig. 14.3. 1-28 1991 Pig. 14.3. 1-29 1991 Fig. 14 3 1-30 1991

14. 3 ~ 1-24 1991 14.3.1-25'4.3.1-26 1991 1991 14.3.1-27 1991 14 3 1-28 1991 14.3.1-29 1991 14.3.1-30 1992 14.3.1-31 1992 14.3.1-32 1992 14.3.1-33 1992 14.3. 1-34 1992 14.3.1-35 1992 14.3.1-36 1992 14.3.1-37 1992 14.3.1-38 1992 14.3 '-39 1992 14 3.1-40 1992 14.3.1-41 1992 14.3.1-42 1992 Table 14.3.1-11 (2pp) 1992 Table 14 3 1-12 1992 Table 14 3.1-13 1992 Table 14.3. 1-14 1992 Table 14.3.1-15 1992 Table 14 3 1-16 (2pp) 1992 Fig. 14 3.1-31 1991 Fig. 14.3.1-32 1991 Fig. 14.3 '-33 1991 Fig. 14.3.1-34 1991 Fig. 14.3.1-35 1991 Fig. 14.3.1-36 1991 Fig. 14.3.1-37 1991 Fig. 14.3.1-38 1991 Pig. 14.3.1-39 1991 Fig. 14 3 1-40 1991 Fig. 14.3.1-41 1991 Fig. 14.3.1-42 1991 Fig. 14.3.1-43 1991 Fig. 14.3.1-44 1991 Fig. 14.3.1-45 1991

VOLUME XI Chapter 14 Safet Anal sis Unit 2 Pacae Date Fig. 14 3 1-46 1991 Fig. 14.3.1-47 1991 Fig. 14.3.1-48 1991 Fig. 14 3.1-49 1991 Fig. 14.3.1-50 1991 Fig. 14 ~ 3 1-51 1991 Fig. 14.3.1-52 1991 Fig. 14.3.1-53 1991 Fig. 14.3.1-54 1991 Fig. 14 '.1-55 1991 Fig. 14 3 1-56 1991 Fig. 14.3 1-57 1991 Fig. 14.3.1-58 1991 Fig. 14.3.1-59 1991 Fig. 14.3.1-60 1991 Fig. 14.3.1-61 1991 Fig. 14.3.1-62 1991 Fig. 14.3.1-63 1991 Fig. 14.3.1-64 1991 Fig. 14.3.1-65 1991 Fig. 14.3.1-66 1991 Fig. 14.3.1-67 1991 Fig. 14.3.1-68 1991 Fig. 14.3.1-69 1991 Fig. 14 3.1-70 1991 Fig. 14.3.1>>71 1991 Fig. 14.3.1-72 1991 Fig. 14.F 1-73 1991 Fig. 14.3.1-74 1991 Fig. 14 3.1-75 1991 Fig. 14.3.1-76 1991 Fig. 14.3.1-77 1991 Fig. 14 3 1-78 1991 Fig. 14.3.1-79 1991 Fig. 14 3.1-80 1991 Fig. 14.3.1-81 1991 Fig. 14.3 1-82 1991 Fig. 14.3 1-83 1991 Fig. 14.3.1-84 1991 Fig. 14 3.1-85 1991 Fig. 14 3.1-86 1991 Fig. 14.3.1-87 1991 Fig. 14 3 1-88 1991 Fig. 14.3 1-89 1991 14.3.2-1 1991 14.3.2-2 1991 14.3.2-3 1991 14.3 '-4 1991 14.3.2-5 1991

VOLUME XI Chapter 14 Safet Anal sis Unit 2 Pacae Date 14.3.2-6 1991 14.3.2-7 1992 14.3.2-7a 1993 14.3.2-8 1993 Table 14.3.2-1 1991 Table 14.3.2-2 1991 Table 14.3.2-3 1992 Table 14.3.2-4 1993 Table 14.3.2-5 1991 Table 14.3.2-6 1991 Table 14.3.2-7 1992 Table 14.3.2-8 1993 Table 14.3.2-9 1991 Fig. 14.3.2-1 1991 Fig. 14.3.2-2 1991 Fig. 14.3.2-3 1991 Fig. 14 3.2-4 1991 Fig. 14 3 2-5 1991 Fig. 14.3.2-6 1991 Fig. 14.3.2-7 1991 Fig. 14.3.2-8 1991 Fig. 14.3.2-9 1991

'ig. 14.3.2-10 1991 Fig. 14.3.2-11 1991 Fig. 14.3.2-12 1991 Fig. 14.3.2-13 1991 Fig. 14.3.2-14 1991 Fig. 14.3.2-15 1991 Fig. 14 3.2-16 1991 Pig. 14.3.2-17 1991 Fig. 14 3 2-18 1991 Fig. 14.3.2-19 1991 Fig. 14.3.2-20 1991 Fig. 14.3.2-21 1991 Fig. 14 3.2-22 1991 Fig. 14.3.2-23 1991 Fig. 14 3.2-24 1991 Fig. 14 3.2-25 1991 Fig. 14.3.2-26 1991 Pig. 14 3.2-27 1991 Fig. 14.3.2-28 1991 Pig. 14 3.2-29 1991 Fig. 14.3.2-30 1991 Fig. 14.3.2-31 1991 Fig. 14.3.2-32 1991 Fig. 14.3.2-33 1991 Fig. 14.3.2-34 1991 Fig. 14.3.2-35 1991 Fig. 14.3.2-36 1991

VOLUME XI Chapter 14 Safet Anal sis Unit 2 Pacae Date Fig. 14.3.2-37 1991 Fig. 14.3.2-38 1991 Fig. 14.3.2-39 1991 Fig. 14 3.2-40 1991 Fig. 14 3.2-41 1991 Fig. 14.3.2-42 1991 Fig. 14.3.2-43 1991 Fig. 14.3.2-44 1991 Fig. 14.3 2-45 1991 Fig. 14.3.2-46 1991 Fig. 14.3.2-47 1991 Fig. 14.3.2-48 1991 Fig. 14.3.2-49 1991 Fig. 14.3.2-50 1991 Fig. 14.3.2-51 1991 Fig. 14.3.2-52 1991 Fig. 14.3.2-53 1991 Fig. 14.3 2-54 1991 Fig. 14 ' '-55 1991 Fig. 14 3 2-56 1991 Fig. 14 3.2-57 1991 Fig. 14 3.2-58 1991 Fig. 14 3.2-59 1991 14.3 3-1 1987 14.3 '-2 1987 14.3.3-3 1987 14.3.3-4 1987 14.3.3-5 1987 14.3.3-6 1987 14 3.4-1 1992 14.3 1993 5-1'4.3.5-2 1993 14.3.5-3 1993 14.3.5-4 1993 Table 14 3 '-1 1993 Table 14.3.5-2 1993 Table 14.3.5-3 1993 Table 14.3.5-4 1993 Table 14.3.5-5 (3 pp) 1993 Table 14.3.5-6 1993 14.3.6-1 1990 14.3 7-1 1993 14.3.7-2 1993

VOLUME ZI Chapter 14 Safet Anal ei.s Unit 2 Pacae Date 14.3 '-3 1993 14.3.7-4 1993 14.3 '-5 1993 14.3.7-6 1993 14 3 7-7 1993 14.3.7-8 1993 14 3 7-9 1993 14.3 7-10 1993 14 3 7-11 1993 Fig. 14.3.7-1 1982 Fig. 14.3.7-2 1982 Fig. 14.3.7-3 1982 Fig. 14.3 '-4 1982 Fig. 14.3.7-5 1982 Fig. 14.3.7-6 1982 Fig. 14 3 '-7 1982 Fig. 14 31,7-8 1982 Fig. 14.3.7-9 1982 Fig. 14.3 7-10 1982 Fig. 14.3 7-11 1982 Fig. 14.3.7-12 1982 Fig. 14.3.7-13 1982 Fig. 14 3 7-14 1982 Fig. 14.3.7-15 1985 Fig. 14.3.7-16 1982 Fig. 14.3.7-17 1982 Fig. 14.3.7-18 1985 Fig. 14.3.7-19 1982 Fig. 14.3.7-20 ,

1982 14.3.8-1 1993 14.3.8-2 1993 14.3.8-3 1993 14.3.8>>4 1993 Fig. 14.3.8-1 1982 Fig. 14.3 '-2 1982 Fig. 14.3.8-3 1982 14.4-1 1991 14.A-1 1989

VOLUME XII chapter 14 Safet nal sis Pacae Date Appendix J ORIG

VOLUME XIII Chapter 14 Safet Anal sis Pacae Date Appendix M ORIG

%%PTER 1 TAELE OP CONTENTS (Cont'd) 1 7 1.7 1 170 Statement of Policy for the Donald C. 1.7 1 Cook Plant Quality Assurance Program 1 7-1 Organisation 1 7-.4

1. 7-2 Quality Assurance Program 1.7-30 1.7 3 Design Control 1. 7-37 1 7.4 Procurement Document Control 1.7-43 175 Instructions, Procedures, and Drawings 1.7-46 1.7.6 Document Control 1.7-49 177 Control of Purchased Items and Services l. 7-51 178 Identification and Control of Ztems 1.7-56 179 Control of SpeciaL Processes 1.7-57 1 7-10 Inspection 1.7-60 1 7'l. Test: Control 1764 17 12. Control. of Measuring and Test Ecgxipment 1 7-66 1 7'13 Handling, Storaget clnd Shipping 1.7-68 1714 Inspection, Test, and Operating Status 1.7-70 1.? 15 Nonconf coming Items 1772 1 7.16 Corrective Action 1.7-73 1 7.17 Quality Assurance Records 1775 1 7 18 Audits 1 7 77 1 7~19 Pire Protection QA Program 1 7 81 AMSX Standard and Regulatory Cuide . 1 7-93 AEPSC/ZQM Exceptions to Operating 1 7 98 Phase - Standards and Regulatory Guides 1 8 ZDENTZPZCATZON OP CONTRACTORS 1.8-1 1 9 PACZLZTY SAPOR CONCLUSIONS 1 9 1-iii July, 1993

CHAP1ZR 1 LIST OF TABLES Comparison of Design Paraaeters Zndec of AEC Caner', Design Criteria References 1 iv July 1991

CHAPTER 2 TABLE OF CONTENTS (Cont'd)

Saetion Title ~Pa e 2.7 ENVIRONMENTAL RADIATION MONITORING 2.7-1 2.7.1 Determination of Maximum Allowable Release 2.7-1 Rates to Air and Water 2.7.2 Sampling Stations 2.7-3 Sampling Lake Water 2.7-3 Sampling of Well Water 2.7-4 Sampling of Milk 2.7-4 Sampling of Food 2.7-5 2 '.3 Stable Element Studies 2.7-5 2.7.4 Measurement of Radioactivity 2.7-6 2.7.5 Operation of the Program 2.7-7 2.7.6 Summary of Preoperational Environmental 2.7-8 Monitoring Program 2.8 PLANT DESIGN BASES DEPENDENT UPON SITE AND 2.8-1 ENVIRONS CHARACTERISTICS 2.8.1 Unit Vent Gas Effluent 2.8-1 2.8.2 Liquid Waste Effluent 2.8-1 2.8.3 Wind Loading Design 2.8-1 2.8.4 Geology 2.8-2 2.8.5 Hydrology 2.8-2 2.8.6 Seismology 2.8-2 2.8.7 Limnology 2.8-2 2.9 PLANT DESIGN CRITERIA FOR STRUCTURES AND 2.9-1 EQUIPMENT 2.9.1 Definition of Seismic Design Classification 2.9-1 Class I 2.9-1 Class II 2.9-1 Class III 2.9-1 2-iii July, 1992

CHAPTER 2 TABLE OF CONTENTS (Cont'd)

~otto Q t~

2.9.2 Classification of Structures and Equipment 2.9-2 2.9.3 Seismic Design Criteria for Seismic 2 '-5 Class I and II Piping 2.9 ' Seismic Design Criteria for Class I, 2.9-8 Class II and Class III Structures Class I 2.9-8 Class II 2.9-8 Class III 2.9-8 For All Structure Seismic Classifications 2.9-9 2.9.5 General Design Considerations for 2.9-10 Building Structures Auxiliary Building 2.9-12 Turbine Building 2.9-15 2.9.6 Seismic Design Criteria for Equipment 2.9-16 2.10 CONCLUSIONS F 10-1 2-iv July 1990

CHAPTER 2 LIST OF TABLES (Cont'd)

Table Title 2.6-3 Summary of Plume Areas, Widths and Volumes 2.6-4 Common and Scientific Names of Fish Species Collected from Cook Plan Study Areas, Southeastern Lake Michigan, 1973-1982 2.6-5 Common Names and Total Estimated Number of Each Species Impinged During 1975-1982 at the Cook Nuclear Plant 2.6-6 Estimates of Annual Entrainment Losses of Fish Larvae and Fish Eggs at the Cook Nuclear Plant 1975-1982 2.7-1 Locations of the Waterborne Surface Sampling Stations 2~7 2 Wells Available from Monitoring Program 2~7 3 Non-Technical Specification Groundwater Wells Steam Generator Storage Facility 2.7-4 Locations of the Milk Sampling Stations 2 7-5 Radiological Environmental Monitoring Program 2.9-1 Loading Condltionss Definitions 2.9-2 Loading Conditions and Stress Limitss Pressure Vessels (Part A) 2.9-2 Loading Conditions and Stress Limits: Pressure Piping (Part B) 2 '-2 Loading Conditions and Stress Limits: Equipment Supports (Part C) 2-vii July, 1993

CHAPTER 2 LIST OF FIGURES

~Fi ure Title 2.1-1 Regional Features 2.1-2 Local Features 2 1-3 Topographic Map of Site 2.1-4 Topographic View of Plant Site 2.1-4a Donald C. Cook Nuclear Plant Sections West and North 2 ~ 1-4b Donald C. Cook Nuclear Plant Topographic Map 2.1<<5 1990 Population Distribution, 0-10 Miles 2 1-6 1975 Population Distribution, 5-60 Miles 2~1 7 2000 Population Distribution, 0-5 Miles 2 1-8 2000 Population Distribution, 5-60 Miles 2.1-9 1972 Dairy Cattle Distribution 0-10 Miles 2.1-10 1971 Transient Population Distribution, 0-8 Miles

2. 2-1 Meteorological Tower 2%2 2 Tornados in the State of Michigan, 1950-1989 2~2 3 Main Tower Wind Rose, January << December 1992 2.2-4 Main Tower Wind Rose, January - Larch 1992 2 2-5 Main Tower Wind Rose, April - June 1992 2~2-6 Main Tower Wind Rose, July - September 1992 2~2 7 Main Tower Hind Rose, October December 1992 2 2-8 Shoreline Tower Wind Rose, January December 1992 2 2-9 Shoreline Tower Wind Rose, January March 1992 2 2-10 Shoreline Tower Wind Rose, April June 1992 2.2-11 Shoreline Tower Wind Rose, July >> September 1992 2>>viii July, 1993

CHAPTER 2 LIST OF FIGURES (Cont'd.)

~Fi ure Title 2.2-12 Shoreline Tower Wind Rose, October December 1992

2. 2-13 Wind Direction Distributions, Turbulence Class IV, 200 Ft. Level 2 2-14 Wind Direction Distributions, Turbulence Class IV, 50 Ft. Level 2 2-15 Wind Direction Distributions, Turbulence Class IV, Satellite 2.2-16 Wind Direction Distributions, All Hours, 200 Ft. Level 2 2-17 Wind Direction Distributions, All Hours 50 Ft. Level 2 2-18 Wind Direction Distributions, All Hours, Satellite 2.2-19 Wind Direction Distributions, Winter 2.2-20 Wind Direction Distributions, Spring 2.2-21 Wind Direction Distributions, Summer 2.2-22 Wind Direction Distributions, Fall 2 ~ 2~23 Monitoring Site Locations 2.3-1 Regional Tectonic Map 2~3 2 Geologic Cross-Section 2 5-1 Epicentral Location Map 2.5-1a Map of Plant Site 2 5-2 Recommended Response Spectra Operating Basis Earthquake 2.5-3 Recommended Response Spectra - Design Basis Earthquake 2.5-3a Site Spectra vs. Modified El Centro '34 Operating Basis Earthquake 2.5-3b Response Spectra Cook Auxiliary Building Floor El.

650'-0" OBE 2.5-3c Response Spectra Cook Auxiliary Building Floor El.

633'-0" OBE 2 5-3d Response Spectra Cook Diesel Generator Building Floor El 609'-0" OBE 2.5-3e Response Spectra Cook Auxiliary Building Floor El.

587'-0" OBE 2 5-3f Site Spectra vs. Modified El Centro '34 DBE 2-ix July, 1993

CHAPTER 2 LIST OF FIGURES (Cont'd.)

~Fi >re Title 2.5-3g Response Spectra Auxiliary Building Floor El.

587'-0" DBE 2.5-3h Response Spectra Diesel Generator Building Floor El.

609'" DBE 2 '-3i Response Spectra Auxiliary Building Floor El.

633'-0" DBE 2.5-35 Response Spectra Auxiliary Building El. 650'-0" 2.6-1 Bathymetric Chart of Lake Michigan 2 6-2 Three Concepts of the Surface Currents of Lake Michigan 2 '-3 Surface Water Temperature 2 '-4 Locations of Current Meters and Temperature Recorders 2.6-5 Schematic of Towed Array 2 ' 6 Region of Lake Influence by Cook Nuclear Plant Discharge 2 6-7 Cook and Palisades Nuclear Plants Meteorological Networks 2.6-8 Present 36-station Cook Nuclear Plant Sampling Grid 2.6-9 Station Locations for the Ma)or Surveys and Short Surveys 2 6-10 Grid of Stations Used in Benthic Sampling Near Cook Nuclear Plant 2.6-11 Map of Souteastern Lake Michigan Showing Plant and Field Fish Larvae Sampling Stations 2 6-12 Species Composition of the Total Number of Fish Impinged Each Year 1975-1982 2 7-1 Air Precipitation, TLD, Well Water Sample, Lake Water Sample Stations 2~7 2 Steam Generator Storage Facility, Non-Technical Specification Groundwater Monitoring Wells 2 ~ 7~3 Air, Precipitation, Lake Water Sample, Milk Sample Stations 2.7-4 TLD Stations Within 2>>5 Mile Plant Radius 2-x July, 1993

CHAPTER 5 LIST 'OF TABLES

~tie Potential Missiles Considered in Class I (Seismic)

Structure Design Wind Velocities and Velocity Pressures Indiana and Michigan Electric Company C. Cook Nuclear Plant Site Soil 'onald Resistivity Measurements Data Taken April, 10 & 11, 1969 Electrical Penetration - Prototype Tests Table of Damping Values Summary of Analyses - Jet Forces Impacting on Internal Structures Summary of Dynamic Motions Dynamic Rotations Ice Condenser Design Parameters Piping Penetrations 5-iii July 1990

CHAPTER 5 LIST OF FIGURES

~fuze ~Ttle

5. 2-1 Location of Resistivity Test 5.2-2 'A Typical Electrical Penetration 5.2-3 Typical Piping Penetrations 5.2-4 Typical Fuel Transfer Tube 5.2-5 Personnel Locks Typical Arrangement & Details 5.2.2-1 Plant Arrangement Sections "G-G", "H-H", "J-J" & "K-K" 5.2.'2-1A Plant Arrangement Sections "L-L & "M-M" 5.2.2-2 Plant Arrangement Mezzanine Floor El. 609'-0" 5.2.2-2A Plant Arrangement Reactor Building Main Floor Elev. 650'-0" 5.2.2-3 Sectional Elevation 5.2.2-4 Containment Building Dome and Wall Reinforcing 5.2.2-4A Containment Building Typical Wall Section 5.2.2-4B Containment Building Re-Bar Anchor Details 5.2.2-5 Typical Expansion Joint Detail 5.2.2-6 Ortographic View of Plant Dynamic Movements 5 '.2-6A Auxiliary and Switchgear Buildings Dynamic Movements 5.2.2-6B Containment and Auxiliary Buildings Dynamic Movements 5.2.2-6C Turbine and Switchgear Buildings Dynamic Movements 5.2.2-6D Turbine, Auxiliary and Containment Buildings Dynamic Movements 5.2.2-7 W Deflection (Inches) 5.2.2-8 W Deflection (Inches)

'5.2.2-9 Pipe Restraint Steam Pipe 5.2.2-10 Wind Funneling Effect 5 '.2-10A Wind Funneling Effect 5.2.2-11 Containment Design Pressures and Temperatures 5.2.2-11A Containment Design Pressures and Temperatures 5.2.2-12 Sect. Elevation Unit No. 1&2 Showing Reactor Containment Thermal Gradients Used For the Design in Summer Operation 5-iv July, 1988

CHAPTER 7 LIST OF TABLES Table Title 7.2-1 List of Reactor Trips and Actuation Means of: ~

Engineered Safety Features, Containment and Steam Line Isolation and Auxiliary Feedwater 7.2-2 Interlock Circuits 7 '-3 Rod Stops 7.2-4 Symbols and Abbreviations 7.2-5 Process Control Block Diagram Drawing 108D087 Index 7.5-1 Process Instrumentation For RPS and ESF Actuation 7.5-2 Engineered Safety Features Equipment Exposed to Harsh Environment 7.8-1 Type "A" Variables Provided the Operator for Manual Functions During and Following an Accident 7.8-2 Type "B" Variables Provided the Operator for Manual Functions During'nd Following an Accident 7 '-3 Type "C" Variables Provided the Operator for Manual Functions During and Following an Accident 7.8-4 Type "D" Variables Provided the Operator for Manual Functions During and Following an Accident 7.8-5 Type "E" Variables Provided the Operator for Manual Functions During and Following an Accident 7-iii July, 1992

CHAPTER 7 LIST OF FIGURES 0

gQ~ure Q tie 7.2-1a Illustration of Overtemperature and Overpower DT Protection Nominal Tavg - 578.7 F, Nominal Pressure 2100 psia 7.2-1b Illustration of Overtemperature and Overpower DT Protection Nominal Tavg - 578.7 F, Nominal Pressure - 2250 psia 7.2-1c Illustration of Overtemperature and Overpower DT Protection 0 2250 psia Nominal Tavg 547 F, Nominal Pressure 7.2-1d Illustration of Overtemperature and Overpower DT Protection Nominal Tavg 547 F, Nominal Pressure - 2100 psia 7.2-2 Reactor Protection Systems 7.2-3 Control Rod Bank Insertion Monitor 7,2-4 'Rod Deviation Comparator 7.2-5 Pressurizer Pressure Protection System

~ 7.2-6 Pressurizer Level Protection 7 '-7 Pressurizer, Sealed Reference Leg Level System 7.2-8 Steam Generator Level Protection 7.2-9 Setpoint Reduction Function For Overpower and Overtemperature DT Trips 7 '-1 Simplified Block Diagram of Reactor Control System 7.5-1 Containment Pressure Protection 7.5-2 Environmental Conditions Inside Containment Loss-of-Coolant Accident 7I5 3 Environmental Conditions Inside Containment Main Steam Line Break 7.6-1 InCore Instrumentation Details 7.6-2 Typical Arrangement of Movable Miniature Neutron Flux Detector System (Elevation View) 7.6-3 Schematic Arrangement of InCore Flux Detectors (Plan View) 7-iv July 1990

CHAPTER 8 TABLE OF CONTENTS Sect o ~Pa e ELECTRICAL SYSTEMS 8.1-1 8.1 DESIGN BASES 8.1-2 General Design Criteria 8.1-2 8.1.2 Functional Criteria 8.1e4 8.2 NETWORK INTERCONNECTIONS 8.2-1 8.3 STATION SERVICE SYSTEMS 8.3-1 8.3.1 4160 Volt System 8.3-1 8.3.2 Low Voltage Power Systems 8.3-2 8.3.3 120 Volt AC Vital Instrument Bus System 8.3-4 8.3.4 250 Volt DC System 8.3-5 8.3.5 250 Volt DC Battery N System 8.3-7 8.3.6 Lighting System 8.3-11 EMERGENCY POWER SYSTEM 8.'4 -1 8.5 DESIGN EVALUATION 8.5-1 8.6 TESTS AND INSPECTION 8.6-1 8-i July, 1988

CHAPTER 8 LIST OF FIGURES Fi~~e Q tie

8. 1-1 Aux. One Line Diagram 8.1-1a Main Auxiliary One-Line Diagram Bus 'A' 'B'ngineered Safety System S.l-lb Main Auxiliary One-Line Diagram Bus 'C' 'D'ngineered Safety System F 1-2 Cook Nuclear Plant Simplified Offsite Power Sources One-Line 8.2-1 Switching Arrangements A Donald C. Cook Nuclear Plant and Neighboring Stations 8.3>>1 Vital Instrument Bus Distribution System 8.3-2 250 V DC Distribution 8.3-3 D. C. Cook Nuclear Plant Turbine Driven Auxiliary Feedwater System One-Line 8.4-1 Emergency Diesel Generator Fuel Oil Supply 8-ii July 199V

CHAPTER 9 TABLE OF CONTENTS

~ect~

9.0 AUXILIARYAND EMERGENCY SYSTEMS 9.1-1 9.1 GENEKLL DESIGN CRITERIA 9.1-2 911 Auxiliary and Emergency Systems Criteria 9.1-2 9 1.2 Related Criteria 9. 1-3 9.2 CHEMICAL AND VOLUME CONTROL SYSTEM 9.2-1 9.2.1 Design Bases 9 2-1 9 2 2 System Design and Operation 9-2-4 9 2 3 System Design Evaluation 9. 2-31 9.2 References 9.2.-38 9.3 RESIDUAL HEAT REMOVAL SYSTEM 9.3-1 9~3.1 Design Bases 9.3-1 9.3 2 System Design and Operation 9.3-2 9.3 3 System Design Evaluation 9.3-6 9.3 4 Malfunction Analysis 9.3-11 9 3 5 Tests: Inspections 9.3-11 9 3.6 Safety Limits and Conditions 9.3-12 9.4 SPENT FUEL POOL COOLING SYSTEM 9.4-1 9.4.1 Design Bases 9. 4-1 9 4 2 System Design and Operation 9.4-2 9 4 3 Design Evaluation 9.4-6 9.4.4 Tests and Inspections 9.4-7 9.5 COMPONENT COOLING SYSTEM 9 5-1 9.5 1 Design Bases 9.5-1 9 5i2 System Design and Operation 9.5-1 9 5.3 Components 9.5 4 9.5.4 System Evaluation 9 5 7 9 ' ' Minimum Operating Conditions 9.5-9 9-i July, 1993

(RAPTER 9 TABLE OF CONTENTS (Cont'd)

Sect on T tie Pacae 9 ' 6 Tests and Inspections 9.5-9 9.6 SAMPLING SYSTEMS 9 6 1 9.6 1 Design Basis 9 6-1 9~6 2 System Design 9.6-2 9 6.3 System Evaluation 9.6-7 9 7 REACTOR COMPONENTS AND FUEL HANDLING SYSTEM 9.7-1 9 7.1 Design Bases 9.7-2 9~7 2 System Design and Operation 9 7-4 973 Design Evaluation 9 7-24 974 Tests and Znspections 9 7-25 9 8 FACZLZTY SERVICE SYSTEMS 9.8-1 981 Fire Protection System 9.8-1 982 Compressed Air System 9.8-21 983 Service Hater Systems 9.8-24 9 9 AUXILIARY BUILDZNG VENTILATION SYSTEM 9 9-1 9.9 1 General Description 9 9-1 9 9 2 Design Basis 9.9 1 9 9 3 System Descriptions ~ 9 9-2 9 9 4 Design Evaluation 9.9-8 9 10 CONTROL ROOM VENTILATION SYSTEM 9. 10-1 9.10 1 General Description 9. 10-1 9 ~ 10 2 Design Basis 9 10 1 9 10~3 System Operation 9 10-2 9~10~4 Design Evaluation 9 10-4

9. 10 ~ 5 Zncident Control 9 10-4, 9o10 ~ 6 Tests and Inspections 9 10 5 9-ii July, 199

CHAPTER 9 LIST OF TABLES gab~e 9 2-1 Chemical and Volume Control System Code Requirements 9 2 2 Chemical and Volume Control System Design Parameters 9.2 3 Principal Component Data Summary 924 Failure Analysis of the Chemical and Volume Control System 9 '-1 Residual Heat Removal System Code Requirements 9 3-2 Residual Heat Removal System Design Parameters 9.3-3 Residual Heat Removal Malfunction Analysis 9 i4-1 Spent Fuel Pool Cooling System Code Requirements 9 4-2 Spent Fuel Pool Cooling System Component Design Data 9 4 3 Spent Fuel Pool Cooling System Malfunction Analysis 951 Component Cooling System Code Requirements 9 5-2 Component Cooling System Minimum Flow Requirements Per Train 9 5-3 Component Cooling System Component Design Data 9 5-4 Component Cooling System Malfunction Analysis J

9.8-1 Fire Pump Starting Sequences 9 8 2 Compressed Air System Descriptive Information 9 8-3 Service Water Systems Components Design Data 9.8-4 Non-Essential Service Water Requirements 9.8-5 Essential Service Water System Minimum Flow Requirements Per Train .

9 8-6 Essential Service Water System Malfunction Analysis 9-iii July, 1993

CHAPTER 9 LIST OF FIGURES

~Fi re ~Ttle 9 2-1 CVCS - Reactor Letdown and Charging 9 2 2 CVCS - Reactor Coolant Demineralisation 9.2-3 CVCS - Boron Make-up 9 2-4 CVCS - Boron Hold-up 9 2 5 Flow Diagram CVCS - Boron Recovery 9.2-6 CVCS - Monitor Tanks 9.3-1 Emergency Core Cooling (RHR)

9. 4-1 Spent Fuel Pit Cooling and Clean-up 9 5-1 Component Cooling 961 Sampling 9 '-2 Posh-Accident Sampling 9 7-1 Typical Fuel Transfer System 9 7-2 Spent Fuel Pool Plan 973 Schematic for SFP Interface Boundary Between Region 1 With Three~t~f-Four Storage Configuration and Region 2 981 Fire Protection Water 9 8-2 Fire Protection CO2 9.8 3 Compressed Air System 9 8 4 Non-Essential Service Water Unit 1 9 8 5 Non-Essential Service Water Unit 2 9.8 6 Non-Essential Service Water Units 1 or 2 9 8 7 Essential Service Water 9o9 1 Auxiliary Building Ventilation Sheet 1 909-2 Auxiliary Building Ventilation Sheet 2

9. 10-,1 Control Room Ventilation 9-iv July 199~

AGQ'TER 11 LIST OF TABLES Tab e ll 1-1 Waste Disposal System Performance Data 11 1-2 Waste Disposal Components Code Requirements ll 1-3 Coaponent Suneag Data ll.1-4 Estimated Liquid Discharge to Waste Disposal Syst: em 11 1-5 Estimated Lipoid Release by Isotope - Two Units

11. 1-6 Estimated Annual Gaseous Release by Isotope

11. 2-1 Plant Zones Classifications 11 2-2 Primary Shielding Design Parameters, Neutron and Gamma Fluxes 11.2-3 Secondary Shield Design Parameters ll 2-4 Accident Shield Design Parameters 11 2-5 Refueling Shield Design Parameters 11 2-6 Principal Auxiliary Shielding 1127 Instantaneous Radi,ation Sources Released To the Containment Following TID-14844 Accident Release-Mev/Sec 11 2-8 Gap Activity Circulating in Residual. Heat Removal Loop, Mev/cc-Sec 11 3-1 Radiation Monitoring System Channel Sensitivities a'nd Detecting Medium 11.3-2 Reactor Coolant Fission and Corrosion Product Activities During Steady State Operation and Plant Shutdown Operation 11.5 1 Design and Measured Equ{,librt.um Reactor Coolant Fission Product Activities for Operating PWR's and Calculated Values for the D.C. Cook Stations ll 5 2 Blowdown Treatment System Components 11-iii July, 1988

CHAPTER 11 LZST OF PZGURES Xi~e 11 1 1 Vents and Drains 1112 Waste Disposal System - Liquids and Solids Sheet l of 3 Units 1 and 2 11.1>>2a Waste Disposal System,- Licpxids and Solids Sheet 2 of 3 11.1-2b Waste Disposal System - Liquids and Solids Sheet 3 of 3 ll 1-3 Waste Disposal System - Gaseous Plow Diagram 11 1-4 Waste Di,sposal System - Gas Supply and Analysis 11.2.-1 Zntegrated Exposure as a Function of Distance from Containment Building 11.4-1 Access Control Area 11.4-2 Containment Access BuiLding 11.4>>3 Hot Laboratory 11.4-4 Chemistry Counting Room 11.5-1 Steam Generator Blowdown System Unit 1 or 2 11.6-1 Organisation and Functional Structure 11.6-2a Fuel Assembly Plow Chart 11.6-2b Movable Miniature Neutron Flux Detector Plow Chart 11.6-2c Fission Chamber Detector Plow Chart

hh CHAPTER 12 TABLE OF CONTENTS

~Sect c ~Tt e ~ae 12 CONDUCT OF OPERATIONS 12.1-1 12.1 ORGANIZATION AND RESPONSIBILITY 12. 1-1 12.2 LICENSED OPERATOR REQUALIFICATION PROGRAM 12.2-1 12.3 EMERGENCY PLAN 12.3-1 12.4 RECORDS 12.4>>1 12 ' REVIEW AND AUDIT OF OPERATIONS 12. 5-3 h

12.6 NUCLEAR DESIGN AND SUPPORT CAPABILITY 12.6-1 12.7 WRITTEN PROCEDURES 12.7-1 12-1 July 1991

CHAPTER 12 LIST OF FIGURES F ure 12.1-1 Donald C. Cook Nuclear Plant Plant Organization 12-ii July, 1988

CHAPTER 14 LIST OF TABLES Table Title

14. 1-1 Unit 1 Design Power Capability Parameters Used in Non-LOCA Safety Analyses 14.1-2 Reactor Trip Points and Time Delays to Trip Assumed in Safety Analyses 14.1-3 Summary of Initial Conditions and Computer Codes Used 14.1-4 Instrumentation Drift and Calorimetric Errors Power Range Neutron Flux 14 1.10-1 Time Sequence of Events (Manual Rod Control) 14 1.10-2 Time Sequence of Events (Automatic Rod Control) 14.1.10-3 Time Sequence of Events (Manual Rod Control) 14.1. 10-4 Time Sequence of Events (Automatic Rod Control) 14.1. 13-1 Potential Turbine - Generator Missiles 14.2 l-l Fuel Handling Accident Auxiliary Building Inventories and Constants of Significant Fission Product Radionuclides 14.2. 1-2 Data and Assumptions for the Evaluation of the Fuel Handl'ing Accident In The Auxiliary Building 14 2 1-3 Nuclear Characteristics of Highest Rated Discharged Assembly Fuel Handling Accident In Containment 14.2 1-4 Activities Zn Hi.ghest Rated Discharged Assembly (Curies At Time Of Reactor Shutdown) Fuel Handling Accident in Containment 14 2 '-5 Parameters For Fuel Handling Accident Zn Containment Dose Calculation 14 2 3-1 Volume Control Tank and Letdown Activiti.es 14.2 3-2 Gas Decay Tank Equilibrium Activity 14 2 5-1 Limiting Steamline Break Statepoint Double Ended Rupture Inside Containment With Offsite Power Available 14.2.6-1 Parameters Used in Analysis of the Rod Cluster Control Assembly Egection Accident 14 2.7-1 Loss of A.C. Power to the Plant Auxiliaries Steam Release 14 2.7-2 Steam Line Break - Steam Release a

14.2.7-3 Steam Generator Tube Rupture Steam Release 14.2.8-1 Time Sequence of Events Unit 1 14-v July, 1993

CHAPTER 14 LZST OF TABLES Table ~ice 14.3 1-1 Large Break LOCA - Results 14e3 1-2 Large Break LOCA - Cases Analysed 14.3 1 3 Large Break Containment Data (Zce Condenser Containment) 14.3. 1-4 Mass and Energy Release Rates Maximum SZ 14.3 1-5 Mass and Energy Release. Rates Minimum SZ 14 3 1-6 Nitrogen Mass and Energy Release Rates 14 3.2-1 Safety Zngection Flow Rate 14 3 2-2 Plant Xnput Parameters Used in Small Break LOCA Analysis 14 3.2-3 Small-Break Loss of Coolant Accident Calculation 14.3 2-4 Time Sequence of Events for Condition ZXX Events 14.3.2 5 Small-Break Loss of Coolant Accident Calculation 14 3.2-6 Time Sequence of Events for Condition ZZZ Events 14.3 2-7 Small-Break Loss of Coolant Accident Calculation Results HHSZ Cross-Tie Valve Closed 14 3 4-1 Cook Nuclear Plant Passive Heat Sinks 14 3.4-2 TMD Flow Path Znput Data 14 3.4-3 1973 Waits Mill Preliminary Test Conditions 14 3e4-4 TMD Volume Znput for AEP 14 3 4-5 (No Heading) 14.3 4-6 Calculated Maximum Peak Pressures in Lower Compartment Elements Assuming Unaugmented Flow 14 3 '-7 Calculated Maximum Peak Pressures Zn the Zce Condenser Compartment Assuming Unaugmented Flow 14 3 4-8 Calculated Maximum Differential Pressures Across the Operating'eck or Lower Crane Wall Assuming Unaugmented Flow 14 3 '-9 Calculated Maximum Differential Pressures Across the Upper Crane Wall Assuming Unaugmented Flow 14.3. 4-10 Sensitivity Studies for Cook Nuclear Plant 14.3 '-11 Cook Nuclear Plant Zce Condenser Design Parameters 14 3 4-12 (No Heading)

Unit 1 14 vi July, 1993

KGLPTER 14 LIST OF TABLES

~ab~e 14 4.2 1 Equipment Required to Shutdown Reactor (For High Energy Pipe Ruptures Outside Containment) 14.4.2-2 High Energy Lines That Were Walked 14.4 2-3 Ultimate Shear Stresses at Distance d From the Supports for e

Two-Way Elements 14 4 2-4 Two-Way Elements 14 4 2-5 Major Postulated High Energy Pipe Breaks 14 4.4-1 Stress Values for Main Steam Leads 1 and 4 Allowable Stress Values: Operational Plus Seismic Stresses < 30,000 PSZ (*+) Thermal Stresses < 18,000 PSZ 14.4.4 2 Stress Values for Main Steam Leads 2 and 3 Allowable Stress Values: Operational Plus Seismic Stresses

< 30,000 PSZ () Thermal Stresses < 18,000 PSZ 14.4 4-3 Supply and Signal Lines to be Protected 14 4.4-4 Stress Values at Postulated Break Locations Allowable Stress ~ 30,000 PSZ 14 4 4-5 Stress Values for Feedwater Lines Allowable Stress Values: Operational Plus Seismic Stresses < 30@000 PSZ Thermal Stresses < 18,000 PSZ 14 4 4-6 Stress Levels Main Steam to Auxiliary Feed Pump Turbine Line Allowable Stress Values:

Operational Plus Seismic Stresses < 30,000 PSI Thermal Stresses < 18,000 PSZ 14.4 6 1 West Steam Enclosure/Main Steam Accessway Vent Area and Volume Inputs to TMD 14.4.6-2 East Steam Enclosure Vent Area and Volume Inputs to TMD Unit 1 14-ix July 1990

CHAPTER 14 LIST OF TABLES

~aha 14i4~6 3 Model. Parameters (West Main Steam Enclosure and Main Steam Accessway) Large Break 14 4 6-3a Model Parameters (West Main Steam Enclosure and Main Steam Accessway) Small Break 14.4.6-4 Model Parameters (East Main Steam Enclosure) Large Break 14 4.6-4a Model Parameters (East Main Steam Enclosure) Small Break 14 4 6-5 Mass and Energy Release for Steam Line Break in Main Steam Enclosure, Large Break 4 14.4.6-5a Mass and Energy Release for Steam Line break in Main steam Enclosure, Small Break 14 4 6-5b Deleted 14.4.6 5c Deleted 14.4.6-6 Feedwater Line Break at, the Containment Penetration (Applicable to East or West. Steam Enclosure) 14467 Feedwater Line Break at the Tee Between the 20" and 30" Lines, 20" Line Running to Steam Generators 2 and 3 (Applicable to Main Steam Accessway) 14o4~6 8 Relation of Node Calculated Pressure to Pressure Capability of Slabs 14 4.6-9 Peak Pressure Differential Main Steam Line Break West Steam Enclosure 14 4 6-10 Peak Differential Pressure Feedwater Zine Break in West Steam Enclosure

$ 4.4. 6-11 Peak Differential Pressure Feedwater Line Break in Main Steam Accessway 14 4.6-12 Peak Differential Pressure Main Steam Line Break in East Steam Enclosure 14 ~ 4 6-13 Peak Differential Pressure Feedwater Line Break in East Steam Enclosure Unit 1 14-x July, 19

TABLE OF CONTENTS (Continued)

~sectic Pacae 14 2 STANDBY SAFEGUARDS ANALYSIS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 14 2-1 1421 RADIOLOGICAL CONSEQUENCES OF FUEL HAILING ACCIDENT ~ ~ 14 2.1-1 14 2.2 POSTULATED RADZOACTZVE RELEASES DUE TO LZQUZD-CONTAINING TANK FAILURES ~ o ~ oo ~ oe ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ oo ~ o ~ ~ ~ .e 14.2.2-1 14o2 REFERENCES o ~ ~ ~ eo ~ ~ ~ ~ ~ ~ e ~ ~ ~ o ~ ~ ~ oooo ~ oo ~ ~ eoo ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ 'l4 2.2-6 14.2.3 ACCXDENTAL WASTE GAS RELEASE ~ ~ o ~ ~ ~ ~ . ~ ~ 14.2.3-1 14 2.4 STEAM GENERATOR TUBE RUPTURE ~ o ~ ~ ~ ~ o ~ ~ ~ ~ 14.2 4-1 14 2 4 1 ZDENTIFZCATION OF CAUSES AND ACCIDENT DESCRIPTION 14.2 4-1 14 2 4 ~ 2 ANALYSIS OF EFFECTS AND CONSEQUENCES ~ ~ - ~ ~ ~ 14.2 4-3 14 2.4.3 CONCLUSZONS ~ ~ ~ ~ o o ~ o o o ~ o ~

~ o ~ o o e o o o ~ a o ~ ~ ~ ~ ~ ~ ~ ~ o ~ e e ~ o o e o 14.2.4-9 1425 RUPTURE OF A STEAMLZNE (STEAMLZNE BREAK) ~ ~ 14 2 5-1 14 2 5 1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION 14.2 5-1 14 2 5 2 ANALYSXS OF EFFECTS AND CONSEQUENCES 14.2 .5-3 14 2.5.3 CONCLUSZONS ~ ~ ~ o ~ oo ~ ~ ~ .oo ~ ~ ~ ~ ooo ~ ~ o ~ o ~ ooo ~ oeoo ~ ~ oooooo ~ o 14.2.5-10 14 2 5 4 REFERENCES ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ o o ~ ~ ~ ~ ~ o e o o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 14 2 5-11 14.2 6 RUPTURE OF CONTROL ROD DRIVE MECHANISM (CRDM)

HOUSING,(RCCA EJECTION) ~ o ~ ~ ~ ~ ee ~ o ~ ~ oe ~ ~ eo ~ o ~ ~ ~ ~ o ~ oooo'4 2.6-1 14.2 ' 1 ZDENTZFXCATZON OF CAUSES AND ACCXDENT DESCRIPTION 14.2 6-1 14 2 6 ~ 2 ANALYSXS OF EFFECTS AND CONSEQUENCES ~ ~ ~ ~ ~ 14 2 .6-7 14263 CONCLUSXONS ~ o ~ o ~ ~ ~ o o ~ o ~ o ~ ~ ~ o o ~ ~ o ~ ~ o ~ ~ o ~ e ~ ~ o ~ ~ 142615 14 2.7 SECONDARY SYSTEMS ACCXDENT ENVIRONMENTAL CONSEQUENCES ~ o o o ~ o ~ o ~ o o ~ ~ ~ ~ o ~ ~ o ~ o o ~ ~ o o ~ ~ ~ o eo ~ ~ ~ ~ ~ o ~ ~ o e 14 2 7-1 14.2 8 MAJOR RUPTURE OF MAXN FEEDWAKGt PIPE (FEEDLZNE BREAR) o ~ ~ ~ o ~ o ~ ~ ~ o ~ o ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ e o ~ ~ o ~ ~ o e ~ o oo o e ~ o o o ~ ~ 14.2. 8-1 14 ~ 2 ~ 8 1 IDENTZFZCATZON OF CAUSES AND ACCIDENT DESCRIPTION ~ ~ 14 2 8-1 14 2 ~ 8 2 ANALYSXS OF EFFECTS AND CONSEQUENCES ~ ~ ~ ~ ~ ~ ~ ~ 14.2 8-3 Unit 2 14-v July 1991

AB 0 CO S ( Continued)

Sect o Pacae 14 2.8 3 CONCLUSIONS e ~ e ~ ~ ~ e ~ o ~ e ~ ~ ~ e ~ ooeoee ~ eoe ~ ~ eeooo ~ oeo ~ oeooo 14 2 8-5 14 2 8 4 REPERENCES o ~ eoo ~ o ~ oo o oo~ oooo ~ o o ~ o o o o o o o o o o ~ o ~ ~ e o ~ o o o o o 14 2 8-6 14.3 REACTOR COOLANT SYSTEM PIPE RUPTURE (LOSS OP COOLANT ACCIDENT) o ~ ~ o ~ ~ ~ oo ~ ooo ~ eel o ~ ~ ~ ~ oe ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 14 3 1-1 1431 LARGE, BREAK LOSS~P~LANT ACCIDENT ANALYSIS ~ ~ ~ ~ ~ ~ 14;3.1-2 14 3 1 1 MAJOR LOCA ANALYSES APPLZCABLE TO WESTINGHOUSE FUEL ~ 14.3 1-2 14 3.1 2 MAJOR LOCA ANALYSES APPLXCABLE TO ANF FUEL 14.3.1-24 14.3 2 LOSS'-COOLANT FROM SMALL RUPTURED PXPES OR FROM CRACKS IN LARGE PIPES WHICH ACTUATES THE EMERGENCY CORE COOLING SYSTEM ~ .e ~ ~ ooooo ~ o ~ ~ ~ ~ ~ o ~ ~ oo ~ e ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ 14 3 2-1 14o3 2 1 ANALYSIS OP EFPECTS AND CONSEQUENCES ~ ~ ao ~ ceo e eaeo oooo e 14.3 2-1 14 3 2 2 CONCLUSIONS ~ e o ~ o o o ~ o a ~ o o o ~ ~ o ~ ~ o o ~ ~ ~ ~ o ~ oe ~ o o oo o oo o ~ o o o 14 3.2-4 1432 REFERENCES oo o a e o o o ~ ~ o. ~ o o e.o.e ~ o o o o 14 3 2-8 1433 ASYMMETRIC LOCA LOADS AND MECHANISTIC'RACTURE, EVALUATZONo~ o ~ o ~ o o o ~ ~ o ~ ~ o o ~ ~ e ~ ~ e e ~ o ~ ~ o ~ ~ o ~ ~ e ~ ~ ~ e o e o o o o o 14 3 3-1 14.3 4 CONTAINMENT ZNTEGRXTY EVALUATION ~ ~ . ~ ~ ~ .. ~ ~ ~ ~ ~

~

14 3 4-1 14

3.4 REFERENCES

~ ~ ooe ~ oo ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ oe. ~ o ~ ~ ~ ~ oee ~ oooo ~ 14 3 4-42 14 3e5 RADZOLOG1CAL CONSEQUENCES OF A LOSS OP COOLANT ACCIDENT AND ~KR EVENTS CONSIDERED ZN SAFETY ANALYSISo ~ e ~ ~ o e ~ e ~ o o o ao ~ o e ~ o o o ~ o e o o~ o o o ~ ~ ~ ~ e e ~ ~ o o o o ~ o oo 14351 14 3.6 HYDROGEN XN THE UNIT 2 CONTAINMENT AFTER A LOSS OP COOLANT ACCZDENT eooeo ~ ~ ~ ~ ~ eooooo,ooo ~ eo ~ 'ooo ~ e 14.3.6-1 14 3.7 LONG TERM COOLZNG o ~ ooooooo ~ ~ ~ ~ ~ oo ~ oo ~ ~ ~ ~ ~ o ~ oo ~ ~ ooooooo 14 3.7-1 Z438 N!TROGEN BLANKETING ~ ooooo ~ ~ ~ ooo ~ o ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ooooooo ~ oa 14.3 8-1 14e4 ENVXRONMEK QUALIFICATION ~ a o o' o ~ ~ o o oe e ~ oo ~ ~ e ~ o o o ooo o 14 4 1 I

APPENDXX 14A RADZATXON SOURCES 14A 1 Unit 2 14-vi July, 199.

CHAPTER 14 LXST OF FIGURES (Cont'd)

~Fl u~e Tte 14.3.2-32 Hot Spot Fluid Temperature (4 Inch) High Temperature, High Pressure 14.3.2-33 Total Break Flow (4 Inch) High Temperature, High Pressure 14.3.2-34 Intact Loop Pumped SX Flow (4 Inch) High Temperature, High Pressure 14.3.2-35 RCS Pressure (4 Inch) Reduced Temperature, High Pressure 14.3.2-36 Core Mixture Height (4 Inch) Reduced Temperature, High Pressure 14.3.2-37 Hot Spot Clad Temperature (4 Inch) Reduced Temperature, High Pressure 14.3.2-38 Core Steam Flowrate (4 Inch) Reduced Temperature, High Pressure 14.3.2-39 Hot Spot Heat Transfer Coefficient (4 Inch) Reduced Temperature, High Pressure 14 3.2-40 Hot Spot Fluid Temperature (4 Inch) Reduced Temperature, High Pressure 14-3.2-41 Total Break Flow (4 Inch) Reduced Temperature, High Pressure 14.3.2-42 Intact Loop Pumped SX Flow (4 Inch) Reduced Temperature, High Pressure 14.3.2-43 RCS Pressure (3 Xnch) High Temperature, Reduced Pressure Cross Ties Closed 14.3.2-44 Core Mixture Height (3 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14 3 2-45 Hot Spot Clad Temperature (3 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14 3 2-46 Core Steam Flowrate (3 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14-3.2-47 Hot Spot Heat Transfer Coefficient (3 Inch) High Temperature, Reduced Pressure 14.3.2-48 Hot Spot Fluid Temperature (3 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14 '.2 49 Total Break Flow (3 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14.3.2-50 Intact Loop Pumped SZ Flow (3 Inch) High Temperature, Reduced Pressure Cross Ties Closed Unit 2 14-xxix July, 1993

CHAPTER 14 LIST OF FIGURES (Cont'd) e 14.3.2-51 RCS Pressure (4 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14.3.2>>52 Core Mature Height (4 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14 3 2-53 Hot Spot Clad Temperature (4 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14 3 2-54 Core Steam Flowrate (4 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14.3.2-55 Hot Spot Heat Transfer Coefficient (4 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14.3.2-56 Hot Spot Fluid Temperatuxe (4 Inch) High Temperature, Reduced Pressure Cross Ties Closed 14.3.2-57 Total Break Flow (4 Inch) High Temperatuxe, Reduced Pressure Cross Ties Closed 14.3 2-58 Intact Loop Pumped SZ Flow (4 Inch) High Temperature, Reduced Pressure Cxoss Ties Closed 14.3.2-59 Hot Rod Power Distribution 3413 MWT Cross Ties Closed 14.3 4-1 AEP Containment Integrity 'Analysis, 2.11E+6 Lbs of Zce/3425 Cross-Tie Closed, System Pressure 14.3 4-2 AEP Containment Zntegrity Analysis, 2.11E+6 Lbs of Zce/3425 MWT/RHR Cross-Tie Closed, Upper Compartment Temperature 14.3.4-3 AEP Containment Integrity Analysis, 2.11E+6 Lbs of Ice/3425 MWT/RHR Cross-Tie Closed, Lower Compartment Temperature 14.3.4-4 AEP Containment Integrity Analysis, 2.11E+6 Lbs of Ice/3425 MWT/RHR Cross-Tie Closed, Active and Inactive Sump Temperatures 14 3.4-5 AEP Containment Integrity Analysis, 2.11E+6 Lbs of Zce/3425 MWT/RHR Cross<<Tie Closed, Ice Melt (Lbs) 14 3.4-6 Plan at Equipment Rooms Elevation 14.3.4-7 Containment Section View 14.3.4-8 Plan View at Zce Condenser. Elevation Zce Condensex Compartments 14.3.4-9 Layout of Containment Shell 14.3.4-10 TMD Code Network Unit 2 14-xxx July, 1993

CHAPTER 14 LIST OF FIGURES (Cont'd)

+i~ es Q~te 14 3.4-11 Upper and Lower Compartment Pressure Transient for Worst Case Break Compartment (Element 6) having a DEHL Break 14.3.4-12 Xllustration of Choked Flow Characteristics 14.3.4-13 Steam Concentration in a Vertical Distribution Channel 14.3.4-14 Peak Compression Pressure Versus Compression Ratio 14.3.4-15 Coolant Temperature at Core inlet 14.3.4-16 Core Reflooding Rate - Vin 14 3.4-17 Carryover Fraction - F out 14.3.4-18 Fraction of Flow Through Broken Loop 14.3.4-19 Post-Blowdown Downcomer and Core Water Height 14 3.4-20 Steam Generator Heat Content 14.3 4-21 Cold Leg Double-Ended Guillotine Full Power mh Transient 14 3.4-22 Cold Leg'ouble-Ended Guillotine Full Power mh Transient 14.3.4-23 Cold Leg Double-Ended Guillotine Full Power mh Transient 14.3.4-24 Cold Leg Double-Ended Guillotine Full Power mh Transient 14.3.4-25 Hot Leg Double-Ended Guillotine Full Power mh Transient 14.3.4-26 Hot Leg Double-Ended Guillotine Full Power mh Transient 14 3 4-27 DECLG: Compartment f1 through 14.3.4-71 14 3.4-72 DEHLG: Compartment 81 through 14.3.4-116 14.3 4'>>117 DEHLG: Compartment P2 through 14.3.4-161 Unit 2 14-xxxi July, 1993

CHAPTER 14 LIST OP FIGURES (Cont'd)

~u~lB 8 14.3.4-162 DEHLGs Compartment P3 through 14 3.4-206 14 3 4-207 DEHLG: Compartment P4 through 14 3.4-251 14 3 4-252 DEHLG: Compartment, 85 through 14 3.4-296 14.3.4-297 DEHLG: Compartment 46 through 14.3.4-341 14.3.4-342 DECLG: Compartment g3 through 14 3.4-386 14 3 4-387 DECLG: Compartment g4 through 14 3.4-393 14.3.4-394 Figure was omitted by Westinghouse Electric Corporation from the

, Amendment No. 78 14.3 4-395 DECLG: Compartment f4 through 14.3.4-431 14 3 4-432 DECLG: Compartment P6 through 14.3.4-476 14 3 4-477 Compartment Temperature 14 3 4 478 Compartment Temperature 14.3.4-479 Pressurizer Enclosure Noding 14 3.4-480 Pressurizer Enclosure Noding Diagram and Plow Paths 14 3.4-481 TMD Code Network 14 3.4-482 TMD Compressible Plow for Pressurizer Enclosure through 14.3.4-486 Unit 2 14-xxxii July, 1993 i

CHAPTER 14 LIST OF FIGURES (Cont d)

~Ti tee Title 14 3 4-487 Pressurizer Enclosure Differential Pressure through 14.3.4-493 14.3.4-494 Steam Generator Enclosure Above Elevation 665 Ft.

14 3.4-495 Steam Generator Enclosure Below Elevation 665 Ft.

14.3.4-496 Steam Generator Enclosure Cut~en View of the Steam Generator Enclosure (Sheet 1 of 2) 14.3.4-497 Worst Break Lower Compaztment Temperature Comparison 14.3.4-498 Upper Compartment Temperature (30% Power Level) 14.3.4-499 Lower Compartment Pressure (30% Power Level) 14.3.4-500 Lowez Compartment Temperature (30% Power Level) 14.3.4-501 Worst Break Lower Compartment Temperature Comparison Generic Analysis 14.3. 7-1 Large Steam Break with Reactor Coolant Pumps Running Reactor Coolant System 14.3.7-2 Large Steam Break with Reactor Coolant: Pumps Running Broken Loop Cold Leg Temperatures Versus Time (Seconds) 14-3 7-3 Large Steam Break with Reactor Coolant Pumps Running Intact Loop Cold Leg Temperature Versus Time (Seconds) 14.3.7-4 Large Steam Line Break with Reactor Coolant Pumps Running 14.3.7-5 Large Steam Break with Reactor Coolant Pumps Tripped - Reactor Coolant System Pressure Versus Time (Seconds) 14.3.7-6 Large Steam Break with Reactor Coolant Pumps Tripped Broken Cold Leg Temperature Versus Time (Seconds) 14.3 '-7 Large Steam Break with Reactor Coolant Pumps Tripped Intact Loop Cold Leg Temperature Versus Time (Seconds) 14.3.7-8 Large Steam Line with Reactor Coolant Pumps Tri.pped 14.3.7-9 Typical Small Break Pressure Transient 14.3.7-10 Energy Removal by Break at Equilibrium Unit 2 14-xxxiii July, 1993

(2DQ'TER 14 LIST OF FIGURES (Cont'd)

~Fc es 14.3 7-11 Equilibrium Pressure Between SI Flow and Break Flow for Saturated Iiquid Discharge from the Break 14.3.7-12 2 Inch Cold Leg break 14.3.7-13 1 Inch Break 14.3.7-14 .615 Inch Break 14.3.7-15 Mixture Height: Abave Bottom of Core, Ft 14.3.7>>16 1 Inch Break 14.3.7-17 .615 Inch Break 14.3.7-18 Charging Flow from One Centrifugal Charging Pump 14.3.7-19 Large Steam Line Break with Reactor Coolant Pumps Running 14.3.7-20 Large Steam Line Break with Reactor Coolant Pumps Running 14 3 8-1 Typical Small Break Pressure Transient 14 3 8-2 Energy Removal by Break Pressure Transient 14.3.8-3 Equilibrium Pressure Between SZ Flow and Break Flow for Saturated Liquid Discharge from the Break Unit 2 14-xxxiv July, 199-

CHAPTER 1 INTRODUCTION AND

1.0 INTRODUCTION

This Updated Final Safety Analysis Report is submitted in accordance with the requirements of 10 CFR 50,71 (e). It is based on the original FSAR, including 84 amendments, which was submitted in support of an application by Indiana & Michigan Electric Company (I&M), whose name is now Indiana Michigan Power Company (the acronym I&M is still used however) for licenses to operate two nuclear power units at its Donald C. Cook Nuclear Plant.

This submittal contains update information for the period up to six months prior to the most recent revision of this document. The update information is of a similar level of detail as that presented in the

~,$

original FSAR. It includes changes necessary to reflect information and analysis submitted to the NRC or prepared pursuant to Commission requirements, and it includes changes describing physical modifications to the plant.

I&M and Westinghouse Electric Corporation have jointly participated in the design and construction of each unit. The plant is operated by I&M.

Each unit employs a pressurized water reactor nuclear steam supply system furnished by Vestinghouse Electric Corporation which is similar in design concept to the majority of the nuclear power plants licensed by the Nuclear Regulatory Commission. Certain components of the auxiliary systems are shared between the two units, but in no case does such sharing result in compromising or impairing the safe and continued operation of either unit. Those systems and components which are shared are identified herein and the effects of the sharing analyzed.

July, 1988

The Unit 1 reactor is currently designed for a power output of 3250 MWt and the Unit 2 reactor is designed for a power output of 3411 MWt, which are their licensed ratings. The approximate gross and net electrical outputs of Unit 1 are 1066 MWe and 1030 MWe and of Unit 2 are 1138 MWe and 1100 MWe, respectively. Containment and engineered safeguards are designed and evaluated for operation at the power rating of 3411 MWt.

Most postulated accidents having off-site dose consequences are analyzed at the power rating of 3411 MWt.

The remainder of Chapter 1 of this report summarizes the principal design features and safety criteria of the nuclear units, pointing out the similarities and differences with respect to other pressurized water nuclear power plants employing the same technology and basic engineering features as the Cook Nuclear Plant.

The research and development program is discussed in Section 1.6. The quality assurance program is discussed in Section 1.7.

Chapter 2 contains a description and evaluation of the site and envi-rons, supporting the suitability of that site for a nuclear plant of the size and type described. Chapters 3 and 4 describe the reactors and the reactor coolant systems, Chapter 5 the containment and related systems, and Chapters 6 through 11 the emergency and other auxiliary systems.

Chapter 12 describes I&M's program for organization and training of plant personnel. Chapter 13 contains an outline and description of the initial tests and operations associated with plant startup.

Chapter 14 is a safety evaluation summarizing the analyses which demon-strate the adequacy of the reactor protection system, and the engineered safety features systems. The consequences of various postulated accidents are within the guidelines set forth in the Nuclear Regulatory Commission regulation 10 CFR 100.

1.0-2 July 1989

by a containment isolation signal derived from the safety injection autanatic activation logic and Phase "B" isolation fran a containment pressure high-high signal.

f) Reliable onmite, diesel-generator power is provided, for the engineered safeguards loads in the event of failure of station auxiliary power. In addition, even if external auxiliary power to, the station is lost concurrent with an accident, power is available for the engineered safeguards frcm onmite diesel.-

generator power to assure protection of the public health and safety for any loss-of-coolant accident.

g) The active ccmponents necessary for the proper operation of the engineered safety features are operable fran the control rocm.

The Engineered Safety Features in this plant are the ECCS, the containment structure, the Ice Condenser Systan, and the, Containment l

Spray System (itaas a,, b, c., d'bove).

1.3.9- SHA'RED PACZLITIES AND EQUIPMENT Separate and similar systems and equipment are'provided for each unit except as noted below. In those instances where ccmponents of' system are shared. by both units, those ccmponents which are shared are either shown, in the following listing or discussed in the applicable Sub-Chapter..

1 ~ 3-'7 July, 1982

a) Chemical and Volume Control System Item Number Shared Boric Acid Tanks Batching Tank Hold-up Tanks ~

3 Recirculation Pump Boric Acid Evaporator Feed Pumps Evaporator Feed Ion Exchangers Boric Acid Evaporator (One temporarily in use as a waste evaporator)

Monitor Tanks 4, Monitor Tank Pumps Evaporator Condensate Demineralizers b) Spent Fuel Pit Cooling System Item Number Shared Spent Fuel Pool Pumps Spent Fuel Pool Demineralizer Spent Fuel Pool Filter Spent Fuel Pool Heat Exchangers Refueling Water Purification Pump c) Fuel Handling System Item Number Shared Spent Fuel Storage Pool New Fuel Storage Area Decontamination Area Spent Fuel Pool Bridge Crane

1. 3-8 July, 1984

1.4.4 RELIABILITYAND TESTABILITY OF PROTECTIVE SYSTEMS Protective systems were designed with a degree of functional reliability and in-service testability which is commensurate with the safety functions to be performed. System design incorporates such features's emergency power availability, preferred failure mode design, redundancy and isolation between control systems and protective systems. In addition, the protective systems were designed such'that no single failure would prevent proper system action when required. 'or design purposes, multiple failures which result from a single event were considered single failures. The proposed criteria of the Institute of Electrical and Electronic Engineers for nuclear power plant protection (IEEE-279) have been utilized in the design of protective systems.

The plant variables monitored and the sensors utilized are identified and discussed at length in Westinghouse proprietary reports submitted in support of the application for an operating license for Donald C. Cook Nuclear Plant and referenced in Chapter 7.

The coincident trip philosophy is carried out to provide a safe and reliable reactor protection system since a single failure will not defeat its function nor cause a spurious reactor trip. Channel independence originates at the process sensor and continues back through the field wiring and containment penetrations to the analog protection racks. The power supplies to the protection sets are fed from instrumentation buses.

Two reactor trip breakers are provided to interrupt power to the control rod drive mechanisms. The breakers main contacts are connected in series. Opening either breaker will interrupt power to all control rod drive mechanisms causing all rods to fall by gravity into the core.

Each reactor trip breaker has an undervoltage trip attachment and a shunt trip attachment. Either attachment trips the breaker. Automatic or manual trip initiation activates both the undervoltage and shunt trip attachments. Each protection channel feeds two logic matrices, one for each undervoltage trip circuit.

July 1991

Each reactor trip channel is designed so that it will go into a trip mode when the channel is de-energized. An open channel or loss of channel power therefore would cause the affected channel'o go into a trip mode. Reliability and independence are obtained by redundancy within each channel, except for back-up reactor trips such as the reactor coolant pump breaker position trip. Reactor trip is implemented by interrupting power to the mechanism on each control rod drive mechanism allowing the rod cluster control assemblies (RCCAs) to be inserted by gravity. The protection system is thus inherently safe in the event of a loss of control rod power.

The components of the protective system are designed and laid out so that the mechanical and thermal environment accompanying any emergency situation in which the components are required to function will not interfere with that function. tl The actuation of the engineered safety features provided for loss-of-coolant accidents (LOCA), e.g,, emergency core cooling'system and containment spray system, is accomplished from redundant signals derived from reactor coolant system, steam flow, and containment instrumentation. Channel independence originates at the process sensor and is carried through to the analog protection racks. De-energizing a channel will cause that channel to go into its trip mode (See Subchapter 7.5).

A comprehensive program of plant testing is executed for equipment vital I to the functioning of engineered safety features. The program consists of performance tests of individual pieces of equipment, and integrated tests of the engineered safety features as a whole, and periodic tests of the actuation circuitry and the performance of mechanical components to assure reliable performance upon demand throughout the plant lifetime.

The following series of periodic tests and checks are conducted to assure that the systems can perform their design functions should they be called on during the plant lifetime.

1 4-14

~ July 1991

design provides for periodic testing of active components for operability and required functional performance as well as incorporating provisions to facil'itate physical inspection of critical components,

3. Heat removal systems are provided within the containment to cool the containment atmosphere under design basis accident conditions.

Two systems of different design principles are provided, the containment spray system and the ice condenser system. These systems have the capacity to adequately cool and reduce the pressure of the containment atmosphere as well as reduce the concentration of halogen fission products.

1.4.8 FUEL AND WASTE STORAGE SYSTEMS Fuel storage and waste handling facilities are designed such that accidental releases of radioactivity will not exceed the guidelines of 10 CFR 100.

During refueling of the reactor, operations are conducted with the spent fuel under water. This provides visual control of the operation at all times and also maintains low radiation levels. The borated refueling water assures subcriticality and also provides adequate cooling for the spent fuel during transfer. Spent fuel is taken from the reactor core, transferred to the refueling cavity, and placed in the fuel transfer canal. Rod cluster control assembly transfer from a spent fuel assembly to a new fuel assembly is accomplished prior to transferring the spent fuel to the spent fuel storage pool. The spent fuel storage I

pool is supplied with a cooling system for the removal of the 'decay heat of the spent fuel. Racks are provided to accommodate the storage of a total of two thousand and fifty fuel assemblies. The storage pool is filled with borated water at a concentration to match that used in the reactor cavity during refueling operations. The spent fuel is stored in a vertical array with sufficient center-to-center 1.4-19 July, 1992

distance between assemblies to assure subcriticality (k ( 0.95) even if unborated water were introduced into the pool. 'heff (3,4) water level maintained in the pool provides sufficient shielding to permit normal occupancy of the area by operating personnel. The spent fuel pool is also provided with systems to maintain water cleanliness and to indicate pool water level. Radiation is continuously monitored and a high radiation level is annunciated in the control room.

Water removed from the spent fuel pool must be pumped out as there are no gravity drains. Spillage or leakage of any liquids from waste facilities 1(

handling within the auxiliary building go to waste drain system floor drains. These floor drains are connected to separate "contaminated" sumps in the auxiliary building.

Postulated accidents involving the release of radioactivity from the fuel and waste storage and handling facilities are shown in Chapter 14 to result in exposures well within the guidelines of 10 CFR 100.

The refueling cavity, the refueling canal, the fuel transfer canal, and the spent fuel storage pool are reinforced concrete structures with a corrosion resistant liner. These structures have been designed to withstand loads due to postulated earthquakes. The fuel transfer tube, which connects the refueling canal and the fuel transfer canal which forms part of the reactor containment, is provided with a valve and a blind flange which closes off the fuel transfer tube when not in use.

1.4.9 EFFLUENTS Gaseous, liquid and solid waste disposal facilities have been designed so that the discharge of effluents and off-site shipments are in accordance with applicable governmental regulations.

1.4-20 July, 1992

American Electric Power Service Corporation 1 Riverside Plaza Columbus. OH 43215 614 223 1000 h~

ANERICAN ELKCTRfC POWER STATEMENT OF POLICY FOR THE DONALD C. COOK NUCLEAR PLANT EQUALITY ASSURANCE PROGRAM POLICY American Electric Power Company Inc., recognizes equality the fundamental importance of controlling the design, modification, and operation of Indiana Michigan Power Company's Donald C. Cook Nucl.ear Plant (Cook Plant) by implementing a planned and documented equality Nuclear Assurance Program, including equality Control, that complies with applicable regulations, codes, and standards.

The Assurance Program has been established to control activities affecting safety-related functions of structures, systems, and components in the Cook Nuclear Plant. The equality Assurance Program supports the goal of maintaining the safety and reliability of the Cook Nuclear Plant at the highest level through a systematic program designed to assure that safety-related items are conducted in compliance with the applicable equality regulations; codes, standards, and established corporate policies and equality practices.

As Chai;rman;;of,"the'!Hoar'd and Chief Executive Officer of American Electric Powet Company", 'I'nc.," I maintain the ultimate responsi bi li ty for the Assurance Program associated with the Cook Nuclear Plant. I have delegated functional responsibility for the Assur ance Program to the American Electric Power Service Corporation (AEPSC) Senior Executive Vice President-Engineering and Construction. He has, with my approval, delegated further responsibilities as outlined in this statement.

IMPLEMENTATION The AEPSC Director-equality Assurance, under the direction of the AEPSC Executive Vice President-Engineering and Construction, has been 'enior assi gned the overal 1 responsibi i ty Director-equality 1 for speci fying the equal i ty Assurance program requirements for the Cook Nuclear Plant and verifying their implementation. The AEPSC Senior Executive Vice President-Engineering and Construction has given the AEPSC Assurance author i ty to stop work on any activity affecting safety-related items that does not meet applicable admi'nistrative, technical, and/or regulatory I

Revised: 2/19/92 1.7-1 July, 1992

Statement of Policy for the Donald C. Cook Nuclear Plant Quality Assurance Program Page 2 requirements. The AEPSC Director-Quality Assurance does not have the authority to stop unit operations, but shall notify appropriate plant and/or corporate management of conditions not meeting the aforementioned cri teria and recommend that unit operations be terminated.

The AEPSC Vice President-Nuclear Operations, under the direction of the AEPSC Senior Executive Vice President-Engineering and Construction, has been delegated responsibility for effectively implementing the Quality Assurance Program. The AEPSC Vice President-Nuclear Operations is the Manager of Nuclear Operations. All other AEPSC divisions and departments, except Quality Assurance, having a supporting role foF the Cook Nuclear Plant are functionally responsible to the Manager of Nuclear Operations.

The Plant Manager, under the direction of the AEPSC Vice President-Nuclear Operations, is delegated the responsibility for establishing the Cook Nuclear Plant Quality Control Program and implementing the Quality Assurance Program at the Cook Nuclear Plant.

The AEPSC Director-Quality Assurance. is responsible for providing technical direction to the Plant Manager for matters relating to the Quality Assurance Program at the Cook Nuclear Plant. The AEPSC Director-Quali ty Assurance is also responsible for maintaining a Quality Assurance Section at the Cook Nuclear Plant to perform required reviews, audi ts, and survei llances, and to provide technical liaison services to the Plant Manager.

The implementation of the Quality Assurance Program is descri bed in the AEPSC General Procedures (GPs) and subtier department/division procedures, Plant Manager's Instructions (PMIs), and subtier Department Head Instructions and Procedures, which in total document the requirements for implementation of the Program.

Each AEPSC and Cook Nuclear Plant organization involved in activities affecting safety-related functions of structures, systems, and components in the Cook Nuclear Plant has'he responsibility to implement the applicable policies and requirements of the Quality Assurance Program.

This responsibility includes being familiar with, and complying with, the requirements of the applicable Quality Assurance Program requi rements .

Revised: 2/1,9/92 July, 1992

Statement of Policy for the equality C. Cook Nuclear Plant Donald Assurance Program Page 3 COMPLIANCE The AEPSC Director-equality Assurance shall monitor compliance with the establ i shed gual i ty Assurance Program. Audi t programs shal 1 be established to ensure that AEPSC and Cook Nuclear Plant activities comply with established program requirements, identify deficiencies or noncompliances and obtain effective and timely corrective actions.

equality Employee's engaged in activities affecting safety-related functions of structures, systems, and components in the Cook Nuclear Plant who believe that the Assurance Program is not being complied with, or that a deficiency in quality exists, should notify their supervisor, the AEPSC Director-equality Assurance, and/or the Plant Manager. If the notification 'does not in the employee's opinion receive prompt or appropriate attention, the employee should contact successively higher levels of management. Employees reporting such conditions shall not be d'iscriminated against by companies of the American Electric Power System.

Discrimination includes discharge or other actions relative to compensation, terms, condi ti ons, or privileges of employment.

R. E. Disbrow Chairman of the Board and Chief Executive Officer American Electric Power Company, Inc..

Revised: 2/19/92 1.7-3 July, 1992

1.7.1 ORGANIZATION 1.7.1.1 SCOPE American Electric Power Service Corporation (AEPSC) is responsible for establishing and implementing the guality Assurance (gA) Program for the operational phase of the Oonald C. Cook Nuclear Plant (Cook Nuclear Plant) . Although authori ty for development and execution of various .

portions of the program may be delegated to others, such as contractors, agents or consultants, AEPSC retains overall responsibility. AEPSC shall evaluate work delegated to such organizations. Evaluations shall be based on the status of safety importance of the activity being performed and shall be initiated early enough to assure effe'cti ve quality assurance duri equality ng the performance of the delegated acti vi ty.

This section of the Assurance Program Oescri ption (gAPD) identifies the AEPSC organizational responsibilities for activities affecting the quality of safety-related nuclear power plant ski':uctures, systems, and components, and describes the authority and duties assigned to them. It addresses responsibilities for both attaining quality objectives and for the functions of establishing the gA Program, and verifying that activities affecting the quality of safety-related items are performed effectively in accordance with gA Program requi rements .

1.7. 1. 2 IMPLEMENTATION 1.7.1.2.1 Source of Authorit The Chairian':,;.o',;;the':.-80ard and Chief Executive Officer of American Electric Power Company, Inc. (AEP) and AEPSC is responsible for safe operation of the Cook Nuclear Plant. Authority and responsibility for effectively implementing the gA Program for plant modifications, operations and maintenance are delegated through the AEPSC Senior Executive Vice President Engineering and Construction, to the AEPSC Vice President - Nuclear Operations (Manager of Nuclear Operations) .

1.7-4 July, 1992

In the operation of a nuclear power plant, the licensee is required to establish clear and direct lines of responsibility, authority and accountability. This requirement is applicable to the organization providing support to the plant, as well as to the plant staff.

The AEPSC corporate support of the Cook Nuclear Plant is the responsibility of the entire organization under the direction of the Manager of Nuclear Operations who maintains primary responsibility for the Cook Nuclear Plant within the corporate organization. The AEPSC Vice President - Nuclear Operations is the Manager of Nuclear Operations. AI-1 other. AEPSC divisions and.departments,. other than the guali ty Assurance: Division, having a supporting .role for nucTear.

operations and for the Cook Nuclear Plant are functionally responsible to the Manager of Nuclear Operations (reference Figure 1.7-1) .;

In order to facilitate a more thorough understanding of the support functions, some of the responsibilities, authorities, and accountabi li ties within the organization are as follows:

1) The responsibilities of the Manager of Nuclear Operations shall be dedicated to the area of Cook Nuclear Plant operations and support..

2) The Manager of Nuclear Operations shall be responsible for, and has the authority to direct, all Cook Nuclear Plant operational and support matters within the corporation and shall make, or concur, in all final decisions regarding significant nuclear safety matters.

3) AEPSC organization managers responsible for Cook Nuclear Plant matters shall be familiar with activities within their scope of responsibility that affect plant safety and reliability. They shall be cognizant of, and sensitive to, internal and external factors that might affect the operations of Cook Nuclear Plant.

1.7-5 July, 1992

4) AEPSC organization managers responsible for Cook Nuclear Plant matters have a commitment to seek and identify problem areas and take corrective action to eliminate unsafe conditions, or to improve trends that will upgrade plant safety and reliability.

5) The Manager of Nuclear Operations shall ensure that Cook Nuclear Plant personnel are not requested to perform inappropriate work or tasks by corporate personnel, and shall control assignments and requests that have the potential for diverting the attention of the Plant Manager from the primary responsibility for safe and reliable plant operation.

6) AEPSC organization managers having Cook Nuclear Plant support responsibilities, as well as the Plant Manager and plant organization managers, shall be familiar with the policy statements from higher management concerning nuclear safety and operational priori ties. They shall be responsible for ensuring that activities under their di rection are performed in accordance with these policies.

1.7. 1.2.2 Res onsibilit for Attainin ualit Ob'ectives in AEPSC Nuclear 0 erations The AEP Chai'i;maj:-.'f~the,",:Bo'ard equality and Chief Executive Officer has delegated the functional responsibility of the equality Assurance Program to the AEPSC Senior Executive Vice President Engineering and Construction.

The AEPSC Director Assurance, under the direction of the AEPSC Senior Executive Vice President Engineering and Construction, is responsible for specifying gA Program requi rements and veri fyi ng their implementation.

The AEPSC Vice President Nuclear Operations, under the di recti on of the AEPSC Senior Executive Vice President Engineering and July, 1992

Construction, is responsible for effectively implementing the gA Program.

The Plan't Hanager, under the direction of the AEPSC Vice President Nuclear Operations, is responsible for establishing the Cook Nuclear Plant guali ty Control Program and implementing the gA Program at the Cook Nuclear Plant.

fiA fi'??'??? 0? fi??'$ fi??fi'i+6 (4 -? Alhfi ? '0fi&?fi'?'i, fi'"3 ?A'ik AEPSC has an independent off-site Nuclear Safety and Design Review Committee (NSDRC) which has been established pursuant to the requirements of the Technical Specifications for the Cook Nuclear Plant.

The function of the NSDRC is to oversee the engineering, design, operation, and maintenance of the Cook Nuclear Plant by performing audits and independent reviews of activities which are specified in the facj;1,:.ity,.',.Technical.-,Specififications;.'?

The Cook Nucl ear Pl ant on-si te revi ew group i s the Indi ana Hi chi gan Power Company (18H) Plant Nuclear Safety Review Committee (PNSRC). This committee has also been established pursuant to the requirements of the Cook Nuclear Plant Technical Specifications. The function of the PNSRC is to review pl.ant operations on a continuing basis and advise the Plant Hanager on matters related to nuclear safety.

1.7. 1.2.3 Cor orate Or ani'zation American Electric Power Com an AEP, the parent holding company, wholl'y owns the common stock of all AEP System subsidiary (operating) companies. The major operating companies and generation subsidiaries are shown in Figure 1.7-2. The Chai'r'man",5f; th'e<'Board and Chief Executive Officer of AEP is the Chief Executive Officer of AEPSC and all operating companies . The responsibility for the functional management of the major operating companies is vested in 1.7-7 July, 1992

the President of each operating company reporting to the AEPSC President and Chief Operating Officer who reports to the AEPSC Chairman of the Board.

American Electric Power Service Cor oration The responsibility for administrative and technical direction of the AEP System and its facilities is delegated to AEPSC. AEPSC provides management and technological services to the various AEP System companies.

0 eratin Com anies The operating 'facilities of the AEP System, are owned and operated by the respective operating companies. The responsibility for executing 'the engineering, design, construction, specialized technical training, and certain operations'upervision is vested in AEPSC, while all, or part, of the administrative functional responsibility is assigned to. the operating companies. In the case of Cook Nuclear Plant, IKN general office staff (headquarters) provides public affairs, accounting, industrial safety direction and procurement support.

The Cook Nuclear Plant is owned and operated by IM which is part of the AEP System.

1.7. 1.2.4 uali t Assurance Res onsi bi lit of AEPSC

1) AEPSC provides the technical direction for the Cook Nuclear Plant, and as such makes the final decisions pertinent to safety-related changes in plant design. Further, AEPSC reviews Nuclear Regulatory Commission (NRC) letters, bulletins, notices, etc., for impact on plant design, and the need for design changes or modifications.

1.7-8

2) AEPSC furnishes quality assurance, engineering, design, construction, licensing, NRC correspondence, fuel management and radiological support activities.

3) AEPSC provides additional service in matters such as supplier qualification, procurement of original equipment and replacement parts, and the process of dedicating commercial grade items or services to safety-related applications.

4) The AEPSC gA Division provides technical direction in quality assurance matters to AEPSC and the Cook Nuclear Plant, and oversees the adequacy, effectiveness and implementation of the gA Program through review and audit activities.

5) Cognizant Engineer (e.g., System Engineer, Equipment Engineer, Lead Engineer, Responsible Engineer, etc.) The cognizant engineer, and/or engineer with the other titles noted, is that AEPSC individual who provides the engineering/design expertise for a particular area of responsibility. Th,is responsibility includes the implementation of the quality assurance and quality control measures for systems, equipment, structures, or functional areas included in that individual's responsibility. The various titles used for the identification of an individual's responsibility and assignment shall be understood to mean the same as cognizant engineer in the respective areas of responsibility. ~

uali t Assurance Res onsibi lit of I&M Cook Nuclear Plant I&M's Cook Nuclear Plant staff operates the .Cook Nuclear Plant in accordance with licensing requirements, including the Technical Specifications ,nd such other commitments as established by the operating licenses. The Plant Mana'ger Instruction (PMI) system and subtier instructions and procedures describe the means by which compliance is achieved and responsibilities are assigned, including interfaces with AEPSC. Figure 1.7-3 indicates the organizational 1.7-9 July, 1992

relationships within the AEP System pertaining to the operation and support of the Cook Nuclear Plant.

1.7. 1.2.5 Or anization AEPSC The Chairman,,of,,the, Board and Chief Executive Officer is ultimately responsible for the QA Program associated with the Cook Nuclear Plant.

This responsibility has been functionally delegated to the AEPSC Senior Executive Vice President Engineering and Construction. The AEPSC Senior Executive Vice President Engineering and Construction has further delegated responsibilities which are administered through the following AEPSC'organizatj.on management personnel:

AEPSC Director Quality Assurance AEPSC Vice President Nuclear Operations AEPSC Vice President Project Management and Construction ualit Assurance Division The AEPSC Director Quality Assurance, reporting to the AEPSC Senior Executive Vice President Engineering and Construction, is responsible for the Quality Assurance Division (QAD) . The QAD consists of the following sections (Figure 1.7-4):

Quality Assurance Engineering Section Nuclear Software Quality Assurance Section Audits and Procurement Section Quality Assurance Support Section Quality Assurance Section (Site)

The QAD is organizationally independent and is responsible to perform the following:

Specify QA Program requirements.

Identify quality problems.

Initiate, recommend, or provide solutions through designated channels.

July, 1992

Veri fy implementation of solutions, as appropriate.

Prepare, issue and maintain gA Program documents, as required.

Verify the implementation of the gA Program through scheduled audits and survei llances.

Verify the implementation of computer software quality assurance through reviews, survei llances and audits.

Audit engineering, design, procurement, construction and operational documents for incorporation of, and compliance with, applicable quality assurance requirements to the extent specified by the AEPSC management-approved gA Program.

~ e Organize and conduct the gA auditor orientation, training, certification and qualification of AEPSC audit personnel.

Provide direction for the collection, storage, maintenance, and retention of quality assurance records.

Maintain, on data base, a i st of suppliers of nuclear (N) i tems 1

and services, plus other selected categories of suppliers.

Identify noncompliances of the established gA Program to the responsible organizations for corrective actions, and report significant occurrences that jeopardize quality,to senior AEPSC management.

Follow up on corrective actions identified by gA during and after disposition implementation.

Review the disposition of conditions adverse to quality to assure that action taken will preclude recurrence.

Conduct in-process gA audits or survei llances at supplier's facilities, as required.

Assist and advise other AEP/AEPSC groups in matters related to the gA Program.

Conduct audits as directed by the NSDRC.

Review AEPSC investigated Problem Reports and associated corrective and preventive action recommendations.

Maintain cognizance of industry and governmental quality assurance requirements such that the gA Program is compatible with requirements, as necessary.

1.7-11 July, 1992

Recommend for revision to, or improvements in, the established gA Program to senior AEPSC management.

Audit dedication plans for commercial grade items and services.

Issue "Stop Work" orders when significant conditions adverse to safety-related items are identified to prevent unsafe conditions from occurring and/or continuing.

Provide AEPSC management with periodic reports concerning the status, adequacy and implementation of the gA Program.

Prepare and conduct special verification and/or surveillance programs on in-house activi ti es, as required or requested.

Routinely attend, and participate in, daily plant work schedule and status meetings.

Provide adequate gA coverage relative to procedural and inspection controls, acceptance criteria, and gA staffing and qualification of personnel to carry out gA assignments.

Determine the acceptability of vendors to supply products and services for safety related applications.

Am li ficati on of S eci fi c Res onsi bi li ties uglification of the AEPSC Director uali t Assurance The AEPSC Director guality Assurance shall possess the following position requirements:

Bachelor's degree in engineering, scientific', or related discipline.

Ten (10) years experience in one of, or a combination of, the following areas: engineering, design, construction, operations, maintenance of fossil or nuclear power generation facilities'r utility facilities'A, of which at least four (4) years must be experience in nuclear quality assurance related activities.

Knowledge of gA regulations, policies, practices and standards.

The same, or higher, organization reporting level as the highest line manager directly responsible for, performing activities affecting the* quality of 1.7-12 July, 1992

safety-related items, such as engineering, procurement, construction and operation, and is sufficiently independent from cost and schedule.

Effective communication channels with other senior management positions.

Responsibility for approval of gA Hanual(s) .

Performance of no other duties or responsibilities unrelated to gA that would prevent full attention to gA matters.

Sto Work Orders The AEPSC gAD is responsible for ensuring that activities affecting the quality of safety-related items are performed in a manner that meets applicable administrative, technical, and regulatory requirements. In order to carry out thi s responsibility, the AEPSC Senior Executive Vice President Engineering and Construction has given the AEPSC Director-

'uali ty Assurance the authority to stop work on any equality activity affecting the quality ofequality safety-related items that does not meet the aforementioned requirements. Stop work'authority has been further delegated by the AEPSC Director Assurance to the AEPSC Assurance Superintendent (si te) .

Director The AEPSC equality Assurance and the AEPSC equality Assurance Superintendent do not have the authority to stop unit operations, but will notify appropriate Cook Nuclear Plant and/or corporate management of conditions which do not meet the aforementioned criteria, and recommend that unit operations be terminated.

July, 1992

A Auditor uglification and Certification Pro ram AEPSC has established and maintains a gA auditor training and certification program for all AEPSC gA auditors.

Problem Idehtification Re ortin and Escalation AEPSC has established mechanisms for the identification, reporting and escalation of problems affecting the quality of safety-related items to a level of. management whereby satisfactory resolutions can be obtained.

Nuclear 0 erations The AEPSC Vice President Nuclear Operations (Manager of Nuclear Operations), reporting to the AEPSC Senior Executive Vice President Engineering and Construction, is responsible for nucleirPojira'ti'o'ns"...";,.

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Formulate policies and practices relati've to safety, licensing, operation, maintenance, fuel management, and radiological support.

Provide the Plant Manager with the technical and managerial guidance, direction and support to ensure the safe operation of the plant.

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Maintain liaison with the AEPSC Director guality Assurance.

Implement the requirements of the AEPSC gA Program.

Maintain knowledge of the latest safety, licensing, and regulatory requirements, codes, standards, and federal regulations applicable to the operation of Cook Nuclear Plant.

1.7-14 July, 1992

Accomplish the procurement, economic, technical, licensing and quality assurance activities dealing with the reactor core and its related fuel assemblies and components.

Prepare bid specifications, evaluate bids, and negotiate and administer contracts for the procurement of all nuclear fuel and related components and services.

maintain a special nuclear material accountability system.

Provide analyses to sopport nuclear steam supply system operation, including reactor physics, fuel economics, fuel mechanical behavior, core thermal hydraulic and LOCA and non-LOCA transient safety analysis and other analysis acti v'i ties as requested, furnish plant Technical Specification changes and other licensing work, and participate in NRC and NSDRC meetings as required by these analyses.

Perform reactor core operation follow-up activities and other reactor core technical support activities as requested, and arrange for support from the fuel fabricator, when needed.

Contract for, arid provide technical support for, disposal of both hi'gh level and low level radioactive waste.

Coordinate the development of neutronics and thermal hydraulic safety codes and conduct safety analyses.

Conduct studies of the Cook Nuclear Plant licensing bases to determine the optimal changes to support unit operations at a lower primary pressure and temperature.

Coordinate NOD computer code development, and provide the interface control for NOD'ith the AEPSC Information System Department and Cook Nuclear Plant.

Obtain and maintain the NRC Operating License and Technical Specifications for the Cook Nuclear Plant.

Act as the communication link between the NRC, AEPSC, and the plant staff.

Perform and coordinate efforts involved in gathering information, performing calculations and generic studi es; preparing criteria, reports, and responses; reviewing items affecting safety; and interpreting regulations.

July, 1992

Review, coordinate, and resolve all matters pertaining to nuclear safety between Cook Nuclear Plant and AEPSC. This includes, but is not limited to: the review of certain plant design changes to ensure that the requirements of 10CFR50.59 are met; the preparation of safety evaluations, or reviews, for any designated subject; the preparation of changes to, and appropriate interpretation of, the plant Technical Specification submi ttals of license amendments; and the analysis of plant compliance with regulatory requirements.

Primary corporate contact for most oral and written communication with the NRC, Provide support in key areas of expertise, such as nuclear engineering, probabi li sti c analysis, thermohydraulic analysis, chemical engineering, mechanical engineering, electrical engineering, and technical writing.

Interface with vendors and other outside organizations on matters connected with the nuclear steam supply system and other areas affecting the safe design and operation of nuclear plants.

Participate, as appropriate, in the review of nuclear plant operating experiences, and relate those experiences to the design and safe operation of Cook Nuclear Plant.

Review, evaluate, and respond to NRC requests for information and Procedures NRC notifications of regulatory changes resulting in plant modifications or new facilities. Such responses are generated in accordance with appropriate AEPSC Administrative .

Develop, specify, and/or review conceptual nuclear safety cri teria for Cook Nuclear Plant in accordance with established regulations.

This includes all information contained in the FSAR, as well as specialized information such as environmental qualification and seismic criteria.

Review and evaluate performance requirements for systems, equipment and materials for compliance with specified safety cri teri a.

Review, on a conceptual basis, plant reports and proposed plant safety-related design changes, to the extent that they a re related 1.7-16 July, 1992

improvements, the ALARA program, the radiation monitoring system, the environmental radiological monitoring and sampling program, dose and shielding analysis, radiochemistry review, .':)Nip1'~ca'0'fojis of::::<federal~i'i!egitl'ati,'ovn's",":

ccCC4v ¹¹Ncc¹,cC'cw)6 VR¹M'.. 4CNhCCv Vv¹CwN and meteorological monitoring.

Cook Nuclear Plant and corporate emergency planning, including procedure development, exercise scheduling, facility procurement and maintenance, and the liaison with off-site emergency planning groups, such as FEMA and the Michigan State Police.

'Review federal codes and regulations as they relate to the development, implementation, revision and distribution of the Modified Amended Security Plan (MASP).

Interface with the plant's security department providing support for the security plan, reviewing security facilities, maintaining secur i ty document files, and developing the employee fitness for duty/background screening program.

Provide Nuclear General Employee Training (NGET) for AEPSC personnel.

Coordinate the development of training for AEPSC personnel who support the operation and maintenance of Cook Nuclear Plant, ensuring a unified, training program meeting annual goals and objectives.

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Serve as technical advisors on plant audits.

Remain cognizant of current decommissioning practices and developments.

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to the ultimate safe operation of th'e plant, for compliance with safety regulations, plant Technical Specifications, the Updated FSAR design basis, and with any other requirements under the Operating License, to determine if there are any unreviewed safety questions as defined in 10CFR50.59.

Perform reviews of Problem Reports and 10CFR21 reviews in accordance with corporate requirements.

Operate the Action Item Tracking System (AIT) for AEPSC internal commitment tracking.

Coordinate design changes for the Cook Nuclear Plant, acting as a focal point within AEPSC. This program primarily involves project

. management responsibilities for scheduling and implementing Request for Changes (RFCs), and includes extensive interfacing with engineering, design, construction, .and Cook Nuclear Plant.

Provide working-level coordination with the Institute of Nuclear Power Opera ti ons (INPO) vari the.-...'ar'e'as'."of.-."INPO,:)tr'alii) oj',"-:,!~seiMiiii';s',"'ndqwarkshops;.".

'0),- .'7:"'b I'X <OX.MAC: This effort includes providing AEPSC access to INPO resources, such as NUCLEAR NETWORK and Nuclear Plant Reliability Data System (NPRDS), and effectively integrating AEPSC and Cook Nuclear Plant efforts towards utilizing INPO recommendations contained in operating experience reports to improve Cook Nuclear Plant performance.

Coordinate daily communication with the Cook Nuclear Plant, provide AEPSC management with a daily plant status report, and make presentations to senior management at regularly scheduled construction staff meetings.

Process incoming vendor information.

Coordinate operations within AEPSC that support the Cook Nuclear Plant Facility Data Base (FDB) .

Contribute to the annual FSAR updates through reviews of Licensee Event Reports, design changes and the Annual Operating Report.

Radiological, emergency and security planning.

Corporate support of the Cook Nuclear Plant's radiation protection and health physics program, technical 're&., ew's and advice on the radiological aspects of desi gn changes, modifications or capital 1.7-17'uly, 1992

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Cook- Nuclear, Plant The Plant Manager reports functionally and administratively to the AEPSC Vice President Nuclear Operations (Manager of Nuclear Operations) and is responsible for the Cook Nuclear Plant activities (Figure 1.7-5) .

ryi..'hvyrv4Viy'Yh:.iy>>- '. ryyr'Vvyyyryh r"Yrr C.:i."CklyN The Cook Nuclear Plant organization is responsible for the following:

Ensure the safety of all facility employees and the general public relative to general plant safety, as well as radiological safety, by maintaining strict compliance with plant Technical Specifications, procedures and instructions.

Recommend facility engineering modification and ini tiate and approve plant improvement requisitions.

Ensure that work practices in all plant departments are consistent with regulatory standards, safety, approved procedures, and plant Technical Specifications.

Provide membership, as required, on the PNSRC.

Maintain close working relationships with the NRC, as well as local, state, and federal government regulatory officials regarding conditions which could affect, or are affected, by Cook Nucl ear Pl ant acti vi ti es.

Set up plant load schedules and arrange for equipment outages.

1.7-23 July, 1992

Develop and efficiently implement all site centralized training acti vi ties.

Administer the centralized facility training complex, simulator, and programs ensuring that program development is consistent with the systematic approach to training, maintain INPO accreditations, regulatory and corporate requirements.

Ensure that human resource activities include employee support programs (i .e., fi tness for duty) consi stent wi th INPO/NUHARC guidelines, company policies, and regulatory requirements and standards..

Administer the NRC approved physical Security Program in compliance with regulatory standards, Hodi fied Amended Security Plan (HASP), and company policy.

Supervise, plan,,and direct the activities related to the maintenance and installation of all Cook Nuclear Plant equipment, structures, grounds, and yards.

Prepare and maintain records and reports pertinent to equipment maintenance and regulatory agency requirements.

Administer contracts and schedule outside contractors'ork ,

forces.

Enforce and coordinate Cook Nuclear Plant regulations, procedures, policies, and objectives to assure safety, efficiency, and conti nui ty in the operation of the Cook Nuclear Plant wi thi n the limits of the operating license and the Technical Specifications and formulation of related policies and procedures.

Plan, schedule, and direct activi ti es relating to the operation of the Cook Nuclear Plant and associated swi tchyards; cooperate in planning and scheduling of work and procedures for refueling and maintenance of the Cook Nuclear Plant; and direct and coordinate fuel loading operations.

Review reports and records, direct general inspection of operating conditions of plant equipment, and investigate any abnormal conditions, making recommendations for repairs. Establish and administer equipment clearance procedures consistent with company, plant, and radiation protection standards; authori ze and arrange

1. 7-24 July, 1992

for equipment outages to meet normal or emergency conditions.

Provide the shift operating crews with appropriate procedures and instructions to assist them in operating the Cook Nuclear Plant safely and efficiently.

Approve operator training programs administered by the Cook Nuclear Plant Training Department designed to provide operating personnel with the knowledge and skill required for safe operation of the facility, and for obtaining and holding NRC operator licenses. Coordinate training programs in plant safety and emergency procedures for Cook Nuclear Plant Operating Department personnel to ensure that each shift group will function properly in the event of injury of personnel, fire, nuclear incident, or civil disorder.

planning and overall conduct-.of scheduled and forced

'dvance outages, including the scheduling and coordination of all plant activities associated with refueling, preventive maintenance, corrective maintenance, equipment overhaul, Technical Specification survei 1.lance, and design change installations.

Coordinate all Cook Nuclear Plant activities associated with the initiation, review, approval, engineering, design, production, examination, inspection, test, turnover, and close out of design changes.

Develop and implement an effective guality Control (gC) Program.

This encompasses, but is not limited to, the planning and directing of quality control activities to assu're that industry codes, NRC regulations, and company instructions and policies regarding quality control for Cook Nuclear Plant are implemented, qualified personnel perform the work, and that these activities are properly documented.

Prepare reports of reportable events which are mandated by the NRC and the Technical Specifications.

Direct the activities of contractor gC/nondestructive examination (NDE) personnel assigned to the Safety and Assessment Department and provide inspections of work performed.

1. 7-25 July, 1992

Prepare statistical reports utilized in NRC Appraisal Meetings and Enforcement Conference.

Coordinate the efforts of outside agencies,, such as American Nuclear Insurers (ANI), INPO, and third-party inspector programs.

Maintain knowledge of developments and changes in NRC requirements, industry standards and codes, regulatory compliance activities, and quality control disciplines and techniques .

~

Stop plant operation in the event that conditions are found which in violation of the Technical Specifications or adverse to

're quality.

Maintain and renew accreditation of training programs .

gualification and certification of I&M personnel performing inspections or tests of major modifications and non-routine maintenance to the requirements of Regulatory Guide 1. 5 and ANSI N45.2.6, except as noted in Appendix B hereto, item 9.

gualification and certification of I&M NDE personnel to the requirements of the AEP NDE Manual.

gualification of I&M personnel performing inspection of normal operating activities to ANSI N18.1.

Proper certification of contractor inspection, test and examination personnel in accordance with Regulatory Guide 1.5, ANSI N45.2.6, ASME B&PV Code and/or SNT-TC-1A, as applicable.

Perform peer inspections of work completed by I&M personnel by independent persons qualified to ANSI N18.7.

Co'nduct of the Inservice Inspection (ISI) Program.

Procurement, receiving, quality control receipt inspection, storage, handling, issue, stock level maintenance, and overall control of stores items.

Provide material service and support in accordance with poli cies and procedures required by AEP Purchasing and Stores, AEPSC gA, and the NRC, which are administered and enforced in a total effort to ensure safety and plant reliability.

Plan and direct engineering and technical studies, nuclear fuel management, equipment performance, instrument and control maintenance, on-site computer systems, Shift Technical Advisors, 1.7-26 July, 1992

and emergency planning for the Cook Nuclear Plant. These activities support daily on-site operations in a safe, reliable, and efficient manner in accordance with all corporate policies, applicable laws, regulations, licenses, and Technical Specification requirements.

Implement station performance testing and monitor programs to ensure optimum plant efficiency.

Direct programs related to on-site fuel management and reactor core physics testing, and ensure satisfactory completion.

Establish testing and preventive maintenance programs related to station instrumentation, electrical systems, and computers.

Recommend alternatives to Cook Nuclear Plant operation, technical or emergency procedures, and design of equipment to improve safety of operations and overall plant efficiency.

Implement the corporate Emergency Plan as it pertains to the Cook Nuclear Plant site.

Provide technical and engineering services in the fields of chemistry, radiation protection, ALARA, and environmental in support of the safe operation of the plant and the health and safety of the employees and the public.

Plan and schedule the activities of the Radi';ation'-':Prot'ection

. AXW r CAw~XCN'Ork'Ci vs% Ã'rA>rYrASFAS D'cpa'rtment of the Cook Nuclear Plant in support of operations and maintenance.

Establish chemistry, radiochemistry, and health physics criteria which ensure maximum equipment life, and the protection of the health and safety of the workers and the public.

Establish sampling and analysis programs which ensure the chemistry, radiochemistry, and health physics criteria are wi thi n the established criteria.

Establish and direct investigations, responses, and corrective actions when outside the established cri teria.

Administer and direct the Cook Nuclear Plant's radioactive waste programs, including volume reduction, packaging and shipping.

Admi ni stration of the gA Records Program.

Maintain the Cook Nuclear Plant Facility Data Base.

1.7-27 July, 1992

Pro 'ect Mana ement and Construction De artment The AEPSC Vice President Project Management and Construction, reporting to the AEPSC Senior Executive Vice President Engineering and Construction, is responsible for the Project Management and Construction Department.

Reporting to the AEPSC Vice President Project Management and Construction are the following:

Site Construction Manager, reporting administratively to the AEPSC Vice President Project Management and Construction, and functionally to the Plant:.,Manager:-:. ,;,Cook::,,'Nuc']ca'rgF3a'nt.'he Project Management and Construction Department is responsible for the following:

Administer and implement construction job orders issued by the Cook Nuclear Plant organization for major modifications, replacement and mai ntenance work with outside contractors.

Administer and monitor contractor's industrial safety programs and performance.

Administer human resources'unctions for site construction organization.

Manage construction labor relations with the .International Building and Construction Trades Unions.

Scope, bid, recommend awards and administer construction labor and services contracts.

Plan, organize and control major construction projects, as assigned by the AEPSC Senior Executive Vice President Engineering and Construction.

Maintain cognizance on matters pertaining to the Cook Nuclear Plant and corporate emergency response organization.

Prepare '.-.;::5 construction labor estimates.

.1.7-28 July, 1992

Provi de constructabi 1 i ty gui dance when requested in support of engineering and design changes.

Participate on the Nuclear Safety Design Review Committee.

Purchasin and Stores De artment (not charted)

The AEPSC Executive Vice President Operations, reporting to the AEPSC P~rcesi dent~and,';Chief.."",-'Operas'kn'g';Office'e":;";: i s responsibl e for the Purchasing and Stores Department through the AEPSC Vice President Purchasing and Haterials Hanagement.

The Purchasing and Stores Department is responsible for the following:

Procurement of safetj-;."related items from only qualified and approved suppliers.

Provide supervision to Cook Nuclear Pl-ant Purchasing Section.

'rovide ordering and stocking descri ptions of sa'fetj,=...ii'il:ite'dr items and include these descriptions in the Cook Nuclear Plant inventory catalog, including necessary communications with suppliers, cognizant engineers, the Cook Nuclear Plant Stores Supervisor and other appropriate personnel.

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Prepare and issue requests for quotations, contracts, service orders and purchase orders for sifetvy.";.rel;ated vXvrrrXCvrCCCi X rX XCNC ) irhvYC WN items.

Establish a system to implement corrective action as described in the AEPSCiGeneral Procedures for the Cook Nuclear Plant.

Establish a system of document keeping and transmittal.

Establish a system of document control for controlled procedures, instructions, and purchasing documents for 'sa'fete...'i'i'.7'at'i.'d wr wivwX.:4vrC w' XvwnXw items.

1.7-29 July, 1992

The maintenance and control, of selected standard procurement document phrases as identified by .the Director guality Assurance, or designee.

Conduct training sessions involving purchasing personnel and others on an annual basis, or more frequently, as required, and ascertain that training sessions include complete responsibilities associated with the purchase of safety-related items.

1.7.2 EQUALITY ASSURANCE PROGRAM 1.7.2. 1 SCOPE Policies that define and establish the Cook Nuclear Plant gA Program are summarized in the individual sections of this document. The program is implemented through procedures and instructions responsive to provisions of the gAPD, and will be carried out for the life of the Cook Nuclear Plant.

guali ty assurance controls apply to activities affecting the quality of safety-related structures, systems and components to an extent based on the importance;";of those structures, systems, components, etc., (items) to safety. Such activities are performed under controlled conditions, including the use of appropriate equipment, environmental condi tions, assignment of qualified personnel, and assurance that all applicable prerequi sites have been met.

Safety-related items are defined as items:

Which are associated with the safe shutdown (hot) of the reactor; or isolation of the reactor; or maintenance of the integrity of the reactor coolant system pressure boundary.

OR Whose failure might cause or increase the severity of a design basis accident as described in the Updated FSAR; or lead to a release of radioactivity in excess of 10CFR100 guidelines.

1.7-30

In general, items are classified as safety-related if they are.: Seismic Class I, or Electrical Class IE; or associated with the Engineered Safety Features Actuation System (ESFAS); or associated with the Reactor Protection System (RPS) .

A special gA Program has been implemented for Fire Protection items (Section 1.7. 19 herein).

The gA Program also includes provision for Radwaste gA in accordance with the requirements of 10CFR71, part H.

gA Program status, scope, adequacy, and comp1tance with IOCFR50, Appendix 8, are regularly reviewed by AEPSC management through reports, meetings, and review of audit results.

The implementation of the gA Program may be accomplished by AEPSC and/or Indiana Michigan Power Company or delegated in whole or in part to other AEP System companies or outside parties. However, AEPSC and/or Indiana Hichigan Power Company retain full responsibility for all activities affecting safety-related items. The performance of the delegated

.organization is evaluated by audit or survei llances on a frequency commensurate with their scope and importance of assigned work.

1.7. 2. 2 IMPLEMENTATION 1.7.2.2.1 Th'8.,:".::Chai,rm'ai:,:::;:i'fi'.th8':.,:Boaj.d;".:",:."an'd!',.Chi ef~Exei'itLv'e.:.:..Offi.cei,. o f AEPSC has stated in a signed, formal "Statement of Policy", that it is the corporate policy to comply with the provisions of applicable codes, standards and regulations pertaining to quality assurance for nuclear power plants as required by the Cook Nuclear Plant operating licenses.

The statement makes this gAPD and the associated implementing procedures and instructions mandatory, and requires compliance by'll responsible organizations and individuals. The statement also identifies the 1.7-31 July, 1992

management positions within the companies vested with responsibility and authority for implementing the program and assuring its effectiveness.

1.7.2.2.2 The QA Program at AEPSC and the Cook Nuclear Plant consist of controls exerci sed by organizations responsible for attaining quality objectives, and by organizations responsible for assurance functions.

The QA Program effectiveness is continually assessed through management review of various reports, NSDRC review of the QA audit program, and shall also be periodically reviewed by independent outside parties as deemed necessary by management.

The QA Program described in this QAPD is intended to apply for the life of the Cook Nuclear Plant.

The QA Program applies to activities affecting the quality of safety-related structures, components, and related consumables during plant operation, maintenance, testing, and all design changes. Safety-related structures, systems and components are identified in the Facility Data Base and other documents. which are developed and maintained for the plant.

necessary by the AEPSC Director As deemed Quality Assurance, or the Plant Hanager, applicable portions of the QA Program controls will be applied to nonsafety-related activities associated with the implementation of the QA Program to ensure that commitments are met (e.g., off-site records storage, training services, etc.) .

1.7.2.2.3 This QAPD, organized to present the QA Program for the Cook Nuclear Plant in the order of the 1 criteria of 10CFR50, Appendix 8, states AEPSC policy for each of the criteria and describes how the controls

'.7-32 July, 1992

pertinent to each are carried out. Any changes made to this gAPD that do not reduce the commitments previously accepted by the NRC must be submitted to the NRC at least annually. Any changes made to this gAPD that do reduce the commitments previously accepted by the NRC must be submitted to the NRC and receive NRC approval prior to implementation.

The submittal of the changes described above shall be made in accordance with the requirements of 10CFR50.54.

The program described in this gAPO will not be intentionally changed in any way that would prevent it from meeting the criteria of 10CFR50, Appendix 8 and other applicable operating license requirements .

1.7.2.2.4 Documents used for implementing the provisions of this gAPO include the following:

Plant Manager Instructions (PMIs) establish the policy at the plant for compliance with specified cri teri a, and assign responsibility to the various departments, as required, for implementation. Plant':.;;Manager';

Proc'e'dures;:-'.,(PcNPc's')",'-'c Department Head Procedures (OHPs); and in'some cases Department- Head Instructions (DHIs), have been prepared to describe the detailed activi ties required to support safe and effective plant operation as per the PMIs.

The PMIs are reviewed by AEPSC gA for concurrence that they will satisfactorily implement regulatory requirements and commitments .

are .'=,;:,,--:! reviewed by the PNSRC prior to approval by the Plant PNIs'nd:.',PMPs Manager.

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1.7-33 July, 1992

AEPSC General Procedures (GPs) are utilized to define corporate policies and requirements for quality assurance, and to implement certain corporate OA Program requirements. AEPSC division/department and/or section procedures are also used to implement gA Program requirements.

GPs may also be used to define policies which are nonprocedural in nature.

When contractors perform work on-site under their own quality assurance programs, the programs are audited for c .appliance and consistency with the applicable requirements of the Cook Nuclear Plant's gA Program and the contract, and are approved by AEPSC gA prior to the start of work.

Implementation of on-site contractor's gA programs, will be audited to assure that the contractor's programs are effective.

1.7.2.2.5 Provisions of the gA Program for the Cook Nuclear Plant apply to activities affecting the quality of safety-related items. Appendix A to this gAPD lists the Regulatory/Safety Guides and ANSI Standards that identify AEPSC's commitment. Appendix 8 describes necessary exceptions and clarifications to the requirements of those documents. The scope of the program, and the extent to which its controls are applied, are established as follows:

a) AEPSC uses the criteria specified in the Cook Nuclear Plant Updated FSAR for identifying structures, systems and components to which the gA Program applies.

b) This identification process results in the Facility Data Base for the Cook Nuclear Plant. This Facility Data Base i s controlled by authorized personnel. Facility Data Base items are determined by 1.7-34 July, 1992

engineering analysis of the function(s) of plant items in relation to safe operation and shutdown.

c) The extent to which controls specified in the gA Program are applied to Facility Data Base items is determined for each item considering its relative importance to safety. Such determinations are based on data in such documents as the Cook Nuclear Plant Technical Specifications and the Updated FSAR.

1.7.2.2.6 Activities affecting safety-related items are accomplished under controlled conditions. Preparations for such activities include consideration of the following:

a) Assigned personnel are qualified.

b) Work has been planned to applicable engineering and/or Technical Specifications.

c) Specified equipment and/or tools are available.

d) Items are in an acceptable status.

e) Items on which work is to be performed are in the proper condition for the task.

Proper approved instructions/procedures for the work are available for use.

g) Items and facilities that could be damaged by the work have been protected, as required.

h) Provisions have been made for special controls, processes, tests and verification methods.

1.7.2.2.7 Responsibility and authority for planning and implementing indoctrination and training of AEPSC and Cook Nuclear Plant staff personnel are specifically designated, as follows:

1.7-35 July, 1992

a) The training and indoctrination program provides for on-going training and periodic familiarization with the gA Program for the Cook Nuclear Plant.

b) Personnel who perform inspection and examination functions are qualified in accordance with requirements of Regulatory Guide 1.8,'NSI N18. 1, Regulatory Guide 1.58, ANSI N45.2.6, the ASME BEPV Code, or SNT-TC-IA, as applicable, and with exceptions as noted in Appendix B hereto.

c) AEPSC gAD auditors are qualified in accordance with Regulatory Guide 1. 146 and ANSI N45.2.23.

d) Personnel assi:gned duties such as special cleaning processes, welding, etc., are qualified in accordance with applicable codes, standards, regulatory guides and/or plant procedures.

e) The training, qualification and certification program includes, as applicable, provisions for retraining, reexamination and recertification to ensure that proficiency 'is maintained.

f) Training, qualification, and certification records including documentation of objectives, waivers/exceptions, attendees and dates of attendance, are maintained at least as long as the personnel involved are performing activities to which the training, qualification and certification is relevant.

g) Personnel responsible for performi ng activities that affect safety-related items are instructed as to the purpose, scope and implementation of the applicable manuals, instructions and procedures.

Hanagement/supe" isory personnel receive functional training to the level necessary to plan, coordinate and administer the day-to-day 1.7-36 July, 1992

verification activities of the gA Program for which they are responsible.

Training of AEPSC and Cook Nuclear Plant personnel is performed employing the following techniques, as applicable: 1) on the job and formal training administered by the department or section the individual works for; 2) formal training conducted by qualified instructors from the Cook Nuclear Plant Training Department or other entities (internal and external to the AEP System); and 3) formal, INPO accredited training conducted by the Cook Nuclear Plant Training Department. Records of training sessions for such training are maintained. Where personnel qualifications or certifications are required, these certifications are performed on a scheduled basis (consi stent with the appropriate code or standard).

Cook Nuclear Plant employees receive introductory training in quality assurance usually within the first two weeks of employment. In addi ti on, AEPSC personnel receive training prior to being allowed unescorted access to the plant. This training includes management's policy for implementation of the gA Program through Plant Manager and Department Head Instructions and Procedures. These instructions also include a description of the gA Program, the use of instructions and procedures, personnel requirements for procedure compliance and the systems and components controlled by the gA Program.

1.7.3 DESIGN CONTROL 1.7.3. 1 SCOPE Design changes are accomplished in accordance with approved design.

Activities to develop such designs are controlled. Depending on the type of design change, these activities include design and field engineering; the performance of physics, seismic, stress, thermal, hydraulic and radiation evaluations,; update of the FSAR; review of accident analyses; the development and control of associated computer programs; studies of material compatibility; accessibility for >nservice 1.7-37 ~u>x, 1>>~

inspection and maintenance; determination of quality standards; and requirement for equipment qualification. The controls apply to preparation and review of design documents, including the correct translation 'of applicable regulatory requirements and design bases into design, procurement and procedural documents.

1. 7.3. 2 IMPLfHfNTATION 1.7.3.2.1 Design changes are c'ontrolled by procedures and instructions and are reviewed as required by 10CFR50.59 and.:.the;:Techni,ca).".',8'peci,':f~.c'ation's:

Safety-related design changes are accomplished by one of two separate processes: Minor Modification (MM), or Request for Change (RFC). Those that do not alter the intended function of the item and can be determined by judgement to have a minimal'verall impact on the item being modified may be implemented via the MM process. All other safety-related design changes, that are not appropriate for MM processing, are implemented via the RFC process.

In cases where design changes could be deemed to be within the scope of RFCs or MMs solely due to possible insignificant adverse seismic effects, the change may be implemented via the Plant Modification (PM) process.

In the case where safety-related items are involved and the change introduces only insignificant adverse seismic effects, the change may be implemented via the Plant Modification (PM) process.

1.7.3.2.2 Oe'sign.-changes 5'V %C 'WPA AC A are reviewed to determine their impact on nuclear safety and to determine if the proposed changes involve an unreviewed safety question as defined by 10CFR50.59. If a design change were to involve 1.7-38 July, 1992

an unreviewed safety question, it would not be approved for implementation until the required NRC approval was received.

RFCs (except those requiring emergency processing), MMs and PMs (having only insignificant seismic effect on safety items) are reviewed and approved prior to implementation, as a minimum, by the cognizant AEPSC s'iiet~ oo',",::and.;:P1'ant", Manager,",:-'j":.'Th'.- PNSRC.':.:a$ so,:,'reije¹is.::;th'ose",'::BFCs"'-,',:"-:-:.',:HHg'.::;.'and PHs:,-,",:f'r:whfc'h,;sa'fi.ty;:,'valuati on's;".':are.:.":,dee'me'0',.":n'e'ces's'ary',::.::,jiu'r'su'ant;"::-:,."t'0

10CFR50.,';59,:a'rid,;",Techni'eel:;-; Speci:;fi.Gati,oA'.:;:::6'-"5,.';l: '6';:

1.7.3.2.3 For RFCs, the Change Contro'l Board established within AEPSC provides an additional review and approval level. The Change Control Board is comprised of members of the Engineering, Design, Nuclear Operations and gA organizations within AEPSC, and is supplemented by other AEPSC organizations or individuals, as required.

The cognizant member of the Change Control Board assigns a lead engineer for each RFC. The lead engineer is responsible for coordinating the RFC activi ties within AEPSC and maintaining close interface with REPS'-;:Si:te Enji'.n'eei'ing',,Suppor't".';:.Propre'ct>'.'Engin'ei'r'i'jg.

1'. 7.3. 2. 4 Proposed RFCs which require emergency processing are originated at the plant, reviewed by the PNSRC, and approved by the Plant Manager. Cook Nuclear Plant management then contacts the AEPSC NOD, and other AEPSC management, as required, describes the change requested, and implements the change only after receiving verbal AEPSC management authorization to proceed. These reviews and approvals are documented and become a part of the RFC Packet.

\

1.7-39 July, 1992

1.7.3.2.5 When RFCs or NNs involve design interfaces between internal or external design organizations, or across technical disciplines, these interfaces are controlled. Procedures are used for the review, approval., release, distribution and revision of documents involving design interfaces to ensure that structures, systems and components are compatible geometrically and functionally with processes and the environment.

Lines of communication are established for controlling the flow of needed design informa'tion across design interfaces, including changes to the information as work progresses. Decisions and problem resolutions involving design interfaces are made by the AEPSC organization having responsibility for engineering direction of the design effort.

1.7.3.2.6

.Checks are performed and documented to verify the dimensional accuracy and completeness of design drawings and specifications.

1.7.3.2.7 RFC design document packages are audited by AEPSC gA to assure that the documents have been prepared, verified, reviewed and approved in accordance with company procedures.

1.7.3.2.8 The extent of, and methods for, design verification are documented. The extent of design verification performed is a function of the importance of the item to safety, design complexity, degree of standardizati on, the state-of-the-art, and similarity with previously proven designs.

Hethods for design verification include evaluation of the applicability of standardized or previously proven designs, alternate calculations, qualification testing and design reviews. These methods may be used 1.7-40 July, 1992

singly or in combination, depending on the needs for the design under consideration.

When design verification is done by evaluating standardized or previously proven designs, the applicability of such designs is confirmed. Any differences from the proven design are documented and evaluated for the intended application.

guali fi cati on testing of prototypes, components, or features is used when the ability of an item to perform an essential safety function cannot otherwise be adequately substantiated. This testing is performed before plant equipment instal-lotion, where possible, but always before reliance upon the item to perform a safety-related function.

gualification testing is performed under conditions that simulate the most adverse design conditions, considering all relevant operating modes. Test requirements, procedures and results are documented.

Results are evaluated to assure that test requirements have been satisfied. Oesign changes shown to be necessary through testing are made, and any necessary retesting or other ver'.fication is performed.

Test configurations are clearly documented.

Oesign reviews are performed by multi-organizational or interdisciplinary groups, or by single individuals. Criteria are established to determine when a formal group review is required, and when review by an individual is sufficient.

Procedures require that minor design changes accomplished by the NH process also receive formal design verification. Applicable design verification activities shall be completed prior to declaring the design change, or portion thereof, operational.

1.7.3.2.9 Persons representing applicable technical 'disciplines are assigned to perform design verifications. These persons are qualified by July, 1992

appropriate education or experience, but are not directly responsible for the design. The designer's immediate supervisor may perform the verification, provided that:

1) The supervisor is the only technically qualified individual.

or

2) The supervisor has .not specified a singular design approach, ruled out design considerations, nor established the design inputs.

and

3) The need is individually documented and approved in advance by the supervisor's management.

and

4) Regularly scheduled gA audits verify conformance to previous items 1 through 3.

Design verification on safety-related design changes shall be completed prior to declaring a design change, or portions thereof, operational.

1.7.3.2.10

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$ 0 Eiig@'e'e'i",i.n'g':..,'Suppor't',,,Pro'ject";~Enj~ ne'er'ing'. Hateri al to per form the desi gn change must meet the specifications established for the original system, or as specified by the lead engineer. For those design changes where testing after completion is required, the testing documentation is reviewed by the organization performing the test and, when specified, by the AEPSC lead engineer or other cognizant engineer(s) . Further, completed design changes are audited/'suizv'ei".l:led by AEPSC gA following installation and testing.

1.7.3.2.11 Changes to design documents, in'eluding field changes, are reviewed, approved and controlled in a manner commensurate with that used for the 1.7-42 July, 1992

ori ginal desi gn. Such changes are evaluated for impact. In formati on on approved changes is transmitted to all affected organizations.

1.7.3.2.12 Error and deficiencies in, and deviations from, approved design documents are identified and dispositioned in accordance with established design control and/or corrective action procedures.

1.7.3.2.13 Established design control procedures provide for:

1) controlled submission of design changes, engineering evaluation, 2)

3) review for impact on nuclear safety,

4) audit by AEPSC gA,

5) design modification,

6) AEPSC managerial review, and

7) approval and record keeping for the implemented design change.

1. 7:4 PROCOREHENT OOCUHENT. CONTROL 1.7.4.1 SCOPE Procurement documents define the characteristics of item(s) to be procured, identify applicable regulatory and industry codes/standards requirements, and specify supplier gA Program requirements to the extent necessary to assure adequate quality.

1. 7.4. 2 IHPLEHENTATION 1.7.4.2.1 Procurement control is established by instructions and procedures.

These documents require that purchase documents be sufficiently detailed to ensure that purchased materials, components and services associated 1.7-43 July, 1992

with safety-related structures or systems are: 1) purchased to specification and code requirements equivalent to those of the original equipment or service (except when the Code of Federal Regulations requires upgrading), 2) properly documented to show compliance with the applicable specifications, codes and standards, and 3) purchased from vendors or contractors who have been evaluated and deemed qualified, or by the commerce'al',-.:grade.:",dedication'yproc'e'sing Procedures establish the review of procurement documents to determine that: quality requirements are correctly stated, inspectable and controllable; there are adequate acceptance criteria; and procurement documents have been prepar'ed, reviewed and approved in accordance with established requirements.

The manager of the originating group, with support of the cognizant AEPSC engineering group, is responsible for assuring that applicable requirements are set forth in procurement documents.

The Cook Nuclear Plant may'request assistance of AEPSC cognizant engineers in any procurement activity..

1.7.4;2.2 The Facility Data Base, in conjunction with other sources, is used for equipment safety classification and procurement grade. AEPSC specifications are usedt to determine requirements, codes or standards that items must fulfill, and define the documentation that must accompany the item to the plant.

Procurement documents for safety-;~related items and services are reviewed to ensure that: correct classification is made; the requirements are properly stated; and that measures have been, or wi 11 be, implemented to assure the requirements are met and adequately provided for .

1.7-44 July, 1992

Purchase requisitions for new safetygrelated items are initiated by the cognizant engineering group which establishes initial requirements.

Replacement/spares are purchased to requirements equivalent to the original unless upgrading is required by Federal Regulations, or deemed necessary by the cognizant engineering group.

1.7.4.2.3 The contents of procurement documents vary according to the item(s) being purchased and its function(s) in the Cook Nuclear Plant.

Provisions of this SWAPO are considered for application to service contractors, also. As applicable, procurement documents'nclude:

a) Scope of work'. to be performed.

b) Technical requirements, with applicable drawings, specifications, codes and standards identified by title, document number, revision and date, with any required procedures, such as special process instructions identified in such a way as to indicate source and need. Imposition of guides/standards on AEPSC/I&M suppliers and subtier suppliers will be on a case-by-case basis depending upon the item or service to be supplied and upon the degree that AEPSC/I8M relies on suppliers to invoke guides/standards.

AEPSC/I&M recognizes that certain suppliers have acceptable 10CFR50, Appendix B gA programs, even through, the suppliers are not committed to Regulatory Guides or industry standards (e.g.

ANSI N45.2.6.). In those cases, in which suppliers are not committed to the same guides/standards as AEPSC/I&M, AEPSC/I8M will assure that (1) the supplier's gA program provides adequate gA controls, regardless of the lack of specific commitment, or (2) controls will be invoked directly by AEPSC/I&M to assure adequate quality of items/services received by suppliers.

c) Regulatory, administrative and reporting requirements.

d) guali ty requi rements appropriate to the complexity and scope of the work, including necessary tests an'd/or> inspections.

1.7-45 July, 1992

e) A requirement for a documented QA Program, subject to QA review and wri tten concurrence prior to the start of work.

A requirement for the supplier to invoke applicable quality requirements on subtier suppliers.

audits g) Provisions for access to supplier, and subtier suppliers',

facilities and records for inspections, survei llances and .

h) Identification of documentation to be provided by the supplier, the schedule of submittals and documents requiring AEPSC approval.

1.7.4.2.4 The AEPSC QA Division performs audits of procurement documents to assure that QA Program requirements have been met. These audits are conducted in accordance with AEPSC QA Division procedures .

1.7.4.2.5 Changes to procurement documents are controlled in a manner commensurate with that used for the original documents.

~

1.7.5 INSTRUCTIONS, PROCEDURES, AND DRAWINGS 1.7.5. 1 SCOPE Activities affecting the quality of safety-related structures, systems and components are accomplished using instructions, procedures and drawings appropriate to the circumstances, including acceptance criteria for determining i f an activity has been satisfactorily completed.

1.7.5.2 IHPLEHENTATION 1.7.5.2.1 Instructions and procedures incorporate: 1) a description of the activity to be accomplished, and 2) appropriate quantitative (such as tolerances and operating limits) and qualitative (such as workmanship and standards) acceptance cri teria sufficient to determine that the

1. 7-46 July, 1992

The PMIs have been classified into the following series:

1000 Pe'r'sonriel,.","=-,.SN'ect'~or(-"";,::PNSM'i'.Pr'or'idtii',es 2000 Administration Document Control, Security, Training, Records, Emergenc j,::,::..P1'assn",-"-','Fi;:re':,;":Pi',otectlo'n';;".;:Cl'ear'ance,,":.Pemi'0'i'..,',!::Ch'e'i,cilia Coon't'ro1,",";~I'nter'nal,:

,)'m::%&xpioa Cl ea'nl,j.'n'cess" i'...'Sp'H 1:;::;Respo'n'se'",,"Styli'idi'ng;:::-'.'".or'ders-;,

's4casoc'.mNY(kA'iNv4 rw'NiÃ'kNv4neri)Nw.as@os.*eserria44 s os...~v'em, Pr&iv&rvNvne...pi4>4wMsi(w Cori<.:ectri:;v'e';.:,.Mai 6 <<wl 6t'xnan'ce.

YS4'h ee G)Nsv( csee':6 mo i i vp 'Ath Procurement, Receiving, Shipping and Storage

'000 4000 Operations, fuel Handling, Surveillance Testing, Test'>>"'Control:s 'ANÃ4amNwe Menem n 0 w'000 Maintenance, Repair, Modi fi cati on, Speci al Processes, Eg and IS I Contr,o.l:jo f,'Contraactoercs 6000 Technical Chemistry/Radiological Controls, Radiations!Prsetecti'on:;-

Performance/Engineering Testing, and Instrument and Control Maintenance and Cal ibrati on, Heassuring..'andeirTe'set,',;Equi jiment 7000 gual r

i ty Assurance, gual i ty Control Program and Condi ti on/Probl em Reporting Instructions and procedures identify the regulatory requirements and commi tments which pertai n to the subject that it will control and establi sh responsibi li ties for implementation. Instructions and procedures may either provide the guidance necessary for the development of supplemental instructions and/or procedures to implement their requirements, or provide comprehensive guidance based on the subject matter.

1.7.5.2.4 Cook Nuclear Plant drawings are produced, controlled and distributed under the control of AEPSC and the Cook Nuclear Plant. AEPSC design drawings are produced by, or under the control of, the AEPSC Nuclear Eengi'neeri'ng,'.Depea'rtemeo't:, under a set of procedures which direct their development and review. These procedures specify requirements for inclusion of quantitative and qualitative acceptance criteria. Speci fic

1. 7-48 July, 1992

acti vi ty has been sati sfactori lyaccompli shed. Hold points for inspection are established when required.

Instructions and procedures pertaining to the specification of, and/or implementation of, the gA Program receive multiple reviews for technical adequacy and inclusion of appropriate quality requirements. Top tier instructions and procedures are reviewed and/or approved by AEPSC gA.

Lower tier documents are reviewed and approved, as a minimum, by management/supervisory personnel trained to the level necessary to plan, coordinate and admini ster those day-to-day verification activities of the gA Program for which they are responsible.

Special procedures may be issued for activities which have short-term applicability.

1.7.5.2.2 AEPSC activities relative to the Cook Nuclear Plant are outlined by procedures which provide the controls for the implementation of these activities. AEPSC has two categories of gA Program implementation procedures: ~

1) General Procedures (GPs) which are applicable to all AEPSC divisions and departments involved with Cook Nuclear Plant.

2) Organizati,'an procedures which apply to the specific division, department or section involved.

1.7.5.2.3 Activities at the Cook Nuclear Plant are controlled using plant procedures.

1.7-47 July, 1992

drawings are reviewed and approved by the cognizant engineering GI;gap7ZRK!ICE ca -, .xac&xcec" S44xama AEPSC has stationed an on-site design staff to provide for the revision of certain types of design drawings to reflect as-built conditions .

1.7.5.2.5 Complex Cook Nuclear Plant procedures are designated as " In Hand" procedures. Examples of " In Hand" procedures are those developed for extensive or complex jobs where reliance on memory cannot be trusted.

Further, those procedures which describe a sequence which cannot be altered, or require the documentation of data during the course of the procedure, are considered. " In Hand" procedures are designated as such by double asterisks (**) which precede the procedure number on the cover sheet, all pages and attachments of a procedure and the corresponding index.

1.7.6 DOCUMENT CONTROL 1.7.6.1 SCOPE Documents controlling activities within the scope defined in 1.7.2 herein are issued and changed according to established procedures.

Documents such as instructions, procedures and drawings, including changes thereto, are reviewed for adequacy, approved for release by authorized personnel, and are distributed and used at the location where a prescribed activity is performed.

Changes to controlled documents are reviewed and approved by the same organizations that performed the original review and approval, or by other qualified, responsible organizations specifically designated in accordance with the procedures governing these documents. Obsolete or superseded documents are controlled to prevent inadvertent use.

1.7-49 July, 1992

1.7.6.2 IMP LBlENTATION 1.7.6.2.1 Controls are established for approval, issue and change of documents in the following categories:

a) Design documents (e.g., calculations, specifications, analyses) b) Drawings and related documents c) Procurement documents d) ., Instructions and procedures e) Updated Final Safety Analysis Report (UFSAR) f) Plant Technical Specifications g) Safeguards documents 1.7.6.2.2 The review, approval, issuance and change of documents are controlled by:

a) Establishment of criteria to ensure that adequate technical and qual i ty requirements are incorporated; b) Identification of the organization responsible for review, approval, issue and maintenance.

c) Review of changes to documents by the organization that performed the initial review and approval, or by the organization designated in accordance with the procedure governing the review and approval of specific types o'f documents.

1.7-50 July, 1992

1.7.6.2.3 Documents are issued and controlled so that:

a) The documents are available prior to commencing work.

b) Obsolete documents are replaced by current documents in a timely manner.

1.7.6.2.4 Paster lists, or equivalent controls, are used to identify the current revision of instructions, procedures, speci fications and drawings.

These control documents are updated and distributed to designated personnel who are responsible for maintaining current copies of the applicable documents. The distribution of controlled documents is performed under procedures requiring receipt acknowledgement and in accordance with established distribution lists.

1.7.6.2.5 In the event a drawing is developed on-site to reflect an as-built configuration, the marked-up drawing is maintained in the Paster Plant File and all holders of the drawing are issued appropriate notification to inform them the revision they hold is not current, cannot be used and, if required, reference must be made to the Haster Plant File drawing.

1.7.6.2.6 Documents prepared for use in training .',",",,-::"; are appropriately marked to indicate that they cannot be used to operate or maintain the facility or to conduct activi ties affecting the quality of safety-related items. At thi Cook Nuclear Plant, unless a document is identified as it is 'automatically assumed 'that the document is

'controlled'r'.":"::,;w'ork~"ng~copy;,,',,'::,only";,

for information l<P use only.

1.7-51 July, 1992

1.7.7 CONTROL OF PURCHASED.ITEMS AND SERVICES 1.7.7. 1 SCOPE Activities that implement approved procurement requests for items and services are controlled to assure conformance with procurement document requirements. Controls include a system of supplier evaluation and selection audits, acceptance of items and documentation upon delivery, and periodic assessment of supplier performance. Objective evidence of quality that demonstrates conformance with specified procurement document requirements is available to the Cook Nuclear Plant site prior to use of equipment, material, or services.

1.7.7.2 IMPLEMENTATION 1.7.7.2.1 AEPSC qualifies suppliers and distributors by performing a documented evaluation of their capability to provide items or services speci fied by procurement documents. Items and services designated as safety-related are purchased from suppliers whose QA programs have been accepted in

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accordance with AEPSC requirements, or from commercial grade suppliers through the AEPSC dedication program. Suppliers of other items/services are;.,'.subj ect ';,ta:.':.eval'uati!on;.';and '/N

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based:,";on.:a'cc8 jta6'ce:.;:'cri',ter'i:a

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Qualification of such suppliers is determined by the AEPSC QA Division.

In the discharge of this responsibility, the AEPSC QA Division may use information generated by other utilities. The supplier, or distributor, must be approved before procurement can be completed. AEPSC is a member of the Nuclear Procurement Issues 5!!??!!<R??X~ Commi:thee (NUPIC), participates in

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joint supplier audits, and shares audit information consistent with NUPIC requirements. The supplier, or distributor, must be acceptable, or acceptable subject to follow-up, before a procurement can be approve and processed. Additional audits will be conducted, as necessary, to meet requirements. Acceptance is not complete until it has been 1.7-52 July, 1992

determined that the suppliers'A program can meet the requirements for the item(s)/service(s) offered.

1.7.7.2.2 For items that are not unique to a nuclear power plant (" Commercial Grade" ) where requirements cannot be imposed in a practical manner at the time of procurement, programs for dedication to safety-related standards are established and accomplished by the AEPSC cognizant engineer prior to the item being accepted for safety-related use.

1.7.7.2.3 In-process audits of suppliers'ctivities during fabrication, inspection, testing and shipment of items are performed when deemed necessary, depending upon supplier qualification status, complexity of the item(s) being furnished, the items'mportance to safety, and/or previous supplier history. These audits are performed by AEPSC gA. The cognizant engineer and/or responsible Cook Nuclear Plant personnel may also participate, if deemed necessary.

1.7.7.2.4 Spare and replacement parts are procured in such a manner that their performance and quality are at least equivalent to those of the parts that will be replaced.

a) Specifications and codes referenced in procurement documents for spare or replacement items are at least equivalent to those for the original items or to properly reviewed and approved revisions.

b) Parts intended as spares or replacement for "off-the-shelf" items, or other items for which quality requirements were not originally specified, are evaluated for performance at least equivalent to the original.

1.7-53 July, 1992

c) Where qual i ty requirements for the original i tems cannot be determined, requirements and controls are established by engineering evaluation performed by qualified individuals. The evaluation assures there is no adverse effect on interfaces, safety, interchangeability, fit, form, function, or compliance with applicable regulatory or code requirements. Evaluation results are documented.

d), Any additional or modified design criteria, imposed after previous procurement of the item(s), are identified and incorporated.

1.7.7.2.5 Instructi ons and procedures address requirements for supplier selection and control, as well as procurement document control. The PHI on receipt inspection of safety-related items addresses the program for inspection of incoming items, including a review of the documentation required under the procurement. Receipt inspection personnel are qualified and certified in accordance with the requirements of ANSI N45.2.6. Provisions for receipt inspection, apply regardless of where the procurement originates. Additional inspections may apply if required by the procurement document.

Where items and/or services are safety-related and procurement is accompli she/ without assistance of AEPSC, supplier selection is limited to those companies identified as being qualified.

1.7.7.2.6 Items:.,recei,ved. at the si,te are: tagged.-, with',,a::,"HOLD"'ag andjor,'-.pl:aced in a -des.ignated area (e.g. new.'ue3) unti 1 receipt. inspected:.. During receipt inspection, designated material characteri sti cs and attri butes are checked, and documentation is checked against the procurement documents. If found acceptable, the "HOLD" tag is removed and replaced with an "ACCEPTED" tag and the item is placed in a designated area of 1.7-54 July, 1992

the storeroom. Item traceability to procurement documents and to end use is maintained through recording of "HOLD" and "ACCEPTED" tag numbers on applicable documents.

Nonconforming items, or missing or questionable documentation results in items being placed on "HOLD" and maintained in a designated, controlled area of the storeroom. If the nonconformance cannot be cleared, the item is either scrapped, returned to manufacturer, or dispositioned through engineering analysis.

1.7.7.2.7 Contractors providing services (on-site) for safety-related components are required to have either a formal quality assurance program and procedures, or they must abide by the Cook Nuclear Plant gA Program and procedures. Prior to their working at the Cook Nuclear Plant, contractor quality assurance programs must be audi ted and approved by AEPSC gA. Contractor procedures must be reviewed and approved by the originating/sponsoring department head. Further, periodic audits of site contractor activities are conducted under the direction ; the AEPSC'uali ty Assurance Superintendent.

1.7.7.2.8 To the extent prescribed in specific procurement documents, suppliers furni sh quality records; documentary evidence that material and equipment either conforms to requi rements or identifies any requirements that have not been met; and descriptions of those nonconformances from the procurement requirements, which have been dispositioned "use-as-is" or "repair." This evidence is retained at the Plant, or at the Service Corporation.

To the extent prescribed in specific procurement agreements, suppliers are required to maintain additional (backup) documents in their record .

system.

1.7-55 July, 1992

In some cases, such as with NSSS, suppliers are designated primary record retention responsibility.

1.7.7.2.9 The capability of suppliers to furni sh valid documeiit'at'i,on is evaluated during procurement document reviews, annual supplier evaluations, and during audits.

1.7.8 IDENTIFICATION AND CONTROL OF ITEMS 1.7.8.1 SCOPE Items are. identified and controlled to prevent their inadvertent use.

Identification of items is maintained either on the items, their storage areas or containers, or on records traceable to the items.

1.7.8.2 IMPLEMENTATION 1.7.8.2.1 Controls are established that provide for the identi fication and control of items (including partially fabricated assemblies).

1.7.8.2.2 Items are identified by physically marking the item or its container, and by maintaining records traceable to the item. The method of i denti ficati on i s such that the qual i ty of 'the i tern i s not degraded.

1.7.8.2.3 Items are traceable to applicable drawings, specifications, or other pertinent documents to ensure that only correct and acceptable items are used. Verification of traceabi li ty is performed and documented prior to release for fabrication, assembly, or installation.

1.7-56 July,. 1992

1.7.8.2.4 Requirements for the identification by use of heat number, part number, serial number, etc., are included in AEPSC Specifications (DCCs) and/or the procurement document.

1.7.8.2.5 Separate storage is provided for incorrect or defective items that are on hold and material which has been accepted for use. All safety-related items are appropriately tagged or identified (stamping, etc.) to provide easy identification as to the items'sage status. Records are li maintained for the issue of items to provide traceabi ty from storage to end use in the Cook Nuclear Plant.

1.7.8.2.6 When materials are subdivided, appropriate identification numbers are transferred to each section of the material, or traceability is maintained through documentation.

1.7.9 CONTROL OF SPECIAL PROCESSES 1.7.9. 1 SCOPE Special processes are controlled and accomplished by qualified personnel using approved procedures and equipment in accordance with applicable codes, standards, specifications, criteria and other special requirements.

1.7.9.2 IMPLEMENTATION 1.7.9.2.1 Processes subject to special process controls are those for which full verification or characterization by direct inspection is impossible or impractical. Such processes include welding, heat treating, chemical 1.7-57 July, 1992

cleaning, application of protective coatings, concrete placement and NDE.

1.7.9.2.2 Special process requirements for chemical cleaning, application of protective coatings and concrete placement are set forth in AEPSC Specifications (DCCs) and/or directives prepared by the responsible AEPSC cognizant engineer. These documents are reviewed and approved by other personnel with the necessary technical competence. AEPSC Specifications are audited by the AEPSC gA Division.

Special process requirements for welding, heat treating and NDE are set" forth in AEPSC Specifications, the AEP Welding and NOE Manuals and plaht procedures. These specifications and manuals are prepared by, or are reviewed and approved by, the AEPSC Cognizant Engineer Welding and NDE Administrator. The administrative controls portion of the NDE Manual is audited by AEPSC gA.

Special process procedures, with the exception of welding and heat treating, are prepared by Cook Nuclear Plant personnel with technical knowledge in the discipline involved. These procedures, which are also reviewed by other personnel with the necessary technical competence, are qualified by testing.

Welding is performed in accordance with procedures contained in the AEP Welding Manual, or by'approved contractor's procedi'i es. These procedures are qualified in accordance with applicable codes, and Procedure gualification Records are prepared. Weld Procedure gualification Records are reviewed and approved by the AEPSC Cognizant Engineer Welding. Weld qualification documentation is retained in the AEP Welding Manual, or the approved contractor's manual.

Contractor welding procedures are qualified by the contractor. These procedures and the qualification documentation are reviewed and approved 1.7-58 July, 1992

by the AEPSC Cognizant Engineer Welding. This documentation is retained by the contractor.

1.7.9.2.3 NDE personnel are qualified and certified by a Cook Nuclear Plant NDE Level III who has been qualified and certified by the designated AEPSC NDE Admini strator. fi Certi cation i s by examination. Personnel qualification is kept current by re-examination at time intervals specified .in.-",,'.'q'ual.:i,:.ficati'on/certficati'on.,';jiro'c'edhr'e's".",.w'hi'ih':,.".i'n accordance with the ASME Code.

Cook Nuclear Plant welders are qualified by the Maintenance Department usinj:.';AEPSC;-;aper'ov'ed procedures. Supervision of Cook Nuclear Plant welder qualifications is performed. by the Maintenance Department.

Examination of specimens is performed under the supervision of the Safety and Assessment Department in accordance wi th the AEP Welding Manual covering welder qualification. Cook Nuclear Plant welder qualification records are maintained for each welder by the Maintenance Department. Contractor and craft welders are qualified by the contractor using procedures approved by the AEPSC Cognizant Engineer Welding in accordance with AEPSC procedures. Contractor and craft welder qualification records are maintained by the contractor.

1.7.9.2.4 gC/NDE Technicians assigned to the Safety and Assessment Department perform nondestructive testing for work performed by Cook Nuclear Plant and contractor personnel. These individuals are qualified to either SNT-TC-1A, or ANSI N45.2.6, and records of the qualifications/

certifications are maintained at Cook Nuclear, Plant.

1.7.9.2.5 For special processes that require qualified equipment, such equipment

's qualified in accordance with applicable codes, standards and specifications.

1.7-59 July, 1992

1.7.9.2.6 Special process qualifications are reviewed during regularly scheduled gA audits. gualification records are maintained in accordance with 1.7.17 herein.

1.7.9.2.7 The documentation resulting from welding and nondestructive testing is reviewed by appropriate personnel.

1. 7. 10 INSPECTION 1.7.10.1 SCOPE Activities affecting the quality of safety-related structures, systems and components are inspected to verify their conformance with requirements. These inspections are performed by personnel other than those who perform the activity. Inspections are performed by qualified personnel utilizing written procedures which establish prerequi sites and provide documentation for evaluating test and inspecti" results.

Direct inspection, process monitoring, or both, are used as necessary.

When applicable, hold points are used to ensure that inspections are-accomplished at the correct points in the sequence of activities.

1.7. 10. 2 IMPLEMENTATION 1.7.10.2.1 Inspections are applied to appropriate activities to assure conformance to. speci fi ed requirements.

Hold points are provided in the sequence of procedures to allow for the inspection, witnessing, exami nati on, measurement, or review necessary to assure that the critical, or irreversible, elements of an activity are 1.7-60 July, 1992

e) Identification of. personnel responsible for performing the inspection.

f) Acceptance and rejection criteria.

g) Recording of the inspection results and the identification of the inspector. r 1.7.10.2.4 Inspectixonvs, are, conducted, using the fol 1 owinYg, pr'ograms:

,a,"..;-.:,'-::;-'.'Peer'. I'n's ection Pro ram;; ..Th'e:::Peer,;-Ins'pection,-,.Progr~am' is'."b'ased::;:.ojr the.;;:,premi s'e" that',- ISA:: 'oersoxnnel'-'.are'aual.i::fi ed.';to', ANSI'-R18 ':1::-'(p9TI);

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being performed as required. Note that hold points may not apply to all procedures, but each must be reviewed for this attribute.

Hold points specify exactly what is to be done (e.g., type of inspection or examination, etc.), acceptance cri teri a, or reference to another procedure, etc., for the satisfactory completion of the hold point.

When included in the sequence of a procedure, the activities required by hold points are completed prior to continuing work beyond that point.

Process monitoring is used in whole, or in part, where .direct inspection alone is impractical or inadequate.

1.7.10.2.2 Training, qualification and certification programs for personnel who perform inspections are established, implemented and documented'in accordance with 1. 7. 2 herein and as described in Appendix B hereto, item 9b, with exceptions as noted therein.

1.7.10.2.3 Inspection requirements are specified in procedures, instructions, drawings, or checklists as applicable. They provide for the following, as appropriate:

a) Identification of applicable revisions of requi red instructions, drawings and specifications.

b) Identification of characteristics and acti vi ti es to be inspected.

c) Inspection methods.

d) Specification of measuring and test equipment having the necessary accuracy.

1.7-61 July, 1992

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1.7.10.2.5 Inspections associated with the packaging and shipment of radioactive waste and materials are conducted using the following program:

a) NRC Licensed Packa in s Inspections of NRC licensed radioactive material packagings shall be performed by individuals independent from the work being performed. The independent inspectors shall be Indiana Michigan Power personnel, qualified in accordance with Regulatory Guide 1.8 and ANSI N18. 1', as a minimum. Additionally, the inspector shall be familiar with the activities being performed.

b) Non-NRC Licensed Packa in s and Containers Inspections of non-NRC licensed radioactive material packagings and containers (shipping and/or burial) shall be performed by Indiana Michigan Power personnel, qualified in accordance with Regulatory Guide 18.

and ANSI N18.1, as a minimum.

c) Trans ortation Vehicles Inspection of transportation vehicles being shipped as "exclusive use", shall be performed by Indiana Michigan Power personnel, qualified in accordance with Regulatory Guide 1.8 and ANSI N18. 1, as a minimum.

d) Other i ns ecti ons and Verification Inspections and verifications of other activities associated with the packaging and shipment of radioactive materials and waste shall be performed by Indiana and 1.7-63 July, 1992

Michigan Power personnel, qualified in accordance with Regulatory Guide 1.8 and ANSI N18.1, as a minimum.

1.7.10.2.6 Inspections are performed, documented, and the results evaluated by designated personnel in order to ensure that the results substantiate the acceptability of the item or work. Evaluation and review results are documented.

1.7.11 TEST CONTROL

1. 7. 11. 1 SCOPE Testing is performed in accordance with established programs to demonstrate that structures, systems and components will perform satisfactorily in service. The testing is performed by qualified personnel in accordance with written procedures that incorporate speci fi ed requirements and acceptance criteria. Types of tests are:

Scheduled Surveillance, preventive maintenance, post-design, qualification.

Unscheduled Pre- and post-maintenance.

Test parameters (including any prerequisites), instrumentation requirements, and environmental conditions are specified in test procedures. Test results are documented and evaluated.

1.7-64 July, 1992

1.7. 11. 2 IMPLEMENTATION 1.7.11.2.1 Tests are performed in accordance with programs, procedures and criteria that designate when tests are required and how they are to be performed.

Such testing includes the following:

a) gualification tests, as applicable, to verify design adequacy'.

b) Acceptance tests of equipment and components to assure their operation prior to delivery or installation.

c) Post-design tests to assure proper and safe operation of systems and equipment prior to unrestricted operation.

d) Surveillance tests to assure continuing proper and safe operation of systems and equipment. The PMI on surveillance testing controls the periodic testing of equipment and systems to fulfill the surveillance requirements established by the Technical Specifications. Controls have been established to identify uncompleted surveillance testing to assure it is rescheduled for completion to meet Technical Specification frequency requirements.

Data taken during surveillance testing is reviewed by appropriate management personnel to assure that acceptance criteria is fulfilled, or corrective action is taken to correct deficiencies.

e) Maintenance tests after preventive or corrective maintenance.

1.7.11.2.2 Test procedures, as required, provide mandatory hold points for witness or revi ew.

1. 7-65 July, 1992

1.7.11.2.3 Testing is accomplished after installation, maintenance, or repair, by surveillance test procedures, or performance tests, which must be satisfactorily completed prior to determining the equipment is in an operable status. All data resulting from these tests is retained at the Cook Nuclear Plant after review by appropriate management personnel.

1.7.12 CONTROL OF.MEASURING AND TEST E(UIPHENT 1.7.12.1 SCOPE Measuring and testing equipment used in activities affecting the quality of safety-related systems, components and structures are properly identified, controlled, calibrated and adjusted at specified intervals to maintain accuracy within necessary limits.

1. 7. 12. 2 IMPLEMENTATION 1.7.12.2.1 Established procedures and instructions are used for calibration and control of measuring and test equipment utilized in the measurement, inspection and monitoring of structures; systems and components . These procedures and instructions describe calibration techniques and frequencies, and maintenance and control of the equipment.

AEPSC gA periodically assesses the effectiveness of the calibration program via the gA audit program.

1.7.12.2.2 Measuring and test equipment is uniquely identified and is traceable to its calibration source.

1.7-66

1.7.12.2.3 A system has been established for attaching, or affixing labels, to measuring and test equipment to display the date calibrated and the next calibration due date, or a control system is used that identifies to potential users any equipment beyond the calibration due date.

1.7.12.2.4 Measuring and test equipment is calibrated at specified intervals.

These intervals are based on the frequency of use, stability characteristics and other conditions that could adversely affect the required measurement accuracy. Calibration standards are traceable to nationally recognized standards; or where such standards do not exist, provisions are established to document the basis for calibration.

The primary standards used to calibrate secondary standards have, except in certain instances, an accuracy of at least four (4) times the required accuracy of the secondary standard. In those cases where the four (4) times accuracy cannot be achieved, the basis for acceptance is documented and is authorized by the responsible manager. The secondary standards have an accuracy that assures equipment being calibrated will be within required tolerances. The basis for acceptance is documented and authorized by the responsible manager.

1.7.12.2.5 Cook Nuclear Plant procedures define the requirements for the control of standards, test equipment and process equipment.

1.7.12.2.6 When measuring and testing equipment used for inspection and testing is found to be outside of required accuracy limits at the time of calibration, evaluations are conducted to determine the validity of the 1.7-67 July, 1992

results obtained since the most recent calibration. Retests, or reinspections, are performed on suspect items. The results of evaluations are documented.

1.7.13 HANDLING, STORAGE, AND SHIPPING 1.7.13.1 SCOPE Activities with the potential for causing contamination or deterioration, by environmental conditions such as temperature or humidity that could adversely affect the ability of an item to perform its safety-related functions and activities necessary to prevent damage or loss, are identified and controlled. These activities are cleaning, packaging, preserving, handling, shipping and storing. Controls are effected through the use of appropriate procedures and instructions.

1. 7. 13. 2 IMPLEMENTATION 1.7.13.2.1 Procedures are used to control the cleaning, handling, storing, packaging, preserving and shipping of materials, components and systems in accordance with designated procurement requirements. These procedures include, but are not limited to, the following functions:

a) Cleaning to assure that required cleanliness levels are achieved and maintained.

b) Packaging and preservation to provide adequate protection against damage or deterioration. When necessary, these procedures provide for special environments, such as inert gas atmosphere, specific moisture content levels and temperature levels.

c) Handling to preclude damage or safety hazards.

1.7-68 July, 1992

d) Storing to minimize the possibility of loss, damage- or deterioration of items in storage, including consumables such as chemicals, reagents and lubricants.

1.7.13.2.2 Controls have been established for limited shelf life items such as "0" rings, epoxy, ubri cants, sol vents and chemi cal s to assure they are 1

correctly identified, stored and controlled to prevent shelf life expired materials from being used in the Cook Nuclear Plant. Controls are established in plant procedures.

1.7.13.2.3 Packaging and shipping requirements are provided to vendors in AEPSC Specifications (DCCs) which are a part of the procurement,.':.,document, or are otherwise specified in. the procurement, document. Controls for receipt i nspection, damaged items and special handling requirements at the Cook Nuclear Plant are established by plant procedures. Special controls are provided to assure that stainless steel components and materials are handled with approved lifting slings.

1.7.13.2.4 Storage and surveillance requirements have been established to assure segregation of storage. Special controls have been implemented for critical, high value, or perishable items. Routine surveillance is

inducted on stored material to provide inspection for damage, rotation of stored pumps and motors, inspection for protection of exposed surfaces and cleanliness of the storage area.

1.7.13.2.5 Special handling procedures have been implemented for the processing of nuclear fuel during refueling outages. These procedures minimize the 1.7-69 July, 1992

risk of damage to the new and spent fuel and the possible release of radioactive material when placing the spent fuel into the spent fuel pool.

1.7.14 INSPECTION, TEST, AND OPERATING STATUS 1.7. 14. 1 SCOPE Operating status of structures, systems and components is indicated by tagging o'f valves and switches, or by other specified means, in such a manner as to prevent inadvertent operation. The status of inspections and tests performed on individual items is clearly indicated by markings and/or logging under strict procedural controls to prevent inadvertent bypassing of such inspections and tests.

1.7. 14. 2 IMPLEMENTATION 1.7.14.2.1 For design change activities, including item fabrication, installation and test, a program exists which specifies the degree of control required for the identification of inspection and test status of structures, systems and components.

Physical identification is used to the extent practical to indicate the status of items requiring inspections, tests, or examinations.

Procedures exist which provide for the use of calibration and rejection stickers, tags, stamps and other forms of identification to indicate test and inspection status. The Clearance Permit System uses various tags to identify equipment and system operability status. Another program establishes a tagging system for lifted leads, etc. For those items requiring cali bration, the program provides for physical indication of calibration status by calibration stickers, or a control system is used.

1.7-70 July, 1992

1.7.14.2.2 Application and removal of inspection and welding stamps, and of such status indicators as tags, markings, labels, etc., is controlled by plant procedures.

The inspection status of materials received at the Cook Nuclear Plant is identified in accordance with established instructions . The status is identified as Hold, Hold for equality Control Clearance, Reject, or Accept.

The inspection status of work in progress is controlled by the use of hold points in procedures. Plant equality Control, or departmental ANSI N18. 1 qualified personnel (reference 1.7.10.2.4 herein), inspect an activity at various stages and sign off the procedural inspection steps.

II The status of welding is controlled through the use of a weld data block which identifies the inspection and NDE status of each weld.

1.7.14.2.3 Required surveillance test procedures are defined in PHIs. These instructions provide for documenting bypassed tests and rescheduling of the test.

The status of testing after minor maintenance is recorded as part of the Job Order. The status of testing after major maintenance is included as part of the procedure, and includes the performance of functional testing and approval of data by supervisory personnel.

Testing, inspection and other operations important to safety are conducted in accordance with properly reviewed and approved procedures.

The PMI for plant procedures requires that procedures be followed as wri tten. Alteration to the sequence of a procedure can only be accomplished by a procedure change which is subject to the same controls 1.7-71 July, 1992

as the original review and approval. When an immediate procedure change is required to continue in-process work or testing and the required complete review and approval process cannot be accomplished, an "On The Spot" change is processed in accordance with the PMI on plant procedures.

1.7.14.2.4 Nonconforming, inoperable, or malfunctioning structures, systems and components are clearly identified by tags, .tickers, stamps, etc., and documented to prevent inadvertent use.

1.7. 15 NONCONFORMING ITEMS 1.7.15.1 SCOPE Materials, parts, or components that do not conform to requirements are controlled in order to prevent their inadvertent use. Nonconforming items are identified, documented, segregated when practical and dispositioned. Affected organizations are noti fi ed of nonconformances .

1.7. 15. 2 IMPLEMENTATION 1.7.15.2.1 Items, services, or activities that are deficient in characteristic, documentation, or procedure, which render the quality unacceptable or indeterminate, are identified as nonconforming and any further use is controlled. Nonconformances are documented and dispositioned, and notification is made to affected organizations. Personnel authorized to disposition, conditionally release and close out nonconformances are designated.

The Job Order System and/or the Condition/Problem Reports (refer to 1.7. 16 herein) are used at Cook Nuclear Plant to identify nonconforming items and ini tiate corrective action'or items which are installed or have been released to the Cook Nuclear Plant. Systems, components, or 1.7-72 July, 1992

materials which require repair or inspection are controlled under the Job Order System. In addition, the various procedures identified in 1.7.14 herein provide for identification, segregation and documentation of nonconforming i tems.

1.7.15.2.2 Nonconforming i tems are identi fied by marking, tagging, segregating, or by documented administrative controls. Documentation describes the nonconformance, the disposition of the nonconformance and the inspection requirements. It also includes signature approval of the disposition.

Completed Job Orders are reviewed by the 'supervisor responsible for accomplishing the work, and the supervisor of the department/section that originated the Job Order. The gA Division periodically audits the Job Order System, and on a sample basis, Job Orders.

1.7.15.2.3 Items that have been repaired or reworked are ;nspected and tested in accordance wi th the original inspection and te;t requirements, or alternatives, that have been documented.

Items that have the disposition of "repai r" or "use-as-is" require documentation justifying acceptability. The changes are recorded to denote the as-built condition.

When required by established procedures, surveillance or operability tests are conducted on an item after rework, repair or replacement.

1.7.15.2.4'isposition of conditional released items are closed out before the items are relied upon to perform safety-related functions.

1.7-73 July, 1992

1. 7. 16 CORRECTIVE 'ACTION 1.7. 16. 1 SCOPE Conditions adverse to quality, such as failures, mal functions, deficiencies, deviations, defective material and equipment, and nonconformances are identified promptly and corrected as soon as practical.

For significant conditions adverse to quality, the cause of the condition is determined, corrective action is taken to correct the immediate problem, and preventive action is implemented to prevent recurrence. In these cases, the condition, cause and corrective action taken is documented and reported to appropriate levels of management.

1.7. 16. 2 IMPLEMENTATION 1.7.16.2.1 Procedures are established that describe the plant and AEPSC corrective action programs. These procedures are reviewed and concurred with by the AEPSC gA Division.

1.7.16.2.2 Condition/Problem Reports provide the mechanism for plant and AEPSC personnel to notify management of conditions adverse to quality.

Condition/Problem Reports are also used to report violations to codes, regulations and the Technical Specifications. Investigations of reported conditions adverse to quality are assigned by management. The Condition/Problem Report is used to document the investigation of a problem; and to identify the need for a design change to correct system or equipment deficiencies, or to i denti fy the need for the i ni ti ation of Job Orders to correct minor deficiencies. Further, Condition/Problem Reports are used to identify those actions necessary to prevent recurrence of the reported condition.

1.7-74 July, 1992

Significant problems, which are so designated on Cor, .tion/Problem Reports, are reviewed by the PNSRC for evaluation of actions taken, or being taken, to correct the deficiency and prevent recurrence.

The AEPSC NSDRC is responsible for assuring that independent reviews of violations (as specified in the Tec'hnical Specifications) are performed.

These violations are considered significant problems which are documented on Condition/Problem Reports. The reviews will provide an independent evaluation of the reported problems and corrective actions.

The AEPSC gA Division periodically audits the corrective action systems for compliance and effectiveness.

r

1. 7. 17 EQUALITY ASSURANCE RECORDS

1. 7. 17. 1 SCOPE Records that furnish evidence of activities affecting the quality of safety-related structures, systems and components are maintained. They are accurate, complete, legible and are protec'.ed against damage, deterioration, or loss. They are identifiable and retrievable.

1. 7,. 17. 2 IMPLEMENTATION 1.7.17.2.1 Documents that furnish evidence of activities affecting the quality of safety-related items are generated and controlled in accordance with the procedure that governs those activities. Upon completion, these documents are considered records. These'ecords include:

a) Results of reviews, inspections, survei llances, tests, audits and material analyses.

b) gualification of personnel, procedures and equipment.

c) Operation logs.

d) Maintenance and modification procedures and related inspection results.

1.7-75 July, 1992

e) Reportable occurrences.

f) Records required by the plant Technical Specifications.

g) Problem Reports.

h) Other documentation such as drawings, specifications, dedication plans, procurement documents, calibration procedures and reports.

Radiographs.

1.7.17.2.2 Instructions and procedures establish the requirements for the identification and preparation of records for systems and equipment under the gA Program, and provide the controls for retention of these records.

Criteria for the storage location of quality related records, and a retention schedule for these records, has been established.

File Indexes have been established to provide direction for filing, and to provide for the retrievability of the records.

Controls have been established for limiting access to the Plant Haster File to prevent unauthorize'd entry, unauthorized removal, and for use of the records under emergency conditions. The Rec0rds,;;:,Ha'ria'geiiieat Supervisor is responsible for the control and operation of the Plant Naster File Room.

1.7.17.2.3 Wi thin AEPSC, each department/di vi sion manager i s responsibl e for the identification', collection, maintenance and storage of records generated by their department/division. Procedures ensure the maintenance of records suffici ent to furnish objective evidence that activi ti es affecting quality are in compliance with the established gA Program.

1.7-76 July, 1992.

When a document becomes a record, it is designated as permanent, or nonp'ermanent, and then transmitted to file. Nonpermanent records have specified retention times. Permanent records are maintained for the life of the plant or equipment, as applicable.

1.7.17.2.5 Only authorized personnel may issue corrections or supplements to records.

1.7.17.2.6 Traceabi 1 i ty between the record and the i tern or activity to which i t applies is provided.

1.7.17.2.7 Except for records that can only be stored as originals, such as radiographs and some strip charts, or micrograohs thereof, records are sto ed in remote, dual facilities to prevent damage, deterioration, or loss due to natural or unnatural causes. When only the single original can be retained, special fire-rated facilities are used.

1.7.18 AUDITS 1.7. 18. 1 SCOPE A comprehensive system of audits is carried out to provide independent evaluation of compliance with, and the effectiveness of, the gA Program, including those elements of the program implemented by suppliers and contractors . Audits are performed in accordance with written procedures or checklists by qualified personnel not having direct responsibility in the areas audited. Audit results are documented and reviewed by management. Follow-up action is taken where indicated.

1.7-77 July, 1992

1.7. 18. 2 IMPLEMENTATION 1.7.18.2.1 AEPSC A Division Res onsibilities .

The basic responsibility for the assessment of the gA Program is vested in the AEPSC gAD. The AEPSC gAO is primarily responsible for ensuring that proper gA programs are established and for verification of their implementation. These responsibilities are discharged in cooperation with the AEPSC and Cook Nuclear Plant management and their staffs.

1.7.18.2.2 Internal audits are performed in accordance with established schedules that reflect the status and importance of safety to the activities being performed. All areas where the requirements of 10CFR50, Appendix B apply are audited within a period of one to two years.

1. 7. 18. 2. 3 The AEPSC gAO conducts audits to verify the adequacy and implementation of the gA Program at the Cook Nuclear Plant and within the AEP System.

gA audit, reports are distributed to th'e appropriate management and the NSORC (all audits) .

1.7.18.2.4 The independent off-site review and audit organization is the AEPSC NSDRC. This committee is composed of AEPSC, 18M and Cook Nuclear Plant management members. An NSDRC Manual has been developed for this committee which contains the NSDRC Charter and procedures. The NSDRC conducts periodic audits of Cook Nuclear Plant operations pursuant to established criteria (Technical Specifications, etc.) .

NSDRC audit reports are submitted for review to the NSDRC membership, the Chairman of the NSORC, and the AEPSC Senior Executive Vice President 1.7-78 July, 1992

Engineering and Construction. Problem Reports provide for the recording of actions taken to correct deficiencies found during these audits.

1.7:18.2.5 The Cook Nuclear Plant on-site review group is the PNSRC. This committee reviews plant operations as a routine evaluation and serves to advise the Plant Manager on matters related to nuclear safety. The composition of the committee is defined in the Technical Speci fi cations .

The PNSRC also reviews instructions, procedures, and design changes for safety-related systems prior to approval by the Plant Manager. In addition, this committee serves to conduct investigations of violations to Technical Specifications, and reviews significant Problem Reports to deter'mine if appropriate action has been taken.

1.7.18.2.6 Audits of suppliers and contractors a'e scheduled based on the status of safety importance of the activities being performed, and are initiated early enough to assure effective quality assurance during design, procurement, manufacturing, construction, installation, inspection and testing.

Principal contractors are required to audit their suppliers systematically in accordance with the criteria established within their quality assurance programs.

1.7.18.2,7 Regularly scheduled aud's are supplemented by "special audits" when significant changes are made in the gA Program, when it is suspected that quality is in jeopardy, or when an independent assessment of program effectiveness is considered nece'ssary.

1.7-79 July, 1992

Audits include an objective evaluation of practices, procedures, instructions, activities and items related to quality; and a review of documents and records to confirm that the gA Program is effective and properly implemented.

1.7;18.2.9 Audit procedures and the scope, plans, checklists and results of individual audits are documented.

1.7.1.2.10 Personnel selected for auditing assignments have experience, or are given training commensurate with the needs of the audit, and have no direct responsibilities in the areas audited.

1.7.18.2.11 Hanagement of the audited organization identifies and takes appropriate action to correct observed deficiencies and to prevent recurrence.

Follow-up is performed by the auditing organization to ensure that the appropriate actions were taken. Such follow-up includes reaudi ts, when necessary.

1.7.18.2.12 The adequacy of the gA Program is regularly assessed by AEPSC management. The following activities constitute formal elements of that assessment:

1.7-80 July, 1992

a) Audit reports, including follow-up on corrective action accomplishment and effectiveness, are distributed to appropriate levels of management.

b) Individuals independent from the gA organization, but knowledgeable in auditing and quality assurance, periodically review the effectiveness of the gA Programs. Conclusions and recommendations are reported to the AEPSC Vice President - Nuclear Operations.

1.7.19 FIRE PROTECTION gA PROGRAM

1. 7.19.1 Introducti on The Cook Nuclear Plant Fire Protection gA Program has been developed using the guidance of NRC Branch Technical Position (APCSB) 9.5-1, equality Appendix A, Section C, "guality Assurance Program," and NRC clarification "Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls, and Assurance," dated June 14, 1977. As such, the Fire Protection gA Program is part of the overall gA Program for the plant. The Fire Protection gA Program encompasses design, procurement, fabrication, construction, surveillance, inspection, operation, maintenance, modification, and audits.

Implementation and assessment of the Fire Protection gA Program is the responsibility of each involved AEPSC and Indiana Michigan Power Company organization.

1.1.19.9 The Fire Protection gA Program is under the management control of AEPSC.

This control consists of:

July, 1992

1) Verifying the effectiveness of the. Fire Protection gA Program through review, surveillance, and audits.

2) Directing formulation, implementation, and assessment of the Fire Protection gA Program by procedural controls.

3) Assuring the gA program is acceptable to the management responsible for fire protection.

The Plant Manager has delegated responsibility to various Cook Nuclear Plant departments for the following fire protection activities:

a) Maintenance of fire protection systems.

b) Testing of fire protection equipment.

c) Fire safety inspections.

d) Fire fighting procedures.

e) Fire drills.

f) Emergency remote shutdown procedures.

g) Emergency repair procedures (10CFR50, Appendix R).

The Fire Protection gA Pr'ogram at the Cook Nuclear Plant aiso provides for inspection of fi re hazards, explosion hazards, and training of fi re brigade and responding fire departments.

The S'a fetj:,;arid:-:."'.A'sse'ssment;".:,.Department':-',,::s-",'.:,;:.Fii X XCNC X v.'??...ChWwXvhNiX?N CweNC XCwNvv??eCV N?X . NNC r'e';,'.:Pryotect'~

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on Shi iW?Ly'ie'vVXCX+WXhvX?iX??v?i?

ft Supervisor on duty, or designee, is designated as the Fire Brigade Leader and coordinates the fire fighting efforts of shift'personnel and the Fire

~ Bri gade. The;.:!Op'er'at] ons: De'p'air tme5t':,p'roti'des"'-:-.",afi~,:.iiid::;vi duil:"':with!:,:pliny hv hhcy hyih' Cvhvi whw?e h ?N h'h ~ vhyhN 4?N ~ hvhvCvWhChvhCCv??C??h (h hC?e?eICCW XWNwweXCCvWW (hwW??N NN'h

'systems',:-'kn'owv1I.-dq'e'."t'o.se'rv'e, as>',;.'an.'-,:. avdhi.'so j<'to:,::,the.:'::Fhcx' i':i'.e',".Bi.';i jade',.'::L'ea'de).",.

?h .. hvwv?Nywh?Sh?i?Nc?ewexv?xv...Nc?WNcNc'.QNhhcixxcvxv'03Nwxveh??&ww?vexv?c Nwecwb  %%4?hyvvwx ckw Ncek>>y vghw wx'yacc?ehvhwxvx" $ x4 1.7. 19.3 Desi n Control and Procurement Document Control guali ty standards are specified in the design documents such as appropriate fire protection codes and standards, and, as necessary, deviations and changes from these quality standards are controlled.

1.7-82 July, 1992

The Cook Nuclear Plant design was reviewed by qualified personnel to ensure inclusion of appropriate fire protection requirements. These reviews include items such as:

1) Verification as to the adequacy of elect ical isolation and cable separation criteria.

2) Verification of appropri ate requirements for room isolation (sealing penetrati ons, floors and other fire barriers) .

3) Determination for increase in fire loadi ngs .

4) Determination for the need of additional fire detection and P'uppression equipment.

Procurement of fire protection equipment and related items are subject to the reoui rements of the fire protecti on procurement documents. A review of these documents is performed to assure fire protection requi rements and quality requirements are correctly stated, verifiable, and controllable,.and that there is adequate acceptance and rejection criteria. Procurement documents must be prepared, reviewed, and approved according to gA Program requirements.

Design and procurement document changes, including field changes and design deviations, are controlled by procedure.

1. 7. 19.4 Instructions Procedures and Drawi n s Inspections, tests, administrative controls, fire drills and trai ni ng that assi st in implementing the fire protection program are prescri bed by approved instructions or procedures.

Indoctrination and training programs for fire prevention and fire fighting are implemented in accordance with approved procedures.

Activities associated wi'th the fire protection systems and fire 1.7-83 "

July, 1992

protection related systems are prescribed and accomplished in accordance with documented instructions, procedures, and drawings. Instructions.

and procedures for design, installati.on, inspection, tests, maintenance, modification and administrative controls are reviewed through audits to assure that the fire protection program is maintained.

Operation and maintenance information has been provided to the plant in the form of System Descriptions and equipment supplier information.

1.7. 19e5 Control of Purchased Items and Services Measures are established to assure that purchased items and services conform to procurement documents. These measures include provisions, as appropriate, for source evaluation and selection, objective evidence- of quality furnished by the contractor, inspections at suppliers, or receipt inspection.

Source or receipt inspection is provided, as a minimum, for those items where quality cannot be verified after installation.

1.7.19.6 ~ins ection A program for independent inspection of the fire protection activities has been established and implemented.

These inspections are performed by personnel other than those responsible for implementation of the activity. The i nspecti ons include:

a) Inspection of installation, maintenance and modification of fire protection systems and equipment.

b) Inspections of penetration seals and fire retardant coating installations to verify the activity is satisfactorily completed in accordance with installation specifications.

1.7-84 July, 1992

c) Inspections of cable routing to verify conformance with design requirements as specified in AEPSC Specifications and/or plant procedures.

d) Inspections to veri fy that appropriate requirements for fire barriers are satisfied following installation, modification, repair or replacement activities.

e) Measures to assure that inspection personnel are independent from the individuals performing the activity being inspected and are knowledgeable in the design and installation requirements for fire protection.

f) =Inspection procedures, instructions or checklists for required inspections.

g) Periodic inspections of fire protection systems, emergency breathing and auxiliary equipment.

h)'eriodic inspections of materials subject to degradation, such 1

as fire stops, seals and fire retardant coating as required by Technical Specifications or manufacturer's recommendations.

1.7. 19.7 Test and Test Control a) Installation testing Following installation, modification, repair, or replacement, sufficient testing i s performed to demonstrate that the fire protection systems and equipment will perform satisfactorily. Written test procedures for installation tests incorporate the requirements and acceptance limits contained in applicable design documents.

b) Periodic testing Peri odic testing occurs to document that fi re protection equipment functions in accordance with its design.

1.7-85 July, 1992

c) Programs have been established to verify the testing of fire protection systems, and to verify that test personnel are effectively trained.

d) Test results are documented, evaluated, and their acceptability determ'ined by a qualified responsible individual or group.

1.7.19.8 Ins ection Test and 0 eratin Status The inspection, test and operating status for plant Technical Specification fire protection systems are performed as described in 1.7. 14 herein.

1. 7. 19. 9 Nonconformin Items Technical Specification fire protection equipment nonconformances are identified and dispositioned as described in 1.7.15 herein.

1.7.19.10 Corrective Action The corrective action mechanism described in 1.7.16 herein applies to the Technical Specification fire protection equipment.

1.7.19.11 Records Records generated

\

to support the fire protection program are controlled as described in 1.7.17 herein.

1.7.19.12 Audi ts Audits are conducted and documented to verify compliance with the Fire Protection gA Program as described in 1.7.1.18 herein.

Audits are periodically performed to verify compliance with the administrative controls and implementation of fi re protect'ion quality 1.7-86 . July, 1992

r assurance criteria. The audits are performed in accordance with pre-established written procedures or checklists. Audit results are documented and reviewed by management having responsibility in the area audited. Follow-up action is taken by responsible management to correct the defi ci enci es reveal ed by the audi t.

1.7-87 July, 1992

AMERICAN ELECTRIC POWER SERVICE CORPORATION Support Organization for the Cook Plant CHAIRMANOF THE BOARD AND CHIEF EXECUTIVE OFFICER

+ PRESIDENT AND CHIEF OPERATING OFFICER SENIOR EXECUTIVE VICE PRESIDENT ENGINEERING AND CONSTRUCTION I

CO CO DIRECTOR VICE PRESIDENT VICE PRESIDENT OTHER AEPSC OTHER AEPSC QUALITY NUCLEAR OPERATIONS

~ PROJECT MANAGEMENT ENGINEERING/DESIGN SUPPORT ASSURANCE AND CONSTRUCTION SUPPORT QUALITY NUCLEAR PROJECT ASSURANCE OPERATIONS MANAGEMENT AND DIVISION DEPARTMENT CONSTRUCTION DEPARTMENT NUCLEAR SITE ENGINEERING CONSTRUCTION DEPARTMENT MANAGER rl ECl, 5

(D PLANT MANAGER COOK NUCLEAR PLANT NOTE S LEGEND

~ 'OT PARTOF AF C ORGANIZATION- SHOWN FOR INFORtAATIONONLY ADtAINISTfIATIVEAND FUNCTIONAL DIRECTION FUNC1 IONAL DIRECTION FOR THE COOK NUCLEAR PLANT

+ EFFECTIVE M I, 1992

Fiaure No. 1.7-3 ORGANIZATIONAL R ELATIONSHIPS WITHIN THE AMERICAN ELECTRIC POWER SYSTEM PERTAINING TO QUALITY ASSURANCE AND QUALITY CONTROL SUPPORT OF THE COOK NUCLEAR PLANT CHAIRMAN OF THE BOARD AND CHIEF EXECUTIVE OFFICER AEPSC, INDIANAMICHIGAN POWER COMPANY AND OTHER AEP SUBSIDIARIES

+ PRESIDENT AND CHIEF OPERATING OFFICER AEPSC, AND VICE PRESIDENT AND DIRECTOR INDIAiNAMICHIGAN POWER COMPANY AND OTHER AEP SUBSIDIARIES SENIOR EXECUTIVE VICE PRESIDENT ENGINEERING & CONSTRUCTION AEPSC, AND VICE PRESIDENT, INDIANA MICHIGAN POWER COMPANY VICE PRESIDENT NUCLEAR OPERATIONS DIRECTOR QUALITY ASSURANCE AEPSC AND AEPSC VICE PRESIDENT, INDIANA MICHIGAN (iUIANAGER OF NUCLEAR OPERATIONS)

NUCLEAR OPERATIONS OTHER AEPSC 8 NUCLEAR ENGINEERING DEPARTMENTS DEPARTMENTS PLANT MANAGER QUALITY ASSURANCE COOK NUCLEAR PLANT SUP ERINTENDEiNT (ON SITE) AEPSC SAFETY & ASSESSMENT LEGEND SUPERINTENDENT Administrative and functional direction Technical direction NOTES ...... Technical liaison

+ EFFECTIVE MARCH 1 1992

~ Functional direction for Cook Nudear Plant Activities July, 1992

1. 7-90

American Electric Power Company

~

AMERICAN ELECTRIC POWER COMPANY American Electric Power Service Corporation Indiana Columbus Appalachian Kingsport I Kentucky Michigan Wheeling Ohio Power Michigan Southern Power Power Power Power Power Company Power Power Company Company Company Company Company Com an Com any I

I I

I Cook I

Nuclear NPlAffV~ Plant administrative, technical, and functional direction

~'KpA8%Itl'~~+j<  ;;:;.'CiiNItai.'j< generating subsidiary

,.Vi)icy,,:P&ek,"',:  !.',:".Op'eratfiijj:,j::

c" CofrIP WYj+%5 "jointly owned with Buckeye Power, Inc.

AEPSC QUALITYASSURANCE DIVISION ORGANIZATION AEPSC DIRECTOR QUALITYASSURANCE SECTION MANAGER SECTION MANAGER SECTION MANAGER SECTION MANAGER QUALITYASSURANCE QUALITYASSURANCE AUDITS AND QUALITYASSURANCE NUCLEAR SOFTWARE SUPERINTENDENT SUPPORT PROCUREMENT ENGINEERING QUALITYASSURANCE (SITE)

Organization for the Cook Nuclear Plant Indiana Michigan Power Company Plant Manager Executive Staff Assistant Plant Safety Chairman Assistant Plant Manager Assistant Plant Manager Assistant Plant Manager Accounting and Production Technical Support Projects Budget Manager Safety and Assessment Superintendent Operations Radiation Protection Integrated Scheduling Superintendent Superintendent Superintendent Quality Control/NDE Maintenance Plant Engineering Computer Sciences Administrative Superintendent Superintendent Superintendent Superintendent Chemistry Nuclear Security Stores Site Construction Manager Superintendent Supervisor Manager n

Training CO C

Emergency Planning Superintendent lD C

C

'C Coordinator z 0'

lO Human Resou ces EO IO Supervi Ql

APPENDIX A REGULATORY AND SAFETY GUIDES ANSI STANDARDS

1. Reg. Guide 1.8 (9/75) Personnel Selection and Training ANSI N18.1 (1971) Selection and Training of Nuclear Power Plant Personnel

2. Reg. Guide 1.14 (8/75) Reactor Coolant Pump Flywheel Integrity

3. Reg. Guide 1.16 (8'/75) Reporting of Operating Information, Appendix A Technical Specifications 4, Safety Guide 30 (8/72) guali ty Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment ANSI N45.2.4 (1972) Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations

5. Reg. Guide 1.33 (02/78) guali ty Assurance Program Requirements (Operation)

ANSI N18.7 (1976) Administrative Controls and guali ty (ANS 3.2 1976) Assurance for the Operational Phase of Nuclear Power Plants ANSI N45.2 (1977) equality Assurance Program Requirements for Nuclear Facilities 1.7-93 July, 1992

equality

6. Reg. Guide 1.37 (3/73) Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants ANSI N45.2. 1 (1973) Cleaning of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants

7. Reg. Guide 1.3 (10/76) gual i ty Assurance Requirements for Packaging, Shipping, Receiving, f

Storage and Handl i ng o I tems for Water-Cooled Nuclear Power Plants ANSI N45.2.2 (1972) Packaging, Shipping, Receiving, S torage and Handl i ng o f I tems for Nuclear Power Plants (During the Construction Phase)

8. Reg. Guide 1.39 (10/76) Housekeeping Requirements for Water-Cooled Nuclear Power Plants ANSI N45.2.3 (1973) Housekeeping During the Construction Phase of Nuclear Power Plants

9. Reg. Guide 1.54 (6/73) equality Assurance Requirements for Protective Coatings Appl,ied to Water-Cooled Nuclear Power Plants ANSI N101. 4 (1972) equality Assurance for Protective Coatings Applied to Nuclear Faci l i ties

10. Reg. Guide 1.58 (9/80) qualification of Nuclear Power Plant In spec ti on, Exami nati on and Testing Personnel 1.7-94 July, 1992

ANSI N45.2.6 (1978) gual i fi cati ons of Inspecti on, Exami-nation, and Testing Personnel for Nuclear Power Pl ants

11. Reg. Guide 1.63 (7/78) Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants

12. Reg. Guide 1.64 (10/73) guality Assurance Requirements for the Design of Nuclear Power Plants ANSI N45.2.11 (1974) equality Assurance Requirements for the Design of Nuclear Power Plants

13. Reg.'Guide 1.74 (2/74) guality Assurance Terms and Definitions ANSI N45.2.10 (1973).. guality Assurance Terms and Definitions

14. Reg. Guide 1.88 (10/76) Collection, Storage, and Maintenance of Nuclear Power Plant guality Assurance Records ANSI N45.2.9 (1974) Requirements for Collection, Storage, and Maintenance of guali ty Assurance Records for Nuclear Power Plants

15. Reg. Guide 1.94 (4/76) guality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 1.7-95 July, 1992

ANS I N45. 2. 5 (1974) Suppl ementary gual i ty Assurance Requirements for Instal l ation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants

16. Reg. Guide 1.108 (8/77) Periodic Testing of Diesel Generator Units used as Onsite Electric Power Systems at Nuclear Power Plants equality

17. Reg. Guide 1.123 (7/77) Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants ANSI N45.2.13 (1976) guality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants

18. Reg. Guide 1.144 (1/79) Audi ting of gual i ty Assurance Programs for Nuclear Power Plants ANSI N45. 2. 12 (1977) Requirements for Auditing of equality Assurance Programs for Nuclear Power Plants equality

19. Reg. Guide 1.146 (8/80) gualification of guality Assurance Program Audit Personnel for Nuclear Power Plants ANSI N45.2. 23 (1978) qualification of Assurance Program Audit Personnel for Nuclear Power Plants 1,7-.96 July, 1992

20. ANSI N45.2.8 (1975) Suppl ementary equal i ty Assurance Requirements for Instal ati on, 1

Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Power Plants

21. ANSI N45.4 (1972) Leakage-Rate Testing of Containment Structures for Nuclear Reactors

1. 7-97 July, 1992

APPENDIX B AEPSC/I&M EXCEPTIONS TO OPERATING PHASE STANDARDS AND REGULATORY GUIDES

1. GENERAL Certain Regulatory Guides invoke, or imply, Regulatory Guides and standards in addition to the standard each primarily endorses.

Certain ANSI Standards invoke, or imply, additional standards.

Exce tion Inter retation The AEPSC/I&M commitment refers to the Regulatory Guides and ANSI Standards specifically identified in Appendix A. Additional Regulatory Guides, ANSI Standards and similar documents implied, or referenced, in those specifically identified are not part of this commitment.

2. N18.7, General Exce tion Inter retation AEPSC and I&M have established both an on-site and off-si te standing committee for independent review activities; together they form the independent review body.

The standard numeric and qualification requirement may not be met by each group individually. Procedures will be established to specify how each group will be involved in review activities. This exception/interpretation is consistent with the plant's Technical Specifications.

1.7-98 July, 1992

2a. Sec. 4.3.1

~Ri t "Personnel assigned responsibility for independent reviews shall be specified in both number and technical disciplines, and shall collectively have the experience and competence required to review problems in the following areas:

Exce tion Inter retation AEPSC Nuclear Safety and Design Review Committee (NSORC) and Plant Nuclear Safety Review Committee (PNSRC) will not have members specified by number, nor by technical disciplines, and its members may not have the experience and competence required to review problems in all areas listed in this section. This exception/interpretation is consistent with the plant's Technical Specifications.

The NSORC and PNSRC will not specifically include a member qualified in nondestructive testing, but will use qualified technical consultants to perform this and other functions as determined necessary by the respective committee chairman.

2b. Sec. 4.3.2.1 "Mhen a standing committee is responsible for the independent review program, it shall be composed of no less than five persons of whom no more than a mi nori ty are members of the on-site operating organization.

Competent alternates are permitted if designated in advance. The use of alternates shall be restricted to legitimate absences of principals'."

Exce tion Inter retation See Item 2a.

1.7-99 July, 1992

2c. Sec. 4.3.3.1 Re uirement

"... recommendations ... shall be disseminated promptly to appropriate members of management having responsibility in the area reviewed."

Exce tion Inter retation Recommendations made as a result of review will generally be conveyed to the on-site, or off-si te, standing committee. Procedures will be maintained specifying how recommendations are to be considered.

2d. Sec. 4.3.4 Re uirement "The following subjects shall be reviewed by the independent review body o ~ ~ ~ ~

Exce ti on Inter retati on Subjects requiring review will be as speci fied in the plant. Technical Specifications.

2e. Sec. 4.3.4(3)

"Changes in the Technical Specifications or License Amendments relating to nuclear safety are to be reviewed by the independent review body prior to implementation, except in those cases where the change is identical to a previously reviewed proposed change."

Exce tion Inter retation Although the usual practice is to meet this requirement, exceptions are made to NSDRC review and approval prior to implementation in rare cases with the permission of the NSDRC Chairman and Secretary. PNSRC review and approval is always done prior to implementation of Technical Specification changes.

1.7-100 July, 1992

2f. Sec. 4.4

~ll "The on-site operating organization shall provide, as part of the normal duties of plant supervisory personnel Exce tion Inter retation Some of the responsibilities of the on-site operating organization described in Section 4.4 may be carried out by the PNSRC and/or NSDRC as described in plant Technical Specifications.

2g. Sec. 5.2.2 "Temporary changes, which. clearly do not change the intent of the approved procedure, shall as a minimum be approved by two members of the plant staff knowledgeable in the areas affected by the procedures. At least one of these individuals shall be the supervisor in charge of the shift and hold a senior operator's license on the uni't affected."

Exce tion Inter retation IKN considers that this requirement applies only to procedures identified in plant Technical Specifications. Temporary changes to these procedures shall be approved as described in plant Techni.cal Speci fi cations .

2h. Sec. 5.2.6 Re uirement

" In cases where required documentary evidence is not available, the associated equipment or materials must be considered nonconforming in accordance with Section 5.2. 14. Until suitable documentary evidence is available to show the equipment or material is in conformance, affected systems shall be considered to be inoperable and reliance shall not be placed on such systems to fulfill their intended safety Functions."

Exce tion Inter retation IM initiates appropriate corrective action when it is discovered that documentary evidence does not exist for a test or inspection which is a 1.7-101 July, 1992

requirement to verify equipment acceptability. This action includes a technical evaluation of the equipment's operability status.,

2i. Sec. 5.2.

"A surveillance testing and inspection program ... shall include the establishment of a master surveillance schedule reflecting the status of all planned in-plant surveillance tests and inspections."

Exce tion Inter retation Separate master schedules may exist for different programs, such as ISI, pump and valve testing, and Technical Specification survei,llance testing.

2j. Sec. 5.2.13.1 "To the extent necessary, procurement documents shall require suppliers to provide a guali ty Assurance Program consistent with the pertinent requirements of ANSI N45.2 1977."

equality Exce tion Inter retation To the extent necessary, procurement documents require that the supplier has a documented Assurance Program consistent with the pertinent requirements of 10CFR50, Appendix B; ANSI N45.2; or other nationally recognized codes and standards.

2k. Sec. 5.2.13.2 Re uirement ANSI N18.7 and N45.2. 13 specify that where required by code, regulation, or contract, documentary evidence that items conform to procurement requirements shall be available at the nuclear power plant site prior to installation or use of such items.

Exce tion Inter retation The required documentary evidence is available at the site prior to use, but not necessarily prior to installation.. This allows installation to 1.7-102 July, 1992

proceed while any missing documents are being obtained, but precludes dependence on the item for safety purposes.

21. Sec. 5.2. 15

~lt "Plant procedures shall be reviewed by an individual knowledgeable in the area affected by the procedure no less frequently than every two years to determine if changes are necessary or desirable."

Exce tion Inter retation Biennial reviews are not performed in that I&M has programmatic control requirements in place that make the biennial review process redundant from a regulatory perspective. These programmatic controls were effected in an effort to ensure that plant instructions and procedures are reviewed for possible revision when pertinent source material is revised, therefore maintaining the procedures current. We believe that this approach, in addition to an annual random sampling of procedures, better addresses the intent of the biennial review process and is more acceptable from both a technical and practical perspective than a static two-year review process.

2m. Sec. 5.2. 16 Records shall be made, and equipment suitably marked, to indicate calibration status.

Exce tion Inter retation See Item 6b.

1. 7-103 July, 1992

2n. Sec. 5.3.5(4)

~lt This section requires that where sections of documents such as vendor manuals, operating and maintenance instructions, or drawings are incorporated directly, or by reference into a maintenance procedure, they shall'eceive the same level of review and approval as operating procedures.

Exce tion Inter retation Such documents are reviewed by appropriately qualified personnel prior to use to'nsure that, when used as instructions, they provide proper and adequate information to ensure the required quality of work. Maintenance procedures which reference these documents receive the same level of review and approval as operating procedures.

3. N45 ~ 2.1, 3a. Sec. 3

~ll N45.2.1 establishes criteria for classifying items into "cleanliness levels," and requires that items be so classified.

Exce tion Inter retation Instead of using the cleanliness level classification system of N45.2. 1, the required cleanliness for specific items and activities is addressed on a case-by-case basis.

Cleanliness is maintained, consistent with the work being performed, so as to prevent the introduction of foreign material. As a minimum, cleanliness inspections are performed prior to closure of "nuclear" systems and equipment. Such inspections are documented.

1.7-104 July, 1992

3b. Sec. 5

~ll "Fitting and tack-welded joints (which will not be immediately sealed by welding) shall be wrapped with polyethylene or other nonhalogenated plastic film until the welds can be completed."

Exce tion Inter retation IKN sometimes uses other nonhalogenated material, compatible with the parent material, since plastic film is subject to damage and does not always provide adequate protection.

4. N45.2.2, General R~i N45.2.2 establishes requirements and criteria for classi fying safety related items into protection levels.

Exce tion Inter retation Instead of classifying safety related items into protection levels, controls over the packaging, shipping, handling and storage of such items are established on a case-by-case basis with due regard for the item's complexity, use and sensitivity to damage. Prior to installation or use, the items are inspected and serviced, as necessary, to assure that no damage or deteriorati on exists which could affect their function.

4a. Sec. 3.9 and Appendix A3.9 "The item and the outside of containers shall be marked."

(Further criteria for marking and tagging are given in the Appendix.)

Exce tion Inter retation These requirements were originally written for items packaged and shipped to construction projects. Full compliance is not always necessary in the case of items shipped to operating plants and may, in some cases, 1.7-105 July, 1992

increase the probability of damage to the item. The requirements are implemented to the extent necessary to assure traceabi li ty and integrity of the item.

4b. Sec. 5.2.2 "Receiving inspections shall. be performed in an area equivalent to the level of storage."

Exce tion Inter retation Receiving inspection area environmental controls may be less stringent than storage environmental requirements for an item. However, such inspections are performed in a manner and in an environment which do not endanger the required quality of the item.

4c. Sec. 6.2.4

~lt "The use or storage of food, drinks and salt tablet dispensers in any storage area shall not be permitted."

Exce tion Inter retation Packaged food for emergency or extended overtime use-may be stored in material stock rooms. The packaging assures that materials are not contaminated. Food will not be "used" in these areas..

4d. Sec. 6.3.4 II "All items and their containers shall be plainly marked so that they are easily identified without excessive handling or unnecessary opening of crates and boxes."

Exce tion Inter retation See N45.2.2, Section 3.9 (Exception 4b.).

1.7-106 July, 1992

4e. Sec. 6.4.1 R~i

" Inspections and examinations shall be performed and documented on a periodic basis to assure that the integrity of the item and its container

... is being maintained."

Exce tion Inter retation The requirement implies tnat all inspections and examinations of items in storage are to be performed on the same schedule. Instead, the inspections and examinations are performed in accordance with material storage procedures which identify the characteristics to be inspected and include the required frequencies. These procedures are based on technical considerations which recognize that inspections and frequencies needed, vary from item to item.

5. N45.2.3, 5a. Sec. 2.1 Cleanliness requirements for housekeeping activities shall be established on the basis of five zone designations.

Exce tion Inter retation P Instead of the five-level zone designation system referenced in ANSI N45. 2.3, I&H bases its controls over housekeeping activities on a consideration of what is necessary and appropriate for the activity involved. The controls are effected through procedures or instructions.

Factors considered in developing the procedures and instructions include cleanliness control, personnel safety, fire prevention and protection, radiation control and securi ty. The procedures and instructions make use of standard janitorial and work practices to the extent possible.

However, in preparing these procedures, consideration is also given to the recommendations of Section 2. 1 of ANSI N45.2.3.

1.7-107 ~

July, 1992

6. N45.2.4, 6a. Sec. 2.2

~ll Section 2.2 establishes prerequisites which must be met before the installation, inspections and testing of instrumentation and electrical equipment may proceed. These prerequisites include personnel qualification, control of design, conforming and protected materials and availability of specified documents..

Exce tion Inter retation During the operations phase, this requirement is consi dered to be applicable to modifications and initial start-up of electrical equipment.

For routine or periodic inspection and testing, the prerequisite conditions will be achieved, as necessary.

6b. Sec. 6.2.1

" Items requiring calibration shall be tagged or labeled on completion, indicating date of calibration and identity of person that performed calibration."

Exce tion Inter retation Frequently, physical size and/or location of installed plant instrumentation precludes attachment of calibration labels or tags.

Instead, each instrument is uniquely identified and is traceable to its calibration record.

A scheduled calibration program assures that each instrument's calibration is current.

7. N45.2.5, 7a. Sec. 2.5.2 "When discrepancies, malfunctions or inaccuracies in inspection and testing equipment are found during cali brati on, all items inspe cted with 1.7-108 July, 1992

that equipment since the, last previous calibration shall be considered unacceptable until an evaluation has been made by the responsible authority and appropriate action taken."

Exce ti on Inter retati on I&A uses the requirements of N18.7, Section 5.2.16, rather than N45.2.5, Section 2.5.2. The N18.7 requirements are more applicable to an operating plant.

7b. Sec. 5.4

~R "Hand torque wrenches used for inspection shall be controlled and must be calibrated at least weekly and more often if deemed necessary. Impact torque wrenches used for inspection must be calibrated at least twice daily."

Exce ti on Inter retati on Torque wrenches are controlled as measuring and test equipment in.

accordance with ANSI N18.7, Section 5.2.16. Calibration intervals are based on use and calibration history rather than as per N45.2.5.

8. N45.2.6, Sec. 1.2 "The requirements of this standard apply to personnel who per form inspections, examinations and tests during fabrication prior to or during receipt of items at the construction site, during construction, during preoperational and start-up testing and during operational phases of nuclear power plants."

Exce ti on Inter retati on Personnel participating in testing who take data or make observations, where special training is not required to perform this function, need not be qualified in accordance with ANSI N45.2.6, but need only be trained to the .extent necessary to perform the assigned function.

1.7-109 July, 1992

9. Re . Guide 1.58 General

~R gualification of nuclear power plant inspection, examination and testing personnel.

9a. C.2.a{7)

Regulatory Guide 1.58 endorses the guidelines of SNT-TC-1A as an acceptable method of training and certifying personnel conducting leak tests.

Exce tion Inter retation 18M takes the position that the "Level" designation guidelines as recommended in SNT-TC-1A, paragraph 4 do not necessarily assure adequate leak test capability. 18M maintains that departmental supervisors are best able to judge whether engineers and other personnel are qualified to direct and/or perform leak tests. Therefore, ISN does not implement the recommended "Level" designation guidelines.

It is I&M's opinion that the training guidelines of SNT-TC-IA, Table I-G, paragraph 5.2 specifically are oriented towards the basic physics involved in leak testing, and further, towards individuals who are not graduate engineers. IN maintains that it meets the essence of these training guideli'nes. The preparation of leak test procedures and the conduct of leak tests at Cook Nuclear Plant is under the direct supervision of Performance Engineers who hold engineering degrees from accredited engineering schools . The basic physics of leak testing have been incorporated into the applicable test procedures. The review and approval of the data obtained from leak tests is performed by department supervisors who are also graduate engineers.

IKH does recognize the need to assure that individuals involved in leak tests are fully cognizant of leak test procedural requirements and thoroughly familiar with the test equipment involved. Plant performance engineers receive routine, informal orientation on testing programs to 1.7-110 July, 1992

ensure that these individuals fully understand the requirements of performing a leak test.

9b. C5, C6, C7, C, C10 Exce tion Inter retation I&M takes the position that the classification of inspection, examination and test personnel (inspection personnel) into "Levels" based on the requirements stated in Section 3.0 of ANSI N45.2.6 does not necessarily assure adequate inspection capability. IEM maintains that departmental and first line supervisors are best able to judge the inspection capability of the personnel under their supervision, and .that "Level" classification would require an overly burdensome administrative work load, could inhibit inspection activities, and provides no assurance of inspection capabilities. Therefore, IN does not implement the "Level" classification concept for inspection, examination and test personnel.

The methodology under which inspections, examinations and tests are conducted at the Cook Nuclear Plant requires the involvement of first line supervisors, engineering personnel, departmental supervisors and plant management. In essence, the last seven (7) project functions shown in Table 1 to ANSI N45.2.6 are assigned to supervisory and engineering personnel, and not to personnel of the inspector category. These management supervisory and engineering personnel, as a minimum, meet the educational and experience requirements of "Level II and Level III" personnel, as required, to meet the criteria of ANSI 1. 1 which exceeds those of ANSI N45.2.6. In IKN's opinion, no useful purpose is served by classification of management, supervisory and engineering personnel into "Levels."

Therefore, IKN takes the following positions relative to regulatory positions C5, 6, 7, and 10 of Regulatory Guide 1.58.

C-5 Based on the discussion in 9b, this position is not applicable to the Cook Nuclear Plant.

1.7-111 July, 1992

C-6 Replacement personnel for Nuclear Plant management, 0

Cook supervisory and engineering positions subject to ANSI 18. 1 will meet the educational and experience requirements of ANSI 18. 1 and therefore, those of ANSI N45.2.6.

Replacement inspection personnel will, as a minimum, meet the educational and experience requirements of ANSI N45.2.6, Section 3.5.1 "Level I."

C-7 I&M, as a general practice, complies with the training recommendations as set forth in this regulatory position.

C-8 All I&M inspection, examination and test personnel are instructed in the normal course of employee training in radiation protection'nd the means to minimize radiation dose exposure.

C-10 I&M maintains documentation to show that inspection personnel meet the minimum requirements of "Level I," and that management, supervisory and engineering personnel meet the minimum requirements of ANSI 18. 1.

10. N45.2.8, 10a. Sec. 2.9e Re uirement Section 2.9e of N45.2.8. lists documents relating to the specific stage of installation activity which are to be available at the construction site.

Exce tion Inter retation All of the documents listed are not necessarily required at the construction site for installation and testing. AEPSC and I&M assure that they are available to the site, as necessary.

July, 1992

10b. Sec. 2.9e

~ll Evidence that engineering or design changes are documented and approved shall be available at the construction site prior to installation.

Exce tion Inter retation Equipment may be installed before final approval of engineering or design changes. However, the system is not placed into service until such changes are documented and approved.

10c. Sec. 4.5.1

" Installed systems and components shall be cleaned, flushed and conditioned according to the requirements of ANSI N45.2. 1. Speci al consideration shall be given to the following requirements:

(Requirements are given for chemical condi'ioning, flushing and process controls.)

Exce tion Inter retation Systems and components are cleaned, flushed and conditioned as determined on a case-by-case basis. Measures are taken to help preclude the need for cleaning, flushing and conditioning through good practices during maintenance or modification activities .

1.7-.113 July, 1992

11. N45.2.9 lla. Sec. 5.4, Item 2 Records shall not be stored loosely. "They shall be firmly attached in binders or placed in folders or envelopes for storage on shelving in containers." Steel file cabinets are preferred.

Exce tion Inter retation Records are suitably stored in steel file cabinets, or on shelving in containers. methods other than binders, folders, or envelopes (for example, dividers) may be used to organize the records for storage.

lib. Sec. 6.2 "A list shall be maintained designating those personnel who shall have access to the files".

Exce tion Inter retation Rules are established governing access to and control of files as provided for in ANSI N45.2.9, Section 5.3, Item 5. These rules do not always include a requirement for a list of personnel who are authorized access. It should be noted that duplicate files and/or microforms may exist for general use.

llc. Sec. 5.6 Re uirement When a single records storage facility is maintained, at least the following features should be considered in its construction: etc.

Exce tion Inter retation The Cook Nuclear Plant Haster File Room and other off-si te record storage facilities comply with the requirements of NUREG-000 (7/81), Section 17.1.17.4.

1.7- 114 July, 1992

12. Re . Guide 1.144 ANSI N45.2.12 12a. Sec. C3a(2)

~ll Applicable elements of an organization's guality Assurance Program for "design and construction phase activities should be audited at least annually or at least once within the life of the activity, whichever is shorter."

Exce tion Inter retation Since most modifications are .straight forward, they are not audited individually. Instead, selected controls over modifications are audited periodically.

12b. Sec. C3b(1)

~ll This section identifies procurement contracts which are exempted from being audited.

Exce tion Inter retation In addition to the exemptions of Reg. Guide 1. 144, AEPSC/IKM considers that the National Institute of Standards and Technology; or other State and Federal Agencies which may provide services to AEPSC/I&M, are not required to be audited.

12c. Sec. 4.5.1

~ll Responses to adverse audit findings, giving results of the review and investigation, shall clearly state the corrective action taken or

" In the event that corrective action planned to prevent recurrence.

cannot be completed within thirty days, the audited organization's response shall include a scheduled date for the corrective action."

1. 7-115 July, 1992

Exce ti on Inter retati on AEPSC/18M take the position that certain circumstances warrant more than thirty (30) days to completely investigate the cause and/or total impact of an adverse finding. For these circumstances, an initial thirty (30) day response will be provided which addresses a schedule for known corrective actions, the reason why additional investigation time is needed, and a schedule for completion of the investigation. These initial responses require the approval of the Director guali ty Assurance.

13. N45.2.13, 13a. Sec. 3.2.2 N45.2. 13 requires that technical requirements be specified in procurement documents b reference to technical requirement documents. Technical requirement documents are to be prepared, reviewed and released under the requirements established by ANSI N45.2. 11.

Exce tion Inter retation For replacement parts and materials, AEPSC/I&M follow ANSI N18.7, Section 5.2. 13, Subitem 1, which states: "Where the. original item or part is found to be commercially 'off the shelf'r without specifically identified gA requirements, spare and replacement parts may be similarly procured, but care shall be exercised to ensure at least equivalent performance."

13b. Sec. 3.2.3 "Procurement documents shall require that the supplier have a documented guality. Assurance Program that implements parts or all of ANSI N45.2 as well as applicable guali ty Assurance Program requirements of other nationally recognized codes and standards."

July, 1992

Exce tion Inter retation Refer to Item 2j.

13c. Sec. 3.3(a)

~R Reviews of procurement documents shall be performed prior to release for bid and contract award.

Exce tion Inter retation Documents may be released for bid or contract award before completing. the necessary reviews. However, these reviews are completed before the item or service is put into service, or before work has progressed beyond the point where it would be impractical to reverse the action taken.

13d. Sec. 3.3(b)

R~i Review of changes to procurement documents 'shall be performed prior to release for bid and contract award.

Exce tion Inter retation This requirement applies only to quality related changes (i.e., changes to the procurement document provisions identified in ANSI N18.7, Section 5.2. 13. 1, Subi tems 1 through 5) . The timing of reviews will be the same as for review of the original procurement documents.

13e. Sec. 10.1 "Where required by code, regulation, or contract requirement, documentary evidence that items conform to procurement documents shall be available at the nuclear power plant site prior to installation or use of such items, regardless of acceptance methods."

1.7-117 July, 1992

Exce tion Inter retation Refer to Item 2j .

"Post-installation test requirements and acceptance documentation shall be mutually established by the purchaser and supplier."

Exce tion Inter retation .

In exercising its ultimate responsibility for its guali ty Assurance Program, AEPSC/18M establishes post-installation test requirements giving due consideration to supplier recommendations.

14. Re . Guide 1. 146 ANSI N45.2.23 and ANSI N45.2.2. 12 14a. ANSI N45.2.23 Sec. 1.1 This standard provides requirements and guidance for the qualification of audi t team leaders, henceforth identified as "lead auditors."

14b. ANSI N45.2. 12, Sec. 4.2.2 A lead auditor shall be appointed team leader.

Exce tion Inter retation The AEPSC audit program is directed by the AEPSC Director guali ty Assurance and is administered by designated gA Division section managers/supervisor who are certified lead auditors.

Audits are, in most cases, conducted by individual auditors, not by "audit teams." These auditors are certified in accordance with established procedures and are assigned by the responsible gA section manager/supervisor based on their demonstrated audit capability and general knowledge of the audit subject. In certain cases, this results in an individual other than a "lead auditor" conducting the actual audit function.

1.7-118 July, 1992

Established AEPSC audit'procedures require that, in all cases, the audit functions of preparation/organization, reporting of audit findings and evaluation of corrective actions be reviewed by gA Division section managers/supervisor, thereby meeting the requirements of ANSI N45.2.23 relative to "lead auditors", and "audit team leaders."

July, 1992

e 2.2 METEOROLOGY Due to the extreme importance of site meteorology, particularly with regard to safety considerations, an extensive meteorological study program was initiated at the site during the summer of 1966.

The meteorological features of the plant site were evaluated primarily on the basis of three years data obtained from the 200 foot tower which was installed on the site in 1966. Satellite aerovane stations at inland and on>>site locations were used to complement the main tower data. Data from the original meteorological study can be found in the original FSAR.

Zn most respects, the meteorological patterns are those of a typical open mid-latitude exposure. The wind speeds are strong, variations in direction are frequent and the overall wind rose shows no marked favoritism for any particular direction. The only unusual feature is J

the low frequency of stable conditions. Both the lapse rate and turbulence class analyses indicate far fewer stable cases than originally anticipated, reaching only 7% over the three year period.

Even in the late spring and early summer when the lake is relatively cold, the frequency of stable cases reaches only 20 to 25%. Even more favorable is the very low frequency of the combination of light winds and stable, on-shore flow. Less than 1% of the 200-foot data and only 2.5% of the satellite data are in this category.

The 1992 meteorological data was obtained from the 10 meter level of the main meteorological tower. Supplemental information from the shoreline tower is also considered. Analysis of the 1992 )oint frequency tables shows that a small percentage (7%) of the year the stability classification indicated a stable state (Pasquill category G). The combination of stable conditions and on-shore wind occurred only about 1% of the time. Both of these figures correspond very closely with the initial data discussed above.

2.2-1 July, 1993

The only major meteorological hazard expected in the site area is the tornado which has recurrence frequency of over 5000 years at the site itself. Ice storms, which would be expected with greater frequency, are not likely to damage essential facilities, but have been considered in developing certain criteria.

2.F 1 SOURCES OF DATA Old Site Meteorolo ical Tower The main source for the inital site meteorological data was a 200-foot meteorological tower which was erected at the site during the summer of 1966 and equipped with meteorological instrumentation (Fig. 2.2-1).

This tower remained in continuous operation from October 1966 until 1978. The tower instruments consisted of the following:

200 ft. level Aerovane and aspirated resistance thermometer.

150 ft. level Climet Bivane (the extremely strong winds at the site had damaged the Bivane, but some data had been obtained).

50 ft. level Aerovane and aspirated resistance thermometer.

Ground-level Resistance thermometer, Dewcell, recording rain gage and recording barometer.

In the Unit 1 Control Room there was instrumentation and a recorder for wind speed, direction and temperature.

2 ' 2 July, 1993

2.2 2 GENERAL METEOROLOGY Southwestern Michigan is typical of the northern lake regions of the United States in most respects..The flat terrain and the frequent passage of well-developed extra-tropical storms create a consistently strong wind flow, as well as rapid changes in both dispersion conditions and wind direction. Some of the meteorological statistics are useful primarily for general planning of the facilities and are therefore reported with a minimum of description. Other data are important in the assessment of safety and these are discussed fully.

Hi h Winds Strong winds are the most important meteorological hazard to the facilities. The region is frequented by relatively strong, gusty winds, usually accompanying the passage of squall lines or thunderstorms and the maximum wind associated with these phenomena is 90 mph on a 100 year recurrency interval.

The tornado presents a very specialized type of hazard involving both violent winds and extremely large, rapid changes in barometric pressure.

The storms are small, unpredictable in detail and rather infrequent, but they undoubtedly represent one of the few environmental factors that could, if ignored in plant design, inflict direct major damage on the facility. Typically, the tornado is a narrow funnel, often only a few hundred yards wide, in which winds may briefly reach 300 mph. Almost instantaneous changes in barometric pressure occur, reaching 3 psi and causing explosion of vulnerable structures. Because of the severity of the phenomena, very few reliable measurements of tornado intensities exist. Zt is therefore difficult to dissociate wind and pressure effects, but the estimates given above are considered fairly reliable maximum values. This portion of Michigan has a signifi-cant tornado probability, as is apparent in the map shown in Figure 2.2-

2. Berrien County has had 25 tornadoes between 1950 and 1989.

2~2 7 July, 1993

Ice Storms Far less destructive, but far more probable, are the ice storms that frequent the north central states. Michigan lies in the belt where such storms are common and in the years from 1970 to 1989, 6 significant ice storms have been reported in this area.

2 2.3 DISPERSION METEOROLOGY Worst Case X values X/g values for 1992 were calculated from data from the main tower's 10 meter instruments using the MIDAS computer code. The data show a worst case X/Q value as 1.06E<<05 sec/m during 1992. The worst case ever computed by the present meteorological system is 1.13E-05 sec/m . Both of these values are well below the established worst case X/Q value of 3.15E-04 sec/m . Table 2.2-3 shows the X/Q values for all sectors at 10 distances.

Atmos heric Stabilit The atmospheric stability for the area is now classified according to the Pasquill categories for use in dispersion calculations. These categories range from A to G, with the G category being the most stable.

Joint frequency tables for 1992 have been compiled and are shown in Tables 2.2-4 through 2.2-11. The data show that a large percentage (33%) of the year is devoted to Category D'. A rather substantial portion of the year (23%) shows an extremely unstable classification (Category A). There is only a small portion of time (7%) devoted to the extremely stable conditions of Category G.

Wind S eed and Direction Wind speeds were moderate in 1992. The predominant wind speed range is 4-7 mph category. The wind speed exceeded 14 mph less than 4% of the time. The wind direction at the main tower varied, with the largest 2.2-8 July, 1993

frequencies occuring both from the North and from the South. 'This can be observed in the wind roses shown in Figures 2.2-3 through 2.2-7.

There was a slight tendency for winds from the West (onshore flow) to occur. The second quarter of the year produced winds mostly from the North, while during the fourth quarter they were from the South.

The wind at the shoreline (measured by the shoreline tower) shows a large contrast. The winds are mostly from the South East. This directional preference can be seen in all four quarters of the year, as shown on the wind roses in Figures 2.2-8 through 2.2-12.

2.2-9 July, 1993

REFERENCES SECTION 2.2

1. Fawbushg Miller and Starrett: A Em irical Method of Forecastin Tornado Develo ent, Bulletin, AMS, 32, 1951.

2. Spohn et. al.t Tornado Climatolo , Monthly Weather Review, Wash.,

D.C , 1962

3. Thorn: Tornado Probabilities, Monthly Weather Review, Wash., D.C.i 1963.

4. Thorn: Distributions of Extreme Winds in the United States, Journal, Struct. Div. ASCE, April, 1960.

5. Singer and Smiths Relation of Gustiness to Other Meteorolo ical Variables, Journal of Met., 1953.

6. Michigan Emergency Management Division, Michi an Hazard Anal sis, September 1992.

2.2-10 July, 1993

2.7 IRONME DIATION MONITORING The radiological environmental radiation monitoring program, described herein, was designed to evaluate the effects that routine and inadvertent radioactive releases from the plant have on the environment. Provisions are made to monitor liquid and gaseous wastes before and/or during their release, and further to monitor the atmosphere, lake water, well water, aquatic organisms, milk and food materials when necessary. Liquid and gaseous wastes are released as continuous releases as well as batch releases from time to time, after appropriate decay, processing and analysis. No foreseeable environmental conditions will restrict the release of wastes.

However, if extreme conditions should indicate the desirability, wastes can be retained until dispersion conditions improve.

2.7 1 DETERMINATION OF MAXIMUM ALLOWABLE RELEASE RATES TO AIR AND WATER The first step in the program is to determine the maximum rate at which radioactive material may be continuously discharged without exceeding the limits set by 10 CFR 20 at the site boundary. The initial estimate is based on the plant design, the anticipated composition of the radioactive material to be released, and the dilution and dispersion characteristics of the air and water into which these materials are discharged.

The unit vent, through which radioactive gaseous waste is routinely released, runs up the outside wall of the containment building. The vent opening is at the top of the containment building, about 160 feet above grade.

Other release pathways for discharge of radioactive gaseous waste include the steam jet air ejectors, gland seal leak-offs, steam generator blowdown, and the auxiliary boiler stack.

2.7-1 July, 1993

Table 2.2-11 shows the predominant wind to be coming from the South. The values calculated for X/Q for all of the wind directions are given in Table 2.2-3. Using the table, the X/Q for winds blowing toward the North (from the South) is calculated to be 9.54E-06 seconds/m . For the waste gas mixture in Table 11.1-6, the weighted average maximum permissable concentration is 1.2E-07 pCi/cc. The maximum rate at which this mixture can be discharged continuously is Q ~ .013 Ci/sec, without exceeding the 10 CFR 20 limit at the site boundary.

Zf suitable for discharge, liquid radioactive wastes are released to the condenser circulating water system. The maximum discharge concentration for licpxids is defined in the Plant Technical Specifications. This defined concentration ensures that the llmi.ts established in 10 CFR 20 are not violated.

2~7 2 July, 1993

2 '.2 SAMPLING STATIONS The stations for sampling airborne particulates, volatiles, and external radiation are placed in two rings about the plant. The inner, or indicator ring, stations are placed where it is estimated that maximum ground concentrations of material released from the plant will occur. Figure 2.7-1 indicates the locations selected for the six indicator stations (shown as Al through'6).

Figure 2.7-3 shows the locations which have been selected for the four background air stations in the outer ring as identified as A. These locations are all about 20 miles from the plant and thus the ground-level concentrations of radioactive material originating from the plant will be less than 1 percent of the concentrations at the indicator stations.

Locations of TLD stations are shown in Figures 2.7-1, 2.7-3 and 2.7-4.

Twelve on-site indicator TLD stations (shown as Al through A12 on Figure 2.7>>1) are located on an approximate 2000 foot radius and eleven off-site monitoring TLD stations are within a 2 to 5 mile radius from the plant (shown as T1 through Tll on Figure 2.7-4). Four background TLD stations located about 20 miles from the plant are identified as A on Figure 2.7-3.

Sam lin Lake Water The locations of the sampling stations for lake water are descr'ibed in Table 2.7-1. Indicator lake samples are taken along the lake front from the condenser cooling water intake and at an approximate distance of 0.1 and 0.2 miles north and 0.1 and 0.3 miles south of the plant centerline.

The sampling of aquatic organisms presents a number of difficulties. Out to a depth of 20 feet or more, the lake bottom is scoured sand and is almost sterile. Attempts to find suitable organisms in sufficient quantities for routine sampling have been unsuccessful.

207-3 July, 1992

I Benthonic organisms occur only at depths greater than twenty feet; such depths occur at 1,000 feet or more from shore. Routine sampling under such circumstances is impractical for extended periods of unfavorable weather.

Fish are collected and analyzed in the program, but fish are a poor sampling medium because they range so widely that it is never certain that they represent the area where they happen to have been caught.

Sam lin of Well Water Well water is the only material in the environmental sampling program that is not likely to be affected by fallout of radioactivity. With well water, and only with well water, is the before and after principle sound. There are 17 wells (13 REMP wells and 4 non technical specification steam generator storage facility groundwater monitoring wells) within the owner controlled area as shown in Tables 2.7-2 and 2.7-3. The orientation of these wells with respect to the plant was chosen as a result of groundwater movement, which was found to be east to west.

Sam lin of Milk The selection of milk sampling locations are, of course, limited to pastures where milk cows graze. The locations shown in Figure 2.7-3 are sub)ect, to change as the location of milk cows change.

2.7-4 July 1992

Sam lin of Food It is now evident that milk alone provides sufficient control of terrestrial pathways. Additional human food materials are not needed in the program unless radioactive materials other than noble gases, tritium and iodine are detected in the plant discharges to the atmosphere.

Because radioactive particulates have been noted in gaseous discharges additional human food crops will be sampled annually.

The noble gases do not enter directly into the food chains. Tritium enters freely into all food chains; however, since almost all tritium occurs as tritiated water, it does not concentrate in food pathways as do other elements. Iodine does concentrate along food pathways and it has been shown that the air-pasture-milk animal-milk pathway is critical and that milk is the best monitoring medium. Radioactive particulates which settle out on the surfaces of crops are adequately monitored by the sampling and analysis of broadleaf vegetation and grapes. There is, consequently, neither need nor justification for monitoring human foods other than milk and selected vegetation in the terrestial environment, and fish in the aquatic environment.

All sampling points have been selected because they are representative of the area and accessible for sampling. Table 2.7-5 describes the current Radiological Environmental Monitoring Program, as defined in the plant Technical Specifications.

2.7.3 STABLE ELEMENT STUDIES The pre-operational phase of the environmental program includes a study of stable element concentrations in the lake water and in selected aquatic organisms. The purposes of these measurements are (1) to put an upper limit on the degree to which radioactive material discharged from the plant into the lake could be concentrated in human food taken from the lake, (2) to find critical pathways and the means for estimating population exposure by these pathways, and (3) to determine 2.7-5 July 1989

the relationship between the concentration factors in fish (and any other human foods taken from the lake) to those in aquatic organisms selected to monitor the water environment.

The principle involved in these stable element studies is that the radioactive isotopes of an element cannot be concentrated more highly than the corresponding stable isotopes of that element by biological, chemical or physical processes in the environment. The general form of these studies is described in the next paragraph.

The radioactive isotopes anticipated in the liquid waste (Table 11.1-5) are examined, as are the data on similar operating reactors. From these one obtains a list of the elements which correspond to all the radioactive isotopes which may contribute to radioactivity in food chains. Samples of lake water, edible portions of fish, and other possible monitoring organisms, if available, are collected and analyzed for each of the elements in the list. The data so obtained give concentration factors from water to fish, and from water to monitor organisms for the stable elements. Radioactive isotopes of these elements cannot be concentrated to factors greater than those for the corresponding stable elements.

2.7.4 MEASUREMENT OF RADIOACTIVITY The pre-operational phase of the environmental program included the collection and analysis of samples for radioactivity; the intensity of the post-operational phase is concerned exclusively with radioactivity released from the plant. This section describes the equipment and techniques that are used to collect and analyze environmental samples for radioactivity.

Direct radiation doses primarily due to radioactive noble 'gases in the environment is measured with thermoluminescent dosimeters. The detection limit of thermoluminescent dosimeters is 1 to 2 mR per month. This 2.7-6 July, 1988

sensitivity corresponds to 2 to 4 percent of the maximum permissible dose to the public from radioactive noble gases.

The air sampling units draw about 6 x 10 8 cc of air per week through a filter. The detection limit of a lithium drifted germanium gamma spectrometer is on the order of 10 -9 Ci/cc, far below the maximum permissible public concentration of any radioactive material the plant could discharge to the atmosphere.

The environmental air sampling units are fitted with charcoal cartridges to collect iodine. If the air passing through this cartridge were at the public maximum concentration of iodine 131 for the entire week, the cartridge would collect 0.06 pCi.

Tritium is measured in a liquid scintillation counter with a nominal sensitivity limit of 10 pCi/cc, which is 0.03 of the permissible drinking water concentration. Analyses will be made by contract with an outside laboratory.

2.7.5 OPERATION OP THE PROGRAM The environmental radiation monitoring program was started some 12 to 18 months before fuel was loaded in Unit 1. During this period, equipment was tested, the suitability of the selected sampling media and sample points were determined, analytical procedures were tested, and some data was accumulated and examined for statistical variability. Modifications that were necessary to attain reliable and coherent data were made during this period.

Prior to any liquid or gaseous release, the concentration of radioactive material that is to be released to the environment is determined. Baaed on the total concentration of radioactive material present, the release flow 2' 7 July, 1993

rate is ad)usted to ensure that the Technical Specification release limits are not exceeded. Zn addition the actual dose to a member of the public resulting from liquid and gaseous releases from the Cook Nuclear Plant is determined to ensure that. Technical Specification requirements are not being exceeded.

2.7 6 SUMHARY OF PREOPERATIONAL ENVIRONHENTAL MONITORING PROGRAH The average monthly LiF Dosimeter Loadings for the quarter of August, 1971 through December, 1971, on site, vary from 3.9 + 1.3 to 11.7 + 0.8 mrem and offsite 3.9 + 1.2 to 13.3 + 1.1 mrem.

Zniti.al water samples taken in the Lake (similar to that shown in Figure 2.7<<

1) and at Bridgman, St. Joseph, Benton Harbor and New Buffalo show a tritium concentration of from 0.562 + 0.036 to 0.583 + .036 picocuries/

milliliter. Gross beta at the above sampling points showed 0.0 + 2.0  !

picocuries/liter to 6.8 + 1.0 picocuries/liter.

The determination of gross beta in the air particulates on site is 0.01 +

0.01 to 0.30 + 0.01 picocuries/cubic meter. The same values for offsite stations are 0.01 + 0.01 to 0.24 + 0.1 picocuries/cubic meter.

2 7-8 July 1989

TABLE 2.7-1 LOCATIONS OF THE WATERBORNE SURFACE SAMPLING STATIONS Indicator Stations Condenser cooling water intake (Ll).

0.3 miles southwest from plant centerline along the lake shore (L2).

0.2 miles northeast from plant centerline along the lake shore (L3).

0.1 miles southwest from plant centerline along the lake shore (L4).

0.1 miles northeast from plant centerline along the lake shore (L5).

(See Figure 2.7-1)

Back round Stations and drinkin water sam le stations Lake Township water intake, 0.4 miles south from the plant (D ).*

St. Joseph municipal water intake, 9 miles northeast from the plant (D ).+

(See Figure 2.7-3)

• D and D A B refer to analysis performed as indicated in Table 2.7-4.

2.7-9 July, 1992

TABLE 2.7-2 WELLS AVAILABLE FROM MONITORING PROGRAM (Refer to Figure 2.7-1 for a map indi.cath.ng the location of these sample points)

Approximate Distance from Direction from Well No. Plant in Feet North Wl 1969 ll W2 2292 63 W3 3279 107 W4 418 301 W5 404 290 W6 424 273 W7 1895 189 W8 1279 58 W9 1447 22 W10 4216 129 Wil 3206 153 W12 2631 162 W13 2152 182 2.7-10 July, 1993

TABLE 2.7-3 (Refer to Figure 2.7-2 for a map indicating the location of these sample points)

Approximate Distance from Direction from SGR-1 4037 95 SGR-2* 3879 92 SGR-4 3699 93 SGR-5 3649 92 These wells are sampled and analyzed quarterly for:

- Gross alpha Activity

- Gross Beta Activity

- Gamma Xsotopic Activity

• No SGR-3 well defined for this'rogram.

July 1991

Table 2.7-4 intentionally deleted.

2.7-12 July, 1992

Page 1 of 4 TABLE 2.7-5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Type & Frequency Sam le Locations Collection Fre enc of Anal sis

1. Airborne Al-A6 (Site) Continuous operation of sampler Radioiodine canister

a. Radioiodine and New Buffalo, South with Sample Collection as Analyzes Weekly for Particulates Bend, Dowagiac, and required by Dust Loading But at I-131 Coloma are Background Least Once Per 7 Days Particulate sample Gross Beta Radio-activity follgwing Filter Change ,

composite (by loca-tion) for gamma isotopic quarterly.

2. Direct Radiation a) Al-A12 (On-Site) At least once per 92 Days Gamma Dose. At Least b) New Buffalo, South (Quarterly) Once Per 92 Days.

Bend, Dowagiac, Coloma c) 11 Off-Site TLD Monitor Locations

3. Waterborne

a. Surface Llg L2g L3g L4g L5 Composite* Sample Over One- Gamma Isotopic Month Period Analysis monthly.

Composite for tritium analysis-quarterly.

• Composite samples shall be collected by collecting an aliquot at intervals not exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

a Particulate sample filters should be analyzed for gross beta 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control samples for any medium, gamma isotopic analysis should be performed on the individual samples.

2 ~ 7 13 July, 1993

Page 2 of 4 TABLE 2.7-5 (Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Type 6 Frequency Sam le Locations Collection Fre enc of Anal sis

b. Ground WI"W13 Quarterly Gamma Isotopic and Tritium analysis quarterly.

c. Drinking St. Joseph Composite* Sample Gross Beta and Gamma Lake Township Collected over a Period of Isotopic Analysis of less than or equal to 31 days each composite sample.

Composite* Sample Over a 2-week Tritium Analysis of Period if I-131 Analysis Performed.

is composite Quarterly.

I-131 analysis on each composite when the dose calculated for the consumption of the water is greater than 1 mrem per year.

d. Sediment from L2, L3~ L4, L5 2/year Gamma Isotopic Shoreline Analysis Semi.-Annually.

4. Ingestion Indicator At Least Once Per 15 Days When Gamma Isotopic and

a. Milk Farms~background Animals are on Pasture. At I-131 Analysis of Farms Least Once Per 31 Days at Other Each Sample.

Times.

• Composite samples shall be collected by collecting an aliquot at intervals not exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• • An indicator farm is defined as the nearest milk producer in each of the land sectors within 8 miles of the plant site who is willing to participate in the radiological environmental monitoring program.

A background farm is defined as a milk producer in one of the less prevalent wind directions at a distance greater than 15 miles but less than 25 miles who is willing to participate in the radiological environmental monitoring program. If at least three indicator milk samples and one background milk sample cannot be obtained, vegetation sampling will be performed as a replacement for the milk sampling and no milk samples will be required.

2.7-14 July, 199:-

I'i'IIIII: ' / I fall

~0NRPSTRICTED AREA

. TIDAL)k. Pl ANT NOH Tll NOR PRPWRTY LINF uk'5

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=-Mj5 YWn l l L I

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==L2 765 kV l VaRn N-SHOAF LlNE PL NT AIO i~-'I A9 A6 ItN

),000 FOOT;:;:/ AOt(jS

.'.r WIO,,:-"'GII j.:.,W12p A Air and TLD Sample Stations 4 (A7 AI2 are TI.D Sample Stations Only)

W Well Water Sample Stations 0 lOOO 2OOQ OOOO )OOO $ P f. f L i.ako w tot and nodimont sampling stations scat E (l.l is lake Water Sampling Stations Only) [

STEAM GENERATOR STORAGE FACILITY NON-TECHNICAL SPEClFlCATION GROUNDWATER MONlTORlNG WELLS FIGURE 2.7-2 SGR-5 SGR-2 STEAM GENERATOR O o~ STORAGE FACILITY 765KV YARD eSGR-4 SGR- > W MONITORING WELLS MONITORING WELL MSGR-3 WAS NC BILLED

In the dynamic piping analyses, vertical seismic spectra equal to 2/3 of the pertinent building base horizontal spectra was computer input with the appropriate building floor horizontal seismic spectra. The effects

'of each seismic spectra input were computed independently and the various modal results were computer combined by the square root of the sum of the squares (SRSS) method. The effects of the vertical and a horizontal seismic run were then computer combined by the SRSS method.

The larger resultant value of the vertical and horizontal seismic run

[(Y + X), or (Y + Z)] at each node was considered to be the critical load and/or stress.

Class I piping smaller than 2h inch nominal diameter with operating 0

temperatures less than 250 F, may be qualified by using either. a simplified analysis (Alternate Analysis) method or a computer dynamic analysis. The Alternate Analysis method developed for the Cook Nuclear Plant considered gravity loads, seismic loads (based on floor acceleration response spectra) and internal pressure loads. The acceptance criteria were based on pipe stress and pipe displacement. A set of instructions, guidelines, tables and graphs reflecting the above, were issued to establish acceptable spacing of supports.

Class II piping with operating temperatures less than 250 F may be qualified by using this Alternate Analysis method. Class II piping with 0

operating temperatures greater than 250 F are qualified using the computer dynamic analysis method. The seismic inputs are taken from the appropriate OBE spectra.

@here a piping system consists of a combination of Class I and/or Class II, and/or Class III piping, the method of analysis is for the higher class piping. The piping model maybe structurally decoupled, to suit the higher class piping, at an anchor or at a point (or points) encompassing restraints in the 3 orthogonal directions.

2.9-7 July 1991

2.9.4 SEISMIC DESIGN CRITERIA FOR CLASS I, CLASS II AND CLASS III STRUCTURES Class I A dynamic analysis was performed using Response Spectrum and Modal Analysis Procedure, as discussed in Appendix F of the original FSAR, "Dynamic Analysis of the Containment Structure for Seismic Loading."

Response spectra were generated from information obtained by a full seismological study of the site. Stress criteria are those of ACI 318-63 Ultimate Strength Design, C ass I An analysis using the procedures of the Uniform Building Code (International Conference of Building Officials) was made. Standard working stresses are used.

i Values of maximum ground acceleration are those used for Class I criteria. The factor applied to the seismic forces from which the values of shear, bending moments, etc. are computed, is taken as that for Zone 3 of the Uniform Building Code multiplied by the ratio of the maximum ground acceleration to a value of 0.30g. The minimum ratio used is one-fourth.

C ass III An analysis using the procedures of the Uniform Building Code (International Conference of Buiilding Officials) was made. Standard working stresses increased by 33 percent are used. Zonal factors of the Uniform Building Code are used.

2.9-8 July, 1982

or Al St cture Se s ic Classi cat ons A vertical component of earthquake acceleration of two-thirds the value of the horizontal component of earthquake is assumed to be acting simultaneously with the horizontal component.

Seismic design criteria for combined structures (i.e., structures having Class I and Class II elements, Class I and Class III elements or Class II and Class III elements) are as follows:

1. Equipment is supported by structural elements equal to or higher than the classification of the equipment.

2. Equipment is surrounded by structural elements equal to or higher than the classification of the equipment.

3. Structural elements are supported by, or framed to, elements equal to or higher than its own classification.

The following example illustrates the design criteria stated above.

The auxiliary feed pumps are Class I equipment but are housed in the turbine building which is essentially a Class III structure. In this case, the Class I equipment is anchored directly to the foundation slab which is designed to Class I criteria. The pumps are surrounded by local structural elements designed to Class I criteria which have been designed to withstand potentially adverse effects of lower class structures in the area.

The superstructure for the turbine room, heater bay and main steam pipe enclosure beyond the steam generator stop valve are Class III structures, which are designed for seismic loading in accordance with the seismic criteria of the Uniform Building Code. The maximum deflection for all conditions of loading were computed for these structures.

2.9-9 July, 1982

These deflections plus an allowance for erection and fabrication tolerances and an additional amount for clearance were designed into these structures to prevent rattling (hammering) effect.

The primary water and condensate tanks are functionally Seismic Class II structures located near Seismic Class I structures, namely, the refueling water storage tank and the containment. The condensate and refueling water storage tanks have been seismically analyzed to insure their structural integrity during a seismic event. The primary water tank was analyzed seismically for the OBE. All three tanks are located in excess of 20 feet from the containment wall. The primary water storage tank is approximately 55 feet from the refueling water storage tank. Analysis indicates that the primary water storage tank will not cause structural damage to the refueling water or condensate storage tanks in the unlikely event that it fails. The condensate storage tank, although not required to be a Seismic Class I structure, was designed as such to insure the structural integrity of the refueling water tank.

2.9.5 GENERAL DESIGN CONSIDERATIONS FOR BUILDING STRUCTURES Those structures considered are the auxiliary, containment, circulating water pump screen house and turbine buildings, and the steam generator stop valves and pipe enclosures outside the containment building.

Building structures were designed to withstand wind forces.

Class I building structures were evaluated with reference to tornado conditions to assure that there would be no loss of function.

The wind velocities and tornado model are discussed in Chapter 5 and Subchapters 1.4 and 2.8.

Tornado loading was not considered coincident with earthquake loading.

However, a 3 psi ambient pressure drop was considered coincident with tornado velocity pressures.

2.9-10 July, 1993

pressure and suction forces together with internal pressure or suction was considered in accordance with the procedure in ASCE Paper No. 3269 "Wind Forces on Structures."

Torsional effects due to tornado loading were considered in evaluating Class I structures.

Maximum torsional loading was determined by using varying diameter tornado "funnels."

Reinforcing was placed so that minimum reinforcing cover provisions are as reccmmended by the Uniform Building Code and ACZ Building Code.

l. 3 in. cover where 'concrete was deposited against the ground (bottom of slab) .

2. 2 in. cover at all formed surfaces exposed to the ground or weather (all exterior surfaces of the structure).

3. 14 in. cover for beams and girders not exposed to the ground or the weather.

4. 1 in. cover for slabs and walls not exposed to the ground or weather.

5. Concrete protection for reinforcement is in all cases at least equal to the diameter of the bars.

Building structures were designed in accordance with the seismic design criteria as stated in Section 2.9.4.

July, 1982

The effects of differential motion considered ~

between the v 'u' various This was necessary both to provide ad equate xldings were t separation between the structures to prevent "banging together" of the structures during a seismic occurrence and to provide for this condition on interconnecting elements.

Both the horizontal and rotational motions of the con containm t t ainmen t structure due to earthquake were analyzed. A plot of displacement vs height was made.

The magnitude of maximum vertical motion due to the DBE was detexmined for the structures, considering each structure as a rigid body.

The maximum magnitude of differential motion was considered to be the absolute value of the peaks of motion between the independent structures, considering each motion to occur. simultaneously with the others.

The effect of static differential settlement was considered additive to the dynamic effects where this resulted in a more severe condition.

A discussion of the design for the auxiliary and turbine building follows.

The design of the containment building is discussed in Chapter 5.

Auxilia Buildin The Auxiliary Building encloses the fuel storage areas (both. new and used fuel), the fuel transfer canal, the containment equipment hatches access areas, control facilities and other equipment.

Seismic considerations for the Auxiliary building were based on the 10%

and 20% Ground Response Curves as indicated in Figures 2.5-2 and 2.5-3.

A dynamic analysis of the building was performed to determine the seismic stresses in Class I portions of the structure. Using a slab-spring model 2.9-12 July, 1982

subjected to independent translational excitation in two perpendicular directions, the modal periods, the forces acting on the slabs, the slab displacement and the loads on major lateral load resisting elements were computed. Consideration was also given to the action of water in the spent fuel pool during a seismic occurrence.

The superstructure is a Class I structure consisting of a structural steel skelton with exterior walls and roof of reinforced concrete.

The structural steel was designed in accordance with the "Specification for the Design, Fabrication and Erection of Structural Steel for Build-ings," adopted April 17, 1963, by the American Institute of Steel Construction.

The roof of the structure is constructed of steel beams and girders supporting a poured concrete roof on steel ribbed decking. The roof varies in thickness, stepped from two feet to seven inches. The thickest portion of the roof is directly over the spent fuel pool area.

The walls are of poured concrete supported, for their vertical load, on the concrete substructure and for their lateral forces by the structural steel columns and struts. The walls vary in thickness from two feet to six inches. The thickest area is the west wall adjacent to the fuel pool and the thinnest portion is at the east end of the structure.

The concrete walls and roof of the auxiliary building were designed to provide protection against potential missiles. The whole structure was designed to withstand the design basis tornado missiles and was also designed to protect the control room and fuel pool against a turbine missile. See Sub-Chapter 1.4 for a discussion on missile protection.

2.9-13 July, 1982

The tornado forces applied to the structure are as outlined in Chapter 5 with the exception that the diameter of the tornado was assumed to vary in the following manner:

a. The diameter is equal to the width of the structure.

b. The diameter is equal to the length of the structure.

c. The diameter is infinite in extent'n the event of a tornado, the pressure within the structure will not differ from the 'outside by more than 1/2 psi in three seconds. This low differential is achieved by the installation of vents in the periphery of the roof which will allow release of internal pressure. However, as an added conservatism, the building roof and walls have been designed to withstand 3/4 psi coincident with tornado wind forces. For forces resulting from tornado winds of 250 mph tangentially and a progression velocity of 50 mph, the auxiliary building steel will not experience stresses in excess of allowable as outlined in the 1963 American Institute of Steel Construction specifications. For tornado winds of 300 mph, tangentially with a progression of 60 mph, coincident with internal pressures of 3/4 psi, steel will remain within yield and no permanent deformation will result.

The auxiliary building, as a Class I structure, has been designed for seismic forces as described in this chapter. A dynamic analysis was made for the OBE and DBE. For the OBE, all stresses in the steel superstructure are within allowables as specified by the 1963 code of the "American Institute of Steel Construction for Buildings." For the DBE, the superstructure steel stresses do not exceed yield and no permanent deformations will result.

For the 1988 Steam Generator Replacement Project, the following changes were made to the auxiliary building. An additional 150 ton single failure proof crane was installed and the existing 150 ton crane was upgraded to a single failure proof design. The building was reanalyzed for the following conditions:

2.9-14 July 1990

a. The two cranes acting in tandem to move steam generator components'during the replacement project,

b. The seismic forces, as described in this chapter, resulting from a single crane with a 60 ton live load acting anywhere in the building.

For the OBE, all stresses in the steel superstructure are within allowable as specified by the code of the "American Institute of Steel Construction for Buildings." Adopted November 1, 1978. For the DBE, the superstructure steel stresses do not exceed yield and no permanent deformation will result.

Turbine Buildin A structure such as the turbine building, which consists of a Class I foundation and a superstructure which is Class I in some areas and Class III in other areas, was designed as follows. The superstructure was designed in accordance with the criteria discussed in Section 2.9.4. The reactions at the base of the superstructure were used as input for the foundation design. The foundation was analyzed for lateral earthquake and a simultaneously acting vertical component, considering the effects on the foundations of the superstructure and any equipment supported directly on the foundation.

For seismic or tornado conditions, the mat was designed in accordance with the stress criteria of ACI Code 318-63 "Ultimate Strength Design". The load equations used were those of Subsection 5.2.2.2, with the elimination of the pressure and temperature items.

For normal load conditions, the mat was designed using Working Stress Design. Stresses and strains for normal loading were held to the limits of ACI Code 318-63 "Working Stress Design."

In the design of Class I structures by ACI Code 318-63 "Ultimate Strength Design" procedure, load reduction factors (P) used for the containment are discussed in Subsection 5.2.2.2. However for structures other than the 2.9-15 July 1990

containment structure and when considering seismic conditions, the load reduction factor for diagonal tension, bond and anchorage in concrete was reduced to 0.75.

2.9.6 SEISMIC DESIGN CRITERIA FOR EQUIPMENT Seismic Class I equipment design generally requires that normal plus DBE stresses do not exceed yield, and rotating or sliding equipment functions do not bind.

The combination of earthquake plus normal stresses for the OBE condition shall not exceed normal allowable, as defined by applicable code. Refer to Table 2.9-1, and Notes thereto, for the definition of loading conditions.

Restraints for both Class I mechanical and electrical equipment were generally designed to accept combined normal plus DBE loading without exceeding 0.9 of the yield stresses.

Class I equipment was designed for earthquake loads represented by the combina-tion of appropriate horizontal and vertical floor responses simultaneously applied. The vertical response was equal to 2/3 of the horizontal response.

N Depending on the relative structural complexity and relative rigidity of the equipment to be evaluated, one of the following methods of seismic qualification was performed:

1. For structurally complex equipment, a dynamic multi;,degree-of-freedom modal analysis which considered frequency, mode shape and modal participation factors in determining seismic response.

2. For structurally simple equipment, a dynamic single degree-of-V freedom analysis which considered fundamental frequency response of the equipment as determined from the floor response spectrum.

3. A simplified dynamic analysis which utilized the peak of the floor response spectrum to determine seismic loading.

4. Testing of identical or similar components using approved procedures to simulate appropriate seismic loads.

2 '-16 July 1990

(Sheet 1 of 3)

TABLE 2.9-1 LOADING CONDITIONS CONDITIONS* EFFECTS CONSIDERED

1. NORMAL Deadweight, Thermal, Pressure (Pressure is considered for vessel and pipe stress only)

2. UPSET Same as 1, OBE

3. EMERGENCY Same as 1, DBE

'(Thermal is considered for supports loads only)

4. FAULTED Same as 1, Postulated Pipe Rupture (Thermal is considered for supports loads only)

5. FAULTED (Including DBE) Same as 3, Postulated Pipe Rupture

• See Note 1.

2.9-17 July 1989

(Sheet 2 of 3)

TABLE 2.9-1 NOTES NOTE 1: Definition of Terms based on the Summer 1968 Addenda to the ASME Boiler and Pressure Vessel Code,Section III.

The Operating Load Combination categories are defined as follows:

(1) Normal Condition - Any condition in the course of system startup, operation in the design power range and system shutdown, in the absence of Upset, Emergency or Faulted Conditions.

(2) U set Condition - Any deviations from Normal Conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational impairment. The Upset Condition includes those transients caused by a fault in a system component requiring its iso-lation from the system, transients due to a loss of load or power and any system upset not result'ing in a forced outage.

The Upset Conditions include the effect of the specified earthquake for which the system must remain operational or must regain its operational status.

(3) Emer enc Condition - Any deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system. The conditions have a low probability of occurrence but are included to provide assurance that no gross loss or structural integrity will result as a concomitant effect of any damage developed in the system. The total number of postulated occurrences for such events shall not exceed twenty-five (25). Among the Emergency Conditions may be a specified earthquake for which safe shutdown is required.

2.9-18 July 1989

(Sheet 3 of 3)

TABLE 2.9-1 (4) Faulted Condition - Those combinations of conditions associated with extremely low probability postulated events whose con-sequences are such that the integrity and operability of the nuclear energy system may be impaired to the extent where considerations of public health and safety are involved. Such considerations require compliance with safety criteria as may be specified by jurisdictional authorities.

2.9-19 July 1989

(Sheet 1 of 5)

TABLE 2.9-2 (Part A)

LOADING CONDITIONS AND STRESS LIMITS'RESSURE VESSELS LOADING CONDITIONS STRESS INTENSITY LIMITS NOTE

1. Normal Conditions (a) P < S (b) P (or PL) + PB < 1.5S 1 (c) P (or PL) + PB + Q < 3.0S 2 2, Upset Condition (a) P m

Sm (b) P (or PL) + PB < 1.5S 1

'or (c) P P ) + P + Q < 3.0S

' 2 m L B

3. Emergency Condition (a) P m

1.2S or S whichever is larger (b) + < 1.5(1.2S

' )

m L B or 1.'5S whichever is larger 3

4. Faulted Condition See Note 4 P - primary general membrane stress intensity PL primary local membrane stress intensity PB

- primary bending stress intensity Q - secondary stress intensity S

m

- stress intensity value from ASME B&PV Code,Section III, Nuclear Vessels - 1968 Edition, Table N-421 S

y minimum specified material yield (ASME B&PV Code, Section III, Nuclear Vessels - 1968 Edition, Table N-424 2.9-20 July 1989

(Sheet 2 of 5)

TABLE 2.9-2 (Part B)

LOADING CONDITIONS AND STRESS LIMITS'RESSURE PIPING LOADING CONDITIONS STRESS LIMITS 1, Normal Conditions (a) P < Sh (b) P + P < Sh (c) SA (1'25 C

+ 0'25Sh) f 2 ~ Upset Conditions (a) P < 1.2Sh (b) PL + PB < 1.2S (c)' < (1,25SC + 0.25Sh)f

3. Emergency Conditions (a) P < 1.2Sh (b) PL + B < 1.5(1.2Sh

4. Faulted Conditions See Note 4.

where:

P m

- primary hoop membrane stress (pressure)

PL - primary longitudinal membrane stress (pressure)

P B

primary longitudinal bending stress (deadweight, seismic)

Sh allowable stress at temperature from USAS B31.1 Code for Pressure Piping, 1967 Edition S

C allowable stress at 70 0 F from USAS B31.1 Code for Pressure Piping, 1967 Edition A

- allowable stress range for expansion stresses (fatigue criteria) f stress range reduction factor for cycling per USAS B 31.1 Code for pressure piping, 1967 Edition 2.9-21 July 1989

(Sheet 3 of 5)

TABLE 2.9-2 (Part C)

LOADING CONDITIONS AND STRESS LIMITS: E UIPMENT SUPPORTS LOADING CONDITIONS STRESS INTENSITY LIMITS

1. Normal Condition Working Stresses or Applicable Factored Load Design Values

2. Upset Condition Working Stresses or Applicable Factored Load Design Values

3. Emergency Condition Within yield after load redistribution

4. Faulted Condition Permanent Deflection of Supports Limited to Maintain Supported Equipment Within Design Limits.

See Note 4 Support loads are combined by algebraic summation, in plus and minus directions of the three orthogonal planes, so as to obtain the maximum positive or maximum negative value of design load.

The thermal load component is not considered when algebraic summation with this load would lessen the support design load.

The seismic load component is considered to have both a positive and a negative sign. The sign of the seismic component is chosen so as to maximize the absolute value of the support design load.

2.9-22 July 1989

(Sheet 4 of 5)

TABLE 2.9-2 NOTES Note 1: The ty limits on local membrane stress intensity (PLL

< 1.5S

' )

and primary membrane plus primary bending stress intensity (P m (or PL) + PB < 1.5S m ) need not be satisfied at a specific location if it can be shown by means of limit analysis or by tests that the specified loadings do not exceed 2/3 or the lower bound collapse load as per paragraph N417.6(b) of the ASME B6PV Code,Section III, Nuclear Vessels - 1968 Edition.

Note 2: In lieu of satisfying the specific requirements for the I

local membrane (P < 1.5S ) or the primary plus secondary stress intensity (P + P + Q < 3S ) at a specific location, the structural action may be calculated on a plastic basis and the design will be considered to be acceptable if shake-down occurs, as opposed to continuing deformation, and if the deformations which occur prior to shakedown do not exceed specified limits, as per paragraph N417.6(a) (2) of the ASME B6PV Code,Section III, Nuclear Vessels - 1968 Edition.

Note 3: The limits on local membrane stress intensity (P < 1.5S )

and primary membrane plus primary bending stress intensity (P m (or PL) + PB < 1.5S ) need not be satisfied at a specific location if it can be shown by means of limit analysis or by tests that the specified loadings do not exceed 120 percent of 2/3 of the lower bound collapse load as per paragraph N417.10(c)'f the ASME B&PV Code,Section III, Nuclear Vessels - 1968 Edition.

2.9-23 July 1989

(Sheet 5 of 5)

TABLE 2.9-2 Cont'd NOTES i

,"= ".. 5' plastic instability analysi's may be. performed for cases considering the actual strain-hardening

'pecific characteristics of the material, but with 'yield'trentgh adjusted to correspond to the tabulated value't , sL the h appropriate temperature in Table N-424 or'N-'425,'as per paragraph N-417.11(c) of the ASME B&PV, Code,'ection III, Nuclear Vessel - 1968 Edition. '

• ~

,e

~ ~

2.9-24 July 1989

3.0 REACTOR 3.1

SUMMARY

DESCRIPTION The Cycle 13 reactor core contains three regions of fuel in a low leakage loading pattern as described in Section 3.5.2. The fuel rods are cold worked, partially annealed Zircaloy tubes containing slightly enriched uranium dioxide fuel.

All fuel rods are pressurized with helium during fabrication to reduce stresses and strains and to increase fatigue life.

The fuel assembly is a canless type with the basic assembly consisting of the RCC guide thimbles fastened to the grids, and to the top and bottom nozzles. The fuel rods are supported at several points along their length by the spring-clip grids.

Full length rod cluster control assemblies are inserted into the guide thimbles of the fuel assemblies. The absorber sections of the control rods are fabricated of silver-indium-cadmium alloy sealed in stainless steel tubes.

The control rod drive mechanisms for the full length RCC assemblies are of the magnetic latch type. The latches are controlled by three magnetic coils. They are so designed that upon a loss of power to the coils, the rod cluster control assembly is released and falls by gravity to shut down the reactor.

The reactor was initially supplied with fuel from Westinghouse Electric Corp. (W). Reload fuel for Cycles 2 through 7 was supplied by Exxon Nuclear Co (ENC). Cycles 8 through 13 reload fuel was supplied by Westinghouse Electric Corp. The latest information regarding the current fuel cycle may be found in Sub-Chapter 3.5.

UNIT 1 3.1-1 - July, 1993

In addition to this summary description, this chapter contains: a description of the mechanical components of the reactor and reactor core, including Cycle 1 W fuel assemblies, reactor internals and control rod mechanisms (Sub-Chapter 3.2); a description of the Cycle 1 nuclear design for the W fuel (Sub-Chapter 3.3); a description of the Cycle 1 thermohydraulic design (Sub-Chapter 3.4); and a description of the current core design (Sub-Chapter 3.5).

The information contained in this chapter is principally concerned with the nuclear fuel and reactor internals design and therefore does not necessarily reflect the same information as that used in the safety analysis. For information concerning safety analysis, Chapter 14 should be consulted.

3.1.1 Performance Objectives The current licensed thermal power limit is 3250 MWt. Calculations indicate that hot channel factors are considerably less than those used for design purposes in this application. The thermal and hydraulic design, and accident analyses (except large break LOCA) in Chapter 14, were performed at 3411 MWt for Cycle 8. These analyses identify design/safety limits for a potential uprating.

The turbine-geherator and plant heat removal systems have been designed for a thermal rating of 3391 MWt. The portions of the safety analysis dependent on heat removal capacity of plant and safeguards systems have assumed the maximum calculated power rating of 3391 MWt, as have the evaluations of activity release and radiation exposure.

The initial reactor core fuel loading was designed to yield the first cycle nominal burnup of 16,666 MWD/MTU, and the Cycle 2 through 7 reload designs yield an nominal cycle burnup of 10,000 MWD/MTU. Reload designs UNIT 1 3. 1-2 July, 1992

for Cycles 8 through 13, yield nominal cycle burnups of between 15,000 and 16,000 MWD/MTU. The fuel rod cladding is designed to maintain its integrity for the anticipated core life. The effects of gas release, fuel dimensi.onal changes, and corrosion-induced or irradiation-induced changes in the mechanical properties of cladding are considered in the design of the fuel assemblies.

Rod control clusters are'mployed to provide sufficient reactivity control to terminate any credible power transient prior to reaching the applicable design minimum departure from nucleate boiling (DNB) ratio (see Section 3.5.3). This is accomplished for the current cycle by ensuring sufficient control cluster worth to shut the reactor down by at least 1.6% in the hot condition with the most reactive control cluster stuck in the fully withdrawn position.

Redundant equipment is provided to add soluble poison to the reactor coolant in the form of boric acid to maintain shutdown margin when the reactor is cooled to ambi.ent temperatures.

Zn addition, the control rod worth in con)unction with the boric acid in)ection from the boric acid in)ection tank is sufficient to prevent return to criticality as a result of the maximum credible steam break (one safety valve stuck fully open) even assuming that the most reactive control rod is in the fully withdrawn position.

Experimental measurements from critical experiments or operating reactors, or bothy are Used to validate the methods employed in the design. During design, nuclear parameters are calculated for various operational phases and, where applicable, are compared with design limits to show that an adequate margin of safety exists.

Zn the thermal hydraulic design of the core, the maximum fuel and clad temperatures during normal reactor operation and at 118% overpower have been conservatively evaluated and found to be consistent with safe operating limitations.

UNXT 1 3. 1-3 July, 1993

3 '.2 PRINCIPAL DESIGN CRITERIA Reactor Core Desi n Criterion: The reactor core with its related controls and protection systems shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits which have been stipulated and justified. The core and related auxiliary system designs shall provide this integrity under all expected conditions of normal operation with appropriate margins for uncertainties and for specified transient situations which can be anticipated.

The reactor core, with its related control and protection systems, is designed to function throughout its design lifetime without exceeding acceptable fuel damage limits. The core design, together with reliable process and decay heat removal systems, provides for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and anticipated transient situations. This includes the effects of the loss of reactor coolant flow, trip of the turbine generator, and loss of normal feedwater and loss of all off-site power.

The reactor control and protection system is designed to actuate a reactor trip for any anticipated combination of plant conditions, when necessary, to ensure a minimum departure from nucleate boiling (DNB) ratio equal to or greater than the applicable design value for the fuel.

The integrity of fuel cladding is ensured by preventing excessive fuel

'swelling, excessive clad heating, and excessive cladding stress and strain.

This is achieved by designing the fuel rods so that the following conservative limits are not exceeded during normal operation or any anticipated transient condition:

a) Minimum DNB ratio equal to or greater than the applicable design value for the fuel. For the current cycle, the design values are given in Section 3.5.3.

b) Fuel center temperature below melting point of UO UNIT 1 3.1-4 July 1990

For p fuel for the initial core and ENC reload fuel, internal gas I

c) pressure less than the nominal external pressure (2250 psia), even at the end of life. For W reload fuel in the current cycle, the rod internal gas pressure shall remain below the value which causes the fuel-cladding diametral gap to increase due to outward cladding creep during steady-state operation.

d) Clad stresses less than the Zircaloy yield strength e) Clad strain less than 1%

f) Cumulative strain fatigue cycles less than 80% of design strain fatigue life for ENC fuel. Cumulative strain fatigue cycles are less than the design fatigue life for W reload fuel in the current cycle.

The ability of fuel designed and operated to these criteria to withstand postulated normal and abnormal service conditions is shown by analyses described in Chapter 14 to satisfy the demands of plant operation well within applicable regulatory limits.

The reactor coolant pumps provided for the plant are supplied with sufficient rotational inertia to maintain an adequate flow coastdown and prevent core damage in the event of a simultaneous loss of power to all pumps.

In the unlikely event of a turbine trip from full power without an immediate reactor trip, the subsequent reactor coolant temperature increase and volume insurge to the pressurizer results in a high pressurizer pressure trip and thereby prevents fuel damage for this transient.

A loss of external electrical load of 50% of full power or less is normally controlled by rod cluster insertion, together with a controlled steam dump to the condenser, to prevent a large temperature and pressure increase in the reactor coolant system. In this case, the overpower-overtemperature UNIT 1 3. 1'-5 July, 1992

protection would guard against any c'ombination of pressure, temperature, and power which could result in a DNB ratio less than the applicable design 0) value during the transient.

En neither the turbine trip nor the loss-of-flow events do the changes in coolant conditions provoke a nuclear power excursion because of the large system thermal inertia and zelatively small void fraction. Protection circuits actuated directly by the coolant conditions identified with core limits aze therefore effective in preventing core damage.

Suppression of Powez Oscillations Criterion: The design of the reactor core with its related controls and protection systems shall ensure that power oscillat'ons the magni.tude oz which could cause damage in excess of acceptable fuel damage limits, aze not possible or can be readily suppressed.

The potential for possible spatial oscillations oz power distribution for this core has been reviewed. Zt is concluded that low zrequency xenon oscillations may occur in the axial dimension, and control rods can be used to supress these oscillations. The core is expec=ed to be stable to xenon oscillations in the X-Y dimension. Out-of-core .'nstrumentation is provided to obtain necessary inzormation concerning power d'stribution. This instrumentation is adequate to enable the operator to monitor and control xenon induced oscillations. (Zn-core instrumentation is used to periodically calibrate and verify the information provided by the out-of-core instrumentation.) The analysis, detection and control oz these oscillations is discussed in Reference 2) of Sub-Chapter 3.3.

Redundancy of Reactivit Control Criterion: Two independent reactivity control systems, preferably oz different principles, shall be provided.

Two independent reactivity control systems are pzovided, one involving rod cluster contzol (RCC) assemblies and the other involving chemical shimming UNlT 1 3.1-6 July 1990

Reactivit Hot Shutdown Ca abilit Criterion: The reactivity control systems provided shall be capable of making and holding the core subcritical from any hot standby or hot operating condition.

The reactivity control systems provided are capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes. The maximum excess reactivity expected for the core occurs for the cold, clean condition at the beginning of life of the initial core.

The rod cluster control (RCC) assemblies are divided into two categories comprising control banks and shutdown banks. The control banks used in combination with chemical shim control provide control of the reactivity changes of the core throughout the life of the core during power operation.

These banks of RCC assemblies are used to compensate for short term reactivity changes at power that might be produced due to variations in reactor power level or in coolant temperature. The chemical shim control is used to compensate for the more slowly occurring changes in reactivity throughout core life, such as those due to fuel depletion and fission product buildup.

Reactivit Shutdown Ca abilit Criterion: One of the reactivity control systems provided shall be capable of making the core subcritical under any antici-pated operating condition (including anticipated opera-tional transients) sufficiently fast enough to prevent exceeding acceptable fuel damage limits. Shutdown margin should assure subcriticality with the most reactive control rod fully withdrawn.

The reactor core, together with the reactor control and protection system is designed so that the applicable minimum allowable DNBR value is satisified and there is no fuel melting during normal operation, including anticipated transients, UNIT 1 ~ 3.1-7 July 1990

The shutdown groups are provided to supplement the control groups of RCC assemblies to make the core at least 1.6 percent subcritical at the hot zero power condition (keff - 0.984) following trip from any credible operating ff condition, assuming the most reactive RCC assembly is in the fully withdrawn position.

Sufficient shutdown capability is also provided to maintain the core subcritical, assuming the most reactive rod to be in the fully withdrawn position for the most severe anticipated cooldown transient associated with a single active failure, e.g., accidental opening of a steam bypass, or relief valve, or safety valve stuck open. This is achieved by the combination of control rods and automatic boric acid addition via the emergency core cooling system. The design minimum shutdown margin is 1.6 percent, assuming the maximum worth control rod is in the fully withdrawn position, and allowing 10% uncertainty in the control rod worth calculations.

Manually controlled boric acid addition is used to maintain the shutdown margin for the long term conditions of xenon decay and plant cooldown.

Redundant equipment is provided to guarantee the capability of adding boric acid to the reactor coolant system.

Reactivit Holddown Ca abilit Criterion: The reactivity control systems provided shall be capable of making the core subcritical under credible accident conditions with appropriate margins for contingencies, and shall be capable of limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public.

Currently, normal reactivity shutdown capability is provided within 2.4 seconds following a trip signal by control rods, with boric acid injection used for the long term xenon decay transient and for plant cooldown. As discussed in response to the previous criteria, the shutdown capability prevents return to critical as a result of the cooldown associated with a safety valve stuck fully open.

UNIT 1 3.1-8 July 1990

Any time that the reactor ia at power, the quantity of boric acid retained in the boric acid tanks and ready for injection always exceeds that quantity required for the normal cold shutdown. This quantity always exceeds the quantity of boric acid required to bring the reactor to hot shutdown and to compensate for subsequent xenon decay. Boric acid is pumped from the boric acid tanks by one of two boric acid transfer pumps to the suction of one of three charging pumps which inject boric acid into the reactor coolant. Any charging pump and either boric acid transfer pump can be operated from diesel generator power on loss of station power. Boric acid can be injected by one pump at a rate which takes the plant to 1% shutdown in the hot condition with no rods inserted in less than 90 minutes. Enough boric acid can be injected to compensate for xenon decay although xenon decay below the equilibrium operating level does not begin until approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after shutdown. If two boric acid pumps are available, these time periods are halved. Additional boric acid injection is employed if it is desired to bring the reactor to cold shutdown conditions.

On the basis of the above, the injection of boric acid is shown to afford backup reactivity shutdown capability, independent of control rod clusters which normally serve this function in the short term situation. Shutdown for long term and reduced temperature conditions can be accomplished with boric acid injection using redundant components, thus achieving the measure of reliability implied by the criterion.

Alternately, boric acid solution at lower concentration can be supplied from the refueling water storage tank. This solution can be transferred directly by the charging pumps or alternately by the safety injection pumps.

The reduced boric acid concentration lengthens the time required to achieve equivalent shutdown.

eact ivit Control S stems Mal funct i on Criterion: The reactor protection systems shall be capable of protecting against any single malfunction of the reactivity control system, such as unplanned, continuous withdrawal (not ejection or dropout) of a control rod, by limiting reactivity transients to avoid exceeding acceptable fuel damage limits.

UNIT 1 3.1-9 July, 1992

The reactor protection systems are capable of protecting against any single credible malfunction of the reactivity control system, by limiting reactivity transients to avoid exceeding acceptable fuel damage limits.

Reactor shutdown with rods is completely independent of the normal rod control functions since the trip breakers completely interrupt the power to the rod mechanisms regardless of existing control signals.

Details of the effects of continuous withdrawal of a control rod and continuous deboration are described in Chapters 14 and 9 respectively.

Maximum Reactivitv Viorth of Control Rods Criterion: Limits, <<hich inc'ude reasonable marg'n, shall be placed on the max.'mum reactivity worth of control rods or elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactiv~ ty cannot (a) rupture the reactor coolant pressure "boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to lose capability of cooling the core.

Limits, which include considerable margin, are p'-"ed on the maximum reactivity worth of control rods or elements and cn rates at which reactivity can be increased to ensure that the po= ntial effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals so as to lose capability to cool the core.

The reactor control system employs control rod clusters. A portion of these are designated shutdown rods and are fully wichdrawn during power operation. The remain'ng rods comprise the control groups which are used to control reactivity changes due to load changes and to control reactor coolant temperature. The rod cluster drive mechanisms are wired into preselected groups, and are therefore prevented f om be'ng withdrawn in other than their respective groups. The rod drive mechanism is of the magnetic latch .type and the coil actuation is sequenced to provide var'able speed rod travel. The maximum reactivity insertion rate is UNIT 1 3.1-10 July 1990 p

analyzed in the detailed plant analysis assuming two of the highest worth groups to be accidentally withdrawn at maximum speed, yielding reactivity

-4 insertion rates of the order of 7.5 x 10 dk/k/sec, which is well within the capability of the overpower-overtemperature protection circuits to prevent core damage.

No single credible mechanical or electrical control system malfunction can cause a rod cluster to be withdrawn at a speed greater than 72 steps per minute (-45 inches per minute).

3.1.3 SAFETY LIMITS The reactor is capable of meeting the performance objective throughout core life under both steady state and transient conditions without violating the integrity of the fuel elements. Thus the release of unacceptable amounts of fission products to the coolant is prevented.

The limiting conditions for operation established in the Technical Specifications specify the functional capacity of performance levels permitted to assure safe operation of the facility.

Design parameters which are pertinent to safety limits are specified below for the nuclear, control, thermal and hydraulic, and mechanical aspects of the design.

Nuclear Limits At full power, the current predicted nuclear heat flux hot channel factor/

F does not exceed 2.15 for W fuel. The equations and curves which show the F limits as a function of power and fuel height are defined in the Core Operating Limits Report and in Section 3.2.2 of the Cook Nuclear Plant Unit 1'echnical Specif ications.

For any condition of power level, coolant temperature, and pressure which is permitted by the control and protection system during normal operation and anticipated transients, the hot channel power distribution is such that the UNIT 1 3.1-11 July, 1992

minimum DNB ratio is greater than or equal to the applicable design value given in Section 3.5.3.

Reactivit Control Limits The control system and the operational procedures provide adequate control of the core reactivity and power distribution. The following control limits are met:

a. A minimum hot shutdown margin as shown in the Technical Specifications is available assuming a 10% uncertainty in the control rod calculation.

b. This shutdown margin is maintained with the most reactive RCCA in the fully withdrawn pos'tion.

The shutdown margin 's maintained at ambient temperature by the use ox soluble poison.

Thermal and H draulic Limits The reactor core is designed to meet the followi"..>> limi.ting thermal and hydraulic criteria:

a. The minimum allowable DNBR duxing normal operation, including anticipated transients, is not less than the applicable DNBR design limit. For the current cycle, design limit is given in Section 3.5.3.

b. No fuel melting 'during any anticipated operating condition.

To maintain fuel rod integrity and prevent fission product release, it is necessary to prevent clad overheating under all operating conditions. This is accomplished by preventing a departure from nucleate boiling (DNB) which causes a large decxease in the heat transfer coefficient between the .fuel

, rods and the reactor coolant resulting in high clad temperatures.

UNIT 1 July 1990

The ratio of the heat flux causing DNB at a particular core location, as predicted by the W-3 and WRB-1 correlations, to the existing heat flux at the same core location is the DNB ratio.'he applicable design limit DNB ratio for W and ENC fuel corresponds to a 95t probability at a 95't confidence level that DNB does not occur and is chosen to maintain an appropriate margin to DNB for all operating conditions.

Mechanical Limits Reactor Internals The reactor internal components are designed to withstand the stresses resulting from startup, steady state operation with any number of pumps running, and shutdown conditions. No damage to the reactor internals occurs as a result of loss of pumping power.

Lateral deflection and torsional rotation of the lower end of the core barrel is limited to prevent excessive movements resulting from seismic disturbances and thus prevent interference with rod control cluster assemblies. Core drop in the event of failure of the normal supports is limited so that the rod cluster control assemblies do not disengage from the fuel assembly guide thimbles'he internals are further designed to mainuain their functional integrity in the event of a major loss-of-coolant accident. The dynamic loading resulting from the pressure oscillations because of a loss-of-coolant accident does not cause sufficient deformation to prevent rod cluster control assembly insertion.

Fuel Assemblies The fuel assemblies are designed to perform satisfactorily throughout their lifetime. The loads, stresses, and strains resulting from the combined effects of flow induced vibrations, earthquakes, reactor pressure, fission gas pressure, fuel growth, thermal strain, and differential expansion during both steady state and transient reactor operating UNIT 1 3.1-13 July 1990

conditions have been considered in the design of the fuel rods and fuel assembly. The assembly is also structurally designed to withstand handling and shipping loads prior to irradiation, and to maintain sufficient integrity at the completion of design burnup to permit safe removal from the core, subsequent handling during cooldown, shipment and fuel reprocessing.

The fuel rods are supported at seven locations along their length within the fuel assemblies by grid assemblies which are designed to maintain contzol of the lateral spacing between the rods throughout the design life of the assemblies. The magnitude of the support loads provided by the grids are established to minimize possible fretting without overstressing the cladding at the points of contact between the gzids and fuel rods and without imposing restraints of sufficient magnitude to result in buckling or distortion of the rods.

The fuel rod cladding is designed to withstand operating pressure loads without rupture and to maintain encapsulation of the fuel throughout the design life.

Rod Cluster Control Assemblies The criteria used for the design of the cladding on the individual absorber rods in the rod cluster control assemblies (RCCA) are similar to those used for the fuel rod cladding. The gladding is designed to be free standing under all operating, conditions and will maintain encapsulation of the absorber material throughout the absorber rod design life. Allowance for wear during operation is included for the RCCA cladding thickness.

Adequate clearance is provided between the absorber rods and the guide thimbles which position the rods within the fuel assemblies so that coolant flow along the length of the absorber rods is sufficient to remove the heat generated without overheating of the absorber cladding.

UNZT 1 3. 3.-14 July 1990 0

The clearance is also sufficient to compensate for any misalignment between the absorber rods and guide thimbles and to prevent mechanical interference between the rods and guide thimbles under any operating conditions.

Control Rod Drive Assembl Each control rod drive assembly is designed as a hermetically sealed unit to prevent leakage of reactor coolant. All pressure-containing components are designed to meet the requirements of the ASME Code,Section III, Nuclear Vessels for Class A vessels.

The control rod drive assemblies for the full length rods provide rod cluster control assembly insertion and withdrawal rates consistent wi.th the required reactivity changes for reactor operational load changes. This rate is based on the worths of the various rod groups, which are established to limit power-peaking flux patterns to design values. The maximum reactivity addition rate is specified to limit the magnitude of a possible nuclear excursion resulting from a control system or operator malfunction. Also, the control rod drive assemblies for the full length rods provide a fast insertion rate'uring a "trip" of the RCC assemblies which results in a rapid shutdown of the reactor for conditions that cannot be handled by the reactor control system.

UNIT 1 3.1-15 July 1990

Reactor Internals Desi n Descri tion The reactor internals are designed to support and orient the reactor core fuel assemblies and contxol rod assemblies, absorb the control rod dynamic loads and transmit these and other loads to the reactor vessel flange, provide a passageway for the reactor coolant, and support in-core instrumentation. The reactor internals are shown in Figure 3.2.1-2.

The internals are designed to withstand the forces due to weight, preload of fuel assemblies, control rod dynamic loading, vibration, and earthquake acceleration. These internals are analyzed in a manner similar to Connecticut Yankee, San Onofre, Zorita, Saxton and Yankee.

Under the loading conditions, including conservative effects of design earthquake loading, the structure satisfies stress values prescribed in Section III, ASME Nuclear Vessel Code.

The reactor internals are equipped with bottom-mounted in-core instrumentation supports. These supports are designed to sustain the applicable loads outlined above.

The components of the reactor internals are divided into three parts consisting of the lower core support structure (including the entire core barrel and thermal shield), the upper core support structure and the in-core instrumentation support structure.

Lower Core Support Structure The major containment and support member of the reactor internals is the lower core support structure, shown in Figure 3.2.1-5.

This support structure assembly consists of the core barrel, the core baffle, and lower core plate and support columns, the thermal shield/

3~2 3 July, 1982

the intermediate diffuser plate and the bottom support plate which is welded to the core barrel. All the major material for this structure is Type 304 Stainless Steel. The core support structure is supported at, its upper flange from a ledge in the reactor vessel head flange and its lower end is restrained in its transverse movement by a radial support system attached'o the vessel wall. Within the core barrel are axial baffle and former plates which are attached to the core barrel wall and form =the enclosure periphery of the assembled core.

The lower core plate is positioned at the bottom level of the core below the baffle plates and provides support and orientation for the fuel assemblies.

The lower core plate is a 2" inch thick member through which the necessary flow distributor holes for each fuel assembly are machined.

Fuel assembly locating pins (two for each assembly) are also inserted into this plate. Columns are placed between this plate and the bottom support plate of the core barrel in order to provide stiffness and to transmit the core load to the bottom support plate. Intermediate between the support plate and lower core support plate is positioned a perforated plate to diffuse uniformly the coolant flowing into the core.

The one piece thermal shield is fixed to the core barrel at the top with rigid bolted connections. The bottom of the thermal shield is connected to the core barrel by means of axial flexures. This bottom support allows for differential axial growth of the shield/core barrel but restricts radial or horizontal movement of the bottom of the shield.

Rectangular tubing in which material samples can be inserted and irradi-ated during reactor operation are welded to the thermal shield and extend to the top of the thermal shield. These samples are held in the rectangular tubing by a preloaded spring device at the top and bottom.

3.2-4 July, 1982

Substantial scale model testing was performed at Westinghouse'his included tests which involved a complete full scale fuel assembly which was operated at reactor flow, temperature and pressure conditions. Tests were run on a 1/7th scale model of the Indian Point Unit 2 reactor. Measurements taken from these tests indicate very little shield movement, on the order of a few mils when scaled up to Indian Point Unit 2. Strain gauge measurements taken on the core barrel also indicate very low stresses. Testing to determine thermal shield excitation due to inlet flow disturbances have been included. Information gathered from these tests was used in the design of the thermal shield and core barrel.

In order to provide further confirmation of the internals design, Indian Point Unit 2 had deflection gauges mounted on the thermal shield top and bottom for the hot-functional tests. Six such gauges were mounted in the top of the thermal shield equidistant between the fixed supports and eight located at the bottom, equidistant between the six flexures, and two next to flexure supports. The internals inspection, just before the hot-functional tests, included. looking at mating bearing surfaces, main welds and welds used on bolt locking devices. At the conclusion of the hot-functional tests, measurement readings were taken from the deflectometers on the shield and the internals were re-examined at all key areas for any evidence of mal-function. It can be concluded from the testing programs, analyses and the experience gained from Indian Point Unit 2, that the design as employed on this plant is adequate.

Core Com onents Core components for the initial core are discribed in the following subsections. The current Westinghouse Company reload fuel is described in Section 3.5.

UNIT 1 3.2-11 July 1990

Desi n Descri tion Westinghouse Fuel Assembly All of the Westinghouse fuel assemblies which have been in the core were of similar design. The overall configuration of the fuel assemblies is shown in Figures 3.2.1-8 and 3.2.1-9. The assemblies are square in cross-section, nominally 8.426 inches on a side, and have an overall height of 160.1 inches.

The fuel rods in a fuel assembly are arranged in a square array with 15 rod locations per side and a nominal centerline-to-centerline pitch of 0.563 inch between rods. Of the total possible 225 rod locations per assembly, 20 are occupied by guide thimbles for the RCCA rods and one for in-core P

instrumentation. The remaining 204 locations contain fuel rods. Xn addition to fuel rods, a fuel assembly is composed of a top nozzle, a bottom nozzle, 7 grid assemblies, 20 absorber rod guide thimbles, and one instrumentation thimble.

I The guide thimbles in conjunction with the grid assemblies and the top and bottom nozzles comprise the basic structural fuel assembly skeleton. The top and bottom ends of the guide thimbles are secured to the top and bottom nozzles respectively. The grid assemblies, in turn, are fastened to the I

guide thimbles at each location along the height of the fuel assembly at which lateral support for the fuel rods is required. Within this skeletal framework the fuel rods are contained and supported and the rod-to-rod centerline spacing is maintained along the assembly.

Bottom Nozzle The bottom nozzle is a square box-like structure which controls the coolant flow distribution to the fuel assembly and functions as the bottom structural element of the fuel assembly. The nozzle, which is square in cross-section, is fabricated from Type 304 stainless steel 3.2-12 July, 1982

The complete drive mechanism, shown in Figure 3.2.1-13, consists of the l.

internal (latch) assembly, the pressure vessel, the operating coil stack, the drive shaft assembly, and the rod position indicator coil stack.

Each assembly is an independent unit which can be dismantled or assembled separately. Each mechanism pressure housing is threaded onto an adaptor on top of the reactor pressure vessel and seal welded . The operating drive assembly is connected to the control rod (directly below) by means of a grooved drive shaft. The upper section of the drive shaft is suspended from the working components of the drive mechanism. The drive shaft and control rod remain connected during reactor operation, including tripping of the rods.

Main coolant fills the pressure containing parts of the drive mechanism.

All working components and the shaft are immersed in the main coolant.

Three magnetic coils, which form a removable electrical unit and surround the rod drive pressure housing induce magnetic flux through the housing wall to operate the working components. They move two sets of latches which lift, lower and hold the grooved drive shaft.

The three magnets are turned on and off in a fixed sequence by solid-state switches for the full length rod assemblies.

The sequencing of the magnets produces step motion over the 144 inches of normal control rod travel.

The mechanism develops a lifting force approximately two times the static lifting load. Therefore, extra lift capacity is available for overcoming mechanical friction between the moving and the stationary parts. Gravity provides the drive force for rod insertion and the weight of the whole rod assembly is available to overcome any resistance.

• A leak in a CRDM lower canopy seal weld was repaired using a mechanical seal clamp.

UNIT 1 3.2-31 July 1990

0 The mechanisms are designed to operate in water at 650 F and 2485 psig. The 0

temperature at the mechanism head adaptor will be much less than 650 F because it is located in a region where there is limited flow of water from the reactor core, while the pressure is the same as in the reactor pressure vessel.

A multi-conductor cable connects the mechanism's operating coils to the 125 volt d-c power supply.

Latch Assembly The latch assembly contains the working components which withdraw and insert the drive shaft and attached control rod. It is located within the pressure housing and consists of the pole pieces for three electromagnets. They actuate two sets of latches which engage the grooved section of the drive shaft.

The upper set of latches move up or down to raise or lower the drive rod by 5/8 inch. The lower set of latches have a maximum 1/16 inch axial movement to shift the weight of the control rod from the upper to the lower latches.

Pressure Vessel The pressure vessel consists of the pressure housing and rod travel housing.

The pressure housing is the lower portion of the vessel and contains the latch assembly. The rod travel housing is the upper portion of the vessel.

It provides space for the drive shaft during its upward movement as the control rod is withdrawn from the core.

3.2-32 July, 1982

e) Pellet-to-pellet gaps All fuel rods are inspected by gamma scanning or other approved methods to ensure that no significant gaps exist between pellets.

f) Gamma Scanning All fuel rods are active gamma scanned to verify enrichment control prior to acceptance for assembly loading.

g) Traceability Traceability of rods and associated rod components is established by Quality Control.

4) Rod Upgrading The rods, upon final inspection, are upgraded and available for fuel assembly loading.

5) Assembly Inspection consists of 100 percent inspection of drawing requirements.

6) Other Inspection The following inspection is performed as part of routine inspection operation:

a) Measurements other than those specified above which are critical to thermal and hydraulic analyses are obtained to enable evaluation of manufacturing variations to a 99.5% confidence level.

UNIT 1 3.2-47 July, 1987

b) Tool and gauge inspection and control including standardization to primary and secondary working standards. Tool inspection is performed at prescribed intervals on all serialized tools. Complete records are kept of calibration and condition of tools.

c) Check audit inspection of all inspection activities and records to assure that prescribed methods are followed and that all records are correct and properly maintained.

d) 'Surveillance of outside contractors, including approval of standards and methods is performed where necessary.

However, all final acceptance is based upon inspection performed at the Westinghouse plant.

To prevent the possibility of mixing enrichments during fuel manufacture and assembly, meticulous process control is exercised.

The UF6 is received from the DOE diffusion plant in 5000 lb cylinders.

These cylinders are tagged with the enrichment of the contents. In addition, samples of the contents are attached. These samples are analyzed by Westinghouse to verify the enrichment of the contents.

Following verifications, the cylinders are moved to the production area, where they are piped in to the UF6 to U02 conversion process equipment and thereafter (during the conversion of the particular region of the core) remain a permanent part of the process equipment.

Upon completion of this conversion, the U02 is placed into 'sealed containers which are color coded to identify the enrichment of the contents.

Hovement of powder from the conversion area to the pellet production area can be made by one authorized group only who direct the powder to the correct pellet production line. All pellet production lines are physically separated from each other and pellets of only a single enrichment and density are produced in a given production line.

UNIT 1 3.2-48 July, 1987

IP Finished pellets are placed on trays having the same color code as the powder containers and transferred to segregated storage racks. Physical barriers prevent mixing of pellets of different densities and enrichments in -this storage area. Unused powder and substandard pellets to be repro-cessed are returned to storage in the original color coded containers.

Loading of the pellets into the cladding is again accomplished in isolated production lines and again only one density and enrichment is loaded on a line at a time.

At the time of loading, the top fuel tube end plug identification character is checked with the density and enrichment identification of the color code of the pellet storage tray. After each fuel tube is seal welded, it is given the same color coding as has been carried throughout the previous processes. The fuel tube remains color coded until just prior to installation in the fuel assembly. The color coding and end plug identification character provide a cross refer-ence of the fuel contained in the fuel rods.

At the time of installation into an assembly, the color coding is removed. After the fuel rods are installed, an inspector verifies that all fuel rods in an assembly have the same end plug identi-fication, and that the top nozzle to be used on the assembly carries the correct identification character describing the fuel enrichment and density for the core region being fabricated. The top nozzle identification then becomes the permanent description of the fuel contained in the assembly.

Burnable Poison Rod Tests and Ins ections The end plug seal welds are checked for integrity by visual inspection and X-ray. The finished rods are helium leak checked.

UNn' 3.2-49 July, 1987

REFERENCES SECTION 3. 2. 1

1. Daniel, R. C., et al, "Effects of High Burnup on Zircaloy-Clad Bulk U02, Plate Fuel Element Samples, "WAPD-263, (September, 1965).

2. Large Closed Cycle Water Reactor Research and Development Program Quarterly Progress Reports for the Period January 1963 through June 1965 (WCAP-3738, 3739, 3743, 3750, 3269-2, 3269-3, 3269-5, 3269-6, 3269-12 and 3269-13).

3. J. S. Moore, WCAP-9000 "Nuclear .Design of Westinghouse PWR's with Burnable Poison Rods", March 1969.

4. WCAP-7072 "Use of Part Length Absorber Rods in Westinghouse Pressurized Water Reactors".

3.2-50 July, 1982

TABLE 3.3.1-1 (cont'd.)

BURNABLE POISON RODS

39. Number and Material 1436 Borosilicate Glass

40. Worth Hot Full Power ~ 9.0X

41. Worth Cold ~ 7.0X KINETIC CHARACTERISTICS

42. Moderator Temperature Coefficient at Full Power -4 (Ap/o F) .3 x 10 to

- 3.2 x 10 -4

-6

43. Moderator Pressure Coefficient (Ap/psi) + .3 x 10 to 4.0 x 10 3 -5

44. Moderator Density Coefficient, hp/yn/cm .1 x 10 to 0.8 x 10 5

-5

45. Doppler Coefficient (dp/oF) 1.0 x 10 to

-5 1.7 x 10

46. Delayed Neutron Fraction, X .51 to .70

47. Prompt Neutron Lifetime, sec. 1.4 x 10 to 2.0 x 10

48. Boron Worth Ap/ppm 1.4 x 10

.09 x 10 UNIT 1 3.3-30 July, 1993

TABLE 3.3.1-2 REACTIVITY REQUIREMENTS FOR CONTROL RODS Per Cent hp Beginning End Re uirements of Life of Life Control Power Defect 1.70 3.05 Rod Insertion Limit 0 70 0.50 Total Control 2.40 3.55 UNIT 1 3.3-31 July 1990

TABLE 3 ~ 5.1-1 Westi ouse 5x15 OFA Desi n Parameters 15xl5 W Optimized Fuel

~Paramete Assembl Des ~Re iaa 14 Fuel Assembly Length, in. 159.785 159.975 Fuel Rod Length, in. 151.85 152. 17 Assembly Envelope, in. 8.426 Compatible with Core Internals Fuel Rod Pitch, in. 0.563 Number of Fuel Rods/Ass'y 204 Number of Guide Thimbles/Ass'y 20 Number of Instrumentation Tube Ass'y Compatible. with Moveable Zn-Core Yes Detector System Fuel Tube Material Zircaloy-4 Fuel Rod Clad OD, in. 0 422 Fuel Rod Clad Thickness, in. 0 0243 Fuel/Clad Gap, mil 7.5 Fuel Pellet Diameter, in. 0.3659 Guide Thimble Material Zircaloy-4 Guide Thimble ZD, ln.+ 0.499 Structural Material - Five Inner Zircaloy-4 Grids Structural Material - Two End Grids Inconel Grid height, in., Outer Straps, 2.25 (Inner Grids)

Valley-to-Valley 1.50 (End Grids)

Bottom Nozzle Reconstitutable Top Nozzle Holddown Springs 3-leaf Above dashpot UNIT 1 3.5.1-19 July, 1992

Table 3.5.1-2 intentionally deleted.

UNIT 1 3.5.1-20

3.5.2 NUCLEAR DESIGN The nuclear design of cores with W OFA is accomplished by using the standard calculational methods as described in the W Reload Safety Evaluation Methodology. In addition to Westinghouse's standard methods, the Westinghouse Advanced Nodal Code (ANC)

(9) was introduced in Cycle 11 to (3) was introduced in perform core neutronics analyses and the PHOENIX-P code Cycle 12 to calculate lattice physics constants.

Each reload core design is evaluated to assure that design and safety limits for the fuel are satisfied according to the W reload safety evaluation methodology. For the evaluation of the worst-case F (2) envelope, axial power shapes are synthesized with the limiting F xy values chosen over three overlapping burnup windows during the cycle.

In order to accommodate potential increases in future feed enrichments, a criticality analysis of the fuel storage areas was performed for nominal enrichments in W 15x15 OFA fuel up to and including 4.55 wt.'4 U-235 for the new fuel storage vault and 4.95 wt.'4 U-235 for the spent fuel pool. These analyses confirm that all current safety criteria applicable to fuel storage are satisfied. (2) 3.5.2.1 Computerized Methods, Codes and Cross Section Data Three principal computer codes have been used in the nuclear design of reactor cores with W OFA; these are PHOENIX-P (two-dimensional), APOLLO (one-dimensional), and ANC (two-dimensional and three-dimensional).

Descriptions and uses for these codes follow.

(3) is a two-dimensional multi-group transport theory code used to I'HOENIX-P calculate lattice physics constants. Microscopic cross section data are based on a 42-energy group structure that has'een derived from the CSRL-V 227 group ENDF/B-V library (4) . It provides the capability for cell lattice modeling on an assembly level. In the core design, PHOENIX-P is used to provide homogenized, two-group cross-sections for nodal calculations and feedback models. It is also used in a special geometry to generate appropriately weighted constants for the baffle/reflector regions.

UNIT 1 3.5.2-1 July, 1992

advanced version of (5) g is a two-group, one-dimensional APOLLO i an PANDA diffusion-depletion code. APOLLO utilizes the burnup dependent radially averaged macroscopic cross sections of the corresponding 3-D model. The APOLLO model is used -as an axial model. APOLLO is utilized to determine axial power and burnup distributions, differential rod worths, and control rod operational limits (insertion limits, return to power limits, etc.).

ANC (6) is an advanced nodal code that is used in two-dimensional and three-dimensional calculations. ANC calculations include power and burnup distributions, critical boron concentrations, reactivity coefficients, control rod worths, and various safety analysis calculations. ANC is used to validate one dimensional results from APOLLO and to provide information about radial (X-Y) peaking factors as a function of axial position. ANC also has the capability of calculating discrete pin powers and pin burnups from the nodal information.

Additional support codes are used for special calculations such as determining fuel temperatures.

3.5.2.2 Neutronic Design of Cook Nuclear Plant Unit 1 Reactor Core 3.5.2.2.1 Analytical Input The neutronics design methods utilized to calculate the data presented herein are consistent with'those described previously with primary reliance upon the ANC code.

For each cycle, the burnup history of each of the fuel assemblies retained from previous cycles for further energy production is calculated by a three-UNIT 1 3.5.2-2 July, 1992

dimensional model which is utilized to simulate operation of the core for previous cycles.

As an example, Cycle 13 core calculations used assembly exposures calculated from the Cycle 12 burnup of 16,541 MWD/MTU.

3.5.2.2.2 Design Bases For each cycle, the nuclear design bases are very similar to those for the example Cycle 13 core as follows:

1~ At core full power, 3250 MWt (not including pump heat), nuclear peaking factors of 2.15 and 1.55 for F~T and F<H N respectively, will not be exceeded. In addition, at any relative power level P (0.0 < P < 1.0),

T N F() and F<H shall not exceed the bases of the plant control and protection system.

2. The moderator temperature coefficient at operating conditions greater 0

than 70% power level is a ramp function limited to +5.0 pcm/ F at 70%

0 power and 0.0 pcm/ F at 100% power. Below 70% power level, the 0

moderator temperature coefficient shall be less than +5.0 pcm/ F.

3 ~ With the most reactive control rod stuck out of the core, the remaining control rods shall be able to shut the reactor down by a sufficient reactivity to reduce the consequences of any credible accident to acceptable levels.

4. The effects of all accident situations in Cycle 13 will be acceptable and compatible with the safety bases of the Final Safety Analysis Report (FSAR), as specified in Reference 7.

5. The fuel loading specified shall be capable of generating approximately 15360 MWD/MTU at normal full power operating conditions during Cycle 13.

UNIT 1 3.5.2-3 July, 1993

3.5.2.2.3 Design Description and Results Each cycle's reactor core consists of 193 W OFA assemblies, each having a 15x15 fuel rod array. A description of the W OFAs is given in Section 3.5.1.

As an example, the Cycle 13 loading pattern is given in Figure 3.5.2-1 which shows the region number, sources, and the burnable absorber configuration.

The core consists of 48 fresh W OFAs with an average enrichment of 3.099 w/o U-235, 32 fresh OFAs with an average enrichment of 3.609 w/o, 80 once burnt OFA assemblies, and 33 twice burnt OFA assemblies. A low leakage loading pattern was developed which results in the scatter-loading of the fresh OFAs throughout the interior of the core. 3328 new ZFBA rods are present in a number of OFAs to control power peaking and MTC. The ZFBA rods contain approximately 0.0018 gm/in of B-10. Pertinent fuel assembly parameters for the Cycle 13 fuel are given in Tables 3.5.1-1 and 3.5.2-1.

Ph sics Characteristics The neutronics characteristics of a reactor core with W OFA fuel are presented in Table 3.5.2-2. These reactivity coefficients are bounded by the coefficients used in the safety analysis. For an example cycle length, Cycle 12 was pro)ected to be 15,360 MWD/MTU at a core power of 3,250 MWt with 10 ppm soluble boron remaining.

Power Distribution Considerations Figure 3.5.2-2 shows the K(Z) function (fuel height limit for normalized F (Z) ). Each cycle's core loading satisfies the envelope shown in Figure Q

3.5.2-2.

UNZT 1 3.5.2-4 July, 1993

Control Rod Reactiv t Re irements The Cook Nuclear Plant Unit 1 Technical Specifications require a minimum shutdown margin of 1,600 pcm in operational Modes 1, 2, 3 and 4 and 1000 pcm in operational Mode 5 at BOC and EOC. As an example, detailed calculations of shutdown margins for Cycle 12 are presented in Table 3.5.2-3. The Cycle 12 analysis indicates excess shutdown margin of 1817 pcm at BOC and 1874 pcm at EOC.

Insertion limits are specified for the control rod groups and are given in the Core Operating Limits Report, as described in Technical Specification 6.9.1.11. The control rod shutdown requirements allow for a HFP D-Bank insertion equivalent to 500 pcm at both BOC and EOC. Table 3.5.2-3 gives the shutdown requirements for the example of Cycle 12.

oderator Tem erature Coefficient Core loadings must satisfy the Technical Specifications requirements that 0

the moderator temperature coefficient be less than or equal to +5 pcm/ F below 70% of rated thermal power and less than or equal to a linear ramp 0 o between +5 pcm/ F at 70% power and 0 pcm/ F at 100% power.

UNIT 1 3.5.2-5 July, 1992

FERENCES SECTION 3. 5. 2

1. Bordelon, F. M., et al., "Westinghouse Reload Safety Evaluation Methodology," WCAP-9272 (Prop.) and WCAP-9273 (Non-Prop.), March 1978.

2~ Alexich, M. P. to Murley, T. E., "Proposed Units 1 and 2 License Conditions and Technical Specifications Changes for Unit 2 Cycle 8 Spent Fuel Pool and New Fuel Storage Vault," AEP:NRC:1071F, December 8, 1989.

3. Nguyen, T. Q., et al. "Qual'.f ication of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, WCAP-11596, November 1987

4. Ford, W ED i et al., "CSRL-V: Processed ENDF/B-V 227-Neutron Group and Point-wise Cross Section Libraries for Criticality Safety, Reactor and Shielding Studi esi NUREG/CR 2306@ ORNL/CSD/TM 160@ ( 1982) ~

5. R. F. Barry, C. C. Emery, and T. D. Knight, "The PANDA Code," WCAP-7048, (April 1967).

6. Y. S. Liu, A. Meliksetion, J. A. Rathkopf, D. C. Little, F. Nakano, and M. J. Poploski, "ANC A Westinghouse Advanced Nodal Computer Code,"

WCAP-10965-P-A (December 1985).

7. Donald C. Cook Nuclear Plant Unit 1 Cycle 13, Reload Safety Evaluation (August 1992)

UNIT 1 3.5.2-6 July, 1993

Figure 3.5.2-1: Exaapla Cora LoadQxg PanarTL:

0. C. COOK UNIT 1 CYCLE 13 R P N M L K J H G F E 0 C B A f80 138 14A 148 14A 148 14A 138 138 LGA LGB L58 15A 158 LGA LGB 158 15A 138 48 48 32 48 32 48 48 D SC D 138 14A 158 148 148 14A 14A 14A 148 148 158 14A 138 48 48 SB SS SB LGA 158 14A 15A L3B 15A L48 LGA LGA 14A 158 15A 48 48 48 48 48 48 SA 8 SA 138 158 148 15A L4A LGA 14A 148 14A 15A 14A LGA 148 158 138 48 48 48 48 48 48 A SA A SS 14A 158 148 138 15A 13A LGA 14A LGA 13A 15A 138 148 158 14A 48 48 48 48 48 48 6 D SD C 0 148 15A 14A LGA 14A LGA 13A 158 13A LGA 14A LGA L4$ 15A 148 32 48 48 48 48 48 32 7 SB SA SB

,gg o'< 14A 158 48 14A 148 148 14A L58 48 138 158 48 14A 148 148 14A 158 48 14A SC C C 8 SC 148 LGA 14A 15A 14A 15A L3A 158 13A LGA 14A 15A '14A 15A 148 32 48 48 48 48 48 32 9 SB SA SB 14A 158 148 138 15A 13A 15A 14A LGA 13A 15A L38 148 158 14A 48 48 48 48 48 48 0 SD C SD 0 138 158 148 15A 14A 15A 14A 148 14A 15A L4A 15A 148 158 L38 48 48 48 48 48 48 SS A SA 15A 158 14A 15A 138 15A 148 15A 138 15A 14A 158 LGA 48 48 48 48 48 48 12 SA 8 SA 138 14A 158 148 148 14A 14A 14A 148 148 158'8 14 A 138 48 A SB SS SB 138 LGA 158 158 LGA 158 LGA 158 158 15A 138 48 48 32 48 32 48 48 14 0 SC 0

. 138 14A 148 14A 148 14A 138 0

X Region tvueosr T/SS Hueoer oP I'Fgg'u< neo 1 ~ Josor oer s/'jeconoary Sources CTL Control Rcc iacstion UNIT 1

Ho~ @gal ~ QhNlfl4 FSCtOf Normalized Operating Envelope, FQ ECCS Umit ~ 5~ 6 Cook Nuclear Plant Un1t 1 Cycle 12 0.0. 1.0 I 6.0. 4.0 I 2 0 3

--c

.40

.20 0.00 12 6

CORE HElCKT (FT)

UNIT 1 July, 1992

NOTES FOR TABLE 3.1-1

[a] These numbers are based on Improved Thermal Design Procedure in Reference 2.

2

[b] The value of 437,800 BTU/hr-ft is associated with a Cycle 1 value of F of 2.32. The Cycle 3 value is 375,500 BTU/hr-ft2 corresponding to a peaking factor of 1.99.

[c] This value of 12.6 Kw/ft is associated with a Cycle 1 value of F of 2.32. The Cycle 3 value is 10.98 Kw/ft associated with a peaking factor of 1.99.

[d] See Section 3.3.2.2.6.

[e] The value of F - 2.32 was the value of F for normal operation reported in the original FSAR. The value for Cycle 3 is 1.99.

[f] The reload feed enrichments for Cycle 8 were 1.5, 3.6 and 4.2 w/o.

[g) These numbers are based on Revised Thermal Design Procedure in Reference 3.

UNIT 2 3.1-11 July 1991

TABLE 3.1-2 ANALYTIC TECHNI UES IN CORE DESIGN Section

~Anal sis ~Techni ne Com uter Code Referenced Mechanical Design of Core Internals Loads, Deflections, and Static and Dynamic Blowdown code, FORCE, 14.3.3 Stress Analysis Modeling Finite element structural analysis code, and others Fuel Rod Design Fuel Performance Characteristics Semi-empirical thermal Westinghouse fuel rod 3.2.1.3.1 (temperature, internal pressure, model of fuel rod with design model 3.3.3.1

'clad stress, etc.) consideration of fuel 3.4.2.2 density changes,-heat 3.4.3.4.2 transfer, fission gas release, etc.

Nuclear Design

1. Cross Sections and Group Microscopic data Modified ENDF/B-V library 3.3.3.2 Constants Macroscopic constants PHOENIX-P 3.3.3.2 for homogenized core regions UNIT 2 3.1-12 July 1991

This scheme of grid fastening is a standard for Westinghouse and has been used successfully since the introduction of Zircaloy guide thimbles in 1969.

The central instrumentation thimble of each fuel assembly is constrained by seating in counterbores in each nozzle. This tube is of constant diameter and guides the incore neutron detectors. This tube is expanded at the top and mid grids in the same manner as the previously discussed expansion of the guide thimbles to the grids.

Grid Assemblies The fuel rods, as shown in Figure 3.2-2, are supported at intervals along their length by grid assemblies which maintain the lateral spacing between the rods. Each fuel rod is supported within each grid by the combination of support dimples and springs. The grid assembly consists of individual slotted straps interlocked and brazed in an "egg-crate" arrangement to join the straps permanently at their points of intersection. The straps contain spring fingers, support dimples and mixing vanes.

The grid material for the top and bottom grids is Inconel-718, chosen because of its corrosion resistance and high strength. The grid material for the mid-grids is Zircaloy, chosen because of its low neutron absorption characteristic and its extensive successful in-reactor use. The magnitude of the grid restraining force on the fuel rod is set high enough to minimize possible fretting, without overstressing the cladding at the points of contact between the grids and fuel rods. The grid assemblies also allow axial thermal expansion of the fuel rods without imposing restraint sufficient to develop buckling or distortion of the fuel rods.

'The intermediate flow mixer (IFM) grids shown in Figure 3.2-2 are located in the three uppermost spans between the Zircaloy-4 mixing vane structural grids and incorporate a similar mixing vane array. The primary function of the IFM grid is to provide mid-span flow mixing in the hottest fuel assembly spans.

Each IFM grid cell contains four dimples which are designed to prevent mid-span channel closure in the spans containing IFMs and fuel rod contact with UNIT 2 3.2-13 July 1991

the IFM mixing vanes. This simplified cell arrangement allows for short grid cells so that the IFM grid can accomplish its flow mixing objective with minimal pressure drop.

The IFM grids are not intended to be structural members. The outer strap configuration was designed similar to current fuel designs to preclude grid hang-up and damage during fuel handling. Additionally, the grid envelope is smaller which further minimizes the potential for damage and reduces calculated forces during seismic/LOCA events. A eoolable geometry is, therefore, assured at the IFM grid elevation, as well as at the structural grid elevation.

3.2.1.3 Desi n Evaluatio 3.2.1.3.1 Fuel Rods The fuel, rods are designed to assure that the design bases are satisfied for Condition I and II events. This assures that the fuel performance and safety criteria (Section 3.2) are satisfied.

Mater als - Fuel Claddin The desired fuel rod clad is a material which has a superior combination of neutron economy (low absorption cross section), high strength (to resist deformation due to differential pressures and mechanical interaction between fuel and clad), high corrosion resistance (to coolant, fuel and fission products), and high reliability. Zircaloy-4 has this desired combination of clad properties. As shown in Reference (4), there is considerable PWR operating experience on the capability of Zircaloy as a clad material. Clad hydriding has not been a significant cause of clad perforation since current

'(4) controls on levels of fuel contained moisture were instituted Materials - Fue Pe lets Sintered, high density uranium dioxide fuel reacts only slightly with the clad at core operating temperatures and pressures. In the event of clad UNIT 2 3.2-14 July 1991

The UO powder is kept in sealed containers. The contents are fully identified both by descriptive tagging and preselected color coding. A Westinghouse identification tag completely describing the contents is affixed to the containers before transfer to powder storage. Isotopic content is confirmed by sample isotopic analysis.

Powder withdrawal from storage can be made by only one authorized group, which directs the powder to the correct pellet production line. All pellet production lines are physically separated from each other and pellets of only a single nominal enrichment and density are produced in a given production line.

Finished pellets are placed on trays identified with the same color code as the powder containers and transferred to segregated storage racks within the confines of the pelletizing area. Samples from each pellet lot 'are tested for isotopic content and impurity levels prior to acceptance by Quality Control. Physical barriers prevent mixing of pellets of different nominal densities and enrichments in this storage area. Unused powder and substandard pellets are returned to storage in the original color coded containers.

Loading of pellets into the clad is performed in isolated production lines and again only one density and enrichment is loaded on a line at a time, except when natural uranium is loaded into axial blankets. Then natural and enriched uranium pellets are separately identified by their different pellet lengths.

A serialized traceability code is placed on each fuel tube which identifies the contract and enrichment. The end plugs are inserted and inert welded to seal the tube. The fuel tube remains coded, and traceability identified until just prior to installation in the fuel assembly. The traceability code provides an identification of the fuel contained in the fuel rods.

UNIT 2 3.2-30 July 1991

At the time of installation into an assembly, the traceability codes are removed and a matrix is generated to identify each rod in its position within a given assembly. After the fuel rods are installed, an inspector verifies that all fuel rods in an assembly carry the correct identification character describing the fuel enrichment and density for the core region being, fabricated. The top nozzle is inscribed with a permanent identification number providing traceability to the fuel contained in the assembly.

Similar traceability is provided for burnable poison, source rods and control rodlets as required.

3,2.1.4.3 Tests and Inspections by Others If any tests and inspections are to be performed on behalf of Westinghouse, Westinghouse will review and approve the quality control procedures, inspection plans, etc. to be utilized to ensure that they are equivalent to the description provided above and are performed properly to meet all Westinghouse requirements.

3:.2.1.4.4 Onsite Inspection Onsite inspection programs for fuel, control rods and internals are based on the NSSS supplier's detailed procedures. In the event reloads or other components are supplied by other suppliers additional programs will be developed based on that supplier's procedures.

Loaded fuel containers, when received on site, are externally inspected to ensure that labels and markings are intact and seals are unbroken. After the containers are opened, the accelerometers are inspected to determine if movement during transit exceeded design limitations.

Following removal of the fuel assembly from the container in accordance with detailed procedures from the fuel fabricator, the polyethylene wrapper is removed and a visual inspection of the entire bundle is performed.

UNIT 2 3.2-31 July 1991

Control rod assemblies are shipped in fuel assemblies and are inspected prior to removal of the fuel assembly from the container.

Surveillance of fuel and reactor performance is routinely conducted by operating personnel. Coolant activity and chemistry are followed to permit early detection of any fuel clad defects.

Visual fuel inspection is routinely conducted during refueling. Additional fuel inspections are dependent on the results of the operational monitoring and the visual inspections.

3.2.2 REACTOR VESSEL. INTERNALS The design bases for the mechanical design of the reactor vessel internals components are as follows:

1. The reactor internals in conjunction with the fuel assemblies shall direct reactor coolant flow through the core to achieve acceptable flow distribution and to restrict bypass flow so that the heat transfer performance requirements are met for all modes of operation. In addition, required cooling for the pressure vessel head shall be provided so that the temperature differences between the vessel flange

'nd head do not result in leakage from the flange during reactor operation.

2, In addition to neutron shielding provided by the reactor coolant, a separate thermal shield is provided to limit the neutron exposure of the pressure vessel material in order to maintain the required ductility of the material for all modes of operation.

3. Provisions shall be made for installing incore instrumentation useful for the plant operation and vessel material test specimens required for a pressure vessel irradiation surveillance program.

UNIT 2 3.2-32 July 1991

4. The reactor internals "shall be designed to withstand mechanical loads arising from operating basis earthquake, safe shutdown earthquake and pipe ruptures and meet the requirement of Item 5 below.

5. The reactor shall have mechanical provisions which are sufficient to adequately support the core and internals and to assure that the core is intact with acceptable heat transfer geometry following transients arising from abnormal operating conditions.

6. Following the design basis accident, the plant shall be capable of being shutdown and cooled in an orderly fashion so that fuel cladding temperature is kept within specified limits. This implies that the deformation of certain critical reactor internals must be kept sufficiently small to prevent overstressing of fuel elements to failure plus allow adequate core cooling.

The functional limitations for the core structures during the design basis accident are shown in Table 3.2-1. To ensure no column loading of rod cluster control guide tubes, the upper core plate deflection is limited to not exceed the value shown in Table 3.2-1.

Details of the dynamic analyses, input forcing functions, and response loadings are presented in Section 14.3.3.

3.2.2.2 Desc it o and Drawin s The reactor vessel internals are described as follows:

The components of the reactor internals are divided into three parts consisting of the lower core support structure (including the entire core barrel and thermal shield), the upper core support structure and the incore instrumentation support structure. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and control rod drive mechanisms, direct coolant flow UNIT 2 3.2-33 July 1991

Part Len th Rod Cluster Control Assembl Part Length Control Rods are not installed in Unit 2 of the Donald C. Cook Nuclear Plant for the following reasons:

a. No credit is taken for their presence in the safety analysis performed by the vendor. Therefore, the decision not to mount them does not constitute a safety issue,

b. The reactor's Operating License and Technical Specifications preempt their use,

c. Unit 1 of the Donald C. Cook Nuclear Plant has successfully load followed and controlled artificially created large xenon oscillations without the use of these rods, and

d. The storage of these rods outside the reactor vessel will prevent their irradiation and subsequent radioactivity.

Anti-rotation devices will be installed in a manner similar to the practice followed during the hot functional tests. Permanent anti-rotation devices similar to those in Unit 1 will be installed later.

Burnable Absorber Assembl Each burnable absorber assembly consists of burnable absorber rods attached to a hold down assembly. Burnable absorber assemblies are shown in Figure 3.2-15.

The absorber rods consist of borosilicate glass tubes contained within Type 304 stainless steel tubular cladding which is plugged and seal welded at the ends to encapsulate the glass. The glass is also supported along the length of its inside diameter by a thin wall tubular inner liner. The top end of the liner is open to permit the diffused helium to pass into the void volume and the liner overhangs the glass. The liner has an outward flange at the bottom end to maintain the position of the liner with the glass. A typical UNIT 2 3.2-52 July 1991

burnable absorber rod is shown in longitudinal and transverse cross sections in Figure 3.2-16.

The absorber rods in each fuel assembly are grouped and attached together at the top end of the rods to a hold down assembly by a flat, perforated retaining plate which fits within the fuel assembly top nozzle and rests on the adaptor plate. The retaining plate (and the poison rods) is held down and restrained against vertical motion through a spring pack which is attached to the plate and is compressed by the upper core plate when the reactor upper internals assembly is lowered into the reactor. This arrangement ensures that the absorber rods cannot be ejected from the core by flow forces. Each rod is permanently attached to the base plate by a nut which is lock welded into place.

The clad in the rod assemblies is 10 percent cold worked Type 304 stainless steel. All other structural materials are Types'04 or 308 stainless steel except for the springs which are Inconel-718. The borosilicate glass tube provides sufficient boron content to meet the criteria discussed in Section 3.3.1.

eutron Source Assembl The purpose of the neutron source assembly is to provide base neutron level to ensure that the detectors are operational and responding to core multiplication neutrons. Since there is very little neutron activity during loading, refueling, shutdown and approach to criticality, a neutron source is placed in the reactor to provide a minimum neutron count of at least 2 counts per second on the source range detectors attributable to core neutrons. The detectors, called source range detectors, are used primarily when the core is subcritical and during special subcritical modes of operations.

The source assembly also permits detection of changes in the core multiplication factor during core loading, refueling and approach to criticality. This can be done since the multiplication factor is related to an inverse function of the detector count rate. Therefore, a change in the multiplication factor can be detected during addition of fuel assemblies UNIT 2 3.2-53 July 1991

holding force cr'eated by a single winding is sufficient to overcome the rundown torque produced by the mechanism load. Therefore, the rod cannot move except under the control of the power supply.

The rotational energy is supplied in sequential pulses to the armature which rotates directionally 15 degrees per pulse as controlled by the power supply.

3.2.3.3 Desi n Evaluatio 3 '.3.3.1 Reactivity Control Components The components are analyzed for loads corresponding to normal, upset, emer'gency and faulted conditions. The analysis performed depends on the mode of operation under consideration.

The scope of the analysis requires many different techniques and methods, both static and dynamic.

Some of the loads that are considered on each component where applicable are as follows:

-1. Control rod trip (equivalent static load).

2. Differential pressure.

3. Spring preloads.

4. Coolant flow forces (static).

5. Temperature gradients'.

Differences in thermal expansion.

a. Due to temperature differences.

b. Due to expansion of different materials.

7. Interference between components.

8. Vibration (mechanically or hydraulically induced).

9. Operational transients.

10. Pump overspeed.

UNIT 2 3.2-66 July, 1982

11. Seismic loads (operating basis earthquake and safe shutdown earthquake).

12. Blowdown forces (due to cold or hot leg break).

13. Material swelling, and gas generation pressure.

The main objective of the analysis is to satisfy allowable stress limits, to assure an adequate design margin, and to establish deformation limits which are concerned primarily with the functioning of the components. The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials.

Standard methods of strength of materials are used to establish the'tresses and deflections of these components. The dynamic behavior'f the reactivity control components has been studied using experimental test data and experience from operating reactors.

The design of incore component rods provides a sufficient cold void 'volume within the burnable absorber and source rods to limit the internal pressures to a value which satisfies the criteria in Section 3.2.3.1. The void volume for the helium in the burnable absorber rods is obtained through the use of glass in tubular form which provides a central void along the length of the rods. Helium gas is not released by the neutron absorber rod material, thus the absorber rod only sustains an external pressure during operating conditions. The internal pressure of source rods continues to increase from ambient until end of life. The stress analysis of reactivity component rods assumes 100 percent gas release to the rod void volume.

Based on available data for properties of the borosilicate glass and on nuclear and thermal calculations for the burnableabsorber,rods, gross swelling or cracking of the glass tubing is not expected during operation. Some minor creep of the glass at the hot spot on the inner surface of the tube could occur but would continue only until the glass came in contact with the inner liner. The wall thickness of the inner UNIT 2 3.2-67 July 1991

3. All clad/end plug welds are checked for integrity by visual inspection, X-ray and are helium leak checked. All the seal welds in the neutron absorber rods, burnable absorber rods and source rods are checked in this manner.

4. To assure proper fitup with the fuel assembly, the rod cluster control, burnable absorber and source assemblies are installed in the fuel assembly without restriction or binding in the dry condition with a force not to exceed 15 pounds. Also a straightness of 0.01 in./ft. is required on the entire inserted length. of each rod assembly.

The full length rod cluster control assemblies are functionally tested, following core loading but prior to criticality to demonstrate reliable operation of the assemblies. Each assembly is operated (and tripped) one time at no flow/cold conditions and one time at full flow/hot conditions. In addition, selected assemblies, amounting to about 15 to 20 percent of the total assemblies are operated at no flow/operating temperature conditions and full flow/ambient conditions. Also the slowest rod and the fastest rod are tripped 10 times at no flow/ambient conditions and at full flow/operating temperature conditions. Thus each assembly is tested a minimum of 2 times or up to 14 times maximum to ensure the assemblies are properly functioning.

3.2.3.4.2 Control Rod Drive Mechanisms Quality assurance procedures during production of control rod drive mechanisms include material selection, process control, mechanism component tests and inspections during production and hydrotests.

S After all manufacturing procedures had been developed, several prototype control rod drive mechanisms, and drive rod assemblies were life tested with the entire drive line under environmental conditions of temperature, UNIT 2 3.2-82 July 1991

pressure and flow. All acceptance tests confirm the 3 x 10 step life 6

capability of the control rod drive mechanism and drive rod assembly.

These tests include verification that the trip time achieved by the full length control rod drive mechanisms meet the design requirement of 2.7 seconds from the beginning of decay of stationary gripper goil voltage to dashpot entry. This trip time requirement will be confirmed for, each control rod drive mechanism prior to initial reactor operation and at periodic intervals after initial reactor operation. In addition, a Technical Specification has been set to ensure that the trip time requirement is met.

It is expected that all control rod drive mechanisms will meet specified operating requirements for the duration of plant life with normal refurbishment. However, a Technical Specification pertaining to an inoperable rod cluster control assembly has been set.

If a rod cluster control assembly cannot be moved by its mechanism, adjustments in the boron concentration ensure that adequate shutdown margin would be achieved following a trip. Thus, inability to move one rod cluster control assembly can be tolerated. More than one inoperable rod cluster control assembly could be tolerated, but would impose additional demands on the plant operator. Therefore, the number of inoperable rod cluster control assemblies has been limited to one.

In order to demonstrate proper operation of the control rod drive mechanism and to ensure acceptable core power distributions during operation partial rod cluster control assembly movement checks are performed on the full length rod cluster control assemblies as described in the Technical Specifications.

In addition, periodic drop tests of the full length rod cluster control assemblies are performed at each refueling shutdown to demonstrate continued

'ability to meet trip time requirements, to ensure core subcriticality after reactor trip, and to limit potential reactivity insertions from a UNIT 2 3.2-83 July 1991

20, Weiner, R. A., et. al., ",Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP-11873-A, August 1988.

UNIT 2 3.2-90 July 1991

TABLE 3.2-1 MAXIMUM DEFLECTIONS ALLOWED FOR REACTOR INTERNAL SUPPORT STRUCTURES No-Loss-Of Allowable Function

~Com onent Deflections in Deflections in Upper Barrel radial inward 8.2 radial outward 1.0 Upper Package 0.10 0.15 Rod Cluster Guide Tubes 1.00 1.75 UNIT 2 3.2-91 July 1991

TABLE 4 1-1 SYSTEM DESIGN AND OPERATING PARAMETERS Plant design life, years 40 Number of heat transfer loops Design pressure, psig 2485 Nominal operating pressure, psig 2235 Total system volume including pressurizer and surge line (ambient conditions), ft3 (estimated) 12,500 System liquid volume, including pressurizer and surge line (ambient conditions), ft3 11,892 System liquid volume, including pressurizer max. guaranteed power, ft 3 (estimated) 11,780 6

Total Reactor heat output (100% power) Btu/hr 11,089 x 10 (Unit 1)

(3250 MWt) 6 11,641 x 10 (Unit 2)

(3411 MWt)

Unit 1 Unit 2 Bounding Conditions for Rerating Lower/Upper Reactor vessel coolant temperature at full power:

Inlet, nominal, o F 514.9/545.2 541.3 Outlet, nominal, 0 F 579.1/607.5 606.4 Coolant temperature rise in vessel at full power, avg., 0 F 64.2/62.3 64.8 6

Total coolant flow rate, lb/hr x 10 139 ~ 0/133 ~ 9 134.6 Steam pressure at full power, psia 618/820 820 o

Steam Temp. 8 full power, F 489.4/521.1 521.1 Total Reactor Coolant Volume at ambient conditions, ft3 12,438 12,470 4.1-25 July 1990

TABLE 4.1-2 REACTOR COOLANT SYSTEM DESIGN PRESSURE SETTINGS Pressure si Unit 2 Design Pressure 2485 2485 Operating Pressure 2235 2235 Safety Valves 2485 2485 Power Relief Valves* 2335 2335 Pressurizer Spray Valves (Begin to Open) 2260 2260 Pressurizer Spray Valves (Full Open) 2310 2310 Pressurizer Pressure High Reactor Trip 2378 2378 High Pressure Alarm 2310 2310 Pressurizer Pressure Low Reactor Trip > 1865 > 1950 Low Pressure Alarm 2135 2135 Pressurizer Pressure Low - Safety Injection ) 1815 1900 Hydrostatic Test Pressure 3106 3106 Backup Heaters On 2185 2185 Proportional Heaters (Begin to Operate) 2250 2250 Proportional Heaters (Full Operation) 2220 2220

• During Start-up and Shut-down when the Reactor Coolant System temperature is below 331<F for Unit 1 and 331<F for Unit 2, a safeguard circuit is manually switched on which allows opening of that Unit's two Power Relief Valves at 435 psig for Unit 1 and 435 psig for Unit 2 for low temperature overpressure protection (LTOP) of the Reactor Vessel. This safeguard circuit ensures that the reactor pressure remains below the ASME Section III, Appendix G "Protection Against Non-ductile Failure" limits in the case of an LTOP event.

4.1-26 July, 1993

The pressurizer relief tank, by means of its connection to the Waste Disposal System, provides a means for removing any non-condensable gases from the Reactor Coolant System which might collect in the pressurizer vessel.

Steam is discharged through a sparger pipe under the water level.

This condenses and cools the steam by mixing it with water that is near ambient temperature. The tank is equipped with an internal spray and a drain which are used to cool the tank following a discharge.

The tank is protected against a discharge exceeding the design value by two rupture discs which discharge into the reactor containment.

The tank is carbon steel with a corrosion-resistant coating on the wetted surfaces. A flanged nozzle is provided on the tank for the pressurizer discharge line connection. This nozzle and the discharge piping and sparger within the vessel are austenitic stainless steel.

The tank design is based on the requirement to condense and cool a discharge of pressurizer steam equal to 110 percent of the volume above the zero-power pressurizer water level set-point. The tank is not designed to accept a continuous discharge from the pressurizer.

The volume of water in the tank is capable of absorbing the heat, from the assumed discharge, with an initial temperature of 120'F and in-creasing to a final temperature of 200'P. If the temperature xn the tank rises above 120'F during plant operation, the tank is cooled by spraying in cool water and draining out the warm mixture to the Waste Disposal System.

The spray rate is designed to cool the tank from 200'P to 120'F in approximately one hour following the design discharge of pressurizer steam. The volume of nitrogen gas in the tank is selected to limit the maximum pressure following a design discharge to 50 psig.

4.2-9. July, 1982

The rupture discs on the relief tank have a relief capacity equal to the combined capacity of the pressurizer safety valves. The tank design pressure is twice the calculated pressure resulting from the maximum safety valve discharge described above. The tank and rupture discs holders are also designed for full vacuum to prevent tank collapse if the contents cool following a discharge without nitrogen being added.

Principal design parameters of the pressurizer relief tank are given in Table 4.1-4.

Dischar e Pi in The discharge piping (from the safety and power-operated relief valves to the pressurizer relief tank) is sized to prevent back-pressure at the safety valves from exceeding 20 percent of the set-point pressure at full flow. The pressurizer safety and power relief valves discharge lines are stainless steel.

4.2.2.4 Steam Generators The steam generators are vertical shell and U-tube heat exchangers with integral moisture separating equipment. The reactor coolant flows through the inverted U-tubes, entering and leaving through the nozzles located in the hemispherical bottom head of the steam generator. The head is divided into inlet and outlet chambers by a vertical partition plate extending from the head to the tube sheet. Manways are provided for access to both sides of the divided head. 'eedwater enters the, steam generators and is distributed thru a feedwater ring located just below the moisture separators. Thermal sleeves are provided in the feedwater piping elbows at the steam generator inlet. Feedwater flow is out of the top of the feedwater ring thru "J" tubes, down between the steam generator shell and tube bundle wrapper and into the tube bundle just above the tube sheet. The "J" t'ubes prevent rapid drainage of the feedwater ring due to a drop in steam generator water 4.2-10. July, 1982

Zndication of valve position for the pressurizer safety and power-operated relief valves is provided by a four channel acoustic flow monitor. There are'our accelerometers, one strapped to the discharge of each of the three pressurizer safety valves and one on the common discharge of the three power relief valves. Flow through any of these valves produces an acoustic energy input to the respective accelerometer and this is amplified on the assigned channel of the monitor which is located in the control room. Zndication on four vertical rows of light emitting diodes represents a bar graph display of relative flow through the monitored valves.

Pressurizer Safet V ives The pressurizer safety valves are totally enclosed pop-type valves. The valves are spring-loaded, self-activated and with back-pressure compensation designed to prevent system pressure from exceeding the design pressure by more than 110 percent, in accordance with the ASME Boiler and Pressure Vessel Code, Section ZZZ. The set pressure of the valves is 2485 psig.

The 6" pipes connecting the pressurizer nozzles to their respective safety valves are shaped in the form of a loop seal. Piping is connected to the bottom of each loop seal to drain any condensate that accumulates in the loop seal. An acoustic flow monitor and a temperature indicator on each valve discharge alerts the operator to the passage of steam due to leakage or valve lifting.

4. 2-21 July 1991

ower Relief Va vea The pressurizer is equipped with 3 power-operated relief valves which limit system pressure for a large power mismatch and thus lessen the likelihood of an actuation of the fixed high-pressure reactor trip. The relief valves operate automatically or by remote manual control. The operation of these valves also limits the undesirable operation of the spring-loaded'afety valves. Remotely operated stop valves are provided to isolate the power-operated relief valves. An acoustic flow monitor and a temperature indicator on the common discharge of the relief valves alerts the operator to the passage of steam due to leakage or valve opening.

During startup and shutdown transient conditions, when the reactor coolant system temperature is below 331 F for Unit 1 and 331'F for Unit 2, a safeguard circuit is manually energized in the control room to allow automatic opening of that unit's two power relief valves at 435 psig for Unit 1, and 435 psig for Unit 2, for low temperature over pressure (LTOP) protection of the reactor vessel. This safeguard circuit ensures that the reactor pressure remains below the ASME Section ZZI, Appendix G "Protection Against Nonductile Failure" limits in the case of an LTOP event.

Design parameters for the pressurizer spray control, safety, and power relief valves are given in Table 4.1-8.

4 2.2.9 Reactor Coolant S stem Su orts

1. Steam Generator Support Each steam generator is supported by a structural system consisting of four vertical support columns and upper and lower lateral restraints approximately 46/ feet apart. The vertical columns have a ball )oint connection at each end to accommodate both the radial growth of the steam generator itself and the radial movement of the vessel from the reactor center.

4.2-22 July, 1993

The lower lateral support consists of an inner frame, keyed and shimmed to the four steam generator support feet to accommodate radial growth of these feet. The inner frame is surrounded by an outer frame which is embedded in both the primary shield and I

crane wall concrete. The connection between the inner and outer frame consists of a series of shimmed points which act as both guides and limit stops to allow for expansion from the center of the reactor. The lower lateral support restrains both torsional and translational movements.

The upper lateral support consists of a ring band which is shimmed to the steam generator at twelve locations around the circumference.

Attached to this'and are lugs 180'part which are shimmed and guided to a structural framing system which is embedded in the crane wall and steam generator enclosure wall concrete. Hydraulic snubbers are also connected 180'part on the band and tied to other embedded frames in a direction coincident with the direction of movement away from the reactor center. The upper lateral support restrains rapid translational movements in all horizontal directions.

2. Reactor Vessel Supports The reactor vessel is supported by four of its eight nozzles by four individual weldments embedded in the primary shield concrete.

Each nozzle pad bears on a shoe, that is supported by a heavy U-shaped weldment which wraps around the shoe. The U-shaped weld-ment is water-cooled at the junction of the outer flange and the web by two continuous welded angles on either side of the web.

The U-shaped weldment bears vertically on two shims and is res-trained horizontally by a series of shims and bearing plates.

These bearing plates and shims are connected to an outer weldment which completely surrounds the U-shaped weldment and is embedded in the concrete.

4.2-23 July, 1982

The reactor support system allows the reactor to expand radially from its vertical centerline but resists rotational motion in all orthogonal planes. The nozzle horizontal centerlines translate in the vertical direction relative ta the shoes.

3. Pressurizer Support The pressurizer is supported on a ring girder which is in turn supported on a concrete slab. Horizontally, the vessel is restrained at two elevations approximately 27 feet apart.

The lower restraint consists of anchor bolts in slightly over-size holes in the ring girder. The upper restraint consists of four individual weldments embedded in concrete that allow the pressurizer to expand radially, but resist torsional and translational horizontal movements.

4. Reactor Coolant Pump Support Each reactor coolant pump is supported vertically by three ball joint ended columns. This structural column system resists both overturning and vertical movement while allowing for expan-sion from the center of reactor. Excessive torsional and hori-zontal translational movements are resisted by a combination of lateral thrust columns anchored into the crane wall concrete.

4.2.3 PRESSURE-RELIEVING DEVICES The Reactor Coolant System is protected against overpressure by control and protective circuits such as the high pressure trip and by relief and safety valves connected to the top head of the pressurizer. The relief and safety valves discharge into the pressurizer relief tank which condenses and collects the valve effluent. The schematic arrang-ement of the relief devices is shown in Figure 4.2-1A, and the valve design parameters are given in Table 4.1-8. The valves are further discussed in Sub-Section 4.2.2.7.

4.2-24 July, 1982

TABLE 4 '-3 STEAM GENERATOR WATER STEAM-SIDE CHEMISTRY SPECIFICATION FOR 100% FULL POWER Cation Conductivity s 0 8 umhos/cm pH 9 254C 8.5 - 9.4 without boric acid h 7.5 with boric acid Chloride 5 20 ppb Sodium 20 ppb Sulfate 5 20 ppb If boric acid is present in the system, the cation conductivity specification will be S [0.8 + 0.03 x (boron conc. in ppm)].

4.2-41 July, 1993

atmosphere, (2) the reactor coolant system, or (3) closed systems inside containment that are assumed vulnerable to accident forces.

One barrier may be the containment isolation valve itself, which is between the containment atmosphere and the test connection.

For closed systems inside containment, which are Seismic Class I design with a low probability of failure, only one barrier is required. Test connections are provided with locked closed valves and/or pipe caps and are administratively controlled.

10. All normally closed locked valves and caps are administratively controlled to ensure containment integrity.

NOTE: A ~eal may be used in lieu of a lock to satisfy the locking requirements discussed in the section above.

Containment Isolation Testin and Reliabilit The containment isolation system is designed to provide such functional reliability and testing facilities as are necessary to avoid undue risk to the health and safety of the public. The air operated isolation valves close on loss of control power or air. The instrumentation and, control circuits are redundant in the sense that a single failure cannot prevent containment isolation. Provision is made for periodic testing of the leak tightness and functioning of the isolation valves.

Test connections are locked closed and/or capped and are administratively controlled to ensure containment integrity. Therefore, no further testing of test connection leak tightness is required. This is consistent with the clarifications of Appendix J requirements discussed with the NRC during the CILRT inspections conducted February 9-15, 1989 (Inspection Report Nos. 50-315/89007 (DRS) and 50-316/89007 (DRS)).

Containment Isolation S stem Protection Adequate protection for containment isolation, including piping, valves, and 5.4-3 July, 1992

vessels, is provt.ded against dynamic effects and missiles which might resultfrom plant equipment failures including a loss-of-coolant accident.

Isolation valves inside the containment are located between the crane wall or some other missile shi.eld and the outside containment wall. Isolation valves, piping or vessels which provide one of the isolation barriers-outside the containment are similarly protected.

Containment Isolation S stem 0 eration No manual operation is required for immediate isolation of the containment.

Automatic trip valves are provided in those lines which must be isolated immediately following an accident. Lines which must remain in service subsequent to certain accidents for safety reasons are provided with at least one remote manual valve, except instrument sensing lines that are provided with one manual valve.

Automatic trip valves may be operated by a manual switch. The position of each automatic trip valve is displayed in the main control room.

The instrumentation and controls for the system are described in more detail in Chapter 7.

Containment Isolation S stem Pi. in Classes The functional classes of piping are used to further define the design bases.

~class Class A piping is open to the outside atmosphere, and is connected to the reactor coolant system, or is open to the containment atmosphere.

Alternatively, Class A piping is Seismic Class III in design and is assumed to be vulnerable to accident forces.

For Class A piping the following is provided, as a minimum, for isolation subsequent to an incidents 0

5 ~ 4-4, July, 1992

Check valves may be employed as one of the two barriers for incoming lines.

Test connections and pressurizing means are provided to test each isolation valve or barrier for leak tightness. Either water or a gas is used as the pressurizing medium depending on the requirements of each case. Where it is necessary to make a quantative leakage test, provision is made to:

a) measure the inflow of the pressurizing medium, or b) collect and measure the leakage, or c) calculate the leakage. from the rate of pressure drop.

The test connections are isolated when not in use by locked closed manual valves and/or caps and administratively controlled to ensure containment integrity.

All isolation valves are missile protected. isolation valves, actuators, and control devices required inside the containment are located between the missile barrier and the containment wall. Isolation valves, actuators and control devices outside the containment are located outside the path of potential missiles or provided with missile protection.

There are two levels of automatic containment isolati n identified as Phase A and Phase B. Phase A isolation closes all lines penetrating the containment except essential lines such as Safety Injection and Containment Spray which are not isolated, and component cooling water to the reactor pumps and service water to the ventilation units which isolates on Phase B. (For Phase A and B initiating signals see Chapter 7 Instrumentation and Control.) All automatic isolation valves are able to be closed from the main control room. Position indicators are provided for each valve near its manual control switch in the main control room.

Specific administrative procedures govern the positioning of all isolation valves except check valves as well as any flanged closures during normal operation, shutdown and incident conditions. Check valves in incoming lines open only when the fluid pressure in the line coming from the outside is l

higher than the pressure on the containment side. Gravity or a spring holds the valve closed in the balanced pressure condition.

5.4-7 July, 1992

5.4.3 DESIGN EVALUATION The containment isolation system provides two barriers to prevent leakage of radioactivity at each containment opening. Either barrier is sufficient to keep the leakage within limits.

5.4.4 TEST AND INSPECTION All valve leak testing for Inservice Inspdction (ISI) and Integrated Leak Rate Test (ILRT) program and surveillance requirements are performed in accordance with Appendix J to 10 CPR 50 for Type A, B and C type testing.

Also certain valves will be tested for operability in accordance with the applicable edition of the ASME Boiler and Pressure Vessel Code,Section XI.

5.4-8 July, 1992

TABLE 5 4-1 SHEET 1 OF 12 PIPING PENETRATIONS Line Size Status of and Isolatio Valves Isolatio Valves Isolation Number Flow Actuation Figure Service Class of Lines Direction Inside Outside ~ei mal /umber gates Gas Analyzer From A 1/2" (1) Out Int. Closed Closed Auto Trip 2 Auto 4.2-1A Pressurizer Trip Relief Tank Primary Water 3" (1) In Int. Closed Closed Check Auto Trip A 4.2-1A Supply to Pressurizer Relief Tank Nitrogen Supply A 3/4" (1) In Int. Closed Closed Check Auto Trip A 4. 2-1A to Pressurizer Relief Tank Reactor Coolant 2" (4) In Open Open Open Check NA 4.2-1A Pumps Seal Water Supplies Reactor Coolant 4" (1) Out Open Open Closed Auto Trip Auto Trip A 9.2-1 Pumps Seal Water

& Excess Letdown Heat Exchanger Discharges Reactor Coolant 8" (1) In Open Open Closed 2 Auto Trip B 9.5-1 Pump Motor and Thermal Barrier Cooling Water Supply Reactor Coolant 8" (1) Out Open Open Closed 2 Auto Trip B 9 '-1 Pump Motor Cooling Water Discharge 5.4-9 July, 1993

TABLE 5.4-1 SHEET 2 OF 12 PIPING PENETRATIONS Line Size Status of and Isolatio Valves Isolation Valves Isolation Number Flow Actuation Figure Service Class of Lines /%rection N reside ontside ~si nal iinmber Notes Letdown Line 2" (1) Out Open Closed Closed Auto Trip Auto Trip A 9.2-1 (CVCS)

Charging Line 3" (1) In Open Closed Open Check NA 9.2-1 (CVCS)

Excess Letdown 4" (1) In Open Closed Closed Auto Trip A 9 5-1 Heat Exchanger Component Cooling Water Inlet Excess Letdown 4" (1) Out Open Closed Closed Auto Trip A 9.5-1 Heat Exchanger Component Cooling Water Outlet Reactor Coolant A 4" (1) Out Int. Int. Closed 2 Auto Trip A 11 l-l Drain Tank Pump Suction Containment Sump C 3" (1) Out Int. Int. Closed 2 Auto Trip A 11.1-2 Pump Discharge to Waste Disposal 8" In Closed Closed Open Check NA 6.3-1 1 Upper Containment (2)

Spray Inlet Lower Containment 6" (2) In Closed Closed Open Check NA 6.3-1 1 Spray Inlet RHR to 8" (2) In Closed Closed If Check NA 6 3-1 1 Containment Spray Needed

1) Check valves held closed by gravity or spring in balanced pressure condition.

5.4-10 July, 1993

TABLE 5.4-1 SHEET 3 OF 12 PIPING PENETRATIONS Line Size Status of and Isolatio gives Isolatio Valves Isolation Number Flow Actuation Figure Service Class of lines glrection Inside cote ide ~si al /amber Notes Residual Heat B 14" (1) Out Closed Open Closed Remote None 6.2-A 2 Removal Inlet to Manual 9.3-1 Pumps (Normal Cooldown)

Residual Heat 8" (2) In Open Closed Open Remote None 6.2-A Removal To Manual 9 3-1 Reactor Coolant Hot Legs-Low Head S,I'esidual Heat D 18" (2) Out Closed Closed Open Remote None 6.2-A 3 Removal Suction Manual 9 '-1 From Sump Safety Injection D 4" (2) In Open Closed Open Remote None 6.2-1 Manual Safety Injection A 3/4" (1) In or Int. Closed Closed 2 Manual NA Test Line and Out (L Ce)

Accumulator Test Line Boron Injection D 3" (1) In Closed Closed Open Remote None 6.2-1 4 Inlet Manual Residual Heat B 12" (1) In Closed Open If Remote None 6.2-A Removal to Needed Manual 9.3-1 Reactor Coolant Cold Legs (Normal Cooldown)

2) Valve administratively locked closed.

3) Open during recirculation mode.

4) Open automatically on Safety Injection Signal.

5.4-11 July, 1993

TABLE 5 '-1 SHEET 4 OF 12 PIPING PENETRATIONS Line Size Status of and Isolation Valves Isolation Valves Isolation Number Flow Actuation Figure Service Class of Lines ~Directio N Inside Outside Ntuiaal Number ~utes Nitrogen to Accumulators 1" (1) In Int. Int. Closed Check 'uto Trip A 6.2-A 9.3-1 Sample Line From A 1/2" (1) Out Int. Closed Closed 2 Auto Trip A 9.6-1 Pressurizer Steam Space Sample Line from A 1/2" (1) Out Int. Closed Closed 2 Auto Trip A 9.6-1 Pressurizer Liquid Space Sample Line from A 1/2" (1) Out Int. Closed Closed 2 Auto Trip A 9.6-1 Hot Legs Sample Lines from A 1/2" (1) Out Int. Closed Closed 2 Auto Trip A 9.6-1 Accumulators Sample Lines from C 1/2" (4) Out Open Closed Closed Auto Trip A 9.6-1 Steam Generator Steam Outlets Steam Generator C 30" (4) Out Open Closed Closed 10 2-1 5 Main Steam Outlets Steam Generator 2" (4) Out Int. Closed Closed Auto Trip A 10.2-1 Blowdown Lines

5) Steam Generator Stop Valves located outside containment also close on steamline isolation signal as described in Chapter 7.

5.4-12 July, 1993

TABLE 5.4-1 SHEET 5 OF 12 PIPING PENETRATIONS Line Size Status of and Isolation Valves Isolatio Valves Isolation Number Flow Actuation Figure Service Class: of I ines girection N ~side Outside ~S1 a ~umber ~ates Steam Generator C - 14" (4) In Open Closed Closed Check NA 10.5-1 Feedwater Supply Steam Generator 6" (4) In Open Int. If Check NA 10.5-1 6 Auxiliary Needed Feedwater Supply Steam Generator C 1/2" (4) In Closed Int. closed Check NA 10.5-1 6 Chemical Feed SuPPlY Non Essential 6" (4) In Open If Closed 2 Auto Trip B 9.8-6 Service Water to Needed Containment 3" (4)

Ventilation Units Non Essential 6" (4) Out Open If Closed 2 Auto Trip B 9 8-6 Service Water Needed from Containment (4)

Ventilation Units Purge Air Inlet A 30" (1) In If If Closed Auto Trip Auto Trip A or 5.5-2 (Containment) 24" (1) Needed Needed CVI Purge Air Outlet A 30" ( 1) Out If If Closed Auto Trip Auto Trip A or 5.5-2 (Containment) 24" (1) Needed Needed CVI

6) No independent containment penetrations. These lines )oin the Feedwater Lines between the penetrations and the isolation valves.

5 '-13 July, 1993

TABLE 5.4-1 SHEET 6 OF 12 PIPING PENETRATIONS Line Size Status of and Isolatio Valves Isolatio Valves Isolation Number Plow Actuation Figure Service Class of Li.oes pl.rection N Inside Outside Silisal Number gates Fuel Transfer A 20" (1) In or Closed Open Closed Blind NA Tube Out Flange Service Air 2" (1) In Closed Open Closed Check Auto Trip A 9.8-3 Instrument Air 1" (2) In Open Open Closed 2 Auto Trip A 9.8>>3 Reactor Coolant 4" (1) Out Open Open Closed 2 Auto Trip B 9.5-1 Pump Thermal Barrier Cooling Water Discharge Gas Analyzer From A 1/2" (1) Out Int. Closed Closed 2 Auto Trip A 11. 1-1 Reactor Coolant Drain Tank Ice Loading Line E 5" (1) In Closed If Closed Blind Blind NA 5 '-2A Needed Flange Flange Containment A 12" (1) Out If If Closed Auto Trip Auto Trip A or 5.5-2 Pressure Relief Needed Needed CVI Line Containment Teat E 5" (1) In Closed If Closed Blind Blind NA 9 '-3 8 Pressurization Needed Flange Flange

7) See Sub-Chapter 5.2 for description of double gasketed seal on the Fuel Transfer Tube.

8) Same physical line as ice loading line.

5.4-14 July, 1993

TABLE 5 '-1 SHEET 7 OF 12 PIPING PENETRATIONS Line Size Status of and Isolatio Valves splat o Valves Isolation Number Flow Actuation Figure Service Class of Lines ~nireotio Inside Outside ~si na 1 ~umber iiotes Ice Loading 5" (1) Out Closed Int. Closed Blind Blind NA 5 '-2A Return Flange Flange Glycol to Ice 3" (1) In Open Open Closed Auto Trip Auto Trip A 5 '-2A Condenser Fan Coolers Glycol from Ice 3" (1) Out Open Open Closed Auto Trip Auto Trip A 5.3-2A Condenser Fan Coolers Bypass Glycol E 3/8" (1) In Open Open Closed Check NA 5.3-2A

-line to Ice Condenser Fan Coolers Bypass Glycol E 3/8" (1) Out Open Open Closed Check NA'.3-2A line from Ice Condenser Fan Coolers Purge Air Inlet 14" (1) In Closed If Closed Auto Trip Auto Trip A or 5.5-2 Sa (Instrumentation Needed CVI Room)

Purge Air Outlet 14" (1) Out Closed If Closed Auto Trip Auto Trip A or 5.5-2 Sa (Instrumentation Needed CVI Room)

Reactor Coolant 1" (1) Out Int. Closed Closed 2 Auto Trip A 11.1-1 Drain Tank &

Press. Relief Tank Vents Sa) For status "N": "Closed" for Unit 2g Unit 1 is "If needed" (for limited purging).

5.4-15 July, 1993

TABLE 5 4-1 SHEET 8 OF 12 PIPING PENETRATIONS Line Size Status of and solatio Valves Isolation Valves Isolation Number Flow Actuation Figure Service Class ~of li,es ~irectio N ~side outside ~Si oal Number iiotes Refueling Water A 2 1/2" (1) In Closed Int. Closed 2 Manual NA 9.4-1 Supply (L C.)

Demineralized 2" (1) In Closed Open Closed 2 Auto Trip A Water Supply Non Essential 3" (4) In Open If Closed 2 Auto Trip B 9. 8-6 Service Water to Needed Reactor Coolant Pump Motor Air Coolers Non Essential A 3" (4) Out Open If Closed 2 Auto Trip B 9.8-6 Service Water Needed from Reactor Coolant Pump Motor Air Coolers Reactor Support C 2 1/2" (1) In Open If Closed Check Auto Trip A 9.5-1 Cooling Inlet Needed Reactor Support C 2 1/2" Out Open If Closed 2 Auto Trip A 9 '-1 Cooling Outlet (1) Needed Refueling Cavity 3" (1) Out Closed If Closed 2 Manual NA 11.1-1 Drain To Needed (LeCe)

Purification System Nitrogen Supply 1" (1) In Open Open Closed Auto Trip & A 11.1-1 9 to Reactor Check Coolant Drain Tank

9) No independent containment penetration. Joins RCDT vent line between penetration and isolation valves.

5.4-16 July, 1993

TABLE 5.4-1 SHEET 9 OF 12 PIPING PENETRATIONS Line Size Status of and Isolatio Valves Isolatio Valves Isolation Number Flow Actuation Figure Service Class of Lines Direction Inside Outside ~Si. oal ~umber Notes Steam Generator C 1/2" (4) Out Int. Closed Closed Auto Trip A 9.6-1 Blowdown Samples Containment Weld D 1/2" (2) In Closed If Open Check NA 5.6-1 10 Channel Needed Pressurization Air Supply Dead Weight Teat E 1/2" (1) Int. Closed Closed Manual NA 4.2-1A Connection Relief Vent 4" (1) In Int. Int. Int. Check NA 4.2-1A Header Ice Condenser and A 3" (1) Out Open Open Closed 2 Auto Trip A 11. 1-1 Containment (Each Line)

Venti.lation Unit 1" (1)

Drain to Drain Header Component Cooling D 1" (4) In Open Open Open Check Manual NA Water to Main Steam Penetrations Component Cooling D 1 1/2" (2) Out Open Open Open Remote None Water from Main Manual Steam Penetrations Component Cooling D 1 1/2" (2) In Closed Closed Open Remote None Water to Pressure Manual Equalizing Fans

10) May be used for Leak Test of Channels.

5.4-17 July, 1993

TABLE 5 4-1 SHEET 10 OF 12 PIPING PENETRATIONS Line Size Status of and Isolation Valves Isolation Valves Isolation Number Flow Actuation Figure Service Class of i,ines direction ~Iside Outside ~ai nai Number Notes Component Cooling D 1 1/2" (2) Out Closed Closed Open Remote None Water from Manual Pressure Equalizing Fans Containment Air 1" (2) Out Open Open Int, 2 Auto Trip B Particulate and Radio Gas Detector Sample Line Containment Air 1" (1) In Open Open Int. Check Auto Trip B Particulate and Raido Gas Detector Sample Return Lower Containment A 1/2" (2) Out If Closed Closed 2 Manual NA Radi.ation Needed Sampling System Upper Containment A 1/2" (2) Out If Closed Closed 2 Manual NA Radiation Needed Sampling System Instrument Room A 1/2" (2) Out If Closed Closed 2 Manual NA Radiation Needed Sampling System

11) May be put in service manually after incident 5.4-18 July, 1993

TABLE 5 '-1 SHEET 11 OF 12 PIPING PENETRATIONS Line Size Status of and Isolation Valves Isolatio Valves Isolation Number Flow Actuation Figure Service Class ~of li es ~Directio ~aside outside ~Si al ~umbsr gates Non Essential A 2 1/2" (2) In Open If Closed 2 Auto Trip B 9.8-6 Service Water to Needed Instrument Ventilation Units Non Essential A 2 1/2" (2) Out Open If Closed 2 Auto Trip B 9.8-6 Service Water Needed from Instrument Room Ventilation Units Sample Lines Hydrogen to D 1/2" (9) Out Closed Closed Int. 2 Auto Trip A 14.3 12A

'- 11 Monitoring System Sample Line D 1/2" (1) In Closed Closed Int. Auto Trip A 14.3.6- 11 Return From 12A Hydrogen Monitoring System Containment E 1/2" (6) Open Open Open Manual NA 12 Pressure Transmitters Containment Sump D 1/2" (1) Out Closed Closed Int. Auto Trip A 9.6-2 11 Sample to Post-Accident Sampling System Post Accident D 1/2" (1) In Closed Closed Int. Check Auto Trip A 9.6-2 ll Sampling System Return ll) May be put in service manually after incident

12) See Fig. 7.5-1 for a functional diagram of these instruments.

5.4-19 July, 1993

TABLE 5.4-1 SHEET 12 OF 12 PIPING PENETRATIONS Line Size Status of and Isolatio Valves Isolation Valves Isolation Number Flow Actuation Figure Service Class of tines @%rection N ~Isolde Outside ~si el /umber Notes Post Accident D 1/2" (1) Out Closed Closed Int. 2 Auto Trip A 9.6-2 13 Sampling System Supply (Gas)

Reactor Vessel E 3/16" (6) Open Open Open Membrane NA Level Barrier Instrumentation System Incore Flux NA 8" (1) Closed If Closed Blind Blind NA 14 Detection System Needed Flange Flange Spare NA 18" (4) Closed Closed Closed Weld Cap Weld Cap NA Penetrations 6ss (4)

Service E 18" (1) Closed If Closed Hinged Hinged NA Penetration Needed Closures Closures

13) Connected to Containment Air Particulate and Radio Gas Detector Sample Line

14) Used for replacement of incore flux instrumentation thimbles.

Nc Normal Int: Intermittent Isolation Actuation Signalss S: shutdown LeCet Locked Closed As Phase A Isolation I: Incident NAt Not Applicable B! Phase B Isolation OVID Containment Ventilation Isolation (initiated by Safety In)ection Signal or High Containment Radiation) 5.4-20 July, 1993