ML20071N568
| ML20071N568 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 05/24/1983 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20071N538 | List: |
| References | |
| 22A7007, NUDOCS 8306070114 | |
| Download: ML20071N568 (515) | |
Text
{{#Wiki_filter:-,m.s- e-- l UNITED STATES 0F AMERICA
. NUCLEAR R E G U.L A T O R Y COMMISSION I In the matter of )
General Electric Company ) Docket No. STN 50-447
- Standard Plant )
AMENDMENT NO. 15 TO APPLICATION FOR REVIEW OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)
- General Electric Company, applicant in the above captioned proceeding, hereby files Amendment No. 15 to the 238 Nuclear Island General Electric Standard Safety Analysis Report (GESSAR II).
1 Amendment No. 15 further amends GESSAR II by:
- 1. Furnishing additional responses in the Commission's letters, dated August 25, 1982, October 5, 1982, November 15, 1982, and January 31, 1983.
- 2. Providing resolution to selected outstanding and confirmatory issues identified by in the GESSAR II Safety Evaluation Report (NUREG-0979).
- 3. Completion of Subsection 1.9.1 which clearly identifies GESSAR II/
FSAR interfaces.
- 4. Clarifying portions of the text where obvious discrepancies exist.
Respectfully submitted, General-Electric Company l l i By: s/Glenn G. Sherwood
- Glenn G. Sherwood, Manager Nuclear Safety & Licensing Operation STATE OF CALIFORNIA )
COUNTY OF SANTA CLARA ) ss: On this 24th day of May in the year 1983, before me, Karen S. Vogelhuber, i Notary Public, personally appeared Glenn G. Sherwood, personally proved to me on the basis of satisfaction evidence to be the person whose name is subscribed to this instrument, and acknowledged that he executed it. t i By: s/ Karen S. Vogelhuber Notary Public - California '
- Santa Clara County My Commission Expires
- December 21, 1984 l 8306070114 830524 175 Curtner Avenue l PDR ADOCK 05000447 San Jose, CA 95125 K PDR GGS:hmm/005243 5/24/83
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 O kJ INSTRUCTIONS FOR FILING AMENDMENT NO. 15 Remove and insert the pages listed below. Dashes (----) in the remove or insert column indicate no action required. Remove Insert Chapter 1 1.2-66, 1.2-67, 1.2-69, 1.2-66, 1.2-67, 1.2-69, 1.7-120, 1.8.6-1, 1.8.6-2, 1.2-69a, 1.7-120, 1.8.44-1/1.8.44-2, 1.8.44-3, 1.8-i through 1.8-x, 1.8.6-1/1.8.6-2, 1.8.44-4, 1.8.52-1, 1.8.52-2, 1.8.44-1/1.8.44-2, 1.8.52-1/1.8.52-2, 1.8.52-3/1.8.52-4, 1.8.63-3, 1.8.63-3, 1.8.82-1/1.8.82-2, 1.8.82-1/1.8.82-2, 1.8.92-4, 1.8.92-4, 1.9-la, 1.9-4.1-11, 1.9-la, 1.9-4.1-11, 1.9-4.3-2, 1.9-4.1-11a, 1.9-4.3-2, 1.9-4.3-2a, 1.9-4.3-3, 1.9-4.5-1/1.9-4.5-2, 1.9-4.3-3, 1.9-4.3-3a, 1.9-4.5-1/ 1.9-4.6-1, 1.9-4.8-2, 1.9-4.9-1, 1.9-4.5-2, 1.9-4.6-1, 1.9-4.8-2, 1.9-4.11-2, 1.9-4.12-2, 1.9-4.14-3, 1.9-4.8-2a, 1.9-4.9-1, 1.9-4.9-la, 1.9-4.14-4, and 1.9-4.19-1 through 1.9-4.11-2, 1.9-4.12-2, 1.9-4.12-2a, 1.9-4.19-9/1.9-4.19-10 1.9-4.14-3, 1.9-4.14-4, 1.9-4.19-1, 1.9-4.19-la, 1.9-4.19-lb, 1.9-4.19-2, 1.9-4.19-2a, 1.9-4.19-2b, 1.9-4.19-3, 1.9-4.19-4, 1.9-4.19-5, 1.9-4.19-6, 1.9-4.19-6a, 1.9-4.19-6b, 1.9-4.19-6c, 1.9-4.19-7, 1.9-4.19-7a through (" 1.9-4.19-7h, 1.9-4.19-8, and 1.9-4.19-9/1.9-4.19-10 Appendix 1A 1A.9-2 1A.9-2 Chapter 3 3.2-22, 3.5-19, 3.6-20, 3.2-22, 3.5-19, 3.6-20, 3.7-15, 3.7-23, 3.7-44, 3.9-37,* 3.7-15, 3.7-23, 3.7-23a, 3.8-61, 3.9-89, 3.9-283, 3.7-44, 3.8-61, 3.9-89, 3.9-284, 3.9-319/3.9-320,- 3.9-283, 3.9-284, 3.10-55, 3.10-57, 3.10-61, 3.9-319/3.9-320, 3.10-55, 3.10-73, 3.10-74, 3.10-75, 3.10-57, 3.10-61, 3.10-73, 3.10-77, and 3.10-83 3.10-74, 3.10-75, 3.10-77, and 3.10-83
- *This page should have been deleted in Amendment 14.
Amendment 15 May 24, 1983
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I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Remove Insert [V3 ~ Appendix 3B 38-1 and 3B-92 3B-1 and 38-92 1 Appendix 3BA - 3BA.8-9/3BA.8-10,* 3BA.12-31, 3BA.12-32, 3BA.12-31, 3BA.12-32, and 3BA.14-28 and 3BA.14-28 Appendix 3H Appendix 3H_ tit 1e page Appendix 3H title page and 3H-i/3H-ii Chapter 4 4.4-11, 4.5-1, 4.5-2, 4.5-2a, 4.4-11, 4.5-1, 4.5-2, 4.5-7, aod 4.5-8 4.5-2a, 4.5-7, and 4.5-8, Chapter 5 A 5.2-39, 5.2-40, and 5.4-1 5.2-39, 5.2-40, 5.4-1, and tj 5.4-la Chapter 6 6.2-v, 6.2-43, 6.2-70, 6.2-v, 6.2-43, 6.2-70, 6.2-101 through 6.2-106, 6.2-101 through 6.2-106, 6.2-116, 6.2-146, 6.2-147, 6.2-106a, 6.2-116, 6.2-146, 6.2-161, 6.2-164, 6.2-195, 6.2-147, 6.2-147a, 6.2-161, 6.2-196, 6.2-212, 6.2-243, 6.2-164, 6.2-195, 6.2-196, 6.3-1, 6.5-v, 6.5-1, 6.5-2, 6.2-212, 6.2-243, 6.3-1, 6.5-3, 6.5-5, 6.5-16, , 6.3-la, 6.5-v, 6.5-1, 6.5-2, 6.5-45 through 6.5-48, 6.5-3, 6.5-5, 6.5-16, 6.5-45 6.5-61/6.5-62, 6.8-i/6.8-ii, .through 6.5-48, 6.5-61/6.5-62, 6.8-1, 6.8-2, 6.8-3, 6.8-9, and '6.8-i/6.8-ii, 6.8-1, 6.8-2, 6.8-10 6.8-3, 6.8-9 and 6.8-10 1 Chapter 7 i 7.2-83a, 7.4-30, 7.2-83a, 7.4-30, and and 7.4-66 7.4-66 9'
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*This page should have been deleted in Amendment 14. -
Amendment 15 May 24, 1983
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V!(_ ! < )_ GESSAR II 238 NUCLEAR ISLAND
-22A7007 Rev. 15
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Remove < Insert Chapter 8
.,~ 8.1-4, 8.1-8,'8.1-9, F. , '8.1-4', 8.1-8, 8.1-9,- _
8.1-14, 8.2-1 through
-8.1-14, 8.2-1, 8.2-2," 8.2-4, 8.3-1, 8.3-3, 8.2-3/8.2-4, 8.3-1, 8.3-3a, 8.3-3b, 8.3-10, f . / '8.3-3, 8.3-3a, 8.3-3b, 8.3-22, 8.3-23,.8.3-26, ~W . "
8.3-10,.8.3-22, 8.3-23, 8.3-28, 8.3-29, 8.3-34,
'p .8.3-26,l8.'3-28, 8.3-29, 8.3-45, 8.3-81, 8.3-84, h
8.3-34,;8.3-45, 8.3-81, 8.3-94, 8.3-94a, 8.3-95, , F 8.3-84,.8.3-94,.8.3-95, 8.3-98, 8.3-99, 8.3-105, . , R- ^8.3-98,'8.3-99, 8.3-105,= 8.3-106, 8.3-130, 8.3-131,
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8.3-106, 8.3-130, 8.3-131,- 8.3-133/8.3-134, 8.3-135,
^ 8.'3-133/8.3-134, 8.3-135, 8.3-136, 8.3-137/8.3-138,
, y 8.3-136, 8.3-137/8.3-138, 8.3-142c, 8.3-169, 8.3-142c, 8.3-169, and 8.3-181/8.3-182, l r [ 8.3"181/8.3-182 8.3-182a and 8.3-182b Chapter 9 ! 9.1-32, 9.~4-1, 9.4-2, 9.1-32, 9.4-1, 9.4-2, 9.4-5, i j 9.4-5, 9.4-10a,. 9.4-10a, 9.4-13, 9.4-14, i '9.4-13, 9.4-14, 9.4-17, 9.4-17, 9.4-18, 9.4-27, i 9.4-18, 9.4-27, 9.4-36, 9.4-36, 9.4-37, 9.4-39, i 9.4-37, 9.4-39, 9.4-47, 9.4-47, 9.4-47a, 9.4-49, ! 9.4-49, 9 4-50, 9.4-55 9.4-50, 9.4-55~ through 9.4-58, *
;' through 9.4-58, S.4-61, 9.4-61, 9.4-63, 9.4-75, ! _. 9.4-63,-9.4-75, 9.4-76, 9.4-76, 9.4-77, 9.4-83, ,
9.4-84, 9.4-87, 9.4-88,
, t n s 9.4-77, 9.M-83, 9f4-87, i( y\ 9 '9.4-88, 9.'4-90, 9.4-95, 9.4-90, 9.4-95, 9.4-146,
? 9.4-146, 9.5-5, 9.5-21, 9.5-5, 9.5-8a, 9.5-21, N Tp e p s9.5-24;o9.5-27, 9.5-53, 9.5-24,_9.5-27, 9.5-53, and ,
- y aad 9.5-54' 9.5-54 L;
3 ' Appendix 9A
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. 47 ( 7\ 7 ! n; L' 9A.'4'-35 through 9A.4-39, 9A.4-35 through 9A.4-39, b' Y
-49A.4-42, 9A.5-24, and 9A.4-39a through 9A.4-39d,
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~9A.5-39/9A.5-40 9A.4-42, 9A.5-24, and 9A.5-39/9A.5-40
- Chapter 10
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,j J Chapter 11 , : ~ ~;a
[o ' '11.3-15, 11.3-26b, 11.3-15, 11.3-26b, 11.4-2, 11.4-2, 11.5-32, 11.4-2a, 11.5-32, and 11.5-34 - %- and 11.5-34 i .< 1 *These pages should have been deleted in Amendment 14. t ,'
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GESS5RII 22A7007 k, -., / C j 238 NUCLEAR ISLAND Rsv.s15 f' , Remove Insert - i 11 4 Chapter 12 r/ . +
'[ ' 12.3-48 12.3-48 T' Chapter 14 14.1-3 through 14.1-25/14.1-26*
Chapter 15 15.7-37, 15.7-39, 15.7-40, 15.7-37, 15.7-39, 15.7-40, and and 15.7-42 15.7-42,
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,' Appendix 15A ( ".
o,I ) ;a - * . 15A.6-20 and 15A.6-20 and 15A.6-38 '
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Chapter 18
, [18A.3-1through /'
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i 19.1.1-1, 19.3-i 19.1.1 1, 19.3-i through through 19.3-iv, 19. 3-xv i f/19. 3-xv i i i , 19. 3.1. 6-2, 19.3.3.52-1/19.3.3.52-2, 19. 3. 3. 52-L throug5 19. 3. 3. 52-18, 1 19.3.3.54-1, 19.3.3.54-2, ' 19.',3.3.54-1/19.3.3.54-2, 19.3,3.80-1/19.3.3.80-2, 19,3.3.80-1/19.3.3.80-2, 19.3.3.82-3, 19.3.3.112-2, 19.3.3.82-3, 19.3.3.112-2, 19.3.3.127-1/19.3.3.127-2, 19.3.3.'127-1/19.3.3.127-2, 19.3.3.139-2, 19.3.3.145-1/ 19.3.3.139-2, 19.3.3.145-1/ 19.3.3.145-2, 19.3.3.148-1/ 19.3.3.145-2, 19.3.3.148-1/ 19.3.3.148-2, 19.3.3.150-1/ 19.3.3.148-2, 19.3.3.150-1/ 19.3.3.150-2, 19.3.3.151-1/ 19.3.3.150-2, 19.3.3.151-1/ 4 19.3.3.151-2, 19.3.4.8-2, 19.3.3.151-2, 19.3.4.8-2, 19.3.5.1-1/19.3.5.1-2, 19.3.5.1-1/19.3.5.1-2,< 19.3.5.11-1, 19.3.6.5-1, 19.3.5.11-1, 19.3.6.5-1,-
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19.3.6.13-1, 11.'3.'6.13-2, 19.3.6.13-1/19.3.6.13-2, , s
;' ,19.3.6.20-1, 19.3.6.21-1, 19.3.6.20-1, 19.3.6.21-1, ,
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*These pages should have been deleted in Amendeent 14.1 i i e ,
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; ,- ~ Amendment 15 'May 24, 1983
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 g Remove Insert Chapter 19 (Continued) 19.3.6.24-1/19.3.6.24-2, 19.3.6.24-1/19.3.6.24-2, 19.3.6.28-1, 19.3.6.28-2, 19.3.6.28-1, 19.3.6.28-2, 19.3.6.28-4, 19.3.6.33-1/ 19.3.6.28-4, 19.3.6.33-1/ 19.3.6.33-2, 19.3.6.35-2, 19.3.6.33-2, 19,3.6.35-2, 19.3.6.48-2, 19.3.6.51-2, 19.3.6.48-2, 19.3.6.51-2, 19.3.6.63-2, 19.3.7.28-5, 19.3.6.63-2, 19.3.7.28-5, 19.3.7.34-1, 19.3.7.42-1, 19.3.7.34-1, 19.3.7.42-1, 19.3.8.4-1, 19.3.8.5-2, 19.3.8.4-1, 19,3.8.5-2, 19.3.8.5-3, 19.3.8.5-4, 19.3.8.5-2a, 19.3.8.5-3, 19.3.8.10-2, 19.3.8.10-3, 19.3.8.5-4, 19.3.8.10-2, 19.3.8.11-1/19,3.8.11-2, 19.3.8.10-3, 19.3.8.11-1/ 19.3.8.14-2, 19.3.8.11-2, 19.3.8.14-2, 19.3.8.18-1/19.3.8.18-2, 19.3.8.18-1/19.3.8.18-2, 19.3.8.25-1, 19.3.8.25-1, 19.3.8.26-1/ 19.3.8.26-1/19,3.8.26-2, 19.3.8.26-2, 19.3.8.27-1, J 19.3.8.27-1, 19.3.8.32-7, 19.3.8.32-7, 19.3.8.32-9, 19.3.8.32-9, 19.3.8.32-25, 19.3.8.32-25, 19.3.8.32-28, 19,3.8.32-28, 19.3.9.19-1/19.3.9.19-2, 19.3.9.19-1/19.3.9.19-2, 19.3.9.43-1/19,3.9.43-2, 19 3.9.43-1/19.3.9.43-2, 19.3.9.56-1 through 19.3.9.56-4,
/3 19.3.9.56-1, 19.3.9.56-2, 19.3.9.58-3, 19.3.9.60-1/19.3.9.60-2, V 19.3.9.56-3/19.3.9.56-4, 19.3.9.58-3, 19.3.9.60-3/19.3.9.60-4, 19.3.9.67-3, 19.3.9.77-2, 19.3.9.60-1/19.3.9.60-2, 19.3.9.88-1/19.3.9.88-2, 19.3.9.60-3/19.3.9.60-4, 19.3.9.92-1/19.3.9.92-2, 19.3.9.67-3, 19,3.9.77-2, 19.3.9.94-2, 19,3.9.102-1/
19.3.9.88-1/19.3.9.88-2, 19.3.9.102-2, 19.3.9.103-1/ 19.3.9.92-1/19.3.9.92-2, 19,3.9.103-2, 19.3.9.110-1/ 19.3.9.94-2, 19.3.9.110-2, 19.3.10.2-1/ 19.3.9.102-1/19.3.9.102-2, 19.3.10.2-2, 19,3.11.9-2, 19.3.9.103-1/19.3 9.103-2, 19.3.15.10-1 through 19.3.9.110-1/19.3.9.110-2, 19.3.15.10-3/19.3.15.10-4, 19.3.11.9-2, and 19.3.15.10-1 19.3.16-1/19,3.16-2, through 19.3.15.10-3/19.3.15.10-4 19.3.18-1/19.3.18-2, and 19,3.19-1/19.3.19-2. (
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GESSAR NI GE Figure Drawing Dwg. Parts List / No. No. Rev. Title Type Item No. 7A.3-16k I-771L 3 DG Rooms and Switchgear Room Heating ELEM 386X988-028 and Vent DIAG 1 M 3 m N 3 n P 3 o Q 3 m P R 3 ." q S 3 @@ em v 4 r T 3 gg 55 $ s U 3 t V 3 $U e u AA 3 E o v BB 3 w CC 2 x DD 3 y F EE 3 o 7A.3-17a I-5 31A 3 Essential Water Service System 386X988-013 b B 3 c C 3 mM d D 3 @y o U e U E 3 e p wo U1 *J W
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GESSAR.II 22A7007 238 NUCLEAR ISLAND Rav. 15 SECTION 1.8 CONTENTS Section Title Page 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1.8.0-1 1.8.0 Summary 1.8.0-1 1.8.1 Regulatory Guide 1.1, Revision 0, Dated November, 1970 1.8.1-1 1.8.2 Regulatory Guide 1.2, Revision 0, Dated November, 1970 1.8.2-1 1.8.3 Regulatory Guide 1.3, Revision 2, Dated June, 1974 1.8.3-1 1.8.4 Regulatory Guide 1.4, Revision 2, Dated June, 1974 1.8.4-1 1.8.5 Regulatory Guide 1.5, Revision 0, Dated March, 1971 1.8.5-1 [~' l.8.6 Regulatory Guide 1.6, Revision 0, Dated March, 1971 1.8.6-1 1.8.7 Regulatory Guide 1.7, Revision 2, Issued in 1978 1.8.7-1 1.8.8 Regulatory Guide 1.8, Revision 1-R, Dated May, 1977 1.8.8-1 , 1.8.9 Regulatory Guide 1.9, Revision 2, Dated December, 1979 1.8.9-1 1.8.10 Regulatory Guide 1.10, Revision 1 (Withdrawn July, 1981) 1.8.10-1 l ! 1.8.11 Regulatory Guide 1.11, Revision 0, Dated March 1971 and Supplement, dated February, 1972 1.8.11-1 ! 1.8.12 Regulatory Guide 1.12, Revision 2, Dated July 1981 1.8.12-1 1.8.13 Regulatory Guide 1.13, Revision 1, Dated December, 1975 1.8.13-1 (N (,) 1.8.14 Regulatory Guide 1.14, Revision 1, Dated August, 1975 1.8.14-1 l 1.8-i I
GESSAR II 22A7007 238 HUCLEAR ISLAND Rev. 15 SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.15 Regulatory Guide 1.15, Revision 1, (Withdrawn July, 1981) 1.8.15-1 1.8.16 Regulatory Guide 1.16, Revision 4, Dated August, 1975 1.8.16-1 1.8.17 Regulatory Guide 1.17, Revision 1, Dated June, 1973 1.3.17-1 1.8.18 Regulatory Guide 1.18, Revision 1, (Withdrawn July, 1981) 1.8.18-1 1.8.19 Regulatory Guide 1.19, Revision 1, (Withdrawn July, 1981) 1.8.19-1 1.8.20 Regulatory Guide 1.20, Revision 2, Dated May, 1976 1.8.20-1 1.8.21 Regulatory Guide 1.21, Revision 1, Dated June, 1974 1.8.21-1 1.8.22 Regulatory Guide 1.22, Revision 0, Dated February, 1972 1.8.22-1 1.8.23 Regulatory Guide 1.23, Revision 0, Dated February, 1972 1.8.23-1 1.8.24 Regulatory Guide 1.24, Revision 0, Dated March, 1972 1.8.24-1 1.8.25 Regulatory Guide 1.25, Revision 0, Dated March, 1972 1.8.25-1 1.8.26 Regulatory Guide 1.26, Revision 3, Dated February, 1976 1.8.26-1 1.8.27 Regulatory Guide 1.27, Revision 2, Dated January, 1976 1.8.27-1 1.8.28 Regulatory Guide 1.28, Revision 2, Dated February, 1979 1.8.28-1 1.8.29 Regulatory Guide 1.29, Revision 3, Dated September, 1978 1.8.29-1 1.8-ii
GESSAR -II 22A7007 238 NUCLEAR ISLAND Rsv. 15 ('h)
~
SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.30 Regulatory Guide 1.30, Revision 0, Dated August, 1972 1.8.30-1 1.8.31 Regulatory Guide 1.31, Revision 3, Dated April, 1978 1.8.31-1 1.8.32 Regulatory Guide 1.32, Revision 2, Dated February, 1977 1.8.32-1 1.8.33 Regulatory Guide 1.33, Revision 2, Dated February, 1978 1.8.33-1 1.8.34 Regulatory Guide 1.34, Revision 0, Dated December, 1972 1.8.34-1 1.8.35 Regulatory Guide 1.35, Revision 2, Dated January, 1976 1.8.35-1 1.8.36 Regulatory Guide 1.36, Revision 0, Dated February, 1973 1.8.36-1 ( 1.8.37 Regulatory Guide 1.37, Revision 0, Dated March, 1973 1.8.37-1 1.8.38 Regulatory Guide 1.38, Revision 1, Dated May, 1977 1.9.38-1 1.8.39 Regulatory Guide 1.39, Revision 2, Dated September, 1977 1.8.39-1 1.8.40 Regulatory Guide 1.40, Revision 0, Dated March, 1973 1.8.40-1 1.8.41 Regulatory Guide 1.41, Revision 0, Dated March, 1973 1.8.41-1 1.8.42 Regulatory Guide 1.42, Revision 0, (Withdrawn March, 1976) 1.8.42-1 1.8.43 Regulatory Guide 1.43, Revision 0, Dated May, 1973 1.8.43-1 1.8.44 Regulatory Guide 1.44, Revision 0, Dated May, 1973 1.8.44-1 0) 'sms 1.8.45 Regulatory Guide 1.45, Revision 0, 1.8.45-1 Dated May, 1973 1.8-iii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.46 Regulatory Guide 1.46, Revision 0, Dated May, 1973 1.8.46-1 1.8.47 Regulatory Guide 1.47, Revision 0, Dated May, 1973 1.8.47-1 1.8.48 Regulatory Guide 1.48, Revision 0, 7ated May, 1973 1.8.48-1 1.8.49 Regulatory Guide 1.49, Revision 1, Dated December, 1973 1.8.49-1 1.8.50 Regulatory Guide 1.50, Revision 0, Dated May, 1973 1.8.50-1 1.8.51 Regulatory Guide 1.51, Revision 0, (Withdrawn July, 1975) 1.8.51-1 1.8.52 Regulatory Guide 1.52, Revision 2, Dated April, 1978 1.8.52-1 1.8.53 Regulatory Guide 1.53, Revision 0, Dated June, 1973 1.8.53-1 1.8.54 Regulatory Guide 1.54, Revision 0, Dated June, 1973 1.8.54-1 1.8.55 Regulatory Guide 1.55, Revision 0, (Withdrawn June, 1981) 1.8.55-1 1.8.56 Regulatory Guide 1.56, Revision 1, Dated July, 1978 1.8.56-1 1.8.57 Regulatory Guide 1.57, Revision 0, Dated June, 1973 1.8.57-1 1.8.58 Regulatory Guide 1.58, Revision 1, Dated September, 1980 1.8.58-1 1.8.59 Regulatory Guide 1.59, Revision 2, Dated August, 1977 1.8.59-1 1.8.60 Regulatory Guide 1.60, Revision 1, Dated October, 1973 1.8.60-1 1.8.61 Regulatory Guide 1.61, Revision 0, Dated October, 1973 1.8.61-1 1.8-iv
GESSAR II ~22A7007 238 NUCLEAR ISLAND Rav. 15 SECTION 1.8 CONTENTS (Continued) Section Title Paae 1.8.62 Regulatory Guide 1.62, Revision 0, Dated October, 1973 1.8.62-1 1.8.63 Regulatory Guide 1.63, Revision 1, Dated May, 1977 1.8.63-1 1.8.63.1 Analysis of Circuits Penetrating Primary Containment 1.8.63-3 1.8.64 Regulatory Guide 1.64, Revision 2, Dated June, 1976 1.8.64-1 1.8.65 Regulatory Guide 1.65, Revision 0, Dated October, 1973 1.8.65-1 1.8.66 Regulatory Guide 1.66, Revision 0, (Withdrawn October, 1977) 1.8.66-1 1.8.67 Regulatory Guide 1.67, Revision O. Dated October, 1973 1.8.67-1 ( 1.8.68 Regulatory Guide 1.68, Revision 2, Dated August, _1978 1.8.68-1 1.8.68.1 Regulatory Guide 1.68.1, Revision 0, Dated December, 1975 1.8.68-7 1.8.68.2 Regulatory Guide 1.68.2, Revision 1, Dated July, 1978 1.8.68-8 1.8.68.3 Regulatory Guide 1.68.3, Revision 0, Dated April, 1982 1.8.68-10 1.8.69 Regulatory Guide 1.69, Revision 0, Dated December, 1973 1.8.69-1 1.8.70 Regulatory Guide 1.70, Revision 3, Dated November, 1976 1.8.70-1 1.8.71 Regulatory Guide 1.71, Revision 0, Dated December, 1973 1.8.71-1 1.8.72 Regulatory Guide 1.72, Revision 2, Dated November, 1978 1.8.72-1 1.8.73 Regulatory Guide 1.73, Revision 0, Dated January, 1974 1.8.73-1 1.8-v
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.74 Regulatory Guide 1.74, Revision 0, Dated February, 1974 1.8.74-1 1.8.75 Regulatory Guide 1.75, Re rision 2, Dated September, 1978 1.8.75-1 1.8.76 Regulatory Guide 1.76, Revision 0, Dated April, 1974 1.8.76-1 1.8.77 Regulatory Guide 1.77, Revision Dated May, 1974 1.8.77-1 1.8.78 Regulatory Guide 1.78, Rccrieion 0, Dated June, 1974 1.8.78-1 1.8.79 Regulatory Guide 1.79, Revision 1, Dated September, 1975 1.8.79-1 1.8.80 Regulatory Guide 1.80, Revision 0, Dated June, 1974 1.8.80-1 h 1.8.81 Regulatory Guide 1.81, Revision 1, Dated January, 1975 1.8.81-1 1.8.82 Regulatory Guide 1.82, Revision 0, Dated June, 1974 1.8.82-1 1.8.83 Regulatory Guide 1.83, Revision 1, Dated July, 1975 1.8.83-1 1.8.84 Regulatory Guide 1.84, ReviLion 18, Dated August, 1981 1.8.84-1 1.8.85 Regulatory Guide 1.85, Revision 18, Dated August, 1981 1.8.85-1 1.8.86 Regulatory Guide 1.86, Revision 0, Dated June, 1974 1.8.86-1 1.8.87 Regulatory Guide 1.87, Revision 1, Dated June, 1975 1.8.87-1 1.8.88 Regulatory Guide 1.88, Revision 2, Dated October, 1976 1.8.88-1 1.8.89 Regulatory Guide 1.89, Revision 0, Dated November, 1974 1.8.89-1 1.8-vi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.90 Regulatory Guide 1.90, Revision 1, Dated August, 1977 1.8.90-1
, 1.8.91 Regulatory Guide 1.91, Revision 1, Dated February, 1978 1.8.91-1 1.8.92 Regulatory Guide 1.92, Revision 1, Dated February, 1976 1.8.92-1 1.8.93 Regulatory Guide 1.93, Revision 0, Dated December, 1974 1.8.93-1 1.8.94 Regulatory Guide 1.94, Revision 1, Dated April, 1976 1.8.94-1 i 1.8.95 Regulatory Guide 1.95, Revision 1, Dated January, 1977 1.8.95-1 1.8.96 Regulatory Guide 1.96, Revision 1,
() 1.8.97 Dated June, 1976 Regulatory Guide 1.97, Revision 1, 1.8.96-1 Dated August, 1977 1.8.97-1 1.8.98 Regulatory Guide 1.98, Revision 0, Dated March, 1976 1.8.98-1 l l 1.8.99 Regulatory Guide 1.99, Revision 1, ! Dated April, 1977 1.8.99-1 1.8.100 Regulatory Guide 1.100, Revision 1, Dated August, 1977 1.8.100-1 l l 1.8.101 Regulatory Guide 1.101, Revision 1, Dated March, 1977 1.8.101-1 1.8.102 Regulatory Guide 1.102, Revision 1, Dated September, 1976 1.8.102-1 ! 1.8.103 Regulatory Guide 1.103, Revision 1, (Withdrawn July, 1981) 1.8.103-1 1.8.104 Regulatory Guide 1.104, Revision 0, (Withdrawn August, 1979) 1.8.104-1 ( l.8.105 Regulatory. Guide 1.105, Revision 1, Dated November, 1976 1.8.105-1 l 1.8-vii
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 15 SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.106 Regulatory Guide 1.106, Revision 1, Dated March, 1977 1.8.106-1 1.8.107 Regulatory Guide 1.107, Revision 1, Dated February, 1977 1.8.107-1 1.8.108 Regulatory Guide 1.108, Revision 1, Dated August, 1977 1.8.108-1 1.8.109 Regulatory Guide 1.109, Revision 1, Dated October, 1977 1.8.109-1 1.8.110 Regulatory Guide 1.110, Revision 0, Dated March, 1976 1.8.110-1 1.8.111 Regulatory Guide 1.111, Revision 1, Dated July, 1977 1.8.111-1 1.8.112 Regulatory Guide 1.112, Revision 0-R, Dated September, 1976 1.8.112-1 1.8.113 Regulatory Guide 1.113, Revision 1, Dated April, 1977 1.8.113-1 1.8.114 Regulatory Guide 1.114, Revision 1, Dated November, 1976 1.8.114-1 1.8.115 Regulatory Guide 1.115, Revision 1, Dated July, 1977 1.8.115-1 1.8.116 Regulatory Guide 1.116, Revision 0-R, Dated May, 1977 1.8.116-1 1.8.117 Regulatory Guide 1.117, Revision 1, Dated April, 1978 1.8.117-1 1.8.118 Regulatory Guide 1.118, Revision 1, Dated November, 1977 1.8.118-1 1.8.119 Regulatory Guide 1.119, Revision 0, (Withdrawn June, 1977) 1.8.119-1 1.8.120 Regulatory Guide 1.120, Revision 1, Dated November, 1977 1.8.120-1 1.8.121 Regulatory Guide 1.121, Revision 0, O Dated August, 1976 1.8.121-1 1.8-viii
G;3SAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 (D ( / SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.122 Regulatory Guide 1.122, Revision 1, Dated February, 1978 1.8.122-1 1.8.123 Regulatory Guide 1.123, Revision 1, Dated July, 1977 1.8.123-1 1.8.124 Regulatory Guide 1.124, Revision 1, Dated January, 1978 1.8.124-1 1.8.125 Regulatory Guide 1.125, Revision 1, Dated October, 1978 1.8.125-1 1.8.126 Regulatory Guide 1.126, Revision 1, Dated March, 1978 1.8.126-1 1.8.127 Regulatory Guide 1.127, Revision 1, Dated March, 1978 1.8.127-1 1.8.128 Regulatory Guide 1.128, Revision 1, () 1.8.129 Dated October, 1978 Regulatory Guide l.129, Revision 0, 1.8.128-1 Dated April, 1977 1.8.129-1 1.8.130 Regulatory Guide 1.130, Revision 1, Dated October, 1978 1.8.130-1 1.8.131 Regulatory Guide 1.131, Revision 0, Dated August, 1977 1.8.131-1 1.8.132 Regulatory Guide 1.132, Revision 1, Dated October, 1978 1.8.132-1 1.8.133 Regulatory Guide 1.133, Revision 1, ' Dated May, 1981 1.8.133-1 1.8.134 Regulatory Guide 1.134, Revision 1, Dated November, 1978 1.8.134-1 1.8.135 Regulatory Guide 1.135, Revision 0, Dated September, 1977 1.8.135-1 j 1.8.136 Regulatory Guide 1.136, Revision 2, Dated June, 1981 1.8.136-1 1.8.137 Regulatory Guide 1.137, Revision 0, Dated January, 1978 1.8.137-1 1.8-ix
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 1.8 CONTENTS (Continued) Section Title Page 1.8.138 Regulatory Guide 1.138, Revision 0, Dated April, 1978 1.8.138-1 1.8.139 Regulatory Guide 1.139, Revision 0, Dated May, 1978 1.8.139-1 1.8.140 Regulatory Guide 1.140, Revision 1, Dated October, 1979 1.8.140-1 1.8.141 Regulatory Guide 1.141, Revision 0, Dated April, 1978 1.8.141-1 1.8.142 Regulatory Guide 1.142, Revision 1, Dated October, 1981 1.8.142-1 1.8.143 Regulatory Guide 1.143, Revision 1, Dated October, 1979 1.8.143-1 1.8.144 Regulatory Guide 1.144, Revision 1, Dated September, 1980 1.8.144-1 1.8.145 Regulatory Guide 1.145, Revision 0, Dated August, 1979 1.8.145-1 1.8.146 Regulatory Guide 1.146, Revision 0, Dated August, 1980 1.8.146-1 1.8.147 Regulatory Guide 1.147, Revision 1, Dated March, 1982 1.8.147-1 1.8.148 Regulatory Guide 1.148, Revision 0, Dated April, 1981 1.8.148-1 1.8.149 Regulatory Guide 1.149, Revision 0, Dated May, 1981 1.8.149-1 1.8.150 Regulatory Guide 1.150, Revision 0, Dated July, 1981 1.8.150-1 1.8.151 Regulatory Guide 8.8, Revision 3, Dated June, 1978 1.8.151-1 1.8.152 Regulatory Guide 8.19, Revision 1, Dated June, 1974 1.8.152-1 O 1.8-x
8 GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 () 1.8.6 Regulatory Guide 1.6, Revision 0, Dated March 1971
Title:
Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems General Design Criterion 17 requires that onsite electrical power systems have sufficient independence to perform their safety functions assuming c single failure. This safety guide describes an acceptable degree of independence between redundant standby (onsite) power sources and between their distribution systems. This guide does not address the suitability of nearby hydro-electric, nuclear, or fossil units as standby power sources at multiple-unit sites. This matter will be evaluated on an indi-vidual case basis. Evaluation The GESSAR II design complies with this safety guide. I O emm i 1.8.6-1/1.8.6-2
A b GESSAR II '22A7007 238 NUCLEAR ISLAND Rov. 15-Regulatory Guide 1.44, Revision 0, Dated May 1973 0 1.8.44
Title:
Control of the Use of Sensitized Stainless Steel 1 This guide describes acceptable methods of implementing the requirements of GDC 1 and 4 of Appendices A and B to 10CFR50, with regard to control of the application and processing of stain-less steel.to avoid'sevire sensitization that could lead to stress corrosion cracking. This guide applies to light-water-cooled reactors. , Evaluation The GESSAR II design conplies with this regulatory guide and with theguidelinesofNUREGOgl3, revision 1aswell. o . 4 MD l.8.44-1/1.8.44-2 c ,
l / GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 -
>n 1.8.52 Regulatory Guide 1.52, Revision 2, Dated April 1978
Title:
Design, Testing, and Maintenance Criteria for Engineered Snfety Feature Atmosphere Cleanup System Air Filtration + and Adsorption Units of Light-Water-Cooled Nuclear Power , Plants This guide' presents methods acceptable to the NRC staff for implementing the Commission's regulations in Appendix A to 10CFR50 with regard to design, testing, and maintenance criteria
' 'i for air filtration and adsorption units,of engineered-safety-feature (ESP) atmosphere cleanup systems in light-water-cooled nucleae power plants. This guide applies only to post-accident engineered-safety-feature atmosphere cleanup systems designed to f
mitigate the consequences of postulated accidents. It addresses
- y' the ESF atmosphere cle'anup system, including the various com-I ponents and ductwork, in the postulated DBA environment.
4 This guide does not apply to atmosphere cleanup systems designed ()
'- ' to collect airborne radioactive materials during normal plant f
doperation, j . including anticipated operational occurrences. c, ' Evaluation This guide is applicable to the Standby Gas Treatment System and i Control Building Outdoor Air Cleanup System (CBOACS) . The SGTS and CBOACS' comply with this regulatory guide and the guidelines l I of Section 6.5.1't-o NUREG-0800.
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s GESSAR II 22A7007
.g., 238 NUCLEAR ISLAND Rev. 15 j
v (O 1.8.63 ' Regulatory! Guide 1.63,' Revision 1, Dated May 1977 (Continued) i; g ., e (3) GE interprets " designed" in Section 1 of the regulatory position first sentence, to mean " designed and applied." This interpretation is necessary to clarify the use of } circuit overload protection.for penetration circuits. Overload protection may'be' external to the penetration ' {I and thus outside the scope of the penetration ( .J 9 designer, l' .I d , 1.8.63.1 Analysis of Circuits Penetrating Primary Containment cf . ,
'y
( d'/ i '. ) .(1) 6.3 kV circuits for recirculation pump motors are pro-
'8' ' /,w tected by two circuit breakers in series in the 6.9 kV supply circuits. The recirculation pump motors are
; al'so fed from the low frequency motor generator sets.
p) This feed is directly protected by one 6.9 kV rated breaker. A second 6.9 kV rated circuit breaker or fuse is provided to protect the containment penetration. l Ch
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 1.8.63.1 Analysis of Circuits Penetrating Primary Containment (Crgntinued) (2) Power circuits for motor control center (MCC) loads are protected by a circuit breaker and a fuse par phase in series. The application of penetration wire protecting devices is shown on the (MCC) single line diagrams. (3) MCC and other control and indication circuits have pro-tection as follows: MCC control circuits have two fusas per control circuit. (4) Specific circuits, having a limited power source, that cannot produce any short circuit current, damaging to R the conductor insulation (i.e., self protecting), do not m' require a protective device. Included in these special circuits are: (a) Thermocoup.'.e circuits (b) Shielded cables for low level signals (4 to 20 mA-LPRM, IRM, SRM, RPIS instrumentation circuits) (c) Annunciator circuits O 1.8.63-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 1.8.82 Regulatory Guide 1.82, Revision 0, Dated June 1974
Title:
Sump for Emergency Core Cooling and Containment Spray Systems
~
This guide describes a method acceptable to the Regulatory staff for implementing the requirements with regard to design, fabrica-tion and testing of sump suction inlet conditions for pumps in the Emergency Core Cooling and Containment Spray Systems. _ Evaluation This regulatory guide pertains to PWRs and is not applicable to the GESSAR II design. O 4 I O 1.8.82-1/1.8.82-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 14 i- 1.8.92. Regulatory Guide 1.92, Revision 1, Dated February 1976 (Continued) i described below in Subsection 1.8.1.02a. Closely spaced medes are combined by the double sum method with absolute sign as described in Subsection 1.8.1.92b. In the time-history method of dyramic analysis, the vector sum at every step is used to calculate the combined response. The use of the time-history method preclude" the need to consider modal spacing. (a) Square Root of the Sum of the Squares The SRSS combination of modal responses is defined mathematically as: 1/2 N R= }] (Rg )2 i=1 i where: R = Corabined response i Rg = Response in the ith
- d*
n = Number of modes considered in the I analysis (b) Procedure of Combining Closely Spaced Modal
Response
The double sum method is used to combine the responses of closely spaced modes when the response O ( l.8.92-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 1.8.92 Regulatory Guide 1.92, Revision 1, Dated February 1976 (Continued) spectrum method of modal dynamic analysis is used. This method is defined mathematically as: _0,5 N N
- E b [ ks ks k=1 s=1 where R is the representative maximum value of a particular response of a given element to a given component of excitation, Rk is the peak value of the response of the element due to the kth mode, and N is the number of significant modes considered in the modal response combination. In addition, R is the peak value of the response of the element h attributed to sth mode. Also,
_ _-1 )
- -2 m;_m.
E ks " l+ S{wk+0 s "s. in which I _ _ 0.5 w'k =w k 1-8 and B' = 8 2 ] l k k k + td"k where wk nd S k are the modal frequency and the i damping ratio in the kth mode, respectively, and td is the duration of the earthquake. l l l 1.8.92-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 1.9 STANDARD DESIGMS v 1.9.1 Interfaces 1.9.1.1 General GESSAR II presents the standard Nuclear Island design incorporat-ing the GW BWR/6 product line with a Mark III containment. The Nuclear Island consists of the Reactor Building, Fuel Building, Auxiliary Building, Control Building, Diesel Generator Buildings, and Radwaste Building including their contained structures, sys-tems, and components as shown in Figures 1.2-2 through l.2-26. The Applicant is responsible for the remaining structures, systems, and components. The Applicant's portion is referred to as the Balance of Plant (BOP). This subsection identifies the information that will be provided
/~ by the Applicant in his FSAR to complete the Regulatory Guide 1.70 \m]- content requirements including responses to selected NRC questions.
' In addition, this subsection identifies the design interfaces between the Nuclear Island and BOP. This GESSAR II/FSAR interface information and Nuclear Island / BOP design interface information is provided in the following subsections. 1.9.1.2 GESSAR II/FSAR Interfaces As discussed in Section 1.1, GESSAR II is written in accordance with Regulatory Guide 1.70. However, the content of this regula-tory guide addresses the entire plant and since the GESSAR II design scope is limited to the Nuclear Island, there are BOP por-tions of the content that will be provided by the Applicant. In addition to this BOP information, there is information within the scope of the Nuclear Island that will not be available until GESSAR II is utilized by an Applicant for a specific plant. This information will also be provided by the Applicant and includes [) V information that is: equipment-vendor dependent; Applicant l.9-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 1.9.1.2 GESSAR II/FSAR Interfaces (Continued) dependent; or site dependent. In addition, GE has chosen for commercial reasons to delay the submittal of certain information until the first Applicant references GESSAR II. All of the information noted above is presented in Tables 1.9-1 through 1.9-19. The columns of these tables are defined below: Item No. This is a chapter-wise interface information item by the NRC and Applicant. Subject Statement as to the subject of the interface. Description A description of the specific information required. Page GESSAR II page number where the interface appears. Subsection GESSAR II subsection, table, or figure where the in'terface appears. Related Question A listing of the related question number (when applicable). GESSAR II question number with NRC question number in parentheses. (See Chapter 19.) 9 1.9-la
O O O Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 1.80 Emergency Respond to the NRC requirement to 1A.74-1 1A.74 3 Preparedness comply with regulations regarding the upgrading of emergency preparedness. 1.81 Emergency Respond to the NRC requirement to 1A.75-1 1A.75 3 Support establish an.onsite technical Facilities support center, an onsite operational support center, and an offsite emergency operations facility. y on H 1.82 Emergency Respond to the NRC requirement to 1A.76-1 1A.76 ua Preparedness upgrade emergency plans in the 3 n un i event of a radiological emergency. U) f $5 P 1.83 Containment Provide information outside of the 1A.77-3 1 A. 77 . 1 M $ Integrity Nuclear Island scope regarding sg g procedures to deal with reduction un of highly radioacti fe fluid leakage. 1.84 Airborne Respond to the NRC requirement to 1A.78-1 1A.78 3 O Iodine provide equipment, training, and Instru- procedures for the accurate mentation assessment of airborne iodine in the plant. 1.84.1 Personnel Supply data on Technical Support 1AA-5 1AA.2 1.10 3 access to Center. Post accident sampling (471.19) control room, station and sample analysis area. gw Technical Sup- 2M port Center, 4>q during 100-day o period. [Oq e 1 J
Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QJESTION CATEGORY 1.85 Waterhammer Conduct a preoperational vibration 1B-5 1B.2.2.3 1 and dynamic effects test program for all Class 1, 2, and 3 and other piping systems and piping restraints. w w CD H Z e CO I O t9 A t"* In t1 CA
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(s f% x . Table 1.9-1 CHAPTER 1 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 1.86 Systems Provide for a systematic visual IB-12 1B.2.4.3 3 Interaction inspection by a multi-disciplinary team to review physical and spatially coupled systems , interactions in the plant. I.87 Seismic Provide qualification summaries 1C.2.2-2 lc.2.2 2 Qualification for equipment, results of plant tests and plans for operational g tests to confirm qualification, w and ind:Oate whether equipment CD 7' has met requirements. g c3 e CM i 1.88 Fire Prevention Establish, train, and equip a 1C.2.3-2 lc.2.3 3 og f site fire brigade, and also b> bd establish surveillance gy M N procedures per 10CFR50, App. R Sections II.C.5 and M F4 r4 II.C.7. 1.89 Hydrant Install hydrant isolation valves 1C.2.3-3 1C.2.3 1 Isolation per 10CFR50, App. R Section III.C. Valves 1.90 Hydrostatic Perform hydrostatic hose tests per 1C.2.3-3 1C.2.3 1 Hose Test 10CFR50, App. R Section III.E. i 1.91 Fire Brigade Establish, train, and equip a site 1C.2.3-4 1C.2.3 3 fire brigade per the specific I requirements of 10CFR50, App. R N Sections III.H and III.I. , g
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Table 1.9-3 CHAPTER 3 GESSAR II/FSAR INTERFACES ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 3.1 GDC 17 Supplement, as required, the 3.1-27 3.1.2.2.8.2 1 evaluation against Criterion 17
- Electric Power Systems.
3.2 Radwaste Drain Provide equipment classification 3.2-26 Table 3.2-1 1 Systems information for radwaste drain systems. 3.3 Plant Electrical Provide equipment classification 3.2-28 Table 3.2-1 1 h) Systems information~for plant electrical $ systems. H O g3 3.4 RCPB Code and Provide reactor coolant pressure 3.2-59 Table 3.2-4 3 g$ o s Addenda boundary Class 1 equipment code and t" U) f and addenda application informa- 3.2-60 $ 2 tion
" 3.5 Collapse of Ensure that collapse of non- 3.3-6 3.3.2.3 1 5" non-Seismic seismic Category I cooling towers Category I or stacks outside of Nuclear Components Nuclear Island would not endanger seismic Category I structures.
3.6 Turbine Describe target areas pertaining 3.5-16 3.5.1.3.1 1 Placement and to the safety-related items Orientation identified in Subsection 3.5.2 not included in Figure 3.5-2. 3.7 Turbine Missile Describe postulated turbine 3.5-16 3.5.1.3.2 1 p, g3 Identification missile properties per R.G. 1.70 MM and , , Characteristics subsection 3.5.1.3, Item 2. $>qa wo ab 4
Table 1.9-3 CHAPTER 3 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 3.8 Target Describe structures and equipment 3.5-16 3.5.1.3.3 1 Identification in subsection 3.5.2 (targets) out-side of the Nuclear Island. 3.9 Probability Provide analysis of strike proba- 3.5-16 3.5.1.3.4 1 Analysis bilities for low-trajectory turbine missiles per R.G. 1.70 subsection 3.5.1.3, Item 4. 3.10 Turbine Describe turbine overspeed 3.5-16 3.5.1.3.5 1 g Overspeed protection system per R.G. 1.70 w Protection subsection 3.5.1.3, Item 5. 03 H ZO 3.11 Turbine Valve Describe turbine valve testing 3.5-16 3.5.1.3.6 1 C P3 [3y e Testing environment per R.G. 1.70 sub-i section 3.5.1.3.6, Item 6. p3 > A
' g X2 w 3.12 Turbine Provide data pertinent to the 3.5-16a 3.5.1.3.7 1 s 8
Characteristics evaluation of its failure char- HH acteristics per R.G. 1.70 Subsection 3.5.1.3, item 7. 3.12.1 Breaks in Provide maximum true lengths to 3.6-9 3.6.1.3.2.2 3.14 [MEB 3 Unrestrained assure that maximum of 30-feet (DSER) Item High Energy distance criterion is not No. 39] Piping exceeded.
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A /% /"'N h Table 1.9-3 CHAPTER 3 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 3.13 Material for Provide a summary of the dynauic 3.6-27 3.6.2.5 2 Operating analyses applicable to high and License Review moderate-energy piping systems and associated supports that j determine the loadings resulting from postulated pipe breaks and $$ 00 cracks per R.G. 1.70 subsection 3.6.2.5. E 3.7.3.12 1 gQM w 3.14 Buried Seismic Describe seismic criteria and 3.7-58 j, Category I methods for buried seismic $$
.. Piping Category I piping systems and NN W tunnels per R.G. 1.70 subsection-HH
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O O O Table 1.9-3 CHAPTER 3 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 3.15 Seismic Category Describe other seismic Category I 3.8-98 3.8.4.1.7 1 I Structure structures not within the Nuclear Descriptions Island. 4 3.16 Codes, Standards Provide applicable codes, standards 3.8-102 3.8.4.2.7 1 and Specifica- and specifications for other t tions seismic Category I structures not within Nuclear Island. N 3.17 Loads and Load Provide loads and load combina- 3.9-114 3.8.4.3.7 1 L> Combinations tions appropriate to Seismic CD Category I structures not within g c) f Nuclear Island Cy g
& un j 3.18 Design and Provide design and analysis 3.8-126 3.8.4.4.7 1 Mg . Analysis procedures for seismic Category I %H b) Procedures structures now within Nuclear b Island. HH 3.18.1 NUREG-0800 Update computer programs and 3.9-la 3.9.1.2 3.151[MEB (DSER)
Compliance on indicate method of verification Item No. 39 Computer and the version used. Programs Al M mN
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Table 1.9-3 CHAPTER 3 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTdH' CE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATE J . i.e Y 3.19 Inservice Provide details of the pump and 3.9-125 3.9.6 3 Testing of valve inservice testing program, and Pumps and including test schedules and 3.9-126 Valves frequencies. Also, applications for written relief from Section XI Addendum requirements, pursuant N) to 10CFR50, Section 55 a(g) (6) (i). y ~ Z 3.20 Materials Provide structural acceptance 3.8-128 3.8.4.5.3 1 fj g f Criteria criteria for Seismic Category I bM A structures not within Nuclear Island. {l @
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h ( O (d kj Table 1.9-3 CHAPTER 3 GESSAR II/FSAR INTEREACES (Continued) 1 ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 3.21 Materials, QA Describe materialc, quality 3.8-130 3.8.4.6.7 1 and Special control, and special construction Construction techiques for Seismic Category I Techniques structures not within Nuclear Island. 3.22 Testing and Describe testing and inservice 3.8-131 3.8.4.7.4 1 Inservice inspection requirements for Seismic Inspection Category I structures not within 83 Nuclear Island, b) o> g 3.23 Verification of Provide verification, through 3.8-142 3.8.6.2 3 go
. Foundation Soil Applicant's soils consulting GM j) engineer, that the Nuclear Island h$
4, foundation soil meets the soil mg g requirements of Appendix 3A, l$ i subsection 3.8.6 and 3.7. s
.4 3.24 Containment Provide a summary of the contain- 3.8-146 Table 3.8-3 1 Vessel Analysis ment vessel analysis.
- Summary 3.25 Major Safety- Provide equipment loading condi- 3.9-177 Table 3.9-11 2,4 Related tions, stress criteria, limiting (2)
Mechanical stress types, allowable and calcu-Components lated stresses on a component-by- 3.9-179 Table 3.9-11 component basis. through (3) 3.9-185 3.9-187 Table 3.9-11 lllj i through (4) <>. 3.9-190
- g3 eo 3.9-191 Table 3.9-11 46 ~J through (5) 3.9-192
O O O Table 1.9-5 CHAPTER 5 GESSAR II/FSAR INTERFACES ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 5.1 Vessel Provide expected effects of 5.3-11 5.3.1.6.3 2 Beltline irradiation on the vessel beltline l Material and weld materials. 5.2 Charpy Test Provide the vessel material 5.3-33 Figure 5.3-1 2 Results and identification, chemical analysis Chemical and fracture toughness properties. Composition w 5.3 Addition of Update to reflect the addition of 5.4-1 5.4 5 03 valves for dual valve barriers for test, vent f' test, vent and drain connection which are hk 0 U2 g3 and drain outside containment and can commun-g connections icate with the containment atmos- E>"
. phere.or suppression pool gW H
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\_ > Tchlc .9-6 CHAPTER 6 GESSAR II/FSAR INTERFACES ITEM RELATED IN'lERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 6.1 Combustible Gas Provide design and performance 6.2-203 Table 6.2-28 2 Control System data for hydrogen mixing compressors and hydrogen recombiners.
6.2 Improved Decay (Optional) Provide the NRC with 6.3-42 6.3.3.7.8 3 Heat Condition a request for exemption from Section I . A.4 of 10CF'.50, App. K referencing GESSAR II Appendix 6A as the technical justification of the exemption. to 6.3 Habitability Verify that food, water, medical ta 6.4-4 6.4.1.1 3 03 Systems Design supplies and sanitary facilities
- p. Baues for sustaining and emergency team of za five persons for a period of d os O U3 f 5 days are available. Verify that Ep"'
a protection such as may be ,, required for eyes and skin is gW r4 available. H e4 r. u) 6.4 foxic Gas Describe potential toxic gas Release Points points in the environs. 6.4-9 6.4.2.2.2 3 h g 6.5 SGTS Provide brake horsepower for 6.5-51 Table 6.5-2 Component exhaust fan and residual heat Descriptions removal fans. 6.6 Instrumentation Provide temperature and pressure 6.5-61 Figure 6.5-1 3 for ESF instrumentation for indication Filters and alarms per NUREG-0800 30 to 6.6.1 In-Service Inspection Provide the details of the in-service plan of class 2 6.6-1 6.6.1 3 yy
. %J and 3 components in the technical specifications. g, @
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c - Table 1.9-6 CIIAPTER 6 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY / 6.7 Examination Indicate the extent to which 6.6-2 6.6.3 3 Techniques the examination techniques and and procedures described in Section XI Procedures of the Code will be used per R.G. 1.70 Subsection 6.6.3. 6.8 Inspection Indicate that an inspection 6.6-2 6.r .4 3 Intervals schedule for class 2 systems components will be developed N per R.G. 1.70 Subsection 6.6.4. s 6.9 Examination Indicate that inservice 6.6-2 6.6.5 3 y@ Categories inspection categories and o us f and Require- requirements for class 2 and yM A ments class 3 are in agreement with the code per R.G. 1.70 > h$ as i W Subsection 6.6.5. s[ N us 6.10 Evaluation of Provide an evaluation of 6.6-2 6.6.6 3 Examination examination results per Results R.G. 1.70 Subsection 6.6.6. 6.11 System Describe the system pressure 6.6-2 6.6.7 3 Pressure testing per R.G. 1.70 Tests Subsection 6.6.7. 6.12 Augmented ISI Provide an augmented inservice 6.6-3 6.6.8 3 to Protect inspection program to protect Against against postulated piping yy Postulated failures per R.G. 1.70 MN Piping Failures Subsection 6.6.8. f {} o HO bM { l l 9 9 9
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{~'s x Table 1.9-8 CHAPTER 8 GESSAR II/FSAR INTERFACES ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 8.1 Utility Grid Provide .a brief description of 8.1-1 8.1.1 1,3 Description the utility grid and its
-interconnection to other grids per R.G. 1.70 Section 8.1.
8.2 Utility Provide the voltage rating for 8.1-1 .8.1.2.1 3 Distribution Utility transmission grid. and Grid 8.1-2 8.3 BOP Power Provide a description of the 8.1-7 8.1.2.2 1 w balance of plant power system per W Systems O' R.G. 1.70 Section 8;1. f 8.4 Offsite Power Provide the safety design basis 8.1-7 8.1.3.1 1
- Design Basis for offsite power per R.G. 1.70 r un f, Section 8.1. ll 8.2-1 8.2.1 1,2 H 7 8.5 Offsite Power Provide a system descriotion of ""
H Systems offsite power and an analysis to Description show compliance with 10CFR50, per R.G. 1.70 Subsection 8.2.1. i 8.6 Offsite Power Present the results of 8.2-1 8.2.2 1 Systems steady-state and transient Analysis stability ana. lyses to show compliance with GDC 17; discuss grid availability per R.G. 1.70 Subsection 8.2.2. 8.7 6.9-KV Demonstrate physical separation 8.3-2 8.3.1.1.1 1 Distribution of the 6.9 KV system feeders to the $ N$ System Nuclear Island. <>
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Table 1.9-8 CHAPTER 8 GESSAR II/FSAR INTERFACES (Continued) TTEM RELATED INTERFACE NO. SUBJECT DESCRIPTION k ..GE SUBSECTION QUESTION CATEGORY 8.7.1 Diesel Gener- Provide diesel generator start- 8.3-26 8.3.1.1.8.5 8.8 2 ator Qualifi- ing test results (430.07) cation Testing 8.8 Reliability Describe the reliability qualifi- 8.3-27 8.3.1.1.8.6 1 Qualification cation testing for the standby AC Testing power supply per R.G. 1.70 Subsection 8.3.1.1. 6) 8.9 W
!iPCS Class 1E Provide physical separation 8.3-31 8.3.1.1.9 4 1 o3 Elcatric details for HPCS Cless lE electric g
Equipment cquipment per R.G. 1.70 Subsection 8.3.1.1. hh O U2 up ta U)
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A 8.10 AC Power Provide analyses of AC power 8.3-37 8.3.1.2.1 1 > gs System systems not within the Nuclear to X3 7 Analyses Island to demonstrate compliance 8.3-51 s[ h) with the GDC's and to show the U2 extent to which regulatory guidus and other criteria are followed per R.G. 1.70 subsection 8.3.1.2. Also, confirm and supplement the analysis of the Nuclear Island portion of the general AC power system. 8.11 NSPS Power Assess NSPS power supply against 8.3-54 8.3.1.2.2.2.6 1 System R.G. 1.41. Compliance y to o N) 8.12 NSPS Power System Assess NSPS power supply against IEEE 338 2975. 8.3-56 8.3.1.2.2.3.3 1 fDo Compliance [$ l i O O O
O O Table 1.9-8 i CHAPTER 8 ) GESSAR II/FSAR INTERFACFS (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY i i' 8.13 Power Systems Describe the method and the 8.3-74 8.3.1.3.1 1 l . materials used for cable and raceway color marking. l l N w CD } h oo
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C'\ f (~ U (/ ( Table 1.9-9 CHAPTER 9 CESSAR II/FSAR INTERFACES ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 9.0 Testing Supply frequency of inspection 9.1-17 9.1.2.4 9.18 3 Inspection and type of sampling used in (281.10) monitoring the Post Accident Sampling system. 9.1 Spent Fuel Describe the spent fuel casks 9.1-28 9.1.4.2.1 1 Cask to be used in the fuel-handling system per R.G. 1.70 Subsection 9.1.4.2. N Overhead Bridge Describe the cask crane per 9.1-28 9.1.4.2.2.2 1 W 9.2 " Cranes R.G. 1.70 to Subsection 9.1.4.2. Fuel Handling Describe any plant specific 9.1-63a 9.1.4.3 3 gO g$ f 9.3 deviations from the fuel-handling e System j, systems safety evaluation W presented in this subsection per R.G. 1.70 Subsection 9.1.4.3. C lu [ 9.4 ESW Safety Provide power generation design 9.2-2 9.2.1.1.2 1 Design Bases bases not.within Nuclear Island per R.G. 1.7L Subsection 9.2.1. 9.5 ESW System Provide a brief description of 9.2-4 9.2.1.2 1
- Description the service water system not within the Nuclear Island per R.G. 1.70 Subsection 9.2.3.
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Table 1.9-9 CHAPTER 9 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 9.6 ESW Failure Provide a failure analysis to 9.2-4a 9.2.1.3.1 1 Analysis demonstrate the capability of the service water system to function during abnormally high or low water levels and for preventing organic fouling per N R.G. 1.70 Subsection 9.2.1. y 9.7 ESW System Confirm that no operator action 9.2-8 9.2.1.3.2 1 Z Safety required, following a LOCA, to
- Evaluation start the ESW system in its
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LOCA operating mode. %y , NN I H HH D in H V O
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Table 1.9 CHAPTER 9 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 9.8 ESW System Furnish a description of chloride 9.2-8 9.2.1.3.2 1 Safety control, as for other parts of Evaluation the ESW system 1 not within the Nuclear Island. 9.9 Potable and Provide a description of the 9.2-19 9.2.4.2 1 Sanitary Water supply system (BOP) for the potable and sanitary water systems per R.G. 1.70 83 Subsection 9.2.4. g 9.10 Ultimate Heat Provide a description of the 9.2-21 9.2.5 1 I" Sink ultimate heat sink to be used to g@o 8 e dissipate traste heat from the t* m [ plant per R.G. 1.70 Subsection g a . 9.2.5. e s E 9.11 Condensate Describe the design of the 9.2-21 9.2.6 1 [H Storage Condensate Storage Facilities Facilities per R.G. 1.70 Subsection 9.2.6. 9.12 Condensate Describe the design of the supply 9.2-23 9.2.6 1 Distribution of the condensate system. .' System 9.13 Heated Water Provide a description of the 9.2-38 9.2.8.1 1 System supply portion of the Heated Water System. 9.14 Chilled Water Provide: chiller capacity and 9.2-69 Table 9.2-7 2 W Ej System motor horsepower; and chiller 9.2-70 Table 9.2-8 k> Components water pump and wondenser water 9.2-71 Table 9.2-9 *y pump motor horsepowers. Fa o ik -a l
f'% Q - G D a Table 1.9-11 CHAPTER 11 GESSAR II/FSAR INTERFACES ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 11.1 Source T<erms Provide the source terms for the 11.1-18 11.1.8.3 1 For Components radiological consequences of the Failures failure of the condensaue utorage tank per R.G. 11.1 Section 11.1. 11.2 Rate of By administrative control, limit ll.2-2a 11.2.1.2 11.10d Discharge to discharge from the single (460.13d) Canal discharge line sourced by the y detergent drain tank and one of w the two excess water tanks to a # H maximum of 5 gpm. This discharge zo e can be increased up to 50 gpm CM i provided it can be demonstrated @$ f H that the discharge will meet the requirements of 10.FR20. g H m HH s 11.3 Cost-Benefit Provide the cost-benefit analysis 11.3-13 11.3.2.17 4 " Ratio of the main condenser offgas treatment system. 11.4 Radioactive Provide all release points of ?.l.3-24 11.2.4.1 3,4 Release Points gaseous waste to the environment per R.G. 1.70 Subsection 11.3.3. 11.5 Estimated Perform site-specific evaluations 11.3-26a 11.3.4.6 3,4 Doses of effluent conce..trations and offsite doses for comparison with the design objectives of Appendix I to 10CFR50 and the limits of 10CFR20. yy
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Table 1.9-11 CHAPTER 11 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE MO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATECARY 11.5.1 Shipment of , Demonstrate to the satis (action '11.4-2 11.4.'1.2 3 kaaioactive of the NRC ataff that adequate Wasto storage space to accommodate
', a.t leastA3 containers of' solid- ,
-~ '
ified wdt _ solid 4asto 7 *,ff,L- a'c
- lease 30 ' day'r;bt '.ofe'~ shipment -to
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Ja 'Li.ce,n sed oitsite 'buria.l site. f
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/
- solid Waste. fas.sciibe the proc'ss and equipment e 11.4-4 11.4.2.1 1,2,3 M
# anagement'r used to handle materials that have been activated duripy M tctor
.op,eration per, R.G.',1;;0 U m
H
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Suossation 11(4.*' .
* //' a-hh e s 11. 'J Dry Wastes Provide for the tabulation, 11.4-7 11.4.2.3.2 1 Om I collection, treatment, and f -disposition of dry solid waste
,,- ,' h
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Wg 7 ,' 11.8 i i [' ss to - Process and Describe the inplant airborne 11.5-1 '11.5 1 m-( Effluent . radiation monitoring system. -
~
11.9 Process . Confirm description of 11.5-26 ll'.t.,5 1,i 3 0 Monitoring'and inspections, tests, calibration,,' throbah 1.1.5.92' l' Sampling Tests and maiutenance to be performed 11.5-2.9 ~ hn6 on the process and effluent ,, f 14 5.5.3
, radiological sampling sys,tems. ,, g 7' 11.10. Process Provide information pertaihing Table 11.5-2 1
' Risdiation to gaseous and airborne moni' tor .', 11.5-33 to Monitoring I 11.5-34 System a no
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.- Describe the offsite radiological 11.11' Offsite- .
j 11.6-1 11.6 1,3 o Radiologic'al monitoring program. Monitor 1r.g [Oy j >
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CHAPTE't 1 ' r u* ,:-_. , GESSAR II/FSAR INTEREACES ( '< l - ,, g f, m ,_
. m 't ITEM RELATED' INTERFACE ' -
NO. ' SUBJECT - DESCRIPTION PAGE' SUBSECTION QUESTION CATEGORY 12.1 ALARA Exposures Describe onsite' inspections to 12.1-2 12.1.1.1 3 Design.and determine that the design and Construction operation keeps radiation
- Policies exposures ALARA, where required.
12.2 ALARA Exposures Describe operational policies to 12.1-2 12.1.1.2 3 Operation maintain occupational doses ALARA Policies per R.G. 1.70 to Subsection 12.1.1. 12.3 R.G. 8.8 Provide if required further 12.1-3 12.1.1.3.1 3 N discussion pertaining to the compliance to R.G. 8.8. e 12.4 R.G. 8.10 Provide an assessment of b o$ f 12.1-3 12.1.1.3.3 3 occupational radiation exposures r$ ut g against R.G. 8.10. % N 4 12.5 R.G. 1.8 Provide,an assessment of 12.1-3 12.1.1.3.3 3 g occupational radiation exposures un against R.G. 1.8. 12.6 Operational Describe operational consider- 12.1-9 12.1.3 3 U Considerations ations to maintain doses to operators ALARA per R.G. 1.70 Subsection 12.1.3. 12.7 Radioactive Provide the radioactive source 12.2-10 12.2.1.2.7.1 1 Sources in data for the main steam system. Main Steam h) 12.8 Equipment Describe recombiners and 12.3-8 12.3.1.1 1 N Design for condensers relative to how they , 23 ALARA Exposures aid in maintaining the exposure c3 of plant personnel during system H c) operation and maintenance ALARA.
Vd Table 1.9-12 ,
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CHAPTER 12 GESSAR II/FSAR INTERFACES (Continued) _-- ~7
', ," 5 ITEM d REtATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSC{' TION QUESTION CATEGORY 12.9 Plant Design Define the location of the 12.3-14 12.3.1.2 3 for ALARA co'.taminatita control areas in Exposures the plant.
12.10 Gaseous haste Describe the gaseous waste system 12.3-24 12.3.1.4.5.2 1 relative to ALARA considerations. i 12.3.2.)
~
12.11 Shielding Describe Turbine Building and a2.3;39 1 general plant area shielding relative to maintaining the s exposure of plant personnel during U system operation and maintenance 03
^#~ O H q%tn b) 12.11.1 Caution signs Provido unique high radiation 12.3-34 12.3.2.3 12.15b. 3 O U2
[ of High Radia- warning signs at fuel transfer s471.14) F $3 . tion Areas cubicle. 23 w 23 Y 12.12 Vc.ntilation Provide the criteria established 12.3-40 12.3.3.1 3 C U) bJ Design for the changeout of the air , I Objectives filters and absorbers in the air cleaning system during plant operation per R.G. 1.70 Subsection 12.3.3. 12.13 Ventilation Provide the details of the air 12.3-40 12.3.3.2 1 Design cleaning unit design per R.G. 1.70 Description Subsection 12.3.3. 12.14 Ventilation Describe the ventilation system 12.3-48 12.3.3.2.9 1 Design design for the turbine building and z s3 12.3.3.2.10 0 bJ environment and the central service facility per R.G. 1.70 .4 3_ Subsection 12.3.3. o wo mM 1 t O O O
O O O Table 1.9-12 CHAPTER 12 CESSAR II/FSAR INTERFACES (Continued) ITEM PELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 12.15 Ventilation Supplement the description of the 11.3-48 12.3.4.1 1 Radiation control room ventilation radiation Monitoring monitoring system per R.G. 1.70 Subsection 12.3.4. w W 03 W 2
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(Q V - Table 1.9-14 CHAPTER 14 GESSAR II/F5AR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTIOt* CATEGORY 14.14 Individual Make the plant startup test 14.2-26 14.2.12 14.42 3 Test specification available to the (640.06) Descriptions site inspector. 14.15 Preoperational Provide balance-of-plant tests 14.2-26 14.2.12.1 14.42 3 Test listed in Table 14.2-1. (640.41) Procedures. 14.15 Preoperationa) Provide the preoperational test Test procedures for the following 3 O 03 g Procedures systems: + 20 cM y a. Reactor Feedwater 14.2-27 14.2.12.1.2 Qg ek M>
- b. Reactor Feedwater Pump Driven 14.2-28 14.2.12.1.3 >W
$a Control N g W ww
- c. Plant Process Sampling 14.2-59 14.2.12.130 m (Radwaste)
- d. Seismic Monitoring 14.2- 14.2.12.1.56 105nn
- e. RHR Service Water 14.2- 14.2.12.1.58 105nn
- f. Condensate Makeup 14.2- 14.2.12.1.59 Demineralizer 105nn WN
- g. General Service Water 14.2- 14.2.12.1.60 $pN 106a . q o
Circulating Water
- h. 14.2-106a 14.2.12.1.61 $$
a t
Table 1.9-14 CHAPTER 14 GESSAR II/FSAR INTERFACUS (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 14.16 i. Main Turbine Electro- 14.2- 14.2.12.1.62 (Cont'd) Hydraulic Control 106e
- j. Condensate 14.2- 14.2.12.1.63 106a
- k. Condensate Polishing 14.2- 14.2.12.1.64 Demineralizer 106a N
- 1. Condensate Storage 14.2- 14.2.12.1.65 w 106a C)
- ZO c m. Emergency Response 14.2- 14.2.12.1.74 14.8 CM Information 106e (640.07) h$
M :p H 14.17 Startup Test Provide the locations to be 14.2- 14.2.12.3.14.3 14.42 3 >M f Number 14 monitored and the predicted 130c (640.41) >4 A displacements for the monitored locations for the ,startup Test Number 14 - System Expansion. 14.18 Startup Test Provide the increase in heat flux 14.2- 14.2.12.3.21.2.4 14.42 Number 21 and dome pressure criteria for the 144b (640.41) full reactor isolation portion of Startup Test Number 21 - Main Steam Isolation Valves. WN ON
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4 Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES 1 1 ITEM RELATED NO. INTERFACE
- SUBJECT- DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY [
i j 19.1 Compliance of Provide a description of
' 19.3.1.6-3 19.3.1.6 1.6 3 Gaseous compliance of the gaseous (460.09)
Radwaste System radwaste system with the geology t to NUREG-0800 and hydrology considerations of 4 Section 15.7.3 NUREG-0800 Section 15.7.3. , 19.2 Containment and Provide periodic leak tests of 19.3.1.7-1 19.3.1.7 1.7a 3 i Primary coolant the containment and primary (460.18a) ha l l Sampling coolant sampling stations to the isolation valves. y & I 2:
= 19.2.1 Item II.F.1.3 Provide response to Item 19.3.1.9 19.3.1.9 1.9
- of NUREG-0737, II.F.1. 3 of NUREG-0737. (471.18) 3 $@
t* un am High Range Radiation yg yy
. . Monitors a
g F4 F4 19.3. ADS Per BWROG-8260, implement 19.3.1. 19.3.1.11 1.11 5
, Modification option 2, option 4 or renegotiate 11-1 (421.13) l NRC approval for another option.
I O 19.3.1 Plant Specific Provide plant specific infor- 19.3.3.6 19.3.3.6 3.6 (220.4) 3 Information mation for Table 3.8-3 -1 l 19.3.2 Vertical Verify that vertical floor 19.3.3.54 19.3.3.54 3.54 3 i Response response spectra is within -1 (220.13)
, Spectra the envelope used-for design i
of equipment, system and component.
- 19.3.3 Pool Dynamic Confirm that structures adja- 19.3.3.61 19.3.3.61 3.61 3 Loads and Their cent to Reactor Bldg. experi- -1 (220.20) i Effect on ence insignificant amount of 28 hJ Reactor impact from pool dynamic loads. k $$
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Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.3.4 Loads used in Specify loads which are considered 19.3.3 19.3.3.111 3.111C 3 Cualification in the qualification testing. .111-2 (271.11) Testing 19.3.5 Compliance to Meet Quality Group "C" require- 19.3.3 19.3.3.118 3.118 3 Quality Group ments for Diesel Generator cool- .118-1 [MEB (DGER)
"C" of Various ing, starting, lubrication, combus- Item No. 6]
Diesel Generator tion, air intake and exhaust w 03 Systems systems. r4 ZO . 19.3.6 Sketches of Provide sketches of postulated 19.3.3 19.3.3.139 3.139 3 c: M
- Postulated Pipe pipe rupture locations and .139-2 [MEB (SER) Qy 4, Rupture related data. Item No. 27] Mp
>N w W 19.3.7 Use of First Provide a list of piping sub- 19.3.3 19.3.3.147 3.147 3 H e
8 Footnote on systems which use first footnote. .147-1 [MEB (DSER) $H p F1 ping Sub- Item No. 35] t' Systems g 19.3.8 Operational Provide operational transient for 19.3.3 19.3.3.160 3.160 3 Transient preoperational testing of non-NSSS .160-1 [HSB (DSER) piping system. Item No. 48] 19.3.9 Annulus Provide results of annulus 19.3.3 19.3.3.171 3.171 3 Pressurization pressurization. .171-3 [MEB (DSER) Loads Item No. 59] W h) o Na
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' O) J b)i Q.
Table 1.9-19 CHAPTER 19
- GESSAR II/FSAR INTERFACES (Continued) i ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY
) 19.3.10 Leak Testing Provide details of the Leak 19.3.3 19.3.3.177 3.177 3 Program Testing Program. .177-3 [MEB (DSER) Item No. 65] 19.3.11 Fuel and GE will resolve the internal 19.3.4.8 19.3.4.8 4.8 3 Poison Rod pressure issue to the satisfaction (490.06) ! Internal of the NRC staff. U$ Pressure DD H ZO
. 19.4 Instrumentation Provide instrumentation for 19.3.6 19.3.6.5 6.Sa 1 c ps j) for ESF Filter measuring flow rates through .5-1 (460.11a) @$
- a. Systeun the ESF filter system in p3
>' W p
- accordance with Regulatory #
[t Guide 1.5.2, Revision 2. s HH y 19.5 Control Room Dascribe the type of recording 19.3.6.5-1 19.3.6.5 f.5b 1 um Recording device for recording pertinent (470.11b) Devices for pressure drop and flow rates gy ESF Filters in the control room. 19.6 ESF Atmosphere Ensure that the ESF atmosphere 19.3.6.5-1 19.3.6.5 6.5d 3 Cleanup System cleanup systems housing leakage (4 70.11d) Housing Leakage test is performed in accordance Tests with the provisions of Section 6 of ANSI N510-1975 and Position C.2.1 of Regulatory Guide 1.52. M ha m ha
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Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.7 Recording Indicate type of recording 19.3.6.5 19.3.6.5 6.5b 3 Device device in the FSAR. -2 (470.11) 19.7.1 ESF Filter Provide an ESF filter setpoint 19.3.6.5-4 19.3.6.5 6.Se 3 Temperature <225'F. (470.11) Setpoint 19.7.2 Containment Containment repressurization 19.3.6 19.3.6.13 6.13 5 [j Inleakage from calculations will be redone using .13-1 (480.08) oo
- g. Positive a less conservative method. Also Seeling System gg
. consider reducing allocated leak- c: M y ages of the isolation valves. pg
- = .1 :p 19.7.3 Use of Leakage Calculation of dose if transport 19.3.6 19.3.6.24 6.24 5 > Xs
[ Control System delay time is less than 20 minutes. .24-1 (480.19) W g, I F4 F4 N 19.7.4 , Type of Spray Specific spray nozzle and 19.3.6 19.3.6.15 6.15c 2 y Nozzle capabilities. .15-1 (480.10) g 19.7.5 RHF. Intake Demonstrate capability to 19.3.6 19.3.6.16 6.16 3 C' Strainer Loads accommodate hydrodynanic loads. .16-1 (480.11) 19.7.6 Suppression Provide details of the strainers. 19.3.6.18 19.3.6.18 6.18 2
-1 (480.13) 19.7.7 Comply with Size of purge lines and valves 19.3.6.28 6.28 5 19.3.6 the Intent of will be provided. .28-4 (480.23)
BTB CSB 6-4
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O O O Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTE ". FACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.7.8 Leaka9e Add to the technical specifica- 19.3.6 19.3.6.42 6.42 3 Integrity tion leakage integrity test pro- .42-1 (480.37) Test visions for containment isolation Provisions valves. 19.7.9 Design of Provide containment atmosphere 19.3.6 19.3.6.44 E.44 3 Containment monitoring design and hydrogen .44-1 (480.39) Atmospheric sampling points. N Monitoring $ and Hydrogen
- Sampling $@
e o us 1 19.7.10 Hydrogen Specify quantity of hydrogen 19.3.6 19.3.6.45 6.45 3 y U2
. Generation generation. .45-1 *
(4EO.40) y,$$ . H $3
- 19.7.11. Gessar II Amend Gessar II tables. 19.3.7.13 19.3.7.13 7.13 5 r, [
w Tables a1 (421.01) un m ; g 19.8 Testability of Confirm that all indicating and 19.3.7 19.3.7.14 7.14 2 E Indicators annunciating functions can be .14-1 (421.02) O tested during normal plant operation. 19.9 Administrative Confirm that administrative 19.3.7 19.3.7.16 7.16e 3 Procedures procedures do not require .16-1 (421.04) immediate operator actions based solely on the bypass indications.
- xs u au 4 $8 Q
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Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.10 Bypasses Provide a list of automatic and 19.3.7. 19.3.7.16 7.16d 3 manual bypasses for the following: 16-1 (421.04) e Essential /HPCS Service Water System o Diesel Generator Control System o HPCS Power Supply System g Also, identify any other BOP systems which require bypass indication za to comply with Reg. Guide 1.47. CM 7' O U) g3 19.11 IE Bulletins, Provide procedures for determining 19.3.7. 19.3.7.24 7.24a,b,c 3 b$ $ i . Notices, applicability, methods for 24-1 (421.14) gj N
- Circulars factoring in, and design modifi- g y cation associated with various HH N
D' IEB, IEC and IEIN appropriate $ to the facility. g O
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% J' Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued)
ITEM RELA 7TD INTERFACE NO. . SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.12 IEB 79-27 Provide a detailed list of busses 19.3.7. 19.3.7.24 7.24d 3 to components used for normal 24-1 (421.14) shutdown path and alarms associated with loss of such busses in response to IEB 79-27. 19.13 IEIN 79-22 Provide a detailed analysis and 19.3.7. 19.3.7.24 7.24e 3 h3 reenits (per IEIN 79-22) to assure 24-1 (421.14) $ consequential control systems 7 failure due to high-energy line break are bounded by accident O un w j g p analysis in Chapter 15. ]
. > ..J Ha 19.14 Manual Scram Provide instructions in Emergency 19.3.7. 19.3.7.26 7.26 3 W
- Procedures for manually scramming 26-1 (421.16) p,[l w by tripping breakers supplying un power to A & B solenoids for a C given group of rods. E C
19.15 Setpoint Provide parametric values required 19.3.7. 19.3.7.28 7.28 3 Methods to generate setpoints including 28-1 (421.18) accuracy, calibration and drift y margins appropriate to facility. 1 In addition, a sample setpoint calculation should be included to illustrate GE methodology. 19.16 Instrument Provide ev'aluation of instrument 19.3.7. 19.3.7.28 7.28 3 Performance in performance under harsh environ- 28-1 (421,18) Harsh mental conditions based on results Environments of environt 11 qualification g program. o s3
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Table 1.9-19 CIIAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.17 Tech Spec Provide trip setpoint and 19.3.7. 19.3.7.28 7.28 5 Changes allowable value for each 28-1 (421.18) transmitter and trip unit under each " loop" line entry in technical specifications. 19.18 Optical Identify and describe any other 19.3.7. 19.3.7.36 7.36 2 N Icolator Test isolation devices besides those 36-1 (421.26) a3 Results furnished by CE. The response f e should address NRC question 421.26 for such devices furnished hO O u) j, by the applicant (if any). [y
>' #2 H 19.19 Instrument The configuration of the 19.3.7. 19.3.7.46 7.46 3 33 j' Piping instrument piping is the respon- 46-1 (421.36) g[
- e. Configuration sibility of the applicant. The Un location of the containment %
penetrations must be such as to 2: accommodate the limited drop C3 specified (see response to 421.36). 19.20 RPS Equipment Provide information on quality 19.3.7. 19.3.7.48 7.48 1 in Turbine and integrity of turbine stop 48-1 (421.38) Building valve and control valve closure sensors and circuits within the turbine building. Include discussion on channel independence and capability for test and calibration. Using detailed drawings, describe the routing g, g and separation for this trip ow circuitry from the sensors < >' in the turbine building, which is $ w c> U1 4 O O O
e a ; Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.20 RPS Equipment not a category 1 structure, to their Cont'd in Turbine interfaces with the class lE routed Building and separated circuits in the (Continued) Nuclear Island buildings, which are category 1 structure. Assure basis for conformance with Section 4-20 of IEEE 279-1971. ks 19.21 Main Steam Assure Main Steam Drain MOVs are 19.3.7. 19.3.7.49 7.49 3 $ s Drain MOV's located in a protected area out 49-1 (421.39) hh
- of the LOCA consequential damage zone. O un i M un .
a g 19.21.1 Remote Shut- Upgrade RSS from the present 19.3.7.60 19.3.7.60 7.60 5 $$ " e down System single-panel system to a -1 (421.50) X8 j, (RSS) redundant safety grade system. ss us 19.22 Remote Shutdown Describe administrative and 19.3.7. 19.3.7.60 7.60 3 h Z Systems (RSS) procedural control on access to 60-1 (421,50c,e) r C3 RSS panel (s). This should assure access via keys shall r.ot be precluded by the event necessitating evacuation of the control room. 19.23 Remote Snutdown Assure proper location of RSS 19.3.7. 19.3.7.60 7.60 3 System (RSS) panel (s) in fire protective 60-1 (421.50c) area (s) in accordance with Appendix R to 10CFR50. 19.24 Remote Shutdown Provide a description of any 19.3.7. 19.3.7.60 7.60 3 System (RSS) communication systems required 60-1 (421.50f) to coordinate operator actions, m s3 including redundancy and separation. gg
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Table 1.9-19 CHAPTER 19 CESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.25 Remote Shutdown Discuss testing to be performed 19.3.7. 19.3.7.60 7.60 3 System (RSS) during plant operation to verify 60-1 (421.50j) the capability of maintaining the plant in a safe shutdown condition from outside the control room. w 19.26 Control Systems Identify control systems whose 19.3.7. 19.3.7.62 7.62 3 $ ~ failure or malfunction could C2-1 (421.52a) - seriously impact plant safety. y@
- This may be a verification and o y>
e. expansion of list provided in 421.52a response. yy pg Z8 e 8 19.27 Control Systems Indicate which, if any, of the 19.3.7. 19.3.7.62 7.62 3 g[ C' control systems identified in 62-1 (421.52b) y) 421.52a receive power from common M power sources. The power sources $ considered should include all power O sources whose failure or malfunction could lead to failure or malfunction of more than one control system and should extend to the effects of cascading power losses due to the failure of higher level distribution panels and load centers. 19.28 Control Systems Indicate which, if any, of the 19.3.7. 19.3.7.62 7.62 3 control systems identified in 62-1 (421.52c) 421.52a receive input signals from common sensors. The sensors considered should include common $' y" taps, hydraulic headers and impulse <> lines feeding pressure, temperature, *y level or other signals to two or so vi 4 more control systems. 9 9 9
~
fm 1_) \. / \s Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.28.1 Technical Provide limiting condition of 19.3.8 19.3.8.5 8.5 3 Specification operations and Surveillance .5-? (430.04 Details requirements. B.l.d.) 19.28.2 Trip Setpoints Provide trip setpoints, allowable 19.3.3 19.3.8.5 8.5 3 and Allowable values for second level voltage .5-3 (430.04 Values protection sensors and time delay B.l.d.) w devices. g H 19.28.3 Bus Voltage analysis to guarantee 19.3.8 19.3.8.5 8.5 3 2: o g Undervoltage grid voltage with maximum .5-1 (430.04) @y i fluctuation of 5% at interface e. u3
- points. p2 >
+ >% H % e 19.28.4 Elimination of Bus ties will be eliminated and 19.3.8 19.3.8.2 8.2 5 H $ Bus Crosstie GESSAR II drawings and text .2-1 (430.01) yj H m updated. i 19.28.5 Minimum Start- Calculation of actual degraded 19.3.8 19.3.8.4 8.4 3 $ ing voltages of voltage trip setting and verifi- .4-1 (430.03) Electric Motors cation of the 90% and 95% conditions. 19.28.6 Diesel Genera- Qualification of diesel generator 19.3.8 19.3.8.11 8.11 2 tor-Qualifica- sets along with pump combination .11-1 (430.10) tion 19.28.7 Cable Marking Color Coding of associated cables 19.3.8 19.3.8.18 8.18 3 external to the control room .18-1 (430.19) to w oN
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Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.20.8 Periodic Testing Testing of AC and DC electrical 19.3.8 19.3.8.23 9.23 3
& Electrical Distribution System .23-1 (430.25)
Distribution System 19.28.9 ha Class lE DC BOP Feeder Circuit Breaker is 19.5.8 19.3.8.24 8.24 3 w BOP Fe_eder key interlocked with the battery .24-1 (430.25) 8 main circuit breaker, zo C tu u) 19.28.10 Total Loading Provide total loading. 19.3.8 19.3.8.27 8.27a 3 hi $ 1 of Motor .27-1 (430.29) MW . Operated Valve >N g N U> a 19.28.11 Class lE Relay trip setpoint drift 19.3.8 19.3.8.30 8.30 3 m Protective H, H
.3C-1 (430.34) e Relays >
19.28.12 Electrical Procedures for usage of jumpers 19.3.8 19.3.8.33 8.33 3 Jumpers on safety related systems .33-1 (430.37) 19.28.13 Diesel Personnel Training on Diesel 19.3.8 19.3.8.37 8.37 3 Generator Generator .37-1 (430.41) Reliability O NJ
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O O O Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 1 19.29 Control Systems Provide analysis to show that any 19.3.7. 19.3.7.62 7.62 3 malfunctions of the control 62-1 '(421.52d) systems identified in 421.52b&c resulting from failures or malfunction of the applicable common power source or sensors including g3 hydraulic components, are bounded w by the analysis in Chapter 15 and D would not require action or response 20 7' beyond the capability of operate or C D3 us 8 safety systems. Where credit is taken for operator action, identify O$
." p2g; H
the, time available for such ll H action. M3 HH kO 19.30 HVAC Trip Provide (in the technical 19.3.7. 19.3.7.63 7.63 3 Logic specification) the procedure to 63-1 (421.53) be followed in case of a downscale a trip in the HVAC radiation monitoring system. This procedure should be compared with the typical procedure given in response 421.53. N O k3
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O O O Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE j NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.30.1 Smaples for Provide the sampling frequency, 19.3.9 19.3.9.14 9.14 3 Monitoring instrument readings, frequency of 14-1 (281.06) Water Purity measurement, operating chemical and radiochemical limits using bases given in Subsection 9.1.3.2. 19.30.2 Post Accident Supply analytical methods, pro- 19.3.9. 19.3.9.17 9.17a 3 y Sampling cedures, cc.pability and exposure 17-2 (28'.09) w i Capability per necessary to comply with NUREG-0737 g N UREG-07 37. gg m 19.30.3 Shielding Provide. analytical methods, pro- O un 19.3.9. 19.3.9.17 9.17b 3 t1 J2 [ Requirements cedures and exposures. 17-2 (281.09)
. of GDC-19, y! %
H Assuming x
- H Reg. Guide 1.3
~J Source Terms $H g
19.30.4 Post Accident Provide evaluation of analytical 19.3.9. Z 19.3.9.17 9.17c 3 o Sampling procedures and compliance to 17-1 (281.09) Compliance Regulatory Guide. with Reg. Guide 1.97 , 19.30.5 Adequacy of Provide timely and reliable power 19.3.9.17 19.3.9.17 9.17d 3 Power Supply source for the Post Accident -4 (261.09) and Time of Sampling System (PASS). Operation of Post Accident Sampling System 19.30.6 Procedure for Provide the procedure. 15.3.9.17 19.3.9.17 9.17h 3 X8 M Relating Radio -5 (281.09) Nuclide Gaseous .$ -J$ and Ionic Species g@ to Estimate Core ui -J Damage f J
Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.30.7 Verification of Provide analytical methods and 1^.3.9.'7 19.3.9.17 9.17h 3 Reactor Coolant procedure -5 (281.09) Dissolved Oxygen Less than 0.1 pam if Chlorides are Greater than 0.15 ppm w 19.30.8 Relationship Provide the procedure 19.3.9.17 19.3.9.17 9.17f 3 ao g between Radio- -5 (281.09) gg . nuclide Gaseous c: m e and Ionic Species O u) I to Estimate Core N $)
- Damages > X2 w %1 e H i 19.30.9 Private Auto- Provide location of PAX system 19.3.9.45 19.3.9.45 9.45 3 ss y matic Exchange phones, jacks, design, scismic -1 (430.46) y)
System Phone category, and integration with g and Phone Jacks interplant communication. g 19.30.10 Communication Estcblish communication between 19.3.9.46 19.3.9.46 9.46 3 between Remote remote shutdown panel and the -1 (430.47) Shutdown Panel plant through PAX System, CAP and Remainder System, the Sound Fowered Phone of the Plant System or with the Radio Control Unit. 19.30.11 Tabulation of Provide discussion and use of 19.3.9.47 19.3.9.47 9.47 3 Communication circuits and various operating -1 (430.48) System conditions Extension A3 h3 O NJ
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4 Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) j ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.30.12 BOP Lighting Provide BOP lighting 19.3.9.55 19.3.9.55 9.55 3 and Lighting -1 (430.56) Circuits 19.30.13 Capacity of Provide capacity NPSH require- 19.3.9.56 19.3.9.56 9.56 3 Diesel ments for diesel engine fuel oil -2 (480.57) Generator booster pump. j Fuel Oil N Day Tank oo i y 19.30.14 Diesel Generator Provide piping, details of componentss and quality classifica-19.3.9.58
-1 19.3.9.58 9.58 (480.59) 3 $@
O us - am Fuel Oil tion on the engine and engine [ $! 7 g Piping mounted components > Xs ] e Interfaces 38 g 8 i HH ! $! 19.30.15 Diesel Verify and provide the heat rate 19.3.9.68 19.3.9.68 9.68 3 us Generator and the margin -1 (430.69) . Engine Jacket Water Cooling C3 i System 19.30.16 Diesel Gener- Provide jacket water outleakage 19.3.9.70 19.3.9.70 9.70 3 ator Engine to the Air Cooler and Governor -1 (430.71) Jacket Water System along with inleakage and Cooling System outleakage 19.30.17 Diesel Generator Provide specific system configur- 19.3.9.72 19.3.9.72 9.72 3 Engine Jacket ation with regard to the air vent -1 (430.73) j Water Cooling and the Air Purging System System ! Xf M i GM
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Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM REIATED INTERFACE NO. SUBJECT DEL,. .sIPTION PAGE SUBSECTION QUESTION CATEGORY 19.30.18 Division 3 Provide details of Air Purge 19.3.9.73 19.3.9.73 9.73 3 Diesel System -1 (430.74) Generator 19.30.19 Diesel Provide details of Cooling Water 19.3.9.74 19.3.9.74 9.74 3 Generator System primary loop (essential -1 (430.75) service water) w 19.30.20 Information on 19.3.9.17 19.3.9.17 9.17i 3 co s Post Accident -6 (281.09) g Sampling System c: o T A Testing, Frequency, Tube py to En gg g and Operator e Training i HH n 19.30.21 Data Supporting Provide the necessary procedure 19.3.9.17 19.3.9.17 9.17 3 yH the Application -9/10 (281.09) > of Selected Z Chemistry U Procedure 19.30.22 Coded-Call Provide location of loud speakers 19.3.9.44 19.3.9.44 9.44 3 Automatic Paging of CAP System, its power supplies -1 (430.45) System and its integration with other (CAP System) communication systems. MM (D M
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O O O Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.30.23 DG Cooling Provide standpipes and expansion 19.3.9.75 19.3.9.75 9.75 2 Water System tank for the Divisions 1, 2 and -2 (430.76) 3 Diesel-Generator Cooling Water System. Indicate sizes and respective capacities. 19.30.24 Diesel- Assure compliance with the DG 3 y Generator specification that the system w Set is capable of unattended opera- " g tion at 100 percent rated load, zo . voltage and frequency during the C D2 y emergency condition for at least h!$ p2 >
- a. 7 days.
. > xs $I 19.30.25 Division 3 Provide NPSH for the jacket 19.3.9.76 19.3.9.76 9.76 3 HH H D-G Cooling water pumps at both normal and -1 (430.77) " y Water System operating water levels in the expansion tank. D-G Jacket Assure specification compliance 19.3.9.77 19.3.9.77 9.77 3 Water Heater for warm standby conditions in -1 (430.78) accordance with quick-start reliability requirements. Describe jacket water system design operation and excess capacity (if any). 19.30.26 I&C for D-G Provide description of additional 19.3.9.78 19.3.9.78 9.78 3 cooling water instruments, controls, sensors -2 (430.79) systems and alarms for the D-G Cooling Water Systems which are outside GE's scope of supply. Describe testing program to assure 23 M reliability of this equipment. @yl
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Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.30.27 I&C for D-G Provide description of additional 19.3.9.79 19.3.9.79 9.79 3 Starting System instruments, control, sensors -1 (430.80) and alarms for the D-G air starting system which are outside GE's scope of supply. Describe testing program to assure reliability of this equipment. Describe inter- N locks (if any). $ H 19.30.28 Diesel Engine Provide detailed description of 19.3.9.80 19.3.9.80 9.80 3 y@ g Air Start the Diesel Engine Air Starting -1 (430.81) o ui i System System shown in Figures 9.5-14 19.3.9.85 19.3.9.85 9.85 3 Qy f and 9.5-15 of GESSAR II. -1 (430.86) y, 33 s 33 to 19.30.29 Divisions 1 Provide description of the air 19.3.9-81 19.3.9.81 9.81 3 g[j $ & 2 D-G starting system downstream of -1 (430.82) us M o Start System the left and right bank air distributes for the Division N 1 and 2 diesel-generators. C7 19.30.30 Divisions 1,2 Supplement Response 9.82 with 19.3.9.82 19.3.9.82 9.82 3
& 3 D-G Start detailed discussion of differ- -1 (430.83)
Systems ences between the two types of air start systems used for the Divisions 1 and 2 and the Division 3 diesels. 19.30.31 Air Dryer for Provide detailed description of 19.3.9.84 19.3.9.84 9.84 3 D-G Start the air drying system used for -1 (430.85) System the Division 3 diesel generator. (Division 3) g3 g oN 19.30.32 Diesel Engine Indicate the design working 19.3.9.86 19.3.9.86 9.86 3 < >', Air Start pressure for the air start -1 (430.87) o System motors for Division 3 and Ho the direct cylinder injection for Divisions 1 and 2 O O O
O e) Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM NO. SUPIECT RELATED INTERFACE DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.30.33 Diesel Engine Provide the time required for 19.3.9.86 19.3.9.86 9.86 3 Air Start the diesel-generator to reach Systum full speed, voltage and
-3 (430.87) frequency end be ready to accept load for each of the five starts.
19.30.34 Diesel Engine State the pressure at which 19.3.9.86 19.3.9.86 9.86 3 Air Ftart the five start capacity is -3 [ System determined; i.e., compressor (430.87) oo w cut-in, cut-out or mid-point. ga
- c: M n un i'
a= 19.30.35 Diesel Engine Indicate the capacity of the 19.3.9.86 19.3.9.86 9.86 3 E "' g Air Start System air receiver. -3 (430.87) >b 38 e g HH
$ 19.30.36 Air Dryers Provide details of the air 19.3.9.89 19.3.9.89 9.89 3 U3 rn for D-G Start dryers, i.e., type, manu- -1 System facturer, model number, (430.90) h 2:
(Divisions 1 capacity, special features, C3 and 2) principal of operation, etc. Show that the dew point will l be maintained below reccmmended minimum in SRP Section 9.5.6. Also provide details of system operation and maintenance procedures. 19.30.37 Air Compressors Provide pertinent characteristics 19.3.9.91 19.3.9.91 9.91 3 for the D-G Air of the air compressors for the -1 (430.92) Start System D-G Air Start System; i.e., rated air flow in CFM at design presnure, p3 g rated duty, motor HP and duty, motor mu voltage and number of operating bypasses. f>q o wo U1 %3
Table 1.9-19 CHAPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SU3 JECT DESCRIPTION / AGE SUBSECTION QUESTION CATEGORY 19.30.38 Lubrication Describe procedure for recharging 19.3.9.95 19.3.9.95 9.95 3 oil System for of lubricating oil as indicated -1 (430.96) D-G in the maintenance and operating manual 19.30.39 Lubrication Describe the process of adding 19.3.9.96 19.3.9.96 9.96 3 Oil System for lube oil to the sumo during -1 (430.97) D-G engine operation. Discuss extra lube oil storage and y inventory. w co r, 19.30.40 Lubricatica Provide description of additional 19.3.9.97 -19.3.9.97 9.97 3 z c3 . Oil System for instruments, controls, sensors -1 & (430.98) & C D2 y D-G and alsrms for the diesel engine 19.3.9.98 19.3.9.98 9.98 3 Q$ a luorication oil, systems which -1 (430.99) bs > + are outside GE's scope of supply. >N
# g
$i Describe testing program to assure reliability of this equipment. HH %J U3 D M 19.30.41 Detail Descrip- Applicant to provide in detail 19.3.9 19.3.9.99 9.99 3 g tion of Diesel Diesel Generator Lube Oil System 99-1 (430.100) gy Generator Lube Oil System 19.30.42 Diesel Genera- The controls, keepwarm lube oil 19.3.9 19.3.9.102 9.102 3 tor Keepwarm pump capacity, heater capacity, 102-1 (430.103) Lube Oil Data oil, temperature and instrumen-tation, etc. will be provided by the applicant. 19.30.43 Diesel Genera- Confirmation of Meeting Require- 19.3.9 19.3.9.104 9.104 3 tor Lubrication ments by NUREG/CR-0600 or MI 9644 104-1 (430.105) System is by the Applicant. Nu 19.30.44 Diesel Genera- All additional controls above 19.3.9 19.3.9.110 9.110 3 CN tor Combustion General Electric's supplied scope .110-1 (430.111) < [ Air Intake and required for the combustion air o Exhaust Systeu intake and exhaust system will be provided by the Applicant. [O O O O
. . . . _ . -. . . _ _ - . - . - . - . _ - - - .. - . . ~ - . . - .
l < 4 Table 1.9-19 CHEPTER 19 GESSAR II/FSAR INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTIvN CATEGORY 19.30.45 Diesel Genera- Provide additional instrumenta- 19.3.9 19.3.9.111 9.111 3 tor Combustion tion controls, sensors, alarms .111-1 (430.112) Air Intake and in the Diesel Generator Combus-Exhaust System tion Air Intake and Exhaust Alarms Systcm. , 19.30.46 Effect of Provide a specific discussion on 19.3.9 19.3.?.114 9.114 3 u Decreasing l 4 Barometric the effect of decreasing barometric .114-1 pressure on diesel generator (430.115) $ Pressure on g Diesel Genera-performance. yQ l
- Om tor Performance i A {g
+ * > ps . 19.31 Charcoal If' filter units are required, 19.3.11. 19.3.11.9 11.9 3 22 w" Absorber provide the charcoal absorber 9-1 (460.12) g H 1 4 i Thickne:.s thickness. us i D" 19.32 Flow Measuring Provide flow measuring devices 19.3.11. 19.3.11.11 11.11c 1 Devices for for the Fuel Building HVAC, 11-1 (460.14c) o Airborne Containment space-refuel mode, Effluent auxiliary building HVAC, control Release building HVAC and plant vent. ! Pathways Ms M fD N l
<>a I O WO U1 4 l
s Table 1.9-19 CHAPTER'29 GESSAR II/FSAn INTERFACES (Continued) ITEM RELATED INTERFACE NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.33 Solidification Provide _ process parameters, 19.3.11 19.3.11.12 11.12d 3 Process Control operating procedures and .12-1 (460.15d !
^
Program sampling rnd test procedures
~
1 and c) necessary to cc:aply with the s solidification process control Tc ' program. Maintain appropriate records showing conformanc( Nith-4 p',- - ha
,^
the established parametoff.' e , )1 < . f , y > , - a
. t
! F" 19.34 Process and Provide technical spec'ificatiaa*, 1.e.3.11 f~~19.3.11.13 11.13a 3 M'O , f e Effluent for radiological eftlucnts for
] [.1Q-1 ' -c'-
(460.16) , I Radiological grab sampling frequency, mininum* t' u) yy,
." Monitoring .apalysis fraquency and sensitivity , t F' and Sampling for the waste samp151ng and analy- , .
N
- Program in conformance with
- g H
.si s; program, Petermine the consis- ~ ,, a -
R.G. oo 1.21. tenc'y of these items with NUREG-0473, Rev. 2. Also providejthe , .
" ' .r
, ' ,q -- rg ur
[
/
rI turbine building monitoring and j - t: sampling provisions. ^'f , 19.35 Liquid Effluent Provide the informati>n requested 19.3.11 19.3.11.13 11.13b 3 Process Streams 4In, question ll.13b (460.166),for s ' .13-1 (460.16b)
- liquid ef fluent and process ,
/ # '
/ st re ams.
s _- 3 19.36 Radiological i ,dustify any' deviations in IrMio-- 19.3.11 19.3.11.13 11.13e 3 _ Effluent- 'e* llogical e':21uent monitors design .13-1 (460.16c) 1,\ 'ariteric Ekom ANSI N13.10 (1974) l ' Monitors 1" Cand Sections C.4 and 0.6 of ~f'i l R.G. 1.3 9 % , Revision #1. ' E j' -fted.- a- ss
.. . .A > a jr . C>
l MO Gi s3~ , I f
^'
a , 1 f s . a w a a u
*~ 'l ;
. , -s
[ ' - e
' u,./
%) ; .J - .
)
r u;
- / q. s
- - s Table 1.9-19 f (' -
e
! CHAPTER 19. 7' GESSAR II/FSAR INTERFACES (Continued) ~
ITEM RELATED INTERFAC2 NO. SUBJECT DESCRIPTION PAGE SUBSECTION QUESTION CATEGORY 19.36.1 Estimated Exposure due to background 19.3.12.17 19.3.12.17 12.17b. 3 Exposure radiation will be supplied -2 (471.17) i Due to Back- by the Applicant , ground Radia- . l g tion N y 19.36.2 Location of Post Accident Location of the post accident sampling and sample analysis 19.3.12.17-2 19.3.12.17 12.17c (471.17) 3 g ao u
+ Sampling and area along with doses received gg ,
l [ Sample Analysis by the personnel c: trl s Area O CD i w t< tn tej > , s 19.3.14.6 14.6 3 ->W s 19.36.3 Preoperational The Applicant will provide 19.3. 14 Test Data After this information. 6-2 (640.05)' W i y Fuel Loading sU tn a " 19.37 Contaminated Provide analyses of the radio- 19.3.15. 19.3.15.8 15.6 4
$I Liquid Release logical consequences resulting 8-1 (460.17) b Radiological from the release of contaminated U o Consequences liquid to the environs due to a postulated failure of the licmid tank including Cs-137 and ti.e site specific geological and hydrological parameters.
t mN
. -a O
HO U1 4 i
GESSAR II' 22A7007 I 238 NUCLEAR ISLAND REV. 7 1A.9 SHIFT RELIEF AND TURNOVER PROCEDUK2S (NUREG-0737 Item I.C.2) (Cont'd) NRC Position
- The licensees shall review and revise as necessary the plant procedure for shift and relief turnover to assure the following:
- a. A checklist shall be provided for the oncoming and offgoing control room operators and the oncoming shift supervisor to complete and sign. -The following i items, as a minimum, shall be included in the checklist.
- 1. Assurance that critical plant parameters are within allowable limits (parameters and allowable limits shall be listed on the checklist).
L
- 2. Assurance of the availability'and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by a check of the .
control console. (What to chP.ck and criteria for acceptable status shall be included on the checklist) _
- 3. Identification of systems and components that are in a degraded mode of operation permitted by the Technical Specifications. For such systems and components, the length of time in the degraded mode shall be compared with the O
lA.9-1 125L3 . _ , . _ . _ _ _ . - . _ _ . - _ _ _ , _ _ _ - _ _ _ _ - . _ _ _ _ _ _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 1A.9 SHIFT RELIEF AND TURNOVER PROCEDURES (NUREG-0737 Item I.C.2) (Cont'd) NRC Position * (Cont'd) Technical Specifications action statement (this shall be recorded as a separate entry on the checklist). .
- b. Checklict or logs shall be provided for completion by the offgoing and ongoing auxiliary operators and technicians. Such checklists or logs shall include any equipment under maintenance or test that by themselves could degrade a system critical to the prevention and mitigation of operational transients and accidents or initiate an operational transient (what to check and criteria for acceptable status shall be included on the checklist); and
- c. A systen shall be established to evaluate the effectiveness of the shift and relief turnover procedure (for example, periodic independent verification of system alignmelits).
Response
The response to this requirement will be supplied by the applicant.
- This position statement is repeated from Reference 10 since it was not provided in detail in either NUREG-0660 or NUREG-0737 1A.9-2
,-~ sm k
m k. Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued) Qualit-1 Group Quality Safety Classi- Assurance Seismic Principal Componenta Classb LocationC .fication d Requiremente Categoryf' Comments XII (Continued)
- 8. Cable with safety function 2 A,X N/A B I XIII High Pressure Core Spray N co
- 1. Piping within outermost 1,2 C A/B B I (g) Z isolation valve CQ w t* cn g 2. Piping - return test line to condensate storage tank beyond Other O D N/A N/A $$
rt 50 [ second isolation valve e HH tn H
- 3. Piping beyond outermost 2 A B B I (g) isolation valve - other* a i
- 4. ? ump 2 A B B I
- 5. Pump motor 2 .A N/A B I
- 6. Valves - outer isolation and 1/2 A,C A/B B I (g) within XIV Leak Detection 'ystem 50 N mN
- 1. Temperature sensors 2 A N/A B I (1) <p
. 4 o
- 2. Temperature switches 2 X N/A B I (1) Ho
* -a
- Pool suction piping, suction piping from condensate storage tank, test line to pool, pump discharge piping and return line to condensate storage tank.
i i
Table 3.2-1 EQUIPMENT CLASSIFICATION (Continued) Quality Group Quality Safet Assurance " Principal Component a Class Location Classi d fication Requirement e Category f Comments XIV (Continued)
- 3. Pressure transmitters 2 C N/A B I (1)
- 4. Pressure switches 2 X N/A B I (1)
M
- 5. Differential pressure trans- 2 A,C N/A B I (1) W mitters (flow)
Z CO W 6. Differential pressure switches 2 X N/A B I (1) O to Ytn
- 7. Square root converters 2 X N/A B I (1) y
- 8. Differential flow summers 2 X N/A B I (1) $N h
- 9. Differential flow switches 2 X N/A B I (1) Z O
- 10. Timer switches 2 X N/A B I (1)
- 11. Pcwer supplies 2 X N/A B I (1)
- 12. Radiation monitor Other C N/A N/A I (1)
- 13. Instrument lines 2 C,A,D B B I (1)
- 14. Sample lines
- 2/Other C,D C,D B I (1)
- 15. N/A N/A (1) M Flow transmitters Other C N/A <: >
*These sample lines are totally within containment and radiation monitoring provides no isolation *j function. g3
- . - . . ~ . . . - -
O O O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 3.5.3.1 Local Damage Prediction The prediction of local damage in the impact area depends on the basic material of construction of the structure or barrier (i.e., conrete or steel). The corresponding procedures are presented separately. Composite barriers are not utilized in the Nuclear Island for missile protection.
- '3.5.3.1.1- Concrete Structures and Barriers The modified Petry formula 3 is applied analytically for missile penetration in concrete.- To prevent perforation, a minimum con- ]
crete thickness of 2.2 times the penetration thickness determined for an infinitely thick concrete slab is employed. In the event that spalling or scabbing is unacceptable, a minimum concrete thickness of 3 times the penetration thickness determined for an infinitely thick concrete slab is provided. These design proce-s-) dures have been substantiated by full-scale inpact tests in which reinforced concrete panels (12 to 24 inches thick, 3000-psi design strength) were impacted by poles, pipes, and rods simulating tornado-borne debris.4 3.5.3.1.2 Steel Structures and Barriers 5 is applied for steel structures and barriers, The Stanford equation 3.5.3.2 Overall Damage Prediction The overall response of a structure or barrier to missile impact l depends largely upon the location of impact (e.g., near mid-span or near a support), dynamic-properties of the structure / barrier and missile, and on the kinetic energy of the missile. In general, it has been assumed that the impact is plastic with all
=of the initial momentum of the missile transferred to the structure
' ks or barrier and only a portion of'the kinetic energy absorbed as strain energy within the structure or barrier. t 3.5-19
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 3.5.3.2 Overall Damage Prediction (Continued) After demonstrating that the missile does not perforate the structure or barrier, an equivalent static load concentrated at the impact area is determined. The structural response to this load, in conjunction with other appropriate design loads, is eval-uated using an analysis procedure similar to that in Reference 6
~
for rigid missiles, and the procedure in Reference 7 for deform- m n able missiles. _ M 3.5.4 BOP Interface 3.5.4.1 External Missile - Natural Phenomena External missiles generated by natural phenomena used in the design of the Nuclear Island structures are given in Subsection 3.5.1.4. 3.5.4.2 Other External Missiles No other external missiles were considered in the design of the Nuclear Island structure. The main turbine generator shall be of an in-line orientation per Regulatory Guide 1.115 to preclude low trajectory turbine missiles from impacting Seismic Category I Nuclear Island structures. 3.5.5 References (1) C. V. Moore, The Design of Barricades for Hazardous Pressure Systems, Nuclear Engineering and Design, Vol. 5, 1967. (2) F. J. Moody, Prediction of Blowdown Thrust and Jet Forces, ASMF Publication 69-HT-31, August 1969. (3) A. Amirikan, Design of Protective Structures, Bureau of Yards and Docks, Publication No. NAVDOCKS P-51, Department of the Navy, Washington, D.C., August 1960. O 3.5-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O r ( 3.6.2.1.4.2 Piping in Containment Penetration Areas (Continued) detailed stress analysis was performed to demonstrate compliance with the stress limits given in items (1) and (2). (8) Guard pipes are constructed in accordance with the rules of Class MC, Subsection NE, ASME Code Section III where the guard pipe is part of the containment boundary. In addition, the entire guard pipe assembly is designed to meet the following requirements and tests. (a) The design pressure and temperature are not less than the maximum operating pressure and temperature < of the enclosed pipe under normal plant conditions. (b) The design stress limits of Paragraph NE-3131(c) e-) are not exceeded under the loading associated with
\m /
the containment design pressure and temperature in combination with the safe shutdown earthquake (SSE). (c) Guard pipe assemblies are pressure tested in accordance with the ASME Code Section III, Article NE-6000. I I l v 3.6-19 [ !
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 3.6.2.1.4.3 For ASME Code,Section III Class 1 Piping Breaks are postulated to occur at the terminal ends
- of the piping run or branch run. In addition, breaks are assumed to occur at any intermediate location between terminal ends where:
1 (1) The maximum stress range between any two loads (includ-ing the zero load set) as calculated by Equation 10 of NB-3653 for normal and upset plant conditions (including n an operating basis earthquake) exceeds 2.4Sm and.the O cumulative usage factor is greater than 0.1. E (2) The calculated maximum stress range of Equation 10 -n N exceeds 2.4Sm and the stress ranges calculated by either 1 Equation 12 or 13 of NB-3653 exceed 2.4Sm or the cumu-- lative usage factor exceeds 0.1. In the event that two or more intermediate locations cannot be O determir.ed by stress or usage factor limits, a total of two intermediate locations are identified on a reasonable basis ** for each piping run or branch run.
- Terminal ends are extremities of piping runs that connect to structures, components, or pipe anchors that are assumed in the piping stress analysis to act as rigid constraints to piping thermal expansion. A branch connection to a main piping run is a terminal end for a branch run except when the branch and main run is modeled as a common piping system during the piping stress analysis.
o
- Reasonable basis is the highest stress location.
h
-m Where more than two such intermediate locations are possible, using the application of the reasonable basis, those two loca-tions possessing the greatest damage potential is used. A break at each end of a fitting ay be classified as two discrete break locations where the strecc analysis is sufficiently detailed to differentiate stress at each postulated break.
3.6-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 (O,) 3.7.2.1.5.1.1 Reactor Building and Reactor Pressure Vessel (Continued)
~
The effect of the concrete in the annulus region is reflected in the response spectra. The effect was to stiffen the structure which consequently shifted some of the responses; but in general, the accelerations were decreased. _ The drywell is modeled by shell elements while the upper pools are modeled by solid elements. For the ehell elements, the steel and concrete are combined into equivalent materials like concrete, but with varying density and elastic modulus. By varying the thick-ness along with the material properties, the equivalent stiffness and mass properties along the drywell are obtained. In the pool area, the modulus of elasticity of the solid elements is modified according to the actual shear area averaged in the two mutually () ' perpendicular directions. The weight of water in the upper pools is lumped along with the concrete and steel in the solid elements. The two steel cylindrical shells and the vertical diaphragm of the shield wall are modeled by an equivalent single shell. The concrete between the steel shells is not considered to contribute any stiffness, but its weight is included in the model. The reactor pressure vessel (RPV) is modeled by shell elements. The stiffness of reactor internals is neglected, but its mass is lumped onto the cylindrical shell. The weight of water in the vessel is also included. This simplified reactor pressure vessel model has been included in the overall Reactor Building model to provide proper interaction with other structures. P The RPV pedestal is modeled by shell elements while the concrete
- pad around it is modeled by solid elements and linked to the
- i_j pedestal node points by stiff, horizontal shell elements. Steel 3.7-15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3.7.2.1.5.1.1 Reactor Building and Reactor Pressure Vessel (Continued) shells at both faces of the pedestal along with the stiffening T-sections are transformed into equivalent concrete sections. Part of the weight of water between the drywell and concrete pad is lumped to the concrete pad elements. Piping and platform weights are lumped to the pedestal. The base mat is modeled by solid elements while the embedded anchorage for the superstructures is modeled by f.lell elements. A' the base of each structure, a very stiff shell element extended into the basemat is used to reflect the stiffening effect of the superstructures. Also very stiff arms are used to distribute the reactions along the mat rather than concentrate them at single nodes. The arm material arbitrarily has an elastic modulus 400 times that of ordinary steel. 3.7.2.1.5.1.2 Other Buildings O The support buildings with their masses concentrated at the floor levels are represented by a series of mass points located at the same levels including the base mat. Each mass point has six degrees of freedom (i.e., three translational and three rotational) . The center of mass, the centroid, and the center of rigidity at each floor do not coincide. A beam element which has only flexural resistance is defined as passing through the centroid of that floor, while another beam element having only torsional resistance is defined as passing through the center of rigidity. These centers are connected by rigid members. The three-dimensional, double-beam model reflects the torsional and eccentric effects. Since the greatest impact of the rotational motion at base and the twisting movement are expected near the exterior walls, especially at the building corners, an additional nodal point is defined at the corner f2rthest from the mass center. This nodal point is h 3.7-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 3.7.2.5 Development of Floor Response Spectra (Continued) are input simultaneously. Consequently,.the response spectra obtained from these responses include the effects of all three earthquake components. The procedure to envelope various site-soil cases is the same as for the Reactor Building, i The response spectra values are computed for the system natural frequencies in addition to the following 76 frequencies. Frequency Range (Hz) Increment (Hz) O.20 - 3.00 0.10 3.15 - 3.60 0.15 3.80 - 5.00 0.20 5.25 - 8.00 0.25 l( ) 8.50 - 15.00 0.50 16.00 - 18.00 1.00 20.00 - 23.50 1.75 25.00 - 34.00 3.00 L Computer program PD53Cl (Appendix 3C), is used to generate response. spectra. This program has the capability to combine corresponding spectra by the SRSS method. The locations'and damping values of the response spectra generated are given in Tables 3.7-22, 3.7-23, and 3.7-24. See the response to Question No. 3.54 for additional information. ], n I 3.7-23 l
- _ __ . , , _ . _ _ . _ . _ , _ , . , . . _ _ , _ . . . _ _ , . . , . , _ , _ . _ . _ .+
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 3.7.2.6 Three Components of Earthquake Motion The three components of earthquake motion are considered in the building seismic analyses. To properly account for the responses of systems subjected to the three-directional excitation, a statistical combination is used to obtain the net response O t l l l t l l l i 3.7-23a
GESSAS II 22A7007 238 NUCLEAR ISLAND Rev. 4 p. ( ,/ 3.7.2.6 Three Components of Earthquake Motion (Continued) according to the SRSS criterion of Regulatory Guide 1.92. The SRSS method accounts for the randomness of magnitude and direction of earthquake motion. The SRSS criterion, applied to the responses associated with the three components of ground earth- ]nM quake motion, is used for seismic stress computation for steel structural design as well as for resultant seismic member force computations for reinforced concrete structural design. 3.7.2.7 Combination of Modal Responses since only the time-history method is used for seismic system analysis, the response spectrum combination of modal responses is not applied. 3 3.7.2.8 Interaction of Non-Category I Structures with Seismic
) Category I Structures The manner in which the seismic analysis of non-Category I struc-tures is treated depends upon whether they are located adjacent to Seismic Category I structures. Any isolated non-Category I structure, whose failure due to a seismic event does not endanger a Seismic Category I structure, is designed in accordance with codes and standards applicable to non-Category I structures. A non-Category I structure whose failure could endanger 4 Seismic Category I structure is designed to assure that interaction during a seismic event does not occur.
On the Nuclear Island, the only non-Category I structures immediately adjacent to a seismic Category I structure are the Turbine Building and the Central Services Building. These build-ings are designed by the Applicant. (Applicant will supply
, details.)
'w) 3.7-24
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 O 0 - e w m
\ms) 3.7.3.5.2 NSSS Subsystems When the natural frequency of a structure or component is unknown, it may be analyzed by applying a static force at the center of mass. In order to conservatively account for the possibility of more than one significant dynamic mode, the static force is calcu-lated as 1.5 times the mass times the maximum spectral acceleration from the floor response spectra of the point of attachments of multispan structures. The factor of 1.5 is adequate for simple beam type structures. For other more complicated structures, the factor used is justified.
3.7.3.6 Three Components of Earthquake Motion
\' The total seismic response is predicted by combining the response calculated from the two horizontal and the vertical analysis.
3.7-43 .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 3.7.3.6 Three components of Earthquake Motion (Continued) h When the response spectrum method is used, the method for combining the responses due to the three orthogonal components of seismic excitation is given as follows:
-1/2 R. = R . (3.7-21) 1 j= 13
}
where R = maximum, coaxial seismic response of interest (e.g., f3 displacement, moment, shear, stress, strain) in directions i due to earthquake excitation in direction j, (j = 1, 2, 3) . Rg = seismic response of interest in i direction for design (e.g., displacement, moment, shear, stress, strain) obtained by the SRSS rule to account for the nonsimultaneous occurrence of the R..'s. 1J 3.7.3.7 Combination of Modal Responses 3.7.3.7.1 Subsystems Other Than NSSS When the response spectra method of modal analysis is used, con-tributions from all modes, except the closely spaced modes (i.e., the difference between any two natural frequencies is equal to or less than 10%) are combined by the square-root-of-the-sum-of-the squares (SRSS) combination'of modal responses. This is defined mathematically as: N R = [ (R1 ) (3.7-22) i=1 3.7-44
GESSAR Il 22A7007 238 NUCLEAR ISLAND Rev. 15 3.8.3.3.1.3 Load Combinations Structural concrete design load combinations are defined in this subsection. Nonstructural concrete and pool liner design load combinations are the same but a load factor of 1.0 is used through-out for the applicable load cases and normal operating conditions including hot and cold shutdowns. 3.8.3.3.1.3.1 Load Combinations for Service Load Conditions The following load combinations are based on the working stress design method: (1) test 1.0D + 1.0L + 1.0Pt+ 1.0Tt i (2) construction 1.0D + 1.0L + 1.0Tg + W; (_/ (3) normal 1.0D + 1.0L + 1.0Tg + 1.0Rg + 1.0P ;.and (4) severe environmental 1.0D + 1.0L + 1.0Tg + 1.0F ego
+ 1.0Rg + 1.0P +R.
3.8.3.3.1.3.2 Load Combinations for Factored Load Conditions (1) Extreme environmental 1.0D + 1.0L + 1.0Tg + 1.0F egs
+ 1.0P y + R g; (2) abnormal 1.0D + 1.0L + 1.5P a + 1.0T a + 1.ORa + 1. 2 5 Ry ; [
A-(3) abnormal 1.0D + 1.0L + 1.0P + 1.0T a + 1.25Ra + 1. 0R ; (4) abnormal / severe environmental 1.0D + 1.0L + 1.25P
+ 1.0T, + .1.25 F ego + 1.0R, + 1.0 (Y r +Y j + Y,)
+ 1.0 R ;
v 3.8-61
GESSAR II .72A7007 238 NUCLEAR ISLAND Rev. 14 3.8.3.3.1.3.2 Load Combinations for Factored Load Conditions (Continued) (5) abnormal / severe enviroamental 1.0D + 1.0L + 1.0 F ego
+ 1.0H + 1.0T g; ]m9m (6) abnormal / extreme environmental 1.0D + 1.0L + 1.0P a
+ 1.0T + 1.0 F a eqs
+ 1.0R a + 1.0 (Y r + Y. +Ym) 3
+ 1.0R ; and v
(7) severe environmental 1.0D + 1.3L + 1.0Tg + 1.5 F ego
+ 1.GR + 1.0P v .
o Maximum values of P a, Ta, R a, Y r, Y., and Ym are applied simul-3 taneously, as approprinte, in the applicable combinations based on a time-history analysis. Local stresses due to Yr,Y., and Ym 3 may exceed the elastic limit alle.wables; however, there is no loss of function and elastic behavior is assured. h Load combinations and stress limits for steel portions, such as the drywell head and the personnel lock, that perform as pressure boundaries are in accordance with the ASME Code. The drywell refueling bellows design is in accordance with the Expansion Joints Manufacture 1a Association Code. The lower portion of the drywell wall is a composite design. The steel plates act compositely with the concrete between them and the entire section tecomes equivalent to a reinforced concrete section; hence, the preceding loads and load combinations apply. 3.8.3.3.2 Weir Wall Refer to Subsection 3.8.3.3.1. 3.8.3.3.3 Refueling Pool and Operating Floor Refer to Subsecticn 3.8.3.3.1. 3.8-62
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 O 3 .9. 3.2.4.3 . Standby Liquid Control Valvc (Explosive Valve) F
.The~ typical SLC Explosive Valve des i gn is qualified by type cest to IEEE 344-1975. The valve body is designed, analyzed and tested per
- the.ASME Code, Section III, Class 1. The qualification test demon-strated the absence of natural frequencies below 33 Hz and the ability to remain operable after the application of horizontal seismic loading equivalent to 6.5 g and a vertical seismic loading equivalent to 4.5 g at 33 Hz.
1 3.9.3.2.4.4 High. Pressure Core Spray Valves The typical HPCS valve body design, analysis and testing is in , accordance with the ASME B&PV Code, Section.III, Class 1 or 2. The l Class lE electrical-motor actuator is qualified by type test in accordance with IEEE 382-1972, as discussed in Subsection _ t'~5 3.11.2.2.2. A mathematical model of this valve is included in the - HPCS piping system analysis. The analysis results are assured not to. exceed 4.5 g horizontal and 4.0 g vertical (this includes accel-eration due to gravity) acting simultaneously for a safe shutdown earthquake - (SSE) , which is considered as an emergency condition.
. 3.9.3.2.4.5 Control Rod Drive Globe Valves ,
The typical globe valves in the CRD scram discharge volume vent and drain lines are evaluated by analysis and test for operability ! under the design loads that envelop the predicted load's during a r. design basis accident and safe shutdown earthquake. The' valve body I is designed, analyzed and' tested in accordance with the ASME B&PV Code Section III, Class 2 requirements. 'The vendor's analysis ] results indicate that the valves will withstand a maximum accelera-4 tion of 4.5 g horizontal and 4.0 g vertical (this includes accelera-tion due to gravity) acting simultaneously for a safe shutdown , I
\m -
N 3.9-89 1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3.9.3.2.4.5 Control Rod Drive Globe Valves (Continued) earthquake. The acceptance criteria for seismic disturbance operability are as follows: Stresses must remain within the limits specified for Upset Conditions under paragraph NB-3223 of Section III of the ASME Boiler and Pressure Vessel Code. The calcu-lated bolt stresses must be less than the normal pre-stress. The bolt prestress must be less than the limits specified for Upset Conditions under Paragraph NB-3233 of Section III. The valve actuator is operated by plant air. However, the valve is designed to be fail-safe; the safety operation of the valve closure does not depend upon the plant air supply or on electrical operation of the con. trolling solenoid valves. When the valve is in its open position, the yoke springs are held in tension. In the event that the solenoid valves that control the globe valves are de-energized, or the plant air supply is interrupted for any reason, the springs are capable of closing the valve. 3.9.3.2.5 Non-NSSS Valves Non-NSSS valves classified as seismic active are tabulated in Table 3.9-16. Safety-related active valves are designed to per-form their mectianical motion in times of an accident. The oper-ability assurance program ensures that these valves will operate during a seismic and hydrodynamic event. 3.9.3.2.5.1 Procedures Qualification tests accompanied by analyses are conducted for all active valves. Procedures for qualifying electrical and instru-mentation components which are depended upon to cause the valve 3.9-90
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 '
/~Ni sj e
Table 3.9-11(26) FUEL POOL HEAT EXCHANGER (Continued) Allowable Nozzle Forces and Moments (force in 1b. , Actual Loading Criteria moment in ft/lb) Nozzle Loads Nozzle Loads The maximum mo- Note 1 (a)
- ments due to pipe esign pressure and reaction and the Note 1 (b) temperature maximum forces Dead weight, thermal shall not exceed expansion, safa the allowable shutdown earthquake limits Pri: nary stress smaller of 0.75 Su cr 2.4 ASIE Code Section III allowable -
NOTES: j 1(a) The following expression relates the allowable combination of forces and moments: i, Fi I t . l Mi Il<1 i Fo ;i Mo 1 - To p1 _____q M1 Mo where Fi = The largest of the three actual external orthogonal forces (Fx, ry, and Fz) Mi = The largest of the three actual external orthogenal moments (Mx , My , and Mz) Fo = The allewable value of Fi when all moments are zero C Mo = The allcwable value of '41 when all forces are zero.
- Applicant to supply 3.9-283 m, , , . . . _ _ _ . _ . . , - - _ - - . , . _ __ ___ _ _ . _ __. - _ _ - . . - . _ . - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Table 3.9-11(26) FUEL POOL HEAT EXCHANGER (Continued) 1(b) Allowable limits (upset) Nl tl2 N3 N4 F o M Allowamle 1:st:s {'aulted)
.?
O a
- Applicant to supply e
O O 3.9-284
I O O O Table 3.9-12(XA3) HVAC DIESEL GENERATOR BUILDING SAFETY-RELATED MECHANICAL EQUIPMENT AND SUPPORTS IDENTIFICATION AND SEISMIC AND HYDRODYNAMIC LOAD QUALIFICATION
SUMMARY
(Continued) Input g Qualificgon Method Motion Item No. Name Equipment Support Equipment Support Qualification Sunnary(
* *
- 3.10-
- XA3-CC017 Battery room exhaust fan 71/72
* *
- 3.10-
- XA3-CC018 Battery room w exhaust fan 71/72 w w $
i z ' W CO H O tg H NOTES: N$ w MW [ 1. Qualification Method HH g <n H o T = Testing % A = Analysis z T/A = Testing and Analysis U j 2. Input Motion Response spectrum if column entry is a figure number.** Static g-level if column is numeric.
- 3. Qualification Summary Source or location of qualification summary. Entries not preceded by Table or Subsection refer to the number of the qualification report provided by the equipment manufacturer.
WN i
- Applicant to Supply jy
.4
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 APPENDIX 3B y ! CONTAINMENT LOADS 4 {_ This. appendix provides the thermal-hydraulic dynamic loading e. ! methodology for the General Electric Company (GE) Mark III pressure j ' suppression containment' system during a loss-of-coolant accident
-(LOCA), safety / relief valve - (SRV) discharge and related dynamic i
! events. Complete numerical information is provided:for the GE l Mark III Reference (238 Standard) Plant. Information is also 1: provided for other.GE Mark III Standard Plants. This informa-tion and guidance is provided to assist the Applica~nt in evaluating the design conditions for the various structures which. form its i containment system. i 6 l- The NRC draft acceptance criteria for LOCA-related Mark III i containment pool dynamic loads
- will be addressed by the Applicant.
3B.1 INTRODUCTION 1 GE has concluded the confirmatory test program for the Mark III containment configuration. These tests support and confirm the l
. pressure suppression loads that result from the postulated LOCA and from SRV operation. The ccafirmatory program includes a series of scaled multivent tests that demonstrate no significant I- vent interaction effects for the LOCA process. Also included is
,2 3
an evaluation of the full-scale Caorso SRV tests t as described in Attachment A. f During a LOCA and events such as SRV actuation, the structures form-ing the containment system and other structures within the Reactor Building experience dynamic phenomena. This appendix provides the numerical information on the' dynamic loads that these phenomena impose on the Mark III containment system structures. l
- Safety Evaluation Report related to the final design approval of
- the GESSAR II-BWR/6 Nuclear Island Design, NUREG 0979, April 1983 _
i' i
- 3B-1
- _ _ _ . . _ _ . . _ _ . . . ~ . _ . . _ . _ . _ _ _ _ _ _ . _ . . _ _ _ _ . , _ _ _ - _______.__ _ _ _ _.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 8 3B.1 INTRODUCTION (Continued) The loading information is based on either observed test data or conservatively calculated peak values. The LOCA loading combina-tions are presented in the form of bar charts for each of the con-tainment system structures. In addition to defining the timing of the LOCA related loads, the bar charts identify other loading conditions such as seismic accelerations, dead-weight, etc. For each bar on the chart, reference is made to the section of the appendix where specific discussion of the load is presented. To provide a better understanding of the various dynamic loads and their interrelationships, Section 3B.2 contains a qualitative description of sequential events for a wide range of postulated accidents. The air-clearing loading phenomena associated with the actuation of a SRV are also described. 3B.l.1 Confirmatory Testing Impact and impingement load specifications for small structures affected by suppression pool swell are based on the results of the Pressure Suppression Test Facility (PSTF) air tests conducted in March 1974 3 The intent of these tests was to provide conserva-tive design data. It was recognized that the data base would require extension beyond that provided by the air tests and, to achieve this, additional impact tests for both small and large structures were included in the PSTF schedule. .These tests involved measurement of pool swell impact forces on a variety of targets representative of small structures found in the Mark III containment annulus and are discussed in Attachment B. This appendix relies on a large experimental test data base from the PSTF program. See Table 3B-1 for a summary of these tests. The scaling of the large-scale and 1/3-area scale PSTF precludes direct application to the prototype Mark III. Conservative inter-pretation of these tests results employing dimensional similitude 3B-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O b s CONTAINMENT DRYWELL SHIELD BLDG <
. . . ,.A . .*. *t.
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- 7- - ,,m.-m --, ,
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 O TifNYEL
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NH $ENING c 3M*?""" O Figure 3B-6. Plan At Elevation 11 Feet 0 Inches 3B-92
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 f3 3BA.12.5.5.2 Statistical Analysis The statistical analysis assured that the sensitivity of MPP to each variable was governed by that sensitivity within a set of data known to be consistent leaving the question of how to recon-cile the three sets of data in one prediction equation to a sep-arate analysis. Accordingly, the multiple linear regression pro-cedure was used on the three sets of data separately. Least-squares estimates of the coefficients were obtained. The stepwise procedure was used in backward stabilization mode, whereby variables are offered for fitting._ All are fitted at the outset, but the fitting criterion is then successively made more discriminating so that in the end only those variables which make a significant contribution to the fit are left. Variables were retained if they were significant at the 1% level or less. In this procedure, as each variable is forced out, the other variables are re-examined () for possibly making a significant contribution to the fit; this is a desirable feature because the inadvertent partial correlation between some pairs of variables means that one variable can, to some extent, play the role of a second whether or not the second variable is significant. There is, in general, no restriction on the form of each variable, a chosen variable in first degree, second degree (squared) or first-degree cross product with other variables is each treated as separate variables in the regression procedure. These possible forms were systematically examined for their significance, and those forms having both statistical and physical significance were retained. Table 3BA-23 shows the data set from which a coefficient for each variable can be estimated. There is duplication in the case of only three variables: coefficients for MNAP, LNTW, and VOT can be esti-mated from both large and small-scale data. Per Subsection 3BA.12. l 5.1. 4 ( 4 ) , estimates from the large-scale data are preferred. For (O j MNAP, the range of values from the small-scale data was so narrow l as to not give a reliable estimate of the coefficients; thus, the I coefficients from the large scale data were used. For LNTW, there I ! 3BA.12-31
1 GESSAR II 22A7007 238 NUCLEAR IGLAND Rev. 15 3BA.12.5.5.2 Statistical Analysis (Continued) , was little to choose between the large and small scale coefficients; ~ the large-scale coefficient was slightly larger, leading to a more _ conservative prediction at higher temperatures, and so was chosen. For VOT, the coefficient for this variable was not significant in the large-scale data and only barely significant in the small-scale data; for conservatism, the coefficient from the small-scale data was used. Second degree terms were significant for MNAP, AWAQ, and WCL reflecting curvature in the data. Such terms describe parabolas which may not be suitable for extrapolation beyond the range of the data. Accordingly, special consideration was given to VAAQ and MNAQ for which prediction outside the range of the data would be necessary. Description in detail of these considerations follows. The second degree terms were called MNQ2, AWQ2, and WCL2, respectively. With all variables being fit simultaneously in order to confirm that the term or terms for each variable are filling the role called for by the data, it is helpful to see the pattern of the data points after adjustment by all terms in the prediction equation excep one. These partially adjusted, observed values (herein called shell residuals in the sense that they form a shell , I for showing the effect on prediction of those terms) are shown in Figures 3BA-65 through 3BA-72. Also shown by smooth curves is the role played by the term (s) for that variable. Conformance of these curves to the shell residuals indicates that the effect of that variable on MPP has been accounted for in the term (s) used in the prediction equation. I Figure 3BA-65 shows the shell residuals for VAAQ with respect to that variable. These are for the small-scale data from which the coefficient for VAAQ was estimated. Also shown are two straight lines, one the horizontal continuation of the other, which show 3BA.12-32
,- J - . . -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f% b Table 3BA--19 QUENCHER BUBBLE PRESSURE SENSITIVITY TO SRVDL AIR VOLUME Bubble Pressure (psid) SRVDL First Actuation Subsequent Actuation Air Volume Maximum Allowable 3 fl/D at 10-in S/40 Pipe P+ P- P+ P-(ft ) 40 1.0 7.9 -5.4 13.6 -6.8 i 44 1.85 8.7 -5.7 14.9 -7.1 48 2.72 9.3 -5.9 15.7 -7.3 52 3.60 10.1 -6.2 17.2 -7.5 56 4.45 10.9 -6.6 le,5 -7.8 60 5.35 11.5 -6.9 19.3 -7.9 Standard conditions Steam flow rate (in.) = 520 metric tons /hr Pool temperature (T ) = 100*F (first actuation) 120'F (subsequent actuation) Water leg (WCL) = 17.8 ft (5.42 m) l
- Valve opening time (VOT) = 20 msec Quencher suLmergence (SUBM) = 13.92 ft (4.24 m)
O 3BA.14-27 '
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Table 3BA-20 DESIGN VALUE EQUATIONS WITH SUBORDINATE EQUATIONS AND TERMS Subsection
- Equations 3BA.12.5.3 MPPDV = MPP DESIGN VALUE = (PRED + CONF x SIFV) x FACT 3BA.12.5.3 FACT 3BA.12.5.4 PRED = CMSA x PRD1 3BA.12.5.5 PRD1 3BA.12.5.6 CMSA 3BA.12.5.7 CONF
! ~
13A.12.5.8 SIFV = VIFV 3BA.12.5.8 VIFV = VPRD + VIND 3BA.12.5.9 VPRD = VPR1 + VPRM 3bA.12.5.9 VPR1 = CMSA x VVP1 3BA.12.5.10 VVP1 3BA.12.5.9 VPP'4 = PRD1 x VVPM 3BA.12.5.11 VVPM 3BA.12.5.12 VIND = (PROR x PRED) 3BA.12.5.13 PROR 3BA.12.5.14 MNPDV = MNP DESIGN VALUE = PINF x MPPDV/(PINF + MPPDV/ FACT) 3BA.12.5.15 PINF = 1.014 + 0.0980 x SUBM _
- Terms are defined and equations derived in the subsections indicated; indexes are defined in Subsection 3BA.12.5.3.
O 3BA.14-28
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 h h I APPENDIX 3H (DELETED) i I i O . l I l } 2 i
-,-.,,--,w
-----,-,,,-,----,--we.------ - - _ , - - ---n- - - - . . . - , - - - - , . . - - - - _ _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () Table 4.4-1 THERMAL AND HYDRAULIC DESIGN CHARACTERISTICS OF THE REACTOR CORE 218-624 238-748 251-800 General Operating Conditions Reference rated thermal output (MWt) 2,894 3,579 3,833 Design power level for engineered safety features (MWt) 3,016 3,730 3,995 Rated steam flow rate, at 420'F final feedwater temperature (millions lb/hr) 12.453 15.400 16.49 Core coolant flow rate (millions lb/hr) 84.5 104.0 112.5 Feedwater flow rate (millions lb/hr) 12.428 15.367 16.46 system pressure, nominal in steam "
-- dome (psia) 1,040 1,040 1,040 a
System pressure, nominal core y design (psia) 1,055 1,055 1,055 4 Coolant saturation temperature at core design pressure ( F) 551 551 S51 Average power density (kW/ liter) 52.4 54.1 54.1 Maximum LHGR (kW/ft) 13.4 13.4 13.4 Aserage LHGR (kW/ft) 5.7 5.9 5.9 2 Core total heat transfer area (ft ) 61,151 73,303 78,398 Maximum he'at flux (Btu /hr-ft2) 361,600 361,600 361,600 : 2 Average heat flux (Btu /hr-ft ) 154,600 159,500 159,800 Design operating MCPR See Table 15.0-1 ], f3
\j 4.4-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 7 Table 4.4-1 h THERMAL AND HYDRAULIC DES 1GN CHARACTERISTICS OF THE REACTOR CORE (Continued) 218-624 238-748 251-800 General Ogerating Conditions (Continued) Core inlet enthalpy at 420*F FFWT (Btu /lb) 527.8 527.7 527.9 Core inlet temperature, at 420 F FFWT, (*F) 533 533 533 Core maximum exit voids within assemblies (%) 76.0 79.0 76.0 Core average void fraction, active coolant 0.411 0.414 0.412 Maximum fuel temperature (*F) 3,435 3,435 3,435 Active coolant flow area per assembly (in.2) 15.164 15.164 15.164 Core average inlet velocity (ft/sec) 6.82 6.98 7.07 Maximum inlet velocity (ft/sec) 7.90 8.54 8.57 Total core pressure drop (psi) 25.26 26.4 26.74 Core support plate pressure drop (psi) 20.84 22.0 22.32 Average orifice pressure drop Central region (psi) 5.41 5.71 5.78 Peripheral region (psi) 17.95 18.68 19.16 l [ Maximum channel pressure loading l (psi) 14.52 15.40 15.59 Average-power assembly channel l pressure loading (bottom) (psi) 13.28 14.1 14.22 Shroud support ring and lower shroud pressure loading (psi) 24.84 25.7 25.12 Upper shroud pressure loading (psi) 4.0 3.7 2.8
.. O 4.4-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 4.5 REACTOR MATERIALS 4.5.1 Control Rod System Structural Materials 4.5.1.1 Material Spec.ifications
- a. -Material List 4
1 f l The following material listing applies to the control rod drive mechanism supplied for this application. The position indicator and minor nonstructural items are omitted. (1) Cylinder, Tube and Flange Assembly Reason ) Allowed Under R.G. 1.44 Flange ASME SA182 Grade F304 (< 200 F) l Plugs ASME SA182 Grade F304 (< 200 F) Cylinder ASTM A269 Grade TP 304 (< 200*F)
) Outer Tube ASTM A269 Grade TP 304 (< 200 F) l Tube ASME SA351 Grade CF-3 Spacer ASME SA351 Grade CF-3 4.5-1
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 15 I 4.5.1.1 Material Specifications (Continued) (2) Piston Tube Assembly Reason ] Allowed Under R.G. 1.44 Piston Tube ASME SA479 or SA 249 Grade XM-19 Nose ASME SA479 Grade XM-19 Base ASME SA479 Grade XM-19 ~ Ind. Tube ASME SA312 Type 316 (<200 F) Cap ASME SA182 Grade F316 (<200 F) (3) Drive Line Assembly Coupling Spud Alloy X-750 Compression ASME SA479 or SA249 Grade XM-19 Cylinder Index Tube ASME SA479 or SA249 Grade XM-19 Piston Head ARMCO 17-4 PH or its equivalent Piston ASTM A312 Grade TP 304 or (<200 F) Coupling ASTM A269 Grade TP 304 Magnet ASTM A312, A249, or A213 (<200 F) Housing TP 316L (4) Collet Assembly Collet Piston ASTM A269 TP 304 or (<200*F) ASTM A312 TP 304 - Finger Alloy X-750 Retainer ASTM A269 TP 304 Guide Cap ASTM A269 TP 304 h 4.5-2
[ GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
. _j. 4.5.1.1 Material Specifications (Continued)
(5) Miscellaneous Parts Stop Piston ARMCO 17-4 PH or its equivalent O-Ring Spacer ASTM A240 Type 304 i Nut ASME SA479 Grade XM-19 Collet Spring Alloy X-750 Ring Flange ASME SA182 Grade F304 Buffer Shaft ARMCO 17-4 PH or its equivalent Buffer Piston ARMCO 17-4 PH or its equivalent Buffer Spring Alloy X-750 Nut (hex) Alloy X-750 The austenitic 300 series stainless steels listed under ASTM /ASME specification number are all in the annealed condition (with the exception of the outer tube in the cylinder, tube and flange assem-(_, bly), and their properties are readily available. The outer tube is approximately 1/8 hard, and has a tensile of 90,000/125,000 psi, yield of 50,000/85,000 psi and minimum elongation of 25%. l I o V 4.5-2a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () 4.5.2.1 Material Specifications (Continued) Shroud, core plate, and grid - ASME SA240, SA182, SA479, SA312, SA249, or SA213 (all Type 304L). Peripheral fuel supports - ASME SA 479, Type 316L, ASTM A213, Type 316L, ASME SA312 Grade Type-304L .. Core plate and top guide studs and nuts, and core plate wedges - ASME SA479, SA193 Grade B8A, SA194 Grade 8A (all Type-304)
~
Control rod drive housing - ASME SA312 TP-316L, SA182 Type-304L, and ASME SB167 Type Alloy 600. _. Control rod guide tube - ASME SA358 Grade 304LN, SA312 Grade TP-304LN; ASTM A358 Grade 304LN, A312 Grade TP-304LN, A351 () Grade CF8, A249 TP-304LN. Orificed fuel support - ASTM A249 TP-316L, A240 TP-316L, A479 TP-316L. - l Materials Employed in Other Reactor Internal Structures. l-l (1) Shroud Eead and Separators Assembly and Steam Dryer Assembly All materials are 304L or 316L stainless steel. Plate, Sheet and Strip ASTM A240, 304L or 316L Forgings ASTM A182 Grade 304L l Bars ASTM A276 316L Pipe ASTM A312 Grade TP-304L - 4.5-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 4.5.2.1 Material Specifications (Continued) O Tube ASTM A269 Grade TP-304L ] Castings ASTM A351 Grade CF8 (2) Jet Pump Assemblies The components in the Jet Pump Assemblies are a Riser, Inlet Mixer, Diffuser, and Riser Brace. Materials used for these components are to the following specifications: Castings ASTM A351 Grade CF8 and ASTM SA351 Grade CF3 ASTM A276 TP-304L, Bars } ASTM A479 TP-316L ASTM A637 Grade 688 O Bolts ASTM A193 Grade B8 or B8M and ASME SA479 TP-316L
~
Sheet and Plate ASME SA240 TP-304L or 316L Pipe ASTM A358 316L and ASME SA312 Grade 316L Forged or Rolled Parts ASME SA182, Grade F316L, , ASTM B166, and ASTM A637 Grade 688. O 4.5-8
GESSAR II' 22A7007 238 NUCLEAR ISLAND Rev. 15 b(,,/ 5.2.3.3.4 Moisture Control for Low Hydrogen, Covered Arc Welding Electrodes (Continued) Electrodes are distributed'from sealed containers or ovens as required. At the end of each work shift, unused electrodes are returned to the storage ovens. Electrodes which are damaged, wet, or contaminated are discarded. If any electrodes are
- inadvertently left out of the ovens for more than one shift, they are discarded or reconditioned in accordance with manufacturer instructions.
5.2.3.4 Fabrication and Processing of Austenitic Stainless Steels 5.2.3.4.1 Avoidance of Stress / Corrosion Cracking 1 5.2.3.4.1.1 Avoidance of Significant Sensitization
~
The GESSAR II design complies with Regulatory Guide 1.44, Rev. O and with the guidelines of NUREG-0313, Rev. 1. 1 'O 5.2-39
= - - - - . -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 l l 1 I 5.2.3.4.1.2 Process Controls to Minimize Exposure to Contaminants Exposure to contaminants capable of causing stress / corrosion cracking of austenitic stainless steel components was avoided by 9 5.2-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
/^%,
5.4 COMPONENT AND SUBSYSTEM DESIGN
~'
This section will be updated to reflect the addition of dual , (series) valve barriers for test, vent and drain connections which are outside containment and can communicate with the containment atmosphere or suppression pool. This update will be provided before the first Applicant references GESSAR II. 5.4.1 Reactor Recirculation Pumps 5.4.1.1 Safety Design Bases The reactor recirculation system has been designed to meet,the following safety design bases, f^x (1) An adequate fuel barrier thermal margin shall be assured during postulated transients. (2) A failure of piping integrity shall not compromise the ability of the reactor vessel internals to provide a refloodable volume. (3) The system shall maintain pressure integrity during adverse combinations of loadings and forces occurring during abnormal, accident, and special event conditions. 5.4.1.2 Power Generation Design Bases The reactor Eecirculation system meets the following power gener-ation d6 sign bases. ,, >,._ s ' a.
\, AJ .,
I ( i
.i (1) The 'sys' tem shall prov,ide suf ficient flow to remove heat 1
,# N, M +2 . f rom tha . fuel. '
i, c
~
s
.A -
's, %<
o ' .;. s , sg , , . A u
- , u .<,
N q? . 5.4-1 (
' ~
*]
, _ \ s
GE5SAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 5.4.1.2 Power Generation Design Bases (Continued) (2) The system shall provide an automatic load following capability over the range of 75 to 100% rated power. (3) System design shall minimize maintenance situations that would require core disassembly and fuel removal. O i l i l 5.4-la
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O r~% 5.4.1.3 Description The reactor recirculation system consists of the two recirculation pump loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps (Figures 5.4-1 and 5.4-2). Each external loop contains one high-capacity motor-driven recirculation pump, a flow control valve, and two motor-operated gate valves (for pump maintenance). Each pump suction line contains a flow-measuring system. The recircu-lation loops are part of the reactor coolant pressure boundary and are located inside the drywell structure. The jet pumps are reactor vessel internals. Their location and mechanical design are dis-cussed in Subsection 3.9.5, Reactor Pressure Vessel Internals. However, certain operational characteristics of the jet pumps are discussed in this subsection. A tabulation of the important design and performance characteristics of the reactor recirculation system {% is shown in Table 5.4-1. The head, NPSH, flow, and. efficiency curves are shown in Figure 5.4-3. Instrumentation and control description is provided in Subsection 7.7.1.3. The recirculated coolant consists of saturated water from the steam separators and dryers that hcs been subcooled by incoming feed-water. This water passes down the annulus between the reactor vessel wall and the core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure flow into an external manifold from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the driven
' flow for the jet pumps. This flow enters the jet pump at suction inlete and is accelerated by the driving flow. The flows, both driving and driven, are mixed in the jet pump throat section and
{'T . result in partial pressure recovery. The balance of recovery is 5.4-2 w 7 .._,e e- - -r -- - - - ~ -+ w
GESSAR II 22A7007 238 NUCLEAn ISLAND Rev. 15
,q is ,) CONTENTS (Continued)
Section Title Page 6.2.4.1 Design Bases 6.2-96 6.2.4.1.1 Safety Design Bases 6.2-96 6.2.4.1.2 Design Requirements 6.2-97 6.2.4.2 System Decign 6.2-99 6.2.4.2.1 Containment Isolation Valve Closure Times 6.2-100 6.2.4.2.2 Instrument Lines Penetrating Drywell 6.2-100 6.2.4.2.3 Compliance with General Design Criteria and Regulatory Guides , 6.2-101 6.2.4.2.4 Operability Assurance, Codes and Standards, and Valve Qualification and Testing 6.2-101 6.2.4.2.5 Valve Operability and Leakage Control 6.2-102 6.2.4.2.6 Redundancy and Modes of Valve _ Actuations 6.2-103 () 6.2.4.3 6.2.4.3.1 Design Evaluation Introduction 6.2-104 6.2-104 6.2.4.3.2 Evaluation Against General Design Criteria 6.2-105 6.2.4.3.2.1 Evaluation Against Criterion 55 6.2-105 i 6.2.4.3.2.1.1 Influent Lines 6.2-106 6.2.4.3.2.1.1.1 Feedwater Line 6.2-106 l 6.2.4.3.2.1.1.2 HPCS Line 6.2-107 i 6.2.4.3.2.1.1.3 LPCI and LPCS Lines 6.2-107 z 6.2.4.3.'.1.1.4 Control Rod Drive Lines 6.2-108 6.2.4.3.2.1.1.5 RHR and RCIC Lines 6.2-108 6.2.4.3.2.1.1.6 Standby Liquid Control System Lines 6.2-108 ! 6.2.4.3.2.1.1.7 Reactor Water Cleanup System Line 6.2-109 6.2.4.3.2.1.1.8 Recirculation Pump Seal Water Supply Line 6.2-109 l 6.2.4.3.2.1.2 Effluent Lines 6.2-110 l 6.2.4.3.2.1.2.1 Main, RCIC and RHR Steam Lines 6.2-110 6.2.4.3.2.1.2.2 Recirculation System Sample Lines 6.2-110 (")]
\- 6.2.4.3.2.1.2.3 RHR Shutdown Cooling Line 6.2-111 6.2.4.3.2.1.2.4 Reactor Water Cleanup System 6.2-111 6.2-v
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) Section Title Page 6.2.4.3.2.1.3 Conclusion on Criterien 55 6.2-112 6.2.4.3.2.2 Evaluation Against Criterion 56 6.2-112 6.2.4.3.2.2.1 Influent Lines to Suppreasion Pool 6.2-113 6.2.4.3.2.2.1.1 IPCS, HPCS, and RHR Test and Pump Minimum Flow Bypass Lines 6.2-113 6.2.4.3.2.2.1.2 RCIC Turbine Exhaust and Pump Minimum Flow Bypass Lines 6.2-114 6.2.4.3.2.2.1.3 EHR Heat Exchanger Vent and Relief Valve Discharge Lines 6.2-114 6.2.4.3.2.2.1.4 SPCU Discharge Line 6.2-114 6.2.4.3.2.2.2 Effluent Lines from Suppression Pool 6.2-115 6.2.4.3.2.2.3 Conclusion on Criterion 56 6.2-115 6.2.4.3.2.3 Evaluation Against Criterion 57 6.2-116 6.2.4.3.2.4 Evaluation Against Regulatory Guide 1.11 6.2-116 6.2.4.3.3 Evaluation of Single Failure 6.2-116 h 6.2.4.4 Tests and Inspections 6.2-117 6.2.5 Combustible Gas Control in Containment 6.2-118 6.2.5.1 Design Bases 6.2-118 6.2.5.2 System Design 6.2-121 6.2.5.2.1 Principles of Operation 6.2-121 6.2.5.2.2 Hydrogen Monitoring 6.2-122 6.2.5.2.3 Drywell-Containment Mixing System 6.2-123 6.2.5.2.4 Hydrogen Recombiner System 6.2-125 6.2.5,2.5 Containment Purge 6.2-127 6.2.5.3 Design Evaluation 6.2-127 6.2.5.3.1 Sources of Hydrogen 6.2-128 6.2.5.3.1.1 Short-Term Hydrogen Generation 6.2-128 6.2.5.3.1.2 Long-Term Hydrogen Generation 6.2-129 6.2.5.3.2 Analysis 6.2-131 6.2.5.3.3 Controlled Purge Site Dose 6.2-133 6.2.5.4 Tests and Inspections 6.2-134 6.2.5.5 Instrumentation Pequirements 6.2-134 6.2-vi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
) 6.2.1.1.5.4 Bypass Capability Without Containment Spray and Heat Sinks (Continued) there will no longer be a pressure differential across the drywell leakage path, so that leakage flow and containment pressurization will cease. Since leakage into the containment is of limited duration, the maximum allowable area of the leakage path is large.
2 . . Assuming a primary system rupture of 3.8 ft , Figure 6.2-21 shows the allowable leakage. flow path could have an A//E of 5.6 ft2 ,
]
1 As the size of the assumed primary system rupture decreases, the
. magnitude of the differential pressure across any leakage. path also decreases. However, smaller breaks result in an increasingly longer. reactor blowdown period, which, in turn, results in longer
. durations of the leakage. flow. The limiting case is a very small reactor system break which will not automatically result in
, reactor depressurization. For this case, it is assumed that the
!( ) response of the plant operators is to shut the reactor down in an orderly manner at 100 F/hr cooldown rate. This would result in
-the reactor being depressurized and the break flow being termina-i ted within approximately 6 hours. During this 6-hr period, the blowdown flow from the reactor primary system would have swept all ,
the drywell air over to the containment. The blowdown steam would be condensed in the suppression pool, but, in order for this to occur, the water level in the vent annulus would have to'be depressed to the top of upper row of vents. This continuous ! pressure differential, combined with a 6-hr duration, results in the most severe drywell leakage requirement. The maximum allow-i able leakage path area under these circumstances is an A//K of l i 0.02 ft2, ' Based on-the above calculated values, the allowable drywell leak-I age rate as established by the small break accident is 200% of the . drywell volume in 6 hrs at 3 psid. The fact that the leak rate is not exceeded will be verified by periodic tests on the same sched-ule as the primary containment leak tests specified in Appendix J l of 10CFR50. 1 6.2-43
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 6.2.1.1.5.4 Bypass Capability Without Containment Spray and Heat Sinks (Continued) A study has been made of potential cracking of the reinforced con-crete drywell due to shrinkage, thermal gradients, seismic events, small break and LOCA accidents, and combinations of these events (Reference 8). This report indicates no significant cracking of the drywell walls. 6.2.1.1.5.5 Bypass Capability With Containment Spray and Heat Sinks An analysis has been performed which evaluates the bypass capability of the containment for small primary system breaks con-sidering containment sprays and containment heat sinks as means of mitigating the effects of bypass leakage. The flow rate of one containment spray loop is 5250 gpm and is assumed to be initiated no sooner than 10 min after the accident. The suppression pool water passes through the RHR heat exchanger and is injected into the upper containment region. The spray will rapidly condense the stratified steam, creating a homogeneous air-steam mixture in the containment. The available containment heat cinks shown in Table 6.2-9 were considered with variable con-vective heat transfer coefficients based on the local instantaneous air-steam ratio. The shutdown rate was assumed to be 100 F/hr, and the maximum design service water temperature as given in Table 6.2-2 was used. The shutdown rate corresponds to the maxi-mum rate which does not thermally cycle the reactor vessel. This analysis results in an allowable drywell leakage capability of A/ /K of 1. 45 f t 2 The corresponding pressure transient is shown in Figure 6.2-22. O 6.2-44
. - .- . . -. - . . - .. . - . - ~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
) 6.2.1.6.1.3 Construction Phase Drywell Tests 1 6.2.1.6.1.3.l' High-Pressure Drywell Structural Proof Test f - This test is started after the temperature inside and outside the drywell has been maintained at 60 F or higher for the previous
- 24 hours. During this test, the temperature inside and outside
!- the drywell must be-maintained above 60 F. l t' A temporary bank of air compressors is used to pressurize the drywell. The initial step in the test is to raise the.drywell pressure to 10 1 psig and hold for at least 1 hour. During ! this period, the drywell-and containment pressure and temperature ! are-monitored along with the air flow-rate from the compressors [ into the drywell. The leak rate is determined in a gross manner , using the air inflow information. A walk-through gross visual and noise inspection is made on the exposed exterior surfaces of the drywell. Particular attention is paid to discontinuities such
-as the personnel' air lock, equipment hatch, and main steam and feedwater line penetrations.
During the second step, the drywell pressure is raised to 20 l 1.0 psig and the 1-hr. hold, measurements, and inspections are performed as described for the 10-psig hold period. The final step is to raise the drywell pressure to its design .
, value of 30 psig. The pressure must be maintained 1.0 psig
- of this value for at least 1 hour while the drywell pressure and temperature and air inflow are monitored and the visual inspection is completed.
i The structural acceptance criterion is the absence of visual evidence of gross structural failure as determined by the ability i of the structure to meet the subsequent drywell leak rate tests. i 6.2-69 ee + = -se-.- --- , + - ,r , r,-+ -
-r, -,e.,4 mey- w ,-- .m.-+y--.+ . , , - , , - . - .,m.- -
-w.-,y-w-.,---+- -ww- ,-- - - - - - -- - ,-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.1.6.1.3.2 High-Pressure Leak-Rate Test Immediately following the high-pressure structural proof test, the drywell pressurization source is shut off and the change in dry-well pressure and temperature is monitored for the next 30 minutes. The drywell pressure and temperature decay information is used to establish that the drywell leak rate is less than the allowable value. The drywell air-flow rate from the 1-hr structural test holding period is used as a gross check on the drywell leak rate. Figure 6.2-37 shows the expected pressure decay rate for the drywell from the 30-psig starting point, the possible effect of temperature, and the calculated allowable and technical specifi-cation limits. The figure demonstrates that adequate accuracy in the drywell leak rate can be obtained by a 30-min test. The acceptance criterion for the high-pressure leak-rate test is demonstration that the drywell has a bypass A/VK of less than 10% of the A/YE value for bypass capability under DBA conditions (i.e., lecs than 10% of 5.6 ft 2 or 0.56 ft2) , - 6.2.1.6.1.4 Post-Construction Drywell Test The drywell is subjected to periodic low pressure integrated leak , rate tests to confirm continuing adequate leak tightness. These tests will be performed during each refueling. The differential pressure selected for the periodic tests is sufficient to simulate controlling SBE conditions, but slightly less than the differen-tial pressure required to clear the top row of horizontal vents. That is, the head of suppression pool water above the top row of horizontal vents, under test conditions, is sufficient to seal the vents without having to install temporary closures. O 6.2-70
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 p
\ms 6.2.4.2.2 Instrument Lines Penetrating Drywell (Continued) instrument line as possible, and a manually-operated isolation valve just outside drywell and inside the containment.
6.2.4.2.3 Compliance With General Design Criteria and Regulatory Guides In general, all requirements of General Design Criteria 54, 55, 56, and Regulatory Guide 1.11 are met in the design of the Containment Isolation System. A case-by-case analysis of all penetrations is given in Subsection 6.2.4.3.2. l Leakage detection capabilities for remote-manual valves are discussed in Subsection 5.2.5. f_ 6.2.4.2.4 Operability Assurance, Codes and Standards, and Valve Qualification and Testing
]
(j 1 l Protection is provided for isolation valves, actuators and controls ( against damage from missiles. All potential sources of missiles I are evaluated. Where possible hazards exist, protection is ( afforded by separation, missile shields or by location outside l the containment. Tornado missile protection is afforded by the fact that all containment isolation valves are inside the missile-proof Shield, Auxiliary, or Fuel Buildings. Internally-generated missiles are discussed in Subsection 3.5.1, and the conclusion is reached that there are no potentially damaging missiles generated within the Nuclear Island. Dynamic effects from pipe break (jet impingement and pipe whip) are discussed in section 3.6. The arrangement of containment isolation valves inside and outside the containment affords sufficient physical separation such that a high energy pipe break will not preclude containment isolation. The Containment Isolation System is designed in accordance with A Seismic Category I requirementr, as defined in Section 3.7 using ( ) the techniques af Subsection 3.9.2. l l 6.2-101
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.4.2.4 Operability Assurance, Codes and Standards, and Valve Qualification and Testing (Continued)
]
Section 3.11 presents a discussion of the environmental conditions, both normal and accidental, for which the Containment Isolation System is designed. The section discusses the qualification tests required to assure the performance of the isolation valves under particular environmental conditions. Containment isolation valves are designed in accordance with the requirements of ASME Code, Section III. Where necessary, a dynamic system analysis which covers the impact effect of rapid valve closures under operating conditions is included in the design specifications of piping systems requiring containment isolation valves. Valve operability assurance testing is dis-cussed in Subsection 3.9.3. Valve Operability and Leakage Control 6.2.4.2.5
]
Provisions for demonstrating the operability of isolation valves are discussed in Subsection 3.9.3. Subsection 6.2.6 describes leakage rate testing of containment isolation barriers. Table 6.2-25 indicates power-operated and automatic isolation valves which are cycled during normal operation to assure their l operability. Provisions of the Containment Isolation Valve Leakage Control Sys-tem are applied to the systems penetrating the containment and l l terminating in the environment. The three leakage control systems are the Main Steam Isolation Valve Leakage Control System, Air l Positive Leakage control System and Water Positive Leakage Control l System. The systems are detailed in Subsection 6.7 and 6.5.3.1.1. O 6.2-102
- - . , _ . ..- - . - -~- .-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () 6.2.4.2.6 Redundancy and Modes of Valve Actuations- ] The main objective of the Containment Isolation System-is to provide environmental protection by preventing releases of radio-active materials. This is accomplished by complete isolation of system lines penetrating.the primary containment. Redundancy is provided in all design aspects to satisfy the requirement that no active failure of a single valve or component prevents containment j isolation. Mechanical components are redundant, in that isolation valve arrangements provide backup in the event of accident conditions. Isolation valve arrangements satisfy all requirements specified in NRC General Design Criteria 54, 55, 56 and 57, and Regulatory
- Guide 1.11.
Isolation _ valve arrangements with appropriate instrumentation are () described in Table 6.2-25 and shown in the P& ids. The isolation valves have redundancy in the mode of actuation, with the primary mode being automatic and the secondary mode being remote manual. i i. A program of testing (Subsection 6.2.4.4) is maintained to ensure valve operability and leaktightness. The design specifications require each isolation valve to be operable under the most severe l operating conditons that it may experience. Each isolation valve is afforded protection by separation and/or adequate barriers from the consequences of potential missiles, i Electrical redundancy is provided in isolation valve arrangements, eliminating dependency.on one power source to attain isolation. Electrical cables for isolation valves in the same line are routed 1 separately. Cables are selected and based on the specific environ-ment to which they may be subjected (e.g. , magnetic fields, high radiation, high temperature and high humidity). d O 6.2-103
. ,-._.,--.. _ __ _.~._-~-~._-, __, - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.4.2.6 Redundancy and Modes of Valve Actuations (Continued) ] Provisions for administrative control and/or locks ensure that the position of all nonpowered isolation 7alves is maintained and known. The position of all power-operated control valves is indicated in the control room. Discussion of instrumentation and controls for the isolation valves is included in Chapter 7. 6.2.4.3 Design Evaluation 6.2.4.3.1 Introduction The main objective of the containment isolation system is to pro-vide protection by preventing releases to the environment of radioactive materials. This is accomplished by complete isolation of system lines penetrating the primary containment. Redundancy is provided in all design aspects to satisfy the requirement that any active failure of a single valve or component does not prevent containment isolation. Mechanical components are redundant, such as isolation valve arrangements to provide backup in the event of accident conditions. Isolation valve arrangements satisfy requirements specified in General Design Criteria 54, 55, 56, and 57, and Regulatory l Guide 1.11, as noted on Table 6.2-25. The arrangements with appropriate instrumentation are described l in Table 6.2-25. The isolation valves have redundancy in the mode of actuation with the primary mode being automatic and the secondary mode being remote manual. A program of testing, described in Subsection 6.2.4.4, is maintained to ensure valve operability and leaktightness. The design specifications require each isolation valve to be operable under the most severe operating conditions that it might experience. Each isolation valve is afforded protection by l l 6.2-104
GESSAR II- 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.4.3.1 Introduction (Continued) separation and/or adequate barriers from the consequences of potential missiles. Electrical' redundancy is provided in isolation valve arrangements which eliminates dependency on one power source to attain isola-tion. Electrical cables for isolation valves in the same line have been routed separately. Cables have been selected and based on the specific environment to which they may be subjected, such as magnetic fields, high radiation, high temperature, and high humidity. 3 Provisions for administrative control and/or locks ensure that the position of all nonpowered isolation valves is maintained and known For all power-operated valves the position is indicated in the main control room. Discussion of instrumentation and con-('s
~ trols for the isolation valves is included in Chapter 7.
6.2.4.3.2 Evaluation Against General Design Criteria I This evaluation will be updated to reflect the addition of dual (series) valve barriers for test, vent and drain connections which are outside containment and can communicate with the containment atmosphere or suppression pool. This evaluation will be provided i before the first Applicant references GESSAR II. . ] 6.2.4.3.2.1 Evaluation Against criterion 55 The reactor coolant pressure boundary (RCPB), as defined in 10CFR50, Section 50.2 (v) , consists of the reactor pressure vessel, pressure retaining appurtenances attached to the vessel, valves and pipeswhichex{pndfromthereactorpressurevesseluptoand 1 () including the putermost isolation valve. The lines of the reactor coolant pressure boundary which penetrate the containment include 6.2-105
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
~
6.2.4.3.2.1 Evaluation Against Criterion 55 (Continued) provisions for isolation of the containment, thereby precluding any significant release of radioactivity. Similarly, for lines which do not penetrate the containment but which form a portion of the reactor coolant pressure boundary, the design ensures that isola-tion of the reactor coolant pressure boundary can be achieved. 6.2.4.3.2.1.1 Influent Lines Influent lines, which penetrate the containment and drywell directly to the RCPB, are equipped with at least two isolation valves, one inside the drywell and the other as close to the external side of the containment as practical. Protection of the environment is afforded by the isolation valves. Where needed, protection of the containment in case of pipe rupture outside the drywell, but inside containment, is further ensured by extending the drywell in the form of guard pipes. The guard pipes protect against overpressurization of the containment in the unlikely event of a pipe rupture in the containment annulus together with a single active failure of the normally open inner isolation valve. Table 6.2-26 lists the influent pipes that comprise the RCPB and penetrate the containment and/or drywell. The table summarizes the design of each line as it satisfies the requirements imposed by NRC Design Criterion 55. 6.2.4.3.2.1.1.1 Feedwater Line The feedwater line is part of the reactor coolant pressure boundary as it penetrates the drywell to connect with the reactor pressure vessel. It has three isolation valves. The isolation valve inside the drywell is a check valve, located as close as practicable to the drywell wall. Outside the containment is another check valve located as close as practicable to the con-6.2-106
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.4.3.2.1.1.1. Feedwater Line (Continued) i containment wall and farther away from the containment is a motor operated gate valve. The check valve outside containment is provided with a spring closing operator which, upon remote manual signal from the main control room, provides additional seating force on the valve disc to assist in long-term leakage protection. Should a break occur in the feedwater line, the check valves pre-vent significant loss of reactor coolant inventory and offer immediate isolation. During the postulated LOCA, it is desirable I n
^%J l
l l l 6.2-106a i I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 \_/ 6.2.4.3.2.2.2 Effluent Lines from Suppression Pool The RHR, RCIC, LPCS and HPCS suction lines contain motor-operated, remote-manually actuated gate valves which provide assurance of isolating these lines in the event of a break. These valves also i provide long-term leakage control. In addition, the suction piping from the suppression pool is considered an extension of contain-ment, since it must be available for long-term usage following a design basis LOCA, and, as such, is designed to the same quality standards as the containment. The ECCS discharge line fill system suction lines have manual valves for operational purposes. These systems are isolated from the containment by the respective ECCS pump suction valves from suppression pool as listed in Table 6.2-25. The SPCU System suction line has two isolation valves, one powered /)l from Division 1 and one powered from Division 2. However, because the penetration is under water, both the isolation valves are located outside containment. The first valve is located as close as possible to the containment, and the second is located so as to provide adequate separat i on from the first. 6.2.4.3.2.2.3 Conclusion on Criterion 56 In order to assure protection against the consequences of accidents involving release of significant amounts of radioactive materials, pipes that penetrate the containment have been demonstrated to previde isolation capabilities on a case-by-case basis in accord-ance with Criterion 56. In addition to meeting iaolation requirements, the pressure retain-ing components of these systems are designed to thc same quality standards as the containment. C)' q 6.2-115
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.4.3.2.3 Evaluation Against Criterion 57 Lines penetrating the primary containment for which neither Criterion 55 nor Criterion 56 govern comprise the closed system isolation valve group. Influent and effluent lines of this group are isolated by auto-raatic or remote-manual isolation valves located as closely as possible to the containment boundary. See Table 6.2-25 for further details. 6.2.4.3.2.4 Evaluation Against Regulatory Guide 1.11 There are no instrument lines which penetrate the containment from the reactor coolant pressure boundary. However, instrument lines which connect to the RCPB and penetrate the drywell have 1/4-in. ' orifices and manual isolation valves, in compliance with Regulatory Guide 1.11 requirements. 6.2.4.3.3 Evaluation of Single Failure A single failure can be defined as a failure of a component (e.g . , a pump , valve , or a utility such as offsite power) of a safety system whose consequences must be evaluated for that safety system. The purpose of the evaluation is to demonstrate that the safety function of the system will be completed even with that single failure. Active components are defined in Regulatory Guide 1.48 as components that must perform a mechanical or elec-trical motion during the course of accomplishing a system safety function. Appendix A to 10CFR50 requires that electrical systems also be designed against passive single failures as well as active single failures. Chapter 3 describes the implementation of these standards as well as General Design Criteria 17, 21, 35, 38, 41, 44, 54, 55 and 56. 6.2-116
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 /) ( ,/ 6.2.6.2.5 Penetration Types and Leakage Rate Test 6.2.6.2.5.1 General
= et Th6 local leakage rate Type B and Type C tests must be completed a m
prior to attempting the Type A test, and the sum of all leakage y rates must be less than 288 scfh. - 6.2.6.2.5.2 Personnel Locks The personnel locks are located at El 11 ft 0 in., azimuth 120' and El 84 ft 7 in., azimuth 119*, The supplier's instruction manual should be consulted for proper setup and operation of the locks prior to the local leakage rate test. The local test should consist of a pressure test of the lock itself (pressurizing between the doors) and a test of the double inflatable seals on each door. \_- 6.2.6.2.5.3 Equipment Hatch The equipment hatch 13 located at El (-) 5 f t 3 in. , azimuth 220* . The double seals on the hatch cover are provided with a test con-nection to leak test the space between the 0-ring seals. This test must be performed in accordance with the supplier's instruction manual. 6.2.6.2.5.4 Fuel Transfer Tube Blind Flange Seals The fuel transfer tube is located below the fuel transfer pool at approximately El 3P. ft. A local leakage rate test should be per-formed on the blind flange to ensure leaktightness prior to the containment test. Provisions for leak testing the flange-to-transfer tube seal are provided for on the transfer tube via an air pressure connection to a pair of O-rings on the lower surface /~' of the blind flange. 6.2-145
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.6.2.5.5 Electrical and Bellows Penetrations Electrical and bellows penetrations are mechanical penetrations with allowable leak rates for testing. The primary seal tor the electrical penetration consists of two concentric O-rings, with a test connection to permit leak testing the space between the 0-ring seals. There is also a total of 43 mechanical penetrations which utilize bellows. These are testable connections with a test fitting to allow for pressurizing between the two laminations of bellows. The leakage rates are included in the sum of Type B and C tests for limits in accordance with 10CFR50, Appendix J. Individual penetrations should be tested in sequence, using a check list to assure complete coverage, based upon Table 6.2-29. 6.2.6.3 Containment Isolation Valve Leakage Rate Tests (Type C) 6.2.6.3.1 General Type C tests are required on all isolation valves. All testing 'd" will be performed pneumatically except hydraulic testing which can .E be performed on isolation valve Type C tests using water as a sealant provided that: 79 Je _ (1) the valves will be demonstrated to exhibi*. leakage rates that do not exceed those specified in the technical specifications, and (2) the isolation valve seal water system inventory (or air-flow rate of the compressor) is sufficient to provide sealing for at least 30 days at a pressure of 1.10P . See Table 6.2-29 for a tabulation of the corresponding isolation valves (Water Positive Leakage Control System and Water Leg Seal). 6.2-146
i < GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () 6.2.6.3.1 General (Continued) The containment isolation valves are listed in Table 6.2-29. Type C tests are performed by local pressurization using either the pressure decay or flow meter method. The test pressure is applied in the same direction as that when the valve is required to perform its safety function, unless it can be shown that results from tests with pressure applied in a different direction are equivalent or conservative. For the pressure decay method, the-test volume is pressurized with air or nitrogen to at least l' P. Thc rate of decay of pressure of the known test volume is a monitored to calculate the leakage rate. For tha flow meter method, the required test pressure is maintained in the test volume by making up air, nitrogen or water (if applicable) through a calibrated flow meter. The flow meter fluid flow rate is the isolation valve (or Type B test volume) -leakage rate. Type C tests on the positive leakage control systems (air, water and main steam) isolation valves will be conducted in the follow-ing manner: 1 (1) Gang test a valve group using the threshold leakage of a single valve (allocated leakage of the smallest single valve of the group) as the acceptance criteria. (2) If the gang test exceeds the above acceptance criteria, partial gang tests will be performed with the same acceptance criteria. (3) If required, individual valve tests will be performed until the valve with excess leakage is located and corrected. . u.) 6.2-147
I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.2.6.3.2 Tests Type C (pneumatic) tests are performed for all isolation valves except as discussed below. Some of the valves listed in Table 6.2-29 g m are Type C tested with water by virtue of a water leg seal, and g pressurized water seal. The 28 valves to be tested with water are listed below: -
- 1. CRD Pump Discharge (3 valves) *
- 2. Demineralized Water to G33-2020 (2 valves)
- 3. RWCU Discharge to Main Condenser (2 valves)
- 4. Skimmer Drain to FPCC (2 valves)
- 5. Drywell CRW Sump to Class Radwaste (2 valves) @
- 6. Drywell DRW Sump to Dirty Radwaste (2 valves) E
- 7. Demineralized Water to FPCU (2 valves)
- 8. RWCU Backwash Drain (2 valves)
- 9. CCW to Containment (2 valves)
- 10. CCW Return from Containment (2 valves) _
O 6.2-147a
GESSAR II 22A7007
-238 NUCLEAR ISLAND Rev. 14
() 6.2.6.3.2 Tests (Continued)
- 11. NI Chilled Wuter from containment (2 valves)
- 12. Chilled Water to Drywell (2 valves) d*
- 13. Chilled Water from Drywell (1 valve)
- 14. Upper Containment Pool to Main Condenser (2 valves) 6.2.6.3.3 Acceptance Criteria The combined leakage rate of all components subject to Type B and Type C (Subsection 6.2.6.3) tests shall not exceed 60% of L . If repairs are required to meet this limit, the results shall be reported in a separate summary to the NRC, to include the structural conditions of the components which contributed to the
, failure. 6.2.6.4 Scheduling and Reporting of Periodic Tests The periodic leakage rate test schedules for Type A, B and C 1 tests are described in Chapter 16. Type B and C tests may be conducted at any time during normal plant operations or during shutdown periods, as long as the time interval between tests for any individual Type B or C test does not exceed the mar.imum allowable interval specified in Chapter 16. j- Each time a Type B or C test is completed, the overall total leakage rate for all required Type B and C tests is updated to reflect the most recent test results. Type A, B and C test l'D U 6.2-148 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Table 6.2-1 CONTAINMENT DESIGN PARAMETERS Drywell Containment A. Drywell and Containment
- 1. Internal Design Pressure .
(psig) 30 15
- 2. External Design Pressure (psid) -21.0 -0.8 .
- 3. Design Temperature ('F) 330 185
- 4. Net Free Volume (ft 3) 274,960 1.139x10 6
- 5. Design Leak Rate (%/ day) --
0.1%
- 6. Maximum allowable leak rate
(%/ day) -- 1.0%
- 7. Suppression Pool Water Volume (ft 3)
Low Level II} 12,075 117,510 () 8. High Level Suppression Pool Surface 12,315 III 120,460 Area (ft 2) 482 5,900
- 9. Suppression pool depth (ft)
Low Level 19.92 19.92 High Level 20.42 20.42
- 10. Upper pool makeup volume (ft3) 37,665 B. Vent System
- 1. Number of Vents 120
- 2. Nominal Vent Diameter (ft) 2.29
- 3. Total Vent Area (ft2) 495
- 4. Vent Centerline Submergence (Low Level), ft Top row 7.0 Middle row 11.5 Bottora row 16.0 f)
N/
- 5. Vent Loss Coefficient (Varies with number of vents open) 2.5 - 3.5 (1) Including horizontal vents.
6.2-161
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 6.2-2 O ENGINEERED SAFETY SYSTEMS INFORMATION FOR CONTAINMENT RESPONSE ANALYSES Containment Analysis Value Full Capacity Case A Case B A. C_entainment Spray
- 1. Number of RHR Pumps 2 0 0
- 2. Number of Lines 2 0 0
- 3. Number of Heaters 2 0 0
- 4. Flow rate (gpm/ pump) 5,250 0 0 B. Containment Cooling System
- 1. Number of RHR Pumps 2 2 1
- 2. Pump Capacity (gpm/ pump) 7,100 7,100 7,100
- 3. RHR Heat Exchangers
- a. Type - Invertcd U-tube, single pass Shall, multipass Tube, vertical mounting
- b. Number 2 2 1
- c. Heat Transfer Area (ft 2/ unit) 34,770 -- --
- d. Overall Heat Transfer Coefficient 2
(Btu /hr-ft _oF/ unit) 190 -- --
- e. Service Water Flow-rate (gpm/ unit) 7,300 7,300 7,300
- f. Service Water Temperature (*F)
Minimum Design 40 -- -- Maximum Design 100 100 100
- g. Containment Heat Removal Capability per unit, using 100 F Service Water and 185 F Pool Temperature 6
(Btu /hr) 186.7 x 10 -- -- O 6.2-162
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. O Table 6.2-3 ACCIDENT ASSUMPTIONS AND INITIAL CONDITIONS FOR CONTAINMENT RESPONSE ANALYSES A. Components of Effective Break Area (Recirculation Line Break) (ft )
- 1. Recirculation Line 2.127
- 2. Cleanup line 0.062
- 3. Jet Pumps 0.461 B. Primary Steam Energy Distribution ( } (10 6
Btu)
- 1. Steam energy 25.65
- 2. Liquid Energy 500
- 3. Sensible Energy
- a. Reactor Vessel 98.41
- b. Reactor internals (less core) 51.81
- c. Primary System Piping 45.46
- d. Fuel ( 7.22
_ C. Other Assumptions Used in Analysis
- 1. MSIV closure time (sec)
- a. Recirculation Line Break 3.5
- b. Main Steamline Break 5.5
- 2. Scram time (sec) <1 III All energy values except fuel are based on a 32*F datum.
(2) Fuel energy is based on a datum of 285*F. O 6.2-163
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Table 6.2-4 INITIAL CONDITIONS EMPLOYED IN CONTAINMENT RESPONSE ANALYSES A. Reactor Coolant System: (at 102% of rated power and normal liquid levels) 1. 2. Reactor power level (MWt) Average coolant pressure (psia) 3,651 1,040
]
- 3. Average coolant temperature ( F) 549
- 4. Mass of reactor coolant system liquid (lbm) 544,540
- 5. Mass of reactor coolant system steam (lb m) 21,530
- 6. Volume of liquid in veesel (ft3) 10,781
- 7. Volume of steam in vessel (ft 3) 7,632
- 8. Volume of liquid in recirculation loops (ft 3) 742 3
- 9. Volume of steam in steam lines (ft ) 1,454
- 10. Volume of liquid in feedwater system (ft ) 9,982
- 11. Volume of liquid in miscellaneous lines (ft3) 84 B. Containment Drywell Containment
- 1. Pressure (psig) 0 0 Air temperature ( F) 135 90 2.
- 3. Relative humidity (%) 40-50 50-60
- 4. Suppression pool water temperature (*F) 100 100 1
- 5. Suppression pool water volume (ftJ) 9,600 117,510
- 6. Top row vent centerline 7.0 7.0
- 7. Upper pool water temperature (*F) --
100
- 8. Upper pool makeup water volume (ft 3) --
37,665 0 6.2-164
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9
V} r 0 V (V 3 Table 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE RATE TEST LIST (Continued) Inboard Isolation Outboard Isolation Barrier Barrier Penetration. Bellows Test
- Barrier Description / Barrier Description /
Number Description Seal Type Valve Number (16) Notes Valve Number (16) _ Notes 143c Chilled Water from No C** P44FF016 - P44FF017 6 Drywell Coolers 144c RHR H Drain (RHR-A) No C** E12F073A 16 E12F074A 16 144 2 RHR H Drain (RHR-B) No C** E12F073B 16 E12F074B 16 M 145c ESW Supply to H2 Yes C P41FF169 - - - Mixing Blower (Div 1) yg
- 145c ESW Return from H2 Yes C P41FF172 P41FF171 h$
M Mixing Blower (Div 1) yy P :o :o H 146c 24-in. Pipe Sparo No Capped - - - HH 147c 12-in. Pipe Spare No Capped - - - fH 148c ADS Pneumatic Supply No P53FF017A P53FF015A Division 1 U 156c Spare No - - - - - 157c - - - - - - - 158c ESW Supply to H2 No C P41FFil5 P41FFil4 Mixing Blower (Div 2) 160c Air to RCIC Turbine No - 5 E51F078, E51F077 5 Exhaust Line 164c RWCU Pump to Filter Yes C G33F053 G33F054 Demineralizer
- o M 165c ESW Return from No C P41FFil7 P41FFil6 mM me Mixing Blower (Div 2) ~ .D o
- Type A test also required unless otherwise noted.
em **Ramains water filled during Type A test. [o w~
$t s I ii I t a e i I s 1 1 6.35 6.35 6.49 6.35 6.34 and 6.35 6.48 6.48
Table 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE RATE TEST LIST (Continued) Inboard Isolation Outboard Isolation Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description / Number Description Seal Type Valve Number (16) Notes Valve Number (16) Notes 166c Drywell Pressure Yes C T41FF035 9 T41FF036 9 Bleedoff C T41FF050 9 T41FF051 9 168c Upper Containment Pool No A G41FF186 12 G41FF187 12 to Main Condenser N w 178c Air Positive Seal to No A Closed System 7 P61FF006B 8 oo E51F063 z co
- O tn
, 189c Annulus Pressure- No A Capped - - - tn M Sensing Instrument y$ u Line WW [ HH H 190c Annulus Pressure- No A Capped - - - Sensing Instrument g Line O 195c Containment Test - - - - - - Connection B 197c1 Spare Yes A Capped - - - 197c 2 Spare Yes A Capped - - - 197c3 Spare Yes A Capped - - - 197c 4 Spare Yes A Capped - - - yM 198c y Spare Yes A Capped - - - o wo ui -a 198c2 Spare Yes A Capped - - -
- t. J 6.34 t i O O O
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
\m/ 6.3 EMERGENCY CORE COOLING SYSTEMS This section will be updated to reflect the addition of dual (series) valve barriers for test, vent and drain connections which are outside containment and can communicate with the con-tainment atmosphere or suppression pool. This update will be provided before the first Applicant references GESSAR II. ,
6.3.1 Design Bases and Summary Description Section 6.3.1 provides the design bases for the Emergency Core Cooling System (ECCS) and a summary description of the several systems as an introduction to the more detailed design descriptions provided in Subsection 6.3.2 and the performance analysis provided in Subsection 6.3.3.
-~
6.3.1.1 Design Bases G 6.3.1.1.1 Performance and Functional Requirements The ECCS is designed to provide protection against postulated loss-of coolant accidents (LOCA) caused by ruptures in primary l system piping. The functional requirements (e.g., coolant delivery I rates) specified in detail in Table 6.3-1 are such that the system performance under all LOCA conditions postulated in the design satisfies the requirements of 10CFR50, paragraph 50.46, (Acceptance Criteria for Emergency Core Cooling System for Light Water-Cooled Nuclear Power Reactors). These requirements are summarized in Subsection 6.3.3.2. In addition, the ECCS is designed to meet the following requirements: (1) Protection is provided for any primary system line (
\'
break up to and including the double-ended break of the largest line. 6.3-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 6.3.1.1.1 Performance and Functional Requirements (Continued) (2) Two independent phenomenological cooling methods (flooding and spraying) are provided to cool the core. O i l l l
- O 6.3-la
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (h
\s ) 6.3.1.1.1 Performance and Functional Requirements (Continued)
(3) One high-pressure cooling system is provided which is capable of maintaining water level above the top of the core and preventing ADS actuation for breaks of lines less than 1 in. nominal diameter. ' (4) No operator action is required until 10 min after an accident to allow for operator assessment and decision. (5) The ECCS is designed to satisfy all criteria specified in Section 6.3 for any normal mode of reactor operation. (6) A sufficient water source and the necessary piping, pumps and other hardware are provided so that the containment and reactor core can be flooded for possible core heat
-s removal following a LOCA.
v 6.3.1.1.2 Reliability Requirements The following reliability requirements apply: (1) The ECCS must conform to all licensing requirements and good design practices of isolation, separation and common mode failure considerations. (2) In order to meet the above requirements, the ECCS network shall have built-in redundancy so that adequate cooling can be provided, even in the event of specified failures. As a minimum, the following equipment shall make up the ECCS: 1 High Pressure Core Spray (HPCS) [) 1 Low Pressure Core Spray (LPCS)
's ' 3 Low Pressure Coolant Injection (LPCI) Loops 1 Automatic Depressurization System (ADS) 6.3-2
. . _ . . . _- _. . _ _ . _ .- . ... ~__ _._ _ . . , _
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 i j() SECTION 6.5. TABLES I Table Title Page
' ~
6.5-1 (Deleted) j 6.5-2 Standby Gas Treatment System Component Description 6.5-49 6.5-3 Standby Gas Treatment System Failure Analysis 6.5-53 6.5-4 Post-LOCA Secondary Containment Flow Rates to SGTS 6.5-55 6.5-5' Allocated Leakages 6.5-57 i- 6.5-6 F&ilure Analysis 6.5-59 ILLUSTRATIONS Figure Title Page 6.5-1 ' Standby Gas Treatment System - P&I Flow Diagram 6.5-61 . 6.5-2 Control Building Outdoor Air Cleanup Filter i Train 6.5-63' l-
- 6.5-3 STGS Filter Train 6.5-65 6.5-4 Standby Gas Treatment System Flow During LOCA 6.5-67'
! 6.5-5 Short-Term Post-LOCA Annulus Pressure versus l Time Curve 6.5-68 i -
-6.5-6 Long-Term Post-LOCA Annulus Pressure versus Time Curve 6.5-69 6.5-7 Short-Term Post-LOCA Annulus Temperature versus Time Curve 6.5-70 6.5-8 Long-Term Post-LOCA Annulus Temperature versus L Time Curve. 6.5-71 6.5-9 Shield Building Annulus Recirculation and Exchange Fan Flow During LOCA 6.5-72 6.5-10 Post-LOCA RWCU and ECCS Pump Rooms Temperature Response 6.5-73 6.5-11 Post-LOCA Fuel Building Pressure Response 6.5-74 6.5-12 Post-LOCA Fuel Building Temperature Response 6.5-75
) 6.5-13 Air Positive Seal Isolation Valve-Leakage Control System 6.5-77 6.5-v
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 ILLUSTRATIONS (Continued) Figure Title Page 6.5-13a Air Positive Seal Leakage Control System PSI Flow Diagram 6.5-13b Air Positive Seal Leakage Control System P&I Flow Diagram 6.5-14 Water Positive Seal Isolation Valve Leakage Control System 6.5-79 O l ( i 6.5-vi 1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 / 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5.1 Engineered Safety Feature Filter Systems The Engineered Safety Feature (ESF) filter systems consist of the Control Building Outdoor Air Cleanup System (CBOACS) filters and the Standby Gas Treatment System (SGTS) filters. The performance of the Control Building outdoor air cleanup filters under postu-lated accident conditions is discussed in Subsection 6.4.3. The function of the complete SGTS is discussed in this section along with the filtration function. 6.5.1.1 Design Bases 6.5.1.1.1 Power Generation Design Basis O bs- The SGTS shall have the capability of filtering Containment, Drywell, and Secondary Containment areas Purge Exhaust if required. 6.5.1.1.2 Safety Design Bases The CBOACS and the SGTS filter trains are designed to accomplish the following l (1) CBOACS only. _ . Ensure that radiation exposure to operating personnel in the Control Building control rooms resulting from a maximum hypothetical accident, as discussed in Section 6.4,'are within the guidelines of 10CFR50, Appendix A, i General Design Criteria 19. l l v 6.5-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 15 6.5.1.1.2 Safety Design Bases (Continued) (2) SGTS only. Ensure that the offsite radiation exposures (site boundary and low population zone) resulting from postulated - accidents, as discussed in Chapter 15, are within the guidelines of 10CFR100. (3) SGTS only. Ensure that negative pressures are maintained in the Secondary Containment by the SGTS for accident requiring Secondary Containment integrity. (4) Ensure that failure of any component of the filtration trains, assuming loss of offsite power, cannot impair the ability of the system to perform its safety function. (5) Remain intact and functional in the event of a safe 9 shutdown earthquake (SSE). (6) Be consistent with the recommendations of NRC Regulatory Guide 1.52 and the guidance of Section 6.5.1 to NUREG-0800. _ The design bases employed for sizing the filter, fans and duct-work for CBOACS are discussed in Subsections 6.5.1.2 and 9.4.1. The design bases employed for sizing the filters, fans and duct-work of the SGTS filter trains are discussed in Subsection 6.5.1.2. The fission product removal capabilities of the subject charcoal filter trains are derived from NRC Regulatory Guide 1.52, based on compliance with the applicable design requirements given in the guide. 6.5-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () 6.5.1.1.2 Safety Design Bases (Continued) The work, equipment, and materials conform to the applicable requirements and recommendations of the guides, codes and standards listed in Section 3.2. 6.5.1.2 System Design l 6.5.1.2.1 General CBOACS is shown in Figures 9.4-la, b and c. The SGTS is shown in Figure 6.5-1. 6.5.1.2.2 Component Description The Control Building outdoor air cleanup filter train consists of () a prefilter, electric heater, high efficiency particulate air (HEPA) filter, a charcoal adsorber with fire detection tempera-ture sensors, a downstream HEPA filter, and a fan. Specific component design parameters are given in Table 9.4-1. ~
~
i The SGTS filter train consists of a moisture eliminator or demi-ster, electric heater, prefilter, high efficiency particulate air (HEPA) filter, a charcoal adsorber with fire detection temperature scnsor, a downstream HEPA filter, main exhaust fans, decay heat removal fan, dampers, and controls and instrumentation. Specific component design parameters are given in Table 6.5-2. m 6.5-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 6.5.1.2.2 Component Description (Continued) The filter housing design provides adequate space for filter inspection and maintenance. The housing is fitted with necessary parts and compartment lights for inspection. Instruments and test connections are also provided. The charcoal adsorber is provided with a two-step high temperature detection system and a water spray system to allow flooding of the charcoal bed in the unlikely event of high temperature in the charcoal to preclude the possibility of iodine desorption. Size of the filter trains and location of components, auxiliary and instrument connections are indicated in Figures 6.5-2 and 6.5-3, for CBOAC and Standby Gas Treatment filter units respectively. 6.5.1.3 Design Evaluation The following evaluation is written to correspond to the design bases of Subsection 6.5.1.1. 6.5.1.3.1 General (1) The performance capability of the Control Building outdoor air cleanup unit is discussed in Subsection 9.4.1. The design of the individual components, which ensure the capability to perform the safety function, is discussed in Subsection 9.4.1. Control room doses resulting from postulated radiological accidents are given in Chapter 15. These doses are within the guideline values of 10CFR50, Appendix A, General Design Criteria 19. O 6.5-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 C\ \s,) 6.5.1.3.1 General (Continued) (2) Component descriptions and safety evaluations for the SGTS filter trains are provided in Tables 6.5-2 and 6.5-3, respectively. Dose analyses of postulated radiological releases involving the site boundary and low population zone are given in Chapter 15. Radiation exposure resulting from these occurrences is shown to be within the guideline values of 10CFR100. (3) The Control Building outdoor air cleanup and SGTS filter trains each consist of two independent and redundant filtration trains. Should any component in one train fail, filtration can be performed by the other train. The electrical devices of the respective trains are powered from separate Class lE electrical buses. Failure modes s and effects analyses are presented in Tables 9.4-2 and ( \ (_) 6.5-3, for CBOACS and SGTS, respectively. (4) The ESF filter systems are designed to Seismic Category I requirements as specified in Section 3.2. The components and supporting structures of any system, equipment, or structure that is not Seismic Category I, and whose collapse could result in loss of safety function of the ESP filters through either impact or flooding, are analytically checked (and upgraded if necessary) to assure that they will not collapse when subjected to seismic loading. (5) The ESF filter systems are designed and constructed to be consistent with the recommendations of NRC Regulatory Guide 1.52 and the guidance of Section 6.5.1 to - NUREG-0800. ,
/~N \_
6.5-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 6.5.1.3.2 SGTS Capability Each SGTS filter and adsorber unit is capable of maintaining negative pressure in the containment areas while handling and treating the following air volumes: (1) Air leakage is secondary containment during any accident, including thermal transients of pressure and temperature impact on volume flow to the SGTS. (2) Normal primary containment, secondary containment, and/or drywell purge. (3) Abnormal (upset) events within the primary and/or secondary containments. The controlling air volume is that described in item (1) with 5625 scfm total. Therefore, the main SGTS unit fan, which is sized for 6000 scfm, can handle any of the above three situations, with consideration of variations in air volume with time of flow to the SGTS. During accident and post-accident conditions, including a LOCA, the STGS initiates automatically and processes air from the secondary containment, which is, the Shield Annulus, the Fuel Building and the ECCS/RWCU compartments. In addition, the system can be started manually from the Control Room if required. I Both the ECCS, RCIC and RWCU Pump Rooms and Fuel Building are kept at (-)S/8-inch of water with respect to the outside environmental atmospheric pressure. The slight negative gage pressure is normally maintained by the heating and ventilating systems. These systems are discussed in Section 9.4.1. After a LOCA, the primary and secondary containment volumes are automati-cally isolated from the normal exhaust air paths. The SGTS is started automatically and exhaust air is drawn through the SGTS 6.5-6 1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
\/ 6.5.1.5 Instrumentation Requirements (Continued)
Each charcoal filter bed is provided with two temperature switches. Any one of the temperature switches actuates an alarm in the Control Room when the increase in filter temperature is beyond the preset value. A temperature controller is provided upstream of the charcoal bed which shuts off the electric heating coil if the temperature rises above the preset value. The status of the ESF filter train monitored equipment is displayed in the C atrol Room during both normal and accident operations. Monitoring instrumentation for ESF filter trains is discussed in Section 7.5. I See response to question 6.5 in Subsection 19.3.6.5 for additional ' m information. 6
- 6.5.1.6 Materials l
The construction materials used in or on the filter systems are compatible with the normal and accident environments postulated for the area in which the equipment is located as well as the l areas served by the upstream air duct work. The CBOACS filter systems are not exposed to accident environments of extreme temperature or radiation that could potentially produce pyroltic or radiolytic decomposition of filter materials, and thus, these filter system decomposition products will not be present. l !. 6.5-16 J
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 6.5.2 Containment Spray Systems O The containment spray system does not provide any fission product removal function in the Mark III containment. O O i 6.5-16a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 O Table 6.5-1 lO (Deleted) l l O = 6.5-45
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
~O Table 6.5-1 (Deleted) l l
~
O l l 6.5.-46 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 0 - Table 6.5-1 l O (Deleted) l
- O 6.5-47
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 l I Table 6.5-1 l ) (Deleted) O 6.5-48
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GESSAR II 22A7007 4 238 NUCLEAR ISLAND Rev. 15
- O SECTION 6.8
-CONTENTS Page Section Title 6.8 PNEUMATIC SUPPLY SYSTEM 6. 8- 1
~
l 6.8.1 Design Bases- 6.8-1 6.8.2 System Description 6.8-2 6.8.3 Safety Evaluation 6.8-4 , 6.8.4 Inspection and Testing Requirements 6.8-5 l 6.8.5 Instrumentation F.equirements 6.8-6 l-1 i TABLES , i Table Title Page () 6.8-1 6.8-2 Pneumatic Supply System Services Pneumatic Supply System Capacity Requirements 6.8-9 (Per Division) 6.8-10 i . ILLUSTRATIONS i' Figure Title Page i 6.8-1 Pneumatic Supply System P&I Flow Diagram 6.8-11 i 3 i l 6.8-i/6.8-il '
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 1 V 6.8 PNEUMATIC SUPPLY SYSTEM 6.8.1 Design Bases The Pneumatic Supply System is divided into two independent divisions, with each division containing a nonsafety-related continuous air supply and a safety-reinted emergency stored air supply. The stored air supply is Safety Class 3, Seismic Category I, designed-for operatio1. of: (1) the ADS SRV; (2) specified valves of the SGTS; (3) the Hydrogen Mixing System; and (4) the Dryyell and Containment Vacuum Relief System. , f'"g The function of the nonsafety-related, continuous Pneumatic Supply
\-- System is:
(1) to recharge, and maintain fully charged, the ADS SRV i accumulators, and i (2) to offset air leakage at valve operators listed above; (3) the system is protected against the entrance of foreign or corrosive materials into the system; and (4) in the event of a small line break accident, the system will provide air to the ADS valves for a period of 100 days. System pressure is 200 psig maximum, 150 psig minimum. 6.8-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rtv. 15 6.8.2 System Description The Pneumatic Supply System provides dry, oil-free, compressed air for the actuators of the ADS SRV, the specified valves of the SGTS, Hydrogen Mixing and Drywell and Containment Vacuum Relief Systems , (see Figure 6.8-1 for details). The Pneumatic Supply System is separated into two independent divisions. Division 1 supplies air to the ADS valves a'nd
~
accumulators on steamlines A and C and to SGTS, Hydrogen Mixing and Drywell and Containment Vacuum Relief Systems. Division 2 supplies air to the ADS valves and accumulators on steamlines B and D an,d to SGTS, Hydrogen Mixing and Drywell and Containment _ Vacuum Relief Systems. Operation of one division is sufficient to meet the requirements of the systems serviced. To ensure adequate air supply under both normal and accident conditions, each division is separated into a non-safety portion and an assured safety portion. The nonsafety portion of each _ division contains an air compressor and associated air intake filter and after cooler to provide a continuous source of compressed air. This air is dried in a parallel flow air dryer, with its associated prefilters and after filters, and is then directed to the air receiver. Supply headers from the receiver service the air valves. During LOCA, LOPP and/or system low pressure, the compressor supply to the receiver is automatically cut off, and the assured safety air supply valve is opened. The safety air supply consists of a bank of high pressure air bottles, with a pressure-reducing valve, piped directly to the air receiver. The bottled supply is adequate to meet emergency air requirements for a period of seven days. The operator must replace some of the bottles after this period O 6.8-2
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.- GESSAR II ^ 22A7007 84TJCLEAR ISIsro P.cv. 15 system Descr1p' tion (Cont'inued)
^ '
6.8.2 ' in the extreme instance that , additional air is needed and the compressor is still-not available, for'the duration of the post-accident condition. .
-The bottles'are mechanically restrained to preclude generation of high-pressure missiles during an SSE. The, bottles are also covered by a heavy steel plate, which serves as a barrier to potential missiles.
Safety air supply bottles shall be inspected periodically and replaced upon low pressure indibation. Replaced units shall be recharged on the plant site ifea high-pressure air bottle recharg-ing facility is airailable, or b'e returned to the bottle supplier for~ recharging.- 'All replachd' air bottles shall be inspected for ,
' leakage follcwing recharging, before returning to the spare bottles storage area.
l
, 'The nonsafety and safety air supplies are located in the Fuel , Building. These systems serve the ADS valves located in the drywell, the drywell vacuum relief valves located in containment, I
the'SGTS valves and containment vacuum relief valves located in the' Fuel Building. Table 6.8-1 lists the valves supplied by the
- Pneumatic Sbpply System
~
Flow. rate and capacity requirements are divided into an initial requirement and a continuous supply, as listed in Table 6.8-2. An initial requirement of 14.4 scf per ADS SRV provides for two actuations of the valve against 70% of maximum drywell pressure. Fif ty- gal' accumulators supplied for each main steam ADS SRV actua-tor fulfill the steam valve requirement. The continuous supply
, is divided i~nto safety and nonsafety portions.
O_ ' E s l 6.8-3 f ..
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 6.8.2 System Description (Continued)
' Compressed air at a rate adequate to make up the air leakage of each serviced valve is provided by the safety portion. This assumes an air leakage rate per valve of 1 scfh for a period of at least seven days, plus the air required for at least one sctuation per valve. The nonsafety portion providea compressed air at a rate adequate to recharge the ADS SRV accumulators within 40 min.
The continuous supply portion of the pneumatic system, extending
, from the compressor to the isolation valve prior to the air receiver, is not safety related. The air receiver, the emergency air bottle storage and supply lines to the valves served along with associated lines, valves and fittings are classified as safety Class 3, Seismic Category I.
Nonsafety piping and valves of the system are designed to ANSI B31.1, Fower Piping Code, and the requirements of Quality
. . Group'D.of Regulatory Guide 1.26. Pressure vessels and heat exchangers are designed to ASMF. Section VIII, Division I.
System design pressure is 200 psig with the system design temperature at 150*F. The air receiver has.the capacity to provide one actuation of all serviced valves in the'aame division as-denoted ir. Table 6.8-1 without requiring operation of the air compressor. 3
-6.8.3 System Evaluation 2.
Vessels, piping and fittings of the safety portion of the system
- i. are designed t,o Seismic Category I, ASME Code III, Class 3, Quality Group C and Quality Assurance B requirements, except for the piping and valves for the containment and drywell penetrations 6.8-4 t
O O O Table 6.8-1
~
PNEUMATIC SUPPLY SYSTEM SERVICES Division Building Equipment or Service Item No. Reference P&ID 1 Drywell Safety-relief valves B21-F041A K-102 B21-F041E B21-F051C B21-F051G
- 1. Containment Drywell vacuum relief valve T41-FF030A K-168 1 Fuel SGTS Air-Operated valves P38-FF001A K-160 .
P39-FF003A g I w 1 " i Fuel Containment vacuum relief valve T41-FF033A K-168 4 z CO i
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9s 2 Drywell Safety / relief valves B21-F041B K-102 , B21-F041F ms B21-F047D $ B21-F047H g . 2 Containment Drywell vacuum relief valve T41-FF030B K-168 2 Fuel SGTS Air-Operated valves P38-FF001B K-160 P38-FF003B i 2 Fuel Containment vacuum relief valve T41-FF033B K-168 ,
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Table 6.8-2 PNEUMATIC SUPPLY SYSTEM CAPACITY REQUIREMENTS (Per Division) Total Capacity or Flow Rate Division-1 Division-2 I INITIAL REQUIREMENT Two actuations each of the ADS SRV against 70% of maximum drywell pressure 57.6 scf 72 scf (Provided by one accumulator for each ADS SRV - 4 valves for division 1 and 5 valves for division 2) II CONTINUOUS SUPPLY A Nonsafety functions Accumulator recharging (< 40 min) 7.5 scfm 7.5 scfm B Safety functions
- ADS SRV leakage (4 valves for division 4 scfh 5 scfh h 1 and 5 valves for Division 2)
SGTS valve leakage (2 valves / division) 2 scfh 2 scfh Containment & drywell vacuum relief 2 scfh 2 scfh valve leakage (2 valves / division) Isolation valve leakage (2 valves / 2 scfh 2 scfh division) III INTERMITTENT SAFETY
- One actuation of ADS SRV (4 valves for 28.8 scf 36.0 scf division 1 and 5 valves for division 2)
Three actuations of SGTS valves (2 valves / 39.6 scf 39.6 scf division) Two actuations of containment and 26.4 scf 26.4 scf drywell relief valves (2 valves / division) -
- Safety functions shall be provided for a period of seven days after LOCA or seismic event.
O 6.8-10
GESSAR II < 238 NUCLEAR ISLAND 22A7007 Rev. 14
'() 7.2.2.2.C.l.j conformance of Regulatory Codes, Guides, and Standards (Continued) certain aspects of the RPS trip channels during periodic test and calibration at shutdown only. During tests of the trip chan-
- nels, proper operation of the mode switch contacts may be easily verified by noting that certain sensors are connected to the RPS logic and that other sensors are bypassed in the RPS logic in an i
appropriate manner dependent on the position of the mode switch. i l
- In the startup and run modes of plant operation, procedures may be used to confirm that scram. discharge volume high water level trip channels cannot be bypassed as a result of the manual bypass switches. In the shutdown and refuel
- modes of plant operation, a similar procedure may be used to l bypass all four scram discharge volume trip channels. Due to the discrete ON/OFF nature of the bypass function, calibration
() is not meaningful. A manual scram switch permits each individual instrument channel and trip logic to be tested on a periodic basis. Testing of each process sensor of the protection system also affords an opportunity to verify proper operation of these components. The time response of the instrument channel and trip logic may be measured by inserting a step function current with a switch provided on the calibrator card. For complete description of the testability of the RPS, see Section 7.2.1.1.D.8. The Applicant's Technical Specifications shall require independent functional testing of the back-up scram valves each refueling outage. In addition, operation of at least one back-up scram valve shall be confirmed following each scram
~
occurrence.
.O 'U i
7.2-83
GESSAR II 22A7007 238 NUCLEAR ISLAND Ro:v. 15 7.2.2.2.C.1.j Conformance of Regulatory Codes, Guides, and Standards (Continued) Independent testing of each back-up scram valve during refueling outages will be performed using one of the following methods: e (Preferred Method) The power supply to one of the back-up scram valves shall be interrupted , by pulling and tagging its fuse. Then, with _ only the other valve functional, the scram buttons shall be depressed and the valve's opera-tion confirmed by observing the subsequent " Low Scram Air Header Pressure" alarm. In addition, an observer should confirm the pressure drop indication on the local scram valve air header pressure gauge. The scram logic can then be reset, the fuse reinstalled, and the tag and annunciator cleared. The procedure should be repeated for the remaining back-up scram valve. e (Alternate Method) An observer shall stand between the back-up scram valves while the operator depresses the manual scram buttons. The operator then confirms the " Low Scram Air Header Pressure" alarm while the observer con-firms air discharges from the exhaust ports of each valve. In addition, the observer should confirm the pressure drop indication on the local scram valve air header pressure gauge. When air discharges have been confirmed on both valves, the test is complete and the scram logic and annunciator can be reset. O 7.2-83a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 , a l 7.4.1.3 Reactor Shutdown Cooling Mode (RHRS) - Instrumentation and Controls (Continued) tested during normal plant operation from the remote manual switches in the main control room. . The logic is tested by automatic pulse testing. The Automatic Pulse Test, APT, the sixth test, discussed in Subsec-tion 7.1.2.1.6 is also applicable here for the Reactor Shutdown Cooling mode function of RHR. l 1 L.- Environmental Considerations The only reactor shutdown cooling control component located inside the drywell that must remain functional in the environment is the control mechanism for the inboard isolation shutdown cooling suction valve. The environmental capabilities l( ) of this valve are discussed in Subsection 7.3.1.2. The control and instrumentation equipment located outside the drywell is selected in consideration of the normal and accident environments , in which it must operate. RHR equipment is seismically qualified and environ-mentally classified as discussed in Sections 3.2, 3.10 and 3.11. t i M. Operational Considerations All controls for reactor shutdown cooling are located , in the main control room. Reactor operator information is pro- ! vided as described in the RHR discussion of the LPCI mode (Subsection 7.3.1.2). N. Reactor Operator Information I Refer to Subsection 7.3.1.1.2 for reactor operator information associated with RHR in general. 7.4-29
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 15 7.4.1.3 Reactor Shutdown Cooling Mode (RHRS) - Instrumentation and Controls (Continued) O. Setpoints There are no setpoints involved in the operation of the shutdown cooling mode of the RHR except that reactor pres-sure and water level setpoints must be satisfied before the oper-ator can begin this mode. 7.4.1.4 Remote Shutdown System
'(NOTE: The RSS will be upgraded from the present single-panel system to a redundant safety-grade system. See Subsec- a tion 19.3.7.60. This subsection will then be revised to e be consistent with the design change, and before its reference by the first Applicant.) _
7.4.1.4.1 General The remote shutdown system provides a means to carry out the reactor shutdown functions from outside the main control room and bring the reactor to cold conditions in a safe and orderly fashion. 7.4.1.4.2 Postulated Conditions Assumed to Exist as the Main Control Room Becomes Inaccessible A. The plant is operating initially at or less than design power. B. The plant is not experiencing any transient situations. Even though the loss of offsite ac power is considered unlikely, the remote shutdown panel or facilities are powered from a Class lE power system bus so that backup ac power would be automati-cally supplied by the plant diesel generator. Manual controls of the diesel generator are also available locally. C. The plant is not experiencing any accident situations. No design basis accident (including a LOCA) shall be assumed, so that complete control of engineered safeguard feature systems from outside the main control room shall not be required. 7.4-30
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ( U 7.4.2.3 Reactor Shutdown Cooling Mode (RHR) - Instrumentation and Controls (Continued) Appendix 15A examines the protective sequences relative to this event and equipment. Chapter 15 considers the operation and the system-level qualitative aspects of this system. Loss of plant instrument air or cooling water will not, by itself, prevent reactor shutdown capability. B. Specific Regulatory Requirements Conformance
- 1. Conformance to NRC Regulatory Guides The regulatory guides as applied for ECCS are also applicable to RHR shutdown cooling; therefore, see Subsec-() tion 7.3.2.1.2.A for conformance.
- 2. Conformance to NRC Regulations - 10CFR50 Appendix A Requirements (a) Criteria 19 through 24 Conformance to these criteria is apparent throughout Subsection 7.3.1.1.2. This system is actually an operating mode of the Residual Heat Removal System.
(b) General Design Criterion 34 - Residual Heat Removal The Reactor Shutdown Cooling System removes residual heat from the reactor when it is shut down and the main steamlines are isolated to maintain the fuel and reactor coolant pressure boundary within design limits. Onsite and offsite
) power are provided in the event that either source is not avail-able when shutdown cooling is needed. Subsection 3.1.2.4.5 7.4-65
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 7.4.2.3 Reactor Shutdown Cooling Mode (RHR) - Instrumentation and Controls (Continued) provides a discussion of the RHR System compliance with General Design Criteria 34. Subsection 5.2.5 provides a discussion of the Leak Detection System and its application to the RHR System. Subsection 15.2.9 discusses a backup method for disposing of residual heat should the normal shutdown line become unavailable during shutdown.
- 3. Conformance to Industry Codes and Standards The IEEE standards as applied for ECCS are also applicable to RHR shutdown cooling; therefore see Subsec-tion 7.3.2.1.2.C for conformance.
7.4.2.4 Remote Shutdown System - Instrumentation and Controls (NOTE: The RSS will be upgraded from the present single-panel system to a redundant safety-grade system. See Sub- o section 19.3.7.60. This subsection will then be revised 9 to be consistent with the design change, and before its b reference by the first Applicant.) _ 7.4.2.4.1 General Functional Requirements Conformance For remote shutdown operation, no abnormal condition is assumed. The remote shutdown capability, by itself, does not perform any safety-related or protective function. This system interfaces with safety-related systems such as RHR and RCIC and meets the design criteria for those systems. No additional design criteria for the remote shutdown capability are necessary since they are already addressed in the respective design requirements sections. l Appendix 15A examines the protective sequences relative to this event and equipment. Chapter 15 considers the operation and the system-level qualitative aspects of this plant capability. f l 7.4-66
1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
- O 8.1. 2.1 Nuclear Island Power System (Continued) j capable of being fed by either the normal preferred, alternate preferred or standby power supplies. Non-Class 1E loads, not supplied by Class lE switchgear, are supplied from unit substations connected to normal feeders through BOP interfaces which are sepa-
, rate from Class lE bus leeders. There is no interconnection J between the non-Class lE and the Class lE substations.
- The 120/240 VAC plant power system provides for non-Class lE plant control devices necessary for plant operation. The 120 VAC regu-lated non-Class lE instrumentation power system provides power for non-class lE control and instrumentation loads.
4 The Class lE 120 VAC regulated instrument power system provides power for Class lE plant controls and instrumentation. The system is separated into Divisions 1, 2 and 3 with separate regulating transformers and distribution panels fed from their respective divisional sources. , The 125 VDC power distribution system provides four independent and redundant onsite sources of power for operation of Class lE DC loads. The Division 3 125 VDC power system provides power for the l various Division 2 (HPCS) system controls and other Division 3 DC loads, including the Division 3 inverter for the Nuclear Systems Protection System - (NSPS) power supply. A separate non-Class lE
-125 VDC power system is provided for nonsafety-related control.
Lighting is supplied from 277/480 VAC, 120/240 VAC systems and 125 VDC systems. Lighting is further described in Subsection 9.5.3. The NSPS power is provided by four uninterruptible 120 VAC buses. The four buses provide the control redundancy for various instru-( ) mentation, logic and trip circuits and solenoid valves. The NSPS power supply is further described in Subsection 8.1.3.2.1.2. 8.1-3
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 15 8.1.2.1.1 Nuclear Island Safety Loads Safety loads include instrumentation, control and motive power for all systems required for safety, as listed in Subsections 7.1.1.2, 7.1.1.3, 7.1.1.4, 7.1.1.5 and 7.1.1.6 and repeated below for con-venient reference. These loads utilize various Class lE AC and/or DC sources for motive or control power or both. Combinations of power sources may be involved in performing a single safety function. For example, low voltage DC power in the control logic may provide an actuttion signal to a 125 VDC load driver which controls a 6.9 kV air circuit breaker to drive a large AC-powered pump motor. Con-currently, the same low voltage DC logic signal may be used to change the state of an AC load driver which controls a 120 VAC con-tactor coil circuit for actuation of a 460 VAC motor-operated injection valve motor. A. Nuclear System Protection System Power Supply B. Core and Containment Cooling Systems (1) Residual Heat Removal System (RHR) (2) Low Pressure Core Spray System (LPCS) (3) High Pressure Core Spray System (HPCS) (4) Leak Detection System (LDS) i (5) Automatic Depressurization System (ADS) C. Containment and Reactor Vessel Isolation and Leakage Control i Systems I (1) Nuclear Steam Supply Shutoff System (2) Containment Isolation Control System l (3) Leakage Control Systems (a) Air portions (b) Water portions l 8.1-4
GESSA2 II 22A7007 238 NUCLEAR ISLAND Rsv. 0 7 t 8.1.2.1.2.2 HPCS Power System Loads The HPCS power system loads consist of the HPCS pump / motor and associated 480 VAC auxiliaries such as motor-operated valves, engine cooling water pumps, engine auxiliary loads and other miscellaneous loads. Table 8.3-3 shows the Division 3 loads required during normal operation, normal shutdown, forced shut-down and LOCA. 8.1.2.2 Balance-of-Plant Power System (Provided by Applicant) 8.1.3 Design Bases 8.1.3.1 Safety Design Bases - Offsite Power O -I (Provided by Applicant ) 8.1.3.2 Safety Design Bases - Onsite Power 8.1.3.2.1 Nuclear Island-General Functional Requirements 8.1.3.2.1.1 Onsite Power Systems - General Each unit's total safety-related load shall be divided into three divisional load groups. Each load group shall be fed by an independent 6.9-kV Class lE bus, and each load group shall have access to two offsite and one onsite power source. Two normally energized power feeders each shall be provided for the Division 1 and Division 2 Class lE systems. The normal pre-ferred feeder is used when operational, with the alternate pre-ferred feeder used when the former is not available. One normally O energized normal preferred feeder which can be supplied from either 8.1-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.1.3.2.1.1 Onsite Power Systems - General (Continued) of two offsite sources shall be provided for the Division 3 (HPCS) system. Additionally, one independent redundant Class lE 6.9-kV diesel-generator and one independent redundant Class 1E 125 VDC system are provided for each divisional load group. A separate 125 VDC system is provided for Division 4 instrumentation and control. The redundant Class lE electrical load groups (Divisions 1 and 2) _ are provided with separate onsite standby AC power supplies, _ electric buses, distribution cables, controls, relays and other , electrical devices. Redundant parts of the system are _ physically separated to the extent that a single credible event, including a single electrical failure, cannot cause loss of power to redundant load groups. Independent raceway systems shall be provided to meet load group cable separation requirements for Divisions 1 and 2, as well as for the HPCS System (Division 3) and NSPS (Divisions 1, 2, 3 and 4). Division 1 and 2 standby AC power supplies shall have sufficient capacity to provide power to all the required Division 1 or 2 loads. Loss of the preferred power supplies, as detected by 6.9-kV Class lE bus under-voltage relay, shall cause the standby power supplies to start and connect automatically, in sufficient time to maintain the reactor in a safe condition, safely shut down the reactor or limit the consequences of a DBA to acceptable limits. The standby power supply shall be capable of being started _ and stopped manually and are not to be stopped automatically during _ emergency operation unless required to preserve integrity. Auto-matic start will also occur on receipt of a LOCA signal. Switchgear circuit breaker control circuits for the GESSAR II design utilize cell switches to assure that control logic is not % altered when the circuit breakers are racked into the test posi- a' tion or removed for maintenance. _ 8.1-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.1.3.2.1.1 Onsite Power Systems - General (Continued) I- structures. Detailed descriptions of equipment seismic design and capabilities are contained in Section 3.10. 8.1.3.2.1.2 NSPS Power Supply System Design Bases In order to provide redundant, reliable power of acceptable quality and availability _to support the NSPS logic and control functions during normal, upset and accident conditions, the following design bases shall apply:
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(1) NSPS power has four separate and independent Class lE inverter power supplies fed from separate Class lE
-batteries.
I (2) Each independent Class lE inverter power supply is V) backed by an alternate Class lE power source. Division 4 shall have access to Division 2 AC power as its alternate source. Provision. shall be made for automatic switching to the alternate source in case of a failure of the inverter power supply. The inverter power supply shall be synchronized in both frequency and phase with the l alternate power supply, so that unacceptable voltage spikes will be avoided in case of an automatic transfer from normal to alternate supply. (3) The Class lE inverter power supply shall supply 120 VAC 2% and 60 1 Hz single-phase power for both a 100% load change and loss of alternate AC input power. (4) The NSPS power supply shall also provide nondivisional power supplied by separate nonessential power supplies. 8.1-9
GESSAR II 22A7007 238 NUCLEAR ISLAND R:v. 0 8.1.3.2.1.2 NSPS Power Supply System Design Bases (Continued) Provision shall be made for automatic switching from the preferred to the alternate source on loss of the preferred source. 8.1.3.2.1.3 High Pressure Core Spray (HPCS) Power Supplies Design Bases The HPCS power system shall supply adequate power for HPCS loads which consist of a 6.9-kV HPCS pump / motor and associated 480 VAC auxiliaries, such as motor-operated valves, engine cooling water pump and miscellaneous engine auxiliary loads. Figure 8.3-14 shows the basic one-line diagram of the system. The HPCS power system shall be self-contained except for access to the preferred power supply, and except for the initiation signal source. It shall be an operable, isolated system independent of h electrical connection to any other system by use of the HPCS diesel generator. Auxiliary equipment such as heaters, air compressors and battery chargers shall be supplied from the same power source as the HPCS motor and be compatible with the plant AC power system. The HPCS diesel generator shall have the capability to restore power quickly to the HPCS pump motors in the event the preferred power supply is unavailable and to provide all power for startup and operation of the HPCS pump motor for safe shutdown of the plant. The HPCS diesel generator shall start automatically on l LOCA signal or 6.9-kV Division 3 bus under-voltage. When the plant preferred AC power supply is not available, the diesel l generator shall be connected to the HPCS bus automatically. The HPCS electric system shall be capable of performing its func-tion when subjected to the effects of a design bases event. In particular, it shall be a Class 1E system and be housed in a Seismic Category I structure. 8.1-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 8.1.3.2.2.2 NRC Regulatory Guides (Continued) (8) Regulatory Guide 1.41 - Preoperational Testing of Redundant Onsite Electric Power Systems to Verify Proper Load Group Assignments (9) Regulatory Guide 1.47 - Bypass and Inoperable Status Indication for Nuclear Power Plant Safety Systems (10) Regulatory Guide 1.53 - Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems (11) Regulatory Guide 1.62 - Manual Initiation of Protective Actions
- (' (12) Regulatory Guide 1.63 - Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants (13) Regulatory Guide 1.73 - Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants (14) Regulatory Guide 1.75 - Physical Independence of Electric Systems (see 8.3.1.4)
(15) Regulatory Guide 1.89 - Qualification of Class lE Equipment for Nuclear Power Plants 8.1.3.2.2.3 IEEE Standards l (1) IEEE Std 279-1971 - IEEE Standard Criteria for Protection () Systems for Nuclear Power Generating Stations l 8.1-13 l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.1.3.2.2.3 IEEE Standards (Continued) (2) IEEE Std 308-1975 - IEEE Standard Criteria for Class lE Electrical Systems for Nuclear Power Generating Stations (3) IEEE Std 338-1971 - IEEE Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Class lE Power and Protection Systems (4) IEEE Std 344-1975 - IEEE Recommended Practices for Seismic Qualification of Class lE Equipment for Nuclear Power Generating Stations (5) IEEE Std 387-1972 - Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Stations O l l I l l 1 l l l l I 8.1-14 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.2 OFFSITE POWER SYSTEMS (Applicant to provide) 8.2.1. Description (Applicant to provide) 8.2.2 Analysis (Applicant to provide) 8.2.3 Nuclear Island - BOP Interface The Applicant shall provide the offsite power system with suffi-cient capability to provide the Nuclear Island as well as the Applicant's BOP loads. The Nuclear Island standard plant power requirements, as well as the Applicant's BOP power requirements, cannot be defined exactly due to vendor differences and site-unique '( characteristics. Therefore, interface control documentation is provided to coordinate the power requirements with the Applicant for a specific plant. Quantitative interface load requirements i provided in this section shall be verified by the Applicant. 8.2.3.1 Design Criteria ! Division 1 and 2 Class lE buses shall each be provided with two immediate access feeder circuits from the transmission network. The two power feeders (referred to as normal preferred and alter-nate preferred) shall be from independent transmission sources. l The Division 3 (HPCS) Class lE bus shall have at least one feeder
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l which is fed from a bus in the Turbine Island. It is a requirement that the Turbine Island bus receive power from two independent offsite ; power sources and that the transfer between the power sources be auto-
-matic. The transient associated with the Turbine Island bus transfer will cause the HPCS bus to be transferred to the diesel-generator.
( ) Transfer back to the Turbine Island bus may be accomplished manually by the operator. This arrangement ensures the availability of the offsite power to the Turbine Island bus and in turn to the HPCS bus within a few minutes following a loss of the normal offsite power supply. , 8.2-1
GESSAR II -22A7007 238 NUCLEAR ISLAND R v. 15 8.2.3.1. Design Criteria (Continued) h Non-Class lE load groups shall be supplied with individual feeders from the auxiliary power system. All feeders shall be 6.9 kV. Both normal preferred and alternate preferred AC power sources to each division of Class lE buses shall be capable of providing power to the Class lE loads on that division in addition to non-Class lE loads. 8.2.3.2 Specific Interfaces Specific Nuclear Island / BOP power system interfaces are given in Table 8.2-1. As discussed before, only qualitative information is provided. See Figures 1.9-Sa and 1.9-5b (E-040A & B) for the physical location of the raceway interfaces which are identified as R-XX. Cable interfaces on Tables 8.2-1 and 8.3-10 are identified as E-XX. O 8.2-2
'N
,J Table 8.2-1 NUCLEAR ISLAND AC POWER SYSTEM / BOP INTERFACES 1
} Maxtmum Allowable Actual Actual Maximum Allowable Interface Steady Inrush SS Load SC Mvh Number
- Interface Description State lead k,VA, J2 5 _ Q) . _ Momentan INT
) E-1 (A-C) Normal Feeder from eOP to AB 6900-400V XFMR 2550 kVA(1) 3,400 1775 kVA 433 409 F-2 (A-P) Normal Feeder from BOP to Bus C 7935 HP(4) 51,574 7935 HP 4'S 450 ' 1780 kVA E-3 (A-C) Normal Feeder f rom %P to AB 6900-480V KrMR 2550 kVAll) 3,400 426 402 a E-4 (A-F) Normal Feeder from BOP to Bus D 7935 HP(4) 51,578 7935 HP 490 461 t E-5 (A-C) Normal Preferred or Alternate Preferrud Feeder 3560 kVA 17,444 83 490 461 l from BOP to HPCS Bus C SWGR E-13 (A-F) Normal Preferred Feeder f rom BOP to Bus E SWCR 8750 kVA(5) 10.423 5907 433 409 E-14 (A-F) Alternate Preferred Feeder from BOP to Bus E SWGR 8750 kVA(St 10.423 5907 433 409 N ! (4 E-15 (4-F) Normal Preferred Feeder from BOP to Bus F SWCR 8750 kvA(5) 10.423 4443 426 402 g i E-16 (A-F) Alternate Preferred Feeder from BOP to Bus F SWGR 8750 kVA(5) 10.423 4(0! 426 402 g ! E-41 (A-C) Normal Feeder from BOP to RW 6900-480V KFMR 1275 kVA(1) 1,700 684.1 433 409 CO g OM e E-43 (A-C) Normal Feeder from BOP to PW 6900-480V 'KFMR 1275 kVA(1) 1,700 250.5 426 402 UM y Mm g E-58 (A-B) 250 VDC Feeder to Inverter and SS "E* $0 kVA --- (4 HH iSee Figure 8.3-1 for interf aces identified in accordance with these numbers. These interf aces are nondivisional. Voltage variation = tSt at the 6900-volt level . Frequency variation = 60 Haf2
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U a i j 1 1 M N A A> t . 4 i O WO Ut 4
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 15 Table 8.2-1 NUCLEAR ISLAND AC POWER SYSTEM / BOP INTERFACES NOTES: (1) Value represents 85% of the transformer "AA" rating. Each transformer has an additional 30% capacity (continuous) under forced air cooling conditions. (2) For unit subtransformer loads, actual inrush kVA is con-sidered as 85% of the transformer "FA" rating. For pumps, inrush kVA = 6.5 x rated horsepower. (3) Values taken from load summary calculations under normal operating conditions (Mode 2). Exceptions: Interfaces E-13, 14, 15 and 16 LPCS and RHR loads are included. (4) Pump nameplate rating. (5) Diesel generator nameplate rating. O O 8.2-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 f~ V 8.3 ONSITE POWER SYSTEMS 8.3.1 AC Power Systems 8.3.1.1 Description The onsite power system is also referred to as the standby power system. The normal preferred supply of AC electric power from the turbine generator, switchyard and the alternate preferred supply from the substation to Class lE distribution systems is furnished by the Applicant. The auxiliary electric power system includes three independent Class lE AC electric power systems for nuclear safety-related loads. The principal elements of the auxiliary AC electric power systems are shown in Figure 8.3-1.
/O O AC power is supplied and utilized at 6.9 kV for motor loads larger than 450 hp and transformed to 480V for smaller loads. The 480V system is further transformed into lower voltages as required for instruments, lighting, 120V motors and controls.
i Class lE AC power loads are divided into three divisions (Divi-
.sions 1, 2 and 3), each fed from an independent 6.9-kV Class lE bus. The loads are listed in Tables 8.3-1, 8.3-2 and 8.3-3. Dur-ing normal operation, Division 1, Division 2 and Division 3 loads are fed from an offsite normal preferred power supply. Division 1 and 2 buses are automatically transferred to an offsite alter-nate preferred power supply when the normal preferred power supply to these buses is lost'. Standby AC power for Class lE buses is _
l supplied by diesel-generators at 6.9 kV and distributed by the Class lE power distribution system after both preferred power sup-s plies to that division are lost. The transfer is discussed in Sub-( ,) section 8.3.1.1.7. Division 3 (HPCS) Class 1E load is supplied by a single feeder from a BOP auxiliary bus which may be served by the_ 8.3-1
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 0 8.3.1.1 Description (Continued) normal preferred or an alternate preferred power source. Each Class lE division has a dedicated diesel generator, which auto-matically starts in case of a LOCA and/or loss of voltage on the division's 6.9 kV bus. Each 6.9-kV Class lE bus feeds its asso-ciated 480V unit substation through a 6.9-kV/480/277V load center transformer. There is no automatic or manual interconnection between the 6.9-kV divisional systems, nor is there any automatic interconnection for the 480V divisional systems. A manual interconnection, however, is provided between redundant Division 1 and Division 2 Class lE 480V buses. The interconnection is guarded by normally open manual breaker at each end of the bus tie and also guarded by key inter-locks which prevent paralleling of the two redundant sources when interconnection is used. Class lE standby distribution systems are not shared netween units of a multiunit plant. 8.3.1.1.1 6.9-kV Distribution System The voltage level of input power to the Nuclear Island is 6.9 kV. The circuits supplying power are shown in Figure 8.3-1. Details l l of the circuit protection devices and relaying for Nuclear Island l switchgear are shown in Figures 8.3-2 and 8.3-3. 1 The 6.9-kV system to the Nuclear Island, which includes normal and alternate preferred power supply feeders, is furnished by the Applicant. Physical separation of the feeders is by the Applicant. O 8.3-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
) 8.3.1.1.2 480V Distribution System Power for 480V auxiliaries is supplied from load centers consisting of 6.9-kV/480V transformers and associated metal clad switchgear.
Class lE 480V load centers supplying Class lE loads are arranged as independent radial systems, with each 480V bus fed by its own power transformer. Each 480V Class lE bus in a division is phys-ically and electrically independent of the other 480V buses in other divisions. A manual crosstie is provided between redundant - buses (Figure 8.3-15 and 8.3-16) and is equipped with a normally open circuit breaker in each substation. The ties are manually initiated and are guarded by key interlock to prevent paralleling of the two divisions. Under normal operation, Division 1 breaker "110A" is closed (Bus 3 El is fed from Bus E), Division 2 breaker "210A" is closed (Bus F1 (_) is fed from Bus F), and the two tie breakers between Divisions 1 and 2 are open (breaker 110F for Division 1 and breaker 210F for Division 2). See Figure 8.3-15. 3 If during plant shutdown, the operator needs to close the tie breakers for maintenance flexibility, the following sequence has to take place. 9
- Trip breaker "110A"/ bus El, and lock it open. m
[ -Remove the key from lock (A4) at breaker 110A/ Bus El. Key is removable only when breaker is locked open.
- Insert key in lock (A4) at breaker 210F (Bus F1).
! - With key (A3) in its respective lock, breaker 210F/ Bus F1 may be closed and a Control Rod indicating light will indicate that breaker 210F is closed.
- Remove key from lock (A5) at breaker 110A/ Bus El.
l -'s - Insert key in lock (A5) at breaker 110F/ Bus El. . d 8.3-3
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.3.1.1.2 480V Distribution System (Continued) <
- With key (A2) in its respective lock, breaker 110F/ Bus El ~
may be closed, and a CR indicating light will indicate that breaker 110F is closed. Main breaker 210A/ Bus F1 is now feeding Buses El and F1 while main breaker 110A/ Bus El is locked open. Similar steps could be taken in order to feed buses El and F1 from main breaker 110A while breaker 210A is locked open. When breaker 110A/ Bus El or breaker 210A/ Bus F1 is open, an indi-cating light will be initiated in the Control Room. The same applies for the tie breakers between nondivisional Buses 9 E2 and F2. Interlocks A2 and A3 are provided to safeguard against personnel coming into contact with live bus in the rear of the cubicle.
]
The operator has to:
- Trip and lock out tie breakers 110F/ Bus El and 210F/ Bus Fl.
Then remove keys A2 and A3.
- Both keys must be inserted into their respective locks on rear door of the cubicle at breaker 110F/ Bus El (or breaker 210F/ Bus Fl) in order to open the rear door to work in the Bus compartment.
The 480V unit substation breakers supply 460Vmotor loads up to and including 400 hp, and motor control centers. Switchgear for the l 480V load centers is of indoor, metal-enclosed type with drawout circuit breakers. Control power is from the Class lE 125VDC power system of the same division. The HPCS 480V auxiliaries are sup-plied from an independent Class lE 6.9-kV bus and transformer in Division 3. 8.3-3a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 (D y ,/ 8.3.1.1.2 480V Distribution System (Continued) The 480 MCCS feed motors 100 hp and smaller, control power trans-formers, process heaters, motor-operated valves and other small e19ctrically operated auxiliaries, including 480-120V and 480-240V transformers. Class lE control centers are isolated in separate load groups corresponding to divisions established by the 480V
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i unit substations. J MCC branch circuit protection for all loads is provided by molded case circuit breakers. Starters for the control of 460V motors 100 hp and less are the MCC-mounted, across-the-line magnetically operated, air break type. The starters are a combination type with circuit breakers of 25,000A, or higher symmetrical interrupting capacity and a mag-netic contractor to provide overload and undervoltage protection. 73
'_,)
( Class 1E MOVs have molded case breakers with thermal magnetic pro-tection since the overload elements of the starter are in the circuit during testing although bypassed during normal plant ope-ration. Circuits leading from the electrical penetration assem-blies into the containment area have a fuse in series with the circuit breakers as a backup protection for a fault current in the penetration in the event of circuit breaker overcurrent or fault protection failure. l i (~h [ N. 8.3-3b
GESSAR II 22A7007 238 NUCLEAR ISLAND R5v. 0 8.3.1.1.2 480V Distribution System (Continued) 4 Starters for the control of 460V motors 100 hp and less are the MCC-mounted, across-the-line magnetically operated, air break type. The starters are a combination type with circuit breakers of 25,000A, symmetrical interrupting capacity and a magnetic contactor to provide overload and undervoltage protection. Class lE MOVs have molded case breakers with thermal magnetic protection, since the overload elements of the starter are in the circuit during testing although bypassed during normal plant operation. Circuits leading from the electrical penetration assemblies into the con-tainment area have a fuse in series with the circuit breakers as a backup protection for a fault current in the penetration in the event of circuit breaker overcurrent or fault protection failure. 8.3.1.1.3 120/240V Distribution System Individual transformers and distribution panels located in the vicinity of the loads supply 120/240V power. This power is used for lighting, 120V receptacles and other 120V loads. 8.3.1.1.4 Instrument Power Supply Systems 8.3.1.1.4.1 120V Safety-Related Instrument Power System Individual regulating trancformers supply 120V instrument power. Each Class lE divisional regulating transformer is supplied from a 480 MCC in the same division. Power is distributed to the individ-ual loads from distribution panels, and to logic level circuits through the Control Room Logic Panels. O 8.3-4
- = = . = -
.n. -. . .. . - _- -- - . . _ . . - . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0
*4
.p
,, 843.1.l'.5.1' Physical Separation and Independence (Continued) m _
- i. __.
, [between divisions. . Electric equipment and wiring for the Class lE
~
systems which are segregated into separate divisions are separated i 'so.that no.d'esign basis event is capable of disabling any ESF
- function.
t j The physical layout of major itiems of Class lE electric equipment in the Nuclear Island is shown in Figure 1.2-4. The methods used for' separating equipment and facilitating the separate routing of
- main raceway systems and feeders are illustrated in Figures 8.3-4 through 8.3-12 for selected areas, as described below.
(1) Figures l8.3-4 and 8.3-5 show the RHR, HPCS, LPCS and RCIC l pumps located-in separate rooms, so that the divisional separationf(i.e. ,' physical independence) is maintained.
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, ~
l (2). Figures 843-8 and 8.3-9' illustrate the layout of switch-
~
gear and MCCs on the electrical floor of the Auxiliary cBuilding,3 The safety-related divisional AC switchgear, 4 load centiers',1 MCCs tattery rooms and DC distributional panels and MCCs are located-to provide separation and electrical isolation amongJthe divisions. Separation is jp provided amongs divisional cables being routed between the ( equipment rooms, the Main Control Room, containment and other processing areas. Equipment in these areas is j l divided into Divisions 1, 2,'3 and 4, sufficiently separated by space and the barrier formed by the steam [ - tunnel walls. The equipment'is located to facilitate divisional separation of cable trays and.to provide accessLto electrical penetration assemblies. m '
;s e /
Diesel Ceherator 1 is in a separate building. Figure 8.3-13
.)'
-illustrbtes the separation between Diesel Generator 2 D.iesel G$ncrator 3 and its associated equipment. -
/
y i -t
- g 8.,3-9 1
, _ - - - - _- . . - , _ . . m_.
l GESSAR II 22A7007 238 NUCLEAR ISLAND Riv. 15 8.3.1.1.5.1 Physical Separation and Independence (Continued) (3) Figure 8.3-12 illustrates the method used to place electrical penetration assemblies in groups at widely spaced points around the containment in order to separate the circuits of the Class lE divisions. The penetration assemblies are located around the periphery of the con-tainment and at different elevations to facilitate rea-sonably direct routing to and from the equipment. No penetration carries cables of more than one division, except for special identified, associated cables (' i . e . , any cable sharing a penetration with a divisional cable becomes identified as divisional associated and continues to run with that div:sion). The electric separation criteria ~are further discussed in subsection 8.3.1.4. (4) Divisional cables to and from the containment and to and from the dedicated divisional equipment in the Auxiliary and Fuel Buildings are routed in separated cable raceways for each division. Routthg is maintained up to the terminal cabinets in the Main Control Room. Wiring for all Class lE equipment indicating lights is an integral part of the Class lE cables used for control ~ of the same equipment and are considered to be Class 1E circuits. (5) Instrumentation control and power wiring circuits are separated'into voltage level groups as shown below. The e' following designated groups are run with their designated divisions but are separate from each other: V1 =-Instrumentation low-level signal and control g (0-55 VDC) analog; 0-12 VDC digital 8.3-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 8.3.1.1.8.1.2 Ratings and Capability (Continued) The size of each of the diesel-generators serving Divisions 1 and 2 satisfies _the requirements of NRC Regulatory Guide 1.9 (March 1971) and' conforms to the following criteria: (1) Each diesel generator is capable of starting, accelerat-ing and supplying its loads in the sequence shown in Tables 8.3-1 and 8.3-2 without exceeding a 5% frequency drop. The generators are capable of recovery to 98% of normal frequency in 2 sec or less. (2) Each diesel generator is capable of starting, accelerat-ing and supplying its loads in their proper sequence without exceeding a 20% voltage drop at its terminals. ]g The generators are capable of recovery to 90% of normal voltage in less than 2 sec. l
\
(3) Each diesel generator is capable of starting, accelerat-ing and running its largest motor at any time after the automatic loading sequence is completed, assuming that the motor had failed to start initially. j (4) Each diesel generator is capable of reaching full speed and voltage within 10 sec after receiving a signal to l ! start, and can be fully loaded within 30 sec following the start signal. (5) The speed of each diesel generator does not exceed 75% of the difference between nominal speed and the over-speed trip setpoint, or 115% of nominal speed, whichever is lower, during recovery from transients caused by disconnection of the largest load. P) t
's - (6) Each diesel generator shall be capable of running for 7
s 7 days under no-load, full-speed conditions. _$ 8.3-21
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 15 8.3.1.1.8.1.3 Starting Circuits and Systems Diesel generators 1 and 2 start automatically on loss of bus voltage logic is one-out-of-two. Two low-water-level switches reactor or high drywell pressure. Undervoltage relays are used to start each diesel engine in the event of a drop in bus voltage below preset values for a predetermined period of time. The under-voltage logic is one-out-of-two. Two low-water-level switches
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and two drywell high-pressure switches in each division are used to initiate diesel start under accident conditions. One-out-of-two-taken-twice logic is used for generating the start signal. The transfer of the Class lE buses to standby power supply is automatic should this become necessary on loss of all preferred power. After the breakers connecting the buses to the preferred power supplies _ are open and the bus voltage is less than 30% of normal, the diesel generator breaker is closed when required generator voltage and frequency are established. Diesel generators 1 and 2 are designed to start and attain rated voltage and frequency within 10 sec. The generator, exciter and voltage regulator are designed to permit the set to accept the load and to accelerate the motors in the sequence and the time requirements listed in Table 8.3-4. The voltage drop caused by starting the large motors does not exceed the requirements set forth in Regulatory Guide 1.9, and proper acceleration of these motors is ensured. Control and timing circuits are provided, as appropriate, to ensure that each load is applied automatically at the correct time not to exceed that as shown in Tables 8.3-4 and 8.3-5. Each diesel-generator set is provided with two independent starting air systems. A detailed discussion of the~ diesel air start systems is presented in Subsection 9.5.6. O 8.3-22
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
" Automatic Shedding Loading and Isolation
. 8.3.1.1.8.1.4 The diesel generator is connected to its Class 1E bus only when the 4 incoming preferred source breakers have been tripped (Subsection
- 8. 3.1.1. 7) . Under this condition, major loads are tripped from the Class 1E bus, except for the Class lE 480V unit substation feeders, before closing the diesel generator breaker.
The large motor loads are later reapplied sequentially and auto-matically as demanded. to the bus after closing of the diesel-generator breaker. 8.3.1.1.8.1.5 Protection Systems The diesel generator is shut down and the generator breaker tripped under the following conditions during all modes of operation and testing operation: (_s (1) engine overspeed trip, and j (2) generator differential relay trip. The generator breaker is tripped under the following conditions during normal operations and testing: (1) generator ground overcurrent; (2) generator voltage restrained overcurrent; (3) bus underfrequency; (4) generator reverse power; and (5) generator loss of field. l In addition, during diesel-generator normal operations or testing, the diesel generator is shut down due to: (1) high jacket water temperature; (2) generator high bearing temperature; ( (3) generator loss of excitation: 8.3-23 i
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 0 8.3.1.1.8.1.5 Protection Systems (Continued) (4) reverse power; (5) low turbo oil pressure; (6) high vibration; (7) high lube oil temperature; (8) low lube oil pressure; (9) high crankcase pressure; and (10) low jacket water pressure. Protective functions (trips) of the engine or the generator breaker and other off-normal conditions are annunciated in the Main Con-trol Room and/or locally. They are: (1) low level - jacket water; (2) low pressure - jacket water; (3) low-low pressure - jacket water; (4) low temperature - jacket water in; . (5) high temperature - jacket water out; (6) high-high temperature - jacket water out; (7) low level - lube oil tank; (8) low temperature - lube oil in; (9) high temperature - lube oil out; (10) high-high temperature - lube oil; (11) high dif ferential pressure - lube oil filter; (12) low pressure - turbo oil right/lef t bank; (13) low-low pressure - turbo oil; (14) low pressure - lube oil; (15) low-low pressure - lube oil; (16) high temperature - bearings; (17) high pressure - crankcase; (18) excessive vibration; (19) overspeed; (20) fuel pump /overspeed f ailure; 8.3-24
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
) 8.3.1.1.8.1.5 Protection Systems (Continued)
(21) barring device engaged; (22) low pressure - starting air; (23) in maintenance mode; (24) unit fails to start; (25) underf requency ; (26) trip unit trouble; (27) out of service; (28) engine tripped; (29) lock out relay operated; (30) emergency start; (31) loss of control power; (32) low-high level - fuel day tank; (33) low pressure - fuel oil; (34)- high differential pressure - fuel filter; (35) low pressure - control air; (36) in local control only; (37) LOCA signal; (38) breaker position; and (39) bus voltage.
- 8.3.1.1.8.2 Local and Remote Control I
l Each diesel generator is cap'able of being started or stopped j manually from the Main Control Room. Start /stop control and bus transfer control may be transferred to a local control station in the diesel generator room by operating key switches at that station. l 8.3.1.1.8.3 Engine Mechanical Systems and Accessories l Descriptions of these systems and accessories are given in l Chapter 9 under the following listed subsections: i (\~ (1) Cooling Water (9.5.5); (2) Lubricating 011 (9.5.7); ( 0.3-25
l GESSAR II 22A7007 238 NUCLEAR ISLAND R@v. 15 8.3.1.1.8.3 Engine Mechanical Systems and Accessories (Continued) i (3) Fuel Oil (9.5.4); and (4) Starting Air (9.5.6). 8.3.1.1.8.4 Interlocks and Testability Each diesel generator, when operating other than in test mode, is totally independent of the preferred power supply (i.e., Interlocks prevent automatic closure of the preferred source breaker). Addi-tional interlocks to the LOCA and LOPP sensing circuits terminate _ parallel operation test and cause the diesel generator to auto-matically revert and reset to its standby mode if either signal appears during a test. A lockout or maintenance mode removes the diesel generator from service. The inoperable status is indicated in the control room. 8.3.1.1.8.5 Prototype Reliability Qualification Testing The qualification tests are performed on one Division 1 or 2 diesel generator per IEEE Std. 387 as modified by Regulatory Guide 1.9 requirements. Test results will be provided by the Applicant. e b I I i 8.3-26
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 14
-s
)
s/ 8.3.1.1.8.5 Prototype Reliability Qualification Testing (Continued) m 8.3.1.1.8.6 Reliability Qualification Testing (Applicant to provide) 8.3.1.1.9 High Pressure Core Spray (HPCS) System Power Supply The HPCS power system is fed by a preferred power supply feeder 7-ly_) at 6.9 kV and is not interconnected with any other electric supply. The system has a diesel generator for standby power. Figure 8.3-14 shows the HPCS power system simplified one-line diagram electrical arrangement, power distribution, protective relaying, and instrumentation. Figure 8.3-17 shows the functional control diagrams of the system. The HPCS power supply system is self-contained except for the initiation signal source and access to the preferred power through the plant AC power distribution system. It has a dedicated diesel-generator and is operable as an isolated system independent of electrical connection to any other system. The auxiliary equip-ment such as heaters, air compressor and battery charger are supplied from the same power source as the HPCS motor. Voltage and frequency of the HPCS diesel-generator is compatible with that
-~ available from the preferred power supply system. 'd 8.3-27
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.3.1.1.9 High Pressure Core Spray (HPCS) System Power Supply (Continued) The HPCS diesel generator has the capability to quickly restore power to the HPCS bus in the event preferred power is not available
~
and to provide all required power for the startup and operation of the HPCS system. The HPCS diesel-generator starts automatically upon a LOCA signal from the plant protection system or the HPCS 6.9 kV bus undervoltage, and will be automatically connected to the HPCS bus when the plant preferred AC power supply is not available. The failure of this unit will not negate the capability of other power sources. There is no provision for automatic paralleling of the HPCS diesel generator with the preferred power or with other s tandby power supplies. Provisions for manual paralleling with preferred power supplies is made for loading the diesel generator during the test mode or for manual bus transfer. An interlock is furnished to prevent inadvertently paralleling. There is no sharing of the HPCS power system with other unit (s) . 8.3.1.1.9.1 Equipment Identification The major HPCS power system equipment (e.g., diesel generator, switchgear, motor control center and transformers) are identified by a nameplate engraved " Nuclear Safety Related, Division 3" . Raceways are identified in accordance with Subsection 8.3.1.4. 8.3.1.1.9.2 HPCS Class lE - Electrical Equipment Capacity HPCS power system electrical apparatus is sized on the basis of the most severe conditions it will be subjected to, for either continuous or intermittent conditions in any mode of operation. Intermittent loads are factored in on the basis of heating (e.g., short time peaks are not added directly to determine total con-tinuous load imposed). Adverse environmental conditions have been 8.3-28
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 O 8.3.1.1.9.2 HPCS Class lE - Electrical Equipment Capacity (Continued) taken into consideration (e.g. , derating of cable for temperatures higher than the basic rated values and use of multipliers on actual service hours for motors operated at higher than normal rated temperatures). The diesel generator can operate without damage for four hours in a no-load . condition, with a subsequent loading of at least 50% of its rated load for a continuous period of approximately 15 min-utes. The diesel-generator is started. during a LOCA condition and kept in running standby (no load) mode when the of fsite power is available to assure diesel-generator availability. The diesel generator remains in an unloaded condition only in this mode. Administrative procedure is employed to load the diesel-generator by means of paralleling the diesel generator with offsite power ( periodically or continuously as necessary. In this way, the diesel-generator availability is fully assured during a total offsite power loss. The HPCS pump load is expected to be above 50% of the diesel-generator continuous rating, and this is suffi- ! cient to avoid degradation of the diesel-generator performance during a LOCA and ~under accident recovery ' conditions. The HPCS unit substation transformer is sized for the largest combination of continuous loads that it may be required to carry plus an allowance for intermittent loads (Table 8.3-3). The switchgear ratings established are consistent with bus loading and interrupting capacity requirements and are compatible with maximum available short-circuit current values at the point where feeders connect to Class lE switchgear. The HPCS Auxiliary motors are designed to start and accelerate
/~' their pump load with 75% voltage applied to the motor terminals. \
w 8.3-29
~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 0 8.3.1.1.9.2 HPCS Class lE - Electrical Equipment Capacity (Continued) The minimum difference between the motor torque and the pump torque at any given speed during acceleration is 10% of motor rated torque. The HVAC fan motors for HPCS are designed to start and accelerate their load with 80% voltage applied to the motor terminals. The HPCS pump motor is provided with thermocouples on bearings and in windings to verify that temperature rise is acceptable. The HPCS motor is initially tested in accordance with NEMA-MG-1. The HPCS diesel generator (Division 3) is capable of starting the HPCS pump motor within the required time although voltage and fre-quency drop will exceed the limits specified in NRC Regu]r. tory Guide 1.9 (Subsection 8.3.1.2.3.2.2). 8.3.1.1.9.3 120 VAC Instrument Power System The Division 3 essential instrument power system provides power to all HPCS diesel-generator system instrumentation. The instrument power system consists of a distribution panel fed through a regulating transformer connected to Division 3 Class lE load centers. There are no bus ties or interconnections between the HPCS distribution panels and other electrical divisions. All equipment associated with the Class lE instrument power system is readily accessible for inspection and testing. Service and testing will be done on a routine basis in accordance with the manufacturer's recommendation. O 8.3-30
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O
\ ,/ 8.3.1.1.9.4.3 HPCS Class lE - Electrical Equipment Circuit Protection (Continued) permit the use of conventional protective relaying practice for isolation of faults. There is no load shedding or load sequencing in the HPCS power system. Emphasis is given in preserving the ESF function in situations of power loss and equipment failure. Over-load relay for HPCS motor and diesel generator give alarm indica-tion only. Major faults are isolated by instantaneous relaying.
The HPCS diesel-generator protection is described in Subsec-tion 8.3.1.1.9.5.4. The unit substation transformer has inverse time and instantaneous protection relaying. The HPCS pump motor has ground indication and both instantaneous and inverse time overcurrent protection relaying. These relays give indications only in case of nominal overloads. The stall condition is moni-tored by a high dropout instantaneous relay in conjunction with the invers,e time overcurrent relay, and trips the motor in case of v incipient stall. In general, relay settings are coordinated in such a way that interference of service is not communicated to a
" higher" level or involve equipment other than that inmediately affected by the fault or overload. This is achieved by selecting trip levels and time delay settings so that faults are not passed through to circuit breakers ahead in a chain leading to the power supply but are relayed off without opening the power supply breaker and thus keeping other loads which share a bus on line. However, l a faulty trip device or circuit breaker trip is protected against ultimate damage by circuit breakers " ahead" of them through coordinated magnitude and time settings.
8.3.l.1.9.4.4 HPCS Class lE - Electrical Equipment Testing j Means are provided for periodically testing the chain of system l elements from sensing devices through driven equipment to assure . ( ty that Class lE equipment is functioning in accordance with design 8.3-33
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 8.3.1.1.9.4.4 HPCS Class lE - Electrical Equipment Testing (Continued) requirements. The drawout feature of protective relays allows substitution of temporarv relays installed while the regular relays are removed for bench tests and calibration. Startup of the standby power supplies can be effected by simulation of a Division 3 LOCA signal or loss of preferred power to the plant auxiliary power system. Connection of HPCS diesel-generator to the HPCS 6.9 kV bus takes place automatically upon loss of pre-ferred power.to the HPCS bus (HPCS 6.9 kV bus undervoltage).
- 8. 3 .1.1. 9 . 5 HPCS Diesel-Generator Set The HPCS diesel generator supplies power to the HPCS 6.9 kV bus in the absence of the preferred power sources. Figure 8.3-14 shows the interconnections between the preferred power system and the h HPCS diesel-generator set, and the HPCS pump and MCC for HPCS valves, and the other HPCS auxiliary loads. Table 8.3-3 shows steady-state loads and continuous and short-time rating of the diesel-generator set.
The generator is rated to have sufficient capacity to start and run the induction motor for the HPCS pump, a unit substation trans-former and several 460V induction motors (<160 hp continuous) which drive the engine cooling water pump and several valve operator motors. There is an additional load of 60 kVA for lighting trans-fo rme rs , instrument transformers and a battery charger. The valve - motors required power for only a short period of time and do not impose a significant load on the generator. The HPCS pump motor has a voltage rating of 6.6 kV. Table 8. 3-3 lists the loads on the Division 3 (HPCS) source. The diesel-generator set has the capacity to start and supply the loads required by the HPCS within the time requirements described in Section 6.3. h 8.3-34
' GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 ]( ) 8.3.1.2.1.2.2 Compliance with Regulatory Guide 1.9 (Continued) Recovery of voltage and frequency to within 10% and 2% of nominal, respectively, has been verified to be accomplished within 40% of
! the sequencing interval of 5 sec. Step loading and disconnection of the largest single load does not cause the diesel generator
~
! to exceed 110% of normal speed, thus precluding an inadvertent overspeed trip. The reliability of the diesels has been substantiated by an extensive test program. The tests verify the following diesel functions: (1) diesel fast start capabilities; (2) load carrying capabilities; () (3) load rejecting capabilities; - i (4) ability of the system to accept and carry the applied loads up to its rated capacity; and (5) long-term no load running of the diesel unit without any i detrimental effects. l The reliability of the system to start and accept loads in a prescribed time interval has been demonstrated by prototype quali- i fication test data and will be verified by preoperational tests. I 8.3.1.2.1.2.3 Compliance with Regulatory Guide 1.32 - Use of IEEE Std 308-1971, " Criteria for Class lE Electric Systems for Nuclear Power Generating Stations" Electric power from the transmission network to the switchyard is the design responsibility of the applicant. Compliance , therefore , O- is not discussed in this section. i l l 8.3-45
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 0 8.3.1.2.1.2.3 Compliance with Regulatory Guide 1.32 (Continued) Compliance of the Class lE battery charger with Regulatory Guide 1.32 is discussed in Subsections 8.3.2.2.1.3.1 and 8.3.2.2.1.2.2. 8.3.1.2.1.2.4 Compliance with Regulatory Guide 1.63 - Electrical Penetration Assemblies in Containment Structures The Nuclear Island design complies with this regulatory guide. 8.3.1.2.1.2.5 Compliance with Regulatory Guide 1.75 - Physical Independence of Electric Systems The Nuclear Island design complies with this regulatory guide, as stated in Subsection 7.1.2.10.18. 8.3.1.2.1.2.6 Compliance with Regulatory Guide 1.93 - Availability e of Electric Power Sources The offsite power sources for a plant referencing this SAR are in the Applicant's scope. Since the above regulatory guide estab-lishes limiting conditions of operation based on combinations of onsite and offsite power, it is the responsibility of the Applicant to address compliance with this guide. 8.3.1.2.1.2.7 Compliance with Regulatory Guide 1.100 - Seismic Qualification of Electric Equipment for Nuclear Power Plants This regulatory guide requires that IEEE Standard 344-1975, with supplemental regulatory requirements, be followed in the seismic qualification procedure of electrical equipment. Compliance with this regulatory guide is discussed in Table 1.8-2. O 8.3-46
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 (} w-8.3.1.4.1.2 Electric Cable Installation (Continued) (4) Cable Fire Protection and Detection - For details of cable fire protection and detection, refer to Subsec-tions 8.3.3 and 9.5.1. (5) Cable and Raceway Markings - All cables (except lighting and nonvital communications) are tagged at their terminations with a unique identifying number. Colors used for identi-fication of cables and raceways are covered in Subsection 8.3.1.3. (6) Spacing of Wiring and Components in PGCC - Separation is accomplished by use of separate steel-enclosed and fire-protected cable raceways in the PGCC floor sections and [ ] separate teamination cabinets or separation of termination cabinet compartments by steel barriers. See NEDO-10466 for detailed documentation of compliance of PGCC with safety criteria, including results of fire tests verify-ing absence of fire propagation properties of PGCC. (7) Spacing of Wiring and Components in Control Boards, Panels and Relay Racks - Separation is accomplished by mounting the redundant devices or other components on physically separated control boards if, from a plant operational point of view, this is feasible. When opera-tional design dictated that redundant equipment be in close proximity, separation is achieved by a barrier or enclosure to retard internal-fire or by a maintained air space in accordance with criteria given in Subsec-tion 8.3.1.4.2. In this case, redundant circuits which serve the same' safety-related function enter the control panel through separated apertures and terminate on sep- [d u arate and separated terminal blocks. Where redundant 8.3-81
GESSAR II 22A7007 ' 238 NUCLEAR ISLAND Rev. 0 8.3.1.4.1.2 Electric Cable Installation (Continued) circuits unavoidably terminate on the same device, bar-riers are provided between the device terminations to ensure circuit separation approved isolators (generally optical) are used. (8) Electric Penetration Assembly - Electric penetration assemblies of different Class lE divisions are separated by distance, separate rooms or barriers and/or by loca-tions on separate floor levels. Grouping of circuits in penetration assemblies follows the same raceway voltage groupings as described in Subsection 8.3.1.4.1. There are a few exceptions, where one assembly consists of circuits in more than one voltage group (V1, V2 or V3) ; in which case, additional flexible ferromagnetic conduits are furnished around each circuit of different voltage g group. W Power circuits going through electric penetration, assemblies are protected against overcurrent by redun-dant overcurrent interrupting devices to avoid penetra-tion damage in the event of failure of any single over-current device to clear a fault within the penetration, or beyond it. 8.3.1.4.1.3 Control of Compliance with Separation Criteria During Design and Installation Compliance with the criteria which insures independence of redundant systems is a supervisory responsibility during both the design and installation phases. The responsibility is discharged by: (1) identifying applicable criteria; (2) issuing working procedure to implement these criteria; 8.3-82 ~
GESSAR II 22A7007 .
, 238 NUCLEAR ISLAND Rev. O l t'
O 8.3.1'.4.1.3 Control of Compliance with Separation Criteria During Design and Installation (Continued) l l (3) modifying procedures to keep them current and workable; (4) checking the manufacturer's drawings and specifications to ensure complaince with procedures;'and (5) controlling installation and-procurement to assure compli-ance with approved and issued drawings and specifications. f' The equipment nomenclature used on standard Reactor Island design is one of the primary mechanisms for ensuring proper separation. 4 Each equipment and/or assembly of equipment carries a single 1 number, (e .g . , the item numbers.for motor drivers are the same j as the machinery drivers). Based on these identification numbers, I each item can be identified as essential or nonessential, and () each essential item can further be identified to its safety separa-tion division. . This is carried through and dictates appropriate treatment at the design level during preparation of the I manufacturer's drawings. Non-Class lE equipment is separated where desired to enhance power 1 generation reliability, although such separation is not a safety consideration. t Once the safety-related equipment has been identified with a Class lE safety division, the divisional assignment dictates a characteristic color (Subsection 8.3.1.3) for positive visual identification. Likewise, the divisional identification of all ! ancillary equipment, cable and associated raceways match the divisional assignment of the system it supports. i l There are certain exceptions to the above where non-Class lE l equipment is: connected to Class lE power sources for functional s_ design reasons (viz. the Standby AC Lighting). This is immediately
- apparent by the absence of essential classification identification 8.3-83 .
, . . _ . . . _ . - _ - . . . _ . . , . . . . _ . . - - _ . - _ . , . . . ~ _ - _ _ -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.3.1.4.1.3 Control of Compliance with Separation Criteria During Design and Installation (Continued) of the connected equipment. The equipment is then designated
" associated" per Regulatory Guide 1.75 (Rev. 0). Cables used to connect such equipment are safety grade and qualified and routed as " associated circuits" and marked as described in Subsection ~
8.3.1.3. 8.3.1.4.2 Independence of Redundant Safety-Related Instrumenta-tion and Control Systems This subsection defines independence criteria applied to safety-related electrical systems and Instrumentation and Control equip-ment. Safety-related systems to which the criteria apply are those necessary to mitigate the effects of anticipated and abnormal operational transients or design basis accidents. This includes all those systems and functions enumerated in Subsections 7.1.1.3, 7.1.1.4, 7.1.1.5, and 7.1.1.6. The term " systems" includes the overall complex of actuated equipment, actuation devices (actuators), logic, instrument channels, controls, and inter-connecting cables which are required to perform system safety functions. The criteria outlines the separation requirements necessary to achieve independence of safety-related functions compatible with the redundant and/or diverse equipment provided and postulated events. l 8.3.1.4.2.1 General Separation of the equipment for the systems referred to in Sub-sections 7.1.1.3, 7.1.1.4, and 7.1.1.6 is accomplished so that they are in compliance with the substance and intent of IEEE 279-1971, IEEE 379-1972, 10CFR50 Appendix A, General Design Criteria 3, 17, 21, and 22, and NRC Regulatory Guides 1.75 and 1.53. 8.3-84
GESSAR II' 22A7007 4 238 NUCLEAR ISLAND Rev. 0
) 8.3.1.4.2.3.2 Other Safety-Related Systems (Continued)
(4) The several systems comprising the ECCS shall have their various sensors, logics, actuating-devices and power l supplies assigned to divisions in accordance with
! Table 8.3-11 so that no single failure can disable a redundant ECCS function. This is accomplished by limit-l- ing' consequences of a single failure to equipment listed in any one division of Table 8.3-11. (See separation specification A62-4350.) The wiring tx) the ADS solenoid valves within the drywell shall_run-in one or more rigid
[. -conduits. ADS conduits'for solenoid A shall-be divi-sionally separated from solenoid B conduits. Short pieces (less than 2 feet) of flexible conduit may be
.used in-the vicinity of the valve solenoids.
(5) Electrical equipment and raceways for systems listed in Table 8.3-11 shall not be located in close_ proximity to l primary steam piping (steam leakage zone), or be designed for-short term exposure to the high tempera-l ture and humidity associated with a steam leak. (6) Any electrical equipment and/or raceways for RPS or ESF located in the suppression pool level swell zone shall be designed to satisfactorily complete their function before being rendered inoperable due to exposure to the environment. created by'the level swell phenomena. This zone includes that space'above the suppression pool normal level which sees the surge of water that could result from a high drywell-to-containment differential pressure. (7) Containment penetrations shall be so arranged that no design basis event can disable cabling in more than one s, ) penetration assembly. Penetrations shall not contain cables of more than one divisional assignment. 8.3-93
. _ . _ _ _ _ _ . _ . _ _ _ _ . . . . _ _ - . _ . _ _ _ - _ - _ _ _ . _ . _ _ . - _ - _ . ~ . - -
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 15 1 8.3.1.4.2.3.2 Other Safety-Related Systems (Continued) (8) Detailed design basis, description, and safety evalua-tion aspects for a power generation control complex (PGCC) System shall be as comprehensively documented and presented in GE Topical Report, Power Generation Control Complex, NEDO-10466A and its amendments. PGCC consists of control room panels, racks, floor sections, and termination cabinets. The floor sections are divided into ducts and the termination cabinets have metallic barriers to separate redundant Class lE wiring. The floor section duces are designed so that each duct acts as a raceway and has. adequate fire barriers and will contain wiring of only one division. The ducts have solid metal walls and floor and a removable solid , metal covers. Cable access to the two PGCC areas is provided through two cable rooms located on either side of the control room. Each cable room contains two divisions; divisional separation is maintained by routing one division in enclosed solid sheet metal cable trays, while the other division is routed in rigid steel conduits which are i completely embedded in the concrete walls or floor to { provide 3-hr fire rated separation. The cable rooms do not contain any high energy equipment, rotating equipment, or piping which could be a potential l source of missiles or pipe whip. No flammable materials are stored in these rooms. Low voltage power cables (V3) l are routed through both cable rooms to provide power for lighting transformers, regulating transformers and instrument buses within the Control Building. The areas are utilized for cable tray and conduit routing only, no other major equipment is housed within the cable rooms. _ 8.3-94
' GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.3.1.4.2.3.2 Other Safety-Related Systems (Continued) (
-See Figures 8.3-30, 8.3-31 and 8.3-32 for physical lay "
l outs of the area. , (9) Annunciator and computer inputs from Class lE equipment i or circuits shall be treated as Class lE'and retain their divisional identification up to and including its input to a Class lE isolation device. The output circuit from this isolation device is classified as nondivisional. 4 Annunciator and computer inputs from non-Class lE equipment or circuits do not require isolation device. 8.3.1.5 Nuclear Island / BOP Interface O The Nuclear Island AC onsite power system provides power to the ESW supply system and the diesel-generator fuel oil storage area and transfer pump. Because of. vendor difference and site-specific .i i. 4 O 1 8.3-94a
GESSAR II. 22A7007 238 NUCLEAR ISLAND Rev. 15 () 8.3.1.5 . Nuclear Island / BOP Interface (Continued)
- l characteristics exact information cannot be provided. Interface control documentation is provided to coordinate power requirements
~ with specific applicant.
8.3.1.5.1 Design Criteria i The design criteria, including codes, standards, general design criteria and regulatory guides for the BOP portion of the Class lE AC onsite electric power system, shall be the same as for Clas's lE power system given in Subsection 8.1. Respective divisions of the ESW and DG Fuel Oil Transfer System i shall be fed from Division 1, 2 and 3 (HPCS) of the Nuclear Island - AC.onsite power system.
)
~
I Motors 450 hp cn: larger shall be fed from 6.9 kV busses. t 8.3.1.5.2 Specific Interfaces
~ Specific Nuclear ~ Island AC onsite power system / BOP system inter-faces _are given in Table 8.3-10. Included are steady state loads,
~
inrush _kVA, nominal voltage, nominal' frequency and allowable fre-quency fluctuation and allowable voltage drop. Reference to the singleline diagram for each interface is also included. See Figures 1.9-Sa and b (E-040A & B) for physical location of the raceway interfaces which are identified as R-XX. Cable interfaces on Tables 8.2-1 and 8.3-10 are identified as E-XX. . 8.3.2 DC Power Systems p 18.3.2.1 Description () The 125'VDC power system is provided for switchgear control, con-
. trol. power, instrumentation, critical motors and emergency lighting i -in control rooms, switchgear rooms and fuel handling areas. The 125 VDC. power system.is not shared between units.
8.3-95'
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 8.3.2.1.1 General Systems Five independent 125 VDC systems are provided to supply Reactor Island normal and emergency DC power for each unit as appropriate. Four of the five 125 VDC systems are Class lE power. The fifth system supplies non-Class lE power. The DC power systems provide adequate power for station emergency auxiliaries and for control and switching during all modes of operation. The operating voltage range of Class lE dc loads is 110V to 140V. The maximum equalizing charge voltage for Class lE batteries is m N 140 VDC. g The DC system minimum discharge voltage at the end of the 2-hr discharge period is 1.83 VDC per cell. The 125 VDC systems provide a realiable control and switching power source for the Class lE systems. All batteries are sized so that required loads will not exceed 80% of nameplate rating, or warranted capacity at end-of-installed-life with 100% design demand. Each 125 VDC battery is provided with two chargers, each of which is capable of recharging its battery from a discharged state to a fully charged state while handling the normal, steady-state DC load. Battery sizes are specified as: (1) Battery E, Division 1 - 1950 A-hr at 8-hr rate; 2080A for 1 min O 8.3-96
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 I') 8.3.2.1.1.1 Balance-of-Plant DC System
)
(Provided by Applicant) 8.3.2.1.1.2 Switchyard DC System (Provided by Applicant) 8.3.2.1.2 Class lE DC Loads Tables 8.3-6 through 8.3-9 list the 125 VDC loads required for the four divisions of 125 VDC power. The 125 VDC Class lE power is required for emergency lighting, diesel-generators field flashing, control and switching functions such as the control of 6.9-kV and 480V switchgear, control relays, meters and indicators, as well as DC components used in the Reactor Core Isolation Cooling System, eN The two divisions that are essential to the safe shutdown of the
- reactor are supplied from two independent 125 VDC systems DC-E, and F.
8.3.2.1.3 Station Batteries and Battery Chargers, General Considerations There are two basic considerations for the 125 VDC requirement of the plant: the Class lE systems and the non-Class lE systems. The four ESF load groups are supplied from the four Class lE 125 VDC systems identified as DC-E, F, G and H. The non-class lE system is supplied by the 125 VDC system identified as DC-J. I Each of the 125 VDC systems (E, F, G, H and J) has a 125 VDC i l battery, two battery chargers and a distribution panel. The main DC distribution buses for E, F, H and J systems are in DC dis-tribution panels with drawout-type breakers and molded case circuit breakers. The main distribution bus for system G is a DC dis-tribution panel with molded case circuit breakers. Local distribu-(_j tion panels and motor control centers are fed from the DC distribution panels. l 8.3-97
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 8.3.2.1.3 Station Battery and Battery Chargers, General Considerations (Continued) The 125 VDC systems E, F, G and H supply DC power to Divisions 1, 2, 3 and 4, respectively, and are designed as Class lE equipment in accordance with applicable clauses of IEEE Std 308-1971. They are designed so that no single failure in any 125 VDC system will result in conditions that prevent safe shutdown of.the plant. The plant design and circuit layout from these DC systems provides physical separation of the equipment, cabling and instrumentation essential to plant safety. Each 125 VDC battery is separately housed in a ventilated room apart from its charger and distribution panel. Each division of the system is located in an area separated physically from other divisions. All the components of Class lE 125 VDC systems are housed in Seismic Category I structures. An emergency eye wash is supplied in each room. All chargers are sized to supply the non-tinous load demand to their associated batteries while restoring batteries to a fully charge state. There are no automatic interconnections between DC systems. A crosstie is provided between systems E and F and equipped with a circuit breaker on each end of the tie in each distribution panel. The crosstie is provided for use during shutdown periods for maintenance and is manually initiated and protected by key interlocks. 8.3.2.1.3.1 125 VDC Systems Configuration Figure 8.3-18 shows the overall 125 VDC system provided for Class lE Divisions 1, 2 and 3. Figure 8.3-19 shows the overall system for Class lE Division 4 and the non-Class lE 125 VDC system. Two divisional battery chargers are used to supply each divisional DC distribution panel bus and its associated battery. _{ l The divisional battery chargers are normally fed from divisional , 1 9.3-98
GESSAR II 22A700'7 238 NUCLEAR ISLAND Rev. 15
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8.3.2.1.3.1 125 VDC Systems Configuration (Continued)
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480V MCC busses. The redundant alternate battery chargers are supplied from nondivisional 480V MCC. buses. _ The nondivisional DC distribution panel (DC-J) has two sections , (sections X and Y) which are each connected to nondivisional battery chargers. Each battery charger is fed from separate non- , divisional 480 VMCC buses for the normal and the redundant alter-nate supplies. . 8.3.2.1.3.2 Battery Capacity Considerations The amp-hr capacity and short-term rating of batteries are in accordance with criteria given in IEEE Std 308-1971, Para-graph 5. 3. 3 (2) . They are adequate to supply all electrical loads. These batteries have sufficient stored' energy to operate connected (~
\- essential loads continuously for at'least two hours without recharging. Tables 8.3-6 through 8.3-9 give loads to be supplied a from the 125 VDC Class lE systems. Each distribution circuit is capable of transmitting sufficient energy to start and operate all required loads in that circuit.
l l An initial composite test of onsite AC and DC power systems is called for as a prerequisite to ini.tial fuel loading. This test will verify that each battery capacity is sufficient to satisfy a j safety load demand profile under the conditions of a LOCA and loss of preferred power. Thereafter, periodic capacity tests may be conducted in accordance with Subs'ection 8.3.2.1.3.4. These tests l l will ensure that the battery has the capacity to continue to meet safety load demands. f k_ e) l 8.3-99 l .-
- ~ GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 8.3.2.1.3.3 Ventilation Battery rooms are v'entilated to remove the minor amounts of gas produced during the charging of batteries. The ventilation system of Class lE battery rooms has a fan on Class lE power and a backup fan on non-Class lE power. The former is a Seismic Category I fan, the latter is not. O O 8.3-99a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 ( 8.3.2.2.1.2.3 Compliance with Regulatory Guide 1.93 - Availability O. of Electric Power Sources The offsite power sources fos a plant referencing this SAR are in the Applicant's scope. Since the abcive ' regulatory guide estab-lish,es limiting conditions of operation based on combinations of I onsite and offsite power, it is the responsibility of the Applicant to addreas compliance'with this guide. 18.3.2.2.1.2.4 Compliance with Regulatory Guide 1.100 - Seismic
; Qualification of Electrical Equipment for Nuclear Power Plants This. egulatory guide requires that IEEE Standard 344-1975, with supp1'emental regulatory requirements, be followed in the seismic q'ualification procedure of electrical equipment. This regulatory guide is complied with for Nuclear Island DC equipment as discussed in Subsection 1.8.100. < -
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O'. 8.3.2.2.1.2.5 Compliance with Regulatory Guide 1.106 - Thermal Overload Protection for Electric Motors on Motor-
! Operated Valves This regulatory guide is not directly applicable to DC systems, and f'Ys is addressed in Subsection 8.3.1.2.1.2".
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+- 8.3.2.2.1.2.6 Compliance with Regulatory Guide 1.108 - Periodic Testing of Diesel-Generator Units Used as Onsite Electric Power Systems at Nuclear' Power Plants This. regulatory guide is not directly applicable to DC systems, N, and is addressed in Subsection 8.3.1.2.1.2.
A 8.3.2.2.1.2.7 Compliance with Regulatory Guide 1.118 - Periodic Testing of Electric Power and Protection Systems
),TheprovisionsofthisGuideareappliedtotheNuclearIsl'andas discussed in Subsection 1.8.118.
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LiL 8.3-105 ,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14
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8.3.2.2.1.2.8 Compliance with Regulatory Guide 1.128 - Installation Design and Installation of Large Storage Batteries for Nuclear Power Plant j The Class IE batteries are specified and located in accordance with IEEE Standard 484-1975, as modified and augmented by Regula-tory Guide 1.128, Revision 1. The three-tier option discussed in the IEEE 484-1981 Standard was used for the DIV 4 battery (Battery H) arrangement, where the height of the 3-steps is equal to or less than the height of the two-tiered batteries. The lower height is possible because of the fact that the Division 4 9 battery (Battery H) is much smaller than the Division 1 and 2 batteries. Space limitation for access necessitated this approach. No deleterious effects are anticipated and maintenance activities were evaluated to be acceptable. Compliance with the Safety, Installation Procedures and Records Section of IEEE 484-1975, as modified and augmented by Regula-tory Guide 1.128 is the responsibility of the Applicant. _ 0 8.3-105a
- - __ . . - . __ _~ . _ _ -_- _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 p-8.3.2.2.1.2.9 Compliance with Regulatory Guide 1.129 - Maintenance, Testing and Replacement of Large Lead Storage Bat-teries for Nuclear Power Plants The Nuclear Island design allows space for maintenance and removal of batteries from the storage battery rooms. See Figure 1.2-4 for the arrangement. Description of the maintenance, testing, and replacement program for large lead storage batteries is the responsibility of the Applicant. 8.3.2.2.1.2.10 Compliance with Regulatory Guide 1.131 - Qualification Tests of Electric Cables, Field Splices, and Connections for Light-Water-Cooled [ ) Nuclear Power Plants The Nuclear Island design complies with this Guide as discussed in Subsection 1.8.131.
]-
l 8.3.2.2.1.3 Compliance with IEEE Standards 8.3.2.2.1.3.1 Compliance with IEEE Standard 308-1971 - Criteria for Class lE Electric Systems for Nuclear Power Generating Stations i See Subsections 8.3.1.2.1.3.1 and 8.3.1.2.1.3.2 for compliance of Class 1E systems with IEEE Standards 279-1971 and 308-1971. 8.3.2.2.1.3.2 Compliance with IEEE Standard 384-1974 - Trial Use Standard Criteria for Separation of Class lE l ' Equipment and Circuits i Each Class lE division has its own 125 VDC battery. Each battery (} is installated in a separate _ room which has fire-resistive walls, i i 8.3-106 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 14 () SEQUENCE OF EVENTS IN AUTOMATIC APPLICATION Table 8.3-5 OF EMERGENCY AC LOADS UPON LOSS OF COOLANT Event Time (sec) Comment Design basis LOCA signal (-10.03 sec) Solid-state drivers 33 ms propagation time Signal to start diesel (-10 sec) Standby and HPCS diesels ready 0 By definition (bus , to load; start LPCS pump and energized) RHR pump C; apply power to LPCS,HPCS, RHR selected 480V auxiliaries Auxiliaries (El o and motor-operated valves and F1 buses) 9 m Start RHR pumps A & B 5 , Start ESW pump 1 10 All ECCS pumps at rated speed 25 Completes ECCS [d)
\ starting sequence Injection valves fully open 40
- Tripped off by LOCA signal.
O l l 8.3-129
Table 8.3-6 125 VDC BUS DC-E LOADS (DIVISION 1) Amperage Requirements Per Time Interval After AC Power Loss Description 0-1 Min 1-2 Min 2-120 Min RCIC motor-operated valves 365 99 - RCIC control system 32 10 10 RCIC gland seal compressor - 130 58 Control Room control circuits 50 30 30 Auxiliary Building control circuits 50 25 25 $ m Diesel-generator control 10 10 10 ga g ** Diesel-generator flashing 78 - A
- Emergency lighting 28 28 28 yy o Indicator lamps 4 4 4 ss MH
*Switchgear 90 - -
p z O NSPS inverter 80 80 80 ESW substation 31 6 6
- Emergency lighting is connected to ESF bus E and circuits connected to the lighting fixtures are treated as associated. The fixtures and lamps are not, themselves, Class IE equipment. However, it is desirable that emergency lighting be available during a LOCA and this justifies the use of buses which are not disconnected during a LOCA.
**Short duration pulse. gU Afield flashing is allowed for 5, 10, 15 and 20 minutes. <>-2 o
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w G J l ! Table 8.3-7 125 VDC BUS DC-F LOADS (DIVISION 2) i Amperage Requirements Per Time Interval After AC Power Loss Description 0-1 Min 1-2 Min 2-120 Min Control Room control circuits 35 35 35 Auxiliary Building control circuits 50 25 25 Diesel generator control 10 10 10 l ** Diesel-generator flashing 78 - A
- Emergency lighting 15 15 15 ya am Y N$
e Indicator lamps 4 4 4 >> w w ww Switchgear 90 - - y[ t NSPS inverter 80 80 80 5 z O i ESW substation 31 6 6 i
- Emergency lighting is connected to ESP bus F anci circuits connected to the lighting fixtures are treated as associated. The fixtures and lamps are not, themselves, Class lE equipment. Ilowever, it is desirable that emergency lighting be available !
during a LOCA and this justifies the use of buses which are not disconnected during i a LOCA.
**Short duration pulse.
Afield flashing is allowed for a 5, 10, 15 and 20 minutes. g>U
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Table 8.3-8 125 VDC BUS DC-G LOADS (HPCS) (DIVISION 3) Amperage Requirements Per Time Interval After AC Power Loss Description 0-1 Min 1-60 Min 60-120 Min Diesel engine control 2 2 2 Generator auxiliary control 2 2 2 Field flashing 20 - - Solenoid valves 2 ro w 2 2 2
- Relays HPCS logic panel z
CQ os O t1 Indicator lamps Control Room panel 2 2 2 yy h Switchgear (breakers closing) 28 - - U Diesel standby fuel pump , 10 10 10 y[ NSPS inverter 60 60 60 $ z O M o>
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I r Table 8.3-9 , 125 VDC BUS DC-H LOADS (DIVISION 4) i Amperage Requirements Per Time Interval After AC. Power Loss t Description 0-1 Min 1-60 Min 60-120 Min i l - i NSPS inverter 80 80 80 i _ Control Building control circuits 10 10 10 l \ Auxiliary Building control circuits 10 10 10 ; j N t W I 00 ** i . z CO ' y OM MM ) m MM 4 N >> < i , ,co WW ,
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Q,/ uj Li Table 8.3-10 NUCLEAR ISLAND ONSITE AC AND DC POWER SYSTEM / BOP INTERFACES Maximum Actual Max Maximuse Interface Reference Allowable Inrush Nom Freq Norber Interface Description Figure Allowable Available SS_HP/lLVA kVA _ Volt. Hs ;OP VO SC Current F-8 (A-C) s900V Feeder from Bus E. SWCR to BOP Figure 8.3-2 2335* 7590 6900 60 24 81 41.8 kA Div 1 FSWS SWCR E-11 (A-C) 6900V Feeder frne Bus F, f.T,R to BOP Fiqure 8.3 2 2335* 7590 6900 60'21 8% gg, ga Div 2 ESWS SWCR E-12 (A-C) 480V Feeder from Bus C1-2 to BOP Figure 8.3-16S 60 296 400 - 60 24 3.75% 25 kA Div 3 (HPCS) ESWS Pump E-17 480V Feeder from Bus El-2, MCC to BOP Figure 8.3-16P 3 25.5 480 60-24 1.5% Day 1 DC Fuel 011 Transfer Pump 15 kA E-18 480V Feeder from Bus El-2, MCC to BOP Figure 8.3-16P 3 25.5 480 60'2% 1.5% 15 kg M Div 1 Back up DC Fuel 011 Transf er Pump W E-20 480V Feeder from Bus F1-2, MCC to B6P Figure 8,3-16P 3 25.5 490 60*2n '3.5% Div 2 DG Fuel Oil Transfer Pump 15 kA E-21 480V Feeder from Bus F1-2, MCC to BOP Figure 8.3-16P 3 25.5 480 60'24 3.5% g5 g3 OO E Div 2 Back up DG Fuel Oil Transfer O E'3 e Pump DM g i W E-23 480V Feeder from Bus Cl-2, MCC to BOP Figure 8.3-16S 3 25.5 480 60t2% 3.754 15 kA .P" l Div 2 (HPCS) DG Fuel Oil TranP(er Pump g F8 W E-24 480V reeder from Bus C1-2, MCC to BOP Figure 8.3-16S 3 25.5 480 60 24 3.754 13 kA yg sJg Div 3 (HPCS) Back up DC Fuel Oil gg Transfer Pump E-25 480V Feeder from 9us C1-2, MCC to BOP Figure 8.3-16S 1 8 480 60 21 3.754 15 kA 1 Div 3 (HPCS) ESWS Strainer Mctor E-44 480V Feeder from Bus C1-2, MC" to BOP Figure 8.3-16S 3 25.5 480 6042t 3.754 15 kA Div 3 (HPCS) ESWS Backwash Strainer Valve E-49 480V Feeder from Bus Cl-2, MCC to BOP Figure 8.3-16S 1 8 480 E0 24 3.75 15 kA Div 3 (HPCS) ESWS Return Header Isolation Valve
*fhTalgure includes an allowance for an additional ESW Pump (1133 HP) which can be connected manually by the operator. Actual steady state load for 6900V feeder from buses E and F is 1202 for each feeder.
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Table 8.3-10 (Continued) NUCLEAR ISLAND ONSITE AC AND DC POWER SYSTEM / BOP INTERFACES Max Maximum Reference SS Inrush Num Freq Allowable Available fnterface Volt. Hz Number,_ Iny rfa_ce_Desgapt Qn _ Fip re HP/kVA yA_ BOP VD _ SC_ Current 480V Fceder from Bus El-2, MCC to BOP Fiqure 8.3-16P 15 -- 480 60 21 3.51 25 kA E-59 Lav 1 Fuel 011 Stora<ie Tank Receptacle 480V Feedc. f rom Bus F1-2, l4CC to BOP Faqure 8.3-16P 15 -- 480 60 24 3. 50 25 kA E-60 Day 2 Fuel 011 Storage Tank Receptacle 480V Feeder f rorr Bus Cl-2, MCC to BOP Fiqure 8.3-16S 15 -- 480 60+24 3.51 25 kA E-61 Day 3 ( AIPCS' *X; Fuel 011 St oraise Tank Heceptacle Figure 8.3-16C 7.5 48 480 63?2% 3.5t 35 kA E 4SOV Feeder from MCC B1-1 to steam-E-55 Inne Drain Valves & D y 60 2t -- -- W E-74 From Recare Pump "A" Current Trans- Faqure 8.3 3a 5A -- -- forner to BOP Switchaear Dev&ce Figure 8.3-3a SA -- -- 60 21 -- -- Z E-76 From Recire Pump "B" Current Trans-former to BOP Switchgear Device CO OM @ Figure 8.3-16S 2 20 480 60t24 3.751 15 kA p (f) E-82 ESW Pump Stat 2on Div i Area Supply M (f) from CE Div 3 MCC-G1-2 W Fagure 8. 3-16F 25 140 480 60 24 51 35 kA
>>y l E-84 (95) Power Feed for Excess Water Pump H A, (B) (Radwaste) G17 from MCC A3-1 3.754 15 kA HH Pnwer Feed for MOV "A" ("B") C17 Figure 8.3-16F 0.5 5 480 60 2t WH E-86 (87) from MCC A3-1 U 1211 Signal -- 120 -- --
E- 9 3 - 120 Volt AC Feeder from Bus J1 to (AC) E-111 Audio Alarre O Figure 8.3-18 350A -- 125 -- 41 20 kA E-34 (A-D) 125 VDC Battery Test Feede.- from Dav 1 Battery (E) to BOP VDC Faaure 8.3-19 150A -- 225 -- At 15 kA E-35 (A-B) 125 VDC Battery h st reeder from Dav 4 Battery (HI to BOP VDC 125 VDC Battery Test Feeder from Figure 8.3-18 250A -- 125 -- 41 15 kA E-36 (A-D) VDC Div 2 Battery (F) to BOP NN (D M
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t , } ) Table 8.3-10 (Continued) l NUCLEAR ISLAND ONSITE AC AND DC POWER SYSTEM / BOP INTERFACES i> s i Max Maximum I j interface Reference SS . Inrush Nom Freq Allowable Available j Number Interface Description Fiqure IIP /kVA kN Volt. Hz BOP VD SC Current ; E-38 (A-DI 46 ' f 125 VDC Battery fest l' ceder from Fagure 8.3-18 100A -- 125 -- Div 3 Battery (C) to Bor VDC l E-79 ESW Pump Station Div I Arc.a supply figure 8.3-16S 2 2fs 480 60 21 3.751 15 kA ran "D" from CE Div 3 MCC cl-2 l E-80 FDH to ESW Pump Station Div 1 CPT FDR - Fiqure 8.3-16S 5 -- 480 60 3.75* l $ E-6 (A-B) 125 VDC Control FDR f a om Bus DC-E Fiqurc 8.3-18 6 31 125 -- SS 20 kA
- i SWGR to ESWG SWGR [
! . N f j E-9 iA-B) 125 VDC Control FDR from Bus DC-r rigure 8.3-18 6 31 125 -- 8' 15 kA (,J ; g SWGR to WSWS SWCR Qll j i h Z
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3 -~s Table 8.3-12 (Continued) BUS CONDITION INDICATION LOCATION DC-J Unde'; voltage, Overvoltage, Ground Fault ND 125 VDC Control Open Battery Main Breaker Bus Trouble Room l Open Battery Disconnect Switch Low Battery Charger DC Volt and Amp Annunciator Low Battery Charger AC Input Volts Panel Bus Voltage Local and Voltmeter Control Rm Bus Ammeter Battery Amp and Volts Volt and Ammeter Local DC-El Undervoltage D1 125 VDC MCC DC-El Trouble N W Ground Fault Control Rm Annunciator m D1 125 VDC Bus DC-E and MCC DC-El Control Rm Z
, and MCC DC-E2 Trouble Status Lights w
i Local and @Q em Bus Voltage and Current Voltmeter & Ammeter H Control to m a Rm >>
*d lC h DC-F1 Undervoltage D2 125 VDC MCC DC-F1 Trouble Control Rm HH Ground Fault Annunciator MH I
D2 125 VDC Bus DC-F and MCC Control Rm DC-F1 Trouble Status Lights 4 Local and a Bus Voltage and Current Voltmeter and Ammeter Control Rm DC-E2 Undervoltage Ground Fault D1 125 VDC MCC DC-E2 Trouble Control Rn Annunciator D1 125 VDC DC-E and MCC DC-El Control Ra and MCC DC-E2 Trouble Status Lights Bus Voltage and Current Local and Voltmeter & Ammeter Control Ra Oc N (D N
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Table 8.3-12 (Continued) SUPPLEMENT TO TABLE 8.3-12 BUS CONDITION INDICATION LOCATION Div. 3 Continuous Bus Voltage Voltmeter Local and (HPCS) Control Room 125 VDC Bus "G" Battery Output Current Ammeter Local Bus "G" Load (Amp) Ammeter Local Bus "G" Ground Fault 125 Vdc System Control Room N Bus "G" Undervoltage Trouble Alarm co Battery Breaker 1 Open (D Battery Breaker 2 Open co O to Y Control Power Failure to DG Cont Pnl Control Power Failure Alarm as "DG Trouble" b cn [ N and Local gg O Battery Charger Input Breaker Tripped /Open HH Battery Charger Failure Battery Charger Trouble Control Room MH (including high voltage and ground fault) Alarm Low Battery Charger Amps g Charger Output Voltage Voltmeter Local Charger Output Current Ammeter Local Charger Ground Fault Ground Indication Light Local ld M to M
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 6 r 9.1.4.2.3.7 Jib Crane 1 The jib crane (Figure 9.1-9) consists of a motor-driven swing boom monorail and a motor-driven trolley with an electric hoist. The jib crane is mounted along the edge of the fuel building fuel storage pool to be used during refuezing operations. Due of the jib crane leaves the refueling platform or fuel-handling platform free to perform general fuel shuffling operations and still permit i uninterrupted fuel preparation in the work area. The hoist has two full-capacity brakes and in-series adjustable up-travel limit _ switches. Upon hoisting, the first of two independently adjustable limit switches automatically stop the hoist cable terminal approx-imately 8 ft below the jib crane base. Continued hoisting is
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possible by depressing a momentary contact (up-travel override pushbutton on the pendant) together with the normal hoisting push r button. The second independently adjustable limit switch automati _ cally-interrupts hoist power at the maximum safe uptravel limit. O' When the jib crane is used in the handling of hazardous radioactive _. materials that must be kept below a specific water level, a fixed mechanical stop is installed on the hoist cable to prevent further - hoisting when that level is reached. The jib crane is normally located adjacent to the fuel storage pool and connected to the , service outlet provided. 9.1.4.2.3.8 Fuel-Handling Platform Refer to Subsection 9.1.4.2.7 for a description of the fuel-handling platform. 9.1.4.2.3.9 Channel-Handling Boom A channel-handling boom (Figure 9.1-10) with a spring-loaded i balance reel is used to assist the operatcr in supporting a portion of the weight of the channel as it is removed from the fuel assem-
- s. ) bly. The boom is set between the fuel preparation machines. With the channel-handling tool attached to the reel, the channel may be conveniently moved between the fuel preparation machines.
9.1-31
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.1.4.2.3.10 Fuel Transfer System The Inclined Fuel Transfer System (Figure 9.1-13a) is used to transfer fuel, control rods, defective fuel storage containers and other small items between the containment and the fuel building pools by means of a carriage traveling in a transfer tube (a 23-in. pools by means of a carriage traveling in a transfer tube (a 23-in. I.D. stainless steel pipe). In the containment upper pool, the transfer tube connects o pool penetration and to a sheave box. Connected to the sheave box is a 24-in, flap valve, a vent pipe, cable enclosures and a fill valve. In the fuel building pool, the transfer tube connevts to a 24-in. gate valve. A bellows connects the building penetration to the valve and transfer tube to prevent water entrapment between the tube and penetration. A 4-in. Weldolet located on the transfer tube approximately 2 ft above the fuel building pool water level and a motor-operated valve are provided for connections to a drain pipe for water level control in the transfer tube. A containment isolation assembly containing a blind flange and a bellows, which connects the containment isolation assembly to the containment penetration, is provided to make containment isolation. A hand-operated 24-in. gate valve is provided to isolate the reactor building pool water from the transfer tube so that the blind flange can be installed. A hydraulically actuated upender is provided in each pool for rotating part of the carriage (tilt tube) to the vertical position for loading and unloading and to the inclined position for trans-fer. The carriage consists of the tilt tube and a follower con-nected with a pivot pin which allows upending of the tilt tube while maintaining the follower in the inclined position. The carriage has rollers and wheels which ride on tracks within the transfer tube and upenders to assure low friction, correct carriage i orientation and smooth transition across valves and between other components. The tilt tube is designed to accept two different inserts - a fuel bundle insert with a two-bundle capacity and a control rod insert for control rods, defective fuel storage container, and other small items. l 9.1-32 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 () 9.4 AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEM 9.4.1 Control Building HVAC System 9.4.1.1 Design Bases 9.4.1.1.1 Safety Design Bases (1) The control building Heating, Ventilating and Air-Conditioning (HVAC) System is designed with sufficient redundancy to ensure operation under emergency conditions assuming the single failure of any one active component. 4 (2) Provisions are made in the system to detect and limit the introduction of airborne radioactive material into the control room. () (3) With the exception of the following components which are not required to be safety-related, the Control Building HVAC System is designed to Seismic Category I requirements: Area exhaust fans and chiller-room exhaust fans and the associated ductwork extending from utility areas to the fan discharge dampers and electric duct reheat coils. ' (4) Provision is made in the system to detect and remove smoke and radioactive material from the control room. (5) The HVAC system is designed to provide a controlled temperature environment to ensure the continued operation of safety-related equipment under accident conditions. l O)
\-
i 9.4-1
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.1.1.1 Safety Design Bases (Continued) (6) The HVAC system and components are located in a Seismic Category I structure that is tornado-missile and flood protected. (7) The HVAC system has been evaluated for the effects of postulated pipe failure and initiation of internally generated missiles. Protection has been provided to the system where necessary to mitigate the consequences of such failures as described in Sections 3.5 and 3.6. (8) For compliance with codes, standards and regulatory guides, see Sections 3.2 and 1.8. 9.4.1.1.2 Power Generation Design Bases (1) The HVAC system is designed to provide an environment with controlled temperature and humidity to ensure both the comfort and safety of the operators. The design conditions for the control room environment are 75 F and 50% relative humidity. (2) The system is designed to permit periodic inspection of the principal system components. (3) The outside design conditions for the control room HVAC system are 95 F during the summer and -10 F during ] the winter. 9.4.1.2 System Description The Control Building is heated, cooled and pressurized by a recir-culating air system with filtered outdoor air for ventilation and pressurization purposes. The recirculated air and the outdoor air 9.4-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 (Oj 9.4.1.2 System Description (Continued) The chiller room exhaust system consists of an operating fan and a standby fan. The fans are powered from the normal power bus. Space cooling units are provided in the chiller rooms. The cool-ing units are powered from the same bus as the chillers. The cooling units provide room cooling if the supply from the air-conditioning unit is shutoff. The filter rooms and the air handling room are heated in winter by electric unit heaters. The unit heaters are powered from the normal power bus. 9.4.1.3 Safety Evaluation The Control Building HVAC System is designed to maintain a habit-() able environment and to ensure the operability of components in the control room. With the exception of those items listed in Subsection 9 . 4 .1.1.1 ( 3 ) , all control room HVAC equipment and surrounding structures are of Seismic Category I design and operable during loss of the offsite power supply. l The ductwork which services these safety functions is termed ESF ductwork, and is of Seismic Category I design. ESF ducting is high pressure safety grade ductwork designed to withstand the
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maximum positive and/or negative pressure to which it can be sub-jected under normal or abnormal conditions. Galvanized steel
~
ASTM A526 or ASTM A527 is used for outdoor air intake and exhaust ducts. All other ducts are welded black steel ASTri A570, Grade A or Grade D. Ductwork and hangers are Seismic Category I. Flanged and welded joints are qualified per ERDA 76-21. . Redundant components are provided where necessary to ensure that a Oj s single failure will not preclude adequate control room ventilation. A system failure analysis is provided in Table 9.4-2. 9.4-5
GES3AR II 22A7007 238 NUCLEAR ISLAND Rav. 0 9.4.1.3 Safety Evaluation (Continued) A radiation monitoring system is provided to detect high radiation in the outside air intake ducts. A radiation monitor is provided in the control room to monitor control room area radiation levels. These monitors alarm in the control room upon detection af high radiation conditions. Isolation of the control room and initiation of the outdoor air cleanup unit fans are accomplished by the following signals: (1) high radiation in the outside air intake duct, and (2) manual isolation. An evaluation of the dose to the operators under various postulated ' accident conditions is presented in Chapter 15. Under normal conditions, sufficient air is supplied to pressurize the control room and exfiltrate to pressurize the Control Building. For other forms of contamination (such as smoke and freon) , provision is made to purge the room, with no recirculation. Details on control room habitability are discussed in Section 6.4. The safety-related isolation valves at the'outside air intake are protected from becoming inoperable due to freezing, icing, or other environmental conditicns. 9.4.1.4 Inspection and Testing Requirements Provisions are made for periodic tests of the oncdoor air cleanup fans and filters. These tests include determinations of differential pressure across the filter and of filter efficiency. Connections for testing, such as injection, sampling and monitor-ing are properly located so that test results are indicative of performance. ' 9.4-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 [) U 9.4.2.1 Design Bases (Continued) installed in the area above and below the operating floor. Each ' will automatically modulate dampers located in the outside air stream to maintain a -1/4 inch water gage pressure in the space. When the negative pressure in the space is higher than -1/4 inch water gage, an alarm will actuate. When the railroad doors are open, the volume control dampers in the railroad car area will be manually closed to hold the remainder of the building at the normal pressures. Isolation valves located in the outdoor air inlet duct will automatically close during a LOCA condition or an indication of high airborne radia-tion in the exhaust' air. All areas normally occupied by personnel are provided with radiation detection instrumentation. Abnormal radioactivity
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'() levels initiate alarms and automatically isolate the ventilation system. M m When high airborne radioactivity is detected, the normal exhaust fan is automatically shutdown, the fan shutoff damper closed, and the standby gas treatment system started to process all exhaust air. l The HVAC intake and exhaust louvers and vents are protected from any projectile by a steel-reinforced concrete structure which obstructs any direct path between an external missile and the HVAC opening. The protecting structure is attached to and a part of the building.
- The corresponding safety and power generation design bases are
_ given in the paragraphs to follow. I)
\- ~
9.4-10a
l l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 9.4.2.1.1 Safety Design Bases i (1) Fuel Building penetrations for the pressure control ducting shall be of ESF quality. (2) Redundant isolation valves shall be used at all HVAC penetrations of the Fuel Building. (3) All air discharged from the fuel-handling area, and other potentially radioactive areas, shall be subject to continuous radiation monitoring. (4) The Fuel Building shall be maintained at a negative pressure with respect to atmosphere. (5) All ducting and HVAC components that are operational during LOCA cr high radiation conditions shall be of ESF quality. (6) The pressure control exhaust system shall be composed of two parallel units. (7) Inlet ducting to the SGTS shall be fitted with butterfly valves to ensure positive isolation. 9.4-10b ]
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 O's / 9.4.2.2.1 Air-Conditioning Units (Continued) j in the fuel pool area with the controller set at 80*F. For the I lower air-conditioning unit, the controller and sensing element ar-located in the FPCCU heat exchanger area with the controller set at 90*F. I-A temperature-indicating switch (low limit) at each unit auto-
.matically shuts the air-conditioning down if the discharge tempera-ture drops below 50 F. Dampers in the discharge side of the ducting prevent backflow of air through either unit, if idle.
Air-operated dampers in each makeup air inlet duct are modulated to maintain Fuel Building pressure at (-)l/4 in, water column with respect to atmosphere. Exceptions are the portions of stair tower not in secondary containment (see Subsection 6.2.3 for detailed definition), and the entire Fuel Building during the brief time when the railroad door is open. A differential
'- pressure controller, sensing the building differential pressure inside and outside with respect to atmosphere, controls each damper to regulate the amount of makeup air admitted.
Two air-operated butterfly valves located in the outdoor air inlet duct close upon a LOCA signal or if high airborne radiation is detected in the exhaust air. Under the same conditions, the pressure control exhaust air system is shut down and all exhaust air is diverted to the SGTS. 9.4.2.2.2 Pressure Control Exhaust Air System The pressure control exhaust air system flow diagram is shown in Figure 9.4-2 (K-169). It consists of two exhaust air fans in parallel. Fan X63-CC002A is operational and is powered from the normal bus. Fan X63-CC002B is standby and is powered from the normal bus. A backflow damper in each fan discharge duct (~) prevents backflow through the idle fan. A LOCA signal or 9.4-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.2.2.2 Pressure Control Exhaust Air System (Continued) detection of high airborne radiation in the exhaust air stream causes the pressure control exhaust fan to be shut down and the air-operated inlet dampers to be closed. All exhaust air is then directed to the SGTS. Under normal conditions, the exhaust air is directed to the plant vent. Startup of the pressure control exhaust air system is by manual l pushbutton station in the main control room. An airflow switch in the discharge duct initiates automatic changeover to the standby unit if the operational unit is not moving sufficient air to keep the contacts of the switch open. An interlock between the exhaust fans and the supply fans (air-conditioning unit) prevents the air-conditioning unit from starting until the exhaust unit is functioning. 9.4.2.2.3 Standby Gas Treatment System Actuation Under emergency conditions such as a LOCA or high airborne radiation, exhaust airflow is diverted to the SGTS. When high radiation occurs in the Fuel Building, it is isolated and 5000 cfm is drawn from it into the SGTS. During a LOCA or general high radiation, the 6000 cfm capacity of the SGTS is dis- } tributed among the buildings involved. Positive isolation and sys-tem functional reliability during emergency are ensured by two butterfly valves in series per division of SGTS . The inlet to the SGTS is provided with parallel pairs of isolation valves. These valves are powered from Division 1 and from Divi-sion 2 for redundancy. The valves are normally closed, and only open when dictated by a LOCA signal or by the detection of high airborne-radiation in the normal exhaust duct. Control of the valves is interlocked with pressure control exhaust air system dampers, the air-conditioning isolation butterfly valves and the ESF ducting butterfly valves that align the system for appropriate 9.4-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () 9.4.2.2.6 Reheat Coils (Continued) (1) upper part of service area (Zone 1); (2) upper part of service area (Zone 2); and (3) upper part of fuel pool area. . Each space has a hot water heating coil installed in the supply duct. Waterflow, and thus space temperature, is modulated by a three-way valve, which is, in turn, controlled by a temperature-indicating controller in each space. A temperature-indicating switch warns of temperature extremes caused by equipment malfunc-tion or failure. The control temperature setpoint is 80 F. 9.4.2.2.7 Cooling Coils Other spaces in the Fuel Building that require individual cooling are-listed below: (1) FPCCU pump and Transfer Pump Room, Division 2; (2) FPCCU Pump Room, Division 1; (3) motor generator set area; (4) Division 1 shield annulus Pan Room; and (5) Division 2 shield annulus Fan Room. These units consist of a cooling coil, coil drain pan and fan. Startup is by remote-manual push button station in the main control room or by temperature-indicating switch. Temperature control is
~
maintained by cycling the fan motor. Setpoint is 100*F. Water
~
in all cases is ESW from the appropriate division, except for Reactor Island chilled water to the motor-generator (MG) set coolers. v) 9.4-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.2.3 Safety Evaluation The ESF portion of the Fuel Building HVAC System is designed to satisfy the following criteria: (1) maintain a suitable environment for the function and operability of any safety-related components and systems (All safety-related components are ESF quality and operable during loss of offsite power. Redundant components are provided where deemed necessary to pre-clude loss of safety function in a single-component failure condition); and (2) detect radiation in the spent fuel handling area and pre-vent release of radioactive substances from the spent fuel handling area to the environment following a fuel-handling accident or other operational transient; and (3) filter contaminants out of the air before exhausting to the environment. The system has no other safety-related function other than as de-
]
fined in section 3.2. Failure of the system does not compromise any safety-related system function and does not prevent safe reactor shut-down. The failure mode and ef fects analysis is given in Table 9.4-4. 9.4.2.4 Testing and Inspection Provisions are made for periodic testing of the safety-related equipment. The tests include determination of differential pressure and filter efficiencies, control setpoints and signals, alarm functioning, modulating valve performance, airflow rates, damper functioning, butterfly valve functioning and thermal per-formance of heating and cooling coils. Test connections are pro-vided for sampling and monitoring the above-noted categories of performance. 9.4-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 () 9.4.3.2.1.2 Auxiliary Building Pressure Control (Continued) fan delivers processed air into the ECCS corridor ductwork, where it is evenly distributed through registers. The pressure control exhaust air system consists of an operating fan and a standby fan to ensure continuous fan availability. Exfiltration air from the ECCS, RCIC and RWCU pump rooms and air from the ECCS corridors is withdrawn by the exhaust fan and exhausted to the plant vent. The RWCU pump room is connected to the ECCS pump rooms by a pipe chase. A differential pressure con-troller modulates the inlet damper in the supply duct to maintain the ECCS CJrridor at a preset negative pressure of (-)l/4 in. of water gage.
?
Isolation of the ECCS, RCIC and RWCU area normal exhaust system is provided by two sets of isolation valves, one set in the ECCS cor-(q~N) ridor ducting and one set in the final leg of the exhaust ducting. All ducting between these valves is ESF quality. Each valve set consists of two valves in series, each of which is closed on signal from the process radiation monitoring system or a LOCA signal. Manual override is also provided. i ( When ventilation air is required in an ECCS pump room, such as during inspection or servicing, a shutoff valve located in the i corridor exhaust ductwork can be closed from a local panel mounted ! switch. Closure of the valve causes exhaust airflow to be diverted through the opened room and out through the room exhaust duct. Supply fan and exhaust fan motor starters are interlocked, so that the supply fan cannot operate unless the exhaust fan is in operation. Discharge air from the exhaust fan is directed to the plant vent during normal conditions. If high airborne radioactivity is v 9.4-27
l GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 0 ' 9.o.3.2.1.2 Auxiliary Building Pressure Control (Continued) l l 1 detected in the exhaust air, an alarm is actuated and the following l sequence of events is automatically initiated: (1) exhaust fan is shut down; (2) supply fan trips out; (3) exhaust air dampers close; (4) supply air dampers close; (5) SGTS starts up; (6) SGTS isolation valves open; and (7) corridor and exhaust duct isolation valves close. Space pressure is maintained negative with respect to ambient by the SGTS. Ventilation air is supplied by the pressure control supply system. Startup is prevented by an interlock, until the exhaust system is operational. Startup is automatic with manual override. Low air-flow causes an airflow switch to activate changeover from the operating to the standby fan. Interlocking damper controls switch the dampers when the fan motors are switched. During the winter heating cycle., inlet air to the pressure control supply air system is sensed by a temperature-indicating switch. When the inlet air temperature falls below 60'F, the switch opens the hot water valve to the coil. A temperature indicating con-troller (located in the ECCS corridor from the unit) modulates face and bypass dampers to maintain the space temperature as specified in Subsection 3.11.1. 9.4-28
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 0 9.4.3.2.1.6 Electrical Switchgear Room Air-Conditioning [)\-- Systems (Continued) air-conditioning unit is the chilled water unit, and the self-contained refrigerant unit is considered standby. Conversely, the chilled water air-conditioning unit in the Division 2 electrical switchgear room is on standby, while the refrigerant unit is oper-ational. These assignments are reversed when the Division 2 Con-trol Building Chilled Water System is operation. Both air-conditioning units located in each electrical switchgear room are connected into a common ductwork system. Backdraft dampers located in the units discharge ductwork prevent airflow through the standby unit. Conditioned air is supplied through ductwork to maintain the space temperature as specified in Subsection 3.11.1. The air is dis-charged from the ductwork through registers or diffusers. Air is returned to the units through a grill installed in the face of the air conditioning unit. Startup of the air conditioners in each switchgear room is by manual pushbutton station in the main control room or by an auto-matic temperature control system that senses the temperature in each area. An airflow switch in the outlet duct of the operating unit initi-ates changeover from the operating to standby unit in case of loss of airflow. Low flow also initiates an alarm in the main control room. In the event of preferred power loss, the unit that was operating again assumes operation once standby power is restored. A temperature-indicating controller with an alarm switch for high temperatures modulates a three-way chilled water valve, when the unit requiring chilled water is operating. The valve mair.tains 9.4-35
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.3.2.1.6 Electrical Switchgear Room Air-Conditioning Systems (Continued) space conditions as specified in Subsection 3.11.1. The self-contained unit has a built-in temperature controller to maintain space conditions. A differential pressure-indicating switch measures the static pressure drop across the filters in the air-conditioning units. An audio / visual alarm is sounded in the main control room in case of high static pressure drop. 9.4.3.2.1.7 Battery Rooms Exhaust System Battery room airflow equals eight changes-per-hour minimum for each battery room. Each battery room is serviced by two 100% fans in parallel. See Figure 9.4-4 (K-16 4) for flow diagram. The battery rooms are ventilated by drawing air in from the elec-trical switchgear rooms and exhausting to atmosphere. The exhaust air is monitored for airborne radioactivity. Two fans in parallel, , one operating and one standby, are employed for this purpose. In-let and outlet backflow dampers are also provided. In addition, backflow dampers are installed in the outlet of each fan to pre-vent backflow through the idle unit. One exhaust fan in the Division 1 Battery Room is on Division 1 power. One fan in each of the Division 2 and Division 4 Battery Rooms is on Division 2 power. The remaining fans are on non-divisional or normal power. Battery room fans are capable of manual start only. The remote pushbutton station is located i1 the main control room. Two stations for each battery room are provided to facilitate manual override of the autocontrol on either fan. A flow switch in the ductwork, common to both fans in each battery room, actuates an 9.4-36
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 (Oj 9.4.3.2.1.7 Battery Rooms Exhaust System (Continued) alarm in the main control room on low flow. Local-start capability is provided by duplicate start stations located in each space. 9.4.3.2.1.8 CRD Maintenance Area Air-Conditioning System Maintenance activities are performed in the CRD maintenance area during normal power plant operations and shutdown. Ventilation must be sufficient for human occupancy as well as for removal of fumes from the cleaning tanks. Air for this purpose is. drawn from the air-conditioning system which serves the Division 2 electrical switchgear room, the cable area and the corridor. Provision is made for obtair.ing air from the system serving the Division 1 electrical switchgear room in emergency. Discharge air () is filtered before its release to the atmosphere. equal to 10 air-changes-per-hour minimum. Air change is The CRD maintenance room area is air conditioned, with air supplied from the cable area, Division 2 electrical switchgear room and corridor air-conditioning system. In case this system is lost, the crossover damper can be opened to obtain air from the system serving the Division 1 electrical switchgear room. l i l The CRD maintenance area exhaust fan draws the air across the cleaning tanks, up through hoods over the tanks, through a pre-filter and high efficiency filter, through the fan and then to
- atmosphere through a backflow damper. The exhaust air is monitored for airborne radioactivity. Manual balancing dampers _
are installed in the hood ducts for balancing the flow over the tanks. Pressure drop through the filters is monitored by differ-ential pressure sensors. The room is maintained at negative dif-() ferential pressure to ensure infiltration of air from the corridor into the CRD maintenance room. To hold the CRD maintenance room 9.4-37
GESSAR II 22A7007 238 NUCLEAR ISLAND R v. 0 9.4.3.2.1.8 CRD Maintenance Area Air-Conditioning System (Continued) negative to the corridors, 10% more air is exhausted than intro-duced through the ductwork. System startup is by remots, pushbutton station in the main control room. 9.4.3.2.1.9 Remote Shutdown Panels Air Conditioning The remote shutdown panel room is used for conducting safe reactor shutdown procedures in case the main control room is not habitable. See Figure 9.4-4 (K-164) for flow diagram. Ventilation air is furnished by the Division 1 electrical switchgear room, corridor and elevator tower air conditioning system. Exhaust air is vented to the corridor area. A hermetic air conditioner and filter are used 4.n the recirculating air-conditioning system. Power for the self-contained air conditioner is obtained from the Division 1 Class lE electrical bus. A self-contained, hermetic air conditioning system is used to cool the remote shutdown panel room. Space temperature is maintained as specified in Subsection 3.11.1. Air is circulated from the room, through a prefilter, through a cooling coil, through the fan and out into the room. Refrigerant is used as the cooling medium in the cooling coil (evaporator) . Condenser cooling is by water from the ESW System. The remote shutdown system possesses manual start capability only. A built-in temperature controller cycles the unit off and on to maintain space temperature, as specified in subsection 3.11.1. Filter static pressure drop is monitored by a local differential pressure in4icator. , O 9.4.38
N GESSAR II' 22A7007 238 NUCLEAR ISLAND Rsv. 15 9.4.3.2.1.10 Electrical Cable Tunnel Smoke Removal System In the event of a fire in the cable tunnel, the smoke removal system clears smoke from the cable tunnel, located in the (-) 6 f t
- l10. in level of the Auxiliary Building.- See Figure 9.4-4 (K-164) for flow diagram. Smoke is drawn from the tunnel by a fan and discharged to atmosphere. No processing of the air is required.
The f an is sized to provide six air-changes-per-hour minimum. Manual starc capability only is provided. Activation of the system is by remote pushbutton station located in the main control room. 9.4.3.2.1.11 Auxiliary Building Smoke Removal System Smoke detectors are located in various areas of the Auxiliary Building to warn the operator of the presence of smoke. The oper-w ator activates the Smoke Removal System from a switch in the main
\s l control room. Status lights inform the operator of the system's operation. Two redundant sets of ductwork and inline fans are ,
provided. Each set evacuates 9,000 cfm from each of the following areas. ECCS Corridor at El. (-) 32 '-0" Corridor Area at El. (-) 6 ' - 10 " Electric Switchgear Rooms at El. 11'-0" HVAC Equipment Rooms at El. 28'-6" The air is exhausted from the Auxiliary Building through louvered roof vents. O V 9.4-39
l GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 9.4.3.2.1.12 Air Positive Seal Compressor Room Cooling System The Air Positive Seal (APS) System prevents leakage of radioactive materials through the outboard valves (or outboard portion of split wedge valves) of lines penetrating containment or secondary con-tainment barriers, as described in Subsection 6.5.3. See Fig-ure 9.4-4 (K-164) for the flow diagram depicting the APS Compressor Room Cooling System features. Heat generated by the compressor is confined to the room until removed by the subject cooling system, which consists of an open-ended fan cooling unit located in the compressor room. The system consists of a cooling coil, coil drain pan and fan. The cooler is designated as Division 2 and takes its power from the Division 2 bus. Cooling water for the coil is Division 2 ESW. Condensate drainage from the coil goes to normal waste. Cooling air in the amount of 1000 cfm is circulated in the room and over the cooling coil by a centrifugal fan. Fan and coil properties and character-istics are described in Table 9.4-5. No ducting is employed. Operation of the cooling system is required only when the compres-sor is operating. Therefore, an interlock assumes simultaneous operation of both compressor and cooling system. Temperature con-trol in the APS compressor room, other than that inherent in the maximum capability of the cooling coil, is not directly exercised. Design temperatures are as specified in Subsection 3.11.1. Instrumentation for the APS Compressor Room Cooling System is limited to pressure and temperature test points located in the ESW line and adjacent to the cooling coil inlet and outlet. Redundancy requirements are satisfied by the redundancy of the APS units, each with its associated cooler unit. O 9.4-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
/ s
( )
\~# 9.4.5.1.2 System Description (Continued) the partition. This air is exhausted to the drywell area. Two supply and two exhaust penetrations through the shield wall are provided, with the supply penetrations 180 apart and the exhaust penetrations at 90 to supply penetrations. See Table 9.4-7 for a description of Reactor Building HVAC System components.
Temperature sensors (thermocouples) are located at strategic points' in the drywell to monitor temperatures. The temperature signals are transmitted to a multipoint recorder located in the main con-trol room for point-by-point readout. An alarm connected to the switch is actuated by a temperature in excess of 160*P from any channel. A thermocouple selector switch parallel to the temper-ature recorder provides input to the temperature-indicating con-troller (TIC). The controller modulates three-way chilled water [\_/ } control valves. Drywell temperatures in excess of 160 F are alarmed in the main control room, alerting the operator to take appropriate corrective action. The TIC readout is located in the main control room, and the chilled water control valves are located outside containment. . 9.4.5.1.3 Safety Evaluation Operation of the drywell cooling system is not a prerequisite to assurance of either of the following: (1) integrity of the reactor coolant pressure boundary, or (2) capability to safely shut down the reactor and to main-f's) (_,/ tain a safe shutdown condition. 9.4-47
GESSAR II 22A7007 283 NUCLEAR ISLAND Rev. 15 9.4.5.1.3 Safety Evaluation (Continued) However, the system does incorporate features that provide reli-ability over the full range of normal plant operation. These features include the installation of redundant principal system components such as: (1) electric power; (2) fan coil units; (3) sources of chilled water; (4) ductwork; (5) controls; and (6) cross connection of all fan coil units. O 9.4-47a
]
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/~N, t $
\/ 9.4.5.1.4 Inspection and Testing Requirements Equipment design includes provisions for periodic testing of functional performance and inspection for system reliability.
Standby components are fitted with test connections so that system effectiveness, except for airflow or static pressure, can be veri-fied without the units being on line. Test connections are pro-vided in the discharge air ducts for verifying calibration of the operating controls. 9.4.5.1.5 Instrumentation Application Drywell cooling unit function is manually controlled from indi-vidual remote pushbutton stations in the main control room. Selec-tion of the operating units is made in accordance with Subsec-tion 9.4.5.1.2. A time-delay relay allows flow switches of each unit to engage in sequence. Airflow failure, sensed at any of the {']' N- operating units, initiates an alarm in the main control room. 9.4.5.2 Drywell Purge System 9.4.5.2.1 Design Bases l 9.4.5.2.1.1 Safety Design Bases All purge control valves shall be series-redundant and designed to Seismic Category I, and Safety Class 2, requirements to ensure the l integrity of the drywell pressure boundary. Purge is initiated only when reactor pressure is less than 150 psia. 9.4.5.2.1.2 Power Generation Design Bases l l l The drywell purge system shall be capable of reducing the level of l [} airborne radioactivity for inspection, maintenance and refueling
after a shutdown.
9.4-48
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 t 9.4.5.2.2 System Description As stated in Subsection 9.4.5.2.1.2, the purge system functions only during plant shutdown. Therefore, this discussion presumes shutdown conditions. See Figures 9.4-6 and 9.4-7 (K-165A&B) and 9.4-8 (K-168) for flow diagrams related to the purge systems. The drywell purge system is a single-stage system, used during and after plant shutdown, as long as airborne radioactivity is at a level precluding personnel access to the drywell. The purge sys-tem can be used during refueling, and, if the air is not radioac- _ tive, it is discharged to the plant vent. During refueling, the drywell may be purged up to 5.5 times the drywell volume per hour. During other operations, the purge rate is one air change per hour, p The drywell purge system and the containment low purge system, usededuring the shutdown procedure, consist of the pressure control supply and the SGTS. The SGTS is described in Sub-section 6.5.1. O 9.4-49
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.5.2.2 System Description (Continued) The containment high purge system, used during post-shutdown, consists of the high purge supply and the high purge exhaust fans, supply units, duct work and controls. _
~
The pressure control supply system is composed of two parallel units to satisfy redundancy requirements. Each unit consists of a pre-o filter, a high efficiency filter, face and bypass dampers, a heat-ing coil, a cooling coil, a coil drain pan and a centrifugal fan. Backdraft dampers in the discharge ducting of each unit prevent backflow of air in the standby unit. One backdraft damper servec both units in the common inlet ducting. Control dampers are installed in both inlets, and a single control damper is installed in the common discharge duct. The damper in the discharge duct is used to control containment pressure (at (-)l/4 in wg) . Startup of the pressure contr:>l supply is from a remote manual pushbutton station located in the main control room. A separate station is provided for each unit. Automatic changeover from operating to standby unit is provided by flow switches in each discharge line that sense airflow. The cooling coils are served by Reactor Island chilled water [ Figure 9.2-9b (K-125B)]. The heating coils are served by the heated water distribution system [ Figure 9.2-10a (K-127A]. t Air from the pressure control supply is directed through a con-tainment penetration which is protected by air-operated butterfly valves on each side of the containment boundary for containment isolation. The conditioned air is then distributed to the con-tainment area through registers equipped with balancing dampers. Circulation between drywell and containment areas in then accom-plished through the Drywell Containment Mixing System described in Subsection 6.2. 5.2. 3. Subsequent to the mixing phase, the air is exhausted by and through the SGTS described in Subsection 6.5.1. 9.4-50
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 1 O b '9.4.5.4.2 System Description The drywell vacuum relief system is composed of two parallel flow paths routed through ESF ducting [ Figure 9.4-8 (K-168)]. Each path contains two valves in series. The inboard valve is air-operated, while the outboard valve is a check. Rapid response to possible negative pressure transients in the drywell is provided by the air operator. Both inboard and outboard valves are located in the containment space. A differential pressure-sensing element pressure switch provides the signal for valve actuation. Allowable pressure is as shown in Subsection 3.11.1. Control of positive pressure fluctuations in the drywell during startup is provided by the bleedoff vent system shown in Fig-ure 9.4-8 (K-168). A 2-in. line connects the drywell and shield annulus air volumes. The line is routed from the drywell, into fys the Auxiliary Building and back into the Shield Building annulus. The two drywell bleed valves in the containment are powered for separate divisions of the Class lE buses. One valve is powered from the Division 2 bus, and one valve is powered from the Division 3 bus. One of the valves outside the containment is powered from the Division 1 ESF bus, and the other valve is a
~
manual, locked closed valve. The Division 1 and Division 2 valves l are used for normal drywell bleedoff. Division 1, 2 and 3 valves receive an isolation signal when a LOCA occurs. The Division 1, 2 _ and 3 valves can be key-lock bypassed to enable the system to per-form as a post-LOCA vent purge backup to the hydrogen recombiner. I n v 9.4-55
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 9.4.5.4.3 Safety Evaluation Operation of the drywell vacuum relief and drywell bleedoff vent systems is not required to assure any of the following conditions: (1) integrity of the reactor coolant pressure boundary, or (2) capability to safely shut down the reactor and maintain it in a safe shutdown condition. However, these systems do incorporate features that ensure reli- 3 able operation when required. Redundant isolation valves in parallel series protect against malfunction from any single active component failure. Opening of the drywell bleed line sooner than 10 days af ter LOCA is not permitted. 9.4.5.4.4 Inspection and Testing Requirements O The vacuum relief and bleedoff vent systems are inspected period-ically to assure that operating equipment and controls are func-tioning properly. Standby components are periodically tested to assure that the standby equipment is operational. 9.4.5.4.5 Instrumentation Application Two differential pressure sensors, one for each of the two drywell vacuum relief systems, measure the pressure differential between the drywell and containment. When containment pressure exceeds drywell pressure by 2.0 psid, the transmitter signals the air operator to open the vacuum relief valve. The valve is also opened if containment pressure exceeds drywell pressure by 0.2 psid and if a high drywell pressure LOCA signal has occurred. Closure
]
of the valve is signaled when differential pressure returns to 0.0 psid or higher, on the drywell side. 9.4-56
GESSAR II 22A7007 238' NUCLEAR ISLAND Rev. 15 -> 9.4.5;5 . Containment Cooling and Dome Recirculation System [. 9.4.5.5.l' Design Bases L 9.4.5.5.1.l' Safety Design Bases (1) Containment cooling is accomplished using the air hand- 3 {- ling and purge units. Seismic Category I isolation ) 2 valves (a part of the containment isolation system) at each containment and Auxiliary Building penetration pre-
- clude the' possibility of escape of airborne radioactiv-ity in the event of a postulated LOCA.
i (2) Radiation detection equipment in the containment cooling ~ and exhaust system is of Seismic Category I design. _
- (~N 9'4.5.5.1.2
. Power Generation Design Bases
.(1) The system is designed to control the. thermodynamic and radiological environment, such that under nonaccident.
, conditions,-the environment does not damage or appre-l ciably reduce the service capability of the equipment in'the containment. Also, this system shall ensure the j safety and comfort of operating and maintenance personnel. (2) Sufficient redundancy is designed into the system to
- l. ' ensure reliable operation during normal power plant operation.
i' .(3) The system is designed to facilitate periodic inspection
- f. and testing of the principal system components.
l (4) The system has the. capability to maintain the contain-ment temperature, as specified in Subsection 3.11.1, f during normal operation and in the various modes of
- operation specified.
9.4-57 ,
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.5.5.2 System Description See Figures 9.4-6 and 9.4-7 (K-165A&B) for flow diagrams depicting the containment cooling system. The system is designed to maintain containment atmosphere at 90 F and 50% to 60% relative humidity during normal power plant operation. Design capacity also includes the capability of holding a maximum temperature of 120*F at 90% rolative humidity during certain operational conditions, such as hot standby. Recirculating f ans draw the hot air from the dome area and main-tain a pattern of air circulation to prevent pocketing of hot air. The containment cooling system consists of two parallel subsystems, each containing three fan coil units. During normal operation, two fan coil units in each subsystem are operational, while the remaining single unit in each subsystem is on standby. Conditioned air is thus normally discharged from four fan coil units, two to a subsystem. Each subsystem distributes air throughout a 180* segment of the containment. Failure of the operating unit initi-ates an alarm in the main control room. The standby unit in the respective subsystem is started manually from the control room. Cooling water is supplied from the Reactor Island Chilled Water System [ Figure 9.2-9c (K-125C)]. The cooling water circulates at a pressure higher than that in the containment space. _ Air discharged from the fan coil units passes through an electric-powered reheat coil for humidity control, before being distributed to the containment space. Each of two branches of the ducting, one for the upper region and one for the lower region, is served by its own reheat coil. Each fan coil unit consists of a cooling coil, a coil drain pan and a centrifugal fan. A backflow damper in each f an outlet 9.4-58
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 () 9.4.5.5.5 Instrument Application (Continued) Individual cooling units control temperatures in the RWCU valve room and the RWCU heat exchanger room. Each unit consists of a cooling coil, a coil-drain pan and a fan. Temperature control is . by means'of a temperature-indicating controller in each room which measures fan discharge air temperature. High room temperature is sensed by a temperature switch, which activates an alarm in the main control room. Unit startup is by remote manual station in the main control room. Controller'and alarm setpoints are as indicated in Subsection 3.11.1. Air change in the amount of 100 cfm infiltrates and is removed by the pressure control exhaust system. The POC detectors are part of the Fire Protection System. 9.4.5.6 Containment Pressure Control and Purge System O See the changes in the response to Question 6.28 pertaining to m the purge penetration size and arrangement. 9.4.5.6.1 Design Bases 9.4.5.6.1.1 Safety Design Bases (1) The containment pressure control system shall supply tempered outdoor air for ventilation purposes and main-tain the containment at less than atmospheric pressure. (2) 'The containment pressure control supply and exhaust and the high flow purge supply and exhaust penetrations - shall be provided with redundant Seismic Category I, ASME, Section III, Class 2 isolation valves at each containment penetration. Radiation detection equipment, () which provides the isolation signal, is also designed to Seismic Category I. 9.4-61
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 9.4.5.6.1.1 Safety Design Bases (Continued) (3) The drywell purge supply and exhaust penetrations shall be provided with a reliable means of automatic isolation. (4) The containment pressure control and purge system shall be designed to maintain containment radiation release below dose limits defined by 10CFR100 and 10CFR50, Appendix I. 9.4.5.6.1.2 Power Generation Design Basis (1) The high flow purge supply and exhaust system may be used for high flow purge of the containment during refueling. (2) The containment pressure control system is designed to permit periodic inspection of its principal components. h 9.4.5.6.2 System Description The containment pressure control system consists of a supply air system and an exhaust air system. [See Figure 9.4-6 and 9.4-7 (K-165A&B) for flow diagram.) These systems, including certain ductwork and isolation valves, are shared with the drywell purge system. The containment pressure control and purge system uses ( the hardware during normal and standby operation, while the drywell l purge system uses the hardware during shutdown. The only dif-ference between these modes is valve alignment to accommodate the l differences between the two operational modes. See Subsec-tion 9.4.5.2 for : description of the hardware. The pressure control supply air system consists of an operating and a standby fan coil unit. One hundred percent outside air is l l 9.4-62
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 (3 q_/ 9.4.5.6.2 System Description (Continued) drawn through a prefilter, face and bypass dampers, a heating and cooling coil section and a fan section. The fan supplies air to , the containment through ductwork. A damper in the supply air duct-work is modulated by a pressure controller to maintain the contain-ment at a negative pressure with respect to atmosphere. Air is distributed in the containment space through ductwork and registers of the double-deflection type. Isolation valves, to maintain the containment integrity, are installed inside and outside containment, where the supply and exhaust air ducts penetrate. Inboard of the inside isolation valve - on each duct is a locked, closed butterfly valve. A bypass loop - containing an isolation valve is installed around both butterfly valves. _ b i ,/ s The bypass loop including the isolation valve is sized to limit the flow rate to 5000 cfm during normal operations. The pressure control exhaust air system consists of an operating fan, a standby fan and a provision for a nonsafety-related charcoal filter for use during normal operation. The exhaust fan draws a constant flow of air from the general containment space and from the RWCU rooms. Air drawn from the RWCU filter deminera-I lizer room is room infiltration air only. Air exhausted from the general containment and from the RWCU rooms is filtered and moni-tored for radioactivity. If radioactivity is detected, the containment isoJation valves are closed by a monitor signal to prevent sign.tficant radioactive release. The exhaust fan is - l l stopped automatically when the isolation valves are closed, b , w_) l 9.4-63 9
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.4.5.6.2 System Description (Continued) The containment pressure control supply fan is interlocked with the containment pressure control exhaust fan to prevent the supply f an from operating when the exhaust fan is shut down. The containment supply air fan can be operated manually during the drywell purge cycle. Outside air intake to the pressure control supply air units con-tains air-operated shutoff dampers. The dampers are interlocked to close when the associated unit is shut down. The damper pre-vents outdoor air from entering the coil section when the fan is shut down. Containment high purge uses the same components as the drywell high purge system. See Subsection 9.4.5.2 for a description of hardware. 9.4.5.6.3 Safety Evaluation Operation of the containment cooling, ventilation and pressure control, and low purge systems is not a prerequisite to assure the following conditions: (1) integrity of the reactor coolant pressure boundary, or (2) capability to safety shut down the reactor and maintain it in a safe shutdown condition. However, these systems do incorporate features that provide reliability over the full range of normal plant operations. A radiation monitoring system is provided to detect high radiation O 9.4-64
,. GESSAR II 22A7007.
- 238 NUCLEAR ISLAND Rev. 15 ,
Radwaste Building CoStrol' Room and Unit Substation
~
9.4.6.2.1 Room (Continued) - x and/or cooling requirements. -The air-conditioningEsystem isCa I . - unit airiconditioner consisting'of a water-cooled' condenser, com-presscr, cooling coil,- heating ' coil, filters and f an. Outdoor air l and recirculating air (are mixed ~and drawn through a' prefilter, a high efficiency filter, a heating coil, and a cooling coil. The
. air is supplied through ductwork, toithe control room .and the' unit -
l substation ' tx) maintain the design conditions similar to those specified in Secti'on 6.4. A-pressure differential' controller regulates the exfiltration from'the control room to maintain it at a positive static. pressure, preventing airborne contamination from entering. ' - t
~
j ' The exhaust air system consists of an -exhaust fan. Exhaust air from'the Control Room and Unit Substation is monitored for air-() borne radioactivity before exhausting to the atmosphere. , 9.4.6.2.2l Radwaste Building HVAC Control System ,
~\
4 The HVAC control system for the remainder of the Radwahte Building
-is'a once-through type. Outdoor air is filtered, tempered and delivered tx) the noncontaminated areas of the building. Theysupply l air. system consists of a prefilter and high efficiency filter, ,
t- heating coil, cooling coil and .two 100% . supply fans. One fan is normally operating and the other fan is on standby. The supply f an - furnishes conditioned air through ductwork and diffusers , or (. registers, to the work- areas of the building. Zone preheat coils l
' installed in the supply air ductwork provide temperature control.
f Air from the work areas is exhausted through the tank and pump ! rooms. Thus, the overall airflow pattern is from the least poten-
- i. .
- tially contaminated areas to the'most contaminated areas.
J The exhaust air system consists- of two exhaust fans, one normally operating and one on standby. Exhaust airlfrom - the " silo, waste j ,
- -9.4-75
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
, 9.4.6.2.2 Radwaste Building HVAC Control System (Continued) filter rooms, oil separator room and the mixing and filling station is monitored for airborne radioactivity. Under normal conditions with no contamination, normal ventilation in the same circuit as the other spaces in the building (except for the control room and -
unit substation room) is maintained. Each of the above-noted spaces is separately monitored. A high level of radioactivity activates an alarm in the main control room, simultaneously isolating the affected space. Infiltration air in the affected space or spaces is exhausted to atmosphere after being cleaned up in a filter train censisting of a prefilter, HEPA filter, activ-ated charcoal filter and a second HEPA filter. The resulting clean air is exhausted to atmosphere by a centrifugal fan. The HVAC equipment room is cooled in summer by supply air fans that blow air through the equipment room and out through wall louvers. The equipment room is heated in winter by hot water unit heaters. The' unit substation room is cooled and heated by the same system that conditions the control room. 9.4.6.3 Safety Evaluation Although the HVAC System is not safety-related as defined in Section 3.2, several features are provided to insure safe opera-tion. A completely separate HVAC System is provided for the con-trol room and unit substation room. Pressure control fans for radwaste areas are redundant, with provision for automatic start of the standby unit. Radiation detectors and isolation dampers are provided to permit isolation and containment of any radio-active leakage. System failure analysis is presented in
- -Table 9.4-10.
9.4.6.4 Tests and Inspections The system is designed to permit periodic inspection of important components, such as fans, motors, belts, coils, filters, ductwork, s piping and valves, to assure the integrity and capability of the 9.4-76
GESSAR II 22A7007 ! 238 NUCLEAk ISLAND R2v. 15 9.4.6.4 Tests and Inspections (Continued) ; system. Local display and/or indicating devices are provided for periodic , inspection of vital parameters suchrassroom temperature, and test' connections are provided in exhaust filter trains and piping for periodic checking of air and wateriflows'for conform-ance to the design requirements. Portable test and monitoring equipment is available to balance the system when required. HEPA filters are tested by using dioctyl phthalate (DOP) aerosol. Charcoal filters are periodically tested with freon for bypasses. 9.4.6.5 Instrumentation Application 9.4.6.5.1 Control Room and Unit Substation Room r' The air-conditioning unit for the cdntrol room is started k-s manually. A temperature indicating controller modulates the air. conditioning system via a three-way hot water valve to maintain _ space conditions. A differential pressure indicating controller modulates dampers in the return air ductwork and the room damper to maintain the positive static rocra pressure. Differential pressure indicators measure the pressure drop across the prefilter bank and the high efficiency. filter bank. A differential pressure-indicating switch measures the' pressure drop across the complete 1 filter banks. The switch actustes an alarm if the pressure exceeds the preset limit. 9.4.6.5.2 Radwaste Work Areas The air exhaust and supply fans for the Radwaste Building are started manually. The fitn inlet dampers open when the fan is started. A flow switch installed in the exhaust fan discharge 3 duct actuates an alarm on' indication of fan failure in the main (\-s and radwaste control rooma,. Ventilation of the Radwaste Building 9.4-77
L GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 0 9.4.6.5.2 Radwaste Work Areas (Continued) must be restarted manually. The exhaust fan is interlocked with the supply fan to prevent the supply fan from operating if the exhaust fan is shut down. Two pressure-indicating controller switches are located in the work space at El. (-) 27 f t 10 in. and at El. (-)6 ft 10 in. One con-troller is normally functioning while the other is on standby. A three-way valve in the output line from the controllers is used to transfer from the operating to the standby controller. The controller modulates variable inlet damper vanes in the supply fan to maintain the area at a negative static pressure with respect to atmosphere. The switch causes an alarm to be actuated if the ncgative pressure falls below the preset limit. Temperature-indicating controllers located in the work areas at each floor elevation and in the oil separator room control and modulate three-way hot water valves to maintain work area tempera-ture requirements. A temperature-indicating controller senses the outdoor air tem-perature. When that temperature falls below the preset limit, the controller opens the hat water valve to the heating coil. A temperature controller installed in the supply air ductwork modulates face and bypass dampers to maintain winter supply air temperature. A temperature-indicating controller is located in the cupply air duct modulating the three-way chilled water valve to maintain summer design air temperature requirements. Differential pressure indicators measure the pressure drop across the prefilter section and the high efficiency filter section. A differential pressure-indicating switch measures the pressure drop 9.4-78
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
) 9.4.7.2 System Description 9.4.7.2.1 Division 1 and 2 Diesel-Generator Buildings Divisions 1 and 2 Diesel-Generator Buildings are identical except for their power bus designations. Thus, for descriptive purposes, they are herein treated as identical. See Figure 9.4-lla for (K-170A) flow diagram. Major equipment is tabulated in Table 9.4-11.
9.4.7.2.1.1 Diesel Generator Rooms The Heating and Ventilating System for each engine room consists of three supply fans, three exhaust fans and two electric unit heaters. one supply fan is interlocked to automatically start when the ~ O- diesel-generator starts. This fan draws air from the outside as well as from the engine room. A backdraft damper in the recir- ; culation air inlet is weighted for a setting of 1/8 in, wg above the suction pressure in the fresh air inlet duct. This establishes the mix ratio of fresh to recirculated air. All ducting associ-ated with the fan is ESF quality to withstand the conditions that cause diesel-generator startup. Fan location is in the mechanical equipment room. 4 The other two supply fans, also located in the mechanical equip-ment room, start up automatically and sequentially, one at 80*F and one at 90*F. They both take inlet air from outside through } the mechanical equipment room and distribute the air ~to the Engine Room and electrical equipment area through ESF quality ducting. Each fan circuit is provided with a backdraft damper to prevent backflow through idle fans. The fan taking air from the outside is equipped with a motorized damper for isolation. O 9.4-83
GESSAR II 22A7007 238 MUCLEAR ISLAND Rav. 15 9.4.7.2.1.1 Diesel Generator Rooms (Continued) Two electric unit heaters are located in the generator room for space heating. These units heat and circulate air within the engine room. Two more electric unit heaters are located in the mechanical equipment room to heat that air. The latter are activated when the room air temperature drops to the value indi-cated in Figures 9.4-lla (K-170A) and 9.4-11b (K-170B). The supply fans are operated only during warm weather, when the heaters are not in operation. Three exhaust fans, located in the engine room, remove building air to the atmosphere. Two of the fans are interlocked to start simultaneously with their respective supply fans. The other is ] started from a remote-manual station in the Main Control Room. All three exhaust fans discharge into a common exhaust plenum through backdraft dampers and a tornado damper. 3 O All fans, both supply and exhaust, are interlocked with the CO 2
~
Fire Protection System to shut down and isolate the building so that a CO 2 blanket is established. - 9.4.7.2.1.2 Day Tank Room One non-Safety Grade fan is used to exhaust air from the day tank room. The air is discharged to the exhaust plenum through a backdraft damper and a tornado damper. Supply air for the day 3 tank room ventilation enters via a backdraft damper from the engine room. Fan startup is by local-manual push button station with an interlock to shut the fan down in case of activation of the CO2 ] blanketing system. Air is al so drawn from the valve pit by the day tank room exhaust fan. ] O 9.4-84
GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 15 O a-4 () 9.4.7.2.2.3 Battery Room (Continued) room through a reversed backdraft damper, which allows only infiltration air to enter the battery room. This also meters the j airflow, so the positige pressure in the switchgear room can be maintained. l Battery room discharge air is routed out of its building through ductwork that terminates in a backflow damper, which, in turn, t admits the air to the exhaust plenum through a tornado damper. ] . All ducting for the battery room exhaust system is ESF quality. Each fan is capable of ventilating the battery room at the rate of i eight air changes /hr. f. 9.4.7.2.2.4 Mechanical Equipment Room F Ventilation for the Division 3 mechanical equipment room is the same as for-Divisions 1 and 2, except for the addition of summer / i winter and summer vent systems units. These are described in i Subsection 9.4.7.2.2.2. r 9.4.7.2.2.5 Day Tank Room Day tank ~ room ventilation is the same for the Division 3 build-ing, as it is for the-Divisions 1 and 2 buildings (Subsec-tion 9.4.7.2.1.2). 9.4.7.3 Safety Evaluation The diesel-generator rooms, ESF electrical switchgear rooms and battery rooms ventilating and cooling systems described in this
~
section are designed.to the requirements specified in Section 3.2. The systems are connected to their corresponding division Class lE O' bus and are operable.after loss of offsite power supply. The t ' 9.4-87
- _ _ . . - . _ . _ . _ _ _ . ~ _ - _ _ . _ _ _ . . _ . _ . .. _ . _ . _ . - _ _ _ _ _ _ _ _ _ _ . - - _ .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.7.3 Safety Evaluation (Continued) diesel-generator rooms, ESP electrical switchgear rooms and battery rooms ventilation systems failure analysis is presented in Table 9.4-12. 9.4.7.4 Tests and Inspections Provisions are made for periodic tests of the safety-related ventilation fans. These tests include determination of control set ] points and signals, alarm functioning, airflow rates, damper functioning and airflow switch operation. Thermal performance of the heaters is also testable. The balance of the system can be proven operable by its performance ] during normal operation. Standby equipment can be tested to ensure its readiness upon demand. O 9.4.7.5 Instrumentation Application 9.4.7.5.1 Diesel-Generator Rooms one supply fan (recirculation fan) is interlocked with the diesel engine starting system, so that the diesel engine start signal also closes the fan motor starter contacts. However, the fan does not start until power generated by the diesel generator is avail- } able. Delay time and the method of delay are described in Chap-ter 8. A temperature-indicating controller modulates the inlet dampers to maintain space temperature. This same controller also activates a high-temperature alarm if the space temperature exceeds 122 F. The associated exhaust fan is manually started from a con-trol station located in the main control room. The damper control point is 70*F. Both fans are interlocked with the CO2 fire fight-ing system, so that the fans shut down and the dampers close if the CO2 system is activated. 9.4-88
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 ( ) 9.4.7.5.1 Diesel-Generator Rooms (Continued) The two sets of supply and exhaust fans are electrically inter-locked to allow starting of each set when required. One set starts at 80'F and the other starts at 90*F. Sensors for these controls are located in the engine room. Remote-manual override from the main control room is provided so that the fans can be started or stopped at any time. An interlock with the CO 2 system causes the fans to be shut down when the CO2 blanketing system is activated. Electric unit heaters are provided for space heating yearround. Temperature indicating controllers sample space temperature and turn on the heaters when the temperature drops to 60*F. s The fume control fan is started manually from a control station (s,) located in the main control room and is run continuously during engine-operation. A temperature indicating-controller starts and modulates the electric heater when the discharge duct air tempera-ture drops to 50*F. The fume control fan is shut off and the inlet damper is closed by air interlock with the CO2 system, when CO2 blanketing is initiated. An alarm is sounded by the tempera-ture controller if the discharge air temperature drops below 40*F. The day tank room exhaust fans are started manually from local push button stations and are also electrically interlocked with the CO2 fire fighting system to shut down when the CO2 system is activated. 9.4.7.5.2 Switchgear Rooms The Division 3 switchgear. room summer / winter vent system is started
-' from a remote-manual pushbutton station located in the main control \-/ room. -No interlock with the CO2 system is provided. Discharge 9.4-89 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.4.7.5.2 Switchgear Rooms (Continued) air temperature control is provided by a temperature-indicating control set to turn the heater on at 50*F. An alarm is sounded if the temperature drops below 40*F. An interlock on the fan motor starter controls the inlet dampers. Startup of the summer vent system is the same as for the summer / winter vent system, except that no heater is provided. An electric unit heater in the switchgear room cycles to maintain space temperature at 60'F. _ The battery room, which receives ventilation air from the switch-gear room, is exhausted by a parallel set of exhaust fans. The operating fan is started from a manual control station located in the main control room. An airflow sensor located in the discharge duct sounds an alarm in the main control room upon lack of airflow. The standby fan is then started manually from the main control room. O 9.4-90
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 , I3 Table 9.4-1
CONTROL BUILDING HEATING, VENTILATING AND AIR-CONDITIONING SYSTEM COMPONENT DESCRIPTION (Continued)
Unit Heaters Division 1 Air Handling Room Unit Heater X93-BB010 & X93-BB012 Quantity 2-50% Capacity Air Flow (scfm) Pressure Drop (in WG) Power Input (kW) 30 Brake Horsepower (hp) Motor Horsepower (hp) Division 2 Air Handling Room Unit Heater X93-BB0ll & X93-BB013 Quantity 2-50% Capacity I Air Flow (scfm)
- Pressure Drop (in WG)
<% Power Input (kW) 30 g ,) Brake Horsepower (hp)
- Motor Horsepower (hp)
Filter Room Unit Heater X93-BB014 & X93-BB015 Quantity 2-100% Capacity 1 Air Flow (scfm)
- Pressure Drop (in WG)
Power Input (kW) 4 Brake Horsepower (hp)
- Motor Horsepower (hp)
- Applicant to Supply O
b 9.4-95
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. O Table 9.4-2 CONTROL BUILDING HVAC SYSTEM SYSTEM FAILURE ANALYSIS Failure Mode Effect on System Detection Remarks Lose a radiation No direct effect Radiation Redundant system is monitoring alarm in started system Control Room Lose a component No direct effect Computer Redundant system FSL status starts in Control automatically Room Radiation Rapid response Radiation Close normal inlets, enters intake alarm in start cleanup system duct Control automatically Room Smoke in No direct effect Smoke Operator switches Control Building alarm in to smoke removal Control mode Room High temperature No immediate Personnel Operator investi-in a room effect detection gates, resets room heater Low temperature No immediate Temper- Operator investi-in main air- effect ature gates, and may conditioning Alarm - start redundant duct Low in system Control Room Low pressure in Possible infil- Pressure Operator starts Control Building tration of Alarm - redundant system unmonitored air Low in Control Room O 9.4-96
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 7 ( ,) 9.5.1.2.3 Codes and Standards (Continued) 10CFR50 - Licensing of Production and Utilization Facilities UL - Underwriters Laboratories Approved Equipment Lists FM - Factory Mutual Approved Materials and Equip-ment Lists ANI - Basic Fire Protection for Nuclear Power Plants, March 1976 ASTM D992 Classification of Flammability Standards ASTM E84 - Method of Test of Surface Burning Character-istics of Building Materials NFPA 10-73 - Portable Fire Extinguishers - Installation ] NFPA 10A-73 - Portable Fire Extinguishers - Maintenance and Use NFPA 12-73 - Carbon Dioxide Systems NFPA 13-73 - Sprinkler Systems
- NFPA 14-73 -
Standpipe and Hose Systems x,,) NFPA 24-73 - Outside Protection NFPA 37-73 - Stationary Combustion Engines and Gas Turbines NFPA 70-73 - National Electric Code NFPA 72D-73 - Proprietary Signaling Systems NFPA 78-73 - Lightning Protection Code
- NFPA 80-73 -
Fire Doors and Windows NFPA 80A-73 - Protection from Exposure Fires l NFPA 90A-73 - Air Conditioning and Ventilating Systems l NFPA 91-73 - Blower and Exhaust Systems NFPA 101-73 - Life Safety Code NFPA 1963 Fire Hose Couplings, Screw Threads NFPA 1961 Fire Hose _ NFPA 251-73 - Fire Test, Building Construction and Materials f NFPA 252-73 - Building Materials, Tests of Surface Burning l Characteristics-NFPA 321-73 - Classification of Flammable Liquids l NFPA 801-73 - Facilities Handling Radioactive Materials () NFPA 802-73 - Nuclear Reactors 9.5-5
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 0 9.5.1.2.3 Codes and Standards (Continued) NRC Regulatory - Housekeeping Requirements for Water-Cooled Guide 1.39 Nuclear Power Plants BTP-APCSB - Appendix A, Guidelines for Fire Protection 9.5-1 For Nuclear Power Plants Docketed Prior to July 1, 1976 9.5.1.2.4 Protection of Operating Units The applicant is responsible for the protection of operating units during construction of additional units. 9.5.1.2.5 General Description of Fire Protection Systems The Nuclear Island buildings are provided with wet standpipe sys-tems, automatic wet pipe sprinkler systems in selected areas, carbon dioxide systems for the Diesel-Generator buildings, Halon 1301 systems in the PGCC modules, ABC multipurpose and Halon 1211 hand extinguishers. Alarm systems, both manual and automatic, are provided in all areas of the plant as passive systems. They alarm without control-ling an extinguishing function. Appendix 9A provides a detailed description and drawings for each area. 1 9.5.1.2.6 Protection and Extinguishing Equipment for Safety-Related Equipment The extinguishing, detection and protection systems for safety-related equipment are identified in Appendix 9A by room or area. 9.5.1.2.7 Design Features of Fire Detection and Suppression Systems Fire detection for all buildings, except the Diesel-Generator Buildings, is provided by the ionization-type product of combustion 9.5-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 O A normally open second pilot valve shall be provided for each line supplying CO2 to the local hazard valve operator. The second pilot valve shall automatically close and be. latched if there is an extended discharge of CO2 due to the failure of the first pilot
' solenoid valve. The second pilot valve shall not prevent succes-sive individual manual or electromanual actuations by the Operator.
A pneumatic actuated time delay relay shall control the second pilot valve. An extended discharge alarm shall be provided as part of the controls for the second pilot valve which shall be supervised by the Fire Protection System. The controls for the second pilot valve should be independent of the controls for the first pilot valve. , f lO l i i i i i O 9.5-8a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 15 () 9.5.3.1.2 Standby Lighting The AC standby lighting systems are fed from Class lE buses _ through separate lighting panels. Fixtures are provided for all safety-related areas (areas where division 1, 2, and 3 system equipment are located) such as areas used during shutdown and accidents, and access areas. The fixtures provide a reduced ]~ lighting level adequate to support personnel movement and observa-tion of equipment after interruption of the normal lighting system. In the event of a LOPP, the standby lighting system is automatic-ally fed from the diesel-generator sets. (See Subsection 8.3.1.1.3 for bus transfer during loss of normal preferred and/or alternate preferred power.) The standby lighting transformers and panels are seismically qualified to keep their structural integrity and stability during and after an SSE. Standby lighting cables up to the lighting 7_ (_, panels are classified as Class lE circuits and are routed in $ Circuits from the lighting panels e seismic Category I raceways. to the individual fixtures are wired and routed by the Applicant. All standby lighting fixtures are seismically supported. _ 9.5.3.2.3 Emergency Lighting I DC Emergency lighting fixtures are installed for stairways, exit l routes and major control areas such as the main control room and y l remote shutdown panel area (all safety-related areas). Each of ] *m. I i the emergency lighting fixtures has two incandescent sealed-beam lamps, a self-contained battery, charger and an initiating switch l whichenergizesthefixturefromthebatteryintheeventofloss] of the AC power supply, and de-energizes the fixture upen return l I of AC power. The power supply AC source is fed from the standby ( g3 lighting system. t ) e..
%J 9.5-21
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 9.5.3.2.3 Emergency Lighting (Continued) In addition to the above, DC emergency lighting fixtures are pro-vided for control rooms, switchgear rooms, diesel-generator
- m buildings and fuel handling areas. The fixtures are connected to y DC distribution panels, which are switched by contactor. The initiating signal is the loss of voltage to a standby AC lighting distribution pancl.
The emergency lighting fixtures provide backup illumination for periods after the loss of preferred power and until the diesel-generators energize the standby lighting systems, as well as in the event of loss of all the AC lighting sources. DC Emergency lighting fixtures are installed for stairways, exit routeu and major control areas such as the main control room and remote shutdown panel area. Each of the emergency lighting fix-tures has two incandescent sealed-beam lamps, a self-contained battery, charger and an initiating switch which energizes the fixture from the battery in the event of loss of the AC power supply, and de-energizes the fixture upon return of AC power. The power supply AC source is fed from the standby lighting system. The self-contained emergency lighting sets are seismically quali-
- i ed to keep their scructural integrity and stability during and $
after an SSE. _ In addition to the above, DC emergency lighting fixtures are pro-vided for control rooms, switchgear rooms and fuel handling areas. The fixtures are fed from the DC distribution panels DC-E and DC-F. ~ The feeder breakers are normally open. The loss of voltage to the standby AC lighting panels will initiate a signal to close the DC g Emergency panels feeder breakers. Cables feeding the DC emergency m panels are Class lE circuits and routed in seismic Category I raceways. 9.5-21a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 14 (h
\-) 9.5.3.3 Inspection and Testing Requirements (Continued) emergency lighting systems will be inspected and tested periodically (as determined by the applicant) to ensure operability of lights and switching circuits.
9.5.4 Diesel-Generator Fuel Oil Storage and Transfer System 9.5.4.1 Design Bases 9.5.4.1.1 Safety Design Bases (1) Each engine shall be supplied by a separate Diesel-Generator Fuel Oil St'orage and Transfer System. All fuel oil transfer equipment shall be designed, fabricated and qualified to Seismic Category I requirements. Failure of any one component could result in loss of fuel supply ("
\~ to only one diesel-generator.
(2) Minimum onsite storage capacity of the system shall be sufficient for operating each diesel-generator for a minimum of seven days while supplying post-LOCA maximum load demands. l l (3) Design and construction of the Diesel-Generator Fuel Oil Storage and Transfer System shall conform to the IEEE Criteria for Class lE Power Systems for Nuclear Power l Generating Stations (IEEE-308); and ASME Code, Section III, Class 3, Quality Group C. Miscellaneous equipment shall conform to applicable standards of NEMA, DEMA, m ASTM, IEEE, ANSI, API, NFPA. ANSI Standard N195 " Fuel Oil Systems for Standby Diesel Generators" shall be used. l (4) The Diesel-Generator Fuel Oil Storage and Transfer shall l(; In addition, the system be of Seismic Category I design. shall be protected from damage by flying debris carried l 9.5-23 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9.5.4.1.1 Safety Design Bases (Continued) by tornados and hurricanes, from external floods, and other environmental factors. (5) System components shall be corrosion resistant. (6) System design shall provide positive protection from damage caused by turbine missiles. 9.5.4.1.2 Power Generation Design Basis The Diesel-Generator Systems are standby systems. The Diesel Fuel Oil Storage and Transfer Systems shall be capable of supporting the instant start requirements of the diesel-generators. The diesel o engine fuel consumption shall not exceed 0.38 pound of fuel per 9 net horsepower hour. 9.5.4.2 System Description There are two Diesel-Generator Buildings. One building houses the Division 1 unit and one houses the Division 2 and 3 units. The Division 1 and 2 units are identical and both are held in reserve to furnish standby AC power in case of an emergency. The Division 3 diesel generator, housed in the same building, but separated from the Division 2 diesel generator, furnishes standby AC power to the HPCS System under emergency conditions (see Subsection 8.3.1 for details). The Diesel-Generator Fuel Oil System (Figures 9. 5-10 and 9.5-11) interfaces with the yard storage and transfer system for the fuel supply for sustained running. (Applicant to provide a description of yard storage and transfer functions.) 9.5-24
~
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 ( ,,) 9.5.4.4 Tests and Inspection (Continued) Fuel oil may normally be stored by a minimum of six months without deterioration. 9.5.4.5 Instrumentation Application I Fuel supply level in the day tanks is indicated both locally and in the main control room. Also, alarms on the local diesel-generator panel annunciate low level and high level in the day tanks. The setting of the low level alarm shall provide fuel at least 60 minutes of DG operation at 100 percent load with 10 percent margin between the alarm and the suction line inlet. A group repeat trouble alarm is also provided in the main control room. Level switches in the day tank signal automatic start and stop of the fuel oil transfer pump. _ n v 9.5.5 Diesel-Generator Cooling Water System 9.5.5.1 Design Bases All essential components of the Diesel-Generator Cooling Water System shall be qualified to Seismic category I requirements and to 10CFR50, Appendix B. All pumps, valves, tanks, piping and heat g exchangers shall be designed in accordance with ASME Code, Section III, Class 3, Quality Group C. Failure of the cooling system in any - one engine shall not affect the readiness or operability of any other engine. The cooling system shall derive from a reliable source, one not affected by a LOPP, the plant Essential Service
. Water (ESW) System. Divisions 1, 2 and 3 diesel-generators are y
located in Seismic Cabegory I structures, protected from tornado- _) generated missiles and flood waters. The Jacket Water Cooling System shall be able to operate at full load for 7 days without any make-up. ~ w 9 m 9.5-27
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 9.5.5.2 System Description Each diesel-generator unit is supplied with a complete closed loop cooling system mounted integrally with the enaine generator pack-age. See Figures 9.5-12 (K-136) and 9. 5- 137) for pertinent flow diagrams. Included in each cooling , _ge are an expansion tank, temperature-regulating valve, lube oil cooler, immersion heater, jacketed manifold and a heat exchanger which is furnished with ESW from the essential portion of the system. ESW supply is from the same division as that of the diesel generator served. - The jacket cooling water passes through a three-way temperature control valve which modulates the flow of water through and/or around Lhe jacket water heat exchangers, as necessary, to maintain required water temperature. Jacket water cools the turbocharger, the governor, the air intercooler, the exhaust manifold and the lube oil heat exchanger. The t'aree-way valve, whose service is crucial, is designed and qualified as stated in Subsection 9.5.5.1. An electric heater is installed in each system for the purpose of keeping the engine jacket water at a temperature near the normal operating level during plant normal operation. The heater water is circulated through the engine to assure temperature uniformity in the engine. Two jacket water circulating pumps are provided to circulate the cooling water through the system during diesel- _ generator operation. Dmring the standby mode, the jacket water temperature is maintained at 120 F based on 60 F normal ambient 9 m temperature. To prevent long-term deterioration of the system internal surfaces, the system is filled with high quality treated water from the Demin-eralized Water System. (See Subsection 9. 2.3 for water quality details.) The ESW side of the system (see Subsection 9.2.1 for water quality details) is designed with the appropriate corrosion allowances on piping, and a fouling factor of 0.002 for heat exchanger tubes. A long interval periodic cleaning of the heat 9.5-28
. . - - =- .- . .. -
O O O . Table 9.5-4 WSP FLOW AND PRESSURE REQUIREMENTS BOP PSIG at GPM - Building Interface BOP Interface Standpipe Sprinklers Total Remarks FIRE WATER OPERATION AND TEST Radwaste P28 Operation 128 500 400 900 Test 131 500 0 500 Div 2 & 3 D/G P29 Operation 131 500 0 500 Test 131 500 0 500 Control P31 Operation 131 500 180 680 [ Test 131 500 0 500 00 Div 1 D/G P32 Operation 131 500 0 500 $o f Test 131 500 0 500 y Fuel and Auxiliary P30 Operation 128 500 500 1000 >$ WW Buildings Test 131 500 0 500 HH Reactor Building - Operation 250 500 750 (1) Future to H . Fuel, Auxiliary and P99 operation (2) 131* 750 500 1250 (2) In FB & AB h g Reactor Building Test (2) 131 750 0 750 (1) Future U Operation (3) 131* 500 500 1000 (3) In RB Test (3) 131 750 0 750 (1) Future ESW OPEPATION Radwaste - - - - - - Div 2 & 3 D/G Div 2 Operation 85 150 0 150 Control Div 1 Operation 95 150 0 150 Div 1 D/G Div 1 Operation 85 75 0 75 $U
<>-J Fuel and Auxiliary Div 1 Operation 95 150 0 150 Building e$
Reactor Building Div 1 Operation 95 150 0 150
- Note: If the future flows are added, the pressure at the BOP interface will be reduced to 128 psig.
Table 9.5-4 WSP FLOW AND PRESSURE REQUIREMENTS (Continued) BOP PSIG at GPM Building Interface BOP Interface Standpipe Sprinklers Total Remarks CONDENSATE OPERATION AND TEST Reactor Building P86 Operation 130 500 0 (4)S00 (4) Not a total Test 134 500 0 (4)500 flow for P86 NOTE: Test data are for standpipe; sprinkler test is not controlling. w LJ o3 Z h Gn I >> xx HH Ch H Z C
%M O
< >M H O wa i j O O O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Ch ( ,) 9A.4.1.ll Building - Auxiliary: El (-)6 ft 10 in. (Continued) (b) provision of raised foundation pads for the equipment; (c)' provision of floor drainage; and (d) Seismic Category I standpipe (rupture unlikely). (12) Fire Containment or Inhibiting Methods Employed: (a) The function is located in a separate fire-resistive enclosure. (b) The means of fire detection, suppression and alarming are provided and accessible. O N/ (13) Remarks - Provisions for core cooling capability are discussed in Subsection 9A.2.5. The pump lube oil is contained in an integral reservoir and has no exposed piping. 9A.4.1.12 Building - Auxiliary: El (-)6 ft 10 in. 9A.4.1.12.1 Zone 1 (1) Space - corr.idor A-06-25, Zone 1 (2) Equipment: Provides Safety- Core Related Cooling Remote shutdown panel (see marks) Yes,D1 Yes , MSIV leakage control panel Yes,D1 No 9A.4-35
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9A.4.1.12.1 Zone 1 (Continued) _ (3) Radioactive Material Present - none. (4) Qualifications of Fire Barriers - Two walls are 3-hr fire-resistive concrete walls, one of which separates the corridor and room areas. There is a 3-hr fire-resistive metal stud-gypsum wall along the cable tunnel. Zone 1 and Zone 2 are separated by a 3-hr fire-resistive metal stud-gypsum wall. The ceiling and floor are 3-hr fire-resistive concrete. Interior walls contain 3-hr fire-resistive A-label doors to the rooms. The exterior wall contains 3-hr fire-resistive A-label doors. The Division 2 cable tunnel smoke vent which passes through the area is of 3-hr fire-resistive construction. _ (5) Combustible Present Fire Loading Total Heat of Combustion (Btu) 6 320 ft of divisional and 433.2 x 10 nondivisional cable trays - containing 9.25 lb/ft of XLPE-FR cable insulation (6) Detection Provided - Class A supervised POC throughout the corridor and manual alarm pull stations at Col. D-12 in the corridor. , (7) Suppression Available: Location Actuation Wet pipe automatic All Automatic sprinklers, ordinary hazard, minimum density 0.15 gpm/ ft2 per head, 100 ft2 per head, waterflow alarm at the Control Room 9A.4-36
GESSAR'II 22A7007 238 NUCLEAR ISLAND Rev. 15. , s 9A.4.1.12.1 Zone 1 (Continued) , J Location Actuation Modified Class III,' Seismic Category I, standpipe and Col.-H-9, D-13 Manual i
)
~ hose reel ,
Manual ABC and Halon 1211 hand extinguishers Col. H-7, D-13
} 5 (8) Fire Protection Design Criteria Employed: i
). , n
'(a) The function-is located in a separate fire-resistive ;
enclosure. -; i , (b) Fire detection and suppression capability provided and accessible. ,. (c) ' Fire-stops are provided for cable tray and piping penetrations. t f [- (9)' Consequences of Fire - Theipostulated fire assumes the - loss of the functions. The Division 1 MSIV leakage control panel is backed up by the Division 2 panel in a separate fire area. The remote shutdown panel is , located in a separate 1-hr fire-resistive structure ~ for security reasons. The. provision of core cooling backup capability is. described in Subsection 9A.2.5. I l i (10) Consequences of Fire Suppression - Suppression extin-() guishes the fire. Water could-pool on the floor.
., c .
l . l
.9A.4-37 i
- - . , - - . . ~ . _ , _ _ ~ . _ . - _ ~ ..;___ - . _ - - _ . . . . . . . ~ _ . . _ _ . - .--
GESSAR II 22A7007 238 NUCLEAR ISLAND R:v. 15 9A.4.1.12.1 Zone 1 (Continued) }h Protection against flooding is provided by equipment on raised pads and by floor drainage through the sump system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System: (a) location of the manual suppression systems on the perimeter of the area; (b) provision of raised foundation pads for the equipment; (c) provision of floor drainage; and (d) Seismic Category I standpipe (rupture unlikely). O (12) Fire Containment or Inhibiting Methods Employed: (a) The area is located in a separate fire-resistive enclosure. (b) The means of fire detection, automatic and manual suppression and alarming are provided and accessible. (c) Fire-stops are provided for cable tray and piping penetrations. (13) Remarks - The corridor provides access to other plant areas. The introduction of combustibles during normal operations and maintenance periods is expected. The h l 9A.4-38
. , :/
.t l ,
j GESSAR II* ' 22A7007
!238" NUCLEAR ISLAND Rsv. 15
s ,
9A.4.1.12.1 -Zone 1 (Continued) f f Applicant is responsible for limiting)the anticipated transitory combustible loading for this area to a load not greater than can be protected by,the ordinary hazard, _ automatic sprinkler system. j.
)
; i. ihU area contains electrical cables in trays. Cable kt. '
insulation in trays is discussed in Subsection 9A.3.4. 9A.4.1.12.2 Zone 2
- (l) Space - corridor Zone 2 (2) Equipment: <> 'I
()) n , Provides
, Safety- Core
, ! 2 Related Cooling MSIV leakage control panel Yes,D2 No
~
RWCU valve corridor isolation valves Yes,Dl,D2 No J (3) Radioactive Material Presen't - non'e. 4 (4) Qualifications of Fire Barriers' - Two walls are 3-hr fire-resistive concrete walls, one of which separates the corridor and room /a'reas. There is a 3-hr fire-resistivemetalstud-gyhsumwallalongthecable tunnel. The ceiling; and floor are 3-hr fire-resistive _ concrete. Interior walls contaik 3-hr fire-resistive
,[ A-label doors to the rooms. The exterior wall contains " ^
3-hr fire-resistive A-label doors. The CRD Maintenance Rhom_is separated from the corridor with 1 1/2-hr
!y ,3-label fire-resistive doors. ,
n , (' [ 9A'.4-39 '
- ??o ,
i A___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ .___ _... _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ . . _ _ _ __.._____.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9A.4.1.12.2 Zone 2 (Continued) (5) Combustible Present Fire Loading Total Heat of Combustion (Btu) 6 32 ft of divisional and 432.2 x 10 nondivisional cable trays containing 9.25 lb/ft of XLPE-FR cable insulation (6) Detection Provided - Class A supervised POC throughout the corridor and manual alarm pull stations at Col. F-1 and J-6 in the corridor. (7) Suppression Available: Location Actuation Wet pipe automatic All Automatic sprinklers, ordinary hazard, minimum density 0.15 gpm/ ft2 per head, 100 ft2 per head, waterflow alarm at the Control Room Modified Class III, Seismic Col. H-2, H-9 Manual Category I, standpipe and hose reel ABC and Halon 1211 hand Col. F-1, H-3 Manual extinguishers (8) Fire Protection Design Criteria Employed: (a) The function is located in a separate fire-resistive enclosure. (b) Fire detection and suppression capability provided and accessible. 9A.4-39a
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
'\
x - 9A.4.1.12.2 Zone 2 (Continued) (c) Fire-stops are provided for cable tray and piping penetrations. (9) Consequences of Fire - The postulated fire assumes the loss of the functions. The Division 2 MSIV leakage control panel is backed up by the Division 1 panel in a separate fire area. The remote shutdown panel is located in a separate 1-hr fire-resistive structure for security reasons. The provision of core cooling backup capability is described in Subsection 9A.2.5. The redundant safety-related power operated valves are separated by a minimum distance of 3 ft 0 in. so that adequate spatial separation is maintained between signal and power cables to the redundant valves.
~ - (10) Consequences of Fire Suppression - Suppression extin-guishes the fire. Water could pool on the floor.
Protection against flooding is provided by equipment on raised pads and by floor drainage through the sump system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System: (a) loca' tion of the manual suppression systems on the perimeter of the area; (b) provision of raised foundation pads for the equipment; () (c) provision of floor drainage; and (d) Seismic Category I standpipe (rupture unlikely). i j 9A.4-39b
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 15 l Zone 2 (Continued) 9A.4.1.12.2 (12) Fire Containment or Inhibiting Methods Employed: (a) The area is located in a separate fire-resistive enclosure. (b) The means of fire detection, automatic and manual suppression and alarming are provided and accessible. (c) Fire-stops are provided for cable tray and piping penetrations. (13) Remarks - The corridor provides access to other plant areas. The introduction of combustibles during normal operations and maintenance periods is expected. The Applicant is responsible for limiting the anticipated transitory combustible loading for this area to a load not greater than can be protected by the ordinary hazard, automatic sprinkler system. Door A-16-25 is built as a 3-hr fire-resistive door. However, other criteria pro-hibit affixing a label. The area contains electrical cables in trays. Cable insulation in trays is discussed in Subsection 9A.3.4. 9A.4.1.13 Building - Auxiliary: El (-)6 ft 10 in. (1) Space - CRD Maintenance Room A-06-9. 9A.4-39c
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 O 9A.4.1.13 Building - Auxiliary: El (-)6 ft 10 in. (Continued) (2) Equipment:
- Safety- Provides Related- Core Cooling Tables and miscellaneous tools for No No performing CRD maintenance HVAC fan nondivisional No No Leak test booster pump No No Flush tank No No CRD ultrasonic cleaner and tank No No (3) Radioactive Material Present - storage and maintenance of contaminated CRDs.
O (4) Qualifications of Fire Barriers - The walls, floor.and ceiling are 2-hour fire-resistive concrete. The doors are 1 1/2-hr fire-resistive B-label. i
- \s /
9A.4-39d
GESSAR II 2?A7007 238 NUCLEAR ISLAND Rnv. 0 9A.4.1.13 Building - Auxiliary: El (-) 6 ft 10 in. (Continued) (5) Combustibles Present: Fire Loading Total Heat of Combustfon None by design. May contain - small quantities of combustible materials necessary for maintenance activities. (6) Detection Provided - Class A supervised POC in the room and a manual alarm pull station at Col. F-1 in the corridor. (7) Suppression Available: Location Actuation Modified Class III, Seismic Col. H-3 Manual Category I, standpipe and hose reel ABC and Halon 1211 hand Col. F-2, F-1 Manual extengu,ishers Ordinary hazard wet pipe All Automatic sprinklers, having a water density of 0.15 gpm/ft 2 and a coverage of 100 ft 2 per head ('s) Fire Protection Design Criteria Employed: (a) The function is located in a separate fire-resistive area. (b) Fire detection and suppression capability is pro-vided and accessible. O 9A.4-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. O hy ,/ 9 A . 4 . l'.13 Building - Auxiliary: El (-) 6 f t 10 in. (Continued) (9) Consequences of Fire - The postulated fire assumes loss of the function. There is no effect on safety-related systems. Fire and smoke could involve low-level con-tamination cleanup. (10) Consequences of Fire Suppression - Suppression extin-guishes the fire. Water could pool on the floor. Pro-tection against flooding is provided by equipment on raised pads and by floor drainage through the sump system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System:
's ,/. (a) location of the manual suppression systems external to the room; (b) provision of floor drainage; and (c) Seismic Category I standpipe (rupture unlikely).
(12) Fire Containment or Inhibiting Methods Employed: (a) The function is located in a separate fire-resistive enclosure. ( (b) The means of fire detection, suppression and alarming are provided and accessible. (13) Remarks - none. O 9A.4-41
i GESSAR II 22A7007 I 238 NUCLEAR ISLAND Rev. 15 1 I 9A.4.1.14 Building - Auxiliary: El (-)6 ft 10 in. (1) Space - Division 2&3 cable tunnel A-06-24. (2) Equipment: Safety- Provides Related Core Cooling None. This space contains Yes,D2,D3 Yes only Division 2 electrical cables in solid-bottom, solid-covered trays. Division 3 electrical cables are in con-duit, embedded in a concrete duct bank in the tunnel along the building exterior wall. Division 2 remote shutdown Yes,D2,D1 Yes panel . (3) Radioactive Material Present - none. (4) Qualifications of Fire Barriers - Two walls are 3-hr fire-resistive concrete, two walls are 3-hr fire-resistive metal stud / gypsum. The floor and ceiling are 3-hr fire-resistive concrete. A single 3-hr A-label door is pro-vided at Col. J-6. The Division 3 cable cabinet at Col. J-12 is 3-hr fire-resistive construction with 3-hr A-label fire-resistive doors. (5) Combustibles Present: Fire Loading Total Heat of Combustion (Btu) 1224 ft of divisional cable 159 x 10 6 trays containing 9.25 lb/ft of XLPE-FR cable insulation (6) Detection Provided - Class A supervised POC is provided for the area and for the inside of Division 3 cable cabi-net. A manual alarm pull station is located at Col. J-6. 9A.4-42
f GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 9A.5.6 Carbon Dioxide Storage (Continued)
- f. After initial discharge, a second discharge for the largest single hazard area must be maintained in the storage tank. Therefore, the Applicant must maintain.a minimum of 11,200 lb of CO I#
2 Diesel-Generator Building fire protection. f In the event of malfunction of the' automatic sequencing for CO 2 discharge to a hazard area, manual activation of the discharge sequence is provided in the control room. 4 9A.5.7 HVAC Systems The majority of the HVAC systems are provided with fire dampers where the duct' penetrates a fire-resistive wall; however, there are some exceptions.- There are some cases where divisional control l valves are in the same fire area. These cases are presented, and 5 - the justification and/or effect on the plant operation relative to reactor safe shutdown is presented, i
-9A.5.7.1 Control' Building The smoke removal systems for the cable rooms and control room are a function of damper arrangement, utilizing the existing air con- -
ditioning system. .The cable room tunnel exhaust ducts are not provided'with fire-dampers. Since separate smoke removal systems are provided for Division 1 and Division-2 areas, a fire in cither ,,,
- . cable room would not preclude safe shutdown utilizing the other f' -division. For a fire in the control room the remote shutdown panel is
- available. The cable-rooms are provided with automatic wet pipe _
a . sprinklers and POC detectors. The cable trays are solid bottom, ! covered metal trays. A postulated electrically initiated XLPE-FR l cable insulation fire in a closed tray or PGCC would evolve little s smoke or heat. The anticipated transitory combustible load, a func- , l tion-of-the Applicant's fire safety program, is expected to be negligible. Inlet ducts are equipped with fire dampers to prevent 9A.5-23
~ - __ -._ _ _. _. , _ _ _ _ . . _ . _ . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 9A.5.7.1 Control Building (Continued) hot smoke or gases from entering the areas from fire sources exterior to the areas. 9A.5.7.2 Division 2 Cable Tunnel - Auxiliary Building The Division 2 cable tunnel is provided with a dedicated smoke removal system. This system is not fitted with a fire damper. The duct which traverses a Division 1 area is of 3-hr fire-resistive construction. POC detection is provided. The cable , trays are solid bottom, covered metal trays. A postulated elec-trically initiated XLPE-FR cable insulation fire in a closed tray would involve little smoke or heat. The anticipated transitory combustible load, a function of the Applicant's fire safety pro-gram, is expected to be negligible. Inlet and exhaust ducts from the normal ventilation system are fitted with fire dampers to pre-vent hot smoke or gases from entering the tunnel via fire sources h exterior to the tunnel. 9A.5.7.3 SGTS Exhaust Stack - Fuel Building The SGTS exhaust stack begins at the (-)5 ft 3 in. level of the Fuel Building and extends through the roof of the building. There are no fire dampers in this stack; however, fire stops are pro-vided where the stack penetrates a fire-resistive floor. The stack is fabricated of 3/8-in. steel plate and is 18 in. in diameter. Since the exhaust gases that enter the stack pass through a charcoal filter bed equipped with water sprays that pre-clude a high temperature condition, and the stack must function to maintain safe plant conditions, fire dampers are not necessary or desirable. The functionality of t he SGTS exhaust stack has no effect on the ability to accomplish safe shutdown. The exhaust 9A.5-24
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& (-) 6 '10" - Reactor, 1 Aux. & Fuel Buildings 9A.5-39/9A.5-40
GESSAR II 22A7007 238 NUCLEAR' ISLAND Rev. 15 f~ (,,h) 11.3.2.20 Seismic Design Equipment and components used to collect, process, or store gaseous radioactive waste are classified as non-Seismic Category I. Conservative analyses presented in Section 15.7 demonstrate that equipment failure will not result in doses exceeding the 0.5-Rem guidelines of Regulatory Guide 1.29.
-The support elements, including the skirts, legs and anchor bolt-ing, for the charcoal adsorber tanks of the offgas system are designed as follows:
(1) the fundamental frequency of the charcoal adsorber tanks including the support elements is greater than 33 Hertz; (2) the charcoal adsorber tanks are mounted on the base mat of the building housing the tanks; () (3) the charcoal adsorber tanks including the support ele-ments are designed with a horizontal static coefficient of 0.15 g; and (4) the stress levels in the support elements of the charcoal adsorber tanks shall not exceed 1.33 times the allowable stress levels permitted by the AISC Manual of Steel Construction, Seventh Edition, 1970.
~
The Applicant will design the vault containing the charcoal tanks to protect the tanks from the effects of structural failure and to resist the OBE as specified in Regulatory Guide 1.143, Section 5'. 2 , as' permitted in Section 5.3. These seismic requirements comply with the requirements of
-Regulatory Guide 1.143. _
11.3 4 21 Quality Control A program will be established that is sufficient to assure that the design, construction, and testing requirements are met. The follow-ing areas will be included in the program: 11.3-15
f GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 11.3.2.21 Quality Control (Continued) (1) design and procurement document control - procedures will be established to ensure that requirements are specified and included in design and procurement documents and that deviations therefrom are controlled; (2) control of purchased material, equipment, and services - procedures will be established to assure that purchased material, equipment, and construction services conform to the procurement documents; (3) inspection - a program for inspection of activities affecting quality will be established and executed by or for the organization performing the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity; O (4) handling, s to rage , and shipping - procedures will be established to control the handling, storage, shipping, cleaning, and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration; (5) inspection, test, and operating status - procedures will be established to provide for the identifications of items which have satisfactorily passed required inspec-tions and tests; and (6) corrective action - procedures will be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defec-tive material and equipment, and nonconformances are promptly identified and corrected, 11.3-16 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 4 (O_,) 11.3.4.6 Estimated Doses (Continued) Tables 12.2-22 and 12.2-23. The expected. total annual release of each radionuclide is 1/4 of the sum of the design basis values for all systems in these tables. Again, site-specific meteorologicaldata are not available. However,
-5 3 (corresponding use of the enveloping X/0 value of 5 x 10 sec/m to an annual whole-body dose equal to the Appendix I to 10CFR50 ,
objective of 5 mrem) provides enveloping estimates of annual average effluent concentrations. The enveloping concentrations based on this X/0 are compared with the concentration limits of Appendix B (Table II, Column 1) to 10CFR20 and Table 11.3-8. When meteorological data becomes available, the Applicant will perform site-specific evaluations of effluent concentrations and off-site doses for comparison with the-design objectives of Appendix I to
- 10CFR50 and the limits of 10CFR20.
x_/ 11.3.4.7 Treated (Delayed) Radioactive Gas Sources The estimated annual release of noble gases during normal operation is given in Table 11.3-1. Values are given in pCi/sec/nuclide and Ci/ year /nuclide. As reviewed in Subsection 11.3.2, iodine and [ particulate releases are essentially zero, i 11.3.5 References I i 1. Browning, W. E., et al., Removal of Fission Product Gases i from Reactor Offgas Streams by Adsorption, June 11, 1959, (ORNL) CF59-6-47. t l
- 2. Miller, C. W., Experimental and Operational Confirmation i
of Offgas System Design Parameters, January 1973,
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NEDO-10751 (General Electric Company Proprietary). ('"N i s-m ll.3-26a- ] t
l GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 15 l 11.3.5 References (Continued)
- 3. Siegwarth, D. P., Measurement of Dynamic Adsorption Coefficients for Noble Gases on Activated Carbon, 12th l
AEC Air Cleaning Conference. l
- 4. Standards for Steam Surface Condensers, Sixth Edition, Heat Exchange Institute, New York, NY (1970). )
- 5. Underhill, Dwight, et al., Design of Fission Gas Holdup l l Systems, Proceedings of the Eleventh AEC Air Cleaning Conference, 1970, p. 217. J
- 6. Nesbitt, L. B., Design Basis for New Gas Systems, July
! 1971, NEDE-lll46 (General Electric Company Proprietary). 1 l
- 7. N66 SJAE Offgas Treatment System - Amendment 1 (supple-ments Licensing Topical Report, August 1978 (NEDE-21056-1-P) (Proprietary).
I l 9 l \ ll.3-26b
GESSAR II 238 NUCLEAR ISLAND h0 (D V 11.4 SOLID WASTE MANAGEMENT SYSTEM The solid waste management system is designed to provide solidifi-cation and packaging for radioactive wastes produced during shut-down, startup, and normal plant operation and to store these wastes until they are shipped offsite for burial. The system is located in the radwaste building. 11.4.1 Design Bases 11.4.1.1 Power Generation Design Bases The solid waste management system provides the capability for solidifying and packaging wastes from the reactor water cleanup g system, the fuel pool cooling and cleanup system, suppression . H pool cleanup system, the liquid radwaste system, resins, and H (N particulate wastes from the condensate cleanup system. Wastes
~s from these systems will consist of spent resin, evaporator bot-toms, diatomaceous earth, and other filtering media.
The solid waste management system also provides a means of com-pacting and packaging miscellaneous dry radioactive materials, such as paper, rags, contaminated clothing, gloves, and shoe coverings and for packaging contaminated metallic materials and incompres-sible solid objectives such as small tools and equipment parts. The solid waste management system is designed so that failure or maintenance of any frequently used component shall not impair sys-tem or plant operation. Storage is provided ahead of process units to allow hold-up in case of delay for maintenance. Drum capping and sample retrieval are performed locally. The
> operating philosophy of the solid radwaste control system is manual start and automatic stop with all functions interlocked to provide
'} a fail-safe mode of operation.
11.4-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 11.4.1.2 Codes and Standards Codes and standards applicable to the solid waste management sys-tem are listed in Table 11.3-6. The solid waste management system is designed and constructed in accordance with quality group D. The solid waste management system components and the structure housing the components are designed to the seismic criteria of Regulatory Guide 1.143. Exterior walls and foundation are seismic Category I; remaining walls and slabs are not. Collection, solidification, packaging, and storage of radioactive wastes will be performed to maintain any potential radiation exposure to plant personnel as low as is reasonably achievable (ALARA) in accordance with the guidelines of Regulatory Guide 8.8 ) and within the dose limits of 10CFR20. Some of the design fea-tures incorporated to maintain ALARA criteria include remote system operation and remotely acruated spraying, flushing, and equipment layout that permits shielding of components containing ll radioactive materials. Proportional amounts of wastes and fixative are incorporated into the solid waste matrix to assure that no free water accumulates in the waste container in compliance with the Branch Technical Posi-tion ETSB 11-3. Packaging and transporting radioactive wastes will be in conform-ance with 10CFR71. Packaged wastes will be shipped in conformance with 49CFR173 dose limits. Additionally, the Applicant will demonstrate to the satisfaction of the NRC staff that the current storage space for solidified " wet" solid waste and the correspond-ing 23 containers (Figure 1.2-14) are adequate or will increase the storage to accommodate at least 28 containers of solidified wet solid wastes for at least 30 days prior to their shipment to a licensed offsite burial site. 11.4-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 11.4.1.2 Codes and Standards (Continued) The solids handling portion of the radwaste system is designed to meet the requirements of 10CFR50 Appendix A by being capable of
- a handling to solid waste generated by all of the occurrences of a the liquid waste system. Additionally, the radiation monitoring .
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V- N . U Table 11.5-1 PROCESS AND EFFLUENT RADIATION MONITORING SYSTEMS Sample Line Upscale Setpoint Monitored No. of Detector or Detector- Channel Warning Incation Range Alarm Trip Scale Process Channels Type A. . Safety-Related Systems Main steamline 4 Gamma Immediately. 1-106 mR/nr above full technical 6 dec. log sensitive downstream of power specification ionization plant main steam- background, chamber line isolation below trip. valve
.0.01 above technical ,4 dec. log Containment !!VAC 4 Geiger-Muller Exhaust duct upstream of to background, specification tubo e thaust venti- 100 mR/hr below trip lation isolation N w
valve co above ' technical 4 dec. log g Auxiliary building 2 Geiger-Muller Exhaust duct 0.01 background, specification ag tube upstream of to g HVAC 100 mR/hr below trip g p3 exhaust venti- ~ g lation. isolation t* U2 M D3 L. valve >>
- above technical 4 dec. log NN W control building 6 Geiger-Muller irtake duct 0.01 H upstream of to background, specification gg HVAC tube below trip, g3 p intake venti- 100 mR/hr 'M lation isolation ;W valve Z at quarterly- technical 4 dec. log U Fuel building HVAC 4 Geiger-Muller Sample line 0.01 to 100 mR/hr tech spec specification tube level 0.01 to above technical 4 dec. log.
Shield annulus HVAC 2 Geiger-Muller -Exhaust duct tube upstream of 100 mR/hr background, specification exhaust vent below trip isolation valve above technical 4 dec. log Standby gas 2 Geiger-Muller SGTS exhaust air 0.01 to specification treatment system tube duct downstream- 100 mR/hr background, of exhaust and below trip heat removal fans and dampers 0.01 to above tecnnical 4 dec. log N Containment space 4 Geiger-Muller locally in upper WN 100 mR/nr background, specification refuel mode tube contaimaent area - below trip Y
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Table 11.5-1 PROCESS AND EFPLUENT RADIATION MONITORING SYSTEMS (Continued) Sample Line lysea l . . t ! ~ s nt Monitored No. of Detector or Detector Channel warnino Process Cnannels Type Location Range Alarm c_r ig 4 4 Afstems_ Required for Plant Operation Liquid radwaste 1 Scintillation Sample lino 10 to 10 6 above t e c hn ic.21 effluent counts / min background, specificatirm below trip 10 to 106
~
Closed cooling 1 Scintillation Sample line above not ; dec. log water system counts / min background applicable Essential service 2 Scintillation Sample line 10 to 106 above not 5 dec. log water system, RIIR counts / min background applicable N Offgas post-treat 2 Geiger-Muller Sample line 10 to 106 above technical 5 dec. log w tubes counts / min background, specification 00 below trip Z Offgas pre-treat 1 Geiger-Muller Sample line 1 to 106 at tech spec not 6 dec. log CO g tubes mR/hr report level applicable hg 6 td CD n Carbon bed vault 1 Geige.:-Muller Carbon bed 1 to 10 above not 6 dec. log y> I tube vault mR/hr background applicable pc sc w HH N Plant vent discharge 1 Geiger-Muller Sample line 10 to 106 at quarterly technical 5 dec. log MH tubes counts / min tech spec specification level h z Radwaste building 8 Geiger-Muller Exhaust ducts 0.01 to above technical 4 dec. log O veut tubes 100 mR/hr background, specification below trip Offgas vent 1 Geiger-Muller Sample line 10 to 106 above technical 5 dec. log tube counts / min background specification 6 Service water 1 Scintillation Sample line JO to 10 above technical 5 dec. log effluent counts / min background specification Battery room 1 Geiger-Muller Exhaust duct 0.01 to above not 4 dec. log exhaust tube 100 mR/hr background applicable CRD maintenance 1 Geiger Muller Exhaust duct 0.01 to above not 4 dec. log area tube 100 mR/hr background applicable A3 M Radwaste building 1 Geiger-Muller Exhaust duct 0.01 to above not 4 dec. log (D M control room and tube 100 mR/hr background applicable <> unit substation j
-o U1 -J I I O 9 9
\_/ v/ v/
. Table 11.5-2 PROCESS RADIATION MONITORING SYSTEM (GASEOUS AND AIRBORNE MONITORS)
Principle Radionuclides Monitor Configuration Type Sensitivity Range Measured Expected Activity Alarms & Trips 3 Offgas posttreat- Offline Part. 0.25 cpm /pci/cm 101 - 106 X -85 (a) Offgas activity Low flow ment radiation Filter epm defined in INOP monitor Iodine Table 11.3-6 High Filter High-IIigh Gas G-M Offgas radiation Adjacent to G-M 100 _ 106 Noble gas Offgas activity INOP/Iow monitor, pretreat- sample mR/hr fission defined in High ment gas - log chamber products Table 11.1-1 High-High scale Main Steamline Adjacent y-lon 3 x 10-10 100 _ 106 Coolant Steamline activity INOP Radiation Monitor amp /R/h to steam chamber mR/hr activation defined in High lines gases Table 11.1-5 High-Ilign m Iow Z CO H Carbon bed Offline G-M 100 - 106 Noble gas Carbon bed inventory Low OM ,P vault mR/hr defined in Table Hjgn h$, g yy b w Offgas vent exhaust radiation Offline Part filter 250 cpm /pci/cm 3 131 - 106 com Kr-05(a) Of fgas activity defined in Low flow INOP HH monitor iodino Table 11.3-3. high CA H filter high-high GM { g Containment off1ine Gil 0.01-100 INOP/high ventilation mR/hr
- low monitor isolate air Plant vent Offline Part. 250 epm /pci/cm3 101 - 106 Kr-85 (a) go,gygy elevated discharge filter epm
- INOP radiation monitor iodine high filter high-high Gtt Radwaste building Offline GM 0.01-100 Kr-8 5 (a ) Negligible activity Low ventilation ruR/hr discussed-in INOP exhaust radiation Section 11.3 High monitor Af bJ Mechanical m>
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LaJ Exhaust mR/hr NN h CRD Maintenance Offline GM 0.01-100 *
- H Area Radwaste Bldg Offline GM 0.01-100 * *
(Control Room mR/hr U and Unit Substation
- Applicant to provide O ha
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. O f O 12.3.3.2.7 Radwaste Building (Continued) in the first zone is maintained slightly above atmospheric, while the air pressure in the second zone is maintained slightly below atmospheric. Air in the second zone is drawn from outside the . building and distributed 1x) various work areas within the building. Air flows from the work areas to' the tank and pump rooms and is then discharged via the radwaste vent. An alarm sounds in the ! control room if the exhaust fan fails. The exhaust flow is monitored for radioactivity, and if a high activity level is detected, the potentially radioactive cells are automatically isolated, but airflow through the work areas continues.
\
If the vent stack high-radiation alarm continues to annunciate after the tank and pump rooms are isolated, the work area branch exhaust ducts are selectively manually is.olated to locate the
/ involved building area. Should this technique fail, because the airborne radiation has spread throughout the building, the control room air conditioning continues, but the air conditioning for the balance of the building is shut down.
The silo, waste filter room, oil separator room, and the mixing
-and filling station exhaust air is drawn.through a filter unit consisting of a particulate filter, a~HEPA filter, a charcoal filter, and then another-HEPA filter, before being discharged to the atmosphere. The air is monitored for radioactivity, and if a high level is detected, supply and exhaust is terminated.
Maintenance provisions for the filters are similar to those for the SGTS and Control Building HVAC System. See Subsection 9.4.6 for a detailed discussion of the Radwaste Building HVAC System. 12.3-47
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 12.3.3.2.8 Diesel-Generator Buildings The Diesel-Generator Buildings do not contain sources of radio-activity, nor is access required during an event involving activity release from another building. Therefore, no radiation protection measures have been taken in these buildings. 12.3.3.2.9 Turbine Building Environmental System (Applicant to supply.) 12.3.3.2.10 Central Service Facility (Applicant to supply.) 12.3.4 Area Radiation and Airborne Radioactivity Monitors
~
The location of airborne radioactivity and area radiation N monitors are given in Subsection 19.3.12.3. ]_d 12.3.4.1 Control Room Ventilation Radiation Monitoring System This system monitors the radiation level exterior to the inlet ducting of the control room ventilation system. The system con-sists of four channels identical to the channels in the contain-ment and drywell ventilaticn radiation monitoring system. The recorder is powered from 125 VDC bus B. l l l Two-out-of-two upscale (high/high)/ inoperative trips in channels A and C initiate shutdown and inboard isolation valve closure of the control room ventilation system and initiate startup of the l emergency air filtration fan (unit A). The same condition for channels B and D initiates shutdown and outboard isolation valve closure of the control room ventilation system and initiates startup of the emergency air filtration fan (unit B). l (Additional information is to be supplied by the Applicant.) . 12.3-48
GESSAR II J2A7007 2 38 NUCLEAR . ISLAND Rev. 15
) Table 15.7-4 GASEOUS RADWASTE SYSTEM FAILURE SYSTEM RUPTURE (DESIGN BASIS ANALYSIS) _
FISSION PRODUCT RELEASE TO ENVIRONMENT
- _
Isotope Ci Iso tope Ci Isotope Ci r
'Cr23 1. 39 E- 2 H3 1. 01E- 3 Rul03 1.07E-7 24 3.02E-2 Cl4 9.81E-5 105 1.08E-6 25 2.96E-2 Na24 4.18E-6 1-6 4.65E-9 i
1131 '9.98E-3 P32 4.01E-8 Ag110m 2.25E-7 132 1.28E-1 Cr51 9.70E-7 Tel29m 2.76E-8 133 -7.43E-2 Mn54 4.69E-8 129 3.98E-6 e 134 2.87E-1 56 1.06E-4 131m 1.86E-7 "j 135 1.19E-1 Fe59 1.47E-7 131 2.76E-7 . CoS8 8.58E-6 132 1.08E-6 Kr83m 9.00E+1 60 3.74E-7 , Co187 4.88E-8 85m 1.54E+2 Ni65 6.84E-7 188 1.72E-8 () 85 87 8.24E-2 4 .19 E-12 Zn65 Rb88 2.71E-9 1.42E+2 Co189 140 6.29 8.44E-8 , 8 aa 8l8 4 . 8 8 E- 12 89 6.80E+1 141 2.84E-8 89 1.52E-13 Br89 2.96E-2 142 4.31E-8 "d 90 1.35E-13 90 4.34E-5 Lal40 1.08E-4 - Xel31m 8.56E-1 91 8.12E-2 142 1.09E-8 , i 133m 8.81 92 1.30E-3 Cel41 2.00E-7 133 2.14E+2 Y90 3.51E-7 143 9.87E-7 + 135m 3.87E+2 91m 1.61E-2 144 7.66E-8 135 6.85E+2 91 1.75E-5 Nd147 1.18E-8 ~ 137 1.87E+3 92 5.31E-6 W187 1.90E-8 138 1.26E+3 93 3.48E-6 Np289 1.81E-3 139 3.54E+3 Zr95 1.83E-7 140 3.17E+3 97 1.73E-6 Nb95 1.60E-4 Mo99 1.69E-6 Te99m 8.97E-4 g-~(
'L/ 101 1.14E-3
- Undergoing reevaluation. Table will be updated in June 1983. ,
L5.7-37 i
-n~ , . . . , , ~ . , . . . ,,.-,-..n,. ,. - , , - . . ., , . . - ~ - ~ . ,
J GESSAR II 22A7007 238 NUCLEAR ISLAND R v. O Table 15.7-4 (Continued) GASEOUS RADWASTE SYSTEM FAILURE SYSTEM RUPTURE (DESIGN BASIS ANALYSIS) FISSION PRODUCT RELEASE TO ENVIRONMENT 3T 9.991E-01 90RB 1.471E 02 106TC 4.864E-02 143XE 1.815E 07 13N2 3.865E 03 90SR 8.272E-05 106RU 2 . r,8 3E- 0 6 143SC 9.297E-04 13AM 9.577E 00 90YM 1.262E-08 110AGM 1.250E-04 143BA 1.067E 00 13NO 5.922E-02 90Y 1.046E-06 129SB r.622C-04 143LA 6.382E-02 14C 3.721E-02 91BR 5.106E-03 129 TEM 3.535E-05 143CE 5.4835-04 16N2 8.299E 06 91KR 1.559E 05 129TE 2.184E-03 143PR 1.490E-04 16AM 3.623E 04 91RB 4.750E 02 1291 7. 410 E-0 9 144XE 1.743E D1 16NO 8.001E 01 91SR 2.394E-01 131SB 1.496E-02 14 4CS 5.810E-01 17N2 1.374E 03 91YM 5.578E-03 131 TEM 7.547E-05 144BA 1.441E 00 17AM 7.677E 00' 91Y 5.226E-05 131TE 1.588E-02 144LA 5.361E-01 17NO 3. 4 2 2E-02 92BR 1.991E-07 131.T '5.545C 00 144CE 4.253E-05 18P 4.9 51E 00 92KR 6.949E 04 1311EM 1.ll6E 02 147ND 6.288E-05 190 7.646E 02 92RB 7.091E 02 132TE 5.977E-04 147PM 6.639E-06 24NA 2.292E-03 92SR 8.3520-01 132I 7.094E 01 149ND 1.517E-03 32P 2.225E-05 92Y 2.876E-03 133SB 8.744E-02 149PM 6.996E-05 51CR 5.391E-04 93KR 5.428E 03 133 TEM 1.195E-02 187W 1.056E-02 54MN 2.604E-05 93RB 1.490E 02 133TE 3.510E-02 2391P 7.297E-01 56MN 5.873C-02 9 3SR 2.6590 00 133IM 1.043E 00 58CO 4.748E-03 9 3Y 1.933E-03 133I 4.*127E 01 59PC 8.181E-05 93ZR 6.383E-ll 133XCM 1.157E 03 60C0 2.080E-04 91NBM 2.307E-12 133XE 2.7980 04 65NI 3.524C-04 94KR 2. 36 6 E-0 5 134TC 2.616E-02 65ZN 1.505E-06 94RB 6.495E-01 134IM 5.703E 00 - 69ZNM 3.4400-05 94$R 8.276E-01 1341 1.3160 02 3.917C-0 2 83AS 94Y 5.011E-02 134XEM 6 . 4 6 5C-01 83SCM 2.ll3C-02 95KP 8.360E-02 ll5K 6.615E 01 3 83SE 1.1890-03 95RB 3.5 39E-0 3 135xEM 1.341C 05 83BR 7.746E 00 95SR 8.794C-01 135XC 8.437E 04 83KRM 1.458E 04 95Y 8.721E-02 135CSM 7.169E-05 84AS 2.611E-02 93ZR 1.0140-04 135CS 4.569C-10 84SE 3.169C-02 95NBM 2.172E-06 136TE 1.565C-01 840RM 2.799C-01 95NB L.863E-05 136IM 2.151E 01 84BR 1.680C 01 97ZR 9.621C-04 136I 3.258C OL 85AS 6.725E-03 97NBM 3.6 54C-01 1371 2.9370 01 85SEM 4.135E-02 97ND 1.193C-02 137XE 9.242E 05 85SC 5.0410-02 99ZR 2.766C-01 137CS 9.029C-05
^5BR 1.644C 01 99NBM 8.237C-02 137BAM 1.985E-04 85KRM 2.172C 04 99NB 4.7500-01 138XE 4.St0C 05 85KR 1.044C 01 99MO 9.380E-04 1 8CSM 3.111C-02 87AS 2.615C-01 99TCM 4.985C-01 128Cs 3.184C 00 87SC 9.006C-02 99TC 1.072E-08 1 % XE 1.953C 06 87BR 2.253E 01 101MO 4.963C-02 139CS 5.0390 01 87KR 7.499E 04 101TC 6.342E-01 139BA 3 ,219 C-01 88SC 6.922E-03 102MO 5.2300-02 140XE 1. 7 5'n: 06 88BR 2.011C 01 102TCM 5.445C-01 140CS 4.404C 02 88KR 7.315E 04 102TC 1.665C-04 140BA 2.551C 2 88RB 9.242E-01 10 3 rc 3.9300-01 140 LA 1.838E-04 895C 1.499C-06 103RU 5.9 39E-05 141XI: 4.342E 04 898R 8.827C 00 10 3RilM 6. 396 C-0 3 141CS 1.800E 02 89KR 7.645C 05 104MO 8.987C-02 141BA 1.409C 00 89RB 1.139C 01 104TC 2.17 8C-01 141LA 4.119E-03 89SR 6.8100-03 105MO 5.886E-02 141CC 1.112C-04 89YM 3.892C-08 105TC 1.213C-01 142XC 3.257C 03 90BR 1.2150 00 105RU 5.9 8 7C-0 4 142CS 9.245E 01 90K1 1.854C 06 10 5 RilM 1.9900-02 124BA ?.336C 00 90kBM 1.161E 01 105RH 8.604C-05 142LA 1.106C-02 l T = Tritiun All = Ammonia No = Mitrogen Oxide N2 : Gaseous Nitrogen 15.7-38
r o GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
, ('~' Table 15.7-5 GASEOUS RADWASTE SYSTEM FAILURE SYSTEM RUPTURE (DESIGN BASIS ANALYSIS) .
OFF-SITE RADIOLOGICAL EFFECTS (mrem) * * . Site Boundary
~
Total Bone Liver Body Thryoid Kidney Lung GI-LLI HalogenI - 1.31E-1 2.69E-1 1.02E-1 30.70 0.44 -- 1.21E-1 Noble Gas 1.67E+3* Other 6.1 15.56 8.29 2.3E-4 7.15 7.60 2.93 Total 6.23 15.83 1.68E+3 30.70 7.59 7.60 3.05 Low Population Zone s Total Bone Liver Body Thyroid Kidney Lung GI-LLI Halogen 6.6E-2 1.35E-1 5.0E-2 15.4 2.20E-1 0.67 Noble Gas 3.61E+2
, -I 'dther 3.07 7.77 4.14 1.2E-4 3.59 3.81 1.46
.:t i i (_ ,1 Total 3.14 7.90 3.65E+2 15.4 3.81 3.81 2.13 s, i,' , s s- ,
. i* Decay in flight accounted for.
** Undergoing reevaluation. Table will be updated in June 1983. ]
- sl t 21 .
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15.7-39 e
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 15 Table 15.7-6 GhSEOUS RADWASTE SYSTEM FAILURE SYSTEM RUPTURE (REALISTIC ANALYSIS) FISSION PRODUCT RELEASE TO ENVIRONMENT * ) Isotope Ci Isotope Ci Isotope Ci Br83 4.65E-3 H3 2.03E-3 Rn103 4.01E-9 84 1.01E-2 014 1.90E-4 104 3.89E-7 85 9.87U-3 Na24 4.13E-6 106 2.84E-9 1131 3.30E-3 P32 4.12E-9 Agl10m 1.24E-7 132 4.32E-2 Cr51 103E-6 Tel29m 1.08E-8 133 2.48E-2 Mn54 8.22E-9 129 1.41E-6 134 7.90E-2 56 1.06E-4 131m 4.54E-9
, 135 3.97E-2 Fe59 1.68E-7 131 9.53E-6 l CoS8 1.05E-5 122 3.62E-7 g Kr83m 8.40 60 9.92E-7 Cel37 1.91E-6 85m 1.25E+1 Ni65 6.34E-7 138 5.86E-1
,, 85 3.17E+2 Zn65 4.07E-9 Br139 1.82E-2 87 4.31E+1 Rb88 2.02E-1 140 4.17E-5 88 4.21E+1 89 1.10 141 8.46E-4 89 4.33E+2 Sr89 3.48E-5 142 1.40E-3 90 ,
1.07E+3 90 4.01E-7 Lal40 1.30E-7 X$131m 6.54E-2 91 3.98E-4 142 6.64E-6 133m 6.4SE-1 92 5.01E-4 Cel41 7.36E-8 133 1.58E+1 Y90 2.52E-9 143 3.39E-7 135m 7.43E+1 91m 9.61E-6 144 4.35E-8 l 135 4.70E+1 91 3.97E-8 Nd147 3.88E-8 137 5.08E+2 92 1.78E-6 W187 6. 34E-6 138 -2.50E+2 93 1.16E-6 Np239 4.39E-9 139 1 llE+3 Zr95 7.35E-8 140 1.03E+3 97 5.77E-7 Nb95 7.27E-8 Mo99 5.67E-7 Te99m 3.03E-4 101 3.81E-4
" Undergoing reevaluation. Table will be updated in June 1983. ]
15.7-40
^
t
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O m Table 15.7-6 GASEOUS RADWASTE SYSTEM FAILURE SYSTEM RUPTURE (REALISTIC ANALYSIS) FISSION PRODUCT RELEASE TO ENVIRONMENT FROM RUPTURE OF SJAE OUTLET LINE ONLY (Continued) 1.141E 07 3 11 1.128E 00 90RB 4.749E 01 106TC 1.621E-02 143XE 3.099E-04 13N2 3.691E 03 90SR 1.442E-04 106RU 1.579E-06 143CS 1.462E-01 13AM 9.577E 00 90YM 4.208E-09 110AGM 6.876E-05 143BA 3.5560-01 13NO 5.922E-02 90Y 1.361E-06 239S8 1.207E-04 143LA 2.127E-02 14C 1.037E-01 91BR 1.702E-03 129 TEM 5.988E-06 143CE 1.830E-04 16N 2 8.270E 06 91ER 5.157C 05 129TE 7.822C-04 143PR 5.178E-05 16AM 3.623E 04 91RB 1.c76E 02 129I 1.574r-08 144XE 5.810E 00 16NO 8.801E 01 91SR 7.977P-02 131SB 4.9 86E-0 3 144CS 1.937C-01 17N2 1.37:E 03 91YM 1.855L-03 131 TEM 2.522E-05 144BA 4.804E-01 17AM 7.679E 00 91Y 2.067C-05 131TE 5.295E-03 144LA 1.787E-01 17No 3.422E-02 92BR 6.636E-08 1311 1.878E 00 144CE 2.418E-05 18F 4.952E 00 92KR 2.316E 04 131XEM 3.481E 01 147ND .' 155C-05 190 7.677E 02 92RH 2.364E 02 132TE 2.010E-04 147PM 5.195C-06
-24NA 2.234E-03 92SR 2.785U-ol 132I 2.399E-01 149ND 5.058E-04 32P 2.288C-05 92Y 9.591U-04 133SB 2.915E-02 149PM 2.3393-05 51CR 5.852E-04 93PR 1.809E 03 133 TEM 3.983E-03 187W 3. 52 3C-0 3 54MN 4.566E-05 93RB 4.965E 01 133TE 1.170E-02 239NP 2.440E-01 56MN 5.873E-02 93SR 8.862E-01 133IM t.476c-01 SRCO 5.822E-03 93Y 6.444E-04 131I 1.3782 01 59FC 9.343E-05 92ZR 6.726E-Il 133XEM 3.430R 02 60C0 5.511E-04 93NBM 7.577E-12 133XE 8.360C 03 65NI 3.524E-04 94KR 7.887E-06 134TE 8.721E-03
) 65ZN 2.263E-06 94RB 2.165E-01 134IM 1.901E 00 sj 69ZNM '3.442E-05 945R 2.759E-01 134I 4. 388E 01 83AS 1.106E-02 94Y 1.670E-02 131XEM 2.155E-01 83SEM 7.043C-03 95KR 2.787E-02 1Mi 2.207E 01 83SE 4.6 31E-0 4 9 ~,RB 1.176E-03 135XEM 3.965L 04 83BR 2.583E 00 95SR 2.9 31E-01 13BXE 2.486C 04 83KRM 4.449E 03 95Y 2.907E-02 135CSM 2.390E-05 84AS 8.704E-03 95ZR 4.084C-05 135CS 1.9 2 2C- 09 84SC 1.056E-02 95NBM 8.9582-07 136TE 5.2170-02 84BRM 9.329E-02 95NM 4.039E-05 136IM 7.170E 00 84BR 5.602E 00 97ZR 3.208E-04 1361 1.086E 01 85AS 2.242E-03 97NDM 1.221E-01 137I 9.789E 00 85SEM 1.378E-02 97NB 3.9 86E-0 3 137XE 2.755E 05 85SC 1.680E-02 99ZR 9.221C-02 137CS 1. 49 3C-0 4 85BR 5.481E 00 93NDM 2.746E-02 137B AM ' 1.705E-04 85KRM 6.625E 03 99NB 1.583C-01 138XE 1.332E ^'
85KR 1.680E 01 99MO 3.149E-04 138CSM 1.037E i 87AS 8.716E-05 9fTCM 1.685E-01 138CS 9.669C-bi 87SC 3.002C-02 99TC 2.382E-08 139 XE 6.076E 05 87BR 74 509E 00 101MO 1.654E-02 139CS 1.597E 01 87KR .2.287C 04 101TC 2.114C-01 139BA 1.078E-01 88SC 2.307E-03 102MO 1.744C-02 140XE 5.723E 05 88BR 6.702E 00 102TCM 1.815C-01 140CS 1.445E 02 88KR 2.231E 04 102TC 5.549E-05 140BA 8.676E-03 88RB 2.873C-01 103TC 1.310E-01 140LA 6.328E-05 89st 4.997E-07 103RU 2.25E-05 141XC 1.447C 04 89BR 2.942C 00 103R!iM 2.442E-03 141CS 6.001E 01 00KR 2.356E 05 104MO 2.996E-02 141BA 4.697C-01 89RB 3.572C 00 104TC 7.260C-02 141LA 1.272E-03 p 89SR 2.533E-03 105MO 1.962C-02 141CE 4.086E-05 105TC 4.045E-02 142XE 1.086C 0 3 (v) 89YM 90DR l'. 29 7C- 0 8 4.051C-01 105RC 1.996C-04 142CS 6.6 37E-0 3 142BA 3.082C 01 7.785E-01 90KR 5.912C 0 5 - 105RIIM 9011B:-l 3.749C 00 103Rll -2.874C-05 142LA 3.6 8 7E-0 3 15.7-41
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 Table 15.7-7 GASEOUS RADWASTE SYSTEM FAILURE (REALISTIC ANALYSIS) FIRST CHARCOAL TANK RUPTURE RADIOLOGICAL EFFECTS (mrem) * * ] Site Boundary Total Bone Liver Body Thyroid Kidney Lung GI-LLI Halogen 4.4E-2 9.05E-2 3.35E-2 1.03 1.47E-1 4.06E-2 Noble Gas 2.88E+2* Other 2.18 6.33 3.66E-2 2.83E-4 2.46E-2 2.33E-2 1.73E-2 Total 2.22 6.42 2.88E+2 1.03 1.71E-1 2.33E-2 5.79E-2 Low Propulsion Zone Total Bone Liver Body Thyroid Kidney Lung GI-LLI Halogen 4.6E-3 1.2E-2 4.4E-3 7.7E-1 1.9E-2 9.00E-3 Noble Gas 37.0* Other 1.09E-2 3.16E-2 1.8E-2 1.42E-4 1.23E-2 1.16E-2 8.6E-3 Total 1.55E-2 4.36E-2 37.0 7.7E-1 3.13E-2 1.16E-2 1.8E-2
- Decay in flight accounted for. y
- Undergoing reevaluation. Table will be updated in June 1983. J O
15.7-42
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O 15A.6.3.3 Event Definitions and Operational Safety O' Evaluations (Continued) If the event occurs when the feedwater system is not active in State D,.a loss in the coolant inventory results in a reactor vessel isolation. The low water level signal initiates reactor vessel isolation. The nuclear system pressure relief system provides pressure relief. Core cooling is accomplished by the RCIC and HPCS systems, which are automatically initiated by the incident detection circuitry (IDC). The automatic depressurization system (ADS) or the manual relief valve system remain as the backup depressurization system, if needed. After the vessel has depressurized, long-term core cooling is accomplished by the LPCI, LPCS and HPCS, which are ini-tiated on low water level by the IDC system or are manually oper-ated. . Containment-suppression pool cooling is manually initiated. O Event 16 - Control Rod Withdrawal Error During Refueling and Startup Operations Because a control rod withdrawal error resulting in an increase of positive reactivity can occur under any operating condition, it must be considered in all operating states. For this specific event situation, only State A and B apply. Refueling No unique safety action is required in operating State A for the withdrawal of one control rod because the core is more than one control rod subcritical. Withdrawal of more than one control rod is precluded by the protection sequence shown in Figure 15A.6-16. During core alterations, the mode switch is normally in the REFUEL position, which allows the refueling equipment to be positioned over the core and also inhibits control rod withdrawal. This O transient, therefore, applies only to operating State A. ! 15A.6-19
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 15A.6.3.3 Event Definitions and Operational Safety Evaluations (Continued) No safety action is required because the total worth (positive reactivity) of one fuel assembly or control rod is not adequate to cause criticality. Moreover, mechanical design of the control rod assembly prevents physical removal without removing the adjacent fuel assemblies. Startup During low power operation (State B), the neutron monitoring sys-tem via the RPS will initiate SCRAM, if necessary (Fig-ure 15A.6-16). Event 17 - Control Rod Withdrawal Error (During Power Operation) Because a control rod withdrawal error resulting in an increase of positive reactivity can occur under any operating condition, it must be considered in all operating states. For this specific event situation, only States C and D apply. During power operation (Power Range State D) , a number of plant protective devices of various designs prohibit the control rod motion before critical levels are reached (Figure 15A.6-17). While in State C, no protective action is needed.
~
Systems in the power range (0 to 100% NBR) prevent the selection of an out-of-sequenced rod movement by using the rod pattern con-trol which uses either Banked Position or Grouped Notch With- , drawal sequences. In addition, the movement of the rod is monitored and limited with acceptqble intervals either by neutronic effects or actual rod motion. The rod pattern con-troller provides movement surveillance. Beyond these rod motion control limits are the fuel / core SCRAM protection systems. While in State C, no protective action is needed. l 15A.6-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 15A.6.5.3 Event Definition and Operational Safety f/)
\_ Evaluations (Continued)
As shown in Figure 15A.5-37, in operating-State C (reactor shut down, but pressurized), a pipe break accident up to the DBA can be accommodated within the nuclear safety operational criteria through the various operations of the MSIVs, emergency core cool-ing systems (HPCS, ADS, LPCI, CSCS and LPCS), containment and reactor vessel isolation control system, reactor / shield / auxiliary buildings, standby gas treatment system, main control room heating, cooling and ventilation system, MSIV-LCS, emergency cooling sys-tems, and the incident detection circuitry. For small pipe breaks inside the containment, pressure relief is effected by the nuclear system pressure relief system, which transfers decay heat to the suppression pool. For large breaks, depressurization takes place through the break itself. In State D (reactor not shut down, but pressurized), the same equipment is required as in State C but, in addition, the reactor protection system and the CRD system must () operate to scram the reactor. The limiting items, on which the operation of the above equipment is based, are the allowahle fuel cladding temperature and the containment pressure capability. The CRD housing supports are considered necessary whenever the system is pressurized to prevent excessive control: rod movement through the bottom of the reactor pressure vessel following the postulated l rupture of one CRD housing (a lesser case of the des'ign basis LOCA and a related preventive of a postulated rod ejection accident) . After complbtion of the automatic action vf the above equipment, manual operation of the RHRS (suppression pool cooling mode) and ADS or relief valves (controlled depressurization) is required to maintain containment pressure and fuel cladding temperature within limits during extended core cooling. 15A.6-37
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 15A.6.5.3 Event Definition and Operational Safety Evaluations (Continued) Events 38, 39, 40 - Loss-of-Coolant Accidents (LOCA) Resulting From Postulated Pipe Breaks - Outside Containment Pipe break accidents outside the containment are assumed to occur any time the nucle'ar system is pressurized (States C and D). This accident is most severe during operation at high power (State D). In State C, this accident becomes a subset of the State D sequence. The protection sequences for the various possible pipe breaks out-side the containment are shown in Figures 15A.6-38. The sequences also show that for small breaks (breaks not requiring immediate action), the reactor operator can use a large number of process indications to identify the break and isolate it. In Operating State D (reactor not shut down, but pressurized) , h scram is accomplished through operation of the reactor protection system and the CRD system. Reactor vessel isolation is accom-plished through operation of the main steamline isolation valves and the containment and reactor vessel isolation control system. For a main steamline break, initial core cooling is accomplished by either the HPCS or the automatic depressurization system ] (ADS) or manual relief valve operation in conjunction with the LPCS, or LPCI. These systems provide parallel paths to effect initial core cooling, thereby satisfying the single-failure criterion. Extended core cooling is accomplihsed by the single-failure proof, parallel combination of LPCS, HPCS and LPCI systems. The ADS or relief valve system operation and the RHRS suppression pool cooling mode (both manually operated) are required to main-tain containment pressure and fuel cladding temperature within limits during extended core cooling. 15A,6-38
GESSAR II- 22A7007 238 NUCLEAR : ISLAND Rsv. 15 j APPENDIX 15D TABLES - Table Title Page 15D.2-1 BWR/6 Water Injection Systems Capability 15D.2-53 15D.2-2 Mark III Long-Term Pressurization 15D.2-54 i 15D.2-3 Containment Structures Drywell Structures 15D.2-55 15D.2-4 BWR/6 PRA Results: Breakdown of the
- Assessed Frequency of Core Damage Per ,
Reactor Year 15D.2-56 15D.2-5 BWR/6 PRA Results: Distribution of Core Damage Sequence Probability 15D.2-57a L 15D.2-6 Summary of Pool Scrubbing Tests from Reference 6 15D.2-58a 15D.2-7 Minimum Supportable and Expected Suppression Pool Decontamination Factors () 15D.2-8 for Iodine and Particulates (Reference 6) Input Parameters for Offsite Dose 15D.2-61a j Calculations 15D.2-62a
-15D.2-9 Acute Bone Marrow Doses (Rem) vs.
Downwind Distance 15D.2-63
- 15D.2-10 Inputs for Sample Case 15D.2-64 15D.2-ll Lifetime Accident Whole Body Dose (Rem) l vs. Downwind Distance vs. DF 15D.2-65 15D.2-12 Potential Acute Exposure Consequences 15D.2-66 15D.2-13 Loss if Life Expectancy Due to Various Causes 15D.2-67 .
-15D.2-14 Long Term Effects 15D.2-69 15D.2-15 Suppression Pool Scrubbing Factors 15D.2-70 1
+ 1 O 15D-ib - 4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 APPENDIX 15D SEVERE ACCIDENTS ILLUSTRATIONS FIGURES PAGE 15D.2-1 Systems to Supply Water to Core 15D.2-71 15D.2-2 Decay Heat Removal 15D.2-72 15D.2-3 Water Level Measurement 15D.2-73 15D.2-4 Containment Passive Heat Sink Capability 15D.2-74 15D.2-5 Planned RCIC Restart Modification (Figure 5.4-8b) 15D.2-75 15D.2-6 Planned RCIC Restart Modification (Figure 7.4-lb) 15D.2-76 15D.2-7 Planned RCIC Restart Modification 15D.2-77a 15D.2-0 Planned RCIC Restart Modification - (Figure 7.4-la) 15D.2-78a 15D.2-9 Planned RCIC Break Detection Modification (Figure 7.4-lb) 15D.2-79a 15D.2-10 Pressure Suppression Containment Banners and Filters 15D.2-80a 15D.2-ll Mark III Fission Product Retention (Halogens and Particulates) 15D.2-81 15D.2-12 Importance of Fission Product Retention in Mark III Pressure Suppression Containment 15D.2-82 15D.2-13 Whole Body Dose with Respect to Distance 15D.2-83 15D.2-14 Calculation Model for DF 15D.2-84 15D.2-15 Particle Size Distribution for Sandia Corium-Steel Experiment 15D.2-85 15D.2-16 Particle Size Distribution for Sandia Corium-Concrete Experiment 15D.2-86 llII 15D-ic < l
d GESSAR II 22A7007 238 NUCLEAR ISLAND Rev.
- O 18A.3 EP-2 CONTAINMENT CONTROL i
PROPRIETARY INFORMATION - provided under separate cover i O 1 I O *Pages 18A.3-3, 18A.3-4, and 18A.3-5 are Revision 15. The remaining pages are Revision 9. 18A.3-1 through 18A.3-12 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 (~N 19.1.1 Chapter 1 - Question / Response Index (s . NRC GESSAR II GESSAR II NRC Question Question y Revision Transmittal Number Number Disposition Number Note 2 100.1 1.1 Tables 1.3-1 through 4 1.3-6 430.17 1.2 Subsections 1.8.7.5 14 Note 3 d n and 19.12 . 430.20 1.3 Subsections 1.8.128 and 8.3.2.2.1.2.8 430.31 1.4 Subsection 1.8.63 430.32 1.5 Subsection 19.1.5 and Figures 7A.8-la through 7A.8-lh Subsection 19.3.1.6 460.09 1.6 460.18a 1.7a Subsection 19.3.1.7 460.18b 1.7b Subsection 19.3.1.7 460.18c 1.7c Section lA.77 471.16 1.8 Table 1AA-2 9 471.18 1.9 Subsection 19.3.1.9 Note 3 471.19 1.10 Section lAA.2 and Subsection lAA.3.3
. Note 4 421.13 1.11 Subsection 19.3.1.11 h 480.51 1.12 Subsection 19.3.1.12
(}
% and Table 1AC.2-1 480.52 1.13 Subsection 19.3.1.13 9 and Table 1AC.2-1 Note 4 480.53 1.14 Subsection 19.3.1.14 U Note 5 271.08 1.15 Subsection 19.3.1.15 14 Chapter 1 - Question / Response Index Notes
- 1. Subsections shown in parentheses reference the corresponding Chapter 19 subsection which details the answer to the question.
- 2. Darrell G. Eisenhut to Glenn G. Sherwood, " Acceptance Review of Application for Final Design Approval for 238 Nuclear i Island," December 9, 1981.
l 3. Thomas M. Novak to Glenn G. Sherwood, " Request for Additional l Information Regarding General Electric Application for an FDA
) for a Standardized Nuclear Island (GESSAR II) ," August 25, l s.)
1982. 19.1.1-1 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 Chapter 1 - Question / Response Index Notes (Continued)
- 4. Frank J. Miraglia to Glenn G. Sherwood, " Request for Additional Information Regarding the General Electric Application for an FDA for a Standardized Nuclear Island (GESSAR II) ," October 5, 1982.
- 5. Cecil O. Thomas, to Glenn G. Sherwood, " Request for Additional Information Regarding the Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Described in the General Electric Application for an FDA for a Standardized Nuclear Island (GESSAR-II) , " January 31, 1983.
O G
~
19.1.1-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 s ,) SECTION 19.3 CONTENTS Section Title Page 19.3 RESPONSES 19.3.0.1-1 19.3.0 Miscellaneous - Responses 19.3.0.1-1 19.3.0.1 Question / Response 0.1 (100.3) 19.3.0.1-1 19.3.1 Chapter 1 - Responses 19.3.1.1-1 19.3.1.1 Question / Response 1.1 (100.1) 19.3.1.1-1 19.3.1.2 Question / Response 1.2 (430.17) 19.3.1.2-1 19.3.1.3 Question / Response 1.3 (430.20) 19.3.1.3-1 19.3.1.4 Question / Response 1.4 (430.31) 19.3.1.4-1 19.3.1.5 Question / Response 1.5 (430.32) 19.3.1.5-1 19.3.1.6 Question / Response 1.6 (430.09) 19.3.1.6-1 19.3.1.7 Question / Response 1.7 (460.18) 19.3.1.7-1 19.3.1.8 Question / Response 1.8 (471.16) 19.3.1.8-1 19.3.1.9 Question / Response 1.9 (471.18) 19.3.1.9-1 19.3.1.10 Question / Response 1.10 (471.19) 19.3.1.10-1 l 19.3.1.11 Question / Response 1.11 (421.13) 19.3.1.11-1 19.3.1.12 Question / Response 1.12 (480.51) 19.3.1.12-1 19.3.1.13 Question / Response 1.13 (480.52) 19.3.1.13-1 19.3.1.14 Question / Response 1.14 (480.53) 19.3.1.14-1 19.3.1.15 Question / Response 1.15 (271.08) 19.3.1.15-1 _
.19.3.2 Chapter 2 - Responses 19.3.2.1-1 19.3.2.1 Question / Response 2.1 (241.1) 19.3.2.1-1 19.3.2.2 Question / Response 2.2 (240.01) 19.3.2.2-1
( ; 19.3.2.3 Question / Response 2.3 (240.02) 19.3.2.3-1 19.3.2.4 Question / Response 2.4 (240.03) 19.3.2.4-1 ~ 1 19.3-i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.2.5 Question / Response 2.5 (240.04) 19.3.2.5-1 19.3.2.6 Question / Response 2.6 (240.05) 19.3.2.6-1 19.3.3 Chapter 3 - Responses 19.3.3.1-1 19.3.3.1 Question / Response 3.1 (210.1) 19.3.3.1-1 19.3.3.2 Question / Response 3.2 (210.2) 19.3.3.2-1 19.3.3.3 Question / Response 3.3 (220.1) 19.3.3.3-1 19.3.3.4 Question / Response 3.4 (220.2) 19.3.3.4-1 19.3.3.5 Question / Response 3.5 (220.3) 19.3.3.5-1 19.3.3.6 Question / Response 3.6 (220.4) 19.3.3.6-1 19.3.3.7 Question / Response 3.7 (220.5) 19.3.3.7-1 g 19.3.3.8 Question / Response 3.8 (220.6) 19.3.3.8-1 19.3.3.9 Question / Response 3.9 (220.7) 19.3.3.9-1 19.3.3.10 Question / Response 3.10 (220.8) 19.3.3.10-1 19.3.3.11 Question / Response 3.11 (241.2) 19.3.3.11-1 19.3.3.12 Quertion/ Response 3.12 (241.3) 19.3.3.12-1 19.3.3.13 Question / Response 3.13 (241.4) 19.3.3.13-1 19.3.3.14 Question / Response 3.14 (241.5) 19.3.3.14-1 19.3.3.15 Question / Response 3.15 (241.6) 19.3.3.15-1 19.3.3.16 Question / Response 3.16 (241.7) 19.3.3.16-1 19.3.3.17 Question / Response 3.17 (241.8) 19.3.3.17-1 19.3.3.18 Question / Response 3.18 (241.9) 19.3.3.18-1 19.3.3.19 Question / Response 3.19 (241.10) 19.3.3.19-1 19.3.3.20 Question / Response 3.20 (241.11) 19.3.3.20-1 19.3.3.21 Question / Response 3.21 (241.12) 19.3.3.21-1 19.3-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.3.22 Question / Response 3.22 (241.13) 19.3.3.22-1 19.3.3.23 Question / Response 3.23 (241.14) 19.3.3.23-1 19.3.3.24 Question / Response 3.24 (241.15) 19.3.3.24-1 19.3.3.25 Question / Response 3.25 (241.46) 19.3.3.25-1 19.3.3.26 Question / Response 3.26 (241.17) 19.3.3.26-1 19.3.3.27 Question / Response 3.27 (241.18) 19.3.3.27-1 19.3.3.28 Question / Response 3.28 (241.19) 19.3.3.28-1 19.3.3.29 Question / Response 3.29 (241.20) 19.3.3.29-1 19.3.3.30 Question / Response 3.30 (241.21) 19.3.3.30-1 19.3.3.31 Question / Response 3.31 (241.22) 19.3.3.31-1 19.3.3.32 Question / Response 3.32 (241.23) 19.3.3.32-1 19.3.3.33 Question / Response 3.33 (241.24) 19.3.3.33-1 19.3.3.34 Question / Response 3.34 (241.25) 19.3.3.34-1 19.3.3.35 Question / Response 3.35 (241.26) 19.3.3.35-1 19.3.3.36 Question / Response 3.36 (251.1) 19.3.3.36-1 19.3.3. " Question / Response 3.37 (270.1) 19.3.3.37-1 l 19.3.3.38 Question / Response ~3.38 (270.2) 19.3.3.38-1 19.3.3.39 Question / Response 3.39 (270.3) 19.3.3.39-1 i j 19.3.3.40 Question / Response 3.40 (270.4) 19.3.3.40-1 19.3.3.41 Question / Response 3.41 (271.1) 19.3.3.41-1 l l 19.3.3.42 Question / Response 3.42 (220.01) 19.3.3.42-1 19.3.3.43 Question / Response 3.43 (220.02) 19.3.3.43-1 19.3.3.44 Question / Response 3.44 (220.03) 19.3.3.44-1
])
19.3.3.45 Question / Response 3.45 (220.04) 19.3.3.45-1 19.3-iii l
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.3.46 Question / Response 3.46 (220.05) 19.3.3.46-1 19.3.3.47 Question / Response 3.47 (220.06) 19.3.3.47-1 19.3.3.48 Question / Response 3.48 (220.07) 19.3.3.48-1 19.3.3.49 Question / Response 3.49 (220.08) 19.3.3.49-1 19.3.3.50 Question / Response 3.50 (220.09) 19.3.3.50-1 19.3.3.51 Question / Response 3.51 (220.10) 19.3.3.51-1 19.3.3.52 Question / Response 3.52 (220.11) 19.3.3.52-1 19.3.3.53 Question / Response 3.53 (220.12) 19.3.3.53-1 19.3.3.54 Question / Response 3.54 (220.13) 19.3.3.54-1 19.3.3.55 Question / Response 3.55 (220.14) 19.3.3.55-1 19.3.3.56 Question / Response 3.56 (220.15) 19.3.3.56-1 19.3.3.57 Question / Response 3.57 (220.16) 19.3.3.57-1 19.3.3.58 Question / Response 3.58 (220.17) 19.3.3.58-1 19.3.3.59 Question / Response 3.59 (220.18) 19.3.3.59-1 19.3.3.60 Question / Response 3.60 (220.19) 19.3.3.60-1 19.3.3.61 Question / Response 3.61 (220.20) 19.3.3.61-1 19.3.3.62 Question / Response 3.62 (220.21) 19.3.3.62-1 19.3.3.63 Question / Response 3.63 (220.22) 19.3.3.63-1 19.3.3.64 Question / Response 3.64 (220.23) 19.3.3.64-1 19.3.3.65 Question / Response 3.65 (220.24) 19.3.3.65-1 19.3.3.66 Question / Response 3.66 (220.25) 19.3.3.66-1 19.3.3.67 Question / Response 3.67 (220.26) 19.3.3.67-1 19.3.3.68 Question / Response 3.68 (220.27) 19.3.3.68-1 19.3.3.69 Question / Response 3.69 (220.28) 19.3.3.69-1 1 19.3-iv
- J GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15
() SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.3.70 Question / Response 3.70 (220.29) 19.3.3.70-1 19.3.3.71 Question / Response 3.71 (220.30) 19.3.3.71-1 19.3.3.72 Question / Response 3.72 (220.31) 19.3.3.72-1 19.3.3.73 Question / Response 3.73 (220.32) 19.3.3.73-1 19.3.3.74 Question / Response 3.74 (220.33) 19.3.3.74-1 19.3.3.75 Question / Response 3.75 (220.34) 19.3.3.75-1 19.3.3.76 Question / Response 3.76 (220.35) 19.3.3.76-1 19.3.3.77 Question / Response 3.77 (220.36) 19.3.3.77-1 19.3.3.78 Question / Response 3.78 (220.37) 19.3.3.78-1 19.3.3.79 Question / Response 3.79 (220.38) 19.3.3.79-1
~~/ 19.3.3.80 Question / Response 3.80 (220.39) 19.3.3.80-1 19.3.3.81 Question / Response 3.81 (220.40) 19.3.3.81-1 19.3.3.82 Question / Response 3.82 (220.41) 19.3.3.82-1 19.3.3.83 Question / Response 3.83 (220.42) 19.3.3.83-1 19.3.3.84 Question / Response 3.84 (220.43) 19.3.3.84-1 19.3.3.85 Question / Response 3.85 (220.44) 19.3.3.85-1 19.3.3.86 Question / Response 3.86 (410.01) 19.3.3.86-1 19.3.3.87 Question / Response 3.87 (410.02) 19.3.3.87-1 i
- 19.3.3.88 Question / Response 3.88 (410.03) 19.3.3.88-1 19.3.3.89 Question / Response 3.89 (410.04) 19.3.3.89-1 19.3.3.90 Question / Response 3.90 (410.05) 19.3.3.90-1 19.3.3.91 Question / Response 3.91 (410.06) 19.3.3.91-1 l /~N 19.3.3.92 Question / Response 3.92 (410.07) 19.3.3.92-1 i
19.3.3.93 Question / Response 3.93 (410.08) 19.3.3.93-1 19.3-v i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.3.94 Question / Response 3.94 (410.09) 19.3.3.94-1 19.3.3.95 Question / Response 3.95 (410.10) 19.3.3.95-1 19.3.3.96 Question / Response 3.96 (410.11) 19.3.3.96-1 19.3.3.97 Question / Response 3.97 (410.12) 19.3.3.97-1 19.3.3.98 Question / Response 3.98 (410.13) 19.3.3.98-1 19.3.3.99 Question / Response 3.99 (410.14) 19.3.3.99-1 19.3.3.100 Question / Response 3.100 (410.15) 19.3.3.100-1 19.3.3.101 Question / Response 3.101 (460.10) 19.3.3.101-1 19.3.3.102 Question / Response 3.102 (271.01) 19.3.3.102-1 19.3.3.103 Question / Response 3.103 (271.02) 19.3.3.103-1 19.3.3.104 Question / Response 3.104 (271.03) 19.3.3.104-1 19.3.3.105 Question / Response 3.105 (271.04) 19.3.3.105-1 19.3.3.106 Question / Response 3.106 (271.05) 19.3.3.106-1 19.3.3.107 Question / Response 3.107 (271.06) 19.3.3.107-1 19.3.3.108 Question / Response 3.108 (271.07) 19.3.3.108-1 19.3.3.109 Question / Response 3.109 (271.09) 19.3.3.109-1 19.3.3.110 Question / Response 3.110 (271.10) 19.3.3.110-1 i l 19.3.3.111 Question / Response 3.111 (271.11) 19.3.3.111-1 1 19.3.3.112 Question / Response 3.112 (271.12) 19.3.3.112-1 19.3.3.113 Question / Response 3.113 [MEB (DSER) l Item No. 1] 19.3.3.113-1 19.3.3.114 Question / Response 3.114 [MEB (DSER) Item No. 2] 19.3.3.114-1 19.3.3.115 Question / Response 3.115 [MEB (DSER) Item No. 31 19.3.3.115-1 1 1 19.3-vi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 , SECTION 19.3 CONTENTS (Continued) Section Title Page l 19.3.3.116 Question / Response 3.116 [MEB (DSER) Item No. 4] 19.3.3.116-1 ' 19.3.3.117 Question / Response 3.117 [MEB (DSER) Item No. 5] 19.3.3.117-1 19.3.3.118 Question / Response 3.118 [MEB (DSER) Item No. 6] 19.3.3.118-1
- 19.3.3.119 Question / Response 3.119 [MEB (DSER) .
Item No. 7] 19.3.3.119-1 ! 19.3.3.120 Question / Response 3.120 [MEB (DSER) Item No. 8] 19.3.3.120-1 ) 19.3.3.121 Question / Response 3.121 [MEB (DSER)
' Item No. 9J 19.3.3.121-1 19.3.3.122 Question / Response 3.122 [MEB (DSER)
() 19.3.3.123 Item No. 10] Question / Response 3.123 [MEB (DSER)
.19.3.3.122-1
' Item No. 11] 19.3.3.123-1 19.3.3.124 Question / Response 3.124 [MEB (DSER)
- Item No. 12] 19.3.3.124-1 19.3.3.125 Question / Response 3.125 [MEB (DSER) j Item No. 13] 19.3.3.125-1 19.3.3.126 Question / Response 3.126 [MEB (DSER)
Item No. 14] 19.3.3.126-1 19.3.3.127 Question / Response 3.127 [MEB (DSER) Item No. 15] 19.3.3.127-1 19.3.3.128 Question / Response 3.128 [MEB (DSER) Item No. 16] 19.3.3.128-1 19.3.3.129' Question / Response 3.129 [MEB (DSER) Item No. 17] 19.3.3.129-1 19.3.3.130 Question / Response 3.130 [MEB (DSER) Item No.~18] 19.3.3.130-1 19.3.3.131 Question / Response 3.131 [MEB (DSER) Item No. 19] 19.3.3.131-1 , 19.3-vii i
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.3.132 Question / Response 3.132 [MEB (DSER) Item No. 20] 19.3.3.132-1 19.3.3.133 Question / Response 3.133 [MEB (DSER) Item No. 21] 19.3.3.133-1 19.3.3.134 Question / Response 3.134 [MEB (DSER) Item No. 22] 19.3.3.134-1 19.3.3.135 Question / Response 3.135 [MEB (DSER) Item No. 23] 19.3.3.135-1 19.3.3.136 Question / Response 3.136 [MEB (DSER) Item No. 24] 19.3.3.136-1 19.3.3.137 Question / Response 3.137 [MEB (DSER) Item No. 25] 19.3.3.137-1 19.3.3.138 Question / Response 3.138 [MEB (DSER) Item No. 26] 19.3.3.138-1 19.3.3.139 Question / Response 3.139 [MEB (DSER) Item No. 27] 19.3.3.139-1 19.3.3.140 Question / Response 3.140 [MEB (DSER) Item No. 28] 19.3.3.140-1 19.3.3.141 Question / Response 3.141 [MEB (DSER) Item No. 29] 19.3.3.141-1 19.3.3.142 Question / Response 3.142 [MEB (DSER Item No. 30] 19.3.3.142-1 19.3.3.143 Question / Response 3.143 [MEB (DSER) l Item No. 31] 19.3.3.143-1 19.3.3.144 Question / Response 3.144 [MEB (DSER) Item No. 32] 19.3.3.144-1 19.3.3.145 Question / Response 3.145 [MEB (DSER) Item No. 33] 19.3.3.145-1 l 19.3.3.146 Question / Response 3.146 [MEB (DSER) Item No. 34] 19.3.3.146-1 19.3-viii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 (~h i SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.3.147 Question / Response 3.147 [MEB (DSER) Item No. 35] 19.3.3.147-1 19.3.3.148 Question / Response 3.148 [MEB (DSER Item No. 36] 19.3.3.148-1 19.3.3.149 Question / Response 3.149 [MEB (DSER) Item No. 37] 19.3.3.149-1 19.3.3.150 QuestiGn/Respo:: c 3.150 !MEB (DSER) Item No. 38] 19.3.3.150-1 19.3.3.151 Quest ion /Pesponse 3.151 {MEB (DSER) Item No. 39] 19.3.3.151-1 19.3.3.152 Question /s sponse 3.152 [MEB (DSER) Item No. 4L] 19.3.3.152-1 19.3.3.153 Question / Response 3.153 [MEB (DSER) Item No. 41] b) I 19.3.3.154 Question / Response 3.154 [MEB (DSER) 19.3.3.153-1 Item No. 42] 19.3.3.154-1 19.3.3.155 Question / Response 3.155 [MEB (DSER) Item No. 43] 19.3.3.155-1 19.3.3.156 Question / Response 3.156 [MEB (DSER) Item No. 44] 19.3.3.156-1 19.3.3.157 Question / Response 3.157 [MEB (DSER) Item No. 45] 19.3.3.157-1 19.3.3.158 Question / Response 3.158 [MEB (DSER) Item No. 46] 19.3.3.158-1 19.3.3.159 Question / Response 3.159 [MEB (DSER) Item No. 57] 19.3.3.159-1 19.3.3.160 Question / Response 3.160 [MEB (DSER) Item No. 48] 19.3.3.160-1 19.3.3.161 Question / Response 3.161 [MEB (DSER) Item No. 49]' 19.3.3.161-1 'sh s/ 19.3.3.162 Question / Response 3.162 [MEB (DSER) Item No. 50] 19.3.3.162-1 19.3-ix
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.3.163 Question / Response 3.163 [MEB (DSER) Item No. 51] 19.3.3.163-1 19.3.3.164 Question / Response 3.164 [MEB (DSER) Item No. 52] 19.3.3.164-1 19.3.3.165 Question / Response 3.165 [MEB (DSER) Item No. 53] 19.3.3.165-1 19.3.3.166 Question / Response 3.166 [MEB (DSER) Item No. 54] 19.3.3.166-1 19.3.3.167 Question / Response 3.167 [MEB (DSER) Item No. 55] 19.3.3.167-1 19.3.3.168 Question / Response 3.168 [MEB (DSER) Item No. 56] 19.3.3.168-1 19.3.3.169 Question / Response 3.169 [MEB (DSER) Item No. 57] 19.3.3.169-1 19.3.3.170 Question / Response 3.170 [MEB (DSER) Item No. 58] 19.3.3.170-1 19.3.3.171 Question / Response 3.171 [MEB (DSER) Item No. 59] 19.3.3.171-1 19.3.3.172 Question / Response 3.172 [MEB (DSER) Item No. 60] 19.3.3.172-1 19.3.3.173 Question /Respoi..me 3.173 [MEB (DSER) Item No. 61] 19.3.3.173-1 19.3.3.174 Question / Response 3.174 [!IEB (DSER) Item No. 62] 19.3.3.174-1 19.3.3.175 Question / Response 3.175 [MEB (DSER) Item No. 63] 19.3.3.175-1 19.3.3.176 Question / Response 3.176 [MEB (DSER) Item No. 64] 19.3.3.176-1 19.3.3.177 Question / Response 3.177 [MEB (DSER) Item No. 65] 19.3.3.177-1 19.3.3.178 Question / Response 3.178 [MEB (DSER) Item No. 66] 19.3.3.178-1 19.3-x
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 [h ( ,/ SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.4 Chapter 4 - Responses 19.3.4.1-1 19.3.4.1 Question / Response 4.1 (492.1) 19.3.4.1-1 19.3.4.2 Question / Response 4.2 (410.16) 19.3.4.2-1 19.3.4.3 Question / Response 4.3 (490.01) 19.3.4.3-1 19.3.4.4 Question / Response 4.4 (490.02) 19.3.4.4-1 19.3.4.5 Question / Response 4.5 (490.03) 19.3.4.5-1 19.3.4.6 Question / Response 4.6 (490.04) 19.3.4.6-1 19.3.4.7 Question / Response 4.7 (490.05) 19.3.4.7-1 19.3.4.8 Question / Response 4.8 (490.06) 19.3.4.8-1 1 19.3.5 Chapter 5 - Responses 19.3.5.1-1 J 19.3.5.1 Question / Response 5.1 (440.1) 19.3.5.1-1 19.3.5.2 Question / Response 5.2 (440.2) 19.3.5.2-1 19.3.5.3 Question / Response 5.3 (281.01) 19.3.5.3-1 19.3.5.4 Question / Response 5.4 (281.02) 19.3.5.4-1 19.3.5.5 Question / Response 5.5.(281.03) 19.3.5.5-1 19.3.5.6 Question / Response 5.6 (410.17) 19.3.5.6-1 19.3.5.7 luestion/ Response 5.7 (440.01) 19.3.5.7-1 13.3.5.8 Question / Response 5.8 (440.02) 19.3.5.8-1 19.3.5.9 Question / Response 5.9 (440.03) 19.3.5.9-1 19.3.5.10 Question / Response 5.10 (440.04) 19.3.5.10-1 19.3.5.11 Question / Response 5.11 (440.05) 19.3.5.11-1 19.3.5.12 Question / Response 5.12 (440.06) 19.3.5.12-1 n.
%d 19.3.5.13 Question / Response 5.13 (440.07) 19.3.5.13-1 19.3-xi
I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.5.14 Question / Response 5.14 (440.08) 19.3.5.14-1 19.3.5.15 Question / Response 5.15 (440.09) 19.3.5.15-1 19.3.6 Chapter 6 - Responses 19.3.6.1-1 19.3.6.1 Question / Response 6.1 (480.1) 19.3.6.1-1 19.3.6.2 Question / Response 6.2 (281.04) 19.3.6.2-1 19.3.6.3 Question / Response 6.3 (281.05) 19.3.6.3-1 19.3.6.4 Question / Response 6.4 (410.18) 19.3.6.4-1 19.3.6.5 Question / Response 6.5 (460.11) 19.3.6.5-1 19.3.6.6 Question / Response 6.6 (480.01) 19.3.6.6-1 19.3.6.7 Question / Response 6.7 (480.02) 19.3.6.7-1 19.3.6.8 Question / Response 6.8 (480.03) 19.3.6.8-1 19.3.6.9 Question / Response 6.9 (480.04) 19.3.6.9-1 19.3.6.10 Question / Response 6.10 (480.05) 19.2.6.10-1 19.3.6.11 Question / Response 6.11 (480.06) 19.3.G.11-1 19.3.6.12 Question / Response 6.12 (480.07) 19.3.6.12-1 19.3.6.13 Question / Response 6.13 (480.08) 19.3.6.13-1 19.3.6.14 Question / Response 6.14 (480.09) 19.3.6.14-1 19.3.6.15 Question / Response 6.15 (480.10) 19.3.6.15-1 19.3.6.16 Question / Response 6.16 (480.11) 19.3.6.16-1 19.3.6.17 Question / Response 6.17 (480.12) 19.3.6.17-1 19.3.6.18 Question / Response 6.18 (480.13) 19.3.6.18-1 19.3.6.19 Question / Response 6.19 (480.14) 19.3.6.19-1 19.3.6.20 Question / Response 6.20 (480.15) 19.3.6.20-1 19.3.6.21 Question / Response 6.21 (480.16) 19.3.6.21-1 19.3-xii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 15 (s /^) SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.6.22 Question / Response 6.22 (480.17) 19.3.6.22-1 19.3.6.23 Question / Response 6.23 (480.18) 19.3.6.23-1 19.3.6.24 Question / Response 6.24 (480.19) 19.3.6.24-1 19.3.6.25 Question / Response 6.25 (480.20) 19.3.6.25-1 19.3.6.26 Question / Response 6.26 (480.21) 19.3.6.26-1 19.3.6.27 Question / Response 6.27 (480.22) 19.3.6.27-1 19.3.6.28 Question / Response 6.28 (480.23) 19.3.6.28-1 19.3.6.29 Question / Response 6.29 (480.24) 19.3.6.29-1 19.3.6.30 Question / Response 6.30 (480.25) 19.3.6.30-1 19.3.6.31 Question / Response 6.31 (480.26) 19.3.6.31-1 19.3.6.32 Question / Response 6.32 (480.27) 19.3.6.32-1 19.3.6.33 Question / Response 6.33 (480.28) 19.3.6.33-1 19.3.6.34 Question / Response 6.34 (480.29) 19.3.6.34-1 19.3.6.35 Question / Response 6.35 (480.30) 19.3.6.35-1 19.3.6.36 Question / Response 6.36 (480.31) 19.3.6.36-1 19.3.6.37 Question / Response 6.37 (480.32) 19.3.6.37-1 19.3.6.38 Question / Response 6.38 (480.33) 19.3.6.38-1 19.3.6.39 Question / Response 6.39 (480.34) 19.3.6.39-1 19.3.6.40 Question / Response 6.40 (480.35) 19.3.6.40-1 19.3.6.41 Question / Response 6.41 (480.36) 19.3.6.41-1 19.3.6.42 Question / Response 6.42 (480.37) 19.3.6.42-1 19.3.6.43 Question / Response 6.43 (480.38) 19.3.6.43-1 g ,/ 19.3.6.44 Question / Response 6.44 (480.39) 19.3.6.44-1 19.3-xiii
GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.6.45 Question / Response 6.45 (480.40) 19.3.6.45-1 19.3.6.46 Question / Response 6.46 (480.41) 19.3.6.46-1 19.3.6.47 Question / Response 6.47 (480.42) 19.3.6.47-1 19.3.6.48 Question / Response 6.48 (480.43) 19.3.6.48-1 19.3.6.49 Question / Response 6.49 (480.44) 19.3.6.49-1 19.3.6.50 Question / Response 6.50 (480.45) 19.3.6.50-1 19.3.6.51 Question / Response 6.51 (480.46) 19.3.6.51-1 19.3.6.52 Question / Response 6.52 (480.47) 19.3.6.52-1 19.3.6.53 Question / Response 6.53 (480.48) 19.3.6.53-1 19.3.6.54 Question / Response 6.54 (480.49) 19.3.6.54-1 f 19.3.6.55 Question / Response 6.55 (480.50) 19.3.6.55-1 19.3.6.56 Question / Response 6.56 (440.10) 19.3.6.56-1 19.3.6.57 Question / Response 6.57 (440.11) 19.3.6.57-1 19.3.6.58 Question / Response 6.58 (440.12) 19.3.6.58-1 19.3.6.59 Question / Response 6.59 (440.13) 19.3.6.59-1 19.3.6.60 Question / Response 6.60 (440.14) 19.3.6.60-1 l 19.3.6.61 Question / Response 6.61 (440.15) 19.3.6.61-1 19.3.6.62 Question / Response 6.62 (440.16) 19.3.6.62-1 19.3.6.63 Question / Response 6.63 (440.17) 19.3.6.63-1 19.3.6.64 Question / Response 6.64 (440.18) 19.3.6.64-1 19.3.7 Chapter 7 - Responses 19.3.7.1-1 19.3.7.1 Question /Responso 7.1 (420.1) 19.3.7.1-1 19.3.7.2 Question / Response 7.2 (420.2) 19.3.7.2-1 19.3-xiv
GESSAR II' 22A7007 ' 238 NUCLEAR ISLAND Rev. 15
/, ,
k SECTION 19.3 ,y CONTENTS - (Continued) . Section Title Page 19.3.'7.3 -Question / Response'7.3 (420.3) 19.3.7.3-1 19.3.7.4 Q'uestion/ Response 7.'4 (420.4) 19.3.7.4-1 19.3.7.5 Questidn/ Response-7.5 (420.5) 19.3.7.5-1 19.3.7.6 Question /Respo'nse ~[.6 (420.6) 19.3.7.6-1 19.3.7.7 Question / Response 7.7 (420.7) 19.3.7.7-1 19.3.7.8 Question /Rosponse'7.8 (420.8) 19.3.7.8-1 19.3.7.9 Question / Response 7.9 (420.9). 19.3.7.9-1 19.3.7.l~0 Question / Response 7.10 (420.10) 19.3.7.10-1 19.3.7.11 0'uestion/ Response 7.11 (420.11) 19.3.7.11-1 l 19.3.7.12 Question / Response 7.12 ('420.12) 19.3.7.12-1 l 19.3.7.13 Question /Fesponse 7.13 (421.01) 19.'3.7.13-1 l 19.3.7.14 Question /liesponse 7.14 (421.02) 19.3.7.14-1 d - 19.3.7.15 Question / Response 7.15 (42L.0D) 19.3.7.15-1 19.3.7.16 Question / Response 7.16 (421.04) 19.3.7.16 19.3.7.17 Question / Response 7.17 (421.05) 19.3.7.17-1 19.3.7.18 Question / Response 7.18 (421.06) 19.3.7.18-1 19.3.7.19 Question / Response 7.19 (421.0'7) 19.3.7.19-1 Question / Response 7.20 (421.08) 19.3.7.20 19.3.7.20 1
- 19.3.7.21 Question / Response 7.21 (421.10) 19.3.;7.21-1 l
19.3.7.22 ~ Question / Response 7.22 '(421.11) 19.3.~/)22-1 ,
't 19.3.7.23 ' yQuestion/ Response 7.23 (421.12) ;_ 19.3.7123-1 19.3.7.24 Question / Response 7.24 (421.14) 19.3.7.24-1 -
19.3.7.25 Question / Response 7.25 (421.15) , 19.3.7'.25-1 , 19.3.7.26 Question / Response 7226 (421.16) 19.3.7.26-1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 SECTION 19.3 CONTENTS (Continued) Section Title Page 19.3.7.27 Question / Response 7.27 (421.17) 19.3.7.27-1 19.3.7.28 Question / Response 7.28 (421.18) 19.3.7.28-1
' '19L3 7 29
. . Question / Response 7.29 (421.19) 19.3.7.29-1 s 19.3.7.30 Question / Response 7.30 (421.20) 19.3.7.30-1
; 19.3.7.31 Question / Response 7.31 (421.21) 19.3.7.31-1 1).3.7.32 Question / Response 7.32 (421.22) 19.3.7.32-1 19.3.7.33 Question / Response 7.33 (421.23) 19.3.7.33-1 19.3.7.34 Question / Response 7.34 (421.24) 19.3.7.34-1 s
- 19. d'. 7 . 3 5 Question / Response 7.35 (421.25) 19.3.7.35-1 19.3.7.36 Question / Response 7.36 (421.26) 19.3.7.36-1 19.3.7.37 Question / Response 7.37 (421.27) 19.3.7.37-1 19.3.7.38 Question / Response 7.38 (421.28) 19.3.7.38-1 19.3.7.39 Question / Response 7.39 (421.29) 19.3.7.39-1 L 19.3.7.40 Question / Response 7.40 (421.30) 19.3.7.40-1 19.3.7.41 Question / Response 7.41 (421.31) 19.3.7.41-1
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19.3.7.42 Question / Response 7.42 (421.32) 19.3.7.42-1 19.3.7.43 Question / Response 7.43 (421.33) 19.3.7.43-1 19.3.7.44 Question / Response 7.44 (421.34) 19.3.7.44-1 19.3.7.45 Question / Response 7.45 (421.35) 19.3.7.45-1 19.3.7.46 Question / Response 7.46 (421.36) 19.3.7.46-1 19.3.7.47 Question / Response 7.47 (421.37) 19.3.7.47-1 19.3.7.48 Question / Response 7.48 (421.38) 19.3.7.48-1 19.3.7.49 Question / Response 4.49 (421.39) 19.3.7.49-1 l 19.3-xvi
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'I' GESSAR II 22A7007 238 NUCLEAR ISLAND i '/ Rev. 15 0(} ' "
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CONTENTS (Continued) Sectioni, t Title Page 19.3.7.50 Question / Response 7.50 (421.40) 19.3.7.50-1 19.3.7.51 Question / Response 7.51 (421.41) 19.3.7.51-1 19.3.7.52 Question / Response 7.52 (421.42) 19.3.7.52-1 19.3.7.53 Question / Response 7.53 (421.43) 19.3.7.53-1 19.3.7.54 Question / Response 7.54 (421.44) 19.3.7.54-1 19.3.7.55 Question / Response 7.55 (421.45) 19.3.7.55-1 19.3.7.56 Question / Response 7.56 (421.46) 19.3.7.56-1 y,,. // p :/t 19.3.7.57 Question / Response 7.57 (421.47) 19.3.7.57-1 f 's < 19.3.7 58> , , Question / Response 7.58 (421.48) 19.3.7.58-1 19.3.7Sh - Question / Response 7.59 (421.49) 19.3.7.59-1 19.3.7.60 Question / Response 7.60 (421.50) 19.3.7.60-1 lb 19.3.7.6f Question / Response 7.61 (421.51) 19.3.7.61-1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 14 () V 19.3.1.6 QUESTION / RESPONSE 1.6 (460.09) i , QUESTION 1.6 Provide a table in Section 1.8 of your FSAR comparing the design features of the liquid, gaseous and solid radwaste systems with each position of Regulatory Guide 1.43, Revision 1 (October 1979). Justify each position for which an exception is taken. If information is provided in other sections of the FSAR for the individual items, cross-references to these sections are acceptable. We consider compliance with Section C.5 of Regulatory Guide 1.143 to be essential. Verify whether you satisfy our acceptance criteria for concentrations of radioactive constituents in accordance with Item II of Section 15.7.3 of the Standard Review Plan (SRP). Our position is that limiting doses to 0.5 rems, as stated in Section 11.3.2.20 of your FSAR, is not an acceptable alterna-tive. (1.8, 11.2, 11.3 and 11.4) RESPONSE 1.6 The offgas system meets the requirements of Position C.2.1.3 of Regulatory Guide 1.143 by the following: (1) The offgas delay tanks have no natural frequencies between 2 and 35 Hz. (2) The stress in the supports, based on a horizontal static equivalent force of 0.15g is less than 1.33 times the allowable stress level of AISC Manual of Steel Construction 7th Edition, 1970. O 19.3.1.6-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 19.3.1.6 QUESTION / RESPONSE 1.6 (460.09) (Continued) (3) The tanks are located on the base mat of the Turbine Building (4) The Applicant will design the vault containing the Charcoal tanks to protect the tanks from the effects of structural failure and to resist the OBE as specified in Regulatory Guide 1.143, Section 5.2, as permitted in Section 5.3. In addition, the entire Offgas System meets the following requirements: (1) The system design pressure is at a minimum 350 psi and in part is as high as 1000 psi while the system operating pressure is 6.7 psig. h (2) The material of construction is required to demonstrate high notched stress ductility. (3) All pressure retaining butt welds are 100% radiographed. (4) The system must pass a 10 5 atm cc/sec helium leak test. (S i The buildings housirg the Gaseous Radioactive Waste Processing System will be designed in accordance with the Uniform Building Code. A conservative analysis of the dose consequences of failure of this system has been provided in topical reports NEDE-21056-P and NEDE-21056-lP. O 19.3.1.6-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 19.3.3.52 QUESTION / RESPONSE 3.52 (220.11) QUESTION 3.52 At the time of this review, Appendix 3H which describes the effect of the concrete between the containment and the shield building on the seismic analysis, is not available. Indicate when this appendix will be provided. This information should be made avail-able prior to the forthcoming structural audit in December.1982. (3.7.2) RESPONSE 3.52 In the Suppression Pool region of the containment vessel the shell has been stiffened by filling the annulus between the Containment and the Shield Building with reinforced concrete. A seismic dynamic analysis was performed to determine the effects of this added concrete on the seismic responses of various structures in the Reactor Building. These structures include the Shield Building, containment vessel, drywell, shield wall and the RPV pedestal. Specifically, the objective of this analysis was to verify that the original seismic envelope curves used in the plant design envelope the seismic response of the Reactor Building structures with the added concrete. Soil Cases Four soil cases were used in the seismic dynamic analysis for the horizontal ground motion. They are the following: CASE NUMBER DESIGNATION 2 GE-75-A-H2 4 GE-75-VP3 6 GE-75-HR-H2 . 7 GE-FB-H2 (Fixed Base) 19.3.3.52-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 15 Two soil cases were used in the analysis for the vertical motion. h They are the following: CASE NUMBER DESIGNATION 11 GE-75-A-V 12 GE-FB-V (Fixed Base) The case numbers above refer to those listed in Table 3A-1. Mathematical Model The mathematical model originally used for the analysis to develop the design loads and building responses did not include the con-crete added to the region between the containment and the Shield Building below elevation (-) 5 ft 3 in. For this analys, solid elements were added to represent the annular concrete. The rest of the model is similar to that used pre-viously, (See Figure 19.3.3.52-1). The computer program for axisymmetric structures (AXIS) was used in the analysis. Dynamic Analysis The horizontal and vertical analyses were performed separately. Shell forces, shell moments and element stresses were obtained for individual soil cases. These results were then enveloped to arrive at a set of final responses for horizontal and vertical mctions respectively. Tables 19.3.3.52-1 through 19.3.3.52-18 depict the final results. These tabulated values were then com-pared with those in Section 3.7. l Response spectra were generated for the soil cases studied. They were enveloped to arrive at a final set of curves. O 19.3.3.52-2 l
GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 15 () Conclusions A review of the seismic forces indicated that in a few locations they are slightly higher than the envelope (in the range of 3%). However, the resulting stresses are within the allowables. The seismic response spectra for the various structures resulting from this analysis were generally within the envelopes previously generated. Since the containment structure was stiffened by the additional concrete and the Shield Building participation increases in the load distribution, some minor shifting in frequency was observed (in the range of 5 to 10%). The revised response spectra are listed below: Figure Title 3.10-1 RPV Floor Response Spectra Horizontal Acceleration () OBE El 54.00, 2 Percent Damping 3.10-3 RPV Floor Response Spectra Horizontal Acceleration OBE El 17.83, 2 Percent Damping 3.10-7 Shield Wall Floor Response Spectra Horizontal Acceleration OBE, El 43.00, 2 Percent Damping 3.10-19 Containment Responce Spectra Horizontal Acceleration OBE, El 149.00, 2 Percent Damping 3.10-20 Containment Response Spectra Vertical Acceleration OBE, El 149.00, 2 Percent Damping 3.10-21 Containment Response Spectra Horizontal A.cceleration OBE, El 110.83, 2 Percent Damping 3.10-23 Containment Response Spectra Horizontal Acceleration OBE, El 93.20, 2 Percent Damping () 3.10-29 Shield Building Response Spectra Horizontal Acceleration OBE, El 115.90, 2 Percent Damping 19.3.3.52-3
GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 15 O
This figure will be provided in June 1983. O Figure 19.3.3.52-1. Mathematical Model For The Annular Concrete 19.3.3.52-4
O O O Table 19.3.3.52-1 CONTAINMENT SEISMIC FORCE ENVELOPE DUE TO OBE HORI7ONTAL EXCITATIONS Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow i No. (ft) (ft-kips /ft) (ft-kips /ft) (kips /ft) (kips /ft) (kips /ft) l S1 160.75 0.07 0.46 2.97 9.50 5.41 S3 158.81 0.53 0.30 3.13 5.75 3.98 i S5 153.16 2.41 0.91 5.78 11.80 7.66 S7 145.63 0.98 0.73 6.71 24.05 11.79 S9 136.71 19.29 4.70 7.30 44.51 19.59 $ S11 115.70 2.84 0.31 17.47 19.50 34.15 g
- P S13 104.20 0.63 0.32 26.00 16.38 40.64 8@
42.30 13.24 48.35 S15 84.58 0.71 'O.37 hh 5W S17 55.00 0.47 0.31 67.27 12.23 57.48 S19 27.67 4.79 1.13 91.95 3.70 63.71 5U E ll? U
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Table 19.3.3.52-2 CONTAINMENT SEISMIC FORCE ENVELOPE DUE TO OBE HORIZONTAL EXCITATION Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow No. (ft) (ft. kip /ft) (ft. kip /ft) (kips /ft) (kips /ft) (kips / f t) C1 152.00 0.00 0.03 0.11 0.05 0.15 C6 143.22 0.21 0.06 0.64 6.71 1.73 C10 122.00 0.35 0.10 0.75 14.77 4.18 w w 5.93 9.31 m C13 93.20 0.10 0.03 5.24
- C16 55.00 0.00 0.00 12.93 1.44 11.47 $o o to e
w C19 27.67 0.00 0.00 18.54 1.23 12.43 y@ C21 11.00 0.01 0.03 22.08 1.25 12.73 $$ h C23 -5.25 0.41 0.06 19.32 7.40 8.85 yH E C24A 0.00 7.67 0.00 12.44
-15.58 0.02 h O
C25A -23.58 0.01 0.00 11.27 0.00 11.85 C26 -27.58 0.02 0.00 17.96- 0.00 12.37 M131 -31.58 0.09 0.01 20.35 1.86 12.33 l l l
$~
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# 9 e
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f~ 's Q) (/ b Table 19.3.3.52-3 , DRYWELL SEISMIC FORCE ENVELOPE DUE TO OBE HORIZONTAL EXCITATIONS 1 Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow No. (ft) (ft-kips /ft) (ft-kips /ft) (kips /ft) (kips /ft) (kips /ft) D1 75.38 0.23 0.80 5.00 16.39 4.91 D3 67.34 0.01 0.00 0.17 3.68 1.33 DS 57.58 2.52 1.32 2.00 20.67 5.59 D11 57.58 32.69 13.50 46.93 51.68 69.35 h D16 W 36.83 2.12 2.90 137,43 26.82 145.73 'zCO
? D18 20.25 12.23 1.04 204.96 17.93 161.89 -o y l
{ D20 D22 4.83
-11.58 4.47 15.03 4.10 2.23 276.22 344.10 10.26 19.54 172.22 M$
b
, 181.49 gg D23 -19.58 71.04 15.49 384.41 71.51 168.99 MH D24 -27.58 245.92 62.31 417.69 113.45 149.24 2 M91 -31.58 454.93 293.74 162.94 284.64 45.15 t
l if d
Table 19.3.3.52-4 SHIELD WALL SEISMIC FORCE ENVELOPE DUE TO OBE HORIZONTAL EXCITATIONS Long Circ Long Circ Shear Node Elev Mom dom Force Force Flow No. (ft) (ft-kips /ft) (ft-kips /ft) (kips /ft) (kips /ft) (kips /f t) SW1 50.38 0.94 0.06 0.14 4.17 1.39 SW3 36.83 0.10 0.17 3.56 2.33 4.79 SW5 20.25 1.24 0.80 14.18 3.93 12.84 SW7 4.84 0.89 0.42 29.78 6.13 15.50 g
; Ea 90 m
- mm w Table 19.3.3.52-5
$ DRYWELL UPPER POOL AND PEDESTAL MASS CONCRETE HH H
SEISMIC STRESS ENVELOPE DUE TO OBE HORIZONTAL EXCITATIONS Radial Longitudinal Circumferential Shear Shear Shear Element Stress Stress Stress (TIZ} (Trt) (Txt)/ft2) No (kips /ft2) (kips /ft2) (kips /ft2) (kips /ft2) (kips / f t_2)_ (kips 56 0.42 0.55 1.53 0.79 0.09 1.88 57 0.04 0.77 1.06 0.34 0.70 0.96 96 1.54 7.64 1.06 4.78 0.68 1.37 97 2.73 1.60 0.75 2.88 1.61 0.30
$U f5 9 O O
O O O Table 19.3.3.52-6 ANNULAR MASS CONCRETE SEISMIC STRESS DUE TO HORIZONTAL EXCITATIONS Radial Longitudinal Circumferential Shear Shear Shear Slament Stress Stress Stress (Trz) (Trt) (Tzt) No. (kips /ft) (kips /ft) (kips /ft) (kips /ft) (kips /ft) (kips /ft) 267 0.79 6.51 9.36 2.67 4.19 6.11 l 268 1.48 4.59 8.59 0.68 7.01 5.96 243 0.33 8.48 1.04 0.94 0.20 11.60 l 244 0.23 13.56 1.39 1.39 0.23 10.74 213 1.10 12.91 3.75 2.63 0.19 10.95 [ 214 2.11 14.28 4.19 1.57 0.07 10.06 g P 201 2.44 15.21 4.23 8.06 9.91 12.74 @@ 202 6.36 .94 0.72 5.37 1.33 4.61 EE
. :d 30 w
$H
- 6 t3 I
5Y
Table 19.3.3,52-7 WEIR WALL SEISMIC FORCE ENVELOPE DUE TO HORIZONTAL EXCITATIONS
~
Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow No. (ft) (ft-kips /ft) (ft-kips /ft) (kips / f t) (kips /ft) (kips /ft) W1 -5.50 0.56 0.03 0.41 7.54 0.67 W2 -12.00 0.47 0.06 0.35 2.70 1.07 W3 -15.00 0.23 0.04 0.37 1.20 1.52 W4 -18.67 0.39 0.08 0.63 0.99 1.67 w $ ? z 88 . Ma m Table 19.3.3.52-8 >> w WW $ LOWER SHIELD BUILDING MASS CONCRETE gg o SEISMIC STRESS ENVELOPE DUE TO.OBE HORIZONTAL EXCITATIONS mH E O Radial Longitudinal Circumferential Shear Shear Shear Element Stress Stress Stress (Trz) (Trt) (Tzt) No. (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft ) 273 4.12 57.64 3.26 4.24 5.27 27.53 248 0.03 30.44 4.25 0.35 0.05 8.36 218 0.06 29.58 6.44 0.45 0.13 8.98 206 0.50 32.10 7.26 0.78 0.35 9.88 gg k$ s G O O O
O O O Table 19.3.3.52-9 RPV PEDESTAL SEISMIC FORCE ENVELOPE DUE TO OBE HORIZONTAL EXCITATIONS , Long Cire Long Cire Shear ! Node Elev Mom Mom Force Force Flow j No. (ft) (ft-kips /ft) (ft-kips /ft) (kips /ft) (kips /ft) (kips /ft) i P2 -1.33 46.26 99.18 65.13 35.82 30.67 PS -11.58 73.73 3.94 205.28 19.91 46.44 P6 -15.75 175.65 35.86 220.93 37.33 29.09 { P7 -21.00 262.56 88.54 133.95 29.01 10.33 y H P8 -26.29 46.63 *
- 17.56 68.29 8.56 10.03
' 2:
j w M41 -31.58 76.81 106.97 24.44 32.22 2.12 gQ
'w Na b EE A --
3 H ms M
- o 1
1 i i S
<tO HO (JR Y )
1
Table 19.3.3.52-10 UPPER SHIELD DUILDING SEISMIC FORCE ENVELOPE DUE TO OBE VERTICAL EXCITATIONS Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow No. (ft) (ft-kips /ft) (ft-kips /ft) (kips /ft) (kips /ft) (kips /ft) S1 160.75 2.21 9.39 10.99 14.77 0.00 S3 158.81 1.03 0.64 10.00 9.23 0.00 SS 153.16 2.31 1.20 9.46 4.96 0.00 g g S7 145.63 0.92 0.94 7.56 6.87 0.00 $ { S9 136.71 14.48 2.90 5.02 17.49 0.00 @g g Sll 115.70 2.84 0.57 6.22 2.36 0.00 0$ M (n m S13 104.20 0.97 0.19 7.15 0.60 0.00 $@ $ S15 84.58 0.35 0.07 8.91 0.36 0.00 HH g en H S17 55.00 0.14 0.03 11.62 0.87 0.00 S19 27.67 0.58 0.12 13.67 0.95 0.00 o N" 5 O O O
O O O Table 19.3.3.52-11 CONTAINMENT SEISMIC FORCE ENVELOPE DUE TO OBE VERTICAL EXCITATIONS Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow N o. (ft) (ft-kip /ft) (ft-kip /ft) (kips /ft) (kips /ft) (kips /ft) C1 152.00 0.00 0.15 1.38 3.65 0.00 C6 143.22 0.21 0.06 0.77 4.78 0.00 C10 122.00 0.13 0.04 0.46 4.69 0.00 Cl3 93.20 0.02 0.01 1.83 0.64 0.00 C16 55.00 0.00 0.00 2.76 0.07 0.00 -
? C19 27.67 0.00 0.00 3.18 0.19 0.00 m CO u
+ C21 11.00 0.00 0.00 3.38 0.48 0.00 Q@
m C23 -5.25 0.02 0.00 2.67 0.39 0.00 $$ WW w 4 C24A -15.58 0.00 0.00 1.65 0.00 0.00 gg I " "" C25A -23.58 0.00 0.00 2.31 0.00 0.00 C26 -27.58 0.00 0.00 3.19 0.00 0.00 E a
- M131 -31.58 0.01 0.00 3.20 0.14 0.00 i
I
- 45 G$
i l
Table 19.3.3.52-12 DRYWELL SEISMIC FORCE ENVELOPE DUE TO OBE VERTICAL EXCITATIONS Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow N o ., (ft) (ft-kips /ft) (ft-kips /ft) (kips /ft) (kips / f t) (kips /ft) D1 75.38 0.09 0.39 22.03 62.01 0.00 D3 67.34 0.01 0.00 1.27 3.97 0.00 DS 57.58 0.64 1.41 1.02 7.16 0.00 Dll 57.58 4.29 0.35 12.64 1.53 0.00 g H D16 36.83 0.78 0.16 28.17 1.39 0.00 $ gg w D18 20.25 0.69 0.14 31.55 1.18 0.00 w D20 4.83 1.12 0.25 33.94 0.91 0.00 OE mm D22 -11.58 3.41 0.85 35.21 0.13 0.00 gg H " D23 -19.58 2.46 0.61 35.58 3.03 0.00 ss tn s D24 -27.58 4.49 1.12 35.84 7.32 0.00 M91 -31.58 33.33 25.33 13.04 6.53 0.00 o
$U
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O O O
O O O Table 19.3.3.52-13
~
SHIELD WALL SEISMIC FORCE ENVELOPE
, DUE TO OBE VERTICAL EXCITATIONS Long Cire Long Circ Shear Node Elev Mom Mom Force Force Flow No. (ft) (ft-kips /ft) (ft-kips /ft) (kips /f t) (kips /ft) (kips /ft)
SW1 50.38 0.00 0.00 0.33 0.04 0.00 SW3 36.83 0.01 0.00 1.21 0.01 0.00 SW5 20.25 0.00 0.00 3.69 0.18 0.00 SW7 4.84 0.48 0.14 5.07 0.61 0.00 U co e z L 88 L Wl2 i m Table 19.3.3.52-14 $$ i Radial Longitudinal Circumferential Shear Shear Shear Element Stress Stress Stress (Trz) (Trt) (Tzt) h a No. (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft ) 56 0.07 0.11 0.04 0.25 0.00 0.00 57 0.00 0.25 0.03 0.13 0.00 0.00 96 0.07 0.67 0.08 0.26 0.00 0.00 97 0.07 0.06 0.09 0.27 0.00 0.00
$U
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Table 19.3.3.52-15 ANNULAR MASS CONCRETE SEISMIC STRESS DUE TO VERTICAL EXCITATIONS Radial Longitudinal Circumferential Shear Shear Shear Element Stress Stress Stress (1rz) (Trt) (Tzt) No. (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft ) (kips /ft (kips /ft ) 267 0.10 0.91 1.09 0.40 0.00 0.00 268 0.17 0.74 1.10 0.16 0.00 0.00 243 0.00 1.53 0.30 0.12 0.00 0.00 w 244 0.00 1.94 0.21 0.18 0.00 0.00 0 213 0.14 2.13 0.14 0.38 0.00 0.00 $o w a to w 214 0.27 1.95 0.13 0.26 0.00 0.00 g@ m 201 0.26 2.50 0.33 1.34 0.00 0.00 $$ 1 m 202 0.52 0.19 0.30 0.10 0.00 0.00 y[ n O
$w
<3 O O -
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! O O O 1
Table 19.3.3.52-16 WEIR WALL SEISMIC FORCE ENVELOPE DUE TO VERTICAL EXCITATIONS Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow (ft/ kips /ft) (ft-kips /ft) (kips /ft) (kips /ft) (kips /ft) No. (ft) W1 -5.50 0.01 0.00 0.03 0.15 0.00 W2 -12.00 0.04 0.01 0.02 0.07 0.00 W3 -15.00 0.06 0.01 0.05 0.19 0.00 W4 -18.67 0.00 0.00 0.05 U.30 0.00 g oo H
- 2 P
EO i y En m Table 19.3.3.52-17 $g l ws 5 LOWER SHIELD BUILDING MASS CONCRETE "H
- 4 SEISMIC STRESS ENVELOPE DUE TO OBE VERTICAL EXCITATIONS Shear E
o i Radial Longitudinal Circumferential Shear Shear Element Stress Stress (Trz) (Trt) (Tzt) (kips /ft ) (kips /ft ) i No. (kips Stres3)
/ft (kips /ft ) _ (kips /ft ) (kips /ft )
0.40 0.68 0.00 0.00 273 0.69 7.87 3.12 0.08 0.04 0.00 0.00 248 0.00 218 0.00 2.71 0.22 0.02 0.00 0.00 j 206 0.03 2.72 0.29 0.01- 0.00 0.00 5Y
?Y l
l
=
l l I
Table 19.3.3.52-18 RPV PEDESTAL SEISMIC FORCE ENVELOPE DUE TO OBE VERTICAL EXCITATIONS Long Circ Long Circ Shear Node Elev Mom Mom Force Force Flow No. (ft) (ft-kips /ft) (ft-kips /ft) (kips /ft) (kips / f t) (kips / f t) P2 -1.33 3.73 9.58 5.11 5.98 0.00 PS -11.58 3.87 0.97 16.06 2.83 0.00 P6 -15.75 0.90 0.23 16.43 3.31 0.00 P7 -21.00 6.42 3.20 9.54 2.01 0.00 [ $ P8 -26.29 0.27 0.51 6.18 1.01 0.00
- Z F M41 -31.58 7.74 12.36 2.18 5.86 0,00 gQ F MN I
5s H HH O OH E O
?5 9 9 9
GESSAR II 22A7007 238 NUCLEAR ISLAND Res. 15 (O) 19.3.3.54 QUESTION / RESPONSE 3.54 (220.13) QUESTION 3.54 It is not clear in the discussion provided in Sections 3.7.2.3 and 3.7.2.5 of your FSAR how you have accounted for the-vertical flexibility of floors in the generation of the vertical response spectra. Accordingly, provide the procedures you have used to account for this phenomenon. (3.7.2) RESPONSE 3.54
~
An analysis which accouats for the vertical flexibility of floors was performed. The procedures and results are summarized as follows: The Auxiliary Building floor at El. 28'-6" was selected x upon review of existing data which indicated that it will produce the maximum vertical amplification due to floor flexibility. Three typical floor panels were modeled by Spring Dashpot Oscillators; they were then added to the mass point at the floor of interest in the mathematical model of the build-ing. A time-history analysis was performed for the soil case expected to provide the maximum response. Vertical response spectra for the selected floor panels and the main building mass point were generated and a comparison with the current corresponding response spectrum curve at the selected floor was made. Since the correlation was not satisfactory for the vertical l direction, GE will require that the Applicant verify that j (~% the floor response is within the seismic envelope used for j design of equipment, systems and components. 19.3.3.54-1/19.3.3.54-2 _ \
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () 19.3.3.80 QUESTION / RESPONSE 3.80 (220.39) QUESTION 3.80 In Section 3.8.4.3.2.3 of four FSAR, the load combination in Equation 3.8-40 includes the SSE. We believe that you actually intend this load combination to include t11e OBE ] instead of the SSE, si.nilar to the combination presented in Item II.3.b(i)(a) of Section 2.8.4 of the SRP. If this 3 question is in error, correct it. If this equation is not, state why you consi. der this load combination. (3.8.4) RESPONSE 3.80 The response to this question is provided in revised Subsections 3.8.4.3.2.3. O f O 19.3.3.80-1/19.3.3.80-2
,w, - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 19.3.3.82 QUESTION / RESPONSE 3.82 (220.41) (Continued) f- ~) NJ Step 3 l When the factored stress calculated in Step 2 is larger than the allowable stress then reanalyze the support frame for 2 percent damping value and check member-stresses to see if they are within allowable stresses. Seventy-eight cable tray supports were investigated. Seventy-three passed in Step 1, five supports passed in Step 2 and none of the supports require reanalysis as described in Step 3. It is concluded that the design approach is justified. The resulta of comparison of support samples are included in Appendix A of this response. (2) Accommodation of differential seismic displacements k between floors in the cable tray and support design. In all Nuclear Island Buildings, the difference in elevation between typical floors is approximately 17'-6" and the maximum ] relative ~ seismic displacement for 17'-6" of all buildings is 0.20 inch. A vertical cable tray running between two floors with a maximum relative displacement is used as a typical example. It is a 24-in. cable tray running through two floors in the Control Building, with four supports cantilevered out from a concrete wall. The system is modeled as a space frame with four supports. A 0.25-in. relative displacement between the top and bottom supports is inputted and analyzed by the STRUDL computer program. The results show that the differential floor displacement induces a bending stress of 4.33 ksi and 2.42 ksi in the cable tray and support, respectively. Since a rounded-up maximum floor displace-
^ ment of all buildings is conservatively used in the analysis, and ~- the resultant stresses are comparatively low, it is concluded that 19.3.3.82-3
l
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GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 14 1 19.3.3.82 QUESTION / RESPONSE 3.82 (220.41) (Continued) the cable trays and supports can withstand the stresses due to floor displacement without flexible seismic joints at floor pene-trations. Flexible joints are used between buildings. The sample analysis and stress calculations are included in Appendix B to this response. O O 19.3.3.82-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 14 () v 19.3.3.112 QUESTION / RESPONSE 3.112 (271.12) QUESTION 3.112a The FSAR implies that tests and/or analysis are performed on assemblies (the pump and drive motor or the valve and actuator). The detailed descriptions however seem to indicate that motor and actuators are more often tested or analyzed separate from the pump cr valve. The FSAR should state precisely whether or not equipment was tested or analyzed as an assembly or as individual components. RESPONSE 3.112a It is the Applicant's responsibility to specify (via the qualification report) as to the method of qualification. This includes whether or not the equipment was tested or I ) analyzed as an assembly or as individual components.
%../
QUESTION 3.112b The sections of the FSAR which present the requirements for documentation of pumps and valves are different in scope. Justify why there is a difference in the documentation requirements between pumps and valves. RESPONSE 3.112b There are no actual differences in documentation requirements between pumps and valves. The documentation requirements for valves have been revised to be the same as those for pumps. 19.3.3.112-1
\.
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s
' GESSAR II i
238 NUCLEAR ISLAND 22A7007' Rev. 1S t ., 4 5 N s s. N, 19.3.3.112 QUESTION / RESPONSE 3.112 (271.12) (Continued)
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QUESTION 3.112c In some instances more quantitative and qualitati've' details are needed in order to understand the intent of a'particular FSAR section. Phrases or terms are used which are unclear or subjective. Phrases such as " operation combinations,." '.'as ~ many motor starts ias possible," should be defined or' quantified.i Such phrases dp sy . pot. convey the informrtion needed to form in~'\ (I] opinion with regard to the acceptability of the approach - presented.
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7 .u '. \ ' RESPONSE 3.11 5 N s '. c s ',, Subsection 3.9' 3.'2.lD,'has,been revised to eliminate these 3
. phrases.
s
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=
19.3.3.112-2
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O ct- \ ' GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15
) ,-
1 .3.3'127 QUESTION / RESPONSE 3.127 [MEB (DSER) ITEM NO. 15]
.~
l QUESTION 3.127
+' The design stress and fatigue limits for class 1 piping in the containment penetration areas are not in compliance with Str.ndard Review Plan 3.6.2 and BTP MEB 3-1. If the maximum
~~'
stress range of Equation (10) exceeds 2.4 Sm, both Equations ( , (12) and (13) must be less than 2.4 Sm. The cumulative usage i
'N (['- fa'ctor must he less than 0.1 even if Equation (1) is less i Ithan 2.4 Sm.-
W I ' RESPONSE 3.127 i' '
- i t.( , . .
, , T1;e response to this question is provided in the revised _
s , ~ , , . subsections 3.6.2.1.4.2 and 3.6.2.1.4.3 which complies with'SRP 3.6.2 and BTP MEB 3-1. Q e f g
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19.3.3.127-1/19.3.3.127-2 3 . l\ i . . - -
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d CiESSAR II 22A7007 e
'g .,'., 238 NUCLEAR ISLAND Rsv. 14
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/
i l19. 3. 3 l139 QUESTION / RESPONSE 3.129 [MEB (DSER) ITEM NO. 27] f, , l (jQUSSTION3.139 ,j i } s. , p. Standard Review Plan requires the review of sketches showing i the locations of the postulated pipe ruptures, including identification of longitudinal and circumferential breaks, structural barriers, if any, restraint locations, and the constra'ined directions in each restraint. Also to be reviewed are the. data developed to select postulated break locations including, for each point, the calculated stress intensity, the calculated cumulative ~ age factor, and the calculated primary plus secondary stress range. Provide these sketches '\ and data in the FSAR. RESPONSE 3.139 i As noted in the 3/31/81 GESSAR FDA submittal letter from ~ G. G. Sherwood (GE) to H. R. Denton'(N9.C), one of the NRC
'1 . concerns regarding FDAs includes anti-trust considerations.
The issue is the potential problem of whether the level of i detail required by the Staff during the FDA review will ; dictate the use of particular equipment vendors in CP/OL applications that reference the FDA. This is of primary I concern particularly in the " buy" area which is mostly outside the NSSS. General Electric has elected to be responsive to anti-trust concerns by identifying specific quantities as " Applicant to Supply" for those quantities which are equipment vendor dependent. 19.3.3.139-1 v , , - . - -.
,- n- - -.e.--- - ,n, . , . ,n,-n,-,- - -- - - - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 19.3.3.139 QUESTION / RESPONSE 3.139 [MEB (DSER) ITEM NO. 27] h (Continued) Further, General Electric has also elected to have tita hydrodynamic loads handled on a site specific basis. Thus, even if anti-trust was not an issue, many of the equipment and piping quantities would not be available generally and these quantities would have to be cupplied by the Applicant. In summary, the sketches and data requested by Question 3.139 and the information requested by Question 3.154 cannot be supplied at this time, and must be suppi<.ed by the Applicant _ in his FSAR. O O 19.3.3.139-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 19.3.3.145 QUESTION / RESPONSE 3.145 [MEB (DSER) ITEM NO. 3?]
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QUESTION 3.145 Provide the basis for asnuring that the feedwater isolation check valve can perform its function following a postulated pipe break of the feedwater line outside containment. RESPONSE 3.145 The design criteria for the feedwater pipelines in the steam tunnel assure that the structural integrity is maintained. The maximum stress occurring between the two isolation check valves is limited to 2.25 Sm. The restraint for a postulated pipe break outside containment is located so that a plastic hinge is not produced, assuring that the desired action of the check valve can occur. The design specification identifies () these checks as isolation valves and requires the supplier to provide operability assurance to be verified by testing. These valves are specified to be selected for non-slam characteristics. The leak integrity of the check valves shall be demonstrated by test or analysis under most severe ] l operating transient conditions. The valve specification requires the feedwater check valves to be ANSI Class 900, tilting disc. The tilting disc check valve is designed to close as quickly as possible to minimize the slamming that is caused when the high velocity reverse flow is allowed to build up before the completion of closing. The disc begins to close before the
- flow reverses and has completed the closing before the I velocity has built up to a dangerous level.
l0 ,V 19.3.3.145-1/19.3.3.145-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 15 () 39.3.3.14e QUESTION / RESPONSE 3.148 [MEB (CSER) ITEM NO. 36] ] QUESTION 3.148 The number of bolt up and unbolt cvents is listed in Table 3.9-1 as 40 each. This is a redu tion from abaut 120 1.isted in past FSARs by GE. Explain the reason for this reduction in the number of cycles considered for thes}}