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Amend 38 to Application for Ol,Consisting of Revision 20 to FSAR
	ML19345D488
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	12/05/1980
From:	Novarro J; LONG ISLAND LIGHTING CO.
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ML19345D485	List: ML19345D484; ML19345D488;
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{{#Wiki_filter:, SNPS-1 FSAR AMENDMENT 38 INSERTION INSTRUCTIONS FOR REVISION 20 TO THE FSAR The following text, tables, and figures are to be inserted in the FSAR. These pages are either replacement pages or new pages as indicated below. All replacement pages which differ from existing pages are identified with the revision number and date in the lower right hand corner. Bars located in the margin of a particular page indicate material which is new in the revision indicated at the bottom of the page. Remove Old (Pages) Insert New (Pages) Volume 1 AM-1 and AM-2 AM-1 and AM-2 EP 1-1 through EP 1-3 EP 1-1 through EP 1-3 EP 2-1 and EP 2-2 EP 2-1 and EP 2-2 EP 5-1 through EP 5-4 EP 5-1 through EP 5-4 EP 6-1 through EP 6-3 EP 6-1 through EP 6-3 (N EP 9-1 and EP 9-2 EP 9-1 and EP 9-2 (_,) EP 11-1 through EP 11-4 EP 11-1 through EP 11-4 EP 13-1 through EP 13-b EP 13-1 through EP 13-6 EP 15-3 and EP 15-4 EP 15-3 and EP 15-4 Table 1.3.1-3 (page 2 of 2) Table 1.3.1-3 (page 2 of 2) Table 1.3.2-1 (pages 5 Table 1.3.2-1 (pages 5 through 17 of 17) through 17 of 17) Volume 2 2.4-17 and 2.4-18 2.4-17 and 2.4-18 Table 2.4.11-2 Table 2.4.11-2 Volume 5 3.1-45 and 3.1-46 3.1-45 and 3.1-46 Volume 6 Figure 3.8.4-8 Figure 3.8.4-8 3B-13 and 3B-14 3B-13 and 3B-14 Ns 1 Revision 20 - November 1980 OO 8032150

SNPS-1 FSAR Remove Old (Pages) Inseart New (Pages) Volume L 7.3-51 and 7.3-52 7.3-51 and 7.3-52 Figure 7.3.1-29A Figure 7.3.1-29A through through Figure 7.3.1-29H Figure 7.3.1-29H Figure 7.3.1-29J through Figure 7.3.1-29J through Figure 7.3.1-29N Figure 7.3.1-2 9N Figure 7.3.1-29P through Figure 7.3.1-29AD Volume 11 Table 8.3.1-1 (page 1 of 4) Table 8.3.1-1 (page 1 of 4) Table d.3.1-1 (pages 3 Table 8.3.1-1 (pages 3 and 4 of 4) and 4 of 4) Ta ble 8 . 3 .1- 2 (page 2 of 2) Table 8.3.1-2 (page 2 of 2) 9.1 and 9-2 9.1 and 9-2 9.2-1 through 9.2-6 9.2-1 through 9.2-b 9.2-6a/b 9.2-6a and 9.2-6b Table 9.2.1-1 Table 9.2.1-1 Table 9.2.1-2 Table 9.2.1-2 Figure 9.2.1-1 Figure 9.2.1-1A Figure 9.2.1-1B 10.2-1 and 10.2-2 10.2-1 and 1G.2-2 Volume 13 i i 14.1-17 and 14.1-18 14.1-17 and 14.1-18 l l Volume 15 420-45 l THESE INSTRUCTIONS ARE TO BE FILED IN FRONT OF VOLUME 1 l O 2 Revision 20 - November 1980 1

SNPS-1 FSAR

/                LIST OF AMENDMENTS TO THE APPLICATION b}                         OPERATING LICENSE STAGE Amendment        Amendment Number           Date                Content Amendment 13     January 1976        Operating License Applicatzon Amendment 14      March 1976          Revision 1 to FSAR Amendment 15      September 1976     Revision 2 to FSAR Amen &nent 16     October 1976        Revision 1 to ER Amendment 17      November 1976       Revision 2 to ER Amendment 18     November 1976        Revision 3 to FSAR Amendment 19      February 1977       Revision 4 to FSAR Amendment 20     March 1977           Revision 5 to FSAR Amendment 21      May 1977            Revision 6 to FSAR

(~ Amendment 22 August 1977 Revision 7 to FSAR - (,- Amendment 23 September 1977 Revision 8 to FSAR Amendment 24 October 1977 Revision 3 to ER Amendment 25 December 1977 Revision 9 to FSAR Amendment 26 February 1978 hevision 10 to FSAR Amendment 27 June 1978 Revision 11 to FSAR Amendment 28 July 1978 Revision 12 to FSAR Amendment 29 September 1978 Revision 13 to FSAR Amendment 30 December 1978 Revision 14 to FSAR Amendment 31 January 1979 Revision 15 to FSAR Amendment 32 April 1979 Revision 16 to FSAR Amendment 33 September 1979 Revision 4 to ER Amendment 34 September 1979 Revision 17 to FSAR I) V Amendment 35 January 1980 Revision to License Application Amendment 36 June 1980 Revision 18 to FSAR AM-1 Revision 18 '- June 1980

SNPS-1 FSAR LIST OF AMENDMENTS 'IO THE APPLICATION OPERATING LICENSE STAGE (CONT

D)

Amendment Amendment Number Date Content Amendment 37 Septembe: 1980 Revision 19 to FSAR Amendment 38 November 1980 Revision 20 to FSAR l O O AM-2 Revision 20 - November 1980

i SNPS-1 FSAR i LIST OF EFFECTIVE PAGES VOLUME 1 Page, Table (T), Revision or Figure (F) Number AM-1 18 AM-2 20 EP 1-1 through EP 1-3 20 EP 2-1 and EP 2-2 20 EP 3-1 and EP 3-2 12 EP 4-1 13 , EP 5-1 thtough EP 5-4 20 EP 6-1 through EP 6-3 20 > EP 7-1 through EP 7-4 19 EP 8-1 19 EP 8-2 through EP 8-4 16 EP 9-1 and EP 9-2 20 l EP 10-1 and EP 10-2 17 EP 11-1 through EP 11-4 20 ! EP 12-1 17 EP 12-2 and EP 12-3 18 EP 12-4 16 EP 13-1 through EP 13-6 20 l s_ EP 14-1 16 EP 14-2 through EP 4-5 19 EP 15-1 17 . EP 15-2 and EP 15-3 19 EP 15-4 20 EP EP-1 16 EP EP-2 15 i 7 11 through vii 12 1-1 1 1-2 Orig. 1-3 1 1-4 3 1 1-1 14 1.1-2 2 1.2-1 through 1.2-20 Orig. 1.2-21 and 1.2-22 2 1.2-23 through 1.2-28 Orig-1.2-29 8 1.2-30 through 1.2-36 Orig. F 1. 2 . 2 -1 16 F1.2.2-2 Orig. 1.3-1 1 1.3-2 Orig. T1.3.1-1 (6 sheets) 5 g' , T1. 3 .1 -2 (2 sheets) Orig.. T1.3.1-3 (sheet 1 of 2) Orig. T1.3.1-3 (sheet 2 of 2) 20

                .                              EP 1-1     Revision 20 - November 1980

SNPS-1 PSAR VOLUME 1 (Cont) Page, Table (T), Revision or Ficure (F) Number T 1. 3 .1 -4 (sheet 1 of 3) 16 T 1. 3 .1 -4 (sheets 2 and 3 of 3) Orig. T1.3.1-5 (4 . sheets) Orig. T1.3.1-6 Orig. T1.3.1-7 (sheet 1 of 2) 1 T1.3.1-7 (sheet 2 of 2) Orig. T1.3.2-1 (sheets 1 through 4 of 17) Orig. T1.3.2-1 (sheets 5 through 17 of 17) 20 1.4-1 through 1.4-4 Orig. 1.5-1 through 1.5-3 Orig. T1.5.1-1 (sheets 1 through 5 of 7) 3 T1.5.1-1 (sheet 6 of 7) 10 T1.5.1-1 (sheet 7 of 7) 3 1.6-1 Orig. T1.6-1 (11 sheets) 10 1.7-1 Orig. F1.7-1 A and F1.7-1B 16 F1.7-2A 16 F1.7-23 Orig. F1.7-3 11 F 1. 7 -4 Orig. 2-1 through 2-6 19 2-7 and 2-S Orig. 2-9 through 2-13 1 2-15 and 2-16 3 2-17 10 2-18 and 2-18a 3 2-19 19 2-20 1 2-21 through 2-22a/b 10 2-23 1 2-24 and 2-25 7 2.1-1 Orig. 2.1-2 1 2.1-3 through 2.1-25 19 T2.1.3-1 through T2.1.3-4 19 T2.1.3-4A 19 T2.1.3-5 (4 sheets) 19 T2.1.3-5A (4 sheets) 19 T2.1.3-4 (2 sheets) 19 T2.1.3-7 (2 sheets) 19 T2.1.3-B (3 sheets) 19 T2 .1. 3 -9 19 F2.1.1-1 and F2.1.1-2 Orig. F2.1.3-1 through F2.1.3-12 19 2.2-1 through 2.2-5 Orig. 2.2-6 and ?.2-7 2 EP 1-2 Revision 20 - November 1980

    . . . .. ~    .. . - _ _ . .               -.               - - _ . - . _            . - - . -                _ - .               . . . -     , ~ .   -        .- --

l .I l' SNPS-1 FSAR i VOLUME 1 (Cont) ) l

Page, Table (T) , Revision j or Fiqure (F) Number l I

2.2-8 1 2.2-9 2 2.2-10 1 i 2.2-11 and 2.2-12 Orig. T2.2.1-1 (2 sheets) Orig. T2.2.1-2 Orig. i F2.2.1-1 and F2.2.1-2 Orig. i 1 i 1 i ) e I I 4 i . f I 1 i i; EP 1-3 Revision 20 - November 1980

SNPS-1 FSAR LIST OF EFFECTIVE PAGES VOLUME 2 Page, Table (T) , Revision or Figure (F) Number i 7 11 through vii 12 2.3-1 and 2.3-2 3 2.3-3 and 2.3-4 4 2.3-5 through 2.3-7 3 2.3-8 through 2.3-12a/b 4 2.3-13 through 2.3-20 3 T2.3.1-1 Orig. T2.3.1-2 and T2.3.1-3 3 T2.3.1-4 through T2.3.1-6 Orig. T2 . 3 .1 -7 3 T2.3.2-1 through T2.3.2-9 Orig. T2.3.2-10 through T2.3.2-14 1 T2.3.2-15 through T2.3.2-17 Orig. T2.3.2-18 and T2.3.2-19 1 T2.3.2-20 and T2.3.2-21 Orig. T2.3.2-22 1 T2.3.2-23 through T2.3.2-26 Orig. T2.3.2-27 and T2.3.2-2 8 1 Os_, T2 . 3 .2 -2 9 Orig. T2.3.2-30 1 T2.3.2-31 through T2.3.2-36 Orig. T2.3.2-37 and T2.3.2-38 1 T2.3.2-39 through T2.3.2-44 Orig. T2.3.2-45 and T2.3.2-46 1 T2.3.2-47 through T2.3.2-56 Orig. T2.3.2-57 through T2.3.2-152 1 T2.3.2-153 through T2.3.2-168 3 T2.3.3-1 Orig, T2.3.3-2 (2 sheets) Orig. T2.3.4-1 through T2.3.4-12 Orig. T2.3.5-1 Orig. F2.3.1-1 Orig. F2.3.2-1 through F2.3.2-7 Orig. F2 . 3 . 2 -8 1 F2.3.2-9A through F2.3.2-9H 1 2.4-1 through 2.4-16 10 2.4-17 20 2.4-18 through 2.4-33 10 T2 . 4 .1 -1 10 T2 .4 .5 -1 Orig. T2.4.'1-1 Orig. 1 T2 .4 . '. '-2 20 ' T2.4.12-1 (4 sheets) 1 4 O T2.4.13-1 Orig. EP 2-1 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 2 (CONT e p} O Page, Table (T) , Revision or Fiqure (F) Number T2.4.13-2 (5 sheets) Orig. T2.4.13-3 and T2.4.13-4 Orig. F2.4.1-1 Orig. F2.4.2-1 Orig. F2.4.2-2A through P2.4.2-2C 16 F2.4.2-3A and F2.4.2-3B 16 F2.4.2-4 16 F2.4.2-5A and F2.4.2-5B 16 F2.4.2-6A through F2.4.2-6D Orig. F2.4.2-7A and F2.4.2-7B 16 F2.4.5-1 and F2.4.5-2 10 F2.4.5-3 16 F2.4.5-4A through F2.4.5-4E 10 F2.4.8-1 10 F2.4.8-2 16 F2. 4. 8 -3 A 10 F2.4.8-3B 16 F2.4.8-3C and F2.4.8-3D 10 F2.4.8-4 and F2.4.8-5 10 F2.4.12-1 Orig. F2.4.13-1 through F2.4.13-5 Orig. 1 EP 2-2 Revision 16 - April 1979 l

St:PS-1 FSAR LIST OF EFFECTIVE PAGES VOLUME 5 Page , Table (T) , Revision or Figure (F) Number _ i 7 11 through vii 12 3-1 and 3-2 Orig. 3-3 10 3-4 and 3-5 14 3-6 and 3-7 Orig. 3-8 1 3-9 16 3-10 8 3-11 lb 3-12 and 3-12a/b 14 3-13 1 3-14 and 3-15 8 3-16 14 3-16a/b 10 3-17 and 3-18 1 4 3-19 and 3-20 3 3-20a/b 4 3-21 16 3-22 10

 \       3-23 and 3-24                                       3 3.1-1 througn 3.1-11                              Orig.

3.1-12 3 3.1-13 through 3.1-33 Urig. 3.1-34 3 3.1-35 and 3.1-36 Orig. 3.1-37 11 3.1-38 througn 3.1-44 vrig. 3.1-45 20 3.1-46 through 3.1-54 Orig. 3.1-55 and 3.1-5o 8 3.1-57 Orig. 3.2-1 and 3.2-2 Orig. T3.2.1-1 (sheet 1 of 21) Orig. T3.2.1-1 (sneet 2 ot 21) 4 T3.2.1-1 (sheet 3 of 21) 19 T3.2.1-1 (sheets 4 through 7 or 21) Orig. 2 T3 . 2 .1 -1 (sneet 8 or 21) 6 T3.2.1-1 (sheets 9 through 11 of 21) Orig. T3.2.1-1 (sheets 12 through 17 of 21) 16 ! T3.2.1-1 (sheets 18 through 20 of 21) Orig. T3.2.1-1 (sheet 21 of 21) 6 T3.2.1-2 (sheet 1 of 2) Ori,. T3.2.1-2 (sheet 2 of 2) 16 p T3.2.1-3 Orig. F3.2.2-1 Orig. 3.3-1 Orig. EP 5-1 Revision 20 - November 1980

I l l SUPS-1 FSAR VOLUME 5 (CONT *D) Paue, Table (T) , or Figure (F) Revision O Number 3.3-2 and 3.3-ea/b 10 3.J-3 through 3.3-5 3 3.4-1 and 3.4-2 10 3.4-3 16 T3.4.1-1 10 F3.4.1-1 througn F3.4.1-3 10 F3.4.1'4 10 3.5-1 and 3.5-2 1 3.5-3 through 3.5-44/b 10 3.5-5 through 3.5-7 3 3.5-8 10 T3.5.1-1 10 T3.5.3-1 and T3.5.3-2 Orig. l T3.5.4-1 Orig. 3.b-1 through 3.6-43 14 l T3.6.1A-1 14 T3.6.4A-1 Orig. F3.6.1A-1 Orig. F3.6.4A-1 through F3.6.4A-22 Orig. l F3.6.5A-1 through F3.6.5A o Oiig. F3.6.5A-7 3 F3.6.5A-8 through F3.6.5A-12 Orig. F3.6.5A-13 and F3.6.5A-14 14 F3.6.5A-15 through F3.b.5A-18 Orig. l F3.6.5B-1 Orig. l 3.7-1 1 3.7-2 and 3.7-3 Crig. 3.7-4 and 3.7-4a/b 1b 3.7-5 enrough 3.7-3 Orig. 3.7-10 a nd 3.7-10a/b 1 3.7-11 and 3.7-12 Orig. 3.7-13 and 3.7-14 2 3.7-15 and 3.7-16 Orig. 3.7-17 and 3.7-18 16 3.7-19 Orig. 3.7-20 and 3.7-21 1 3.7-22 Orig. 3.7-23 3 3.7-24 through 3.7-28 Orig. l 3.7-29 and 3.7-30 lb i 3.7-31 through 3.7-52 Orig. I 3.7-53 1 3.7-54 through 3.7-56 Orig. T3.7.1A-1 Orig. T3.7.2A-1 Orig. T3.7.2A-2 1 T3.7.2A-3 (2 sheets) 2 T3.7.2A-4 through T3.7.2A-6 2 T3.7.3A-1 through T3.7.3A-5 Orig. l EP 5-2 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 5 (CONT ' D) Page, Table (T) , Revision or Figure (F) Number T3.7.13-1 Orig. T3.7.2B-1 and T3.7.2B-2 Orig. T3.7.3B-1 and T3.7.3B-2 Orig. F3.7.1 A-1 through F3.7.1A-5 Orig. F3.7.2A-1 through F3.7.2A-11 Orig. F2.7.2A-12 through F3.7.2A-14 2 F3.7.3A-1 througn F3.7.3A-5 Orig. F3.7.2B-1 and F3.7.2B-2 Orig. F3.7.2B-3A and F3.7.2B-3B Orig. F3.7.38-1 Orig. 3.8-1 10 3.8-2 and 3.b-3 Orig. 3.8-4 1 3.8-5 and 3.8-6 16 3.b-7 through 3.8-11 Orig. 3.8-12 through 3.8-14b 1 3.8-15 through 3.8-17 Orig. 3.8-18 14 3.b-19 and 3.8-20 Orig. 3.u-21 3 3.8-22 4 3.8-224 and 3.8-22b 19 O- 3.8-22c/d 4 3.0-23 Orig. 3.8-24 16 3.8-25 through 3.8-28 Orig. 3.8-29 through 3.8-30a/b 10 3.8-31 and 3.8-32 Orig. 3.8-33 3 3.u-34 through 3.8-38 Orig. 3.u-39 through 3.8-40a/o 2 3.d-41 Orig. 3.8-42 and 3.8-42a/b 18 3.8-43 through 3.8-52 Orig. 3.8-53 and 3.8-54 1 T3.8.1-1 (4 sheets) Orig. T3.8.1-2 througn T3.8.1-6 Orig. T3.8.1-7 (2 sheets) 16 T3.8.1-8 through T3.8.1-13 1 T3.8.5-1 3 F3.8.1-1 through F3.8.1-6 16 F3.8.1-7 through F3.8.1-10 Urig. F3 . u .1 -1 1 16 F3.8.1-12 14 F3.8.1-13 and F3.8.1-14 Orig. F3.8.1-15 16 F3.u.1-16 through F3.8.1-18 Orig, ()/ s, F3.8.1-19 through F3.8.1-26 1 F3.8.1-27A and F3.8.1-27B 4 EP 5-3 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 5 (CONT *D) Page, Table (T) , Revision or Figure (F) Number F3.6.3-1 16 F3.8.3-2 through F3.8.3-6 Orig. F3 . 8 . 3 -7 16 F3.8.3-d 14 F3.8.3-9 and F3.d.3-10 Orig. l l l l O l l l l l l l l l EP 5-4 Revision 20 - NovemDer 1980

SNPS-1 FSAR LIST OF EFFECTIVE PAGES VOLUME 6 Page, Table (T) , Revision or Figure (F) Number i 7 11 through vii 12 , F3.8.4-1 through F3.8. 4- 7 16 F.3.8.4-8 20 F.3.8.4-9 through F.3.8.4-19 16 F3.8.4-20 through F3.8.4-22 Orig. F3.8.4-23 1o F3.8.4-24 through F3.8.4-28 Orig. F3.8.4-29 16 F3.8.4-30 2 F3.8.4-31 10 F3.8.5-1 through F3.8.5-5 Orig. 3.9-1 1 3.9-2 through 3.9-4 Orig. 3.9-5 and 3.9-6 16 3.9-7 through 3.9-8a/b 1 3.9-9 through 3.9-16 Orig. 3.9-17 through 3.9-20 10 O 3.9-21 through 3.9-26 3.9-27 Orig. 10 T3.9.1A-1 and T3.9.1A-2 Orig. T3.9.2A-1 (2 sheets) Orig. T3.9.2B-1 (3 sheets) Orig. T3.9.2B-2 (4 sheets) Orig. T3.9.2B-3 and T3.9.2B-4 Orig. T3.9.2B-5 (2 sheets) Orig. T3.9.2B-6 (15 sheets) Orig. T3.9.2B-7 (5 sheets) Orig. T3.9.2B-8 (3 sheets) Orig. T3.9.2B-9 (2 sheets) Orig. T3.9.2B-10 Orig. T3.9.2B-11 (2 sheets) Orig. T3.9.2B-12 Orig. T3.9.2B-13 (2 sheets) Orig. T3.9.2B-14 (2 sheets) Orig. l T3.9.2B-15 (2 sheets) Orig. T3.9.2B-16 (2 sheets) Orig. T3.9.2B-17 (2 sheets) Orig. T3.9.2B-18 (3 sheets) Orig. l T3.9.2B-19 (2 sheets) Orig. T3.9.2B-20 (2 sheets) Orig. T3.9.2B-21 Orig. l

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T3.9.2B-22 (3 sheets) Orig. l F3.9.1A-1 through F3.9.1A-11 Orig. F3.9.1B-1 and F3.9.1B-2 Orig. 3.10-1 and 3.10-2 Orig. l EP 6-1 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 6 (CONT

D)

Pcge, Table (T) , Revision or Figure (F) Number 3.10-3 through 3.10-16 1 T3 .10.1B-1 (10 sheets) 8 T3.10.2B-1 (2 sheets) 1 T3.10.2B-2 (3 sheets) 1 T3.10.2B-3 1 T3 .10. 2B -4 (sheets 1 and 2 of 4) 1 T3.10.2B-4 (sheet 3 of 4) 8 T3.10.2B-4 (sheet 4 of 4) 16 F3.10.2B-1 through F3.10.2B-9 1 3.11-1 8 3.11-2 Orig. 3.11-3 8 3.11-4 and 3.11-5 1 3.11-6 through 3.11-13 8 T3.11.1-1 (3 sheets) 8 T3.11.1-2 (3 sheets) 8 T3.11.2-1 (O'eet 1 of 3) Orig. T3.11.2-1 (sheet 2 of 3) 8 T3.11.2-1 (sheet 3 of 3) Orig. T3.11.2-2 1 3.12-1 through 3.12-10 Orig. 3.12-11 16 3.12-12 through 3.12-15 Orig. T3.12.3-1 and T3.12.3-2 Orig. T3.12.3-3 (2 sheets) Orig. F3.12.3-1 through F3.12.3-9 Orig. Appendix 3A Title Page Orig. 3A-i 14 3A-ii throcgh 3A-v Orig. 3A-1 through 3A-23 Orig. 3A-24 through 3A-28 14 T3 A-1/2 Orig. T3A-3/4 Orig. T3A-5/6 Orig. T3A-7/8 Orig. T3A-9/10/11 Orig. T3 A-12/13 Orig. T3 A-14/15 Orig. T3A-16/17/18 Orig. T3A-19 Orig. T3A-20 Orig. T3A-21 (2 sheets) Orig. T3A-22 through T3A-25 Orig. F3A-1 through F3A-24 Orig. Appendix 3B Title Page Orig. 3B-1 and 3B-2 Orig. 3B-3 and 3B-4 4 3B-5 3 EP 6-2 Revision 20 - November 1980

SNPS-1 FSAR VOLUwI 6 (CONT ' D) Page, Table (T) , Revision gy Fiqure (F) Eunber 3B-6 17 33-7 4 33-8 +.hrough 3B-10a/b 14 33-11 and 3B-12 4 33-13 20 l 33-14 4 33-15 7 33-16 through 33-20 4 33-21 and 33-22 16 Appendix 3C Title Page 14 3C-i through 3C-viii la 3C.1-1 14 3C.2-1 and 3C.2-2 la F3C.2-1 through F3C.2-13 14 3C.3-1 through 3C.3-22 14 T3C.3-1 (2 sheets) 14 T3C.3-2 through T3C.~3-7 14 T3C.3-8 (2 sheets) 14 T3C.3-9 and T3C.3 ',0 14 F3C.3-1 14 O 3C.4-1 through 3C.4-28 T3C.4-1 through T3C.4-7 14 14 T3C.4-8 (2 sheets) 14 F3C.4-1 through F3C.4-14 la F3C.4-15A through F3C.4-15C 14 F3C.4-16 14 3C.5-1 through 3C.5-9 14 T3C.5-1 14 3C.6-1 through 3C.6-4 14 T3C.6-1 and T3C.6-2 14 Attachment 3C.1 Title Page la A3C.1-1 through A3C.1-6 14 Attachment 3C.2 Title Pace 14 A3C.2-1 through A3C.2-10 14 FA3C.2-1 14 Attachment 3C.3 Title Page 14 A3C.3-1 14 Attachment 3C.4 Title Page '14 A3C.4-1 through A3C.4-10 la FA30.4 -1 14 i o EP 6-3 Revision 20 - November 1980

SLPS-1 FSAR LIST OF EFFECTIVE PAGES VOLUME 9 Page, Table (T), Revision

or Figure (F) Number

! i 7 ii through vii 12 7.3-1 through 7.3-11 Orig. 7.3-12 16 7.3-13 through 7.3-21 Orig. 7.3-22 and 7.3-23 16 7.3-24 through 7.3-27 5 7.3-28 16 7.3-29 through 7.3-34 Orig. 7.3-35 through 7.3-40 1 7.3-41 and 7.3-4 2 Orig. 7.3-43 through 7.3-50 16 7.3-51 20

       '_3-52 and 7.3-52a/b                                            16 7.3-53 through 7.3-54a/b                                          3 7.3-55 through 7.3-56b                                          16 7.3-57                                                         Orig.

7.3-58 5 7.3-59 through 7.3-66 3 O' 7.3-67 and 7.3-68 Orig. 7.3-69 through 7.3-70a/b 3 7.3-71 through 7.3-82 Orig. 7.3-83 and 7.3-84 5 7.3-85 Orig. 7.3-86 through 7.3-88 5 7.3-89 through 7.3-100 Orig. 7.3-101 through 7.3-110 3 7.3-111 through 7.3-113 Orig. 7.3-114 through 7.3-116a/h 3 7.3-117 through 7.3-120 1 7.3-121 and 7.3-122 16 7.3-123 1 7.3-124 and 7.3-124a/b 3 7.3-125 3 7.3-126 and 7.3-127 16 7.3-128 through 7.3-135 1 7.3-136 through 7.3-139 16 T7.3.1-1 through T7.3.1-3 Orig. T7 . 3 .1 -4 5 T7.3.1-5 through T7.3.1-7 Orig., T7.3.1-8 5 T7.3.1-9 (2 sheets) Orig. T7.3.1-10 Orig.

/~]

(_/ T7.3.2-1 (sheet 1 of 2) T7.3.2-1 (sheet 2 of 2) 16 3 EP 9-1 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 9 (CONT 'D) Page, Table (T) , Revision or Figure (F) Number F7.3.1-1A and F7.3.1-1B Orig. F7.3.1-2A through F7.3.1-2E Orig. 47.3.1-3 5 F7.3.1-4 Orig. F7.3.1-5A and F7.3.1-5B , Orig. F7 . 3 .1 -6 Orig. F7.3.1-7A and F7.3.1-7B Orig. F7.3.1-8A through F7.3.1-8D Orig. F7.3.1-9A through F7.3.1-9D 1 F7.3.1-10A through F7.3.1-10E Orig. F7.3.1-11 through F7.3.1-15 Orig. F7.3.1-16A through F7.3.1-16C 16 F7.3.1-17A through F7.3.1-17H 16 F7.3.1-18A and F7.3.1-18B 16 F7.3.1-19A through F7.3.1-19D 16 F7.3.1-20A through F7.3.1-20H 16 F7.3.1-20J and F7.3.1-20K 16 F7.3.1-21 Orig. F7.3.1-22A through F7.3.1-22H 16 F7.3.1-22J 16 F7.3.1-23A through F7.3.1-23H 16 F7.3.1-23J through F7.3.1-23L 16 F7.3.1-24A through F7.3.1-24H 16 F7.3.1-24J through F7.3.1-24M 16 F7.3.1-25 16 F7.3.1-2 6A through F7.3.1-26F 16 F7.3.1-27A and F7.3.1-27B 16 F7.3.1-28A through F7.3.1-28F 16 l F7.3.1-29A through F7.3.1-29H 20 l F7.3.1-29J through F7.3.1-29N 20 l F7.3.1-29P through F7.3.1-29AD 20 i F7.3.1-30 through F7.3.1-34 10 l F7.3.1-35A through F7.3.1-35F 16 l l 1 l I l EP 9-2 Revision 20 - November 1980 l l

SNPS-1 FSAR LIST OF EFFECTTVE PAGES (} VOLUME 11 Page, Table (T) , Revision or Fiqure (F) Number i 7 11 through vii 12 8-1 through 8-4 17 8.1-1 through 8.1-4 17 T8.1.4-1 (2 sheets) 17 T8 .1. 6 -1 17 T8.1.7-1 (sheet 1 of 4) 3 T8.1.7-1 (sheets 2 through 4 of 4) 17 8.2-1 through 8.2-9 17 F8.2.1-1 through F8.2.1-3A 17 F8.2.1-4 7 F8.2.1-5A and F8.2.1-5B 7 F8.2.1-6 17 F8.2.1-7 Orig. F8.2.1-8A and F8.2.1-8B 7 F8.2.1-9 through F8.2.1-11 17 8.3-1 through 8.3-45 17 T8.3.1-1 (sheet 1 of 4) 20 (s~' T8.3.1-1 (sheet 2 of 4) 17 T8.3.1-1 (sheets 3 and 4 of 4) 20 T8.3.1-2 (sheet 1 of 2) ;7 T8.3.1-2 (sheet 2 of 2) 20 T8.3.1-3 and T8.3.1-4 3 T8.3.1-5 and T8.3.1-6 17 T8.3.1 -7A and T8.3.1-7B 17 T8 .3 .2 -1 (3 sheets) 17 T8.3.2-2 Orig. F8.3.1-1 through F8.3.1-8 17 F8.3.1-9 1 F8 .3 .1 -10 17 F8.3.1-11 and F8.3.1-12 6 F8.3.2-1 through F8.3.2-6 17 F8.3.2-7A through F8.3.2-7C Orig. 9-1 20 l 9-2 Orig. 9-3 through 9-5 16 9-6 8 9-7 through 9-9 16 9.1-1 and 9.1-2 7 9.1-3 through 9.1-121/j 8 9.1-13 Orig. 9.1-14 and 9.1-14a/b 3 9.1-15 through 9.1-21 Orig. l () T9.1.2-1 T9.1.3-1 (2 sheets) 8 8 EP 11-1 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 11 (Cont) l'a g o , Table (T) , Revision or Piqure (F) Number T9 .1. 4 -1 Orig. T9.1.4-2 (16 sheets) 16 F9.1.1-1 Orig. F9.1.2-1 through F9.1.2-6 8 F9.1.3-1 and F9.1.3-2 16 F9.1.4-1 through F9.1.4-3 Orig. F9.1.4-4 A and F9.1.4-4B Orig. F9.1.4-5A and F9.1.4-5B Orig. F9.1.4-6 through F9.1.4-8 Orig. F9.1.4-9 and F9.1.4-10 3 l 9.2-1 through 9.2-6b 20 9.2-7 16 9.2-8 Orig. 9.2-9 16 0 9-10 through 9.2-13 Orig. 9 .2 -14 16 9.2-15 through 9.2-22 Orig. 9.2-23 through 9.2-26 16 l T9.2.1-1 and T9.2.1-2 20 T9.2.2-1 16 T9.2.7-1 Orig. l F9.2.1-1A and F9.2.1-1B 20 F9.2.2-1A and F9.2.2-1B 16 F9.2.3-1A and F9.2.3-1B 16 F9.2.4-1 and F9.2.4-2 Orig. F9.2.5-1 16 F9.2.7-1A and F9.2.7-1B 16 F9. 2.P.-1 A and F9.2.8-1B 16 F9.2.9-1 16 9.3-1 through 9.3-3 16 9.3-4 Orig. 9.3-5 and 9.3-6 16 9.3-7 through 9.3-10 Orig. T9.3.2-1 (sheet 1 of 4) Orig. T9.3.2-1 (sheets 2 through 4 )f 4) 16 T9.3.3-1 Orig. FP.3.1-1A through F9.3.1-1E 16 F9.3.2-1A through F9.3.2-1D 16 F9.3.3-1A through F9.3.3-1D 16 9.4-1 16 9.4-2 3 9.4-3 through 9.4-4a/b 16 9.4-5 Orig. 9.4-6 1 9.4-7 and 9.4-8 Orig. 9.4-9 16 9.4-10 and 9.4-10a/b 2 EP 11-2 Revision 20 - November 1980

SNPS-1 FSAR [^ > VOLUME 11 (Cont) N. , Page, Table (T) , Revision or Figure (F) _ _ _ Number 9.4-11 through 9,4-13 Orig. 9.4-15 1 9.4-16 Orig. 9.4-17 and 9.4-18 1 9.4-19 and 9.4-20 Orig. 9.4-21 through 9.4-23 16 9.4-24 19 9.4-25 and 9.4-26 16 F9 .4 .1 -1 16 F9.4.1-2 3 F9.4.2-1 16 F9.4.3-1A and F9.4.3-1B 16 F9 . 4 . 4 -1 16 F9.4.5-1 16 F9.4.6-1 16 F9.4.7-1 16 F9.4.8-1A and F9.4.8-1B 16 F9.4.8-2 and FP :.8-3 16 F9.4.9-1 16 F9.4.10-1 16 ('") F9.4.11-1 9.5-1 16 18 9.5.2 and 9.5-3 Orig. 9.5-4 and 9. 5-4a/b 18 9.5-5 through 9.5-8 Orig. 9.5-9 1 9.5-10 Orig. 9.5-11 16 9.5-12 through 9.5-16 Orig. 9.5-17 16 9.5-18 Orig. 9.5-19 16 0.5-20 and 9.5-21 Orig. 9.5-22 14 9.5-23 and 9.5-24 Orig. 9.5-25 through 9.5-28a/b 3 9.5-29 Orig. 9.5-30 and 9.5-30a/b 3 9.5-31 Orig. 9.5-32 through 9.5-34 2 9.5-35 through 9.5-38a/b 16 9.5-39 Orig. 9.5-40 18 9.5-41 1 9.5-42 through 9.5-45 10 T9.5.9-1 Orig. F9.5.1-1 and F9.5.1-2 16 EP 11-3 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 11 (Cont) Page, Table (T) , Revision or Fiq'2re (F) Number F9 . 5 .1 -3 19 F9.5.1-4 16 F9.5.4-1A and F9.5.4-1B 16 F9 . 5 . 5 -1 Orig. F9.5.6-1 Orig. F9.5.7-1 18 Appendix 9A Title Page 8 9A-1 through 9A-10 8 T9A-1 through T9 A-3 8 T9A-4 (2 sheets) 8 F9A-1 through F9A-7 8 10-1 through 10-4 3 10.1-1 and 10.1-2 Orig. T10.1-1 (sheet 1 of 3) Orig. T10.1-1 (sheet 2 of 3) 16 T10.1-1 (shect 3 of 3) Orig. F10.1-1 through F10.1-4 16 10.2-1 Orig. 10.2-2 20 10.2-3 through 10.4 -7 Orig. 10.2-8 16 10.2-9 through 10.2-11 3 T10.2.3-1 Orig. F10.2.3-1 through F10.2.3-4 Orig. 10.3-1 16 10.3-2 Orig. F10.3.2-1 16 10.4-1 through 10-4-3 Orig. 10.4-4 through 10.4-6 16 l l 10.4-7 1 l 10.4-8 and 10.4-9 Orig. l 10.4-10 16 10.4 -11 through 10.4-14 Orig. 10.4-15 through 10.4-16a/b 1 10 .4 -17 Orig. 10.4-1R 19 10.4-19 Orig. T10.4.1-1 and T10.4.1-2 Orig. T10.4.1-3 (2 sheets) Orig. T10.4.6-1 and T19.4.6-2 Orig. F10.4.2-1 16 F10.4.3-1 16 F10.4.5-1 16 F10.4.6-1A <nd F10.4.6-1B 16 F10.4.7-1 14d F10.4.7-2 16 O l EP 11-4 Revision 20 - November 1980

SNPS-1 FSAR LIST OF EFFECTIVE PAGES VOLUME 13 Page, Table (T) , Revision or Figure (F) Number 1 7 11 through vii 12 14-1 and 14-2 Orig. 14-3 19 14-4 17 14-5 and 14-6 18 14-7 and 14-8 6 14.1-1 and 14.1-2 Orig. 14.1-3 10 14.1-4 and 14.1-5 Orig. 14.1-6 19 14.1-7 and 14.1-8 Orig. 14.1-9 12 14.1-10 7 14.1-11 Orig. 14.1-12 7 14.1-13 2 14.1-14 Orig. , O,s 14.1-15 andough 14.1-16 4 14.1-17 20 14.1-18 4 14.1-19 throur,a 14.1-29 Orig. 14.1-30 through 14.1-32a/b 4 14.1-33 and 14.1-34 Orig. 14.1-35 and 14.1-36 19 14.1-37 and 14.1-38 Orig. 14.1-39 and 14.1-40 19 14.1-41 through 14.1-43 Orig. 14.1-44 10 14.1-45 through 14.1-48 19 14.1-49 and 14.1-50 11 14.1-51 through 14.1-55 Orig. 14.1-56 4 14.1-57 through 14.1-58d 6 14.1-58e 16 14.1-58f through 14.1-581 6 14.1-58j 7 14.1-59 and 14.1-60 7 14.1-61 through 14.1-65 Orig. 14.1-66 and 14.1-67 19 14.1-68 and 14.1-69 Orig. 14.1-70 through 14.1-72 17 14.1-73 Orig. 17 O' 14.1-74 through 14.1-78 14.1-79 through 14.1-80a/b 19 14.1-81 through 14.1-88 17 EP 13-1 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 13 (CONT 8D) Page, Table (T) , Revision or Figure (F) Number 14.1-89 through 14.1-90a/b 18 14.1-91 through 14.1-92c/d 17 14.1-93 Orig. 14.1-94 through 14.1-96 18 T14.1.1-1 (sheet 1 of 2) 17 T14.1.1-1 (sheet 2 of 2) 18 P14.1.1-1 through F14.1.1-3 Orig. F14.1.4-1 17 14.2-1 through 14.2-4 Orig. 14.3 -1 Orig. P14.3.1-1 4 15-1 16 15-2 5 15-3 16 15-4 5 15-5 12 15-6 16 15-7 5 15-8 8 15-9 5 15-10 16 15-11 5 15.1-1 through 15.1-2a/b 17 15.1-3 through 15.1-5 5 l 15.1-6 16 15.1-7 through 15.1-10 5 15.1-11 through 15.1-13 16 15.1-14 and 15.1-15 5 15.1-16 and 15.1-17 16 15.1-18 8 15.1-19 5 15.1-20 through 15.1-22 16 15.1-23 through 15.1-28 5 15.1-29 and 15.1-30 16 l 15.1-31 and 15.1-32 5 15.1-33 16 l 15.1-34 and 15.1-35 5 15.1-36 16 15.1-37 through 15.1-50 5 15.1-51 through 15.1-59 16 15.1-60 5 15.1-61 16 15.1-62 5 15.1-63 16 15.1-64 through 15.1-90 5 l 15.1-91 through 15.1-94a/b 8 15 1-95 through 15.1-97 5 15.1-98 6 EP 13 -2 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 13 (CONT'D) Page, Table (T) , Revision or Fiqure (F) Number 15.1-99 through 15.1-114 5 15.1-115 17 T15.1-1 (sheets 1 and 2 of 3) Orig. T15.1-1 (sheet 3 of 3) 5 T15.1-2 Orig. T15.1-3 (2 sheets) 16 T15.1.1-1 and T15.1.1-2 16 T15.1.2-1 and T15.1.2-2 16 T15.1.4-1 16 T15.1.5-1 5 T15.1.7-1 16 T15.1. 8-1 and T15.1.8 -2 5 T15.1.10-1 5 T15.1.11-1 5 T15.1.14-1 (2 sheets) Orig. T15.1.16-1 and T15.1.16-2 5 T15.1.17-1 (2 sheets) Orig. T15.1.18-1 16 16 (' N-T15.1.19-1.ind T15.1.19-2 T15.1.20-1 and T15.1.20-2 T15.1.21-1 16 5 T15.1.21 -2 16 , T15.1.21-3 11 T15.1.22-1 16 T15.1.24-1 5 T 15.1.25-1 5 T15.1.76-1 (2 sheets) Orig. T15.1.27 -1 (2 shtets) Orig. T15.1.30-1 (2 sheets) 2 T15.1.30 -2 2 T15.1.31-1 and T15. , .31-2 Orig. T15.1.31-3 and T15.1.31-4 1 T15.1.33-1 (2 sheets) Orig. T15.1.33-2 and T15.1.33-3 Orig. T15.1.34-1 (2 sheets) 8

T15.1.34-2 Orig.

l T15.1.34-3 8 T15.1.34-4 2 T15.1.35-1 through T15.1.35-3 Orig. T15.1.36-1 (2 sheets) Orig. T15.1. 3 6-2 an't T15.1.3 6-3 Orig. T15.1.37-1 and T15.1.3'l-2 Orig. T15.1.38-1 and T15.1.38-2 Orig. F15.1-1 8 ("N F15.1-2 5

 \           F15.1.1-1 and F15.1.1-2                        16 F15.1.2-1 and F15.1.2-2                        16 F15.1.4-1                                      16 I

EP 13-3 Revision 20 - November 1980

SNPS-1 FSAR VOLUMP. 13 (CONT

D1 Page, Table CP) , Revision or Figure (F)_ Number ,

F15.1.5-1 5 F15.1.7-1 16 F15.1.8-1 and F15.1.8-2 5 F15.1.10-1 5 F15.1.11-1 through F15.1.11-5 5 F15.1.16-1 5 F15.1.18-1 16 F15.1.19-1 and F15.1.19-2 16 F15.1.20-1 and F15.1.20-2 16 F 15.1. 21 - 1 16 F15.1.22-1 16 F 15.1. 2 4 -1 5 F15.1.25-1 5 F15.1.30-1 and F15.1.30-2 2 Appendix 15A Title Page 12 15A-i 16 15A-ii 12 15A .iii and 15A-iv 16 15A-v 12 15A-vi 16 & 15A-vii 12 W 15A-1 through 15A-2a/b 17 15A-3 and 15A-4 12 15A-5 16 15A-6 through 15A-9 12 15A-10 and 15A-11 16 15A-12 and 15A-13 12 15A- 14 and 15A-15 16 15A-16 and 15A-17 11 15A-18 through 15A-20 16 15A-21 through 15A-26 12 15A-27 and 15A-28 16 15A-29 and 15A-30 12 15A-31 16 15A-32 through 15A-43 12 15A-44 16 15A-45 12 15A-46 through 15A-48 16 15A-49 and 15A-50 12 15A-51 through 15A-56 16 15A-57 through 15A-63 12 15A-64 17 T15A.1-1 (3 sheets) 12 T15A.1-2 12 T 15A .1 -3 (2 sheets) 16 T15A.1.1-1 and T15A.1.1-2 16 T15A.1. 2-1 and T15A.1. 2-2 16 T15A.1.4-1 16 EP 13-4 Revision 20 - November 1980

I SNPS-1 FSAR VOLUME 13 (CONT so) Page, Table (T) , Revision or Fiqure (F) Number T15A.1.5-1 12 T15A .1.7 -1 16 T15A.1.8-1 and T15A.1.8-2 12 T15A.1.10-1 12 T15A.1.11-1 12 T15A.1.16-1 and T15A.1.16-2 12 T15A.1.18-1 16 T* SA.1.19-1 and T15A.1.19-2 16 T15A.1.20-1 and T15A.1.20-2 16 T15A.1.21-1 12 T15A.1.21-2 16 T15A.1.21-3 12 T15A.1.22-1 16 T15A.1.24-1 12 T15A.1.25-1 12 F15A.1-1 and F15A.1-2 12 F15A.1.1-1 and F15A.1,1-2 16 F15A.1.2-1 and F15A.1.2-2 16 F 15A.1.4 -1 16 F15A.1.5-1 12 16 O.- F15A.1.7-1 F15A.1.8 -1 and F15A.1.8-2 12 F15A.1.10-1 12 F15A.1.11-T through F15A.1.11-5 12 F15A.1.16 *. 12 F15A.1.18-1 16 F15A.1.19-1 and F15L.1.19-2 16 F15A.1.20-1 and F15A.1.20-2 16 F15A.1.21-1 16 i F15A.1.22-1 16 l F15A.1.24-1 12 F15A,1.25-1 12 Chapter 16 Title Page Orig. 17-1 through 17-5 17 17.1-1 and 17.1-2 17 l 17.1-3 Orig. 17.1-4 17 1 17.1-5 through 17.1-6a/b 19 17.1-7 through 17.1-55 17 , T17.1-1 17 T17.1.5A-1 (7 sheets) 17 l T 16 .1. 6 A- 1 17 T17.1.2B-1 (2 sheets) Orig. T17.1.2B-2 (8 sheets) 17 T17.1.2B-3 (sheet 1 of 6) 17 T 17.1. 2B-3 (sheet 2 of 6) 13 . O' T17.1.2B-3 (sheets 3 through 6 of 6) 17 ( T17.1.2B-4 (4 sheets) 13 l EP 13-5 Revision 20 - November 1980

SNPS-1 FSAR VOLUME 13 (CONT *D) g Page, Table (T) , Revision or Figure (F) Number T17.1.2B-5 (2 sheets) 17 T17.1.2B-6 13 T17.1.18B-1 17 T17.1C-1 17 F17.1-1 17 F17.1.1A-1 17 F17.1.1B-1 and F17.1.1B-2 17 F17.1.1D-1 17 F17.1.1D-2 19 17.2-1 through 17.2-27 17 T17.2.5-1 (5 sheets) 17 T17.2.5-2 (5 sheets) 17 T17.2.6-1 (2 sheets) 17 F17.2.1-1 19 l O EP 13-6 Revision 20 - November 1980

i SNPS-1 FSAR () VOLUME 15 (CONT 'D) Page, Table (T) , Revision or Figure (F) Number 324-3 through 324-5 1 T324.5 (2 sheets) 1 324-6 14 324-7 7 T324.7-1 and T324.7-2 7 F324.7 7 324-8 through 324-8c 7 T324.8-1 and T324.8-2 7 T324.8-3 (3 sheets) 7 T324.8-4 7 F324.8-1 through F324.8-5 7 324-9 through 324-9g 7 T324.9-1 7 F324.9 7 324-10 and 324-10a 7 324-11 and 324-11a 7 F324.11 7 324-12 and 324-12a 7 T324.12-1 (4 sheets) 7 T324.12-2 (2 sheets) 7 O> F324.12-1 through F324.12-5 7 324-13 7 T324.13-1 (4 sheets) 7 T324.13-2 (5 sheets) 7 324-14 through 324-15c 7 T324.15-1 (3 sheets) 7 T324 15-2 (3 sheets) 7 F324.15-1 through F324.15-7 7 324-16 through 324-16b 7 T324.16 7 F324.16 7 324-17 through 324-17c 7 T324.17-1 (2 sheets) 7 T324.17-2 and T324.17-3 7 F324.17-1 and F324.17-2 7 324-18 7 T324.18 7 331-1 through 331-5 1 331-6 through 331-15 2 331-16 through 331-23 4 331-24 through 331-25b 18 331-26 through 331--28a 18 F331.28 's8 331-29 through 331-31a 18 ("N 371-1 through 371-7b 10 (_) 371-8 through 371-15 412-1 and 412-2 10 1 412-3 through 412-6 6 EP 15-3 Revision 19 - September 1980

SNPS-1 FSAR VOLUME 15 (CONT *D) Page, Table (T), Revision or Figure (F) Number 412-7 through 412-12a 2 412-13 through 412-15 4 412-16 14 413-1 through 413-6b 4 413-7 through 413-7b 4 413-8 through 413-12a 4 413-13 and 413-13a 4 413-14 through 413-19 4 413-20 6 413-21 through 413-25 4 413-26 and 413-26a 6 413-27 and 413-27a 6 413-28 through 413-31 6 413-32 7 420-1 through 420-4 4 420-5 2 420-6 through 420-8 4 420-9 and 420-10 2 420-11 through 420-13a 4 420-14 through 420-16 4 420-17 2 420-18 through 420-34a 4 420-35 through 420-36a 4 420-37 and 420-38 4 420-39 through 420-39b 18 420-40 and 420-41 18 420-42 through 420-44 19 420-45 20 422-1 through 422-7 3 422-8 and 422-9 11 423-1 18 430-1 2 431-1 through 431-10 2 431-11 and 431-12 4 432-1 and 432-2 2 432-3 4 O EP 15-4 Revision 20 - November 1980

CNPS-1 FSAR TABLE 1.3.1-3 (CONT'D) Cooper , Shoreham Hatch Station Isrowns Ferry B. AUXILIARY SYSTEMS (CONT'D) Unit 1 Unit 1 Unit 1 Units 1 and 2 Service %8ater System Flow Rate (qpm/ pump) 8,600 8,000 4,000 4,500 , Number of Ptanps sta) g a g l Reactor Core Isolation Coolinq system Flow Rate (gpm) 400 at 400 at 416 at 616 at 1,120 paid 1,120 paid 1,200 psid 1,120 psid Fuel Pool Coolino and Cleanup System Capacity of Cooling Subsystem (Stu/hr) 7.0 x 10* 5.7 x IO* 3.4 x 106 8.8 x 106 i

    '. Capacity during reactor flooding mode with three or four pumps running.

can Safety related pumps; there are three additional nonsafety related pumps. l b i I i t h n n 2 of 2 Revision 20 - November 1980 L i

-e

SNPS-1 FSAR

   /~h                      TABLE 1.3.2-1   (CONT *D)

U Significant Changes Since PSAR FSAR Section

17. Fuel pool cooling and cleanup system 9.1.3 has been changed from a cumbined system to two separate and independent systems.

The cooling system is safety related. This provides total independence and flexibility in operation for cooling and cleanup functions.

18. An additional set of pumps has been 9.2.2 provided in tne reactor building closed loop cooling water system to supply cooling water to the M-G set fluid coupling coolers. Placement of the M-G sets outside of the reactor building made separate cooling provision manda-tory. Also the system has 4 loops instead of 2, and essential equipment is safety related. This precludes exit and return of safety related ex piping through the reactor building.

k 19. The separate station service water 9.2.1 and RHR service water systems were rearranged into separate reactor building and turbine building subsystems. The four reactor building service water pumps provide cooling for normal and emergency reactor building heat loads including the RHR heat exchangers. These pumps also provide emergency diesel generator cooling and cooling requirements for RBSVS and CRAC. The remaining three pumps are devoted ex-clusively to nonsafety turbine building heat loads. This arrangement provides a simplified operational sequence.

20. The maximum number of fuel assemblies 9.1.1 the new fuel storage vault can hold was changed from 180 to 170. Also, the center to center spacing between racks was changed to 11 1/2 in. from 11 in.

O V 5 of 17 hevision 20 - November 1980

SNPS-1 FSAR TABLE 1.3.2-1 (CONTap) O Significant Changes Since PSAR FSAR Section

21. The reactor building crane capacity 9.1.4 was changed from 100 ton to 125 ton to enable using larger spent fuel casks. The reactor building crane will De a dual load path crane.

22. Portions of the reactor building 6.2.3 stand-by ventilation system (RBSVS) 9.2.9 and reactor building normal ventila- 9.4.2 tion system (RBNVS) , previously sepa-rate systems, have been combined:

Exhaust ducts and exhaust fans are now common to both systems. Redundant sets of dampers are provided for mode conversion Revisions to the RBNVS include the use of common tilters, heating coils and fans for the supply D air, and a common exhaust system

 \~sJ          for the retueling level and the lover leveln of the secondary containment.

The commonality of equipment and ducting provides flexibility of operation, simplification for isolation, elimination of a large secondary containment penetration at the refueling level and reduction in space and equipment requirements. Recirculation rate for the RBSVS l increased from 30,000 cfm to

45,000 cfm. The increased flow rate improves the building mixing and provides additional dilution ror the discharge.

, The RBSVS charcoal filter exnaust air rate has been increased from 700 cfm to 1160 cfm. The increased flow rate ensures better control 3 of the negative pressure in the l ,i reactor building during RBSVS l operation. 6 of 17 Revision 20 - November 1980 l

SNPS-1 FSAR TABLE 1.3.2-1 (CONT

D) v Significant Changes Since PSAR FSAR Section A chilled water system has been added to the RBSVS. This system is also used to provide chilled water to the control room air conditioning system. The RBSVS and control room a.c. chilled water system ensures a posi-tive heat sinx for sensible and latent heat in the reactor building. It pro-vides for maintaining the secondary containment at the design negative pressure following the postulatea DBA.

It also provides cooling for the control room during nor-mal and design basis accident conditions.

      . 23. The following changes have been                  9.5.1 made to the station fire protec-
  %            tion system:

Focm solutions will not be used in order to eliminate foam cleanup and disposal problems. The personnel passage way (which contains some non-safety related cable trays) between the reactor and turbine buildings will not be serviced with carbon dioxide in order to eliminate asphyxiation danger to personnel. Smoke detectors for the primary containment have been deleted. Reliability is gained by use of temperature elements, since smoke detectors are unreliable in a radiation field. Fixed water spray fire protection for the turbine high pressure oil piping under the operating 4 floor has been deleted. High {~Ns_,) pressure piping is run inside the gravity return line and 7 of 17 Revision 20 - November 1980

SNPS-1 FSAR TABLE 1.3.2-1 (CONT *D1 l Significant Changes Since PSAR FSAR Section the areas are protected by automatic sprinklers.

24. Station equipment and floor 9.3.3 drainage systems - all turbine 11.2 building floor drainage, except drains carrying oil from selected areas, are routed to radwaste.

This provides a means to monitor the drainage and treat it, if necessary.

25. Primary containment purge exhaust - 9.4.9 charcoal filters have been added to die primary containment purge exhaust to provide the capability for the removal of iodines.

26. Station potable and sanitary water 9.2.4 systems - decontamination shower 9.3.3 and laundry room wastes will not 11.2

/ ) be directed to a holdup tank for \l monitoring and discharge to radwaste or sanitary sewer. They now go to floor drainage for direc-tion to radwaste only. This prevents the possibility of conta-minated discharge to the sanitary sewer.

27. Main control room ventilation - 9.4.1 charcoal and Hepa filters have been increased in capacity and the arrange-ment altered to filter outside air and a portion of the recirculated air in lieu of filtering only the outside air.

This reduces the potential doses to personnel in the control room for the period during which occupancy is required subsequent to a DBA.

28. Containment seismic analysis - the 2.5 soil spring constant has been revised 3.7 to 13,000 psi 225 percent. Additional soil boring data permitted this modification.

(j q_ S

29. Quality group and code classifications -

numerous enanges have been made to 3.2 upgrade the SNPS-1 design to comply 8 of 17 Revision 20 - Novemoer 1980

SNPS-1 FSAR ( ( TABLE 1.3.2-1 (CONT *D) Significant Changes Since PSAR FSAR Section with Regulatory Guides 1.26 and 1.29 and 10CFR50.55a.

30. Seismic analysis - the method 3.7 for calculating amplified response spectra has been based on the time history method to comply with recent NRC regulatory positions.

31. Quality assurance program - numerous 17 changes have been made to upgrade the quality assurance program to 10CFR50 Appendix B and recent Regulatory Guides.

32. Instrument lines - restricting 6.2.4 orifices were added to instrument lines and the rege trement for remote operation of shutoff valves was f)/

x_, deleted in complying with Regula-tory Guide 1.11. The excess flow check valves will pass a flow of between 0.7 and 0.7 gpm at 1000 psi differenti. to assure opening after isolation for system resto-ration. The excess flow check valves will have remote position indication.

33. 138 KV generator circuit breakers 8.2 and connections to the switchyard -

the main step-up transrormer's power is transmitted to the 138 kV switch-yard via two 3000 amp circuit breakers and a double circuit transmission line (in lieu of one 4000 amp circuit breaker and one transmission line as previously described) . The two circuits are bussed together prior to terminating at the r'dtch-yard. The use of two circus'. breakers instead of one "1)I nat ' result in complete shu+dt'q the unit as would occur Af uw-tenance on a single brec.Aer were i -w required. C/ 9 of 17 Revision 20 - November 1980 l

SNPS-1 FSAR TABLE 1.3.2-1 (CONT'D) ('N)

 \_J Significant Changes Since PSAR                         FSAR Section

34. Electrical loads - loads have been 8.3 revised and redistributed as a re-sult of various system changes.

35. 480 V switchgear emergency busses - 8.3 bus ties and automatic trans-f ers between 480V emergency busses have been eliminated to comply with Regulatory Guide 1.6.

36. 480 V switchgear circuit breakers - 8.3 circuit breakers are equipped with static trip devices, instead of electromechanical trip devices, for better protection and coordination of low voltage system. Static trip devices are more accurate than electro-mechanical trip devices and provide ready means of incorporating ground f ault protection in addition to

(/) s_ phase overcurrent protection.

37. 125 V d-c station battery system - 8.3 the station battery system was changed from a two to three battery system. Each battery has its own ch; ger and distribucion board.

              . -'_smatic transfer of direct current load groups from one battery to the other is eliminated to comply with Regulatory Guide 1.6.

Battery chargers have sufficient capacity to restore the battery from the design minimum charge to its full charge in a maximum of 24 hr while supplying normal and post cacident steady state loads to comply with Regulatory Guide 1.32. i 38. 120 V a-c uninterruptible power 8.3 supply - the 120/240 V single phase 60 Hz uninterruptible power for vital instruments and controls are r^ provided by static inverters in-( ,N) stead of dual drive motor generator sets. Two static inverters are pro-10 of 17 Revision 20 - November 1980

SNPS-1 ESAR f TABLE 1.3.2-1 (CONT *D) \-- Significant Changes Since PSAR FSAR Section vided, one for v$tal instrumenta-tion, controls, and essential con-trol room lighting, and the other for the process computer. Static inverters have several advantages as compared to rotating (motor gene-rator) types. They are smaller, lighter, easier to install and maintain, have higher overall efficiency, and give better voltage and frequency response to transients.

39. Standby diesel generator system - 8.3 9.5.4 The rating of each diesel generator was increased from 2850 Kw to 3500 Kw.

[~h x-The day tank capacity of each unit is 550 gal which will provide fuel requirements for approximately 2 hr (instead of the 4 hr previously specified) to comply with National Fire Protection Association (NFPA) Code 37-1971, Chapter 5, Section 52. The diesel fuel oil system is safety related and is designed and installed in accordance with the ASME III, Code Class 3 requirements. This action upgrades the fuel oil system with an accompanying improvement in diesel reliabi11ty even though the day tank capacity has been reduced to comply with NFPA Code. The day tank capacity is well above the minimum 1 hr. fuel supply recommendation of the diesel manufacturer. Two 100 percent capacity transfer pumps are pro-vided for automatically filling the day tank from its respective storage tank.

40. 5 kV power cables - All 5 kV power 8.3 f-~g _ cables are single conductor or tri-( i plexed, shielded, with thermosetting,

~#

flame retardant insulation and 11 of 17 Revision 20 - November 1980

SNPS-1 FSAR ' TABLE 1.3.2-1 (CONT

D)

Significant Changes Since PSAR FSAR Section thermosetting, flame retardant jacket over each individual conductor (instead or three conductor, inter-locked, armor cable for tray instal-lation as previously described) . The type of cable used provides mechanical protection and is fire , resisting and retardant. Addition-ally, this construction is more suitable since the majority of the cable runs are partly underground.

41. Main condenser system - materials 10.4.1 of fabrication and size have been modified:

Condenser surface has been in-creased to 420,000 ft2 curing optimization of plant cycle. () Tube sheets are Muntz metal instead of 90/10 Cu-N1. Tubes are Titanium instead ot' 90/10 Cu-Ni. Water boxes are cast iron instead of 90/10 Cu-Ni.

42. Maximum Flood Levels - new maximum 2.4.2 water levels were calculated using 2.4.5 the "Bathystrophic Storm Tide Theory." The recalculation was per-formed at the suggestion of the AEC j staff consultants. Results of the l

recalculation (repor t to AEC, dated l January, 1970) indicate that flood protection to 25 ft above mean low water, as originally planned, is adequate for componenets essential j for safe station shutdown. l 43. Core spray system - the system 3.2 design has been upgraded to the 6.3 1969 General Electric product line.

      )      Piping has been designed to ASME

[~J N III instead of B31.1. 12 of 17 Revision 20 - November 1980

SNPS-1 FSAR TABLE 1.3.2-1 (CONT 'D) (} Significant Changes Since PSAR FSAR Section

44. A rod sequence control system has 4.3.2 been incorporated to render the effects of control rod drop to results acceptable to NRC. This limits the result of a control rod drop to peak fuel enthalples less than 280 cal /g.

45. The fracture toughness capabilities 5.2.4 of the pressure vessel are evaluated in accordance with the requirements of Paragraph G-2115 (for normal heatup and cooldown condition) and G-2410 (f or inservice hydrostatic testing) ot Appendix G of the Summer 1972 addenda to Section III of ASME Boiler and Pressure Vesu Code, 1971 edition. Proper heatt, and cooldown rates plus operating

(-^S limitation during startup and ( ,/ shutdown will be established.

46. Separation of wiring and switches 3.12 in panels involving the reactor 7.2 protection system and engineered 7.3 safety features are designed to ensure that common mode failure will not occur. Testability during operation of the systems is provided.

47. Combustible gas control in con- 6.2.5 tainment- systems to control the combustible gas concentration in the primary containment tollowing a LOCA have been incorporated to comply with the NRC Containment Systems Branch Technical Position CSB6-2.

48. The inservice inspection program S.2.8 will comply with ASME Section XI, Summer 1972 addenda.

49. The feedwater sparger design has been 4.3

   )      improved to reduce power flux                   4.4

(/ w asymmetries. 13 of 17 Revision 20 - November 1980

SNPS-1 FSAR i TABLE 1.3.2-1 (CONT *D) O Significant Changes Since PSAR FSAR Section

50. ADS control power supply - the 7.3 automatic power transfer circuits for ADS have been deleted. Each ADS cabinet has its own individual and separate d-c control power feed.

51. Reactor water cleanup system - auto- 5.2.7 matic isolation of RWCU on area high 5.5.8 temperature /high flow has been incorpor- 7.6.1 ated to ensure automatic isolation in case of pipe break in the RNCU system.

52. ECCS performance has been evaluated 6.3 in accordance with 10CFR50.46.

53. RCIC and HPCI system elbow flow 7.3.1 devices have been located to ensure 7.4.1 that main steam flow does not affect proper functioning of the elbow flow devices.

54. The RCPB leakage detection system 5.2.7 has been upgraded to comply with 7.6.2 General Design Criteria 30 and Regulatory Guide 1.45.

55. Protective system instrumentation 7.1.2 and engineered safety features are 7.2.2 constructed to meet the requirements 7.3.2 or IEEE-279, 1971 edition (10CFR50.55a). 7.4.2 7.5.2 7.6.2

56. Strong motion accelerographs, peak 3.7.4A recording accelerographs, and a peak deflection recorder will be in-stalled to properly monitor and record any seismic disturbance in accordance with Regulatory Guide 1.12.

57. Safety and relief valves are sized 3.9.2A and installed, in the number required, to reflect ASME III and General Electric design criteria.

() 58. The instrument air system is designed such that its loss will not affect 9.3.1 14 of 17 Revision 20 - November 1980

SNPS-1 FSAR

   s                       TABLE 1.3.2-1  (CONT 8D)

O Significant Changes Since PSAR FSAR Section the proper functioning of engi-neered safety features. Air-operated valves will fail to a safe position or be operated by accumulators on loss of instrument air.

59. The control room will be provided 7.2.2 with an indication and annunciation 7.3.2 when any protective action bypass 7.4.2 occurs.

60. Failure to SCRAM (ATWS) - Recirculation 15.1.27 4

pump trips have been implemented as a backup to SCRAM.

61. A bottom drain temperat te monitor has been installed to s . low control of the reactor vessel bottom head temperature, thus preventing over-stressing of the bottom nead.

x/ 62. Test jacks and test lights are 7.3.1 installed to tacilitate testing and to indicate the operational status of ECCS systems to ensure that adequate ECCS are available when needed.

63. Condenser loss of vacuum protection - 7.2 low vacuum scram was removed and 10.4.1 low vacuum isolation is installed to provide an improved plant pro-tection system. This requires closure of MSLIVs on low vacuum.

64. Flexible type SS connectors are used to ensure against fatigue failure of reliet valves with air lines.

65. Discharge of safety valves is piped to 5.2.2 the suppression pool.

66. MSL isolation logic was redesigned to prevent possible water hammer damage resulting from a single violation of

(~'S an operating procedure. L.) 15 of 17 Revision 20 - November 1980

SNPS-1 FSAR s TABLE 1.3.2-1 (CONT

D)

Significant Changes Since PSAR FSAR Section

67. The control circuits for the standby 7.4.1.2 liquid control system pump starters are arranged so that taking one pump out for service will not disable the other pump.

68. The logic for isolation of reactor 7.3.1 water sample lines has been changed to permit easier sampling of reactor water so that it can be analyzed before venting of the drywell. The only

    ,         required 1 solation signals are low reactor level and high steam line radiation and these will be maintained.

69. A solid state reactor manual control 7.7.1 system has been used. This provides the capability for automatic on-line testing.

O \m,/ 70. The LPCI loop selection logic has been 7.3.1 modified so that single failure will not prevent the LPCI function.

71. No nitrogen strengthened grades 5.2.3 of stainless steel (i .e . , 304 N and 316 N) are used.

72. mine fill pumps have been added to ECCS 6.3 and RCIC to maintain pump discharge lines full to prevent water hammer curing operation.

73. The turnine building gaseous radwaste 3.2 treatment area is designed and analyzed (not designated) as Seismic Category I. This ensures the integrity of Category I structures adjacent to non-Category I structures.

74. The fuel design has been changed from a 4.1 7 x 7 array to an 8 x 8 array in order 4.2 to improve performance and thermal margin. 4.3 Fuel channel thickness increased trom 4.4 80 to 100 mils.

I \ 16 of 17 Revision 20 - November 1980

                 , . .                              .         .~.        ,

SNPS-1 FSAR s TABLE 1.3.2-1 (CONT 'D) O Significant Changee Since PSAR FSAR Section

75. A decoupler will be installed between 5.5.1.4 the recirculat3'n pump and motor to eliminate the recirculation pump missile hazard due to suction line breaks.

76. A Prompt (Pressure) Relief Trip system 7.6.1.4 has been incorporated to mitigate the 15.1.1 consequences of overpressure during 15.1.2 ,

end-of-cycle turbine trip and load rejection incidents.

77. A remote shutdown panel has been 7.4.1.4 added to provide the capability to 7.5.1.4 shut down the reactor from outside the main control room.

78. The low population zone has been 2.1.3 reduced from 5 to 2 mi.

79. A second personnel hatch has been 3.8.1 installed in the drywell (in the O)

\_ center of the equipment hatch) to comply with OSIIA requirements (to provide a second egress). 17 of 17 Revision 20 - November 1980

SNPS-1 FSAR Jefferson, N.Y. indicates a return period of 50 years for the C1 lowest observed water elevation. 2.4.11.4 Puture Control Consideration of future control of the cooling water source is unnecessary since the plant uses Long Island Sound. The use of water from the Sound by future users will not affect the cooling water supply because of the abundance of water available. 2.4.11.5 Plant Requirements The required minimum safety related cooling water flow is 12,800 gpm supplied by two service water pumps; the maximum required total station service water flow during normal operation is 42,765 gpm, supplied by three reactor building (RB) service water pumps and two turbine building (TB) service water pumps. The circulating water system supplies a maximum of 573,500 gpm through four pumps. The service water system, circulating water system, and the ultimate heat sink are described in detail in Sections 9.2.1, 10.4.5, and 9.2.5, respectively. The location of the service water pumps within the screenwell is described in Section 9.2.1. Figure 3.8.4-8 shows the configuration of the screenwell and the sump elevation to be -21 m1w. The centerline elevation of the lower stage impeller of the ( RB service water pumps is at approximately -14.9 mlw. -Table l 2.4.11-2 lists the net positive suction head (NPSH) available for the delivered flow at the RB service water pumps for normal l operation and loss of coolant conditions at three water elevations. The elevation of -3.5 mlw corresponds to the minimum recorded elevation of the Long Island Sound at the site. l 2.4.11.6 Heat Sink Dependability Requirements i Since the plant uses Long Island Sound as the cooling water ! source, there are two types of events which could potentially endanger the water supply, an extremely low tide or an accident at the intake canal. The jetties and the accident considerations l are described in Sections 2.2, 2.4.8.1, 2.5.5, and 9.2.5. The historical low water at Port Jefferson is -3.5 m1w and the computed low water elevation due to the effects of the PMH at the i site is -5.9 m1w. The plant flow requirements and the available l RB service water pump NPSH for various submergence conditions are i listed in Table 2.4.11-2. The table shows that the RB service ( water pumps have sufficient submergence to operate during the low l water elevation of -5.9 m1w caused ny the PMH. l Since the source of the plant water for fire pngtection is groundwater, the water availability is unaffected by the short term water elevation fluctuations in Long Island Sound. Three wells, each equipped witha 250 gpm pump, supply all the fresh-water. needs of the plant and are described in Section 2.4.13.1. Fire water is stored in two 350,000 gal tanks. The fire ! protection system is described in Section 9.5. l 2.4-17 Revision 20 - November 1980

SNPS-1 FSAR 2.4.12 Environmental Acceptance of Effluents The sink for accidental releases of liquid effluents is Long laland Sound. There are essentially two pathways for accidentally spilled liquids to the Sound: liquid seepage to the groundwater, and liquid flow through the diffuser discharge system by way of the storm drains to the intake canal. Any liquid which would reach the groundwater on the site would not affect groundwater users, since the groundwater enters the marshes on the east and west of the site and Long Island Sound on the north without intersecting wells of existing users. Groundwater flow and use are discussed in Section 2.4.13. Accidentally released liquid reaching the marshes and Long Island Sound would affect those who use this water for recreation or food sources. This section discusses only the general aspects of accidental liquid releases, such as methods of computation, travel time, dilution, dispersion, and ion exchange of radioactive isotopes in the soil. Any accidentally released liquid from tanks which are constructed to contain liquid at an elevation above the surrounding grade elevation, or from partly or completely underground tanks constructed adjacent to an ;.ea with a grade elevation less than the elevation of the liquid surf ace, has the potential to enter the intake canal through the storm drains. The accidentally released liquid would mix with the cooling water and travel through the circulating water and cooling systems of the plant to enter Long Island Sound through the diffuser. The circulating water system is described in Section 10.4.5. The ratio of the flow of water entrained by the diffuser to the water released through the diffuser, called the uilution ratio, was oF tained from model studies (67) under conditions of varying tidal currents and current reversal. The results of the study are shown on Fig. 2.4.12-1, which shows the variation of the dilution ratio in the near field, based on a 6.5 acre surface mixing zone, over the tidal cycle. The dilution ratio, which varies from a minimum of 12 to a maximum of about 29, depends on the strength of the tidal currents and on the tidal level. If a liquid spill were to enter the drains on the east side of the radwaste buil ding, the dilution ratio of the flow in the intake canal to the flow in the drain would be 32. Accidentally released liquid which would enter the groundwater in the vicinity of the condensate storage tanks or the radwaste building, shown on Fig. 9.2.6-1, would eventually reach Long Island Sound by seeping into the salt marsh and Wading River Creek on the east of the plant without affecting any users of groundwater. In order to make a general assessment of the behavior of accidentally released liquid, groundwater flow from the location of the condensate storage tanks was analyzed. The equations and parameters used in the assessment of the effects of accidental release of radioactive liquid to the environment are contained in Table 2.4.12-1. 2.4-18 Revision 10 - February 1978

SNPS-1 FSAd Table 2.4.11-2 s SUBMr.Rtif.NCE CONDITIONS OF Rt.At~rOR BUILDING SERVICE hATt.k PbMP6 l Loss of Coolant Normal Operation Accident Total flow Required (gpm) 11, tu.5 12, ts 00 humoer or Pumps Operating 2 2 Flow Delivered per Pump (gpm) 5,145 ts ,550 l hequtred hP5H (It) at flow delivered 33 34 l Available NPSH (It) at 0.0 m1w 46.9 46.9

    -3.5 m1w                          43.4                 4s.4
    -5.9 m1w                          41.0                 41.0 O

i t i f 1 or 1 Revision 20 - hovemcer 1980 i 1

SNPS-1 FSAR The service water system provides cooling water for removal of (/ (_,

heat from the structures, systems, and components important to safety during all abnormal and accident conditions. The station service water system supplies cooling water to the RHR neat exchangers, standby diesel generatars, RBSVS and CRAC water chillers, drywell cooling booster heet exchangers, and RBCLCW l heat exchangers. The RBCLCW supplies - pump seal and bearing cooling water to the control rod drive pcmps, CS pumps, and RHR pumps, and tne reactor coolant recirculation motcr pumps and pump coolers. The service water system and portions of the RBCLCW system are designed to Seismic Category I requirements. Redundant safety related components served by these systems are supplied through the redundant supply headers and returned through redundant discharge or return lines. Electric power for operation of redundant safety related components of these systems is supplied from separate independent offsite and redundant onsite standby power sources. No single failure renders thes a systems incapable of performing their safety functions. Referenced sections are as follows:

1. A-c Power Systems 8.3

, 2. Service Water System 9.2 l

3. Reactor Building Closed Cooling Water System 9.2

4. Ultimate Heat Sink 9.2

() 3.1.2.45 Inspection of Cooling Water System JCriterion 45) Criterion l The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system. Design Conformance l The service water and the RBCLCW systems are designed to permit appropriate periodic inspection in order to assure the integrity of system cc ponents. l Referenced sections are as follows: l

1. Service Water System 9.2

2. Reactor Building Closed Loop Cooling 9.2 Water System l

us l l l 3.1-45 Revision 20 - November 1980

SNPS-1 FSAR 3.1.2.46 Testing of Cooling Water System (Criterion 46) Criterion The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the systen, and (3) the operability of the system as a whole and, under conditions as close to design as prretical, the performance of full operational sequence that brings the system into operation for reactor shutdown and for loss of roolant accidents, including operation of applicable portions of the protection systems and the transfer between normal and emergency power sources. Design Conformance The service water and RBCLCW systems are in operation duridh normal plant operation and shutdown. Thus, component performance is continucusly demonstrated. These systems are designed to the extent praticable to permit demonstration of operability of the systems as required for operation during a LOCA or a loss of offsite power. Referenced sections are as follows:

1. Service Water System 9.2

2. Reactor Building Closed Loop Cooling Water System 9.2 3.1.2.47 Criterion 47 This criterion has not yet been promulgated by the NRC.

3.1.2.48 Criterion 48 This criterion has not yet been promulgated by the NRC. 3.1.2.49 Criterion 49 This criterion has not yet been promulgated by the NRC. 3.1.2.50 Containment Design Basis (Criterion 50) Criterion The reactor containment s tructure , including access o;enings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and, with sufficient margin, the calculated pressure and temperature conditions resulting from any loss of coolant 3.1-46

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7*"' FIG. 3.8.4-8 s-e n SCREEN WELL PLANS AND SECTIONS SHOREHAM NUCLEAR POWER; STATION-UNITI FINAL SAFETY ANALYS'3 REPORT PiVISION 20 NOVEMBER 1980

SNPS-1 FSAR

 /~    38-1.52  Design, Testing, and Maintenance Criteria for Atmosphere

(_j} Cleanup, System Air Filtration, and Absorption Units of Light-Water-Cooled Nuclear Power Plants (6/73L The Reactor Building Standby Ventilmtion System filter trains and the Control Room Air Conditioning filter trains have been designed within the guidelines of Regulatory Guide 1.52. l Reference Section 6.2.3 3B-1.53 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems (6/73) The application of the single f ailure criterion to the protection system complies with Regulatory Guide 1.53. Reference Section 7.3.2 3B-1.54 Quality Assurance Requirements for Protective Coatings Apolied to Water-Cooled Nuclear Power Plants (6/73) Quality assurance requirements for protective coatings comply with Regulatory Guide 1.54. l Reference Sections 6.2.1 and 6.3.2 and Chapter 17 The Operational QA Program conforms to the guidan.? provided in [\ -) Regu3atory Guide 1.54 dated 6/73 during the operational phase. 3B-1.55 Concrete Placement in Category I Structures (6/73) Concrete placement in Category I structures complies with Regulatory Guide 1.55 except that all reinforcing shop detall drawings other than for the containment mat, pedestal, primary wall and drywell wall are checked by engineers at the Jobsite. Reference Section 3.8 and Cnapter 17 3B-1.56 Maintenance of Water Purity in Boiling Water Reactors J7/78) The plant design complies with Regulatory Guide 1.56. Operating limits are contained in the technical specifications. Referen'e Sections 5.2.3.4 and 10.4.6 and Chapter 16 3B-1.57 Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components ( 6/731 Regulatory Guide 1.57 is not applicable to this plant, s3nce the primary containment design is not the metal primary rea ctor

 /~    containment type.

( }/ 3B-13 Revision 20 - November 1980

SNPS-1 FSAR 3B-1.58 Qualification of Nuclear Power Plant Inspection, Examin-O ation, and Testing Personnel (8/73) The qualification of personnel performing inspection, examination, and testing to verify conformance of work activities to quality requirements complies with Regulatory Guide 1.p8. Reference Chapter 17 The Operational QA Program conforms to the guidance provided in Regulatory Guide 1.58 dated 8/73 during the operational phase as it pertains to the certification and qualification of personnel who verify conformance of work activities to quality requiremonts. The qualifications of the plant operating organization (on site personnel) concerned with the day to day operation, maintenance, and certain technical services shall be controlled by Regulatory Guide 1.8. 38-1.59 Design Basis Floods for Nuclear Power Plants (8/73) Within the limits of its applicability to coastal floodin s, Regulatory Guide 1.59 has been followed in determining the maximum flood levels. Reference Sections 2.4.2, 2.4.3, and 3.4 h 38-1.60 Design Response Spectra for Seismic Design of Nuclear Power Plants (12/73) The design response spectra were developed several years prior to the issuance of Regulatory Guide 's .60 to standards acceptable at that time. Reference Section 3.7 1 3B-1.61 Damping Values for Seismic Desian of Nuclear Power Plants (10/73) Damping values for the seismic analysis of Category I structures were assumed to be 5 and 7 percent for the Operating Basis Earthquake and Design Basis Earthquake, respectively. Regulatory l Guide 1.61, Table 1, recommends the use of 4 and 7 percent for ! reinforced concrete structures. The mathematical models for l Seismic Category I structures of this plant do not separate the effects of structural damping and subsoil damping. Since soil damping tends to increase the overall damping level of a combined soil-structure model, the use of 5 percent rather than 4 percent damping for the OBE is justifiable in this case. Reference Section 3.7 l l 3B-14 Revision 4 - February 1977

SNPS-1 FSAR the main control room so that the operator may initiate or shut ('-'~) down the PCAC system as the needs of the plant may require. Scheduled tests will verify the operability of the PCAC system and its actuation controls. 7.3.1.8 Service Water System Instrumentation and Controls 7 . 3.1. 8 .1 System Identitication The instrumentation and controls for the reactor building (EB) l service water system are required to function following a loss of coolant accident (LOCA), a loss of offsite power (LOOP), and a LOCA coincident with LOOP. The system design is described in Section 9.2.1. 7.3.1.8.2 Power Sources The power for the safety related instrumentation and controls of each train is supplied from separate standby a-c and d-c buses. The RB service water pumps C and D and their associated discharge l valves are powered from Div. III. 7.3.1.8.3 Equipment Design Control switches and indicating lights for the motors of the service water pumos and the motor operated valves are provided in the main control room. The tour RB service water pumps start (s lD automatically if there is a start signal from the emergency bus l program. Instrumentation in the main control room allows monitoring of the:

1. Discharge header pressure for each loop.

2. Flow to each RRR heat exchanger.

A high radiation alarm is provided in the discharge of each RHR heat exchanger. Automatic indication, accompanied by an audible alarm, is provided in the main control room when the system has been deliberately rendered inoperative by intentional operator action. The motor operated valves used to isolate noncritical portions of the system and the two redundant portions of the system are closed automatically on a LOCA signal or LOP. The service water system logic is shown on Figs. 7.3.1-29A through AD. l 7.3.1.b.4 Environmental Considerations Tae safety related instrumentation and controls of the service water system are designed to remain functional in the O,, environmental conditions as discussed in Section 3.11. 7.3-51 Revision 20 - November 1980

SNPS-1 FSAR 7.3.1.8.5 Operational Considerations The portion of the instrumentation and controls of the service water system used during plant operation is verified for operability by their normal use. The operability of the standby diesel generator controls is proven whenever the standby diesel generators are tested. The remainder of the system actuated by automatic signals is given a full operational test during every retueluag per iod. The operability of the utstr umentation required tor saie shutdown is verified by periodic tests. 7.3.1.9 Containment Spray and Suppression Pool Cooling bubsystem Instrinnentation and Controls 7.3.1.9.1 System Identification The contauunent spray / cooling mode is an integral part or 1he residual heat removal (RHR) system anu is used to aid in redur.ing drywell pressure and mixing contauiment air tollowing a LOCA. The containment spray /cooluig mode is initiated manually atter the LPCI cooling requirements have been satistled. An interlock is provided so that the main control room operator does not Anadvertently initiate containment spray betore LPCI requirements are met. 7.3.1.9.2 Power Sources The power supply to the instrumentation and control of the O contauncent spray / cooling mode of the RliR system 1s taken trom the 120 V a-c instrument buses A and B, which are supplied trom the standby a-c buses. 7.3.1.9.3 tquipment Design With the KilR system in the contaulment spray / cooling mode or operation, the hHR pumps transrer water from the suppression pool through the residual heat exchangers where heat is removed by the service water. The cooled water can be diverted to two redundant suppression chamber spray headers and two redundant drywell spray headers. The spray in the suppression chamber cools any noncond2 nsable gases collected in the free volume above the suppression pool. The spray in the drywell condenses any steam that may exist in the drywell, thereby lowering containment pressure (Figs. 7. 3.1-7A and b, and 7.3.1-uA through D) .

1. Initiating Circuits The containment spray cooling mode is initiated only ny manual action. The system cannot be actuated unless certaul requiretients are met, as described in the paragraph on interlocks.

O 7.3-52 Revision 16 - April 1979

SOURCE CONDITION MONITOR CONTR1 iP41 AMOV03 t A 03 DISCHARGE VALVE [c CLOSED /

R. A. SERVICE WATER SW-S R I m A/D SYSTEM "A" PRESSURE IP41 A P003A i

               , pq                          l           LOW < 40 PS IG       j       [A           START i              -PT 00lA                                       REACTOR BUILDING                 P41      SW SERVICE WATER SYSTEM           -Pl        IP41 AP003A "A" PRESSURE                    00lA CR   START
                                                      / START SIGNAL FROM FIG. T.3.1-190              l ENGINEERED SAFEGUARD 015                                        EMElt. BUS 101 PRGM.

SW-SR IP41(P003A AUTO 86 [MOTORPROTECTION ELECTRICAL LOCKt;JT RELAY OPERATED SW -S R IP41pP003A STOP SIGNAL FROM DIVISION 1 FIG. 7. 3.1-19 A ENGINEtRED SAFEGUAR O6 ENER. BUS 101PLGM.]D 41hP003A ERVICE WATER PUMP ] STOP I IP4t d P003A ) a MOTOR OVERCitRRENT j PB-S R IP414MOV03l A OPEN T .D .

                                                                                                     ?    20 SEC.

P4thP-003A PB-SR 52 PUMP RUNNING IP414 H0V03lA l CLOSE NOTES I.. REACTOR BUILDING SERVICE WATER PUMP IP41 TP003A AND DISCHARGE VALVE g . por _ IP41 &MOV03t A FOR DIVISION I SHOWN. REACTOR BUILDING SERVICE WATER PUMP IP41 A P003C AND DISCHARGE VALVE IP41f MOV031C FOR DIVISION DI SlHILAR. PORTIONS [ 2. VALVE CLOSED POSITION TO BE ADJUSTED TO APPR0XIMATELY 10 DEGREES OPEM. t

 . ACTION                                                                  RESULTANT                                            MONITOR R
                                               -                                                                           /,           GB
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3 R. E. SERVICE WATER PUMP IP41hP003A (NOTE 1) l F I G. i.3.1 - 29 D,F,H,K,W O P0 TO R.E.S6.W. VALVES SE Q - IP41tMOV031A OPEN s' R D R.R. SERVICE WATER PUMP DISCHARGE VALVE IP41NMOV031 A (NOTE 1)

       \       ~

t [GE q , IP4ftMOV035A . v- CLOSE e- G F I G. 7. 3.1 - 29 A ! SERVICE WATER SYSTEM ARE NUCLEAR SAFETY RELATED LOGIC DIAGRAM p l SHOREH AM NUCLEAR POWER STATION-UNIT I I FINAL SAFETY ANALYSIS REPORT [ j' R EV ISIO N 20 - NOVEMBER 1980 1

SOURCE MONITOR CONDIT10M CONTROL A IP4lt MOV031 B 33 (DISCHARGE VALVE CLOSED j SW IP41dP003B \ START A SW D A/D [R..B. SYSTEMSERVICE WATER "B" PRESSURE RSP TRANSFER gpgi gpgi LOW < 40 PSlG NORMAL

     -PT            >                           -PI 001                                        0018 CR                                     SW-SR

[ REACT 9RBUILDING SERVICE WATER SYSTEM IP4ltP0038 "B" PRESSURE START SW-S R IP4lt P0038 AUTO [STARTSIGNALFROM ENGINEERED SAFEGUARD Cl3 DIVISION 11 ENER. BUS 102 PRGM SW IP41(P0038 START SW RSP TRANSFER EMERGENCY NOTOR PROTECTION 86 (ELECTRICALLOCK0UTRELAY86ENER SW IP41 AP0038 STOP Fi g, 7,3,g .19 0 STOP SIGNAL FROM ENGINEERED SAFEGUARD -- O22 DIV I S I ON .IL EMER. BUS 102 PRGM. A

                                                       \    R.8. SERVICE WTR.PP..           SW 51
                                             "              IP414.P003B                     IP41%P0038 STOP (MOTOROVERCURRENT SW-SR IP414 P0038 STOP SW RSP TRANSFER NORMAL NOTES: 1. REACTOR BUILDING SERVICE WATER PlNP IP41V P003B FOR DIVISION 3I SHOWN, REACTOR BUILDING SERVICE WATER PUHP                                REACTOR BUILDING SERVICE 1 IP41 AP003D FOR DIVISION III SIMILAR.

., 2. R$f - REMOTE SHUTDOWN PANEL. PORTIONS ARE NUCL 4

CTION REGULTANT MON'.0R GB

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g l lATER PUMP IP414 P0038 (NOTE 1) FIG. 7. 3.1-29 8 l SERVICE WATER SYSTEM ' l LOGIC DIAGRAM /AR SAFETY RELATED SHOREHAM NUCLEAR POWER STATION-UNIT l l FINAL SAFETY ANALYSIS REPORT t i ! REVISION 20 - NOVEMBER 1980 i

                                 ,m   y. np m              - - ~ .         - - - -                  y-fe

CONDITION CONTROL ACTION SOURCE

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  'l                                                                                SW s                                                                               RSP TRANSFER NORMAL PB-SR 1P4i AMOV031 B OPEN I.Q.

D 20 SEC. - SW-SR IP41AMOV031B OPEN 3 IP4 F$P0038

                                                                    ^               RSP TRANSFER 52                PLMP SUNNING EMERGENCY SW-S R IP41AMOV0318 CLOSE
                                                                                         > NOT

( f PB-SR IP41KMOV031B CLOSE SW RSP TRANSFER NORMAL R.ll. SERVICE WATEI NOTES

1. REACTOR BUILDING SERVICE WATER PUMP IP41-$P0038 DISCHARGE VALVE IP41% MOV0318 SHOWN, REACTORBUILDINGSERVICEWATERPUMPIP41$P003DDISCHARGE VALVE IP41k MOV031D SIMILAR.

2. R$2 - REMOTE SHUTDOWN PANEL.

3. VALVE CLOSED POSITION TO BE ADJUSTER, TO APPR0XIMATELY 10 DEGREES OPER.

{ (

RESULTANT I A E AND D

                                                                                                  /
                  ~

y R O SE q v I P41 ANOV031B OPEN ,, v 3 RE EE ( j AND D J "" = AND ~ s E 'G GE s Q IP414 NOV0318 a g - CLOSE (NOTE 3)

                  ?                                                                           M RE j ca AND            D j B2 PUNP DISCHARGE VALVE IP414 NOV0318 (NOTE 1)

NUCLEAR SAFETY RELATED FI G. 7. 3.1 - 29 C SERVICE WATER SYSTEM LOGIC DIAGRAM y SHOREHAM NUCLEAR POWER STATION-UNIT I FINAL SAFETY ANALYSIS REPORT l-R EVISION 20 - NOVEMBER 1980

SOURCE ICMITOR CONDITION CONTROL ACTION

 .                                                                               gy (NOTE 1)-      IP4lh M0V034A THROTTLE MODE

[G P41 Aq PB -SR I49A G HR HX IElli E034A (NOTES 2,3) IP41AMOV034A lElI} FT [ SERVICE WATER FLOW OPEN O _ A 1p41 fPNI SW \ 034A G P41AMOV034A (NOTE 1) IP4I A MOV034A \ gpgl _ MNT NORMAL MODE [g 034A (VALVEPOSITION PB-SR IP41A MOV034A OVERRIDE [g flG. 7.3.1 - 19B 0CA CORE SPRAY I [IMITIATIONSIGNAL O23 PRESENT _

                 . FIG. 7.3.1-29 A 3 0F     ER                      p    0
                                                                             --9      R IP41 AM0'!034A 33                            (CLOSED                  N AND   l
                                                                  -e-NOTES:

L. REYLOCKED SWITCH - KEY REMOVEABLE IN

          " NORMAL MODE" POSIT 10K.

SW

I

2. VALVE MAY BE THROTTLED BY HOLDING THE PUSHBUTTON IN THE DEPRESSED POSITION L M0 E [G WITH THE MODE SELECTOR SWITCH IN THE PB4R -

l

          " THROTTLE MODE" POSITION UNTil DESIREC j                                                                (N0sES2.3)       IP41 ( MOV034A          i FLOW RATE IS ESTABLISHER.                                              CLOSE              2 .

! 3. THE MOTOR OVERLOAD INTERLOCK IS BYPASSED l WHEN HOLDING THE PUSHBUTTON IN THE 'sw DEPRESSED POSIT 10R. (NOTEl) IP41i MOV034A I 4. PUCYJUTTON USED TO ESTABLISH FLOW RATE THROTTLE MODE [G i REQUIRED DURING HOT STANDBY OR MORMAL ! OPERATION WITH SUPPRESSION POOL COOLING. PB-SR l 5. IP411MOV034A (NOTES 3,tj) HOT STBY/ POOL COOL t 6. # BY G.E. CO. SW

 .                                                              (NOTE 1)         IP41AMOV034A
                                                                                                  \

1 NORMAL MODE [G ( I

RESULTANT WONITOR 9, AND ) D- IP414 NOV034A D NOT ? AND --9. THROTTLE OPEN 9-AND

g. MOT _ _9, C -

W IP414MOV034A G GE 49 MOT ? AND --9 OPEN R AND ? 0 e GR 9 _ O NOT -@

        -9. NOT -9,R                                        s s

d AND NOT

                                            ?-
                                                                     ?

IP41 AMOV034A OR CLOSE G

GR

4' MOT D IP41$MOV034A

                                                            ? MOT    D  AND  --D'    INTERMEDIATE     A POSITION               gg 9

y AND 9 IP414 MOY034A AND ? THROTTLE CLOSE 9 NUCLEAR SAFETY RELATED 9' AND FIG. 7.3.1-29 D SERVICE WATER SYSTEM 9' LOGIC DIAGRAM RESIDUAL HEAT REMOVAL h4 E034A PWM MM-N i OUTLE7 VALVE IP4t

H0V034 A FINAL SAFETY ANALYSIS REPORT ,

REVISION 20 - NOVEMBER 1980

4 CONDITION CONTROL Acil0p SOURCE MON' TOR I y

 ,                                                           (NOTE 1)     IP41(MOV0348             p MORMAL MODE         GE AND Pgi                                      PB-SR m                                         IP41fMOV0348 OPEN y

IEli 1998 GR HR MX 1 lit, 0348 [GB

   #                                         SERVICE WATER EFT                                                         (NOTES 2.3) 0068                             FLOW P41
                               $PMI IP41               0348 GE       P41 A MOV0348 APNT                          ,  V.tLVE POSITION 0348

( FIG. l.3.1-1 G B ERGENCY M 102 LOSS OF POWER

                                                                 }                                               D  0 O2                                                   Y.                                             _p  R IP41( MOV0348 33                            CLOSED                .                              p AND  -

P8 -SR IP41%MOV0348 y@ OVERRIDE

                                                                                                 -D-0CA CORE SPRAY 2 T                                       AND       C  0

[IMITIATIONSIGNAL ' D 4 Oll PRESENT C g i 4NOT NOTES: L. KEYLOCKED - REY REWVEABLE IN *MORMAL MODE" POSITION. IP41 MOY0348 ? OPEN

2. VALVE MAY BE THROTTLED BY HOLDING THE

[GE AND PUSH 8UTTON IN THE DEPRESSED POSITION $y WITH TME DODE SELECTOR SwlTCH IN THE IP41/sMOV0348

                                                                                           \       D             A
        " THROTTLE MDDE" POSITION UNTil DESIRED FLOW RATE IS ESTABLISHED.

(NOTE 1) NORMAL MODE

                                                                                           /CE AND

3. THE MOTOR OVERLOAD INTERLDCK IS BYPASSED Ps.SR \

WMEN HOLDING THE PUSMBUTTON IN THE DEPRESSED POSIT 104. IP41 E MOY0348 CLOSE h4

4. PUSH 80TTON USED TO ESTABLISH FLOW RATE D REQUIRED DURING H0T STA4D81 OR NORMAL gw OPERATION WITH SUPPRESSION POOL COOLING. (NOTE l} IP41tMOV0348 -

AND THROTTLE IODE GR gy RESID, 6 I BY 6.E. CO. P'a TRANSFER NORMAL R$f ( (

RE S ULT ANT MONITOR RSP TRANSFER ' D8ERGENCY Uf AND --e _ Sw IP41E N0V0348 R OMN Mf IP41f M0V0348 U D' OR j v . 0PE N "' V

                                    ?  MT           U                                                                     R uf AND 4 D  NOT          D Sw RSP TRANSFER             @

NORMAL RSE -9 ' Sw (NOTE 1) IP414 MOV0348 r--- NORMAL NODE G AND -9 IP41 AMOY0348 PB-SR AhD D INTERMEDi&TE - A (NOTES 354) IP41R MOV0358 - POSITION G HOT STBY / POOL COOL G y D NOT > AND + :) ~ IP41iMOV0348 + NOT 4 NOT > AND THROTTLE OPEN AND "I Sw I P41 A MOV0348 O CLOSE UP SND D Sw RSP TRANSFER -

                                                                                                                        - G EMERGENCY            ME                                   IP41( MOV034B                    /       G OR     CLOSE                    -o I                             ~

Of v AND --E) s= IP4tf MOV034B

                                              \                 ?
                           ==t                /ue IP41IlMOV0348 O  THROTTLE CL0$E JAL HEAT REMOVAL HI E0348 OUTLET VALVE IP41 AMOV0348 NUCLEAR SAFETY .1 ELATED F IG. 7.3.1 - 29 E                                ,

SERVICE WATER SYSTEM LOGIC DIAGRAM SHOREHAM NUCLEAR POWER STATION-UNITI . FINAL SAFETY ANALYSIS REPORT REVISION 20 - NOVEMBER 1980

SOURCE HONITOR COND ITION CONTROL ACTK P41 (NARR0w RANSE)

                            *n    8
      / P41                   GEA             CLCw H1
                                      ' IP42tE0ll A SERVICE (AFT                    IP41         WATER FLOW t                         ffMI gpg;               037A (3          PSIAM0v037A VALVE POSITION
                                                              ]

4 + P NT ,

037A j 5, LS [RBCLCW HEAD IP421TK026 A LEVEL Tk >j 0 (LCa-L0w <5'9' A .B .L NOTE 2) P42 "RSCLCW MEAD TK

4 LS 012A 2

IP42 gTK026B LEVEL -

h d (L0w-LCw<5'9'A.B.L./ [ENERGENCY BUS 101 - l \ III I3I-2SA I LOSS OF P0nER ? A o N R IP41AH0V037A - 33 . OPEN N AND FB-SR IF414 C V037A \ p ? O V E R 8.l C E

i Q. - i AND 4 0 OCA M WAY I -- 4 923 F IG. 7.3.1 - 19 D INITIATICM SIGNAL 3 3 g PRESENT NOT

        %CTES:

I.. VALVE MAY BE THROTTLED BY NOLDING THE PUSHSUTTON IN THE DEPRESSED POSITION WITH THE H0DE SELECTOR SalTCH IN THE " THROTTLE HODE' POSITICM UNTil DESIRED FL0w RATE IS ESTABLISHEQ.

2. HEAD TANK LEVELS MEASURED FROM BOTTON BEND LINE OF TANK.

3. THE MOTOR OVERLOAD INTERLOCK IS BYPASSED WHEN HOLDING THE PUSMBUTT04 IN THE DEPRESSED POSITION.

4

5. KEYLOCED SalTCH - REY RENOVEABLE IN "NCRMAL MODE' POSITIOK.

f

O RESULTANT MONITOR (NOTE 5) SW IP4lthov037A \ ? THROTTLE MODE [ CF , D NOT C- IP4lXMOV037A AND D THROTTLE CLOSE

                                                      ?   NOT         ?
                                                                      ?

PB-SR (NOTES (83) IP41A H0V037A > CLOSE CH

SW (NOTE 5) IP41%MOV037A \ ? IP411MOV037A y NORMAL MODE [C8 ? CLOSE G CH D NOT -9 _ AND ,

?-

3 NOT -9

SW IP41AMOV037A \(NOTE 5)C- I P4lh MOV037A ' / NORMAL MODE [CB AND D INTERMEDIATE . A N POSITION (g

-k W                             - PB-S R
                                                       \(NOTES 183)
  /         GR                      IP41+ MOV037A                     D INTERMEDIATE       [CB NOT  -D D

AND y I P4 t X MOV037A v SW

                                                                          -   OR        OPEN                p   R IP41 %MOV037A NORMAL MODE (NOTE 5)                              CB          ,_ g PB -S R                     9 IP4 H- H0V037A              >

(NOTES 183) OPEN 9. IP41 MOV037A AND ? Sw THROTTLE OPEN (NOTE 5) IP41kMOV037A THROTTLE MODE jCR FIG. 7. 3.1 - 2 9 F , RBCLCW HEAT EXCHANGER OUTLET VALVE IP414 MOV037A SERVICE WATER SYSTEM LOGIC DIAGRAM NUCLEAR SAFETY RELATED ' SHOREHAM NUCLEAR POWER STATION UNITl FIN AL SAFETY ANALYSIS REPORT REVISION 20 - NOVEMBER 1980 l

SCURCE MONITOR CONDITION CONTROL Action _ fP41 (MARROWRANGE) tpgl 1468 gg [N5GLCWMA RSP TRANSFER D j , 77 1 IP42 *E0llB SERVICE

MCRMAL [Bif AMR R0W 1468 spgg f

   +
    ,               y spyy                                                  (NOTE 1) Sw                   \            AND IP41* MOV0378          \      ~

0378 (g [l P41 t MOV')378 IP41 THROTTLE MODE I VALVE POSITION [CR JEU ( PB-SR N

(NOTES 2.4) IP41 k MOV0378 o CLOSE ,8 4 P4 RBCLCW HEAD TK (NOTE 1) SW IP4tt MOV0378 \ h

AND tLS 2M .

                                  --        iP42 TK026 t     A LEVEL         ;  ,          NORMAL MODE           [CR
      .012B                                 LOW-LOW < 5'9" A.B.L.

Pg BCLCW HEAD TK 41t MOV0378 ?

LS Pu2 ATK026B LEVEL MORMAL [R32 OR 013C OW-LOW < 5' 9" A .B .L .

                                                                                                                                ?  O EMERGENCY BUS 102                                                         D' FlG' T'3'l-MB            LOS$ OF POWER                 N                                               AND  --.ca R 2

4 PB-SR IP41A MOV0378 IP41 A MOV037B og 33 OPEN OVERRIDE CR AND -9 0 [LOCACORESPRAY2 lNITIATIOP SIGNAL i h ^ ~6* R O' LSK-24-4.4E PRESENT j 0--9 NOT NOT ES: I.. KEYLOCKED - KEY REMOVEABLE IN " NORMAL MODE" POSITION.

2. VALVE MAY BE THROTTLED BY HOLDING THE PUSHBUTTON IN THE DEPRESSED POSITION WITH THE MODE SELECTOR SWITCH IN THE " THROTTLE MODE" POSITION UNTIL (NOTES 2.4)

DESIRED FLOW RATE IS ESTA3LISHED. PB-SR

3. HEAD TANK LEVELS MEASURED FROM BOTTOM BEND LINE IP41fMOV0378 a 0F TAMK. OPEN GR

4. THE MOTOR OVERLOAD INTERLOCK MAY BE BYPASSED 4 BY HOLDING THE PUSMBUTTON IN THE DEPRESSED POSITION. (NOTEI) i SW IP4titMOV0378 \ s THROTTLE MODE

[CR AND RBCl SW RSP TRANSFER

                                                                                                            \              D NORMAL               [ ESE i

i (

RESULTANT I D NOT U IP414 M0v0378

                           --e    NDT                        p     AND                           > THROTTLE CLOSE SW l

IP41t MOV0378 5 CLOSE E AND ? SW s RSP TRANSFER IP411POV0378 - C8 EMERGENCY _$f OR > CLOSE D - G s D NOT U AND ? d NOT Y D' SW RSP TRANSFER NORMAL ESE IP48t MOV0378 AND

                                                                                                 >   lNTERNEDIATE      -     A PB-SR                                                              POSITION          /         Ngg O
    /         CB                  IP41AMOV0378                 -    (NOTES 2.4)

INTERMEDIATE CJ

4 NOT 4 sw 9.

IP41 AMOV0378 , (NOTE 1) AND i-NORMAL MODE gg 9 AND 9 AND -Z) SW RSP TRANSFER

                                                      \,                        >                                         -    R NORMAL              / R$2                                           IP41 *M0v0378 GR-
                                                                                           ,OR       O PEN          +'

SW IP415M0v0378 - R OPEN R$f AND ? IP41AM0y0378 EE

                                                                                               --@    THROTTLE OPEN
                              /Sw RSP TRANSFER R$E
                              \ EMERGENCY

.CW HEAT EXCHANGER OUTLET VALVE IP41E MOV0378 NUCLEAR SAFETY RELATED FIG. 7.3.1 - 29G SERVICE WATER SYSTEM i LOGIC DIAGRAM SHOREHAM NUCLEAR POWER STATION UNIT I FINAL SAFETY ANALYSIS REPORT R EVISION 20 - NOVEM BER 1980

a I

SOURCE CONDITION CONTROL ACTION PB-SR (NOTE 1) IP41hMOV032A OPEN [Cg SW-SR (NOTE 2) MOV032A OVERRIDE D CLOCKWISE [Cg AND C' ) f J g, y,3,j _ J9 g LOCA CORE SPRAY I m 23 INITIATION SIGNAL PRESENT O lFIG. 7.3.1 - 2 9 ALOSS [EMER. OF POWERBUS!OI -

                                                        )                                                        ._

IP41AMOV032A 33 (CLO3ED l IP411 M0YO32A

                                                                                           \ (NOTE 1) l                                                                         CLOSE

[CE REACTOR BUILb'NG SERVICE WATER I NOTES

1. THE MOTOR OVERLOAD INTERLOCK MAY BE BYPASSED BY HOLDING THE PUSHBUTTON IN THE DEPRESSED POSITION.

2. KEYLOCKED SWITCH - KEY REMOVEABLE IN " COUNTERCLOCKWISE" POSITION. OVERRIDE IS TO BE USED ONLY WHEN IT IS REQUIRED

TO SUPPLY TWO OUT OF FOUR RBSYS AND CRAC CONDENSING WATER

! PUMPS WITH ONLY ONE SERVICE WATER PUMP AVAILABLE.

l I

t a RESULTANT MONITOR D D NOT C IP41 AH0V032A / AND --C' OPEN v R CE w ca D 0 4 NOT 4 D NOT D MOT D : R AND l' s I p, D 0 D D R Q - I IP41 AMOV032A CLOSE G

? '

(EADER ISOLATION VALVE IP41 (MOV032A NUCLEAR SAFETY RELATED FIG. 7.3.1- 29 H SERVICE WATER SYSTEM t' LOGIC DIAGRAM SHOREHAM NUCLEAR POWER STATION-UNITI ' FIN AL S/ FETY AN ALYSiS REPORT ( R EVISION 20 - NOVEMBER 1980

SOURCE CONDIT10N CONTROL ACTION SW

 ,                                                                I P41 + M0YO32B
                                                                                      \                 ?
  )                                                              OPEN                 / EXE i                                                                                                         AND 6                                                               SW N

RSP TRANSFER. EMERGENCY ' N [ R$E PB-SR ( NOTE 1) I P414 MOV0328 OPEN [ CE SW \ RSP TRANSFER \ NORMAL [ ESE SW-SR (NOTE 2) M0v0328 OVERRIDE ; CLOCKWISE [ CH AND [ 4 Q NOT q [LOCACORESPRAY2 INITIATION SIGNAL CIl PRESENT j 2 FIC. 7. 5.l-29BLOSS hMER. BUS 102 OF POWER A _ C P41 *MOV0328 l 33 CLOSED j j PB-SR I P414 kOY0328 \(NOTE 1) l CLOSE [ CE l SW RSP TPANSFER NORNA: [ESE SW NOTES RSP TRANSFER

l. THE MOTOR OVERLOAD INTERLOCK MAY BE BYPASSED ENERGENCY

[ ESE BY PhDING THE PUSMBUTTON IN THE DEPRESSED AND P%ITION. SW

2. KEYLOCKED SWITCH - KEY REMOVEABLE IN M0v032B f ?

g CLOSE

            " COUNTERCLOCKWISE" POSITION. OVERRIDE IS TO BE    \                     / ESE USED ONLY WHEN IT IS REQUIRED TO SUPPLY TWO OUT                                REACTOR BUILDING SERVl!
                                                                                            ~ ~ ~ ~             -

0F FOUR RBSYS AND CRAC CONDENSING WATER PUMPS WITH ONLY ONE SERVICE WATER PUMP AVAILABLE. { I

RE ULTANT MONITOR D IP48t MOV0328

                                                  >           OR   4 OPEN                J    -    R CR
                                                  >                                                R AND 4 D NOT 2 4 W C8 i,O C        C NOT     d               4   NOT AND        J D               5 R

AND d'

                                                  >                   IP41A MOV0328
                                                                                                     /

I I OR 4CLOSE ;; - G

                                                                                            '          CE
                                                                                                     /

G B12 D , D ATER HEADER ISOLATION VALVE IP4t d MOV0328 FIG. 7.3.1 29 J SERVICE WATER SYSTEM NUCLEAR SAFETY RELATED LOGIC DIAGRAM SHOREHAM NUCLEAR POWER STATION-UNITI . FINAL S AFETY ANALYSIS REPORT REVISION 20- NOVEMBER 1980

500ltCE MONITOR CONDIT10N COC t i SW-SR IP41&M0 OPEN T'lWB. BLOG. lP41AMOV035A ISOL. VA. 33 [NOTFULLYCLOSED ] "A OPEN L ) 7 FIG. T.3.1 19 B [LOCASIGNAL I INITIATION CORE SPRAY l j O23 PRESENT M .B !01 FIG 7*3.1-29 A a g LOSS OF E0WER a 4 NOT Y U 0 0 ED ~ AND

SW-SR 7 [D!ESELGENERATOR BREAKER ACBl0!-8 52 3 s pgl 4 Mo1 OPEN \ CLOSE g TURBl> [ o0R1 l r j l l NOTES

1. THE MOTOR OVERLOAD INTERLOCK MAY BE BYPASSED BY l HOLDING THE CONTROL SWITCH IN POSITION.

l 2. l 3. KEYLOCKED SWITCH - MEY REMOVEABLE IN MID-POSITION. I { l < l I

1 i TROL Action RESULTANT MONITOR i /035A \ (NOTES I & 3) D

           / ca                                ,

IP41 MMOV035A D N0r ) AND-9lOPEN --

                                                                                         ? R G8 NOT 3.
                                                       ?

I P414, NOV035A N

                                                       ?, OR   9   CLOSE                Cy g
                                                                                        /-           gg

'035A \ (NOTES I & 3)

           /cs 1 BUILDING SERVICE WATER ISOLATION VALVE iP414HOV035A

[l0NS ARE NUCLEAR SAFETY RELATED FIG. 7.3.1 2 9 K SERVICE WATER SYSTEM LOGIC DI AGRAM 8 SHOREHAM NUCLEAR POWER STATION-UNIT l FIN AL SAFETY ANALYSIS REPORT REVIS10tl 20 - fl0VEMBER 1980

i CONDITION C0k TROL A MONITOR f SOURCE SW RSP TRAN

  \                                                                                    EWERGENC
  \

SW IP41\MC OPEN SW.S,R I P41 +, M( OPEN SW TURS. BLDG. RSP TRAL IP41%M0y0358 ISOL. VA. NORMAL e A OPEN 33 NOT FULLY CLOSED L J 7 x [LOCACORESPRAY2 IMITIATION SIGNAL O11 PRESENT j O2 FIC.T.3.1 298 [EMER. BUS 102 (Y LOSS OF POWER

                                               )
                                                          ?   NOT  -9           --9 o lP41 % MOV0358                                         g 33                 [ CLOSED                                   D                  SW.Sg AND            g pgg 4 g CLOSE DIESEL GENERATOR
                                                                     ~

52 BREAKER ACBl02-8 (OPEN ] SW RSP TRAI NORMAL SW I P41 A M CLOSE SW RSP TRAI ENERGEN WOTES

1. THE MOTOR OVERLOAD INTERLOCK MAY BE BYPASSED BY HOLDING THE TURBINE B CONTROL SWITCH IN POSITION.

2. PORTIO 8 3. KEYLOCKED SWITCH - KEY REMOVEABLE IN MID-POSITION.

1 l

RESULTANT MONITOR CTION SFER \ y i y

          /ase AND                                 D v0358               rs
          / Ese "i q             >

IP41*MOV035B OPEN +- R (NOTES I & 3) - CB 035B " '

         /ca                                                                                           ,

BSE ISFER y g

          / Ese 9   NOT             D
                                            > NOT  U
                                      -9 D

v (NOTES I & 3) p0358 D AND --9 l CE l q s IP41\MOV035B lSFER g3g g - ctOSE - o G )V0358 p. --

          /Ese AND                                y l

iSFER N by

           / Rie JILDING SERVICE WATER ISOLATION VALVE IP41 +MOV0358 F I G. 7. 3.1 - 2 9 L l

NS ARE NUCLEAR SAFETY RELATED SERVICE WATER SYSTEM LOGIC DIAGRAM SHOREHAM NUCLEAR POWER STATION-UNIT I FINAL SAFETY ANALYSIS REPORT ) REVISION 20 - NOVEMBER 19 80

i SOURCE CONDITION CONTROL u LOCA COWE SPRAY l fl ^ ~ *H' sl 23 INITIATION SIGNAL ' PRESENT O FIC.T.3.129 A 0 0 PO ; O

                           \
                                                                             >t R O' NOT 4                         pg I P41 t M04036 A 33                       CLOSED                               -
                                                                           ~

AND [lESELGENERATOR 52 ' BREAKER ACB101-8 " OPEN VENTILATION t )

LCTION R ES'J LTA NT MONITOR -SR 4l ( HOV036A ~ r.N CR IP41* M0v036A / AND -? OPEN yR

                                                                                       /          CR C  MOT      ?

_ _o A IP41 AMOV036A N CLOSE ' CR

=SR 41QMOV036A                               y OSE

[ CR l HILLED CATER SYSTEM SUPPLY ISOLATION VALVE iP414 M0v036A_ NUCLEAR S AFETY R ELATED FIG. 7. 3.1 - 29 M SERVICE WATER SYSTEM LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNIT I FIN AL SAFETY ANALYSIS REPORT REVISION 20- NCVEMBER 1980

[ l SOURCE CONDITION CON IPW OPE OCA CORE SPRAY 2 O INITIATION SIGNAL PRESENT j 3. 3 FIG. T.3.1 - 2 9 B [EMER. BUS 102 2 LOSS OF POWER , v o N O'R

                                                       --h NOT 4 l P41 A H0V036 g                                       PS-33                     [ CLOSED                                                   IP4 j             D.AND CLO DIESEL GENERATOR 52                        BREAKER ACB102-8                 D-(OPEN                 ]                    VENTILATION CHilLQ ei

i TROL ACTION RESULTANT MONITOR , SR )G MOV0368

              !                                     I P414 MOV0368 AND      ? OPEN                    g
                           ?  NC T W D

QOR - IP414M0v0368 CLOSE MG Y

                                                                       /        CR BR I $NOV0368                              0 BE

[C1 ) WATER SYSTEM SUPPLY ISOLATION VALVE IP41%M0v0368 NUCLEAR SAFETY RELATED FIG. 7.3.1 - 29 N .* SERVICE WATER SYSTEM LOGIC DIAGRAM , SHOREHAM NUCLEAR POWER STATION-UNITI FINAL SAFETY ANALYSIS REPORT REVISION 20 - NOVEMBER 1980

 'l
  ?

SOURCE CONDITION COPT PB-IP4 OPE FIC. T.3.1-198 [LOCACORESPRAYl INITIATION SlWAL v O23 PRFSENT LOCA CORE SPRAY 2 INITIATION SIGNAL ( D OII PRESENT j k OR 3 FIG. T.3.1 - 2 9 A. B t 3S 0 ER D 0

A,R 4'NOT N ' IP41 %,MOV036C E8" CLOSED ^ IP4 k ] AND LC DIESEL GENERATOR D VENTILATit (OPENBREAKER ACBl03-8 ] i 7 1 1

I R0L ACTION RESULTANT MONITOR BR I AM0v036C D n

               /2                                              IPul KM0v036C AND               ?  OPEN                   R D NOT       ?

N m ) O IP41 AMOV036C D CLOSE -- G SR h$MOV036C g, SE [G el CHILLED WATER SYSTEM SUPPLY ISOLATION VALVE IP414M0f036C NUCLEAR SAFETY RELATED FIG. 7.3.1 - 29 P SERVICE WATER SYSTEM i LOGIC DI AGR AM { SHOREHAM NUCLEAR POWER STATION-UNITI ' FINAL SAFETY ANALYSIS REPORT REVISION 20 - NOVEM BER 1980

I. I

SOURCE CONDITION CONTROL ACTION PB-SR IP414 MOVl29A OPEN [CR FIC* 7*3*l-198 [LOCACORESPRAYI 23 lNITIATION SIGNAL kPRESENT PB.SR \ (p IP41 AMOV129A CLO3E /CE DRYWELL BOOSTER HEAT EXCHANGER If i NUCL NOTES

1. DRYWELL 800 STER HEAT EXCHANGER IP42 AEll7A SERVICE WATER OUTLET VALVE IP414 MOV129A SHOWK.

DRYWELL 800 STER HEAT EXCHANGER IP421 Ell 78 SERVICE WATER QUTLET VALVE IP41 %MOV1295 SIMILAR.

2. THE MOTOR OVERLOAD INTERLOCK MAY BE BYPASSED IN THE CLOSING DIRECTION BY HOLDING THE CLOSE FUSMBUTTON IN THE DEPRESSED POSITION.

1 4

e

                                                                                                        ,I RESULTANT          HONITOR C

IP41+ MOV129A / 4MD D OPEN g CB D NOT I m D pOR , IP41tMOV129A C) CLOSE G

V / CB CTE2) 1 $2AEll7A SERVICE WATER OUTLET VALVE !P41 AMOV129A (NOTE l) EAR SAFETY RELATED FI G. 7. 3.1 - 2 9 0 ' SERVICE WATER SYSTEM LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNITl FIN AL SAFETY AN ALYSIS REPORT l REVISION 20 - NOVEMBER 1980

CONTROL ACTION

I SW

 %                                                                                   IP41A MOV033A OPEN SW IP41 (M0v033A EMER. OPEN SW IP41t MOV033A CLOSE PB-SR IP414 M0v039A OPEN
                                                                                  ' PB-SR IP41%MOV039A CLOSE g-~  NOTES

\ 1. ULTIMATE COOLING SEA WATER VALVE IP41 XM0v033A SHJWN. $y ULTIMATE COOLING SEA WATER VALVE IP41 AM0v033B SINILAR. IP41 MOV042A ULTIMATE C00LlWG SEA WATER VALVE IP41 AM0v033C AND OPEN IP414MOV0330 ARE ALSO SIMILAR EXCEPT THAT REDUNDANT INDICATING LIGHTS ARE LOCATED OM G.E. MIMIC. 3,

2. KEYLOCKED SWITCH-KEY REMOVEABLE IN CLOSE POSITION. IP48mMOV042A

3. ULTIMATE COOLING DRAIN VALVE IP41 KNOV039A SHOWN, EMER. OPEN ULTIMATE COOLING DRAIN VALVE IP41 AMOV039B SIMILAR.

EMERGENCY SERVICE WATER TO FUEL POOL DRAIN VALVE IP414MOV043 SIMILAR. SW

4. EMERGENCY SERVICE WATER TO FUEL POOL VALVE IP411M0v042A SHOWN, IP41 )(MOV042A CLOSE EMERGENCY SERVICE WATER TO FUEL POOL VALVE IP41 A MOV042B SIMILAR.

                                                                                 /SW IP41 FM0V042A
                                                                                 \EMER.CLOSE
 )

RESULTANT MONITOR I \ J CB I IP416 MOV033A (NOTE 2) n. OPEN , g [ CE (NOTE 2) IP41 KMOV033A

 \                                                         D    CLOSE                                      j    G ULTIMATE COOLING SEA WAFER VALVE (NOTE 1)

IP41n A 4039A / \ O OPEN g [CH _ CE P41(MOY039A D yG

CLOSE gg CE ULTIMATE COOLING ORAIN VALVE (NOTE 3) \ ~ CR qOR (NOTE 2) IP41$MOV042A OPEN ' [ Ca 1 \ m [CB (NOTE 2) IP41T MOY042A I D CLOSE pG ' \ 1 [CE i EMERGENCY SERVICE WATER TO FUEL POOL VALVE (NOTE 4) NUCLEAR SAFETY RELATED FIG. 7. 3.1 - 2 9 R SERVICE WATER SYSTEM LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNIT l < FIN AL S AFETY AN ALYSIS REPORT REVISION 20 - NOVEMBER 1980

SOURCE HOMITOR CONDiTI1M CONTROL ACT100

                                                                                                 / SW   IP41-S001A
 .                                                                                                \ HAND l    IP41                                                 IP41 500lA
    -PDT                >     A/D s'
 -    Ol0A                                             (> STRAINER 4 PSID           [6P IP41-500lA 42                                                STRAINER NOT ip4    (RUMMING
                                             $ PDI 150A        Ip41.Soolg IP41 STRAINER
     *PDT                                 ,

SOA DIFF. PRES $URE SW R R.B. SERVICE WATER I P41 -Sf)0 l A g

                         >    A/D STRAINER 6P HIGil > 4 PSID AUTO IP41
                                               -PDI
                                            \010A           IP41-300lA
                   ;                                       STRAINER
      .PDT                                        '

Ol0A (DIFF. PRESSURE T.D. C IP41-500lA

                    < W    .

A/D

STRAINER 6P $w IP41-S00lA (HIGH>8PSID MAND

                                                           ~R.B. SERVICE WATER
                         >    A/D
                                                       ,   STRAINER 6P
                                                        \ HIGH > B PSID                               Sy IP41-$00lA 0FF 42 P41-S00lA

[ STRAINER } D T.D. 10 MIN. ; RUNNING _ lP41-S001A IP41

      -PDT               ?   A/D                       [ STRAINER 6P             --

010 _

                                                           < 4 PSID 41-500lA

[' CONTROL POWER DIRECT AVAILABLE AUTO ll0TES

                                                                                                 /SW  IP41-S00lA

1. R.B. SERVICE WATER PUMP IP414- P003A STRAINER IP41-S00! A AND \0FF BACKWASH VALVE IP41-MOV040A SHOWN. R.B. SERVICE WATER PUMPS e

IP4t + P0038.C. AND D STRAIMERS IP41-$00lB.C, AND D AND REACTOR BUILDING SEl BACKW ASH V ALVES IP41-HOV0408.C. AND D SIMILAR. ( 2. NEMORY 15 NON-RETENTIVE. OUTPUT IS LOST ON LOSS GF CONTROL POWER.

3. MOTOR OVERLOAD PROTECT 10M(NOT SHOWN) IS PROVIDED FOR STRAINERS IP41-S00lA. B. AND C.

RESULT AN T NONITOR D L , C - 1 IP41-500lA D START AND

Oj OR IP41-MOV0404

                                                                                                                                                                                                                                                             ,     D   OPEN m ,

IP41-$00lA D ? START SIGNAL 4R AND D 0

? "

4'R (NOTE 2) /t PORTIONS ARE NUCLEAR SAFETY RELATED

                                                                                    ~
                                                                      ~

s-L l D -t> NOT 4 i O

.~

h --9 llP41-S00lA STOP

L IP41-N0V040A

                                                                                                                                                                                                                                                           --@ CLOSE

\ ~ /' IVlCE CATER STRAINER AND BACKWASH VALVE (NOTE I) 1 FIG. 7.3.1 - 29 S SERVICE WATER SYSTEM ' l LOGIC DI AGRAM , I SHOREHAM NUCLEAR POWER STATION-UNITI , j FIN AL SAFETY ANALYSIS REPORT 1 REVISION ,20 , NOVEMBER 1980

SOURCE CONDITION CONTROL ACTION t SW-SR i (NOTE 3) I P41'fMOV102A f k CLOSE L g g;y, ; ) 23 F 1 ' 7.5.l- 19 B [LOCACORESPRAYl INTITIATION SIGNAL PRESENT j RETURN FLOW IP41X A/0 FROM RAD. MONITOR Fil51A LOW'.(LTR)GPM [ RNR HX OUTLET VV. T\ m 33 1 i IP4l#MOV034A " (NOTFULLYCLOSED j 5. SW-SR

                                                                                                 \ NOTE 3)  '

IPUMMOV102A f OPEN /g D T.R. m (LTR) SEC.

                                                                                   ?    NOT D

AND RADIATION MONITOR 33 I SOL.VV. l P41/ MOVl 02 " D OPEN 1 IP41MMOV102A SHOWN-RADI ATION HONITOR ISOLATION VALVE. IP41MMOV102B SIMILAR.

2. AUTO START /STOP SIGNAL LOGIC FOR SW RADIATION HONITOR PUMP IDil-P251A SHOWN, AUTO START /STOP SIGNAL LOGIC FOR SW RADI ATION MONITOR PUMP IDil-P2518 SIMILAR.

3. MOTOR OVERLOAD (NOT SF0WN) MAY BE MANUALLY BYPASSED BY HOLDING CONTHOL SWITCH IN POSITION.

f 4 ISSUE NO. 5 INCORPORATES CHANGE CONTROL FORM P41/9. I,

5. LOCA INITI ATION SIGNAL BYPASSES THE MOTOR OVERLOAD.

RESULTANT N0NITOR e D (NOTE 5)

h IP4tWHOV102A Q "- CLOSE AND 4 ft) SEC, NOT % I OR IP41%H0Vl02A i AND D OPEN G D RADIATION MONITOR ISOLATION VALVE (NOTE l} AUTO START SIGNAL i N TO

                                                    ?    TO SW RAD. MONITOR f                                 MMC SUMP 10ll-P251A NOTE 2                              ) DwG. NO.

AUTO STOP SIGNAL (LATER)

                       ?  MOT                        D   TO SW RAD. MONITOR                                    f PU4P IDil-P251A
                                                                                                                              /

PORTIONS ARE NUCLEAR SAFETY RELATED FIG. 7.3.1 - 2 9 T SERVICE WATER SYSTEM LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNITI ,' FIN AL SAFETY ANALYSIS REPORT REVISION 20 - NOVEMBER 1980

SOURCE MONITOR CONDITION (ONTRCL Ai

                                                        ' DISCHARGE VALVE IP41-H0Vil2A OSFD                                    SW-SR

, y / IP41-P271A A START iP41-P271A 51

                         "                                MOTOR OVERCURRENT                                 SW
                              ~                                                        (NOTE 2) .

AUTO STANDBY C IP41-P27lA \IP41-P271A STOPPED 52 ] [SW-SRIP41-P271A T \ AUTO (AFTE IP41-P271B 52 STOPPED - AND O @;lf 3-V ( IP41-P271B J' /SW-SR IP41-P27lA ELECTRICAL STOP 86

                                                      ' PROTECTION TRIP IP41-P271C 86                                                     ELECTRICAL PROTECTION TRIP MOTOR PROTECTION ELECTRICAL LOCKOUT 86 RELAY OPERATED     j P41-P271A 27                                                     BUS UNDERVOLTAGE PB-SR IP41-MOVil2A OPEN IP41-P271A D  NOT PB-SR IP41-MOVil2A CLOSE I. TURBINE BUILDING SERVICE WATER PUMP IP41-P27l A AND DISCHARGE VALVE IP41-H0Vil2A SHOWN, TURBINE BUILDING SERVICE WATER                               TURBINE BUILDIN

[ PUMPS IP4l-P2718. IP41-P271C AND DISCHARGE VALVES IP41-MOVil28, { I P41 -M0'il l 2C S I MI LAR.

2. FOUR POSITION AUTO STANDBY SWITCH WITH " PUMP A" " PUMP B" -

             " PUMP C"       "0FF" POSITIONS, C0H40N TO PUMPS IP41-P271 A.B AND C.

VION HONITOR x sf i

gg [c2

m") -

            ~
      /ce '                                   IP41-P271A
                                                              ~>                                      gg 3                          AND      > START s
                                                                       ~
 >    /ca                                                              v 3                                                          ,m_

AND C AUTO START p 9

                )  NOT     3:
      \                    =
      /Ca                        ]             IP41-P271A y
                                          ?    STOP CB C

g T.D. 3 SEC. NOT NUCLEAR SAFETY RELATED

       \
        \

IP41-NOVII2A g [C8 q ? OPEN T .D . 2 MIN. s h _ gg N IP41-NOVil2A s

CP CLOSE " G g

       /ca                                               FIG. 7.3.1 - 29 U hSERVICEWATERPUNPANDDISCHARGEVALVE(NOTEI)                SERVICE WATER SYSTEM                                    E LOGIC DI AGRAM                                          k-SHOREHAM NUCLEAR POWER STATION-UNITI FIN AL SAFETY ANALYSIS REPORT REVISION 20 - NOVEMBER 1980

SOURCE MONITOR CONDITION CONTROL ACV PB-SR

IP41-MOVillA OPEN f

I A PB-SR IP41-M0 Villa CLOSE

                                          ,                                      TBCLCW HX SERVI T.B. SERVICE WTR.
                     ---@      A/D

[HEADERPRESSURE LOW < IP41

       -PT           '

116 [TURBINEBUILDING SERVICE WATER I P41 HEADER PRESSURE

                                           <    -Pl ll6X CE IP41                              ,
                                                  -PI Il6Y L       T.B. SERVICE WTR.

52 PUMP IP41-P271A RUNNING

                                                               .B. SERVICE WTR.

52 PUMP IP41-P271B RUNNING T.B. SERVICE WATER 52 PUMP IP41-P271C RUNNING NOTES

1. TBCLCW HX SERVICE WATER OUTLET VALVE IP41-M0 Villa SHOWN, TBCLCW HX SERVICE WATER OUTLET VALVE iP48-MOVillB SIMILAR.

(' T.B. SERVICE WATER STRAIMERS IP41-Sl45A AND IP41-Sl45B INLET VALVES IP41-MOVil3A AND IP41-MOVil3B ALSO SlHILAR. {

ION RESULTANT M0NITOR IP41-MOVillA

                                                             ?    OPEN                       >R
    /a                                                                                                 a i

~ IP41-MOVillA y CLOSE DG

    !"                                                                                                 G CE WATER OUTLET VALVE IP41-MOVillA (NOTE 1)
                ?  .AND                           ?

THREE PUMPS RUNNING

T .D . I P41 -HOV 120 OR 2 MIN. 4 OPEN

m p

                 ? 2/3 TWO OUT OF THREE
                ?            PUMPS RUNNING NOT NUCLEAR SAFETY RELATED M
                 ? I/3                            ?

ONE OUT OF THREE PUMPS RUNNING IP41-MOV120 N CLOSE 9' NOT ? 9' not C'AND S' THREE PUMPS STOPPED E) NOT ?- FIG. 7.3.1 - 29V SERVICE WATER SYSTEM ' LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNITI ' FIN AL SAFETY AN ALYSIS REPORT REVISION 20 - NOVEMBER 1980 1

SOURCE MONITOR C0NDI TION CONTROL ACTIO 9 /SWIP41-Si45A HAND gpgg P41-S145A

        -PDT               N    A/D                            STRAIMER AP ll5A IP41-H0Vil3A 33 (NOTFULLYCLOSEDSTRAIMERIN
                                                                                         /SW
                                                /    /         T.B. SERVICE WTR.            IP41-S145A
                           ~

A/D STRAIMER AP AUTO IP41 IGH >

                                              , -PDI IISA           IP41-S145A lpyg                                        (
          -PDT                              '                   STRAINER 6P ll5A                                             k IP41-Sl45A
                           '    A/D
                                            "                   STRAINER AP                SW HIGH >                      IP41-Sl45A
 ,                                                                                         HAND IP41-H0Vil3A 33                     .

STRAINER INLET , VALVE CLOSED SW IP41-Sl45A AUTO IP41-S145A HOTOR OVERLOAD g 41-Sl45A [0NTROLPOWER OFF DIRECT (AVAILABLE x k iP41 P48-Si45A T. .D. .

         -PDT              O ,

A/D STRAIMER 6P C 3 HIN. IISA NTERVAL TlHER O \" ) SW f NOTES IP41-Sl45A AUTO g 1. TURBINE BUILDING SERVICE WATER STRAINER IP41 -Sl45A AND BACKWASH VALVE IP41-H0V12iA SHOWN. TURBINE BUILDING SI TURBlVE BUILDING SERVICE WATER STRAINER IP41-Sl45B A;10 BACKWASH VALVE IP41440V12iB SIMILAR. i

I RESULTANT HONITOR

     \                                                ~                                                        <

[L . D BACKWASH BACKWASH D SIGNAL ? AND D PRESENT L O I P'il H0V12tA D OPEN

     /t b-            ilNTERVAL TlHER AND
                     -   ON TlHE 3 HINUTES        :   )
             ;,          OFFTIME 7 HINUTES INTERVAL TIMER s

D NOT ? 0FF - A N k s d s I P41 -Sl 45 A g AND v START v NOT v D s AND

     \
     !L                       Q                                          s v

I P41 -S l 45 A l STOP

     \                      ~

9

/t

~

Q Q s IP41-NOVl2iA L g - g - CLOSE F ~ AND U l NOT NUCLEAR SAFETY RELATED

     \                                  3                  FIG. 7.3.1 - 29 W

[L SERVICE WATER SYSTEM ' RVICE WATER STRAINER AND BACKWASH VALVE (NOTE I) LOGIC DI AGRAM l SHOREHAM NUCLEAR POWER STATION-UNIT l l FIN AL SAFETY AN ALYSIS REPORT REVISION 20 - NOVEMBER 1980 l

SOURCE CONDITION MONITOR C { Q FIG. T.3.1 -29Y P4 0 LOSS OF CONTROL 0-0. EC . FIG' T'3'1- 29Y T .D. 0 4 [SmlCE WAW NM IP41 A P003C D,0.5 SEC. I d LOSS OF CONTROL ( YWELL E00 STER HX.011T.

                                                                                            ~                C T

49X , VV. IP41 A K)V129A D-20 . LOSS OF (X)NTROL ENT CHILLED WTR.SYS. T .D . 49X SPLY.15)L. VALVE IP41 A D 20 SEC. K)V036A LDSS OF CONTROL s C VENT CHILLED WTR.SG. T .D . 49X SPLY.lSOL.VALVEIP41A N 20 SEC. - (M)v036C LOSS OF CONTR)L/ [SVCE.WTR.HEADERISOL. T. 4. 49X D 20 SEC. { VALVE IP41*K)v032A (LOSS Of CX)NTROL VCE.WTR PP. IP41 X' T .D . " 49X P-003ADIS01.VV.IP41% D,20 SEC* 1A LOSS OF CONTROL SVCE MTR.PP. IP414 T .D . 49X g P-003C DISCH.VV. IP41 A D 20 SEC. Nt()VCllC LOSS OF CONTROL 49X ULT M TE COOLING S Q WTR.VV. IP41kM0VC33A N T .D . a 70 LOSS OF CONTROL 20 SEC. T .D . 49X [ULTNTECOOLINGDRAIN VV. IP41 A K)V039A D LOSS OF CONTROL j 20 SEC. c q49X (<mRcENCYSERviCEWTR.3 TO FUEL POOL W. lWI t D*20 1.D . SEC. ' (H)V042A LOSS OF (DNTR]OL _ R.H A. HEAT EXCHANGER T .D. 49X OUI.VV.IWI$MOV034A D 20 SEC. LOSS OF CONTROL 349X g.CLCW8x.0UruT VALVE IP41$s D 1.0. 20 SEC. C j LOSS OF CONTR0L MOV037A]S ULT N TE COOLING SEA y ,0 .

 )                                                                         D 49X

[WTR.VV.lP41(M)V03 VOSS OF Coma 20 SEC-T .D . RAD. MON.ISOL. VALVE ] 49X I P41 + MOV 102A D' 20 SEC. LOSS OF CONTROL

CONTROL ACTl0N mMITOR I D

 \                                                                           [

l D-C- D ' N v s '

                                               ^   REACTOR BUILDING SERVICE WATER
                                            ~

SYSTEM "A"

@O

[g DEGRADED l D D

m, D*

1 FIG. 7.3.1 - 29 X l SERVICE WATER SYSTEM s. LOGIC DI AGRAM l ~~ SHOREHAM NUCLEAR POWER STATION-UNIT l C FIN AL SAFETY AN ALYSIS REPORT a REVISION 20 - NOVEMBER 1980 l

SOURCE CONDITION f" 49X [SVCE.WTR.HDR. VALVE IP41A10V032A lSOL. LOSS OF (DNTROL SVCE.WTR. HOR.ISOL. 33 VALVE IP41 M OV032A NOT FULLY CLOSED j [VCE.WTR. HOR.lSOL. _I VALVE IP4lh0V0328 qg* FIG. T.3.1-29I (LOSS OF CONTROL j 33 (SVCE.WTR.HDR.lSOL. VALVE IP41A W V0318 NOT FULLY CLOSED Sw IP41 K P-003A NOT D IP414P003A NORMAL $ 74 , SERVICE WTR. PG ? ? og LDSS OF CONTROL SW lP41$ P-003A D [SVCE.WTR.PP.lP41AP003C PULL TO LOCK [CE 49X OlSCH. VV. IP41KlOV031C LDSS OF CONTROL SW lP41%P-003C m NOT J NORMAL $ IP413P003C 74 i SERVICE WTR. PI M ? OR LDSS OF CDMTROL /SW

                                                                / IP41AP-0030                              s

[SVCE.WTR.PP.IP41AF003A 49X DIS 0t. VV. IP41 t 60V031 A LOSS OF CONTROL DlWELL B0OSTER HX.0UT. VALVE IP41 + MDV129A 49X LOSS OF CONTROL j VENT OtlLLED WTR.SYS, SPLY. lSOL.VV. lP4140V129A 33 NOT FULLY CLOSED RYWELL BDOSTER HX. T s. OUT. VALVE lP4ltt0V036A (LDSS OF CONTROL VENT OtlLLED WTR.SYS. 33 SPLY.lSOL.VV.lP41 Al0V036A > NOT FULLY CLOSED VENT OllLLED WTR.SYS. 49X SPLY. ISOL.VV. IP414 60VCD6B ? LOSS OF CONTROL AND D VENT CHILLED WIR. SYS. f SPLY. lSOL.VV. lP41460VCD6B ? 33 NOT FULLY CLOSED j VENT CHILLED WTR.SYS. OR - 49X SPLY. ISOL.VV. IP41 Etev036C --Gu-VENT DilLLED WTR. SYS. SS OF CONTROL AND - --gu N 33 ~ (SPLY. NOT FULLYISOL.VV. CLOSED IP41 $ 60VCD6Cj

CONTROL ACTION RESULTANT MONITOR I

                    --3:                                                                                -

T.D. AND D 20 SEC. ) j

U SERVICE WATER PUMP

                                                                ?   IP41A.P003A                 ,    3 T.D.                                                       LOSS OF CONTROL d%                                           ?                                                  LSK 9 71 0.5 SEC.

AND C

T.D. OR 3

   ?  20 SEC.                ?

lSERVICEWATERPUMP

                                                                 ?   IP41*,P003C                  -   y T.D.                                                         LOSS OF CONTROL D                                                 LSK-9 7X

+8 0.5 SEC. m AND

                             ?,                                               REACTOR BUILDING T.D.                       OR       3      -

SERVICE WATER

   ?   20 SEC.               D SYSTEM "A" dog        [A        IMOPERATIVE
                             =

T.D. AND 4 20 SEC. M D NOT NUCLEAR SAFETY RELATED SV SYSTEW

                   \                                   3 IN0PERATINE    [

T. D. AND N 20 SEC. FIG. 7.3.1 - 29 Y SERVICE WATER SYSTEM LOGLC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNIT I FIN AL SAFETY AN ALYSIS REPORT REVISION 20 - NOVEMBER 1980

SOURCE CONDITION CONTROL ACTl [ SERVICE WATER PUMP T.D. 5 F I G. 7.3.1- 2 9 A A / IP414 P003B C : 0,5 SEC. LOSS OF CONTROL ERVICE WATER PUMP T.D. l 6 I P41 %P003D ? 0.5 SEC. g FIG.T.3.1-29 A A 1 (LOSSOFCONTROL [ULTIMATECOOLINGDR T.D. l VV. IP41A M0"039B 49X C : 20 SEC. LOSS OF CONTROL ,/ EMER. SERVICE WT' T.D. [FUELPOOLVV. IPO $ / 20 SEC. HOV0428 LOSS OF C6. RHR HX OUTLET VALVE T.D. 49X IP41A M0v0348 C 20 SEC. LOSS OF CONTROL T.D. 49X BCLCWHXOUTLETVV.} IP41% M0v0378 20 SEC. LOSS OF CONTROL [EMER. SERV.WTR.TO T.D. 49X FUEL POOL DRN.VV.lP4lj C : 20 SEC. AMOV043 LOSS OF CNTy DRYWELL BOOSTER HX T.D. 49y OUT.VV. IP41AMOV1298 ?:20 SEC. 1.0SS OF CONTROL j FMT CHILLED WTR.SY T.D. 49X SPLY. lSOL.VV. IP41 WW6 D 20 SEC. kOSSOFCONTROL TURB. BLDG. SERV.WTR. T.D. 49X ISOL.VV. IP415 @ 035A C'20 SEC. LOSS OF M L T.D. 49X [TURB. BLDG. SERV.WTR. C 20 SEC. (NOTE 1) (LDSS OF (NTROLISOL.VV. IPyt tKr.0358] VCE.WTR.HDR.lSOL. T.D. (NOTE 1) ggg VV. IP41$H0v0328 D 20 SEC. OSS OF CONTROL j. [SVCE.WTR.PP. IP4l k TD (NOTE l)

                                / P-003B DISCH.VV.lP4lk                                      C'20 SEC*

H0v03tBLOSSOFCNTL) SVCE.WTR.PP. IP41% T.D. P-003D DISCH.VV.lP41 ? 20 SEC. 49X MOV0310 LOSS OF CNT ULTIMATECOOLINGSEh 1.D. WTR.vv. IP41 AMOV033B) 0 20 SEC. 49X OSS OF CONTROL j T TULTIMATE COOLING SE T.D.

             \                                               [ WTR.VV. IP41 i MOV033D                      D   20 SEC.

{ LOSS OF CONTROL RAD.J40K. ISOL.VV . T .D. 49X IP4t p H0V102B D' 20 SEC. OSS OF CONTROL

1N MONITOR C C D l ? C C C

               .                        s 1

v X: C

<,                                      D
   ~~

C C y C 7 REACTOR BUILDING SERVICE WATER O SYSTEM "B" 7 h DEGRADED

       ' C C
                                    \   ~

NOTES 7 1. LOSS OF CONTROL RELAY 49X WILL BE ACTUATED WHEN ASSOCIATED VALVE REMOTE SHUTDOWN PANEL

      -                                             TRANSFER SWITCH IS PLACED IN"EWERGENCY" POSITION
                      ~                D 1

l> y NOT NUCLEAR SAFETY RELATED - C C

                      "              p C
                                        ?

FIG. 7.3.1 - 29 Z C SERVICE WATER SYSTEM LOGIC DI AGRAM

           '                           O                                                                                    '
                 ,                                                            SHOREHAM NUCLEAR POWER STATION-UNITI FIN AL SAFETY AN ALYSIS REPORT C

n y I REVISION 20 - NOVEMBER 1980

 -   SOURCE                        CONDITION                      CONTROL ACTIO SW RSP TRANSFER EMERGENCY

{

  $                                                             SW IP41*P003B PULL TO LOCK lP411 P0038 (NOTE 1) 74               [ SERVICE WTR. PUMP ]

LOSS OF CONTROL SW IP4t h P003B NORMAL SW iP41 4P003D NORMAL SW I P411 P0030 PULL TO LOCK (NOTE 1) IP41 AP003D 74 , SERVICE WTR. PUMP LOSS OF CONTROL SW RSP TRANSFER EMERGENCY VENT CHILLED WTR.SYS. SPLY. ISOL.VV. IP414 60W36B 49X LOSS OF CONTROL ENT OllLLED WTR.SYS. 33 SPLY.lSOL.VV.IP41XHDv0368 NOT FULLY CLOSED VENT CHILLED WIR.SYS. ggy I SPLY.ISOL.VV.IP41AF0V036A D-LOSS OF CONTROL AND D [VENTCHILLEDWTR.SYS. SPLY. lSOL.VV. lP41 A K)V036A N 33 ~ NOT FULLY CLOSED ENT CHILED WTR.SYS. 49X SPLY. ISOL.VV. IP41 K Hr436C D-SS OF (X)NTROL AND C VENT CHILLED WTR.SYS. 33 SPLY.lSOL.VV.lP41 V 0V036C D NOT FULLY CLOSED

9 RESULTANT

 \
           )
 /a2
 \
 /CR
           ?Q OR SERVICE WTR. PUMP D-lP419,P0038                       -5 LOSS OF CONTROL FIG.T.5.1-29 Z, AB
 \   ? NOT ?
 /s
 \   ? NOT D
 /s
 \         ?q                                            SERVICE WTR. PUMP                p D

[CE V)OR IP41 $ P003D LOSS OF CONTROL FIG. T.3.1-29Z, AB 0, NOTE

 \         s I. LOSS OF CONTROL RELAY 74 WILL BE ACTUATED WHEN ASSOCI ATED PUMP REMOTE SHUTDOWN
 /E2                      PANEL SWITCH TRANSFER IS PLACED IN " EMERGENCY" POSITION. AN AUXILIARY LATCHING RELAY LOCATED AT SWITCHGEAR HUST BE MANUALLY RESET TO D              RE-ENERGlZE RELAY 74 WHEN RSP TRANSFER SWITCH IS PLACED IN " NORMAL" POSITION, D

VENT CHILLED WTR. SYS. AND ? S P LY . ISOL. VALVES ?7 LOSS OF CONTROL FIG. T.3.1-29 AB NOT NUCLEAR SAFETY RELATED

FIG. 7.3.1 - 29 A A SERVICE WATER SYSTEM LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNITI FIN AL SAFETY AN ALYSIS REPORT ~ REVISION 20 - NOVEMBER 1980

SOUkCL CO uti10N CONTRC' J T.D. 5 FI G. T.3.1 - 29 A A --[ SERVICE IP41Y P0038WATER PUMP v 0.5 SEC. LOSS OF CONTROL

VCE.WTR.PP. IWI A FOO3D T.D.

ggx DISCH.W.lP41AM) @ lD D 20 SEC. ? LDSS OF CONTROL { y flC. T. 3.1-29 A A ' T.D. P 03 LOSS OF CONTROL VCE.WTR.PP. iWI AE003 T.D. 2~ 20 SEC. 2 49X (DISCH.VV. IP414 M)v031B] LDSS OF CONTROL SVCE.W T R.H DR. l SO L. ggg . VALVE IP41 ANOV032A N LOSS OF CONTROL SVCE.WTR.HDR. ISOL. 33 VALVE IP41% MOV032A D NOT FULLY CLOSED AND [S VCE.W T R .H DR. l SO L. (NOTEI) I VALVE IP41k MOV032B . 49X 7 (LOSSOFCONTROL SVCE.WTR.HDR.lSCL. 33 VALVE IP41 AMOV0328 D (NOTFULLYCLOSED DRYWELL BOOSTER HX. ) 49X [OUT VV. IP4l+M)V1299 ; LOSS OF CONTROL AND DRYWELL BOOSTER HX. 33 OUT vv. INI ASCV1295 D NOT FULLY CLOSED [ VENT.CHILLEDWTR.SYS. FIG. T.3.1-2 9 A A / SPLY. ISOL. VALVES t 7 LOSS OF (X)MTROL T.B. SERV.WTR.ISOL. 49X VALVE IP41* MDV0354 D l LOSS OF CDNT1DL T.B. SERV.WTR. lSOL. 33 VALVE IP41AM)V035A N NOT FULLY CLOSED AND (NOTE I) [T.B. SERV.WTR.lSOL. ggy : VALVE IP411t M)V0358 D LOSS OF (DNTROL T.B. SERV.WTR.lSOL. 1 VALVE iP41%Mh0356 N 33 (NCT FULLY CLOSED j ~

CTION HONITOR

,s!

AND ?

g - t AND D-rh 3 n ,a T.D.

    ? 20 SEC.             3' REACTOR BUILDING

/ sa SERVICE WATER SYSTEM B OR 'A SYSTEM *B" IMOPERATIVE [G d \lNOPERATIVE T. D. NOTES

1. LOSS OF CONTROL RELAY 49X WILL BE ACTUATED WHEN ASSOCI ATED VALVE REMOTE SHUTD0%N PANEL TRAMSFER SWITCH IS PLACED IN

EMERGENCY
POSITION.

T.D.

    ?  20 SEC.            U NOT NUCLEAR S AFETY RELATED T.D.              s
    - 20 SEC.              9                                FIG. 7. 3.1 - 2 9 A B SERVICE WATER SYSTEM LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNITI               ,

FIN AL SAFETY ANALYSIS REPORT REVISION 20 - NOVEMBER 1980

lJ l ( SOURCE MONITOR CONDITION

iNOT B

                                                 /        M                             lP41% MOV031A 49                                                                  [ MOTOR                l (NOTE 1)                              (OVERLOAD llCTES

1. MOTOR OVERLOAD MONITORING FOR PUMP ir41+ P003A DISCHARGE VALVE IP4f AMOV03f A SHOWN, MOTOR OVERLOAD MONITORING FOR THE FOLLOWING VALVES SIMILAR:

A. IP41 AMOV031B, PUMP IP414P003B DISCHARGE VALVE (NOTE 2) B. IP41 kMOV0310, PUMP IP41 AP003C DISCHARGE VALVE C. IP41%M0v0310, PUMP IP41-AP003D DISCHARGE VALVE D. IP4l-* MOV033A,B,C, AND D: ULTIMATE COOLING SEA WATER VALVES E. IP41 AMOV039A, AND B; ULTIMATE COOLING DRAIN VALVES F. IP41s MOV034A, RHR HEAT EXCHANGER E034A OUTLET VALVE G. IP4l4 MOV034B, RHR HEAT EXCHANGER E034B OUTLET VALVE H. IP41 DMOV036A,B AND C: VENT. CHILLED WATER SYSTEM SUPPLY IS8.LATl0N VALVES ' J. IP41-AMOV032A, REACTOR BUILDING SERVICE WATER HEADER "A" IS0ALTION VALVE K. IP41)M0v0328, REACTOR BUILDING SERVICE WATER HEADER "B" ISOLATION VALVE (NOTE 2) L. IP41%M0YO35A, TURBINE BUILDING SERVICE WATER ISOLATION VALVE J M. IP4l(MOV0358. TURBINE BUILDING SERVICE WATER ISOLATION VALVE (NOTE 2) ' N. IP414MOV037A, RBCLCW HX E0ll A OUTLET VALVE P. IP41-* MOV0378, RBCLCW HX E0llB OUTLET VALVE Q. IP41-AMOV042A AND B, EMERGENCY SERVICE WATER TO FUEL POOL VALVES R. IP414MOV043 EMERGENCY SERVICE WATER TO FUEL POOL DRAIN VALVE S. IP41-AMOv129A, AND B DRYWELL B0OSTER HEAT EXCHANGERS OUTLET VALVES

2. BLUE LIGHT FOR IP414 MOV03l B, 0328. AND 035B WILL BE EXTINGUISHED WHEN ASSOCI ATED VALVE REMOTE SHUTDOWN PANEL TRANSFER SWITCH IS PLACED IN

                      " EMERGENCY" POSITION,

3. VIBRATION MONITORING FOR REACTOR BUILDING SERVICE WATER PUMP IP41# P003A SHOWN, VIBRATION MONITORING FOR REACTOR BUILDING SERVICE WATER PUMPS IP4ltP0038, IP414 P003C, AND IP4l k P003D SIMILAR. VlBRATION MONITORING FOR TURBINE BUILDING SERVICE WATER PUMPS IP41-P271A, IP41-P271B, AND IP41-P271C ALSO SIMILAR.

4. VIBRATION INDICATION FOR R.B. SERVICE WATER PUMP IP41 t P003A, ,$P003B, 4 P003C, (P003D. MAIN CHILLED WATER UNIT IM60 WC00l A, OR IM60-WC00lB CAN BE SELECTED FOR DISPLAY ON METER IP65-VBl505.

5. TEMPERATURE MONITORING FOR REACTOR BUILDING SERVICE WATER PUMP IP41 AP003A.SHO TEMPERATURE MONITORING FOR REACTOR BUILDING SERVICE WATER PUMPS IP414P0038, IP41 \P003C, IP4t tP0030, AND TURBINE BUILDlhG SERVICE WATER PUMPS IP41-P27t A, IP41-P2718, IP41-F271C SIMILAR.

6. B - VIBRATION MONITORING PANEL.

h

1 r SOURCE HOMITOR CONDITION VB (NOTE 4) IP41 \ SOS yM R.B. SERVICE WTR. -VBE ' PUMP IP41 $ P003A 092A / VIBRATION VlBRATION A WARNING E A/D , [R.B.SERVICEWTR. PUNP IP41%P003A - R VIBPATION VIB. HI > EM

                     ' g ' DANGER
                           /                   .B. SERVICE WTR.               '
          ? A/0   ;

g PUNP IP41 & P003A - R

(VIB. HI-HI > 18 C \ R.B. SERV.WTR.PP. T/C - IP414 P003A UPPER I kEARINGTEMP. C .B. SERV.WTR.PP. T/C a

IP41 AP003A LOWER > (NOTE 5) BEARING TEMP. C R.B. SERV.WTR.PP.

                "                             iP41 4 P003A MOTOR TD INDING TEMP.

p PORTIONS ARE NUCLEAR SAFETY RELATED FIG. 7. 3.1 - 2 9 AC SERVICE WATER SYSTEM LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNIT l 6 FIN AL SAFETY AN ALYSIS REPORT REVISION 20 - NOVEMBER 1980

f' SOURCE CONDITION MONITOR R.B. SERVICE WTR. k 52 PUMP IP41 AP003A j t STOPPED

IP41 A P003A CONTROL AND CW I SWliCH IN "AFTER START" POSITION IP41 + P003A MOTOR 86 PROTECTION RELAY N OPERATED > (NOTE I R.B. SERVICE WTR. 52 ~ PUMP 1P41% P003A STOPPED A AND ---o IP41Y P003A CONTROL

I - SWITCH IN "AFTER TART" POSITION j V O j NOTES

1. AUTO TRIP MONITORING FOR REACTOR BUILDING SERVICE WATER PUMP iP41 AP003A SHOWN, AUTO TRIP MONITORING FOR REACTOR BUILDING SERVICE WATER PUMP IP41 A P003C SIMILAR. AUTO TRIP MONITORING FOR TURBINE B ILDING SERVICE WATER PUMPS IP41-P271A, IP41-P2718, AND IP41-P271C ALSO SIMILAR EXCEPT THAT AUTO TRIP ANNUNCI ATION IS COW 40N.

2. AUTO TRIP MONITORING FOR REACTOR BUILDING SERVICE WATER PUMP lP41 AP003B SHOWN, AUTO TRIP HONITORING FOR REACTOR BUILDING SERVICE WATER PUMP IP41% P003D SIMILAR.

I

SOURCE CONDITION MONITOR

           .B. SERVICE WTR.

52 PUMP iP41t P003B v STOPPE0 ' P41 i P003B CONTROL < l SWITCH IN "AFTER ? hTART" POSITION ] OR -- AND W P41( P0038 MOTOR - CE 86 PROTECTION RELAY ? ) PERATED

43 [RSPTRANSFER SWITCH IN " NORMAL"

                               )  w                                            >(NOTE 2)

POSITION j A 52 [R.B.SERVICEWTR. PUMP IP4lt P003B D W AND STOPPED

           ."il A P003B CONTROL J

i SWITCH IN "AFTER s TART" POSITION PORTIONS ARE NUCLEAR SAFLTY RELATED FI G. 7. 3. ,' - 2 9 A D SERVICE WATER SYSTEM ' LOGIC DI AGRAM SHOREHAM NUCLEAR POWER STATION-UNITI FIN AL SAFETY ANALYSIS REPORT L REVISION 20 - NOVEMBER 1980

     /                                                        O i].J -                                                    k                                                              b'N SNPS-1 FSAR TABLE 8.3.1-1 EMERGENCY DIESEL GENERATOR SYSTEM REQUIRED LOADS AND MAXIMUM COINCIDENT DEMAND Number Required Nameplate      Total     Design isasis Loss         M as or      Maximum Coincident Rating        Plant    of Coolant Accident Offsite Power Demand (Kilowatt) ( a >(a a Pimetion             (Hp)        Number 0-10 Min 10 Min on         (Hot Standby) DG-101     DG-102    DG-103 Core Spray Pump              1250            2       1(*)          1             -

998 998 - Residual Heat Removal Pump 1250 4 2(*> 1 2 999 999 1998t** Service Water Pump 450 4 2 2 3883 322 322 644 l RBSVS and CRAC Water Chiller 292 4 2 2 2 236 236 472(*) RBSVS and CRAC Water Chiller Lube Oil Pump .25 4 2 2 2 0.2 0.2 0.u(*3 RBSVS Chiller Circ. Water Pump 75 (34 BHP) 4 2 2 2 28.8 28.8 57.o(*8 RBSVS Chiller Cond. Water 20 4 2 2 2 17 17 33.9(** RBSVS Unit Cooler 30 8 4 4

10 1.7 101.7 -

RBSVS Exhaust Fan 100 3 2 2 2 84.7 84.7 us.7 Reactor Building Exhaust Booster Fan 7.5 2 1 1 1 6 6 - RBSVS Filter Reheat Coil 6.6 kW 2 1 1 1 6.6 6.6 - RBCLCW Circ. Pump 100 (85 BHP) 3 2 2 2 72 72 - l Diesel Generator Air Compressor 10 6 - - - 17 17 17 Diesel Generator Air Compresser Dryer .25 6 - - -

                                                                                                .42        .42        .42 Diesel Generator Fuel Oil Transfer Pump          .5              6       2             2             2         0.4        0.4       0.4 Diesel Generator Jacket Water Heater               36 kW           6       -             -              -

72ft23 72(aan 72(128 Diesel Generator Jacket Water Keep Warm Pump 2.5 3 - - - 2(*31 24133 2(aa > Diesel Generator Lube Oil Heater 20 kW 3 - - - 20(53) 20(na) 20(**> i Diesel Generator Before

     -6 After. Lube Oil Pump      5               3       -             -              -

4(tas atta) 4(na n Diesel Generator Heater 4.2 KW 3 - - - 4.2(13) 4.2(533 4.2(ma> Battery Charger (125 V) 60 KVa 3 2 2 2 48 46 48 120 V ac Instrument Power 75 kva 3 2 2 2 60 61 20 (25 kva DG-103) Diesel Generator Room Vent Supply Fan 20 3 2 2 2 17 17 17 Battery Room Vent Supply Fan 1 3 2 2 2 .8 .8 .o Control Room Air Condition-ing Unit 40 2 1 1 1 33,9 33,9 - Control Room Vent Booster Fan 7.5 2 1 1 1 6.4 e.4 - 1 of 4 Revision 20 - November 1980

(% % % SNP3-1 FSAR TABLE 8.3.1-1 (CONT 'D) Number Required Nameplate Total Design Basis Loss Loss ot Itaximu:n Colncident Rating Plant of Coolant Accident Ottsite Power Demand (Kilowatti t a na > Function (Hp) Number 0- 10 Mi n 10 Min on (Hot StandDY) DG-101 DG-102 uG-103 Feedwater Turbine Turning Gear Oil Pumpt** 10 2 - - 2 8.54=3 u.Sta> - Standby Liquid Controi Pump 40 2 - - - 33.9(ep 33,9te> - Standby Liquid Control Main Heater (5) 10 kW 1 - - - - 10(*) - Standby Liquid Control Mixing Heater ( 5 3 45 kW 1 - - - 45(*3 - - Standby Liquid Control Heac Tracing 3 kva - - - - 3(*) - - 480 V M-G Set 200(1a) 4 2 2 2 23 23 40 Refueling Jib Crane 3.25 2 - - - 2.6t** 2.6te) - Refueling Platfom Assemnly 3.5 1 - - - 2.8(*8 - - Motor Operated Valves 133 - X(?) - - 227 226 138.2 Nonoperating MOV*s(u 3 209.5 - - - - 304 214 1 Total Connected Ioads 4539 4301 4028 l Minus Note 11 Ioads - 304 - 214 - 1 4 2 3 '> 4087 4027 l Minus Note 8 Loads - 530 - 414 - 320 3705 3673 3707 l Minus Note 10 Loads - 21 - 21 - 0 3684 3652 3707 l Minus Not.e 9 Loads - 155 - 176 - 0 3529 3476 3707 ! Minus Note 13 Loads - 102 - 102 - 102 Total kW (One min approx) 3u27 3374 3605 l Minus Motor Operated Valves - 227 - 226 - 138 Total kW (10 min approx) 3200 31e8 3=b7 l Minus Note 4 Loads - 0 - 0 -1281 3200 3148 2186 l 3 of 4 Revision 20 - Nove:ater 19s0

g g

  /
   %                                                                                                                         '% J SNPS-1 FSAR TABLE ts . 3 .1 - 1 (CONT'D)

(** Maximum coincident demand shown occurs during the 0-10 minute period after a design basis loss of coolant accident. (a) Kilowatt loads given are from manuf acturer's data f or the CS, RHR, and . service water pumps, motor generator sets, and the RBSVS chiller "'its. The remaining kilowatt loads are based on the following: xW = .746 x Hg kW = kVa x pf where For MOVs RW = [Tx .4 60 x .41.PA x pt, where pr = .7 0.88 (Ef f .) pf = 0.0 (3) On loss of offsite power, it is necessary to go to a cold shutdown condition if DG-103 does not start, since the three required service water pumps will not be avallaole. Note that only two service water pumps are required f or a design basis loss of coolant accident condition. (Only one pump is connected automatically to DG-103, the other may be connected manually only.) (*) Two units are started on DG-103. One unit is shut down wnen it is determined which section of the system will be l used. (5) These nonclass IE components are not required for a saf e shutdom. Loading indicated for various modes of operation is desirable, although not essential. All remaining components are Class IE. (*) Minimu:n safe shutdown requirements for a suction line break. Actual pump requirements depend on break location (see Section 6.3.3) . (*) X indicates load required. (*) These loads are tripped intentionally (automatically) on a loss of coolant accident (LOCA) . (*) These saf ety related loads are not normally operating and receive no automatic start signal atter a LOCA. (*** These nonsafety related loads have seal-in type control circuits that drop out on a loss of orfsite power prior to connecting to the diesel generators. t * * > These MOV's are connected to their respective diesel tuses but do not operate upon LOCA. t

a > Tne load to be carried by the M-G Sets consist of certain motor-operated valves whose loads are already accounted

' for under the heading, " Motor-Operated Valves." The only additional load on the diesels is due to tne inefficiencies of the M-G Set. On Unit 103, one set operates at full load and one set operates unloaded. (**3 These loads are automatically tripped when diesel generator starts. f

                                                                         " Of "                         Revision 20 - November 1960

 ,O                                                           [
 'N                                                          k SNPS-1 FSAh i

TABLE 8.3.1-2 (CONT *D)

Time Rated brake Starting Running Cumulative Sequence (Sec) (*3 Horsepower Horsepower kVa(28 kW kW Diesel 103

1. Initial Load - 0 -

734 kVa 5,127 564 564 Motor land, including MOV's,Ltg,etc. on 1,000/1,333 kva, 4,160-480 V, transformer with 81 impedance

2. Start RHR Pump 2 1,250 1,250 b,734 999 1,563

3. Start RHR Pump 7 1,250 1,250 6,734 999 2,562

4. Start two Service Water Pumps 12(38 450 400 2,238 644 -

l and two RBSVS Chillers - - 584 2,u00 471 3,355

5. MOVs Complete Operation - - - -

138 3,217

6. Manually stop loads not 600-On - - - - -

required and add additional loads as required within the rating of the diesel generator (t) All large motors and a majority of small motors are squirrel cage induction motors. (a) Starting kva for RHR, CS, Service Water Pumps, 480 V MG sets, and RBSVS chillers are manuf acturers' data. Starting kva for the remaining 480 V loads are based on the f ollowing: Motors - 6 SkVa per HP MOV*S - 10 SxVa per HP Transformers - 1.5 SkVa per kVa Resistive Loads - 1 SkVa per kW (83 Service water pumps and RBSVS chillers receive start signals f rom the bus program at 12 see as shown. However, other interlocks may delay motor start beyond this point. (*D The time shown in table is time after the diesel generator breaxer closes to its associated a XV bus. 2 of 2 Revision 20 - November 1980

SNPS-1 FSAR CHAPTER 9 AUXILIARY SYSTEMS TABLE OF CONTENTS Section Title Page 9.1 FUEL STORAGE AND HANDLING 9.1-1 9.1.1 New Fuel Storage 9.1-1 9.1.1.1 Design Bases 9.1-1 9.1.1.2 Facilities Description 9.1-1 9.1.1.3 Safety Evaluation 9.1-2 9.1.1.4 Inspections 9.1-3 9.1.2 Spent Fuel Storage 9.1-3 9.1.2.1 Design Bases 9.1-3 9.1.2.2 Facilities Description 9.1-4 9.1.2.3 Safety Evaluation 9.1-10 9.1.2.4 Tests and Inspections 9.1-12a 9.1.2.5 Radiological Considerations 9.1-12a i 9.1.3 Fuel Pool Cooling and Cleanup System 9.1-12a 9.1.3.1 Design Bases 9.1-12a 9.1.3.2 Systen Description 9.1-12c 9.1.3.3 Safety Evaluation 9.1-12f 9.1.3.4 Tests and Inspections 9.1-12h Cs 9.1.3.5 Instrumentation Applications 9.1-12h 9.1.3.6 Radiological Considerations 9.1-12i/j 9.1.4 Fuel Handling System 9.1-121/j 9.1.4.1 Design Bases 9.1-12i/j 9.1.4.2 Equipment Description 9.1-12i/j 9.1.4.3 Description of Fuel Transfer 9.1-19 9.2 WATER SYSTEMS 9.2-1 9.2.1 Service Water System 9.2-1 9.2.1.1 Design Bases 9.2-1 9.2.1.2 System Description 9.2-1 9.2.1.3 Safety Evaluation 9.2-4 9.2.1.4 Tests and Inspections 9 . 2-6 a 9.2.1.5 Instrumentation Application 9.2-6a 9.2.2 Reactor Building Closed Isop Cooling Water System 9.2-6 a 9.2.2.1 Design Bases 9.2-6 a 9.2.2.2 System Description 9.2-6 b l 9.2.2.3 Safety Evaluation 9.2-9 9.2.2.4 Tests and Inspections 9.2-10 9.2.2.5 Instrumentation Application 9.2-10 l 9.2.3 Makeup Water Demineralizer System 9.2-10 l 9.2.3.1 Design Bases 9.2-10 l 9.2.3.2 System Description 9.2-11 9.2.3.3 Safety Evaluation 9.2-12 9.2.3.4 Tests and Inspections 9.2-13 9.2.3.5 Instrumentation Application 9.2-13 9-1 Revision 20 - November 1980

SNPS-1 FSAR TABLE OF CONTENTS (COtrPSD) Section Title Page 9.2.4 Potable and Sanitary Water Systems 9.2-13 9.2.4.1 Design Bases 9.2-13 9.2.4.2 System Description 9.2-14 9.2.4.3 Safet.y Evaluation 9.2-14 9.2.4.4 Tests and Inspections 9.2-14 9.2.4.5 Instrumentation Application 9.2-15 9.2.5 Ultimate Heat Sink 9.2-15 9.2.5.1 Design Bases 9.2-15 9.2.5.2 System Description 9.2-15 9.2.5.3 Safety Evaluation 9.2-16 9.2.6 Condensate Storage Facilities 9.2-17 9.2.6.1 Design Bases 9.2-17 9.2.6.2 System Description 9.2-17 9.2.6.3 Safety Evaluation 9.2-18 9.2.6.4 Tests and Inspections 9.2-18 9.2.6.5 Instrumentation Application 9.2-19 9.2.7 Turbine Building Closed Icop Cooling Water System 9.2-19 9.2.7.1 Design Bases 9.2-19 9.2.7.2 System Description 9.2-19 9.2.7.3 Safety Evaluation 9.2-20 9.2.7.4 Tests and Inspections 9.2-20 9.2.7.5 Instrumentation Application 9.2-20 9.2.8 Hain Chilled Water System 9.2-21 9.2.8.1 Design Bases 9.2-21 9.2.8.2 System Description 9.2-21 9.2.8.3 Safety Evaluation 9.2-21 9.2.8.4 Tests and Inspections 9.2-22 9.2.8.5 Instrumentation Application 9.2-22 9.2.9 Reactor Building Standby Ventilation System (RBSVS) and Control Room Air Conditioning (CRAC) Chilled Water System 9.2-22 9.2.9.1 Design Bases 9.2-22 9.2.9.2 System Description 9.2-23 9.2.9.3 Safety Evaluation 9.2-23 9.2.9.4 Tests and Inspections 9.2-25 9.2.9.5 Instrumentation Application 9.2-25 9.3 PROCESS AUXILIARIES 9.3-1 9.3.1 Compressed Air Systems 9.3-1 9.3.1.1 Design Bases 9.3-1 9.3.1.1 Systen Description 9.3-1 9.3.1.3 Safety Evaluation 9.3-1 9.3.1.4 Tests and Inspections 9.3-3 9.3.1.5 Instrumentation Application 9.3-3 9.3.2 Process Sampling System 9.3-3 9.3.2.1 Design Bases 9.3-3 9.3.2.2 System Description 9.3-4 9-2

SNPS-1 FSAR p 9.2 WATER SYSTEMS \,j Service Water System 9.2.1 9.2.1.1 Design Bases During normal operation, the service water system is designed to provide cooling water to the secondary side of the reactor building closed loop cooling water (:RECLCW) heat exchangers, the drywell cooling booster heat exchangers, the secondary side of the turbine building closed loop cooling water (TBCLCW) heat exchangers, the RBSVS and control room air conditioning (CRAC) chilled water condensers, and to the main chilled water condensers. The service water system is also designed to provide cooling water to the residual heat removal (RHR) heat exchangers, to dissipate reactor decay heat during a scheduled shutdown or accident conditions. The system also provides cooling water to the diesel engine coolers, emergency makeup water to the spent I fuel pool, and emergency cooling water to the ultimate cooling connection. The service water system is split into two independent subsystems: reactor building (RB) and turbine building (TB) .

  ) With the exception of the branches supplying cooling water to the

(~'/ x_ TB and to the main chilled water, the service water supply system is a safety related system and is ASME III Code Class 3 and Seismic Category I. 9.2.1.2 System Description Reactor Building Service Water Subsystem The RB service water subsystem is shown on Fig. 9.2.1-1A. It consists of four 50 percent capacity, motor driven, vertical wet pit centrifugal pumps, associated piping, valves, and instrumentation. Each pump has a capacity of 8600 gpm and taxes suction at the screenwell. An automatic, selt-cleaning, rotary type strainer is provided at the discharge of each pump. The four RB service water pumps discharge into a common header, which l is provided with two normally open, motor operated valves that can isolate either half of the system from the other to ensure that at least 50 percent of the pumping capacity will be available at all times. Two redundant supply lines, one taken from each side of the double isolation valves on the pump discharge header, are used to provide a reliable source of service water to equipment required for safe shutdown and cooldown. (r) A list of components served by the RB service water subsystem is provided in Table 9.2.1-1. l 9.2-1 Revision 20 - November 1980

SNPS-1 FSAR Since water to supply the service water system comes from Long Island Sound, the supply of service water is inexhaustible and , limited only by the capacity of the pumps and heat exchangers. I Water temperature in Long Island Sound varies f rom 29 to 77 F; the design temperature used for the service water system is 77 F. The mean tide range at the screenwell is 5.9 ft, and the spring tide range is 7.0 ft with the mean tide level at 2.9 ft above mean low water level; an extreme low water level of S.9 ft below mean low water level is used to determine minimum pump submergence and net positive suction head. The highest tide on record was 11.5 ft above mean low water level and the maximum design basis stillwater level has been set at 26.0 ft above mean low water level; all equipment necessary to maintain the service water system operational nas been protected to 26.0 ft above mean low water level and associated waves (see Section 3.4) . High and low water level considerations are discussed in Section 2.4. l Service water supply consists of two separate lines which penetrate the secondary containment independently. Eacn supply header branches off to supply >n e of the two RBCLCW heat l exchangers, one of two drywell cooling booster heat exchangers, and one of the two RHR heat exchangers. A separate return line l for each RHR and RBCLCW heat exchanger, each provided with an unobstructed stand pipe, ensures that a continuous 11ow or cooling water can be passed through the exchangers at any time. Admission of water to either RHR heat exchanger, wnen necessary to remove decay heat, is accomplished by opening a motor operated valve. The flow of service water through any of the heat exchangers in the RB can be stopped remotely by closing the motor operated valve on the exchanger discharge pipe. Water trom the l heat exchangers in the RB is discharged to the circulating water discharge tunnel. A single pipe connected to each service water supply line in the RB is used to provide emergency cooling water to the ultimate cooling connection. An operator action is necessary to open two motor operated valves, installed in series, so that no service water can accidentally be supplied to the ultimate cooling connection. A similar arrangement is used to supply emergency makeup water to the spent fuel pool. Similarly, service water cannot be sent to the fuel pool except ny intentional operator action. A testable check valve is provided in the pipe supplying emergency service water to the ultimate cooling connections for the purpose of preventing back flow of potentially radioactive water into the service water system. The check valve is tested manually at regular intervals to verify that the service water supply to the ultimate cooling connection remains unobstructed. Two redundant water supply pipes are used to provide each of the three diesel engine coolers with two separate sources of service l water. These lines also supply the tour RBSVS and CRAC chilled 9.2-2 Revision 20 - November 1980

SNPS-1 FSAR water condensers. One of the supply pipes provides cooling water to two chilled water condensers while the remaining pipe supplies the other two. A separate, return pipe per diesel cooler ensures that a continuous tiow of service water can be passed through the coolers at all times. Tne tull flow of water through eacn cooler is controlled by an air operated valve. Two separate return pipes are used to discharge cooling water from the R8SVS chilled water condensers through temperature control valves. Service water from the diesel engine coolers and the RBSVS and CRAC chilled water system is discnarged to the circulating water discharge tunnel. A single pipe connected to each of the supply pipes to the RbSVS and CRAC chilled water condensers is used to provide water to the main chilled water condensers. Service water supply to the main chilled water condensers is not sarety related. During an accident, service water supply to the main ch111ed water condensers is automatically isolated by double isolation volves. During normal operation, two or the rour Rb service water pumps are in operation to supply cooling water to one of tne RECLCW heat exchangers, one of the four RBSVS and CRAC chillec water condensers, one of the two drywell cooling booster heat exchangers, and to the main chilled water condensers. No service water is required for residual heat removal or diesel engine cooling during normal operation. The two remaining reactor (g Du11 ding service water pumps are kept on automatic standby.

'%)

During a normal or scheduled shutdown, two or three Rb service water pumps operate to supply cooling water to one of the RBCLCW heat exchangers, to one or both RHR heat exchangers, to one or the RBSVS and CRAC cn111ed water condensers, to one or the drywell cooling booster heat exchangers, and to the main chilled Wdter Condensers. The tourth pump remains in automatic standby. Radiation monitors are installed on the secondary (service water) side of the RHR neat exchangers, on tne primary (demineralized l water) side or the RBCLCW heat exchangers, and on the service water drains subsystem. Eftluents from the service water system i are diluted in tne circulating water discharge tunnel. The ! environmental acceptance of .these effluents is uiscusse d in

Section 2.4.12.

Since the service water system utilizes seawater directly Irom Long Island Sound, the only pr&ctical means of corrosion control

is the selection of suitable materials. Thus, all safety related

! piping and the internals of all service water system components 2 are fabricated from corrosion resistant materlaJs tnat are i designed for service in a seawater environment. l 1 9.2-3 Revision 20 - November 1980 i L

SMPS-1 FSAR Turbine Building Service Water Subsystem The turbine building service water subsysten is shown on Figure 9.2.1-1B. Service water supply to the turbine building consists or three 50 percent capacity motor-driven pumps each rated at 3,000 gpm and associated piping, valves, and instrumentation. The three pumps discharge to a common hedder that branches off to each of the two TBCLCW heat exchangers. Service water supply to the TB is not safety related. Service water supply to the TB is isolated Irom the Rb service water system by double isolation motor operated valves. During normal operation these valves are locked closed. In an absolute necessity, RB service water flow could be supplied by the TB system. However, the double isolation motor-operated valves receive signals to close during a loss-of-coolant accident or a loss of offsite power. During normal operation, two of three TB service water pumps are in operation to supply cooling water to the TBCLCW heat exchangers. The taird pump remains in automatic standby. During a normal or seneduled shutdown, one of the three TB service water pumps is in operation to supply cooling water to the TBCLCW heat exchangers. One of the two non-operating pumps remains in automatic standby. A list of components by the TB service water system is provided in Table 9.2.1-1. 9.2.1.3 Safety evaluation The RB service water subsystem is capable of cooling essential equipment through two redundan+. headers. It is designed so that no single failure of any component will prevent the system from performing its intended safety function. During a design basis accident, each supply pipe is capable of providing sufficient cooling water to the follcwing equipment, essential to the safe shutdown of the plant:

1. One residual heat removal (RHR) exchanger.

I O 9.2-4 Revision 20 - November 1980

SNPS-1 FSAR One reactor building closed loop cooling water (RBCLCW) O 2. heat exchanger.

3. All three diesel engine coolers.

4. Two reactor building standby ventilation system (RBSVS) and control room air conditioning (CRAC) chilled water l condensers.

The RB service water subsystem is designed so that, following an l accident, not less than two service water pumps will be in operation to cool the equipment listed above. No service water is supplied to the TB or the main chilled water system following an accident. ! The design margin for the above mentioned components is based on the maximum duty required for each. In addition, sufficient redundancy is provided in the individual systems to accommodate loss of a component. On an accident signal, such as low reactor coolant level or nigh primary containment pressure, tne following automatic operations take place:

1. The isolation valves on the service water pump discharge

() header close, dividing the nain supply pipes.

2. Service water supply to the RHR and drywell cooling booster heat exchangers and the main enilled water system is isolated.

3. The TB service water system is isolated from the RB service water system.

4. Tne motor operated valves in the supply pipes to both RBCLCW heat exchangers are interlocked in the open position.

5. All RB service water pumps start.

6. All RBSVS and CRAC condensing water pumps start.

After the above automatic operations, operator action is required to supply service water to the RHR heat excnanger when it becomes necessary. Upon loss of all offsite power without LOCA, the following operations take place automatically:

1. Service water pumps are tripped.

O (_,/ 2. The standby diesel generators are started. i 9.2-5 Revision 20 - November 1980 l

SNPS-1 FSAR

3. When diesel generator operating voltage is reached, the diesel generators are connected to their respective buses. (Saf ety related instrumentation is supplied trom the emergency buses as soon as power from standby diesel generators is available.)

l 4. The RB service water pumps on each emergency bus start 12 sec atter the diesel generator is connected to the bus.

5. Valve operators move automatically into the position for post-accident condition with the exception of the drywell cooling booster heat exchangers which remain in position.

In the event of an accident, followed by loss or offsite power, all operations described above take place automatically. In addition, the following events take place:

1. Valve operators move au*.omatically into the proper position for pr.st-accident condition.

2. Operator action is necessary to supply service water to the RilR heat exchangers.

l The FB service water subsystem is designed so that it is capable of accommodating any single failure of a component within the system or of a component in another related system without affecting the overall system capability of effecting plant shutdown and cooldown or post-accident heat dissipa tion . Operator actions may be required to isolate a given failed component from the remainder of the service water system and to transfer cooling to tz.e redundant cooling subsystem. A single active or passive failure of a component in the service water system initiates an alarm, e.g., diesel generator trip, flow, level, pressure or temperature condition in the control room. Upon annunciation, the operator responds by remote manual initiation of the necessary valve action to isolate the failea component and to make use of the proper independent redundant subsystem. Following a design basis accident with coincident loss of offsite power, the maximum heat load on the service water system, with minimum safeguard equipment operating, would be 129,300,000 Btu /hr resulting from the following safety related sources: O 9.2-6 Revision 20 - November 1980

SNPS-1 FSAR System Approx Max Heat Load RHR 89,300,000 Btu /hr RBCLCW 10,000,000 Btu /hr RBSVS & CRAC Chilled Water 8,000,000 Btu /hr Diesel Generator Cooling 22,000,000 Btu /hr Total 129,300,000 btu /hr The RB service water subsystem has been sized to continuously l provide sufficient cooling water flow to these components to remove the above heat rejection rates with the maximum service water inlet temperature of 80 F. The auxiliary system components have also been sized to accommodate the accident heat rejection requirements (see Table 9.2.1-2) . As stated in Section 9.2.5, the ultimate heat sink for the station is the Long Island Sound. By observation, the Sound is an essentially infinite cooling water source and heat sink in comparison to the requirements of tne service water system. /~) Comparison of the available and required NPSH of the service \ms/ water pumps is provided in Table 2.4.11-2.

9. 4.1. 4 Tests and Inspections The service water pumps operate during normal plant operation.

The standby service water pumps are started at regular intervals to assure operability. Isolation valves are tested periodically. During a refueling shutdown, any system component, subsystem, etc., can be isolated, tested, or disassembled by making use of its redundant counterpart. 9.2.1.5 Instrumentation Application Pressure at the pump discharge header and through the heat excnangers is continuously monitored. Radiation monitors are installed at tne oischarge of the residual heat removal heat exchangers. Numerous local pressure and temperature indicators are provided. Instrumentation for the service water system is described in more detail in Section 7.3. 9.2.2 Reactor Building Closed Loop Cooling Water System 9.2.2.1 Design Bases

-    The RBCLCW system is designed to provide both safety and non-(w)

\' safety related systems and equipment with cooling water. normal operation of the plant, the RBCLCW system provide s cooling During 9.2-6a Revision 20 - November 1980

SNPS-1 FSAR water to reactor auxiliary equipment and other equipment inside the RB. That portion of the RBCLCW system designed to supply cooling water to safety related systems and equipment during accident conditions is safety related and Seismic Category I. RBCLCW supply to the reactor recirculating pumps seal coolers, motor bearings and winding coolers, although not required during an accident, is safety related and Seismic Category I. Remotely actuated isolation valves are provided for the purpose of isolating sarety related portiong of the RBCLCW system from nonsafety related portions of the system following an accident signal. 9.2.2.2 System Description The RBCLCW system is a redundant closed loop system, providing safety related components with a reliable source of cooling water. Branches taken off either redundant loop serve nonsatety related components. The RBCLCW system is shown on Figs. 9.2.2-1A and 18. The RBCLCW system is designed to provide cooling water to the following equipment: Safety Related Equipment

1. Residual heat removal pumps seal coolers

2. Spent fuel pool cooling water heat exchangers

3. Reactor recirculation pumps seal coolers, motor winding coolers, and motor bearing coolers. (Equipment operation not required during an accident) .

O 9.2-6b Revision 20 - November 1980

A %/ ShPS-1 PSAR TABLE 9.2.1-1 SrRVICE WATER SYSTEM COMPONt.NT DESIGN DATA Nominal Capacity Normal op-Each hormal eration Hot Normal Actual Flow Provided Vuantity Nom) 0;wration Standtv Shutdown IDCA heactor Building Service Water Pumps 14 6.600 (2)12,290 (.s) 2 5, b60 (J) 24,920 (2)17,090 Wrbine Building Service Water Pumps 3 o,000 (2316,9 u O (2) 1e ,905 (1) 4,940 -- 29,230 42,765 3e,doo 17,090 Beactor Building Subsystem Components: Reactor Building Service Water Strainers 4 250 (2) 500 (3) 750 (3) 750 (2) 500 Diesel Generator Jacket Coolers 3 700 -- -- -- (3) 2,100 Residual Heat hemo 'l Heat Exchangers 2 b,000 -- (2)16,000 (2)16,000 (1) 8,000 Reactor Building closed Loop Cooling 2 6,370 (1) 6,300 (1) Note A (1) 4,100 (1) 1,150 Water heat Exchangers keactor Building Standby Ventilation 4 525 (1) 525 (1) 525 (1) 525 (4) 1,050 System Water Chiller Condensers Main Chilled Water System Water Chillea 3 1,500 (2) 3,000 (2) 3,000 (1) 400 -- Condensers Drywell Cooling Booster heat Exchangers 2 1,460 (1) 1,500 (1) 1,500 (1) 1,500 -- 11,825 Note A 23,275 12,800 Wrbine Building Subsystem Components: Turbine Building Service Water Strainers 2 #20 (1) 420 (1) 420 (1) 420 -- Cire Water Pump Bearing Cboling 4 b (4) 32 -- (4) 32 -- Fish Retention Pool 1 185 (1) 185 (1) 185 (1) 185 --

    'narbine Building Closed Loop Cooling               2      14,200        (1)16,000      (1)16,000             (1) 9,000              --

Water Heat Exchangers Vacuum Priming Pinnps Seal Water Coolers 3 100 (3) 300 (3) 300 (3) 300 -- 16,937 16,905 9,937 -- Total System Flow Requirements 2d,762 Note A 33,212 12,d00

1. Note A: Flow requirements dependent on plant auxiliaries.

2. Numbers in parentheses are number of components operating.

1 of 1 hevision 20 - November 19d0

l l l i i SNPS-1 FSAR TABLE 9.2.1-2 i , AUXILIARf SYSTEMS COMPONE?fP DESIGN DATA Approx. Post-LOCA Design Design Service Heat Load (Btu /hr) Component Heat Load (Btu /hr) Water Flow (apm) per Component (a> RHR Heat Excnanger (each) 41,400,000(a> 8,000 89,300,000ta> RBCLCW Heat Exchanger (each) 40,000,000 6,370 10,003,000 RBSVS & CRAC Chiller Condensers teach) 4,065,000 525 4,000,000 Diesel Generator Coolers (each) 11,790,000 700 11,000,000 O DFor minimtsu safeguard equipraent operating. (a 3 Design heat load based on shutdown mode with 125 F Hx inlet. Post-LOCA heat load based on 193 F 3x inlet (See Fig. S.5.7-1A) . 1 of 1 Revision 20 - November 1980

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                                                    ' ' 2. 5- I A F I G. 9. 2.1 - 1 B SERVICE WATER SYSTEM SHOREHAM NUCLEAR POWER STATION-UNIT I FINAL SAFETY AN ALYSIS REPORT REVISION 20-NOVEMBER 1980

i SNPS-1 FSAR 10.2 TURBINE-GENERATOR 10.2.1 Design Bases The turbine is an 1,800 rpm, tandem compound, four-flow, two-stage reheat steam turbine with 43 in. last-stage buckets. The maximum capability of the turbine is 880,256 kW gross with valves wide open and inlet steam conditions of 965 psia and 99.6 percent quality while exhausting at 1.5 in. Hg abs with 0 percent makeup, and extracting for six stages of feedwater heaters. The generator is a 978,000 kva, 1,800 rpm, direct connected, 3-phase wye connected, 60 Hz, 24,000 V, liquid cooled stator, hydrogen cooled rotor, synchronous generator rated at 0.90 pf, 0.58 short circuit ratio at a maximum hydrogen pressure of 60 psig. The compact Alterrex excitation system consists of a 60 Hz, 1,800 rpm air cooled Alterrex generator and liquid cooled rectifiers with static regulation equipment. The exciter is rated for a maximum output of 1860 kW at 500 v. l The turbine generator unit control is effected through an electrohydraulic control (EHC) system capable of controlling the speed, load, pressure, and steam flow under steady state and transient conditions. The turbine generator is normally base loaded; however, the design allows for the unit to operate at reduced load. The turbine-generator unit, a General Electric Company (GE) design, is built in accordance with GE standards and codes. The moisture ceparator/ reheaters are built in accordance with ASME j Section VIII. i The turbine generator unit and the associated steam and power [ conversion system is capable of a 25 percent load reduction without producing a reactor trip by dumping steam into the j condenser through the turbine bypass system (Section 10.4.4) . A l 10 percent step load change or a five percent per minute ramp i load change is accommodated by the reactor control system without l bypassing the steam into the condenser. { Sudden loss of generator load due to an external system fault results in a rapid closure of the main steam control valves and the combined intermediate valves. I 10.2.2 Description The turbine consists of one double flow high pressure cylinder and two double flow low pressure cylinders. It includes two moisture separator / reheaters located on the operating floor on l \_/ l ! 10.2-1

SNPS-1 FSAR either side of the turbine. Main steam is used to heat the exhaust steam from the high pressure turbine ^ylinder while baffles are used to remove moisture from the exhautt steam. The turbine shells are equipped with extraction nozzles to provide extraction steam for six stages of feedwater heaters. Each high pressure steam line to the high pressure cylinder contains a main steam turbine stop valve plus an automatically controlleu main steam turbine control valve. Combined intermediate valves (each comprised of an intercept valve and an intermediate stop valve) are provided in the discharge line from the moisture separator / reheaters to the low pressure turbine cylinders. The turbine uses an electrohydraulic control (EHC) system consisting o' conventional governing devices (two initial pressure regulators, speed governor, and startup control devices) , emergency devices for turbina and plant protection, motoring protection, thrust bearing wear detector, electrical f ault protection relays, and special control and test devices. The EHC system operates the main stop valves, control valves, bypass valves, combined intermediate valves, and other protective devices. Tne EHC system uses solid state control techniques in combination witn a high pressure hydraulic system which is completely independent of the turbine lubricating system. The turbine generator unit is protected against dangerous overspeed by redundant speed control systems. The EHC system controls the speed during normal and transient conditions. If the EHC system speed controls fall, either a mechanical overspeed or a backup overspeed system trips the turbine generator unit. The peak transient speed, with failure of the normal EHC system and the mechanical overspeed trip but with proper operation of the backup overspeed trip, is approximately 111 percent of rated speed. For overpressure protection of the turbine exhaust hoods and the condenser snells, two rupture diaphragms are provided in each low pressure turbine exhaust hood which rupture at approximately 5 psig. Condensate sprayed into the exhaust hood is used for l overtemperature protection. Additional protective devices include exhaust hood high temperature alarm and trip and a pilot dump valve for protective closing of extraction nonreturn valves. The turbine lubricating oil system supplies oil for lubricating the bearings. A bypass stream of turbine lubricating oil flows continuously through an oil conditioner to remove water and other impurities. The generator is sized to accept the gross output of the turbine. The generator is equipped with an excitation system, hydrogen control system, and stator liquid cooling system. 10.2-2 Revision 20 - November 1930

SNPS-1 FSAR Acceptance Criteria

1. The applicable general acceptance criteria, as listed in Section 14.1.3.6, will be met.

2. Battery chargers will provide battery float and equal-izing charge while maintaining the normal amount of load.

3. The batteries will maintain normal mount of load upon loss of battery chargers.

4. Batteries will meet their rated amp-hr capacities and the battery chargers will restore battery charge in the raed recovery time. Battery parameters remain within acceptance limits during the test.

14.1.3.7.7 Service Water System Preoperational Test obiective To verify the capability of the Service Water system to deliver cooling water to safety related components serviced by this system. Heat removal capability w211 be verified under separate testing during startup testing. Prerequisites

1. The applicable general prerequisites, as listed in Sec-tion 14.1.3.4, will be met.

2. The appropriate sections of the rollowing systems necessary to this test will be operational:

a. Reactor building closed loop cooling system

b. Turbine building closed loop cooling system

c. Residual heat removal heat exchangers

d. Reactor building , ventilation system chilled water condensers

e. Emergency diesel generators

f. Drywell cooling booster heat exchangers l Test Method

1. Normal flow paths will be tested to verify that design flow rates are available and flow balancing restrictions are functioning properly.

2. Normal system functions including automatic strainer backflushing and standby-start features will be verified.

Acceptance Criteria f-% \ s/

1. The applicable general acceptance criteria, as listed in Section 14.1.3.6, will be met.

14.1-17 Revision 20 - November 1980

SNPS-1 FSAR

2. System automatic functions will meet design require-ments.

14 .1.3.7 .8 120 V A-C RPS-MG Sets Preoperational Test Obiective Tb demonstrate the ability of the RPS-MG sets to provide 115 V a-c to the reactor protection system buses. Prerequisites

1. The applicable general prerequisites, as listed in Sec-tion 14.1.3.4, will be met.

2. The appropriate sections of the 480 V a-c power supply system necessary to this test shall be operational.

3. Sufficient equipment installed to provide at least partial load on buses during this test.

Test Method

1. The RPS buses will be energized from the RPS-MG sets.

2. Voltage and frequency regulation of each RPS-MG set will be checked under varying load conditions.

3. Power delivery during short term interruption of normal power supply will be checked.

4. Interlocks between RPS-MG sets and the alternate power supply will be tested.

5. The ability of the alternate power supply to deliver proper voltage will be verified.

Acceptance Criteria

1. The applicable general acceptance criteria, as listed in Section 14.1.3.6, will be met.

2. The RPS-MG sets will deliver and maintain RPS bus voltage and frequency in accordance with design requirements.

3. The alternate power supply will provide RPS bus voltage in accordance with design requirements.

4. Transfer between the RPS-MG sets and the alternate power supply will function in accordance with design require-ments.

O 14.1-18 Revision 4 - February 1977}}

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