LD-90-046, Forwards Addl Copies of Amend G to CESSAR-DC
| ML20055H289 | |
| Person / Time | |
|---|---|
| Site: | 05200002, 05000470 |
| Issue date: | 07/12/1990 |
| From: | Erin Kennedy ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PROJECT-675A LD-90-046, LD-90-46, NUDOCS 9007260010 | |
| Download: ML20055H289 (9) | |
Text
{{#Wiki_filter:__ ___ ABB ASEA BROWN BOVEFil July 12, 1990 LD-90-046-p Project No. 675 U..S. Nuclear Regulatory Commission Attnf Document Control Desk
. Washington, D.C. -20555
Subject:
Combustion Engineerin Standard Safety Analysis { , Report - Design Certi ication, Amendment G O Referonce: Letter LD-90-031, E. H. Kennedy (C-E) to , T. E. Murley (NRC), dated April 30, 1990 m
Dear Sirs:
[ A copy of Amendment G to-the Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC)-was submitted prevjously to initiate an early review of that material (Reference). This letter transmits the thirty-seven (37) formally printed copies of Amendment G and the affidavit, as required by 10CFR50.4(b) and 10CFR50.30(b). If you have any questions, please call me or Mr. S. E. h Ritterbusch of my staff at (203) 285-5206.
- g. Very truly yours,
_ COMBUSTION ENGINEERING, INC. s
/
, . /92/' --
E E. H. K nn dy Manager 5 c-- Nuclear Systems icensing r EHK:lw / s
+ ,
Enclosures:
As Stated y cc: . W/o
Enclosures:
i o C. Miller (NRC) O
=
z[c$ T. Murley (NRC I
-g c4 F. Ross (DOE - Germantown)
T. Wambach (NRC) (g% F 3gk rw17() g
$9 n.a..o ABB Combustion Engineering Nuclear Power Combustion Engtneering Inc. Prospt, t H Road Te e ne (203) 6881911 Windsor. Connectcut 06095-0500 TeleA 99291 COMBEN WSOR
e UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
$l In the Matter of: ) $o9
) le Combustion Engineering, Inc. ) $$
) y Standard Plant Design )
i APPLICATION FOR REVIEW OF j
" COMBUSTION ENGIMEERING STANDARD SAFETY ANALYSIS REPORT -
DESIGN CERTIFICATION Shelby T. Brewer, being duly sworn, states that h9 is the President, ABB Combustion Engineering Nuclear Power, of Combustion Engineering, Inc.; that he is authorized on the part j of said corporation to sign and file with the Nuclear Regulatory Commission this document; and that all' statements made and matters' set'forth therein are true and correct to the best of his knowledge, information and belief. COMBUSTION ENGINEERING, INC. Y By: Shelby T. Brewer President ABB Combustion Engineering Nuclear Power Subscribed and swor to before me this # day of 0 telv. , 1990 .
#. di (UA1J - f L.)
~ Notary Publgc My Conwission Expires: 3 S/ y
)
ABB ASEA BROWN BOVERI - 1 1 0 l July 12, 1990 LD-90-046 l j Project No. 675 J U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Combustion Engineering Standard Safety Analysis Report - Design Certification, Amendment G
Reference:
Letter LD-90-031, E. H. Kennedy (C-E) to T. E. Murley (NRC) , dated April 30, 1990
Dear Sirs:
1 A copy of Amendment G to the Combustion Engineering Standard Safety Analysis Report Design Certification (CESSAR-DC) was submitted previously to initiate an early review of that
. material (Reference). This letter transmits the thirty-seven -
(37) formally-printed copies of Amendment G and the affidavit, as required by 10CFR50.4 (b) and 10CFR50.30(b). If you have any questions, please call me or Mr. S. E. Ritterbusch of.my staff at-(203).285-5206. Very truly yours, COMBUSTION ENGINEERING, INC.
}$ / .,, , ,
E. H. Kennedy Manager / Nuclear Systems Licensing EHK:lw'
Enclosures:
- As Stated cc: w/o
Enclosures:
C. Miller (NRC) , T. Murley (NRC F. Ross (DOE - Germantown) T. Wambach (NRC) a ABB Combustion Engineering Nuclear Power Comt:uston Ereneenna inc 1000 Proscect Hol Road Te;echene (203) 6881911 Post Omce Boe 500 Fax (203) 285-9512 Winaso . Ccnnechcut 06095 0500 Tee 4 9929'l COYBEN WSOR
31 Page 1 of 6 JCESSAR-DC. AMENDMENT G
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- . INSTRUCTIONS FOR INSERTION
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^i FRONT MATTER Remove:
Listing of Amendments, Amendment E Insert: { New Listing of Amendments, Amendment G IQLUE_ft l TAB 6.0 3 r Remove: ;
. Effective Page Listing (8 sheets)
{ Table of Contents, List of Tables, List of Figures (pages i through xxix) r Insert: i New Effective Page Listing (Sheets 1 through 13) New Table of Contents, List of Tebles, List of Figures (pages i , through xxix) :
(:s -< Page 2 of 6'. TAB 6.2- ! EtmRY1: > Section 6.2.1 in its entirety (See Note *] (i .e., Pages 6.2-1 through 6.2-28, Tables 6.2.1-1 through 6.2.1-38,- Figures 6.2.1-1 through 6.2.1.26) Insert: i
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New Section 6.2.1 l TAB 6.3' Remove: Page 6.3-21 Insert: - New pages 6.3-21 and 6.'3-22 VOLUME'7 Remove:
;1 i Pages 6.3-22'through 6.3-26 [See Note *]
i Tables 6.3.3.2-1 through 6.3.3.2-6 (See Note *] Figures 6.3.3.2-1A through 6.3.3.2-11 [See Note *] Insert: t I~ L New pages 6.'3-23 through 6.3 New Tables 6.3.3.2-1 through 6.3.3.2-6 " New Figures 6.3.3.2-1A through 6.3.3.2-11 L L ) L 1-
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TAB A:. CLOSURE OF UNRESOLVED.AND GENERIC SAFETY ISSUES 1
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Pages A-11.and A-12 Insert: 3 New pages A-11 through A-12 Remove: l t i I Pages A-45 through A-48 1 Insert: 1 l New pages A-45-through A-48 l Remove:
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New pages A-63 through A-64 - i New pages A-66a through A-66d, after page A-66 ' New pages A-100a through A-100d, after page A-100 J; New pages A-102a through A-102d, after'page A-102 f
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1 Pages A-123 and A-124 , Insert: s New pages.A-123 through A-124 - New pages A-150a through A-150h, after page 'A-150 New pages A-159 through A-164, after page A-158 l i
* ' Some CESSAR-DC sets include CESSAR-F material. [CESSAR-F material is easily identified: amendments are numbered; for CESSAR-DC, amendments are lettered.] If you do not have CESSAR-F material, just skip this instruction step.
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O TCESSAR !!n1Picans i LISTING OF AMENDBENTS Amendment No.- Date l I i A September 11, 1987 l i B March 31, 1988- ; 1 C June 30, 1988' , i D- September 30, 1988 E December 30, 1988 F December 15, 1989 G- April 30, 1990 -F i e i f 0 Amendment G April 30, 1990
-kCDTDYW Shlff'
CESSART#tuh m (Shoot-1 of-13) i ~ EFFECTIVE PAGE LISTING CHAPTER 6 ; Table of Contents . Engt Amendment i ' 11 G iii G iv G { v E vi E ; vii G viii D- ;
-ix- E ,
x E ; Xi -E xii G xiii- G xiv xv E
,xvi G !
xvii -G ! xviii' G xix C
.xx xxi xxii G ;
-xxiii G '
xxiv
.xxv 3 xxvi xxvil '
xxviii xxix D 113.t Rags Amendment 6.1-1 D 6.1-2 6.1-3 D
-6.1-4 D 6.1-5 D 6.2-1 G e
Amendment G April 30, 1990
1 CESSAR?!nL m - (Shant 2'of 13). EFFECTIVE PAGE LISTING- (Cont'd) CHAPTER 6 lagt (Cont'd) Enga amendment
'6.2 -
G 6.'2-3 G1 r 6.2-4 G 6.2-5 G 6.2-6 G ; 6.2-7 G ; 6.2-8 G i
'6.2-9 G 6.2-10 G 6.2-11 G 6.2-12 G '
6.2-13 'G 6.2-14 G 6.2-15 G 6;2-16 G - 6.2-17 G ' 6.2-18 G - 6~.2-19 G 6.2-20 G 6.2-21 G 6.2-22 G 6.2-23 G-6.2-24 G 6.2-25~ G 6.2-26 G 6.2-27 G i
.6.2-28 G 6.2-28a G 6.2-28b G 6.2-28c G 6.2-28d G
'6.2-28e G 6.2-28f G 6.2-29 E 6.2-30 E 6.2-31 E 6.2-32 E 6.2-33 E 6.2-34 E 6.2-35 E 6.'2-36 E 6.2-37 E 6.2-38 E Amendraent G April 30, 1990
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- CESSAR Einineui: (Sh$$t 3 of 13) 1 1
1 Y R- l O l EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 6 l Text (Cont'd) 1 Page Amendment J i
'6.2 E 6.2-40. E 6.2-41 E.
6.3 C 'I
'6.3-2 C /
6.3-3 E 6.3-4 C 6.3-5 C 6.3-6 -E ) 6.3-7' C
.6.3-8~ C 6.3 C b 6'.3-10 C- i 6.3-11 E !
6.3-12 E j l l' 6.3-12a E 6'3-13
. E 6.3 C 6.3 E l 6.3-16 E 6.3-17 E 6.3-18 E.
6.3-19 C 6.3-20 C 6.3 C 6.3-22' 6.3-23 G 6.'3-24 G ' 6.3-25 G 6.3-26 G 6.3-27 10
~6.3-28 4 6.3-29 6.3-30 10 6.3-31 1 6.3 4 l; 6.3-33 9 6.3-33a 9 6.3-34 9 6.3-35 9
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6.3-36 6.3-37 l l
- 6.3-38 Amendment G ,
April 30, 1990 l L
CESSARSINLm.
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o (Shaat 4 of 13)' EFFECTIVE PAGE LISTING (Cont'd) QEAPTER 6
-T331 (Cont'd) haendment 4
6.3-39 6.3-40a- G 6.3-40b 4 6.3-40c G 6.3-40d G 6.3-40e G 6.3-41 D 6.3-42 E-
'6.3-43 E 6.3-44 D 6.4-1 .E 6.4-2 'E 6.4-3 E 6.4 E 6.5-1 E 6.5-2 E 6.5-3' E ..
6.5-4 E 6.5-5 E 6.5-6 E 6'.5-7 E
-6.5 E-6.5-9 E 6.5-10 -E 6.5-11 E 6.5-12 -E 6.5-13 E 6 5-14 E 6.5-15 E 6.5-16 E 6.-5-17 E 6.6-1 D 6.6-2 E 6.6-3 E 6-7-1
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- : '6.7-2 D 6.7-3 D 6.7-4 D 3.7-5 D 6.7-6 D 6.7-7 D 6.7-8 D 6.7-9 D 6.7-10 D 6.7-11 D Amendment G April 30, 1990
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.CESSART!Encm1= (Sh$$t 5 of 23) s f'N EFFECTIVE PAGE LISTING (Cont'd) '
CHAPTER 6 . [ Text (Cont'd) naapdaent ' 6.7-12. D-6;7-13 D . I Tables Amendment 6.1-1 (Sheet 1) ,
-6.1 1 (Sheet 2) '
6.1 D 6.1-3 6 .' l- 4 D 6.2.1-1 (Sheet 1) G .} 6.2.1-1 (Sheet 2) G ' 6.2.1-l' (Sheet 3) G 6.2.1-2 'G 1;
'6.2.1-3 G 6.2.1-4-(Sheet 1) G.
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6.2.'l-4 (Sheet 7) G 6.2.1-4 (Sheet 8)- G 6.2.1-4'(Sheet 9) G l 6.2.1-5 (Sheet 1) G 6.2.1-5 (Sheet'2)' _G 6.2.1-5 (Sheet 3) -G 6.'2.1-5 (Sheet 4) .G 6.2.1-5 (Sheet 5) G q 6.2.1-5 (Sheet 6) G 6.2.1-5 (Sheet 7). G 6.2.1-5 (Sheet 8) G 6.2.1-5 (Sheet 9) G 6.2.1-6 (Sheet-1) G 6.2.1-6.(Sheet 2) G
'6.2.1-6 (Sheet 3) G 6.2.1-6 (Sheet 4) G 6.2.1-6 (Sheet 5) G 6.2.1-6 (Sheet 6) G 6.2.1-6 (Sheet 7) G 6.'2.1-6 (Sheet 8) G 6.2.1-6 (Sheet 9) G C 6.2.1-6-(Sheet 10) G . ! ,. 6.2.1-6 (Sheet 11) G 6.2.1-6 (Sheet 12) G Amendment G April 30, 1990
CESSAR !!nific.m . '(Shast 6.of 13) EFFECTIVE PAGE LISTING (Cont'd). CHAPTER 6 Tables (Cont'd)- Amendment 6.2.1-6 (Sheet 13) G 6.2.1-6 (Sheet 14) G 6.2.1-7 (Sheet 1) G 6.2.1-7 (3heet 2) G 6.2.1-7 (Sh'. set 3) G 6,2.1-7 (Sheet-4) G 6.2.1-7 (Sheet 5) G 6.2.1-7 (Sheet 6) G 6.2.1-7 ' Sheet 7). G 6.2.1-7 (Sheet'8) G 6.2.1-7 (Sheet 9) G 6.2.1-7 (Sheet 10) G 6.2.1-7 (Sheet 11) G
-6.2.1-7 (Sheet'12) G 6.2.1-7 (Sheet 13). G 6.2.1-7 (Sheet 14) G 6.2.1-8 (Sheet 1) G 6.2.1-8 (Sheet 2) G
'6.2.1-8 (Sheet 3). G 6.2.1-8 (Sheet 4) G 6.2.1-9 (Sheet 1) G 6.2.1-9 (Sheet 2) G 6.2.1-9 (Sheet 3). G 6.2.1-9 -(Sheet 4) G 6'2.1-9 (Sheet 5)
. G 6.2.1 (Sheet 6) G 6.2.1-9 (Sheet 7) G 6.2.1-10 (Sheet'1) G 6.2.1-10 (Sheet 2) G 6.2.1-10 (Sheet 3) G 6.2.1-10 (Sheet 4). G 6.2.1-10 (Sheet 5) G 6.2.1-10 (Sheet 6) G 6.2.1-10 (Sheet 7) G 6.2.1-11 (Sheet 1) G 6.2.1-11 (Sheet 2) G 6.2.1-11 (Sheet 3) G 6.2.1-11.(Sheet 4) G 6.2.1-11 (Sheet 5) G 6.2.1-11 (Sheet 6) G
-6.2.1-11 (Sheet 7) G 6.2.1-12 (Sheet 1) G 6.2.1-12 (Sheet 2) G 6.2.1-12 (Sheet 3) G 6.2.1-12 (Sheet 4) G Amendment G Apr.tl 3 0, 1990
H i j CESSAR !!!Oicam - (Sh20t 7 of 13) i l Ii [f*N l Y, 1 EFFECTIVE PAGE LISTING (Cont'd). CHAPTER 6 ! l l Tables (cont'd) haendment l 6.2.1-12 (Sheet 5) G 6.2.1-12 (Sheet 6) G ! 6.2.1-12 (Sheet 7) G 6.2.1-13 (Sheet 1) G ' 6.2.1-13 (Sheet 2) G 6.2.1-13 (Sheet 3) G~ i 6.2.1-13 (Sheet 4) G 6.2.1-13 (Sheet 5) G 6.2.1-13 (Sheet 6) G 6.2.1-13'(Sheet 7) G 6.2.1-13 (Sheet 8) G 6.2.1-13 (Sheet 9) G 6.2.1-14 -(Sheet 1) G-6.2.1-14 (Sheet 2) G ' 6.2.1-14 (Sheet 3) G 6.2.1-14 (Sheet 4) G !
. 6.2.1-14 (Sheet 5) G
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- 6. 2.1 (Sheet 6)- G
\ 6.2.1-14 (Sheet 7) G ,
6.2.1-14 (Sheet 8) -G 6.2.1-14 (Sheet 9) G 6.2.1-15 (Sheet 1) G ; 6.2.1-15 (Sheet 2)- G 6.2.1-15 (Sheet 3) G 6.2.1-15 (Sheet 4) G 6.2.1-15 (Sheet 5) G 6.2.1-15 (Sheet 6) G 6.2.1-15 (Sheet 7) G
.6.2.1-16 (Sheet 1) G i 6.2.1-16 (Sheet'2) G 6.2.1-16 (Sheet 3) G 6.2.1-16 (Sheet 4) G 6.2.1-16 (Sheet 5) G
-6.2.1-16 (Sheet 6) G 6.2.1-16 (Sheet 7) G 6.2.1-17 (Sheet 1) G 6.2.1-17 (Sheet 2) G 6.2.1-18 G 6.2.1-19 (Sheet 1) G 6.2.1-19 (Sheet 2) G 6.2.1-20 G 6.2.1-21 (Sheet 1) G 6.2.1-21 (Sheet 2) G O, 6.2.1-21 (Sheet 3) G 6.2.1-21 (Sheet 4) G Amendment G April 30, 1990
CESSAR !!Nace 2,. (Shaat 8 of 13) O, EFFECTIVE PAGE LISTING (Cont'd) ' CHAPTER 6
-l Tables l(cont'd) Amendment ;
1 6.2.1-21,'(Sheet 5). G l 6.2.1-21_(Shuet 6) G L 6.2.1-21 (Sheet 7) G i 6.2.1-22 G. l
.6.2.1-23 (Sheet 1) G !
- 6. 2.1-2 3 . (Sheet 2) G l (Sheet 3)
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'6.2.1-23 G 6.2.1-23 (Sheet 4) G [
6.2.1-23'(Sheet 5)- G' l 6.2~.1-23 (Sheet 6) G 6.2.1-23 (Sheet 7) G 6.2.1-23 (Sheet 8) G , 6. 2.1-2 3 ' (Sheet- 9)- G l 6.2.1-23 (Sheet-10) G 6.2.1-23 -(Sheet 11) G 6.2.1-23 (Sheet 12) G I
- 6. 2.1 (Sheet - 1) G 6.2.1-24 (Sheet 2) G -
6.'2.1-24 (Sheet 3) G -
'6.2.1-24 (Sheet'4) G 6.2.1-24 (Sheet 5) G 6.2.1-24 (Sheet 6) G -;
-6.2.1-24 (Sheet 7) G 6'.2.1-24 (Sheet-8) G 6.2.1-24 -(Sheet 9) .G .;
6.2.1-24 (Sheet 10) G ; 6.2.1-24 _(Sheet 11) G-6.2.1-24-(Sheet 12) G ; 6.2.1-24 (Sheet 13) G 3 6.2'.1-24 (Sheet 14) G
- 6. 2.1 (Sheet 15) G 6.2.1-25 G 6'.2.1-26 (Sheet 1) G 6.'2.1-26 (Sheet 2) G !
'6.2.1-26 (Sheet 3) G 6.2.4-1 E 6.3.2-1 (Sheet 1) E 6.3.2-1 (Sheet 2) E 6.3.2-1 (Sheet-3) E (Sheet 1)
~6.3.2-2 E 6.3.2-2 (Sheet 2) E 6.3.2-2 (Sheet.3) E 6.3.2-2 (Sheet 4) E -
6.3.2-2 (Sheet 5) E 6.3.2-2 (Sheet 6) E Amendment G April 30, 1990
CESSARi!n%uc. (Shoot 9 of:13) .
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EFFECTIVE PAGE LISTING (Cont'd) , CHAPTER 6 Tablea-- (Cont'd) Amendment 6.3.2-2 (Sheet.7)- E
.6.3.2-2 (Sheet 8) .E !
6.3.2-2 (Sheet'9) E 6.3.2-3 C 6.3.2-4 C ' 6.3.'3.2-1 G 6.3.3.2-2 G 6.3.3.2-3 ! 6.3.3.2-4 G 6.3.3.2-5 G j 6.3.3.2-6 G 6.3.3.3 t 6.3'.3.3-2 4 6.3.3.3-3 6.3.3.3-4 , 6.3.3.3-5 4 l 6.3.3.3-6 10 O-D 6.3.3,5-1 (Sheet 1) 6.3,3.5-1:(Sheet 2) 10
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6.3.3.5-2-(Sheet 1) 3 l 6.3.3.5-2 (Sheet 7) !, 6.3.3.5-3 ..; l 6.3.3.6-1 6.3.3.7-1 1 6.5-1 (Sheet 1) E 6.5-1 (Sheet 2) E 6.5-1 (Sheet 3) E 6.5-2 E l
._6.5-3 (Sheet 1) E 6.5-3 (Sheet 2) E 6.5-3 (Sheet 3) E s
- 6. 5-3l- (Sheet - 4 ) E 6.5-3 (Sheet 5) E 6.7-1 D 6.7-2 D 6.7-3 (Sheet 1) D !
6.7-3 (Sheet 2) D Ficures Amendment 6.2.1-1 (Sheet 1) G ;
.D: 6.2.1-1 (Sheet 2) G i ? 6.2.1-2 (Sheet 1) G 6.2.1-2 (Sheet 2)
G Amendment G April 30, 1990
-CESSARin b re. (Shsst 10'of 13)
Q EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 6 Figures (Cont'd) haendment 6.2.1-3 (Sheet 1) G
- 6. 2 .'l-3 (Sheet 2) G 6.2.1-4 (Sheet 1) G 6.2.1-4 (Sheet 2) G 6.2.1-5 (Sheet 1) G 6.2.1-5 (Sheet 2) G 6.2.1-6 (Sheet 1). G
-6.2.1-6 (Sheet 2) G 6.2.1-7 (Sheet 1). G.
6.2.1-7 (Sheet 2) G 6.2.1-8 (Sheet 1) G 6.2.1-8 (Sheet 2) G 6.2.1-9 (Sheet 1) G 6.2.1-9 (Sheet 2) G 6.2.1-10 (Sheet 1) G 6.2.1-10 (Sheet 2) G 6.2.1-11 (Sheet 1) G 6.2.1-11 (Sheet;2) G : 6.2.1-12 (Sheet 1)- G -
; 6 '. 2 .1-12 (Sheet 2) G 6.2.1-13 (Sheet 1) G 6.2.1-13 (Sheet 2) G
-6.2.1-14 G 6.2.1-15 G 6.2.1-16 G
- 6. 2.1 (Sheet 1) G 6.2.1-17 (Sheet 2) G 6.'2.1-18 G 6.2.1-19 G 6.2.1-20 G 6.2.1-21 G 6.2.1-22 G 6.2.1-23 (Sheet 1) G 6.2.1-23 (Sheet 2) G 6.2.1-24 (Sheet 1) G 6.2.1-24 (Sheet 2) G 6.2.1-25 (Sheet 1) G 6.2.1-25 (Sheet 2) G 6.'2.1-2 6 ' (Sheet 1) G 6.2.1-26 (Sheet 2) G 6.2._-27 (Sheet 1) G 6.2.1-27 (Sheet 2) G 6.2.1-28 (Sheet 1) G -
6.2.1-28 (Sheet 2) G 6.2.1-29 (Sheet 1) G 1 Amendment G l April 30, 1990 1
I e CESSAR !!! Gem: (sh=$t 11 or 13) t l'h L Q-EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 6 l Ficrureg. (Cont'd) Amendment ( 6'.2.1-29 (Sheet 2) G l 6.2.1-30 (Sheet 1) G ; s 6.2.1-30 (Sheet 2) G l 6.2.1-31 G 6.2.1-32 G 6 . 2.1-33 G 6.2.1-34 G , 6.2.1 G i 6.2.1-36 G 6.2.4-1A ' 6.2.4-1B 10 6.2.4-1C 6.3.2-1A C 6.3.~2-1B C l
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G G Amendment G April 30, 1990
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.l CESSAR ML",e.m.
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(Shsst 12 of 13)' L EFFECTIVE PAGE LISTING (Cont'd) . CHAPTER 6 Figures (Cont'd) Amendment q 6.3.3.2-3G G-6.3.3'.'2-3H~ G ' 6.3.3.2-4A G-6.3.3.2-4B
, G 6.3.3.2-4C G 6.3.3.2-4D.1 G 6.3.3.2-4D.2' G- ,
' 6 .' 3 '. 3 . 2 - 4 E G l 6.3.3.'2-4F G 6.3.3.2-4G G 6.3.3.2-4H G 6.3.3.2-5A' G 6.3.3.'2-5B- G 16.3'.3.2-5C G-6.3.3.2-5D.1 G j 6.3.3.2-5D.2 G
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6.3.3.2-5F G .' 6.3.3.2-5G G - 6.3.3.2-5H G 6.3.3.2-5I-: G 6.3.3.2-5J G
'6.3.3.2-5K G 6.3.3.2-5L G i 6.3.3 2-SM . ~G 6.3.3.2-5N G i 6.3.3.2-50 G 6.3.3.2-5P G fl 6.3.'3.2-5Q' G 6.-3.3.2-5R G 6.3.3.2-5S G ,
6.3.3.2-6A G 6.3.3~.2-6B G i 6.3.3.2-6C G _; 6.3.3.2-6D.1 G 6.3.3.2-6D.2 G 6.3.3.2-6E G ! 6.3.3.2-6F G 6.3.3.2-6G G 6.3.3.2-6H G 6.3.3.2-7A G 6.3.3.2-7B G 6.3.3.2-7C G - 6.3.3.2-7D.1 G - 6.3.3.2-7D.2 G Amendment G l April 30, 1990
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LCESSARE!nLm. (Shsst 13 of 13) q dsc.. EFFECTIVE PAGE LISTING (Cont'd) I CHAPTER 6 EiSDM:.41 (Cont'd). Amendment 6.3.3'.2-7E G J 6.3.3.2-7F G I 6.3.3.2-7G G 6.3.3.'2-7H G- ? 6.3.3.2-8A' G 6.3.3.2-8B G 6.3.3.2-8C G 6.3.3.2-8D.1 G 6.3.3.2-8D.2 G 6.3.3.2-8E G 6.3.3.2-8F G 6.3.3.2-8G G 6.3.3.2-8H G 6.3.3.2-9A G -l 6.3.3.2-9B G 6,3.3.2-9C G 6.3.3.2-9D.1 G t ! 6.3.3.2-90.2 G L 6.3.3.2-9E G- I 6.3.3.2-9F G 6.3.3.2-9G G
'6.3.3.2-9H G
-6.3.3.2-10 G i 6.3.3.2-11 G '
6.3.3.3 (Sheets lA-lH)- l (Sheets 2A-2H) i (Sheets 3A-3E) l (Sheet 3F) 4 (Sneets 3G and 3H) (Sheets 4A-4H) i (Sheets'5A-5H)
'(Sheets 6A-6H)
(Sheet 7) 6.3.3.4-1 9 ;) 6.3.3.~4-2 6.3.3.4-3 6.3.3.4-4 ) 6.3.3.4 4 -l 6.3.3.4-6 6.3.3.5 (Sheets lA-lF) 6.7-1 D l c0 Amendment G
- April 30, 1990 l
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r CESSAR ';!nific m., ' ([3) TABLE OF CONTENTS p ! CHAPTER 6 Section Subiect Pace No. 6.0 ENGINEERED SAFETY FEATURES 6.1-1~ 6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.1-2 6.1.1- METALLIC MATERIALS 6.1-2 1 6.1.1.1- Katerials Selection and 6.1-2 Fabrication 6.1.1.1.1 Specifications for Principal 6.1-2 ESF Pressure Retaining Materials 6.1.1.1.2 Engineered Safety Feature 6.1-2 Construction Materials 6.1.1.1.3 Integrity of ESF Components 6.1-2 During Manufacture and Construction
\' 6.1.1.1.3.1 Control of Sensitized 6.1-2 Stainless Steel 6.1.1.1.3.2 Cleaning and 6.1-3 Contamination Protection Procedures-6.1.1.1.3.3 Cold Worked Stainless 6.1-4 j Steel 1 6.1.1.1.3.4 Non-Metallic' Insulation 6.1 )
1 6.1.1.1.4 Weld Fabrication and Assembly 6.1-4 ' of Stainless Steel ESF , components j 6.1.2 ORGANIC MATERIALS 6.1-5 6.1.2.1 Protective Coatinas 6.1-5 6.1.2.2 Other Materials 6.1-5 l 6.2 CONTAINMENT SYSTEMS 6.2-1 6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2-1 O i l l l v v
'CESSARlin% - ,
O TABLE OF CONTENTS (Cont'd) CHAPTER 6 j Section Subiect Pace No. 6.2.1.1 Containment Structure 6.2-1 6.2.1.1.1 Design Bases 6.2-1 1 6.2.1.1.1 1. Postulated Accident 6.2-1 Conditions 6.2.1.1.1.2 Mass and Energy Release 6.2-2 l 6.2.1.1.1.3 Effects of the Containment- 6.2-2 i Spray System on Energy Removal 6.2.1.1.1.4 Effects of the Containemnt 6.2-2 System on Pressure Reduction 6.2.1.1.1.5 Basis for Containment 6.2-3 Depressurization Rate 6.2.1.1.1.6 Bases for Analysis of 6.2-3 Containment Mininum Pressure . 6.2.1.1.2 Design Features 6.2-3 6.2.1;1.2.1 Protection from the Dynamic 6.2-3 g Effects of Postulated Accidents ! l 6.2.1.1.2.2 Codes and Standards _6.2-4 ' 6.2.1.1.2.3 Protection Against External 6.2-4 I Pressure Loads 6.2.1.1.2.4 Potential Water Traps Inside 6.2-4 Containment 6.2.1.1.2.5 Containment Cooling and 6.2-5 Ventilation Systems 6.2.1.1.3 Design Evaluation 6.2-5 6.2.1.1.3.1 Analysis of Pressure and 6.2-5 Temperature Response of Containment to LOCAs 6.2.1.1.3.1.1 Postulated Breaks in 6.2-5 Primary Coolant System 6.2.1.1.3.1.2 Pressure and Temperature 6.2-5 Response of Containment to Postulated LOCAs Amendment G 11 April 30, 1990
CESSAR Kncuca O TABLE OF CONTENTS (Cont'd) CHAPTER 6 i Section Subiect Pace No. 6.2.1.1.3.2 Analytical Techniques 6.2-6 6.2.1.1.3.2.1 Initial Conditions- 6.2-6 6.2.1.1.3.2.2 Containment Pressure 6.2-6 Analysis
.6.2.1.1.3.3 Failure Mode and Effects 6.2-7 Analysis 6.2.1.1.3.4 Analyses of Pressure and 6.2-8 Temperatue Response of Containment to Secondary System Postulated Pipe Breaks 6.2.1.1.3.5 Containment Passive-Heat 6.2-9 Sinks 6.2.1.1.3.6 Inadvertent Operation of the 6.2-10 Containment Heat Removal
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\- 6.2.1.1.3.7 Systems Sequence of Accident Events 6.2-11 6.2.1.1.3.8 Energy Inventories and 6.2-11 Distribution 6'.2.1.1.3.9 Long-Term' Containment 6.2-11 G Pressure and Temperature 6.2.1.1.3.10- Functional Capability 6.2-12 of Containment Normal Ventilation Systems .
6.2.1.1.3.11 Post-Accident Containment 6.2-13 ! Pressure and Temperature Monitoring 6.2.1.2 Containment Subcompartments 6.2-13 6.2.1.3 Mass cnd Enerav Release Analyses 6.2-13 for Postulated-Loss-of-Coolant L Accidents I 6.2.1.3.1 Mass and Energy Release: Data 6.2-14 6.2.1.3.2 Energy Sources 6.2-15 6.2.1.3.3 Description of Blowdown 6.2-15 Model 6.2.1.3.4 Description of Core Reflood 6.2-17 Model f('
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6.2.1.3.5 Description of Post Reflood 6.2-19 Model I Amendment G L 111 April 30, 1990 L
CESSARan%ma , O TABLE OF CONTENTS (Cont'd) CHAPTER 6 Section Section Page No. 6.2.1.3.6 Description'of Long-Term 6.2-19 Cooling Model. 6.2.1.3.7 Single Active Failure 6.2-21 Analysis 6.2.1.3.8 Metal-Water-Reaction 6.2-21 6.2.1.3.9 Energy Inventories 6.2-21 6.2.1.3.10 Additional Information 6.2-21 6.2.1.4 Mass and Enerav Release Analysis 6.2-22 for Postulated Secondary System Pine Ruotures Inside Containment 6.2.1.4.1 Mass and Energy Release Data 6.2-23 6.2.1.4.2 Single Failure Analysis 6.2-24 6.2.1.4.3 Initial Conditions 6.2-24 6.2.1.4.4 Description of Blowdown Model 6.2-25 6.2.1.4.5 Energy Inventories- 6.2-27
'.. 2.1.4.6 Additional Information 6.2-27 6.2.1.5 Minimum Containment Pressure 6.2-28 Analysis for Performance.
Cacability Studies on Safety Iniection System G 6.2.1.5.1 Introduction and Summary 6.2-28 6.2.1.5.2 Method of Calculation 6.2-28 6.2.1.5.3 Input Parameters 6.2-28 6.2.1.5.3.1 Mass and Energy Release 6.2-28 Data 6.2.1.5.3.2 Initial Containment 6.2-28a Internal Conditions 6.2.1.5.3.3 Containment Volume 6.2-28a 6.2.1.5.3.4 Active Heat Sinks 6.2-28a 6.2.1.5.3.5 Steam Water Mixing 6.2-28a 6.2.1.5.3.6 Passive Heat Sinks 6.2-28b 6.2.1.5.3.7 Heat Transfer to 6.2-28b Passive Heat Sinks 6.2.1.5.3.8 Containment Purge 6.2-28b System 6.2.1.5.4 Results 6.2-28b 6.2.1.6 Testino and Insoection 6.2-28b Amendment G iv April 30, 1990
CESSAR'JnL mm 1 l TABLE OF CONTENTS (Cont'd) CHAPTER 6 Section. Subiect Page No. 6.2.1.7 Instrumentation Acolications 6.2-28c t 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS 6.2-29 6.2.2.1 Desian.Pases 6.2-29 6.2.2.1.1 Summary Description 6.2-29 6.2.2.1.2 Functional Design Basis 6.2-29 6.2.2.2 System Desian 6.2-29 ' 6.2.2.2.1 ~ System Schematic 6.2-29 6.2.2.2.2 Component Dcecription 6.2-29 ! 6.2.2.2.3 Overprescore Protection 6.2-29 3 6.2.2.2.4 Applice.ble Codes and 6.2-29 Classifications 6.2.2.2.5 Systen Reliability-Considerations- 6.2-30
- 6.2.2.2.6 System Operation 6.2-30 6.2.2.3 Dasian Evaluation 6.2-30 6.2.2.4~ Preonerational Testina 6.2-30 6.2.3 SECONDARY CONTAINMENT FUNCTIONAL 6.2-30 E-DESIGN -
6.2.4 CONTAINMENT ISOLATION SYSTEM 6.2-30 6.2.4.1 Desian Bases 6.2-31 6.2.4.1.1 Overall Requirements 6.2-31 6.2.4.1.2 Design Features 6.2-32 6.2.4.2 System Description 6.2-33 -3 6.2.4.3 Safety Evaluation 6.2-36 6.2.4.4 Testino and Inspection 6.2-37 6.2.4.5 Instrumentation Recuirements 6.2-38 6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT 6.2-39 6.2.6 CONTAINMENT LEAKAGE TESTING 6.2-39 Amendment E v December 30, 1988 4
CESSAR fB#, car... - 9 T& BIT OF CONTENTS (Cont'd) CRPTER 6 Section subiect Page No. 6.2.6.1 containment Intearated Leak 6.2-39 Rate Test 6.2.6.2 containment Penetration Leakace 6.2-a0 Rate Test 6.2.6.3 Containment Isolatip M'A3X9, 6.2-41 Leakace Rate Test 4 6.2.6.4 Schedulina and Reoortinc of 6.2-41 Periodic Tests 6.2.6.5 Soecial Testina Reauirements 6.2-4x
- 6. 3 - SAFETY INJECTION SYSTEM 6.3-1 6.3.1 DESIGN BASES 6.3-1 6.3.1.1 Summary Descriotion 6.3-1 6.3.1.2 Criteria 6.3-1 6.3.1.2.1 Functional Design Bases 6.3-1 6.3.1.2.2 Reliability Design Bases 6.3-2 6.3.1.3 Interface Recuirements 6.3-2 6.3.2 SYSTEM DESIGN 6.3-11 6.3.2.1 System Scher.atic 6.3-11 6.3.2.2 Component Descriotion 6.3-12 6.3.2.2.1 Incorkainment Refueling Water 6.3-12 Storage Tank 6.3.2.2.2 Safety Injection Tanks 6.3-12a 6.3.2.2.3 Safety Injection Pumps 6.3-13 C 6.3.2.2.4 Piping 6.3-15 6.3.2.2.5 Valves 6.3-15 6.3.2.3 Aeolicable Codes and 6.3-16 Classifications O
Amendment E vi December 30, 1988
C E S S A R H m ?,c. m O TABLE OF CONTENTS (Coat'd) , CEAPTER 6 section Subiect Pace No 6.3.2.4 Materiale Soecifications and 6.3-16 Comontibility 6.3.2.5 system Reliability 6.3-16 6.3.2.5.1 Safety Injection Tanks 6.3-16 C 6.3.2.5.2 Safety Injection Subsystems 6 . .' - 17 , 6.3.2.5.3 Power Sources 6.3-18 6.3.2.5.4 Capacity to Maintain Cooling 6.3-18 ; Following a Single Failure 6.3.2.6 Protection Provisions 6.3-20 6.3.2.6.1 Capability to Withstand 6.3-20 l Design Bases Environment i 6.3.2.6.2 Missile Protection 6.3-20 6.3.2.6.3 Seismic Design 6.3-21 6.3.2.7 Reauired Manual Actions 6.3-21 i 6.3.3 PERFORMANCE EVALUATION 6.3-23 L 6.3.3.1 Introduction and Summarv 6.3-23 6.3.3.2 Larae Break Analysis 6.3-24 6.3.3.2.1 Mathematical Model 6.3-24 6.3.3.2.2 Safety Injection System 6.3-25 Assumptions 6.3.3.2.3 Core and System Parameters 6.3-25 6.3.3.2.4 Containment Parameters 6.3-26 6.3.3.2.5 Break Spectrum 6.3-26 6.3.3.2.6 Results and Conclusions 6.3-26 G 6.3.3.3 Small Break Analygla 6.3-27 1 6.3.3.3.1 Evaluation Model 6.3-27 6.3.3.3.2 Safety Injection System 6.3-27 Assumptions ' 6.3.3.3.3 Core and System Parameters 6.3-28 6.3.3.3.4 Containment Parameters 6.3-28 6.3.3.3.5 Break Spectrum 6.3-28 6.3.3.3.6 Results 6.3-29 G- 6.3.3.3.7 Instrument Tube Rupture 6.3-30 Amendment G vii April 30, 1990
CESSAR tiMneu.. O TABLE OF CONTENTS (Cont'd) CEhPTER 5 Sectien Subiest Page _Mo .. 6.3.3.4 Post-LOCA Lona-Term Coolina 6.3-33 6.3.3.4.1 General Plan 6.3-33 6.3.3.4.2 Assumptions Used in the 6.3-33a Performance Evaluation of the LTC Plan 6.3.3.4.3 Paramatars Used in the 6.3-34
- Performance Evaluation of the LTC Plan 6.3.3.4.4 Results of the LTC 6.3-35 Performance Evaluation 6.3.3.5 Secuence of Event and Systems 6.3-36 OoeratiQD 6.3.3.6 Radioloaical Consequences 6.3-39 6.3.3.7 Chanter 15 Accident Analysis 6.3-40b 6.3.4 TESTS AND INSPECTIONS 6.3-41 6.3.4.1 SIS Performance Tesig 6.3-41 D 6.3.4.2 Reliability Tests and Inspections 6.3-41 6.3.4.2.1 System Level Testa 6.3-41 6.3.4.2.2 Component Testing 6.3-41 6.3.5 INSTRUMENTATION 6.3-41 6.3.5.1 Desian criteria 6.3-41 6.3.5.2 System Actuation Sianals 6.3-42 6.3.5.2.1 Safety Injection Actuation 6.3-42 Signal (SIAS) 6.3.5.3 Instrumentation Durina coeration 6.3-43 6.3.5.3.1 Temperature 6.3-43 6.3.5.3.2 Pressurc 6.3-43 6.3.5.3.3 Valve Position 6.3-43 6.3.5.3.4 Level 6.3-44 6.3.5.3.5 Flow 6.3-44 Amendment D viii September 30, 1988
CESSAR tRnnem.. O TABLE OF CONTENTS (Cont'd) CEAPTER 6 Section Subient Page No. 6.3.5.4 Post-Accident Instrumentation 6.3-44 6.4 ABITABILITY SYSTEMS D 6.4-1 6.4.1 DESIGN BASES 6.4-1 6.4.2 SYSTEM DESCRIPTION 6.4-1 6.4.2.1 General 6.4-1 6.4.2.2 System Ooeration 6.4-2 6.4.3 SAFETY EVALUATION 6.4-2 6.4.4 INSPECTION AND TESTING REQUIREMENTS 6.4-4 E 6.5 CONTAINMENT SPRAY SYSTEM 6.5-1 6.5.1 DESIGN BASES 6.5-1 6.5.1.1 Summary Descriotion 6.5-1 6.5.1.2 Functional Desian Bases 6.5-1 6.5.1.3 Interface Requirements 6.5-2 6.5.2 SYSTEM DESIGN 6.5-11 6.5.2.1 System Schematic 6.5-11 6.5.2.2 Comoonent Descriotion 6.5-11 0.5.2.2.1 Containment Spray Pumps 6.5-11 6.5.2.2.2 Containment Spray Heat Exchangers 6.5-12 6.5.2.2.3 Valves 6.5-12 6.5.2.2.3.1 Actuator-operated Valves 6.5-12 6.5.2.2.3.2 Manually-Operated Valves 6.5-12 6.5.2.2.3.3 Relief Valves 6.5-13 6.5.2.2.4 Spray Nozzlea 6.5-13 6.5.2.2.5 In-containment Refueling Water 6.5-13 Storage Tank O 6.5.2.2.6 6.5.2.2.7 Piping Instrumentation 6.5-13 6.5-14 Amendment E ix December 30, 1988
y CESSARtR L ; i O TABLE OF COMTENTE (Cont'd) CEAPTER 6 EAf%iRn subiect Page No. ; 6.5.2.3 Overoressure Protection 6.5-14 6.5.2.4 Aeolicable Codes and Classi- 6.5-14 fications System Reliability Considerations E 6.5.2.5 6.5-15 , 6.5.2.6 System Operation 6.5-16 6.5.2.6.1 Normal Operation 6.5-16 6.5.2.6.2 Post-Accident Operation 6.5-16 6.5.2.6.3 Plant Shutdown (Startup) 6.5-16 , 6.5.3 DESIGN EVALUATION 6.5-17 6.5.4 PREOPERATIONAL TESTING 6.5-17 , 6.6 INSERVICE INSPECTION OF CLASS 2 &3 6.6-1 COMPONENTS , 6.6.1 COMPONENTS SUBJECT TO EXAMINATION 6.6-1 6.6.2 ACCESSIBILITY 6.6-1 6.7 SAFETY DEPRESSURIZATION SYSTEM 6.7-1 6.7.1 DESIGN BASIS 6.7-1 6.7.1.1 Summary Descrintion 6.7-1 t 6.7.1.2 Criterie. 6.7-1 0 6.7.1.2.1 Functional Design Basis 6.7-1 6.7.1.2.2 Reliability Design Data 6.7-5 6.7.1.2.3 Interface Requirements 6.7-5 6.7.2 SYSTEM DESIGN 6.7-8 6.7.2.1 System Schematic 6.7-9 6.7.2.1.1 Reactor Coolant Gas Vent Function 6.7-9 6.7.2.1.2 Rapid Depressurization (Bleed) 6.7-10 Function l Amendment E x December 30, 1988
CESSAR m#co.. : TABLE OF CONTENTS (Cont'd) CEAPTER 6 Section subiect Pace No. l 6.7.2.2 Comoonent Descriotion 6.7-11 , 6.7.2.2.1 Reactor Coolant Gas Vent Function 6.7-11 . Valves 6.7.2.2.2 Rapid Depressurization (Bleed) 6.7-11 Valves D 6.7.2.3 Aeolicable Codes and 6.7-12 Classification ' 6.7.2.4 System Reliability 6.7-12 6.7.2.5 Protection Provisions 6.7-12 6.7.2.6 Recuired Manual Actions 6.7-13 O e O , Amendment E xi December 30, 1988
CESSARtiMem.. O LIST OF TABLES CEAPTER 6 i Table subient 6.1-1 Principal ESF Pressure Retaining Materials 6.1-2 Engineered Safety Features Structural Materials That Could Be Exposed To Core Cooling Water or containment spray In The Event of A LOCA 6[1-3 Coating Materials Used In Containment 6.1-4 Other Organic Materials In Containment 6.2.1-1 Spectrum of Postulated Accidents 4 6.2.1-2 Calculated Values for Containment Pressure Parameters 6.2.1-3 Principal Containment D3 sign Parameters 6.2.1-4 Double-Ended Suption Leg Slot Break - Maximum SIS Flow (9.8175 ft Total Area) 6.2.1-5 Double-Ended Sugtion Leg Slot Break - Minimum SIS Flow (9.8175 ft Total Area) 6.2.1-6 Double-Ended Dischafge Leg Slot Break - Maximum i SIS Flow (9.8175 ft Total Area) G 6.2.1-7 Double-Ended Dischafge Leg Slot Break - Minimum i SIS Flow (9.8175 ft Total Area) 6.2.1-8 Double-Ended Hot Leg Slot Break (19.2423 ft , Total Area) 6.2.1-9 Main Steam Line B3eak, 102% Power - Loss of One CSS Train (8.72 ft Total Area) 6.2.1-10 Main Stepm Line Break, 102% Power - MSIV Failure (8.72 ft Total Area) 6.2.1-11 Main Steam Ling Break, 50% Power - Loss of One CSS Train (8.72 ft Total Area) O Amendment G xii April 30, 1990
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@ESSAR fBr?.co . l O l LIST OF TABLES (Cont'd)
CRAPTER 5 Table subiect i i 6.2.1-12 Main Stepm Line Break, 50% Powcr - MSIV Failure l (8.72 ft Total Area) l 6.2.1-13 Main Steam Ling Break, 20% Power - Loss of One CSS i Train (8.72 ft Total Area) l 1 6.2.1-14 Main Stepa Line Break, 20% Power - MSIV Failure I (8.72 ft Total Area) 6.2.1-15 Main Steam Ling Break, 0% Power - Loss of One CSS Train (4.50 ft Total Area) l 6.2.1-16 Main Stepm Line Break, 0% Power - MSIV Failure , (4.50 ft Total Area) G 1 6.2.1-17 Summary Results of Postulated Pipe Rupture l fs Analysis 6.2.1-18 Initial Conditions for Containment Peak Pressure l Analysis l 6.2.1-19 ESF. Systems Parameters for Containment Peak Pressure Analysis ! 6.2.1-20 Containment Spray Pump Activation Characteristics 6.2.1-21 Typical Passive Heat Sink Data ) l 6.2.1-22 Initial Conditions for Containment Minimum l Pressure Analysis 6.2.1-23 Long-Term Mass and Energy Release 6.2.1-24 Energy Balances 6.2.1-25 Primary Side Resistance Factors, FLOOD MOD 2 CODE 6.2.1-26 Blowdown And Reflood Mass And Energy Release for the Minimum Containment Pressure Analysis 1 l 6.2.4-1 Containment Isolation System i 6.3.2-1 Safety Injection System Component Parameters C l Amendment G ' xiii April 30, 1990
C E S S A R ti M ieu . O LIST OF TABLES (Cont'd) CHAPTER 6 Table Subiect 6.3.2-2 Safety Injection System Frilure a Modes and Effects Analysis 6.3.2-3 Safety Injection Pump NFSH Requirements 6.3.2-4 Safety Injection System Head Loss Requirements 6.3.3.2-1 Time Sequence of Important Events for a Spectrum of Large Break LOCAs (Seconds After Break) 6.3.3.2-2 General System Parameters and Initial Conditions, Large Break SIS Performance 6.3.3.2-3 Large Break Spectrum 6.3.3.2-4 Peak Clad Tertperature and Oxidation Percentage for the Large Break Spectrum 6.3.3.2-5 Variables Plotted as a Function of Time for Each Large Break in the Spectrum 6.3.3.2-6 Additional Variables Plotted as a Function of Time for the Limiting Large Break 6.3.3.3-1 Safety Inject 30n Pumps Minimum Delivered Flow to RCS (Assuming One Emergency Generator Failed) 6.3.3.3-2 General System Parameters 6.3.3.3-3 Small Break Spectrum 6.3.3.3-4 Variables Plotted as a Function of Time for Each Large Break in the Spectrum 6.3.3.3-5 Fuel Rod Performance Summary 6.3.3.3-6 Times of Inl.arest for Small Breaks (Seconds) 6.3.3.5-1 Sequence of Events for Representative Large and Small Break LOCAs 6.3.3.5-2 Disposition of Norlaally Operating Systems for Large and Small Break LOCA Analyses xiv
CESSAR inancu. O LIST OF TABLES (Cont'd) CHAPTER 6 Table Subiect 6.3.3.5-3 Utilization of Safety Systems for Representative Small Break (0.02 ft') 6.3.3.5-4 Utilization of Safety Systems for Representative Large Break (0.8 DEG/PD) 6.3.3.6-1 Parameters Used in the Radiological Consequences of a LOCA 6.3.3.7-1 Chapter 15 Limiting Events Which Actuate the Safety Injection System 6.5-1 Containment Spray System Design Parameters , 6.5-2 Containment Spray System Display Instrumentation E 6.5-3 Containme.nt Spray System Failure Modes and Effects Analysis v 6.7-1 Safety Depressurization System - Active Valve List 6.7-2 Safety Depressurization System - Safety Class 6.7-3 Failure Modes and Effects Analysis Safety Depressurization System 0 l-l O Amendment E xv December 30, 1988
L CESSAR !!! Nim.. _ i i O LIST OF FIGURES , CHAPTER 6 Figure Subiect 6.2.1-1 Double-Ended Suction Leg Slot Break - Maximum SIS Flow, C:..t&Inmei.t Pressure and Atmosphere Temperature vs. Time 6.2.1-2 Double-Ended Suction Leg Slot Break - Minimum SIS Flow, Containment Pressure and 7.Luosphere Temperature vs. Time 6.2.1-3 Double-Ended Discharge Leg Slot treak - Maximum SIS Flow, Containment Pressure and Atmosphere Temperature vs. Time 6.2.1-4 Double-Ended Discharge Leg Slot Break - Minimum SIS Flow, Containment Pressure and Atmosphere Temperature vs. Time 6.2.1-5 Double-Ended Hot Leg Slot Break, Containment Pressure and Atmosphere Temperature vs. Time g , 6.2.1-6 Main Steam Line Break - 102% Power, Loss of One CSS Train, Containment Pressure and Atmosphere Temperature vs. Time 6.2.1-7 Main Steam Line Break - 102% Power, MSIV Failure, Containment Pressure and Atmosphere Temperature vs. Time 6.2.1-8 Main Steam Line Break - 50% Power, Loss of One CSS Train, Containment Pressure and Atmosphere Temperature vs. Time 6.2.1-9 Main Steam Line Break - 50% Power, MSIV Failure, containment Pressure and Atmosphere Temperature vs. Time , 6.2.1-10 Main Steam Line Break - 20% Power, Loss of One CSS Train, Containment Pressure and Atmosphere l Temperature vs. Time 6.2.1-11 Main Ceam Line Break - 20% Power, MSIV Failure, Containment Pressure and Atmosphere Temperature l vs. Time Ol Amendment G xvi April 30,_1990
CESSAR E L.. LIST OF FIGURES (Cont'd) CHAPTER 6 ; Eiqure subject i 6.2.1-12 Main Steam Line Break - 0% Power, Loss of One CSS Train, Containment Pressure and Atmosphere , Temperature vs. Time ' 6.2.1-13 Main Steam Line Break - 0% Power, MSIV Failure, ' Containment Pressure and Atmosphere Temperature vs. Time 6.2.1-14 Long-Term Condensing Heat Transfer Coefficient vs. Time 6.2.1-15 Long-Term Containment Pressure Response 6.2.1-16 Long-Term Containment Temperature Response 6.2.1-17 Normalized Decay Heat Curve 6.2.1-18 Double-Ended Suction Leg Slot Break - Maximum SIS Flow, Safety Injection Flow vs. Time G 6.2.1-19 Double-Ended Suction Leg Slot Break - Minimum SIS-Flow, Safety Injection Flow vs. Time 6.2.1-20 Double-Ended Discharge Leg Slot Break - Maximum SIS Flow, Safety Injection Flow vs. Time 6.2.1-21 Double-Ended Discharge Leg Slot Break - Minimum SIS Flow, Safety Injection Flow vs. Time 6.2.1-22 Main Steam Line Break Model 6.2.1-23 Main Steam Line Break - 102% Power, Loss of One JSS Train, Feedwater Addition vs. Time 6.2.1-24 Main Steam Line Break - 102% Power, MSIV Failure, Feedwater Addition vs. Time 6.2.1-25 Main Steam Line Break - 50% Power, Loss of One CSS Train, Feedwater Addition vs. Time 6.2.1-26 Main Steam Line Break - 50% Power, MSIV Failure, Feedwater Addition vs. Time 6.2.1-27 Main Steam Line Break - 20% Power, Loss of One CSS Train, Feedwater Addition vs. Time Amendment G xvii April 30, 1990
CESSAR !!!nnca. i LIST OF FIGURES (Cont'd) 1 CEAPTER 6 Ficrure Bubiect 6.2.1-28 Main Steam Line Break - 20% Power, MSIV Failure, Feedwater Addition vs. Time ; 6.2.1-29 Main Steam Line Break 'J% Power, Loss of One CSS Train, Feedwater Additir,n vs. Time G 6.2.1-30 Main Steam Line Break - 0% Power, MSIV Failure, Feedwater Addition vs. Time 6.2.1-31 Heat Removal Capacity of the Fan Coolers vs. Containment Atmospheric Temperature 6.2.1-32 Combined Spillage and Spray into Containment 6.2.1-33 Condensing Heat Transfer Coefficients for Static Heat Sinks 6.2.1-34 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Minimum Containment Pressure for ECCS Performance Analysis 6.2.1-35 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, containment Atmosphere Temperature 6.2.1-36 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Containment Sump Temperature 6.2.4-1A Containment Isolation Valve Arrangement 6.2.4.1B Containment Isolation Valve Arrangement 6.2.4.1C Containment Isolation Valve Arrangement 6.3.2-1A Safety Injection System Piping and Instrumentation Diagram 6.3.2-1B Safety Injection System Piping and Instrumentation Diagram 6.3.2-1C Safety Injection System Flow Diagram, Short-Term Mode C 6.3.2-1D Safety Injection System Flow Diagram, Short-Term Mode Amendment G xviii April 30, 1990
CESSARUnhi= 0 LIST OF FIGURES (Cont'd) CHAPTER 6 Fiwure Subject 6.3.2-1E Safety Injection System Flow Diagram, Long-Term Cooling Mode 6.3.2-1F Safety Injection System Flow Diagram, Long-Term C , Cooling Mode i 6.3.2-2 Engineered Safeguards System Schematic Diagram 6.3.2-3 Safety Injection Pump Head and NPSH Curves (Typical) 6.3.3.2-1A 1.0 x Double-Ended Slot Break in Pump Discharge Leg, Core Power 1 6.3.3.2-1B 1.0 x Double-Ended Slot Break in Pump Discharge I Leg, Pressure in Center Hot Assembly Hode ! 6.3.'3.2-1C 1.0 x Double-Ended S)ot Break in Pump Discharge 1 Leg, Leak Flow 6.3.3.2-lD.1 1.0 x Double-Ended Slot Break in Pump Discharge j Leg, Flow in Hot Assembly - Path 16, Below Hot Spot 6.3.3.2-lD.2 1.0 x Double-Fnded Slot Break in Pump Discharge Leg, Flow in Hot Assembly - Path 17, Above Hot Spot 6.3c3.2-lE 1.0 x Double-Ended Slot Break in Pump Discharge Le?, Hot Assembly Quality i 6.3.3.2-lF 1.0 x Double-Ended Slot Break in Pump Discharge Leg, Containment Pressure i 6.3.3.2-1G 1.0 x Double-Ended Slot Break in Pump Discharge Leg, Mass Added to Core During Reflood 6.3.3.2-lH 1.0 x Double-Ended Slot Break in Pump Discharge Leg, Peak Clad Temperature 6.3.3.2-2A 0.8 x Double-Ended Slot Break in Pump Discharge Leg, Core Power 2 t Amendment C xix June 30, 1988 m
C E S S A R tl & co.. ; O; LIST OF FIGURES (Cont'd) - CEAPTER 6 Figure subiect 6.3.3.2-2B 0.8 x Double-Ended Slot Break in Pump Discharge Leg, F. essure in Center Hot Assembly Node 6.3.3.2-2C 0.8 x Double-Ended Slot Break in Pump Discharge Leg, Leak Flow 6.3.3.2-2D.1 0.8 x Double-Ended Slot Break in Pump Discharge Leg, Flow in Hot Assembly - Path 16, Below Hot Spot 6.3.3.2-2D.2 0.8 x Double-Ended Slot Break in Pump Discharge Leg, Flow in Hot Assembly - Path 17, Above Hot Spot 6.3.3.2-2E 0.8 x Double-Ended Slot Break in Pump Discharge Leg, Hot Assembly Quality - 6.3.3.2-2F 0.8 x Double-Ended Slot Break in Pump Discharge
- Leg, Containment Pressure l 6.3.3.2-2G 0.8 x Double-Ended Slot Break in Pump Discharge ,
Leg, Mass Added to Core During Reflood 6.3.3.2-2H 0.8 x Double-Ended Slot Break in Pump Discharge Leg, Peak Clad Temperature l 6.3.3.2-3A 0.6 x Double-Ended Slot Break in Pump Discharge Leg, Core Power 6.3.3.2-3B 0.6 x Double-Ended Slot Break in Pump Discharge Leg, Pressure in Center Hot Assembly Node l 6.3.3.2-3C 0.6 x Double-Ended Slot Break in Pump Discharge Leg, Leak Flow 6.3.3.2-3D.1 0.6 x Double-Ended Slot Break in Pump Discharge ,. Leg, Flow in Hot Assembly - Path 16, Below Hot ! Spot 6.3.3.2-3D.2 0.6 x Double-Ended Slot Break in Pump Discharge Leg, Flow in Hot Assembly - Path 17, Above Hot Spot 6.3.3.2-3E 0.6 x Double-Ended Slot Break in Pump Discharge Leg, Hot Assembly Quality xx
CESSAR tit #,co..
~
l. O' LIST OF TIGURES (Cont'd) CEAPTER 6 Figure subiect 6.3.3.2-3F 0.6 x Double-Ended Slot Break in Pump Discharge Leg, Containment Pressure , 6.3.3.2-3G 0.6 x Double-Ended Slot Break in Pump Discharge Leg, Mass Added to Core During Reflood 6.3.3.2-3H 0.6 x Double-Ended Slot break in Pump Discharge Leg, Peak Clad Temperature 6.3.3.2-4A 0.5 Ft 2 Slot Break in Pump Discharge Leg, Core Power 2 6.3.3.2-4B 0.5 Ft Slot Break in Pump Discharge Leg, Pressure in Center Hot Assembly Node
's . 3 . 3 . 2 - 4 C 0.5 Ft Slot Break in Pump Discharge Leg, Leak Flow 2
b.3.3.2-4D.1 0.5 Ft Slot Break in Pump Discharge Leg, Flow in Hot Assembly - Path 16, Below Hot Spot 2 6.3.3.2-4D.2 0.5 Ft Slot Break in Pump Discharge Leg, Flow in Hot Assembly Path 17, Above Hot Spot 6.3.3.2-4E 0.5 Ft Slot Break in Pump Discharge Leg, Hot Assembly Quality 2 6.3.3.2-4F 0.5 Ft Slot Break in Pump Discharge Leg, Containment Pressure 6.3.3.2-4G 0.5 Ft Slot Break in Pump Discharge Leg, Mass Added to Core During Reflood 6.3.3.2-4H 0.5 Ft 2 Slot Break in Pump Discharge Leg, Peak Clad Temperature 6.3.3.2-5A 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Core Power 6.3.3.2-5B 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Pressure in Center Hot Assembly Node 6.3.3.2-SC 1.0 x Double-Ended Guillotine Break in Pump l Discharge Leg, Leak Flow 1 xx1 1
.+s a- A CESSAR WW.co.
III LIST OF FIGURES (Cont'd) CHAPTER 6 Fim1re subiect 6.3.3.2-5D.1 1.0 x Double Ended Guillotine Break in Pump - Discharge Leg, Flow in Hot Assembly-Path 16, Below , Hot Spot ; 6.3.3.2-5D.2 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Flow in Hot Assembly-Path 17, Above , Hot Spot 6.3.3.2-5E 1.0 x Dcuble-Ended Guillotine Break in Pump Discharge Leg, Hot Assembly Quality 6.3.3.2-5F 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Containment Pressure 6.3.3.2-5G 1.0 x Double-Ended Guillotine Break in Pump-Discharge Leg, Mass Added to Core During Reflood 6.3.3.2-5H 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Peak Clad Temperature 6.3.3.2-SI 1.0 x. Double-Ended Guillotine Break in Pump Discharge Leg, Mid Annulus Flow 6.3.3.2-5J 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Qualities Above and Below the Core ! 6.3.3.2-5K 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Core Pressure Drop 6.3.3.2-5L 1.0 x Double-Ended. Guillotine Break in Pump _ Discharge Leg, Safety Injection Flow Into Reactor g Vessel
-6.3.3.2-5M 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Water Level in Downcomer During Reflood 1
x Double-Ended Guillotine Break 6.3.3.2-5N 1.0 in Pump l Discharge Leg, Hot Spot Gap Conductance 6.3.3.2-50 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Local Clad oxidation t i l-l Amendment G l xxii April 30, 1990
@ESSAR tiirvie.v... %.J LIST OF FIGURES (Cont'd)
CHAPTER 6 Eiggga Subioet 6.3.3.2-5P 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg; Clad, Centerline, Average Fuel and ! Coolant Temperature for Hottest Node 6.3.3.2-SQ 1.0 x Double-Ended Guillotine Break in Dump Discharge Leg, Hot Spot Heat Transfer coefficient 6.3.3.2-5R 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Hot Rod Internal. Gas Pressure G 6.3.3.2-SS 1.0 x Double-Ended Guillotine Break in Pump Discharge Leg, Core Bulk Channel Flow Rate 6.3.3.2-6A 0.E x Double-Ended Guillotine Break in Pump Discharge Leg, Core Power
, 6.3.3.2-6B 0.8 x Double-Ended Guillotine Break in Pump (N Discharge Leg, Pressure in Center Hot Assembly Node r
6.3.3.2-6C 0.8 x Double-Ended Guillotine Break in Pump Discharge Leg, Leak Flow 6.3.3.2-6D.1 0.8 x Double-Ended Guillotine Break in Pump Discharge Leg, Flow in Hot Assembly-Path 16, Below Hot Spot 6.3.3.2-6D.2 0.8 x Double-Ended Guillotine Break in Pump Discharge Leg, Flow in Hot Assembly-Path 17, Above Hot Spot 6.3.3.2-6E 0.8 x Double-Ended Guillotine Break in Pump Discharge Leg, Hot Assembly Quality 6.3.3.2-6F 0.8 x Double-Ended Guillotine Break in Pump Discharge Leg, Containment Pressure 6.3.3.2-6G 0.9 x Double-Ended Guillotine Break in Pump Discharge Leg, Mass Added to Core During Reflood 6.3.3.2-6H 0.8 x Double-Ended Guillotine Break in Pump Discharge Leg, Peak Clad Temperature 6.3.3.2-7A 0.6 x Double-Ended Guillotine Break in Pump Discharge Leg, Core Power Amendment G xxiii April 30, 1990
1 l)lb!hhh k!I SkN$IICATION c I O LIST OF FIGURES (Cont'd) CEAPTER 4 Figure Jubiect i i 6.3.3.2-7B 0.6 x Double-Ended Guillotine Break in Pump Discharge Leg, Pressure in Center- Hot Assembly i Node 1 6.3.3.2-7C 0.6 x Double-Ended Guillotine Break in Pump j Discharge Leg, Leak Flow i 6.3.3.2-7D.1 0.6 x Double-Ended Guillotine Break in Pump I Discharge Leg, Flow in Hot Assembly-Path 16, Below Hot Spot ! 6.3.3.2-7D.2 0.6 x Double-Ended Guillotine Break in Pump Discharge Leg, Flow in Hot Assembly-Path 17, Above Hot Spot i 6.3.3.2-7E 0.6 x Double-Ended Guillotine Break in Pump Discharge Leg, Hot Assembly Quality . 6.3.3.2-7F 0.6 x Double-Ended Guillotine Break in Pump Discharge Leg, Containment Pressure i 6.3.3.2-7G 0.6 x Double-Ended Guillotine Break in Pump Discharge Leg, Mass Added to Core During Reflood 6.3.3.2-7H 0.6 x Double-Ended Guillotine Break in Pump Discharge Leg, Peak Clad Temperature 6.3.3.2-8A 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Core Power 6.3.3.2-8B 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Pressure in Center Hot Assembly Node 6.3.3.2-8C 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Leak Flow 6.3.3.2-8D.1 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Flow in Hot Assembly-Path 16, Below Hot Spot 6.3.3.2-8D.2 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Flow in Hot Assembly-Path 17, Above Hot Spot xxiv
CESSAR !!!W.co. n U , LIST OF FIGURES (Cont'd) CHAPTER 6 Fleure subiect 6.3.3.2-8E 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Hot Assembly Quality 2 6.3.3.2-8F 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Containment Pressure > 6.3.3.2-8G 1.0 x Double-Ended Guillotine Break in Pump Suction Leg, Mass Added to Core-During Reflood 6.3.3.2-8H 1.0 x Double-Ended Guillotine Break in Pump Ouction Leg, Peak Clad Temperature 6.3.3.2-9A 1.0 x Double-Ended Guillotine Break in Hot Leg, Core Power 6.3.3.2-9B 1.0 x Double-Ended Guillotine Break in Hot Leg, Pressure in Center Hot Assembly Node L 6.3.3.2-9C 1.0 x Double-Ended Guillotine Break in Hot Leg, Leak Flow 6.3.3.2-9D.1 1.0 x Double-Ended Guillotine Dreak in Hot Leg, Flow in Hot Assembly - Path - 16, Below Hot Spot 6.3.3.2-9D.2 1.0 x Double-Ended Guillotine Break in Hot Leg, Flow in Hot Assembly-Path 17, Above Hot Spot 6.3.3.2-9E 1.0 x Double-Ended Guillotine Break in Hot Leg, Hot Assembly Quality 6.3,3.2-9F 1.0 x Double-Ended Guillotine Break in Hot Leg, Containment Pressure 6.3.3.2-9G 1.0 x Double-Ended Guillotine Break in Hot Leg, Mass Added to Core During Reflood 6.3.3.2-9H 1.0 x Double-Ended Guillotine Break in Hot Leg, Peak Clad Temperature 6.3.3.2-10 Peak Clad Temperature vs Break Area 6.3.3.2-11 1.0 x Double-Ended Guillotine Break in Pump O\ Discharge Leg, Peak Clad Temperature and Local Oxidation vs Rod Average Burnup Peak xxV
l CESSAR !!inne n.. O LIST OF FIGURES (Cont'd) CHAPTER 5 rigura Subiasi 6.3.3.3-1A 0.5. Ft.2 Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-1B 0.5 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-lc 0.5 Ft.2 Cold Leg Break at Pump Discharge-Break Flow Rate 6.3.3.3-lD 0.5 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Inlet Flow Rate 6.3.3.3-lE 0.5 Ft. Cold Leg Break at Pump Discharge-Inner Vessel Two-Phase Mixture Volume 6.3.3.3-lF 0.5 Ft.2 Cold Leg Break, at Pump Discharge-Heat Transfer coefficient at Hot Spot 6.3.3.3-lG 0.5 Ft.2 Cold Leg Break at Pump Discharge-Coolant Temperature at Hot Spot 6.3.3.3-1H 0.5 Ft.2 Cold Leg Break at Pump Discharge-Hot Spot Clad Surface Temperature 6.3.3.3-2A 0.35 Ft.2 Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-2B 0.35 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-2C 0.35 Ft.2 Cold Leg Break at Pump _ Discharge-Break Flow Rate 6.3.3.3-2D 0.35 Ft.2 Cold Log Break at Pump Discharge-Inner Vessel Inlet Flow Rate 6.3.3.3-2E 0.35 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Two-Phase Mixture Volume 6.3.3.3-2F 0.35 Ft.2 Cold Leg Break at Pump Discharge-Heat Transfer Coefficient at Hot Spot 6.3.3.3-2G 0.35 Ft. Cold Leg Break at Pump Discharge-Coolant Temperature at Hot Spot xxvi
CESSAR Mne.non ; LIST OF FIGURES (Cont'd) CHAPTER 6 , Fior. . 4..t ) 6.3.3.3-2H 0.35 Ft.2 Cold Leg Break at Pump Discharge-Hot Spot Clad Surface Temperature l 1 6.3.3.3-3A 0.2 Ft.2 Cold Leg Break at Pump Discharge-Normalized Total Core Power 1 6.3.3.3-3B 0.2 Ft.2 Cold Leg Break at Pump Discharge-Inner vessel Pressure i I 6.3.3.3-3C 0.2 Ft. Cold Leg Break at Pump Discharge-Break Flow Rate ] 6.3.3.3-3D 0.2 Ft.2 Cold Leg Break at Pump Discharge-Inner i Vessel Inlet Flow Rate 6.3.3.3-3E 0.2 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Two-Phase Mixture Volume Ft.2 Cold Leg Break at Pump Discharge-Heat 6.3.3.3-3F 0.2 Transfer Coefficient at Hot Spot 6.3.3.3-3G 0. 2 Ft.2 Cold Leg Break at Pump Discharge-Coolant i Temperature at Hot Spot 6.3.3.3-3H 0.2 Ft.2 Cold Leg Break at Pump Discharge-Hot Spot Clad Surface Temperature 6.3.3.3-4A 0.05 Ft.2 Cold Leg Break at Pump Discharge-Normalized Total Core Power 1 6.3.3.3-4B 0.05 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-4C 0.05 Ft.2 Cold Leg Break at Pump Discharge-Break Flow Rate 6.3.3.3-4D 0.05 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Inlet Flow Rate 6.3.3.3-4E 0.05 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel TwoPhase Mixture Volume 6.3.3.3-4F 0.05 Ft.2 Cold Leg Break at Pump Discharge-Heat C Transfer Coefficient at Hot Spot xxvii
7 f C E S S A R !R Wicuion O LIST OF FIGURES (Cont'd) CHAPTER 6 Figure Subject 6.3.3'.3-4G 0.05 Ft.2 Cold Leg Break at Pump Discharge-Coolant Temperature at Hot Spot 6.3.3.3-4H 0.05 Ft.2 Cold Leg Break at Pump Discharge-Hot Spot-Clad Surface Temperature 6.3.'3.3-5A 0.02 Ft.2 Cold Leg Break at Pump Discharge-Normalized Total Core Power 6.3.3.3-5B 0.02 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Pressure 6.3.3.3-5C 0.02 Ft.2 Cold Leg Break at Pump Discharge-Break Flow Rate 6.3.3.3-5D 0.02 Ft.2 Cold Leg Break at Pump Discharge-Inner Vessel Inlet Flow Rate 1 6.3.3.3-5E 0.02 Ft.2 Cold Leg Break at Pump Discharge-Inner - Vessel Two-Phase Mixture Volume 6.3.3.3-5F 0.02 Ft.2 Cold Leg Break at Pump Discharge-Heat I Transfer Coefficient at Hot Spot 6.3.3.3-5G 0.02 Ft.2 Cold Leg Break at Pump Discharge-Coolant Temperature at Hot Spot l 6.3.3.3-5H 0.02 Ft. Cold Leg Break at Pump Discharge-Hot Spot Clad Surface Temperature 6.3.3.3-6A 0.03 Ft.2 Break at Top of Pressurizer-Normalized Total Core Power 6.3.3.3-6B 0.03 Ft.2 Break at Top of Pressurizer-Inner Vessel Pressure 6.3.3.3-6C 0.03 Ft.2 Break at Top of Pressurizer-heak Flow Rate 6.3.3.3-6D 0.03 Ft.2 Break at Top of Pressurizer-Inner Vessel Inlet Flow Rate 6.3.3.3-6E 0.03 Ft.2 Break at Top of Pressurizer-Inner Vessel TwoPhase Mixture Volume xxviii
CESSAR 'enWenio L
.%J LIST OF FIGURES (Cont'd)
CHAPTER 6 !~ Ficture Subiect 6.3.3.3-6F 0.03 Ft.2 Break at Top of Pressurizer-Heat i Transfer Coefficient at Hot Spot I 6.3.3.3-6G 0.03 Ft.2 Break at Top of Pressurizer-Coolant Temperature at Hot Spot 6.3.3.3-6H 0.03 Ft.2 Break at Top of Pressurizer-Hot Spot Clad Surface Temperature 6.3.3.3-7 Maximum Hot Spot Clad Temperature vs Break Size 6.3.3.4-1 Long Term Cooling Plan , 6.3.3.4-2 Core Flush by Hot Side Injection For 9.8 Ft Cold Leg Break 6.3.3.4-3 Inner Vessel Boric Acid Concentration vs Time 6.3.3.4-4 RCS Refill Time Versus Break Area
'6.3.3.4-5 Overlap of Acceptable LTC Modes In Terms of Cold i Leg Break Size 6.3.3.4-6 RCS Pressure After Refill vs Break Area 6.3.3.5-1A Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-1B Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-1C Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-lD Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-lE Sequence of Events Diagram for Large and Small Break LOCAs 6.3.3.5-lF Sequence of Events Diagram for Large and Small Break LOCAs 6.7-1 Safety Depressorization System Flow Diagram D Amendment D xxix September 30, 1988
4_ 4 p_mM- - = * - . . u.+- ...a,.~+s + -aa.,-- am.- . a..=,.. -e--m.ao. - . - - -.- - -- -----.-- - --- -,- i
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E CESSAR Ennen.. n 6.2 CONTAINMENT SYSTEMS The containment systems include the containment building, the Containment Spray System (CSS), the containment air purification and cleanup systems, the containment isolation system, and the g containment combustible gas control system. i The containmant systems design basis accident (DBA) is the most
- severe of a spectrum of hypothetical loss-of-coolant accidents l (LOCA) and secondary system breaks.
This section provides the design limits and evaluations necessary to demonstrate that the systems listed above will function throughout the station operating lifetime. 6.2.1 CONTAINMENT FUNCTIONAL DESIGN , This section pertains to those aspects of containment building design, testing, and evaluation that relate to the accident mitigation - function. A physical description of the containment and the design criteria relating to construction techniques, static loads, and seismic loads are provided or referenced in Section 3.8. V 6.2.1.1 Containment Structure 6.2.1.1.1 Design Basis ; The safety design basis for the containment is the requirement that the release of radioactive materials subsequent to an accident does not result in doses in excess of the values ; specified in 10 CFR 100. The containment must withstand the l pressure and temperatures of the DBA without exceeding the design l leakage rate of 0.50% volume for the first 24 hours and the l volume thereafter is based on a leak rate associated with half of I the peak pressure assuming 0.50% volume lesk rate at peak pressure. Containment leakage testing is desccibed in Section 6.2.6. The radiological consequence of the DBA are presented in Chapter 15. l 6.2.1.1.1.1 Postulated Accident Conditions l The spectrum of postulated accidents considered in determining
-the design containment peak pressure and temperature and the I containment minimum pressure are summarized in Table 6.2.1-1.
Additional information on the effects of pipe breaks is provided i in Section 3.6. I Containment pressure and temperature analyses assumed that: l Amendment G 6.2-1 April 30, 1990
l l CESSARHuhm. ! i l O A. Each accident is concurrent with a loss of offsite power and the most limiting single active failure of engineered safety features (ESFs) but in cases of secondary system breaks, offsite power is available, and B. No two accidents occur simultaneously or consecutively. The DBAs fer containment peak pressure, subcompartment pressure, and containment minimum pressure are the most severe accidents in the spectrum listed in Table 6.2.1-1. The DBA and calculated values for containment pressure parameters are given in Table 6.2.1-2. Containment design parameters are summarized in Table 6.2.1-3. 6.2.1.1.1.2 Mass and Energy Release The sources and amount of mass and energy release for the j accidents listed in Table 6.2.1-1 are given in Tables 6.2.1-4 ! through 6.2.1-16. The computer codes and assumptions used in l deriving each of the mass and energy release tables are discussed in Sections 6.2.1.3 and 6.2.1.4.
- 6. 2.1. L 1. 3 Effects of the Containment Spray System on Energy Removal
, Energy released to the containment during a postulated accident G l is removed from the containment atmosphere and the In-containment ! Refueling Water Storage Tank (IRWST) by the Containment Spray System (CSS).. The CSS also transfers heat from the containment atmosphere to the IRWST. There are two complete CSS trains, each consisting of one spray train. Each CSS train is cooled by the Component Cooling Water System through the Containment Spray Heat Exchanger (CSHX). This j energy is then rejected to the ultimate heat sink (Section 9.2.5) via the Station Service Water System, 6.2.1.1.1.4 Effects of the Containment Spray System on Pressure Reduction For the purpose of containment peak pressure analysis, the CSS is assumed to be afi'ected by the most restrictive single active failure resulting in the minimum heat removal capability. Assuming this failure, the CSS is still capable of reducing the , containment peak pressure below the containment design pressure and reducing the post-accident pressures to less than 50% of the containment peak . pressure within 24 hours following the postulated accidents. Amendment G 6.2-2 April 30, 1990 l
CESSAR n!#,cua ! i r% ! V) ( l Further descriptions about the single failure application tu ESF systems are presented in Sections 6.2.1.1.3.2, 6.2.1.1.3.3, 6.2.1.3.7 and 6.2.1.4.2. g 6.2.1.1.1.5 Hasis for Containment Depressurization Rate l l The containment depressurization rate following the DBA LOCA I satisfies the requirements of General Design Criterion 50 for offsite doses less than the guideline values of 10 CFR 100 at the containment design leakage rate. The containment design leakage was established as the minimum practicable rate based on . consideration of reactor power level, site characteristics, type of containment, iodine removal capability, constructability, and testability. The validity of the established design leakage rate is verified by analysis of the offsite radiological consequences of the design-basis LOCA as discussed in Section 15.6. 6.2.1.1.1.6 Bases for Analysis of Containment Minimum Pressure - s The analysis of containment minimum pressure following a LOCA is ( based on confirming Safety Injection System (SIS) core reflood
\ capability under the conservative assumptions that maximized the heat removal effectiveness of ESF systems, structural heat sinks, and _ other potential heat removal processes. These assumptions are discussed in Section 6.2.1.5.
6.2.1.1.2 Design Features The containment consists of an essentially leaktight spherical steel containment pressure vessel housed in a reinforced concrete shield building. Section 3.8 presents a detailed description of the containment pressure boundary and shield structure. The principal containment design parameters are shown in Table 6.2.1-3. 6.2.1.1.2.1 Protection from the Dynamic Effects of Postulated Accidents The containment structure, subcompartments, and ESF are protected from loss of safety function due to the dynamic effects- of postulated accidents. The containment design has provided separation and inclusion of barriers and restraints when required to protect essential structures, systems, and components from accident generated missile, pipe whip, and jet impingement l' forcec. The detailed criteria, locations, and descriptions of l systems used for protection are given in Sections 3.5 and 3.6. Amendment G 6.2-3 April 30, 1990 1 -- __ _
CESSARMnAm. . l O, 6.2.1.1.2.2 Codes and Standards ; Codes and standards applied to the design, fabrication, and lt> erection of containment and internal. structures are given in
-Table 3.8.3-1 and Section 3.8. In each case, the _ . codes and standards used are consistent with the equipment safety function. l 1
6.2.1.1.2.3 Protection Against External Pressure Loads l No special-provisions for protection'against. loss of containment !
! integrity under external loading conditions are required.
Consideration given to inadvertent operation of containment heat removal systems and;other possible modes of plant operation that could potentially result in significant external structural loading has- resulted in pressures lower than the design ' containment external pressure. Details of this evaluation are
;provided in Section 6.2.1.1.3,6. -
6.2.1.1.2.4 ' Potential Water Traps Inside Containment The design of the containment minimizes potential trapping of safety injection and' containment spray water which would prevent the. return of the water to the IRWST. The IRWST design is discussed in-Section 6.3.2.2. i The SIS is. designed so that adequate net positive suction head g (NPSH) is provided to system pumps. In addition to considering the static head and suction line pressure drop, the calculation of_available NPSH assumes that the vapor pressure of _ the liquid in the IRWST.is equal to the containment ambient pressure. This essures that the actual availat:s NPSH is always greater than the calculated NPSH. The calculation of available NPSH for the safety injection and containment spray pumps is as follows: (NPSH) actual = (h) ambient pressure
- (h) vapor pressure + (h) static head - (h) loss (NPSH) calculated = (h) static head (h) loss ad(quate NPSH is shown to be available- for the safety injection and containment spray pumps as discussed in Sections 6.3 and 6 .~2 . 2 , respectively.
O Amendment G 6.2-4 April 30, 1990
I CESSARinancam I 6.2.1.1.2.~5 ' Containment Cooling and Ventilation Systems During normal operation, the containment atmosphere is maintained below 110'F by operation of the Containment Cooling and ; Ventilation System which is described in Section 9.4.6. During inspection, testing, maintenance and refueling operations, the ! Containment Purge Ventilation System, described in Section 9.4'.5 is also used. 6.2.1.1.3 Design-Evaluation 6.2.1.1.3.1 Analysis of Pressure and Temperature Response of Containment to IDCAs 6.2.1.1.3.1.1 Postulated Breaks-in Primary Coolant System 1 The spectrum, size, and location of the postulated breaks in the primary coolant system which were analyzed to determine the l containment internal pressure design basis accident are tabulated ; in Table 6.2.1-1. The blowdown data for these breaks are given > in Tables 6.2.1-4 through 6.2.1-8. The calculation of these data is described in Section 6.2.1.3. 6.2.1.1.3.1.2 Pressure and Temperature Response of Containment to Postulated LOCAs In the. event of LOCA in the containment, much of the released reactor coolant will flash to steam. This release of mass and energy raises the pressure and temperature of the atmosphere a within the containment. The severity of the pressure and temperature peaks depends-upon the nature,. size, and location of' the postulated rupture within the RCS. The postulated LOCAs in the reactor coolant system,. described in Table 6. 2.1-1, were analyzed to' determine the most significant LOCA DBA in containment design. Mass'and energy release data for the hot leg break were not generated after the end of the blowdown phase. For the hot leg case, most of the reflood fluid does not pass through a steam generator prior to release tc the containment. .Hence, in contrast to a cold leg break, there is no physical' mechanism to rapidly remove the residual secondary system energy in the steam generators during or after reflood phase, and the mass and energy release rate from the break will be less than for a cold leg break. O Amendment G I 6.2-5 April 30, 1990
.CESSARlna mn
'In order to identify-the worst case, the nypothetical accidents 9:
listed in Table 6.2.1-1, with the-initial conditions specified in Table 6.2.1-18, have been analyzed by the computer Code CONTRANS (Reference 1) to predict the pressure and temperature transients in the. containment following the rupture. The analytical model is described in Reference 1.- The containment pressures and temperatures as a function of time
'for the spectrum of breaks are illustrated in Figures 6.2.1-1 through 6.2.1-5 and the results are summarized in Table 6.2.1-17..
The DBA LOCA is a double-ended hot leg slot break with loss of 1 one containment spray train. The peak calculated pressure is 45.60 psig and the peak containment atmosphere temperature is j 269'F. l 6.2.1.1.3.2 Analytical Techniques 6.2.1.1.3.2.1 Initial Conditions ~ l A conservative prediction of cortainment response for the , spectrum of accidents was assured Ly considering the upper or lower bounding values of containment initial conditions, geoutric parameters, and thermodynamic properties to produce the I maximum pressure results, p The initial conditions within the containment system and the reactor. coolant system prior to accidental initiation are given in Table 6.2.1-18. The containment is assumed to - be at the naximum pressure, maximum inside and outside temperature and minimum design humidity at normal operating condition, to g minimize the calculated heat transfer and maximize the-calculated peak pressure during the postulated accidents. Accident , chronologies for LOCA and MSLB are presented in Part C of Tables 6.2.1-4' through 6.2.1-16 and operating assumptions for the engineered safety feature systems are provided in Table 6.2.1-19 and 6.2.1-20. 1 j 6.2.1.1.3.2.2 Containment Pressure Ar.alysis 1 p The containment pressure analyses are performed using the CONTRANS computer program (Reference 1). The CONTRANS model l predicts the pressure and temperahlrs withir. the containment l' regions and the temperatures in the containment structures. It is assumed that separate blowdown and core thermcl beha viar studies have been made to determine mass and/or energy inp at rates from sources such as the release of reactor coolart, chemical reactions, decay energy, and sensible heat release t'4at O Amendment G 6.2-6 April 30, 1990
CESSAR1!nL m. O may cause heating or boil-off of residual water in the reactor { vessel-or superheating of steam as it passes through the reactor ! system and enters the containment through the postulated point of an RCS rupture. The CONTRANS model analyzes both active and passive heat transfer ! within the containment following a LOCA The CSS and the CSHX models enable an accurate calculation of sheir effect in reducing i containment pressure and temperature. ; CONTRANS calculates a pressure-time transient with stepwise integrations between the thermodynamic state points. The 4 integrations are based on the laws-of conservation of mass and mmy together with their thermodynamic relationships. 1 m wtoosition of heat input functions is assumed so that any i conW, nation of coolant release, metal-water reaction, decay heat , generation, and sensible heat release can be used with ! appropriate ESF features to determine the containment pressure time history associated with a LOCA. The program uses a two region containment model consisting of the containment atmospheres (vapor region), the sump (liquid region), and a primary system model (reactor vessel) which _ is used to calculate mass and energy release data after the LOCA post reflood- period (see Section 6.2.1.3). Mass and energy is transferred between the liquid and vapor regions by boiling, condensation, and evaporation. The heat and mass transfer < coefficients between these two regions are calculated in CONTRANS 1 (Reference 1) . Each region is assumed to be homogeneous, but a G temperature difference can exist between regions. Any moisture condensed in the vapor region during a time increment is assumed to' fall immediately into the liquid region. Noncondensible gases are included in-the vapor region. The thermodynamic assumptions used in CONTRANS and the modeling of the atmosphere and sump regions, reactor vessel region (long term release), safety injection system, containment spray system and containment spray heat exchanger are discussed in Reference 1. The modeling of heat transfer surfaces and coefficients are also discussed in Reference 1. 6.2.1.1.3.3 Failure Mode and Effects Analysis The failure mode and effects analyses (FMEA) of the SIS and the CSS demonstrate that the redundancy of equipment permits at least one train of the systems to function after experiencing either a single active or passive failure. O Amendment G 6.2-7 April 30, 1990
- CESSART!ni?,ce..
I O; The most restrictive failure'in the CSS has been determined to be ' the loss of one ' diesel or one containment spray pump. FMEAs'of G the~ CSS and. SIS- are -discussed in Sections 6.2.2 and 6 '. 3 , respectively. 6.2.1.1.3.4 Analyses of Pressure and Temperature Response of Containment to Secondary System Postulated l Pipe Breaks In addition to the reactor coolant pipe ruptures, an analysis of . secondary system pipe rupture has been performed. For a' main ! ' steam line break (MSLB), following isolation and blowdown of the ruptured. steam generator unit, the decay heat is transferred to the intact unit which will vent to the atmosphere when its safety valves open. Boiloff of emergency feedwr.ter after ruptured steam generator dryout was included. ; For the main feedwater line breaks (MFLB), the're .are no-significant effects on containment pressurizatior. .since the effective break creas are limited by the. steam generator- , internals design and the fluid enthalpy of MFLB is less than the ! enthalpy of the fluid 'in the MSLB. Consequently, breaks in the main feedwater piping would result in blowdown that is less limiting than the MSLB. Therefore MFLBs were not analyzed. A detailed mass and energy release analysis is presented in Section 6.2.1.4 and the blowdown data used to evaluate the t containment response to the main steam line breaks is given. in Tables 6.2.1-9 through 6.2.1-16. The methods and assumptions. used to analy.e the pressure and temperature of the containment from the postulated MSLB are l' described in Section 6.2.1.1.3.2. The most limiting single ( active failure considered is a loss of ane CSS train or an MSIV l- failure. l The most restrictive secondary system pipe break with respect to pressure is a double-ended rupture of main steam line at 0% power concurrent with a loss of one CSS train. The peak pressure of this break-is 48.34 psig. The maximum peak atmospheric temp which occurs for the , case of MSLB at 102% power with ,erature a single failure of MSIV is ! 405.71'F. l> 0 Amendment G
- 6. 2 -8 April 30, 1990
CESSAR %"camn 1 l l The:results of MSLBs are summarized in Table 6.2.1-17 and Figures i 6.2.1-6 through -6.2.1 show the containment pressure and temperature response to these accidents. 6.2.1.1.3.5 Containment Passive Heat Sinks Containment heat sink data, provided in Table 6.2.1-21, consists of the component heat sink data, simplified heat sink modeling data and thermophysical property data of passive heat sink materials used in pressure and temperature analysis. A simplified model is used for the CONTRANS computer code input. The node spacing.used for heat sinks is fine enough to ensure an -
-accurate representation of the thermal gradient in each slab.
The nodal spacing selection is consistent with Reference 1. All containment floors or other larger horizontal surfaces inside the containment above the centainment floor . are designed to be - self-draining and allow all water released into the building to accumulate in the hold up volume of the IRWST, and then return to the IRWST. The spherical steel contair. ment vessel is free-standing above
- O V
elevation 91.75 feet. Below this elevation, the steel J containment is sandwiched between reinforced concrete with a 3 foot thick wall both inside and outside containment. The 3 foot-thick concrete wall inside containment serves as the outer wall of the In-containment Refueling Water Storage. Tank (IRWST). .The concrete wall outside of ' containment forms the boundary of the G subsphere region of the Reactor Building Complex. Heat transfer , to the subsphere region is conservatively assumed to.be zero. Above elevation 91.75 feet, the containment vessel is shielded from the outside environment by a 3 foot thick reinforced l concrete wall and dome. There is an annulus with a minimum of 5 feet clearance between the containment and the inside of- the concrete dome or wall. Heat transfer from the containment to the annulus region is conservatively assumed to be zero. The initial temperature distribution in these heat sinks is l establishgd by assuming a heat-transfer coefficient of 0.0 Btu /hr-ft *F exists at outside surfaces. Following the LOCA or main steamline break. (MSLB), Tagami or Uchida condensing ' heat-transfer coefficients are applied to the inside surfaces ! while a value of 0.0 Btu /hr-ft 'F is used on the outside j surfaces. l Amendment G 6.2-9 April 30, 1990 l I
CESSARnnamn All other heat sinks are entirely within the containment. The inillai containment temperature is the design maximum for reactor power operation (110*F). The initial- relative humidity is conservatively assumed as 10%. The.-thermophysical data used in the analysis are given-in Table 6.2.1-21. The. thermal properties of metal, concrete, and protective coating materials are typical values for temperature observed. Figure 6.2.1-14 shows the condensing heat transfer coefficient as a function of time for the-double-ended pump discharge leg slot break with maximum SIS. During the blowdown phase, the - Tagami heat transfer coefficient was used and thereafter Uchida heat transfer coefficient was used to minimize heat removal to containment heat sink. Detailed description of the heat transfer correlations are discussed and justified in Reference 1. 6.2.1.1.3.6 Inadvertent Operation of the Containment Heat Removal-Systems The containment systems which have the capability of. developing a
. negative pressure within the containment .. are the spray system, the purgo system, and the fan cooler system. The DBA for the minimum containment' pressure design has been. determined to be the inadvertent actuation of the CSS. Consideration was also given
'to misoperation of the containment normal purging system, i.e.,
- operation of the exhaust train with the supply train isolated, but the maximum feasible internal vacuum for this case is limited to a' few inches- of water (gauge) based on the exhaust fan-
. operating curve.
An inadvertent actuation of the CSS can result in a reduction in the containment internal pressure. During normal operation, or during shutdown, the containment is purged - using either the low-volume containment purge or the- high-volume containment purge. An inadvertent actuation of the spray system with the containment purge valves open will result in a negligible decrease in- the containment internal pressure. When- the containment happens to be sealed and'all purge valves are closed, at ' the time ' of the spray actuation, a significant pressure reduction can occur. The main parameters affecting the resultant negative pressure are the pressure, initial containment temperature, relative humidity, and the spray water temperature. O Amendment G 6.2-10 April 30, 1990
CESSARs!nsem x
.i t The analysis of minimum containment pressurer as a result of -f inadvertent spray. actuation, is based upgn a conservative l calculation that assumes no heat t::ansfer from the containment structure, no containment volume reditction due to the addition of ;
spray water and disregard _ of the source of heat within the containment. Initial containment parameters are specified in Table 6.2.1-22 and final containment temperature is taken as the spray water temperature. Dalton's law is applied to determine ' the final containment pressure. In the analysis'of this incident, the containment is assumed to be initially _ at the maximum normal design _ conditions as 110'F. The temperature of the IRWST, the source of the spray water, was .
-conservatively assumed to be equal to 80'F, the lowest allowable i temperature for a containment atmosphere temperature- of 110'F.
The CSS flow rate has a small impact on the minimum pressure. The minimum calculated pressure 'is -1.83 psig. A nominal pressure of'-2.0 psig has been used for the containment design. j 6.2.1.1.3.7 Sequence of Accident Events ' The . accident chronologies of the postulated LOCA and MSLB are
-tabulated in Part C of Table 6.2.1-4 through Table 6.2.1-16.
'6.2.1.1.3.8 Energy Inventories and. Distribution The energy inventories and distribution in the containment for the most severe LOCA and secondary system pipe ruptures are a tabulated in Table 6.2.1-24.
6.2.1.1.3.9 Long-Term Containment Pressure and Temperature The- principal mechanisms that provide ' reduction in the r post-accident pressure' are (1) the heat absorbed by . heat i containment sinks inside the containment (2) the cooling provided
'by the containment sprays, and (3) the effectiveness of the containment spray heat exchangers (CSHX). The analytical .
modeling of these containment heat-absorbing systems is described ! in Reference 1. The parameters describing the heat sinks credited with heat absorption in the containment analysis are shown in Table
=6.2.1-21. The heat sinks act as a temporary repository for a part of the accident energy release, absorbing energy so long as the containment atmosphere is at a temperature greater than that of the exposed heat sink surfaces. Once other heat removal systems, i.e., CSSs and CSHXs lower the containment atmosphere temperature below that of the heat sinks, energy stored in the C
( Amendment G 6.2-11 April 30, 1990
CESSAR !!!nnema i i heat sinks flows back into the containment atmosphere and IRWST 9; ! regions where it is rejected to the outside environment through ' the action of the active heat removal systems mentioned above. G , The CSS is described in Section 6.2.2. The spray system transfers energy- from the containment atmosphere to the IRWST (liquid water region) through heat transfer to finely divided water droplets originating at spray nozzles inside and at the top of the containment building. The IRWST serves as a source of spray water. The CSHXs are described in Section 6.2.2. These heat exchangers serve to transfer energy from the IRWST to the outside environment. The major fraction of the accident energy release is ultimately rejected through CSHXs. k The capability of systems to reduce the containment pressure i after an accident has been conservatively estimated. In the analysis, it is assumed that only one of the two spray systems is operational. It is also assumed that the IRWST water is'at a-J temperature of 110*F. The conservative condensing heat-transfer i coefficients of Tagami and Uchida are used in accordance with the USNRC-acceplea methodology. Heat transfer between the ' containment atmosphere and IRWST water is not considered. Only one train of the CSHX is assumed in operation. The - long-term - mass and energy release data used for the DBA LOCA are listed in Tab 1c 6.2.1-23. The containment pressure response for the the double-ended. pump discharge leg slot LOCA break with maximum SIS flow is shown for a time period of one million seconds in Figure 6.2.1-15. As can be seen from this figure, the containment pressure has bcen reduced to below 22 psig within 24 hours (86,400 seconds) following the accident. The containment atmosphere temperature response of the break, as a function of time, appears in Figure 6.2.1-16. 6.2.1.1.3.10 Functional Capability of Containment Normal Ventilation Systems Containment maximum and minimum design pressures are based on conservative assumptions of initial pressures and temperatures within .the containment. The functional capability of the containment normal ventilation systems to maintain initial containment atmospheric conditions within the range of temperature and pressure defined for normal plant operation is discussed in Section 9.4.6. The maximum average containment temperature during normal plant operations is 110*F. O Amendment G l 6.2-12 April 30, 1990 1
z CESSAR EEncam. (D 6.2.1.1.3.11' Post-Accident Containment Pressure and Temperature Monitoring Both channels of containment pressure ~are monitored in the main control room. Containment atmosphere temperature and IRWST . temperature are monitored in the main control room. Section 7.5 contains a detailed discussion of range, accuracy, and response of the instrumentation used and the type and ~ accessibility of
' recorders provided. The tests conducted to qualify the instruments for use in the~ post-accident containment environment are discussed in-Section 3.11.
6.2.1.2 Containment Subcompartments I subcompartments within the containment are the reactor
~
I The
- cavity, the ' pressurizer compartment and the IRWST. The Leak
- Before Break (LBB) concept applies- to the RCS piping and the surge line. Therefore, the dynamic effects of pipe ruptures in '
the reactor. cavity and pressurizer compartment are not considered. There are no high energy lines in the IRWST. 6.2.1.3 Mass and Energy Release Analyses for Postulated .; Loss-of-coolant Accidents i
- (
LOCA : mass / energy release analyses can be classified into the following phases: blowdown, refill, reflood, post reflood, and long term. The blowdown period extends from time zero until the primary system depressurizes to essentially the containment ; ' pressure. During . blowdown, most of the initial primary coolant is released to the containment as a two phase mixture. Following G blowdown, the water for releases is provided by the Safety , Injection System (SIS).
-There is an important distinction between hot leg breaks and cold leg. breaks for LOCA post blowdown analyses. For a hot leg break, the majority of the SIS supplied water leaving the core can vent ,
directly to the containment without passing through a steam generator. Therefore, since there is no mechanism for releasing f the steam generator energy to the containment for a hot leg-break, only the blowdown period must be considered. Conversely, for cold' leg breaks, the water can pass through a steam generator before reaching the containment so that post blowdown releases to the containment must be considered for cold leg breaks. , O Amendment G 6.2-13 April 30, 1990 i r_.--_-__.__---___-__-_____ -
3CESSAR ?!%mt i The first post blowdown period is- refill. During refill, the e i~ SIS water refills the bottom of the reactor vessel to the bottom of. the core. This period is conservatively omitted from the analysis.
'The second post blowdown period is the reflood period. 'During reflood, SIS water floods the core. Reflood is assumed to end when the liquid level in the core is 2 feet below the top of the active core. During reflood, a significant. amount of the SIS water entering the core is postulated to be carried out; of the core.by.the steaming action of the core to coolant heat transfer ,
process. This fluid then passes through a steam generator where ' reverse (i.e., secondary to primary) heat transfer heats it before it reaches the containment. The residual steam generator secondary energy is sufficient to convert all of this fluid to superheated steam during the initial part of the reflood period.
. Subsequently, as the generators are cooled by this process, there is not enough heat transfer to boil all of the : fluid passing.
through the tubes. This causes the break flow to change from pure steam to two phase. In time, as the entire-NSSS cools, the flow to - the containment will be subcooled since the safety injection water.is subcooled. -The onset of the two phase release to the containment may or may not occur before the end of. reflood; typically, this occurs close to the end of reflood. The . potential release of subcooled fluid to the containment does not occur during .reflood when conservative system parameters are. > utilized. G The third post blowdown period is the post reflood period. During this time frame, the dominant process is the continued cooling of the steam' generators by the SIS water leaving the core. The release to the containment 'during this time frame ia generally two phase due to the cooling of the steam generators. The post reflood ends when the affected steam generator has essentially reached the containment temperature. The final post blowdown period is the long term cooling period, l l which begins at the end of post reflood. During long term, the ; dominant mechanisms for release rates are the decay heat and the l L cooling of all NSSS metal. Long term ends when the containment , pressure and the environment pressure are essentially equal. 6.2.1.3.1 Mass and Energy Release Data ll Mass and energy release data for the suction leg, discharge leg, and hot leg break cases are given in Part A of Tables 6.2.1-4 through 6.2.1-8. For cold leg breaks (pump suction and L- discharge), some of the post blowdown SIS water is postulated to l 9 Amendment G 6.2-14 April 30, 1990 l
y L u CESSAR !!Micari:n ; L f) lV L : spill directly to the containment floor whenever the reactor l vessel annulus is full. The vessel spillage data associated with these breaks are also given in _ Part A of Tables 6.2.1-4 through 7.2.1-8. 6.2.1.3.2 Energy Sources The following sources of generated and stored energy in the , reactor coolant system and secondary coolant system are # considered: primary coolant, primary walls (including reactor internals), secondary coolant, secondary walls,. safety injection water, core power transient and decay. heat, and steam generator forward-and reverse heat transfer. The initial reactor coolant system water volumes are conservatively calculated based on maximum manufacturing tolerances for the reactor vessel and steam generator tubes. Expansion of the loop components from cold to hot operating - conditions is also considered. The pressurizer water volume includes an allowance for level instrumentation error. Initial conditions in the reactor coolant system are given in f Table 6.2.1-18. A tabulation of sources and amounts of stored j energy is given in Table 6.2.1-24. Figure 6.2.1-17 shows the normalized decay heat curve as a fraction of the initial power level following the accident; a 20% conservatism factor is used for the first 1000 seconds, followed by a 10% factor thereafter. G The initial power level assumed- in the analyses is 3876 MWt. l This is 102% of its nominal 3800 MWt to account for instrumentation error. The higher power-level is cmnservative l for LOCA containment pressure calculations. 6.2.1.3.3 Description of Blowdown Model Blowdown mass and energy release rates are calculated using the CEFLASH4A computer code (Reference 2). A description of the CEFLASH4A code including the conservatisms in modeling- is given below. This section includes justification of-the heat transfer correlations. The following assumptions are made in selecting l 1 input data for the code. A. The CEFLASH4A code model of the heat transfer in a node I allows only one wall per node. Accordingly, the thickness used for the "U" factor for each node wall is selected such i that the energy released from the system is conservatively modeled. 1 O l l i Amendment G J 6.2-15 April 30, 1990 l
CESSAR1!nh a i o i B._ The- CEFLASH4A wall representation uses the total heat capacitance of_all the walls-in the reactor coolant ~ system that actually face a given - node. This. is conservative-since, in' reality, some of the walls will not participate as effectively as others _in the heat transfer' process. For ! example, the geometry of the_ flow path is,such as to_ allow l some components to partially shield _otaers from the flow. This'effect is conservatively omitted f3om the modeling. l C. Although much of the steel facing the c >olant in the reactor- . coolant system is stainless cladding (} = 12 ' Btu /hr-f t *F) ,- ' a conservative . carbon steel condtctivity (K ' =; 26 Btu /hr-ft *F), is used for .the entire wall. . This ~ conservatively overpredicts the -energy released from all such walls. I D. Wall surface heat transfer coefficients are assumed to be infinite. E. .All primary water volumes are conservatively increased from their nominal design values in order to obtain an - upper bound: for the available mass and energy in the system prior to LOCA. The pressurizer water volume includes an allowance for level instrumentation error.- Pressure and temperature expansion of the reactor coolant system.and steam generator L to the normal operating condition is~ included. l G. I F. An accepted (Jens Lottes) two pilase heat- transfer correlation is used'for the core to coolant-heat transfer -; whenever the flow.through the core is not pure steam.- i G. Heat transfer across the steam generator tubes is modeled with the same heat transfer coefficient in both the-forward and reverse directions. This is conservative since' it ' maintains a nucleate' boiling heat transfer coefficient . on the secondary side during the LOCA blowdown. In reality, i the reactor trip following the LOCA would result in a L turbine trip which would close the turbine stop' valves and the secondary side heat transfer coefficient would decrease to a small natural convection coefficient. Using the initial steady state full power overall heat transfer. coefficient conservatively maximizes the reverse heat transfer. l H. The turbine stop valves are assumed to close at 0.01 ! seconds. This is conservative since it keeps energy within L the NSSS which in turn is a source of energy for containment I pressurization. O l l Amendment G l 6.2-16 April 30, 1990
,. . n CESSAR sinhm. 1 I. The Main Feedwater Isolation Valves are assumed to close , only after the generation of a Main Steam. Isolation Signal- 1 (MSIS) of 4 psig containment pressure. This signal occurs very rapidly (~1 second). Main Faedwater I s o l a t i o n V a l v e -- closure is assumed to take 5 seconds (step function), with. !. an additional allowance of-1.35 seconds.for the MSIS signal delay. Feedwater flow and enthalpy are kept.at their normal values due to4 the short. times involved. As -an additional conservatism, the feedwater is assumed to be added at'the end.-of' blowdown, so that ' the . steam generator secondary temperatures (~500*F) during the blowdown are not lowered b)
, the relatively cold feedwater (~450*F). . Note that th r feedwater is hot relative to potential peak LOCA containment temperatures (-275*F) , so that feedwater addition at the erd of blowdown is conservative both for blowdown' and' for reflood calculations.
J; -Emergency feedwater flow is conservatively omitted since it
'is cold (4 0
- F - to - 120'F) relative to -both blowdown- and-reflood conditions.
6.2.1.3.4 Description of Core Reflood Model
; ; 1Reflood . mass and energy release-rates are calculated using the FLOOD-MOD 2 computer code (Reference 3). Heat transfer is i conservatively modeled for core, vessel walls, vessel internals, loop metal,. steam generator tubes, steam generator secondaries, and- steam generator secondary walls. The FLOOD-MOD 2 code hydraulics calculates flow rates and pressure. The heat transfers G process. predicts fluid enthalpies.. Fluid densitics- are calc 61ated as functions of pressures -and enthalpies. The.
conservatisms in the model are as follows: A. The reflood and post reflood mass and energy release rates are dependent on containment pressure. The- transient containment pressures for input-to FLOOD-MOD 2 were obtained by iteratively running FLOOD-MOD 2 and the -containment
-pressure code CONTRANS until the containment pressures agreed.
B. A one-dimensional heat transfer model is used for all wall heat transfer calculations. This is demonstrated in Reference 4 where comparisons of one-dimensional models and otherwise identical two-dimensional models show that one-dimensional modeling is conservative. O Amendment G 6.2-17 April 30, 1990
'CESSARiin h w !
1 e C. A.nucleage. boiling heat transfer coefficient of 10000 Btu /hr-ft *F is used to model - the heat transfer from the steam : generator tubes to the fluid. This coefficient ,
-represents an upper: limit, and is conservatively used at all I times throughout the tubes. )
l D. During reflood, calculations are made on the steam generator secondaries to predict the liquid levels. These i calculations show that_a conservatively calculated fraction' l (-25%);of the tube heat transfer area is in. contact with the l secondary steam; the remainder of the tubes is . in contact I with the_ -secondary liquid. A conservative Nusgelt l condensation heat transfer coefficient of 2250 Btu /hr-ft 'F 1 is used in conjunc, tion with the tube area exposed tg steam; ' a natural circulation coefficient of 300 Btu /hr-ft *F is used for the rest of the tube area.
~E. The thermal resistance corrfspon_dgng to the steam generator tubes is 0.00037 (Btu /hr-ft 'F) . This value is also used '
in calculating secondary to primary heat transfer.
-F. The carryover rate fraction (CRF) used- during reflood -is J 0.05 up to the-18-inch core level, increases to 0.8 at the '
24-inch core level, and is kept at 0.8 until the 10.5 foot level-is reached. 10.5 feet is 2 feet below the level of-the top-of the active core. Other-variables, such as core. inlet-temperature, pressure, flow rate, linear heat rate, or other experimental-data are not used to determine the CRF. G. Reflood is assumed to terminate when the 10.5 foot _ quench level in the core is reached. G H. 120% of the standard decay heat (Figure 6.2.1-17) curve is used as a conservatism for the available energy sources. I., During reflood, credit is taken for the condensation of steam in the annulus by the cold SIS. water.- As a conservatism, credit is not taken unless the reactor vessel annulus is full. The SIS flow is injected directly into the , annulus. Also, as an additional conservatism, credit is not ' taken when' the SIS rate is too low to thermodynamically condense all of the steam in the annulus. The percentage of the total steam flow condensed varies slightly with time for each case. For suction leg and discharge leg cases, credit is taken for the condensation c,f approximately 42% of the-total steam flow when the annulus full and the thermodynamic criteria are simultaneously met. No credit for condensation is taken after the SITS empty. Amendment G 6.2-18 April 30, 1990
i L CESSAR:sinL mr l l y 6.2.'1.3.5 Description of Post Reflood Model l The Post Reflood Model is identical to the reflood model except L that, at the end of reflood, the carryout rate fraction (CRF) is changed- from 0.8 to 1.0. This conservatively increases the system flow rates due to the increased CRF. The flow rates are further enhanced by the fact that the core liquid height is-now i constrained at the 10.5 foot level, which maximizes the available- ' driving -head between the annulus level and the core in the l
' FLOOD-MOD 2 flooding equation. All heat transfer coefficients are j kept at the values used for the reflood analysis. Condensation l is - analyzed as :previously described; however, there is not sufficient spillage to completely thermodynamically condense the steam so that credit for condensation has not been taken. j 6.2.1.3.6 Description of Long-Term Cooling Model The heat generation rate from shutdown fissions, heavy isotope decay, and fission product decay is shown in Figure _ 6. 2.1-17. I For conservatism, the long-term analysis assumes that decay heat l is added.to the reactor vessel water at a 20% greater rate than '
that predicted by the decay heat curve up to time 1000 seconds, l after which a 10% margin is used. Following the post reflood period outlined above, the mass / energy source terms for long-term containment analysis are computed concurrently with' the containment back pressure in- the , containment code. The steam flows out the break will be a. I function _of the,depressurization of the containment, decay heat l (plus margin) 'and primary metal-to-primary fluid heat transfer. G l The steam generator secondary fluid, tube, thick and thin metal stored energy _are used to superheat the steam prior to discharge into the containment. The long-term energy release for the double-ended discharge leg LOCA with maximum SIS flow is given in Table 6.2.1-23. This long-term data is based on the containment pressure and the following calculational method: The reactor coolant system is assumed to be a vessel containing a constant mass of saturated water. The pressure in the vessel is assumed to be the containment pressure. SIS water is injected
'into the vessel. Steam is formed at a rate determined by de:ay -
heat, RCS metal-to-coolant heat transfer and the rate of containment depressurization. Uince the water in the vesse3 is saturated, boiling will occur even without decay heat or matal heat transfer as the containment pressure decreases. The difference between the SIS injection rate and the steaming rate is the spillage rate to the IRWST. It is conservatively assumed Amendment G 6.2-19 April 30, 1990
CESSAR:ERWicciar
.. L that all of the decay and RCS metal heat transfer goes into-creating. steam. The spillage flow is assumed to have- the same enthalpy-as.the SIS' injection enthalpy.
The rate:of steam production is du f
~
9RCS ilt "sta- h g -h in Where: m ata =: steam flow, lbm/sec qRCS-
* "Y "" " " " ** * **' !'"
du M dt -= rate of change of internal energy in RCS as a-result of depressurization, Btu /sec h - saturated steam enthalpy at containment . 9 pressure, Btu /lbm h = SIS in e tion enthalpy, Btu /lbm in G The steam created from the decay.and RCS metal heat is saturated. For cold leg breaks it is conservatively' assumed that -all the steam passes through the steam . generators and leaves at the secondary side temperature. The long-term-energy release to.the containment is given by
-(m h) Release " "stm h g +g gg Where:
(m h) Release = rate of steam energy released to containment, Btu /sec ggg = rate -? heat transfer from steam ge m ators to RCS steam, Btu /sec O Amendment G 6.2-20 April 30, 1990
= _ _ _ _ _ . . _ . . .
i LCESS'AR l;necuer o m-An energy balance-is made on the steam generator secondary side. Secondary side metal heat transfer is included, dH SG " 144 dP dt 9SG Walls. ~9SG + 778 dt Where: 1 H gg = steam generator total enthalpy, Btu ggg y,yy, = rate of energy addition to secondary fluid from steam generator walls,-Btu /sec , o P gg = pressure in steam generator secondary side, l psia q gg. = rate of heat transfer to RCS steam, Btu /sec 6.2.1.3.7 Single Active Failure Analysis . i The cases presented in Tables 6.2.1-4 through 6.2.1-7 show the ; mass / energy source terms with maximum safety injection (no pump 4 or power source faalure) and minimum safety injection (failure of= ( 1 diesel). 6.2.1.3.8 Metal-Water Reaction Energy addition to the' containment atmosphere resulting from the maximum allowable 1% zirconium water reaction is ' based on a zirconium mass in the active . core of' 58498 lbm. Using a molecular-weight of 91.22 for zirconium and a reaction energy of 25g900 Btu /lbm mole, the 1% metal-water reaction' produces 1.622 x ;' 10 Btu. This energy is not included ~in the mass / energy source terms of Tables 6.2.1-4 through . 6. 2.1-8. Note that this energy. will have a very small . ef fect on the containment pressures. Also, with the SIS flow' rates to the core with the conservative LOCA mass and energy release model, no core damage will occur. 6.2.1.3.9 Energy Inventories Energy balances for the most severe LOCA cases listed in Table ' 6.2.1-1 are provided in Table 6.2.1-24. 6.2.1.3.10 AdlIttonal Information Reactor vessel pressure versus time for the LOCA cases listed in Table 6.2.1-1 are given in Part B of Tables 6.2.1-4 through 6.2.1-8. O Amendment G 6.2-21 April 30, 1990
CESSAR%BL m. O Chronology of events for the LOCA cases listed in Tables 6.2.1-1 are given in Part C of Tables 6.2.1-4 through 6.2.1-8. The' primary side resistance f actors for the FLOOD-MOD 2 code are shown in Table 6.2.1-25. Curves of Safety Injection Flow vs time are provided in Figures 6.2.1-18 through 6.2.1-21. 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures Inside Containment-Following a postulated - main steam line break (MSLB) or ' a main feedwater line break (MFLB) inside the containment, the' contents of one steam generator (affected) will be released to the containment._ Most of the contents of the other steam generator (unaffected) will be isolated by the main steam isolation valves (MSIV) and main feedwater isolation valves (MFIV). Containment pressurization following a secondary side rupture depends on how much of the break fluid enters the containment atmosphere as steam.. MSLB flows can be pure steam or two-phase. MFLB . flows are two-phase. With a pure steam blowdown, all of the_ break flow enters the. containment atmosphere. With two-phase blowdown, part-of the liquid-in the break flow boils off in the containment and-is also added to the atmosphere, while the rest falls to the sump and contributes nothing to containment' pressurization. .For MSLB cases with large break areas, steam cannot escape fast enough a from the two-phase region of the affected steam generator, and the two-phase level rises rapidly to the steam line nozzle. A two-phase- blowdown results.' The duration of this blowdown is short;.therefore little primary-to-secondary heat transfer takes place and the break flow is largely liquid. For MSLB cases with small break areas, steam can escape fast enough from the two-phase region of the affected steam generator so that the level swell does not reach the steam line nozzle. A pure steam blowdown results._ Because of the pressure reducing effects of active and passive containment heat sinks, the highest peak containment pressure resulting from.a MSLB for a - an set of initial steam. generator conditions occurs for that # ca where the break area is.the maximum at which a pure steam blc:. town can occur. The potent'ial for steam generator two-phase level swell follow 1ng a MSLB increases as power level decreases; therefore, a spectrum of power levels must be analyzed to determine which one results in the peak MSLB containment pressures. O Amendment G 6.2-22 April 30, 1990
l LCESSARtinincm. - 4 l ". b-V - l The feedwater distribution box is below the steam generator water level; therefore, MFLB cases always result'in two-phase blowdowns and do:not produce peak' containment pressures as severe as MSLB'
~
l cases. To permit a determination of the effect of MSLB upon containment pressure, analyses are performed with SGNIII (described in Appendix 6B of Reference 5) at 102, 50, 20 and 0 percent power. The largest breaks at which a pure steam blowdown can occur are determined. The breaks are conservatively assumed to be at the nozzle of one of the steam generators. The cases analyzed are listed in Table 6.2.1-1. The System 80+ plant has integral flow restrictors in-the nozzles 1 of the steam generators. Credit for the flow ' restrictors is , taken in the analysis. In the plant, th'e main steam isolation signal (MSIS)' of the engineered safety features actuation system (ESFAS) closes the MSIVs: and the 'MFIVs. MSIS is generated either by a steam generator low pressure signal or a containment high pressure l signal. .The'MSIVs close in 5.0 seconds. The valve closures have been considered in the analysis. , The emergency feedwater system functions automatically during 4 l- MSLB to ensure that a heat sink is always available to the reactor coolant system by supplying cold feedwater to maintain an o adequate water inventory in the unaffected steam generator. Flow to the affected steam generator is modeled. No credit 'for emergency feedwater flow to the unaffected steam generator is take9 in the MSLB analysis. The total volume of fluid between the MSIVs and each steam generator is assumed to be 2000 cubic feet (total for two' steam lines). The volume of fluid between the MSIVs.and the turbine stop valves is assumed to be 14000 cubic feet maximum. The maximum volumes are considered in the MSLB analysis. There are , two MFIVs in each feedwater line. The maximum volume of fluid between the upstream MFIV and each steam generator is assumed to ; be 700 cubic feet. The flashing of this fluid into the affected steam generator and-then into the containment is considered in the analysis. These assumed volumes conservatively exceed the System 80+ volumes. 6.2.1.4.1 Mass and Energy Release Data Mass / energy release data for the MSLB cases listed in Table 6.2.1-1 are given in Part A of Tables 6.2.1-9 through 6.2.1-16. 1 U Amendment G 6.2-23 April 30, 1990 l
CESSAR REncam. O 6.2.1.4.2 Single Failure Analysis The availability of non-emergency power is conservatively assumed since it allows the continuation of reactor' coolant pump operation. This maximizes the rate of heat transfer to the affected. steam generator which maximizes the rate of mass / energy release. With non-emergency power, a diesel failure need not be postulated. .There is an MSIV in each main steam line. The MSIVs have been designed to close based on a conservative calculation which maximizes the dynamic pressure loading on the valve for all possible flow rates and qualities. Each valve has dual solenoid-valves to assure closure even with a single failure in the control system.. Single failure of the actuation signal will not prevent valve closure since both trains of MSIS actuation are provided to each. MSIV. Any failure would result in. the valve going to the closed position.so that no additional steam could be added to the containment. The other MSIV isolates the unaffected ~ steam generator. Each valve is tested periodically. Therefore, the failure of.an MSIV is not considered to be a credible event; however, MSIV failure events have been considered. There are two MFIVs in' series in each main feedwater line.- If cne MFIV. fails, the second MFIV would provide isolation. All cases analyzed considered.the flashing of the fluid in the lines from. the' upstream- MFIVs to the- affected steam generator; therefore, there. is no need to do- a separate analysis assuming MFIV failure. G The MSLB data in Tables 6.2.1-9, 6.2.1-11, 6.2.1-13 and 6.2.1-15 are based on a loss of one CSS train. The data in Tables 6.2.1-10, 6.2.1-12, 6.2.1-14 and 6.2.1-16 are based on an MSIV failure. 6.2.1.4.3 Initial Conditions Nominal full load for System 80+ is 3800 Mwt. Reactor coolant system parameters at 102 percent of full power are given in Table 6.2.1-18. The steam generator pressure varies from 1021 psia (nominal full load) to 1100 psia (no load). The initial steam generator inventory is calculated assuming manufacturing tolerances which maximize the initial inventory. The increase in the initial inventory resulting from thermal expansion of the steam generator is included. Amendment G 6.2-24 April 30, 1990
CESSAR EHL"ic== 1 O ! 6.2.1.4.4 Description of Blowdown Model [ The SGNIII digital computer code described = in Appendix 6B of Reference 5 is used'for the secondary system pipe break analysis. ! All significant equations, including those for the calculation of j primary-to-secondary, core-to-coolant, and metal-to-coolant heat transfer, and for the calculation of steam. separation and i moisture carryover, are discussed in Appendix 6B of Reference 5. Experimental justification for all heat transfer coefficients, } steam separation velocities, and two-phase - flow correlations is i provided in Appendix 6B of Reference 5.
~ Steam line capacity is modeled by performing mass, energy and e volume balances on'a steam line node. Figure 6.2.1-22 shows the flow paths into and out of the steam line node. -The-mass, energy, :
and volume balances for the steam line node are given below. N = Im S = Imh } V=0 ! Where: Im = my+m2 + "4 - "B ~ "T o for slot breaks, and 1 Im=my+m2 ~ "T ~ "B2 for guillotine breaks (see Figure 6.2.1-22 for subscript definition). ; I The , contribution to containment pressure of feedwater flow is handled by feedwater flow addition' to the affected steam generator and the boiling off of the feedwater by primary to secondary heat transfer. The feedwater flow is the sum of'the pumped feedwater ' flow prior to isolation plus the expansion of the fluid in' the feedwater line between the affected steam generator and its MFIV. The feedwater flow pumped to' the affected steam generator is conservatively modeled as 200% of the initial feedwater flow to account for spiking. No degradation of the feedwater flow occurs until the closure of the MFIVs. For consistency, no feedwater is added to the unaffected steam generator. Following closure of the MFIVs, there is an inventory of feedwater between the MFIVs and the affected steam generator. As O the affected steam generator depressurizes, this inventory starts Amendment G 6.2-25 April 30, 1990
CESSAR!ah mt O to' boil. As' steam in the line expands, this.feedwater inventory
~
is pushed into-the steam generator and_is boiled off by primary 4 to ~ secondary heat transfer. The expansion' of the feedwater inventory into the affected steam generator has - been considered in the analysis. The expansion is assumed to~be isentropic. The,isentropic-expansion'of'the feedwater downstream of the MFIV - is determined in SGNIII. As the affected steam generator depressurizes, the feedwater expands. At first the feedwater is-subcooled and the fluid which expands into the'affected steam generator is pure - liquid, once the steam generator pressure
- drops below the saturation pressure of the ~ feedwater, flashing starts to-occur and then the fluid which expands into the steam generator is two-phase. The equations for each . phase of the
- isentropie e::pansion process are given below:
l Subcooled
, s y(PSG, T) =s o
M{vy(Pgg, T) = V MI ~ M I
,, t- t - At
" # 1(P gg, T)
Saturated G
+ gYg (PSG) " V Mfgv (PSG}
Msf f(Pgg) +Msq q(Pgg) M
=s 0 Mf+ g (Mr+ M) (Mf+ M) gt gt- Q t
Mhf f(Pgg) + Mgqh (Pgg) mh = m g., , , g I 9 Symbols M Feedwater mass, lbm V Volume of feedwater downstream of MFIV, ft 3 P Pressure of affected SG, psia SG T Feedwater temperature, 'F h Specific enthalpy, Btu /lbm v Specific volume, cubic feet /lbm Amend: sit G 6.2-26 April 3', 1990
LCESSARE!nAma u s Specific entropy, Btu /lba 4 m Feedwatar'flowrate-from isentropic expansion,-lbm/sec mh Feedwater enthalpy rate from isentropic expansion, i Btu /sec
+ t Time, seconds i at-' Time step, seconds !
-Subscripts 1 Subcooled liquid.
f- Saturated liquid i
- g. Saturated steam '
O Initial The feedwater flow from isentropic expansion is added to .the pumped _feedwater flow. The ' pumped feedwater flow is conservatively ~ assumed to be a constant 200% of the initial feedwater flow until the-MFIVs close. The MSLB mass / energy data given in Part A of Tables 6.2.1-9 through 6.2.1-16 represent the-total release to the-containment. : The mass / energy contributions from.the steam lines and.feedwater l lines are included in Tables 6.2.1-9.through 6.2.1-16. Tbsre are- l
;\ no additional releases. G.
6.2.1.4.5 Energy Inventories - An energy balance for.the most severe MSLB. cases listed in Table 6.2.1-1~are provided in Tables 6.2.1-24. 6.2.1.4.6 Additional'Information The flow area of the main steam lines is 4.28 square feet. For the MSLB analysis, the postulated rupture is assumed to occur at ' the nozzle of one of the steam generators; therefore, the fL/D-from the-affected steam generator to_the break is zero. In the MSLB- analysis, the fL/D from the unaffected steam generator to , the break is conservatively assumed to be 10 since the fL/D in ' the System 80+ plant is greater than '10. The flow restrictor area'is 1.28 square feet. Pressures in the affected and unaffected steam generators for l
- each: MSLB case listed in Table 6.2.1-1 are given in Part B of l- Tables 6.2.1-9 through 6.2.1-16.
I Chronology of events for the MSLB cases listed in Table 6.2.1-1 are given in Part C of Tables 6.2.1-9 through 6.2.1-16. lO Amendment G 6.2-27 April 30, 1990
- CESSARanh a Oa Feedwater: flow to the e.ffected.stean generator for each MSLB case listed -in Table 6.2.1 are shown on Figures 6.2.1-23 through 6.2.1-30.
6.2.1.5 Minimum Containment Pressure Analysis for i 1 Performance Capability Studies on Safety Injection (' System 6.2.1.5.1 Introduction and Smmmanj Appendix K to 10 CFR 50 lists the required and acceptable-features of Emergency- Core Cooling ' System (ECCS) evaluation models (Reference 6). . Included in this list is the requirement , that the- containment pressure used in the evaluation of Safety Injection System (SIS) performance not- exceed a. pressure calculated conservatively for that . purpose. This section presents the minimum containment pressure used in the SIS performance analysis for the System 80+ Standard Design, which is presented in Section 6.3.3. 6.2.1.5.2 Method of Calculation The calculations reported in this section- use the methods ' described ~ in Reference 7 and approved in Reference 9. - In . this ' . method, : the CEFLASH-4 A (Reference 2) computer program-determines the mass and' energy released to the containment during the ; blowdown phase of a postulated LOCA. The COMPERC-II . computer program .(Reference 8) determines both the mass and energy released.to the containment during the refill /reflood phase and'- the minimum cc.1tainment pressure response used to evaluate the o effectiveness of the SIS. 6.2.1.5.3 Input Parameters The input for the minimum containment pressure analysis for the System 80+ Standard Design presented herein is consistent with the SIS performance analysis of Section 6.3.3 which uses - the results of this section. \c 6,2.1.5.3.1 Mass and Energy Releise Data The mass and energy released to the containment for the limiting
'large break LOCA, 1.0 x DEG/PD, is listed as a function of time in Table 6 2.1-26.
. The quantity of safety injection fluid that spills from the break is discussed in Section 6.2.1.5.3.5.
l l O' l Amendment G 6.2-28 April 30, 1990 r -
a i CESSAR !!nhu- ! ((()
- i. ' 6.2.1.5.3.2- Initial Containment Internal Conditions -i The initial containment conditions used for this analysis are:
Temperature 60*F (minimum) j Pressure 14.3 psia (minimum) ! Relative Humidity 100% (maximum) 4 i For - each parameter, the conservative direction with respect' to > minimizing the-containment pressure appears in parentheses.. ! 6.2.1.5.3.3- Containment Volume The net- frge containment volume used for this analysis is , 3,547,000 ft (maximum). , 6.2.1.5.3.4 Active Heat Sinks In order to conservatively maximize the heat removal capacity ofE i ithe containment active heat sinks, the containment sprays and fan coolers-are actuated in the shortest-possible time following.the break, and operate at their maximum capacity using the minimum temperature for both the spray water and cooling water-. 1 O The operating parameters used for the containment sprays are as follows: Flow rate 13,000 gpm.(total, all pumps) G Temperature 60*F during-SIS Injection Mode The heat removal capacity of the fan coolers used in the analysis is shown in Figure 6.2.1-31. . The containment sprays and fan coolers are conservatively modeled to be actuated coincident with break opening. 6.2.1.5.3.5 Steam Water Mixing The effect on' containment pressure due to condensing containment steam with spilled SIS water is calculated in the manner described in Section III.D.2 of Reference 7. The effective SIS spillage rate is shown in Figure 6.2.1-32. The spillage rate is conservatively determined using the maximum flow rate from all four SI pumps and initiating the SI pump injection when the downcomer has been filled by the SITS. O Amendment G l 6.2-28a April 30, 1990 l
l CESSARiintncam. 1 l e-6.2.1.5.3.6 Passive Heat Sinks' l The surface areas and thickness of all exposed containment -1 passive heat sinks are listed in Table 6. 2.1-21. The materia G properties used for this analysis are also listed in Table 6.2.1-21.
/ 6.2'.1.5.3.7 Heat Transfer to Passive Heat Sinks
, The condensing heat transfer coefficients between the containment atmosphere and the passive . heat sinks are calculated in - the manner described in Section III.D.2 and Figure III.D.2 of Reference 7. The variation of the condensing heat- transfer coefficients as a function of time is shown quantitatively in
-Figure 6.2.1-33. -j
(. 1 l/ 6.2.1.5.3.8 Containment Purge System l The minimum containment pressure analysis assumes that the 8-inch l' diameter purge system. is operating from the time of the LOCA~ 1- initiation until-the isolation valves close, after a containment i l isolation:-actuation signal. It is conservatively assumed that only. dry' air is purged from the containment. l l 6.2.1.5.4 Results 4 ll For the limiting large break LOCA, 1.0 x DEG/PD, the minimum containment pressure response used in analyzing the effectiveness-of the SIS is shown in Figure 6.2.1-34. The responses of the 4 containment. atmosphere and containment sump temperatures are shown'in Figures 6.2.1-35 and 6.2.1-36, respectively.
~
The containment response is used in the SIS performance analysis presented in Section 6.3.3. 6.2.1.6 Testine and Inspection 7 L Testing and inspection requirements for the containment are i discussed in Section 6.2.6. Testing and inspection requirements ! for other engineered safety features that interface with the containment structure are discussed with the applicable system descriptions. O' Amendment G 6.2-28b April 30, 1990
1
.CESSAR inWicam,. i l
6.2.1.7 Instrumentation Applications I The containment pressure is measured by independent pressure transmitters located' at . widely separated points within_ the j containment. Refer to Section 7.3 for a discussion of pressure ) as an input to the engineered safety features actuation system j (ESFAS). Refer to Section 7.5 for a discussion of the display j
. instrumentation associated with pressure. G
-l The containment airborne radioactivity is monitored by- the i airborne radioactivity monitoring system. Hydrogen concentration ;
is' monitored in the containment by the hydrogen monitoring system discussed in Section 6.2.5. Temperature sensors are positioned at appropriate locations throughout -the containment. The temperature is displayed in - the main control room along with high-temperature alarms. 1 i O a 6
.i O
Amendment G 6.2-28c April 30, 1990
@ESSAR BBWico O
REFEREMCE3 FOR SECTION 6.2.1
- 1. CONTRANS, " Description of the CONTRANS Digital Computer Code for Containment Pressure and Temperature Transient Analysis," CENPD-140-A, June 1976. g
- 2. "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis," CENPC-133P, August 1974 (Proprietary).
"CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis (Modifications) , " CENPD-133P, Supplement 2, February 1975 (Proprietary).
o "CEFLASH-AA, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis," CENPD-133, Supplement 4-P, April 1977 (Pr oprietary) .
"CErLASH-4A, A FORTRAN 77 Digital Computer Program for Reactor Blowdown Analysis," CENPD-133, Supplement 5-P, June 1985 (Proprietary).
- 3. " FLOOD-MOD 2 - A Code to Determine the Core Reflood Rate for a PWR Plant with Two Core Vessel Outlet Legs and Four Core l h vessel Inlet Legs," Interim Report, Aerojet Nuclear Company, November 2, 1972.
- 4. F. Kreith, " Principles of Heat Transfer." International Textbook Company, 1958.
- 5. CESSAR, " Combustion Engineering Standard Safety Analysis Report," Combustion Engineering, Inc., docketed December 19, 1973.
- 6. "ECCS Evaluation Models," Apper. dix K to Part 50, 39 FR 1003, January 4, 1974.
- 7. " Calculative Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132, August 1974 (Proprietary).
" Updated Calculative Methois for the C-E Large Break LOCA Evaluation Model," CENPD-132, Supplement 1, February 1975 (Proprietary).
" Calculational Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132, Supplement 2, July 1975 (Proprietary).
O Amendment G i 6.2-28d April 30, 1990 i 1
l CESSAR BRW,can.= 0 " Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS," CENPD-132, Supplement 3-P-A (Proprietary).
- 8. "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," CENPD-134, August 1974 (Proprietary).
"COMPERC-II, A Progrha for Emergency Refill-Reflood of the Core (Modification) ," CENPD-134, Supplement 1, February 1975 G (Proprietary).
"COMPERC-II, A Program for Emergency Refill-Reflood of the core," CENPD-13_4, Supplement 2, June !985.
- 9. Letter, O. D. Parr (NRC) to F. M. Stern (C-E), June 13, 1975. Letter, O. D. Parr (NRC) to A. E. Scherer (C-E),
December 9, 1975. Letter, D. M. Crutchfield (NRC) to A. E. Scherer (C-E) , July 31, 1986. O l I l 4 / Amendment G j 6.2-28e April 30, 1990 l
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)
,-~ :
TABLE 6.2.1-1 , (Sheet 1 of 3) SPECTRUM OF POSTUIATED ACCIDENTS Loss-of-Coolant Accidents (LOCA) i l Double-Ended Suction Leg Slot (DESLS) i Maximum SIS Flow, 9.8175 Square Feet Double-Ended Suction Leg Slot (DESLS) Minimum SIS Flow, 9.8175 Square Feet [ ' \ Double-Ended Discharge Leg Slot Break (DEDLS) G Maximum SIS Flow, 9.8175 Squate Feet , Deuble-Ended Discharge Leg Slot (DEDLS) Minimum SIS Flow, 9.8175 Square Feet Double-Ended Hot Leg Slot (DEHLS) 19.2423 Square Feet O Amendment G April 30, 1990
CESSAR c'If#ico O TABLE 6.2.1-1 (Cont'd) (Sheet 2 of 3) SPECTRUN OF POSTULATED ACCIDENTS Main Steam Line Breaks (MSLB) 4 102% Power, 8.72 Square Feet, Loss of One CSS Train 2 102% Pover, 8.72 Square Feet MSIV Failure 50% Power, 8.72 Square Feet Loss of One CSS Train SO% Power, 8.72 Square Feet O1 G l MSIV Failure 20% Power, 8.72 Square Feet Loss of One CSS Train 20% Power, 8.72 Square Feet MSIV Failure 0% Power, 4.50 Square Feet Loss of One CSS Train 0% Power, 4.50 Square Feet MSIV Failure O. Amendment G d April 30, 1990
. _ ~ - - - - - - _ _ _ _ _ . - __ -.
CESSAR E h O TABLE 6.2.1-1 (Cont'd) (Sheet 3 of 3) SPECTRUM OF POSTULATED ACCIDENTS Subcompartment Pipe Ruptures There are no Subcompartment Pipe Ruptures G . Minimum containment Pressure Inadvertent Operation of Coltainment Spray System l O i I i O , Amendment G April 30, 1990
CESSAR !RWeu... i O TABLE 6.2.1-2 CAICULATED VALUES FOR CONTAINMENT PRESSURE PARAMETERS 1 Peak Calculated l Parameter Design-Basis Accident Value Containment Peak 0% Power MSLB with 48.34 Pressure, psig Loss of One CSS Train Peak Subcompartment There are no Sub-Differential Pressure compartment Pipe G psid Ruptures External Containment Inadvertent Operation of -1.83 i Pressure, psig Containment Spray System I i O I i l l-- 1 l LO Amendment G April 30, 1990
r i
~
I:IEllllAlltRNnine.n.. ([]) ! TABLE 6.2.1-3 PRINCIPAL CONTAIMMENT DESIGN PARAMETERS i Parameter Desian Value : Internal Design Pressure, psig 49.0 G External Design Pressure, psig -2.0 Peak Temperature, 'F 290 Containment (steel) Design Temperature, 'F 290 Internal Dimension Diameter of Sphere,3ft 200 4 Net Free Volume, ft 3.377 x 10 Design Leak Rate, percent free volume / day First 24 hours 0.50 () After i day 0.25 t e Amendment G April 30, 1990
i CESSARIM&u. ; 1 s TABLE 6.2.1-4 (Sheet 1 of 9) i DOUBLE-ENDED SUCTION LEG SIDT BREAK - MAXINUM SIS FIDN 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data 1 l 1
^
Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lbe/sec) (Btu /sec) 0.000 0.00000E+00 0.00000E+00 0.028 7.55833E+04 4.17879E+07 0.053 7.47554E+04 4.12574E+07 j 0.102 7.54066E+04 4.15729E+07 1 0.150 7.74168E+04 4.27427E+07 0.202 7.77315E+04 4.29866E+07 i 0.253 9.03920E+04 5.00891E+07 1 0.300 9.01567E+04 5.00007E+07 i t 8.78716E+04 o~ 0.352 0.501 8.72964E+04 4.88015E+07 4.86059E+07 0.651 7.91404E+04 4.41690E+07 0.801 9.19291E+04 5.14526E+07 0.951- 9.09209E+04 5.10498E+07 1.202 8.54119E+04 4.82250E+07 1.400 8.11399E+04 4.60272E+07 1.602 8.08481E+04 4.60648E+07 G 1.800 7.93604E+04 4.53922E+07 2.000 7.74451E+04 4.44592E+07 3.004 6.48128E+04 3.77221E+07 4.015 5.44915E+04 3.25994E+07 5.015 4.07658E+04 2.63472E+07 6.01! 3.57029E+04 2.35677E+07 7.015 3.38417E+04 2.23953E+07 8.015 3.17027E+04 2.11818E+07 9.015 2.95990E+04 2.00184E+07 10.015 2.69626E+04 1.86044E+07 11.015 2.39726E+04 1.70944E+07 12.015 2.00996E+04 1.51119E+07 13.015 1.59931E+04 1.29885E+07 14.015 1.35005E+04 1.14895E+07 15.015 1.12545E+04 9.86448E+06 16.006 8.02625E+03 6.85176E+06 16.200 7.37928E+03 6.64032E+06 16.400 O 16.600 6.69674E+03 6.14117E+03 6.08618E+06 5.60001E+06 Amendment G April 30, 1990
L CESSARIMWien. I O TABLE 6.2.1-4 (Cont'd) (Sheet 2 of 9) DOUBLE-ENDED SUCTION LEG S WP BREAK - MAXIMUM SIS PLOW 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1ba/sec) (Btu /sec) 16.8 5.665^3E+03 5.17392E+06 17.0 5.218iSE+03 4.77524E+06 17.2 4.82674E+03 4.42423E+06 17.4 4.51073E+03 4.12148E+06 17.6 4.21515E+03 3.83064E+06 17.8 3.94891E+03 3.56781E+06 18.0 3.72144E+03 3.35152E+06 18.2 3.53694E+03 3.18325E+06 18.4 3.37361E+03 3.03081E+06 18.6 3.20218E+03 2.87402E+06 1848 3.11835E+03 2.79793E+06 19.0 4.02552E+03 3.32500E+06 19.2 4.60621E+03 3.76169E+06 19.4 5.14842E+03 4.20051E+06 19.6 5.59701E+03 4.50276E+06 19.8 6.06622E+03 4.70455E+06 20.0 6.36098E+03 4.83597E+06 20.2 6.70677E+03 4.92749E+06 g 20.4 6.81109E+03 4.94384E+06 20.6 6.99119E+03 4.96332E+06 20.8 7.01604E+03 4.92808E+06 21.0 7.02967E+03 4.87504E+06 21.2 7.02094E+03 4. 81564 E4 06 21.4 6.93118E+03 4.72114E+06 21.6 6.92645E+03 4.61569E+06 21.8 6.71463E+03 4.46738E+06 22.0 6.64975E+03 4.36366E+06 22.2 6.58945E+03 4.23874E+06 22.4 6.30562E+03 4.07136E+06 22.6 6.14675E+03 3.92726E+06 22.8 5.81248E+03 3.74008E+06 23.0 5.54480E+03 3.57513E+06 23.2 6.44593E+03 3.88707E+06 23.4 6.90779E-03 4.08909E+06 23.6 7.44546E+03 4.26589E+06 23.8 8.03797E+03 4.46482E+06 Amendment G April 30, 1990
CESSARsi&Ln m V TABLE 6.2.1-4 (Cont'd) (Sheet 3 of 9) ) DOUBLE-ENDED SUCTION LEG SIMF BREAK - MAXINUM SIS FLOW (9.8175 ft Total Area) PART A: Mass / Energy Release Data ; j Break Mass Break Energy Time Flow Rate Flow Rate (sec) (Iba/sec) (Btu /sec) 24.0 8.48076E+03 4.55964E+06 24.2 9.05228E+03 4.69236E+06 24.4 9.72404E+03 4.84813E+06 24.6 1.05776E+04 5.03851E+06 24.8 1.04542E+04 4.90047E+06 25.41 9.44470E+02 1.18814E+06 26.41 1.34648E+03 1.68953E+06 26.91 1.54143E+03 1.93351E+06 O 26.92 27.51 8.94029E+02 9.25842E+02 1.12144E+06 1.15972E+06 l 28.61 9.23650E+02 1.15494E+06 29.61 9.20187E+02 1.14951E+06 30.71 9.16348E+02 1 14373E+06 31.81 9.12021E+02 1.13741E+06 I 32.81 9.08077E+02 1.13168E+06 33.91 9.03744E+02 1.12541E+06 35.01 8.99470E+02 7. 11919E+06 36.01' 8.95630E+02 1.11358E+06 37.11 8.91443E+02 1.10745E+06 g 38.21 8.87278E+02 1.10136E+06 39.21 8.83560E+02 1.09591E+06 40.31 8.79431E+02 1.08988E+06 41.41 8.75324E+02 1.08390E+06 42.41 8.71607E+02 1.07849E+06 43.51 8.67535E+02 1.07257E+06 44.61 8.63492E+02 1.06670E+06 45.61 8.59821E+02 1.06138E+06 46.71 8.55802E+02 1.05556E+06 47.71 8.52165E+02 1.05029E+06 48.81 8.48175E+02 1.04454E+06 49.91 8.44207E+02 1.03881E+06 50.91 8.40611E+02 1.03364E+06 , 52.01 8.36673E+02 1.02797E+06 O 53.11 54.11 55.21 8.32799E+02 8.29307E+02 8.25491E+02 1.02235E+06 1.01728E+06 1.01174E+06 Amenament G April 30, 1990
r ,. CESSARHELm.. 9: TABLE 6.2.1-4 (Cont'd) (Sheet 4 of 9) ' DOUBLE-ENDED SUCTION LEG SLOT BREAK - MAXIMUM SIS FIDW 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy i Time Flow Rate Flow Rate ! (sec) (Ibe/sec) (Btu /sec) 56.31 8.21692E+02 1.00622E+06 57.31 8.18252E+02 1.00124E+06 l 58.41 8.14482E+02 9.95785E+05 59.51 8.10730E+02 9.90356E+05 60.51 8.07325E+02 9.85449E+05 61.61 8.03602E+02 9.80078E+05 , 62.71 7.99890E+02 9.74731E+05 J 63.71 7.96526E+02 9.69893E+05 - 64.81 65.91 7.92837E+02 7.8916TE+02 9.64604E+05 9.59337E+05 h, 66.91 7.85842E+02 9.54570E+05 68.01 7 . 8 2.1.9 4 E+02 9.49356E+05 l 69.01 7.78694E+02 9.44640E+05 70.01 7 . 7 'i2 7 4 E+02 9.39478E+05 71.21 7.71713E+02 9.34368E+05 j 72.21 7 . (< 812 3 E+ 02 9.2C?.99E+05 g 73.31 7.64637E+02 9.24282E+05
- 74.41 7.61175E+02 9.19300E+05 l 75.41 7.58037E+02 9.14788E+05 i 76.51 7.54603E+02 9.09852E+05 t 77.61 7.51181E+02 9.04939E+05 l 78.61 7.48020E+02 9.00769E+05 l
79.71 7.44616E+02 8.95885E+05 80.81 7.41182E+02 8.91315E+05 81.81 7.38120E+02 8.86919E+05 1 82.91 7.34767E+02- 8.82105E+05 l 84.01 7.31403E+02 8.77279E+05 85.01 7.04868E+02 8.31552E+05 86.11 6.61409E+02 7.80378E+05 87.21 6.25130E+02 7.37667E+05 88.21 5.94384E+02 7.01458E+05 89.31 5.62762E+02 6.64216E+05 90.31 5.37834E+02 6.34839E+05 O Amendment G April 30, 1990
CESSAR ERWien O TABLE 6.2.1-4, (Cont'd) (Sheet 5 of 9) ' DOUBLE-ENDED SUCTION LEG SIM BREAK - MAXIMUM SIS FIM 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate ! (sec) _(lba/sec) (Btu /sec) 91.41 8.bj829E+02 1.04327E+06 92.51 8.42850E+02 9.94900E+05 G 93.51 8.08250E+02 9.54039E+05 ! 94.61 7.74260E+02 0.13870E+05 ; 95.71 7.42079E+02 8.75829E+05 96.71 7.14550E+02 8.43270E+05 97.81 6.86129E+02 8.09639E+05 98.91 6.59439E+02 7.78039E+05 99.97 6.36500E+02 7.50879E+05 102.41 5.83915E+02 6.88548E+05 107.41 5.84589E+02 6.88267E+05 111.96 4.64917E+02 5.47177E+05 116.51 3.71572E+02 4.37277E+05 I 121.51 2.14004E+02 2.51831E+03 i 126.51 1.45002E+02 1.70623E+05 131.51 1.19874E+02 1.41352E+05 l
\
k O Amendment G April 30, 1990
C E S S A R RH W.c m . O TABLE 6.2.1-4 (Cont'd) (Sheet 6 of 9) SPILLAGE DATA Time, sec. Mass - lha Energy - 10 6 Btu End of Blowdown 24.8 0.0 0.0 End of Reflood 112.01 227,105 69.599 G End of Post Reflood 131.51 303,369 89.365 O l l l l l I O l Amendment G April 30, 1990 l
. ~ .
CESSAR E h ' TABLE 6.2.1-4 (Cont'd) (Sheet t of 9) DOUBLE-ENDED SUCTION LEG SIAT BREAK - MAXIMUM SIS FIM (9.8175 ft2Total Area) PART B: Reactor Vessel Pressure vs. Time Reactor Vessel Time (sec) Pressure (psia) 0.000 2275
- 0.027 2258 0.052 2045 ,
0.102 1701 0.149 1696 0.202 1689 0.252 1687 0.351 1672 0.500 1643 O.650 1628 \- 0.801 1620 < 0.951 1606 1.201 1576 1.400 1556 2.000 1482 3.000 1380 l 4.012 1298 5.012 1203 0.012 1140 G 7.012 1098- i 8.012 1062 I 9.012 1025 10.012 987 11.012 942 l 1 12.012 874 14.012 731 16.012 462 18.000 234 20.000 304 22.000 258 l 22.400 237 O Amendment G April 30, 1990
CESSAR %iR* fica.. O TABLE 6.2.1-4 (Cont'd) (Sheet 8 of 9) DOUBLE-ENDED SUCTION LEG SIAT BREAK - MAXIMUM SIS FIDW 2 (9.8175 ft Total Area) PART B: Reactor Vessel Pressure vs. Time Reactor Vessel Time (sec) Pressure (psia) 24.91 74.10 25.91 100.32 26.91 129.88 27.91 131.11 28.91 130.47 30.91 129.12 32.91 127.77 34.91 126.45 36.91 125.15 38.91 123.87 - 40.91 122.60 42.91 121.36 44.91 120.14 46.91 118.94 48.91 117.75 50.91 116.64 52.91 115.49 54.91 114.35 G 56.91 113.24 58.91 112.14 60.91 111.05 , 65.91 108.41 I 70.91 105.88 75.91 103.41 80.91 101.07 86.91 98.44 88.91 97.07 99.91 91.62 114.71 95.69 134.71 59.02 O Amendment G April 30, 1990 t
CESSAR tim?eue (v'~) s TABLE 6.2.1-4 (Cont'd) (Sheet 9 of 9) DOUBLE-ENDED SUCTION LEG SIOT BREAK - MAXIMUM SIS FLOW 2 (9.8175 ft Total Area)
- PART C: Chronology of Events t
Time (seconds) Event 0.00 Break Occurs 14.87 Start Safety Injection Tank Injection 22.68 Peak Containment Pressure (Blowdown) 24.80 Start SIS Injection Phase 24.80 End of Blowdown () 26.91 Downcomer Full 79.90 Start Spray Injection 89.90 Peak Containment Pressure Subsequent To end of Blowdown 0 114.71 End of Core Reflood 129.01 Safety Injection Tank Empty
-131.51 End of Post Reflood Amendment G April 30, 1990
@ESSAR Mnm I
($3) : TABLE 6.2.1-5 t (Sheet 1 of 9) , DOUBLE-ENDED SUCTION LEG SI.OT BREAK - MINIMUN SIS FLOW (9.8175 ft 2 Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1ba/sec) (Btu /sec) l 0.000 0.00000E+00 0.00000E+00. ) 0.028 7.55833E+04 4.17879E+07 ) 0.053 7.47554E+04 4.12574E+07 ' 0.102 7.54066E+04 4.15729E+07 O.150 7.74168E+04 4.27427E+07 , 0.202 7.77315E+04 4.29866E+07 0.253 9.03920E+04 5.00891E+07 0.300 9.01567E+04 5.00007E+07 i' 0.352 8.78716E+04 4.88015E+07 I O.501 8.72964E+04 4.86059E+07 0.651 7.91404E+04 4.41690E+07-0.801 9.19291E+04 5.14526E+07 0.951 9.09209E+04 5.10498E+07 ' 1.202 8.54119E+04 4.82250E+07 1.400 8.11399E+04 4.60272E+07 1.602 8.08481E+04 4.60648E+07 1.800 7.93604E+04 4.53922E+07 G 2.000 7.74451E+04 4.44592E+07 3.004 6.48128E+04 3.77221E+07 4.015 5.44915E+04 3.25994E+07 5.015 4.07658E+04 2.63472E+07 6.015 3.57029E+04 2.35677E+07 7.015 3.38417E+04 2.23953E+07 8.015 3.17027E+04 2.11818E+07 9.015 2.95990E+04 2.00184E+07 10.015 2.69626E+04 1.86044E+07 11.015 2.39726E+04 1.70944E+07 12.015 2.00996E+04 1.51119E+07 13.015 1.59931E+04 1.29885E+07 14.015 1.35005E+04 1.14895E+07 15.015 1.12545E+04 9.86448E+06 16.006 8.02625E+03 6.85176E+06 16.200 7.37928E+03 6.64032E+06 O 16.400 16.600 6.69674E+03 6.14117".+03 6.08618E+06 5.60001E+06 Amendment G - April 30, 1990
CESSAR WW.co... O TABLE 6.2.1-5 (Cont'd) (Sheet 2 of 9) DOUBLE-ENDED SUCTION LEG SIDT BREAK - MINIMUM SIS FIDW (9.8175 ft2Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lbe/sec) (Btu /sec) 16.8 5.66563E+03 5.17392E+06 17.0 5.21836E+03 4.77524E+06 17.2 4.82674E+03 4.42423E+06 17.4 4.51073E+03 4.12148E+06 17.6 4.21515E+03 3.83064E+06 17.8 3.94891E+03 3.56781E+06-18.0 3.72144E+03 3.35152EA06 1 18.2 3.53694E+03 3.18325E3.6 > 18.4 3.37361E+03 3.03081E+06-10.6 3.20218E+03 2.87402E+06 - 18.8 3.11835E+03 2.79793E+06 19.0 4.02552E+03 3.32500E+06 19.2 4.60621E+03 3.76169E+06 19.4 5.14842E+03 4.20051E+06 19.6 5.59701E+03 4.50276E+06 19.8 6.06622E+03 4.70455E+06 20.0 6.36098E+03 4.83597E+06 20.2 6.70677E+03- 4.92749E+06 G l 20.4 6.81109E+03 4.94384E+06 20.6 6.99119E+03 4.96322E+06 c 20.8 7.0PJ04E+03 4.92808E+06 21.0 7.02967E+03 4.87504E+06 21.2 7.02094E+03 4.81564E+06 21.4 6.93118E+03 4.72114E+06 21.6 6.92645E+03 4.61569E+06 21.8 6.71463E+03 4.46738E+06 22.0 6.64975E+03 4.36366E+06 22.2 6.58945E+03 4.23874E+06 22.4 6.30562E+03 4.07136E+06 22.6 6.14675E+03 3.92726E+06 22.8 5.81248E+03 3.74008E+06 23.0 5.54480E+03 3.57513E+06 23.2 6.44593E+03 3.88707E+06 23.4 6.90779E+03 4.08909E+06 23.6 7.44546E+03 4.26589E+06 . 23.8 8.03797E+03 4.46482E+06 Amendment G April 30, 1990
CESSAR WW.co.. lO TABLE 6.2.1-5 (Cont'd) (Sheet 3 of 9) DOUBLE-ENDED SUCTION LEG SIDT BPJtAK - MININUM SIS FIAMf 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break-Energy Time Flow Rate Flow Rate (sec) (1bs/sec) (Btu /sec) 24.0 8.48076E+03 4.55964E+06 24.2 9.05228E+03 4.69236E+06 24.4 9.72404E+03 4.84813E+06 24.6 1.05776E+04 5.03851E+06 24.8 1.04542E+04 4.90047E+06 25.41 9.15640E+02 1.15213E+06 26.51 1.33977E+03 1.68096E+06 27.51 1.59603E+03 1.99969E+06 O 27.91 27.92 1.59585E+03 9.25593E+02 1.99776E+06 1.15870E+06 l 28.61 9.23992E+02 1.15549E+06 l 29.71 9.20240E+02 1.14956E+06 ! 30.71 9.16789E+02 1.14435E+06 31.81 9.12468E+02 1.13804E+06 32.91 9.08123E+02 1.13173E+06 33.91 9.04185E+02 1.12602E+06 35.01 8.99899E+02 1.11978E+06 36.11 8.95671E+02 1.11360E+06 g 37.11 8.91860E+02 1.10803E+06 38.21 8.87690E+02 1.10192E+06 39.31 8.83589E+02 1.09592E+06 40.31 8.79831E+02 1.09044E+06 41.41 8.75719E+02 1.08445E+06 42.51 8.71624E+02 1.07849E+06 43.51 8.67918E+02 1.07310E+06 44.61 8.63869E+02 1.06722E+06 45.71 8.59827E+02 1.06136E+06 46.71 8.56167E+02 1.05606E+06 47.81 8.52159E+02 1.05026E+06 48.91 8.48163E+02 1.04450E+06 49.91 8.44550E+02 1.03929E+06 51.01 8.40594E+02 1.03359E+06 52.11 8.36650E+02 1.02792E+06 O 53.11 54.21 55.31 8.33118E+02 8.29278E+02 8.25456E+02 1.02281E+06 1.01722E+06 1.01167E+06 Amendment G l April 30, 1990 l l 1
CESSARHMieu. 9 , TABLE 6.2.1-5 (Cont'd) (Sheet 4 of 9) DOUBLE-ENDED SUCTION LBG SIDT BREAK - MININUM SIS FIAN (9.8175 ft 2 Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lba/sec) (Btu /sec) i 56.31 8.21993E+02 1.00665E+06 57.41 8.18206E+02 1.00116E+06 58.51 8.14430E+02 9.95703E+05 59.51 8.11014E+02 9.90762E+05 60.61 8.07267E+02 9.85356E+05 l 61.71 8.03532E+02 9.79974E+05 i 62.71 8.00151E+02 9.75108E+05 l 63.81 7.96450E+02 9.69783E+05 64.91 7.92756E+02 9.64488E+05 l 65.91 7.89415E+02 9.59691E+05 { 67.01 7.85749E+02 9.54442E+05 68.11 7.82C '+ 9 2 9.49222E+05 69.11 7.78763
- 9.44501E+05 70.21 7.75170Ev 9.39333E+05 71.31 7.71603E+0/ 9.34218E+05 72.31 7.68007E+02 9.29143E+05 G 73.41 7.64515E+02 9.24126E+05 74.51 7.61047E+02 9.19132E+05 75.51 7.57909E+02 9.14614E+05 76.61 7.54470E+02 9.09672E+05 77.71 7.51042E+02 9.04759E+05 78.71 7.47875E+02 9.00578E+05 79.81 7.44465E+02 8.95688E+05 80.91 7.41025E+02 8.91118E+05 81.91 7.37957E+02 8.86716E+05 i 83.01 7.34605E+02 8.81896E+05 l 84.11 7.31229E+02 8.77064E+05 I l
85.11 7.03378E+02 8.29783E+05 86.21 1.13940E+03 1.34433E+06 87.31 1.08217E+03 1.27691E+06 88.31 1.03451E+03 1.22072E+06 l 89.41 9.85829E+02 1.16330E+06 l 90.51 9.42589E+02 1.11226E+06 l l 91.51 9.02489E+02 1.06492E+06 l 92.61 8.62300E+02 1.01747E+06 I 93.71 8.24960E+02 9.73360E+05 ' Amendment G April 30, 1990
C E S S A R !!I Wico... ; O TABLE 6.2.1-5 (Cont'd) (Sheet 5 of 9) DOUBLE-IDIDED SUCTION LEG SIOT BREAK - MININUM SIS FIAMI (9.8175 ft2 Total Area) PART At Mass /2n'argy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1be/sec) (Btu /sec) 94.71 7.93429E+02 9.36100E+05 95.81 7.61120E+02 8.97910E+05 96.91 7.31039E+02 8.62339E+05 97.91 7.05420E+02 8.32020E+05 99.01 6.78929E+02 8.00660E+05 102.41 6.06130E+02 7.14469E+05 l 107.41 5.66202E+02 6.66769E+05 ! 112.06 5.19020E+02 6.10721E+05 l 116.71 4.19474E+02 4.93551E+05 121.71 3.20914E+02 3.77557E+05 i 126.71 1.65732E+02 1.94972E+05 131.71 2.06498E+02 2.44767E+05 136.71 2.46834E+02 2.91180E+05 141.71 2.02410E+02 2.38074E+05 j 2.02275E+05 146.71 1.71992S+02 151.71 1.49148E+02 1.75393E+05 g 156.71 1.31396E+02 1.54508E+05 161.71 1.18124E+02 1.38889E+05 166.71- 1.08516E+02 1.27584E+05 : l
.)
l Amendment G l April 30, 1990 l
CESSAR ifW,co ! l l l 9 TABLE 6.2.1-5 (Cont'd) ] (Sheet 6 of 9) DOUBLE-ENDED SUCTION LEG SI&f BREAK - MINIMUM SIS FIDW SPILLAGE DATA l Time, Sec Mass, llan Energy, 10 6 Btu End of Blowdown 24.80 0.0 0.0 End of Reflood'114.21 196,276 64.111 End of Post Reflood 166.71 259,582 80.451 l l l l O Amendment G April 30, 1990
)! h h k!I !bb!flCATION I
l i l-TABLE 6.2.1-5 (Cont'd) (Sheet 7 of 9) j l DOUBLE-ENDED SUCTION LEG SIDT BREAK - MININUM SIS FIDW l 2 (9.8175 ft Total Area) 1 PART B:
~
Reactor Vessel Pressure vs. Time Reactor Vessel l Tlee (sec) Pressure (psia) ; 1 0.000 2275 0.027 2258 ' O.052 2045 0.102 1701 0.149 1696' O.202 1689 0.252 1687 0.351 1672 0.500 1643 i , g 0.650 1628 0.801 1620 0.951 1606 ^ 1.201 1576 1.400 1556 2.000 1482 3.000 1380 4.012 1298 G 5.012 1205 0.012 1140 7.012 1098 ! 8.012 1062 9.012 1025 10.012 987 11.012 942 12.012 874 14.012 731 ' 16.012 462 1 18.000 234 20.000 304 22.000 258 22.400 237 l Amendment G L April 30, 1990 l L -
1 CESSAR!HW.co TABLE 6.2.1-5 (Cont'd) j (Sheet 8 of 9) 1 DOUBLE-ENDED SUCTION LEG SIDT BREAK - MINIMUN SIS FIANI 2 (9.8175 ft Total Area) PART B: Reactor Vessel Pressure vs. Time l Reactor Vessel i Time (sec) Pressure (psia) ) I 24.91 77.54 25.91 101.14-
- 26.91 126.80 27.91 133.57 28.91 132.99 30.91 131.83 32.91 130.61 34.91 129.41 36.91 128.23' 38.91 127.07 40.91 125.93 42.91 124.81 '
44.91 123.71 46.91 122.62 48.91 121.55 i 50.91 120.50 , 52.91 119.47 54.91 118.46 G 7 56.91 117.46 , 58'.91 116.47 60.91 115.51 65.91 113.15 70.91 110.90 75.91 108.71 80.91 106.71 86.91 103.78 88.91 102.48 99.91 93.31 114.21 85.70 134.21 59.07 154.21 56.73 l Amendment G April 30, 1990 l I J
CESSAR!!nnui I l 1 l TABLE 6.2.1-5 (Cont'd) . (Sheet 9 of 9) ; DOUBLE-BIDED SUCTION LEG SIDT BREAK - MINIMUBE SIS FIDit 2 (9.8175 ft Total Area) PART C: Chronology of Events Time (seconds) _ Event , 0.0 Break Occurs 14.87 Start Safety Injection Tank Injection 22.68 Peak Containment Pressure (Blowdown) 24.80 Start SIS Injection Phase 24.80 End of Blowdown () 26.91 Downcomer Full 79.90 Start Spray Injection 89.90 Peak Containment Pressure Subsequent t To end of Blowdown G ) 114.21 End of Core Reflood j 129.21 Safety Injection Tank Empty 166.71 End of Post Reflood i l l l
)
i Amendment G l April 30, 1990 l
I
!hhk !! r$FICATCN i
i I gx 1 TABLE 6.2.1-6 l (Sheet 1 of 14) DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MAXIMUM SIS FIDW (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lba/sec) (Btu /sec) 0.000 0.00000E+00 0.00000E+00 0.028 7.54399E+04 4.16739E+07 0.053 7.49763E+04 4.13519E+07 0.101 7.65384E+04 4.21770E+07 0.155 1.03933E+05 5.73492E+07 0.202 1.03370E+05 5.70807E+07 0.257 1.02327E+05 5.654CJE+07 0.308 1.02142E+05 5.64862E+07 0' 0.351 O.504 0.662 1.01734E+05 9.96669E+04 9.78153E+04 5.62937E+07 5.52347E+07 5.42757E+07 0.812 9.61317E+04 5.33980E+07 j 0.952 9.42525E+04 5.24081E+07 1.212 9.16780E+04 5.10959E+C7 G 1.412 9.03275E+04 5.04767E+07 1.612 8.80764E+04 4.93908E+07 1.812 8.53110E+04 4.80292E+07 i 2.012 8.19683E+04 4.63246E+07 l 3.012
~
4783E+04 4.13978E+07 1 4.012 6.21422E+04 3.56171E+07 5.012 5.82256E+04 3.34663E+07 6.006 5.25307E+04 3.07886E+07 7.012 4.11874E+04 2.60995E+07 8.004 2.86077E+04 2.13623E+07 9.001 2.34231E+04 1.89132E+07 10.00 2.07247E+04 1.73225E+07 11.00 1.63386E+04 1.50402E+07 12.00 1.10760E+04 1.18028E+07 13.00 1.30962E+04 9.43083E+06 14.00 1.35235E+04 7.75199E+06 15.00 1.18662E+04 6.04711E+06 16.00 8.35923E+03 3.98625E+06 16.20 7.61847E+03 3.60785E+06 16.40 7.09407E+03 3.31586E+06 16.60 6.48957E+03 3.00631E+06 Amendment G April 30, 1990
. _- _ _ A
CESSAR timne.- O TABLE 6.2.1-6 (Cont'd) (Sheet 2 of 14) { DOUBLE-ENDED DISCHARGE L8G SIDT BREAK - MAXIMUN SIS FLOW (9.8175 ft Total Area) PART A: Mass / Energy Release Data ; l Break Mass Break Energy Time Flow Rate Flow Rate !< (sec) (1be/sec) (Btu /sec) 16.80 3.37722E+03 1.55052E+06 l 17.00 3.35939E+03 1.53579E+06 j 17.20 3.34360E+03 1.52408E+06 i 17.40 3.29544E+03 1.50291E+06 l 1 17.60 3.21787E+03 1.47673E+06 17.80 2.85056E+03 1.31226E+06 ; 18.00 1.54325E+03 1.17246E+06 > 18.20 2.24486E+03 1.03981E+06 18.40 1.94972E+03 9.11453E+05 ; 18.60 1.65285E'03 7.85437E+05 ! 18.80 1.34658E+03 6.58857E+05 18.81 0.00000E+00 0.00000E+00 1 18.91 1.54910E+02 2.00460E+05 19.61 3.14500E+02 4.09020E+05 20.21 3.61050E+02 4.69770E+05 g , 20.71 4.92570E+02 6.41270E+05 1 21.31 6.25290E+02 8.13960E+05 21.91 7.42770E+02 9.66380E+05 22.41 8.33360E+02 1.08380E+06 3 22.71 8.42330E+02 1.09522E+06 22.72 4.88551E+02 6.35228E+05 23.01 4.88389E+02 6.34955E+05 23.61 4.88029E+02 6.34334E+05 24.11 4.87728E+02 6.33888E+05 24.71 4.87299E+02 6.33250E+05 25.31 4.86823E+n2 6.32565E+05 25.81 4.86481E+02 6.32171E+05 ; 26.41 4.85970E+02 6,31452E+05 27.01 4.85454E+02 6.30727E+05 27.51 4.85019E+02 6.30118E+05 28.11 4.84503E+02 6.29387E+05 28.71 4.83993E+02 6.28662E+05 29.21 4.83610E+02 6.28198E+05 29.81 4.83105E+02 6.27485E+05 30.41 4.82566E+02 6.26713E+05 Amendment G Anril 30, 1990
,j l
CESSARHELm-O ; TABLE 6.2.1-6 (Cont'd) . (Sheet 3 of 14) DOUBLE-ENDED DISCHARGE LEG SLCT BREAK - MAXIMUM SIS FLOW 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1ba/pec) (Stu/sec) l 30.91 4.82142E+02 6.26122E+05 m 31.51 4.81638E+02 6.25484E+05 32.11 4.81185E+02 6.24828E+05 32.61 4.80768E+02 6.24237E+05 33.'21 4.80263E+02 6.23523E+05 33.81 4.79764E+02 6.22816E+05 34.31 4.79347E+02 6.22224E+05 34.91 4.78848E+02 6.21511E+05 4.78343E+02 6.20803E+05
' O: 35.51-36.01 4.77926E+02 6.20211E+05
'36.61 4.77427E+02 6.19498E+05 37.21 4.7692;E+02 6.18790E+05 .
37.71 4.76S05E+02- 6.18199E+05 38.31 4.76006E+02 6.17485E+05 38.91 4.75501E+02 6.16778E+05-39.41 4.75084E+02- 6.16186E+05 4.74579E+02 6.15479E+05 40.01 4.74080E+02 6.14765E+05 40.61 G 41,11 4.73657E+02 6.14174E+05 41.71 4.73158E+02 6.13466E+05 & ~ C. 31 4.72654E+02 6.12753E+05 42.81 4.72236E+02 6.12161E+05 13.41 4.71731E+02 6.11453E+05 44.01 4.71227E+02 6.10740E+05 44.51 4.70809E+02 6.10148E+05 45.11 4.70328E+02 6.09470E+05
-45.71- 4.69881E+02 C.08838E+05 =
46.21 4.69516E+02 6.08310E+05 46.81- 4.69069E+02 6.07678E+05 47'41.
. 4.68628E+02 6.07045E+05 47.91 4.68257E+02 6.06518E+05 48.51- 4.67811E+02 6.05885E+05 _
49.01 4.67445E+02 6.05358E+05 , O 49.61 50.21 4.66999E+02 4.66552E+02 6.04725E+05 6.04093E+05 Amendment G April 30, 1990
1
; CESSAR Mut?, cue.
Oi TABLE 6.2.1-6 (Cont'd) l (3heet 4 of 14) DOUBLE-ENDED DISCHARGE LEG SIAT BREAK - MAXIMUM SIS FIAMi l 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data
'l Break Mass Break Energy Time Flow Rate Flow Rate (sec) ____
(lba/sec) (Btu /sec) 50.71 4.66181E+02 6.03565E+05 i 51.31 4.65740E+02 6.02927E+05-51.91 4.65293E+02 6.02295E+05 52.41 4.64922E+02 6.01767E+05 53.01 4.64476E+02 6.01135E+05 53.61 4.64029E+02 6.00503E+C5 54.11 4.63658E+02 5.99975E+05 54.71 4.63211E+02 5.99337E+05 55.31 4.62765E+02 5.98705E+05 h 55.81 4.62393E+02 5.98177E+05 56.41 4.61947E+02 5.97545E+05 57.01 4.61500E+02 5.96913E+05 57.51 4.61135E+02 5.96385E+05 58.11 4.60694E+02 5.95753E+05 58.71 4.60247E+02 5.95126E+05 G 59.21 4.59882E+02 5.94599E+05
'59.81 4.59441E+02 5.93972E+05
- i. 60.41 4.59000E+02 5.93346E+05 60 S1 4.58629E+02 -
5.92818E+05 61.51 4.58188E+02 5.92192E+05 W 62.11 4.57748E+02 5.91559E+05 l' 62.61 4.57376E+02 5.91037E+05 63.21 4.56936E+02 5.90405E+05 63.81 4.56489E+02 5.89773E+05 64.31 4.56124E+02 5.89251E+05 64.91 4.55677E+02 5.88619E+05 65.51 4.55236E+02 5.87987E+05 66.01 4.54865E+02 5.87465E+05 66.61 4.54418E+02 5.86832E+05 67.21 4.53972E+02 5.86200E+05 67.71 4.53606E+02 5.85678E+05 68.31 4.53160E+02 5.85046E+05 l 68.91 4.52713E+02 5.84414E+05 69.41 4.52342E+02 5.83886E+05 70.01 4.51895E+02 5.83260E+05 l Amendment G April 30, 1990
l l h hh k! b kflCAfl201-l l j' l
\_) l TABLE 6.2.1-6 (Cont'd)
(Sheet 5 of 14) :l DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MAXIMUN SIS FIDW (9.8175 ft Total Area) l PART A: Mass / Energy Release Data
'1 Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lbm/sec) (Btu /sec) 70.61 4.51449E+02 5.82627E+05 !
71.11 4.51078E+02 5.82100E+05 71'.71 4.50631E+02 5.81467E+05
'72.31 4.50184E+02 5.80835E+05 72.81 4.49813E+02 5.80307E+05 73.41 4.49367E+02 5.79675E+05 74.01 4.48920E+02 5.79043E+05 J
74.51 4.48549E+02 5.78515E+05 e 75.11 4.48096E+02 5.77883E+05
'" 75.60 4.47725E+02 5.77355E+05 75.61 7.71940E+02 9.95440E+05 79.61 7.45540E+02 9.61080E+05 83.51 7.21020E+02 9.29210E+05 ' '
87.41' 6.98280E+02 8.99660E+05 ' 91.31- 6.76770E+02 8.71740E+05 95.21 6.56150E+02 8.44970E+05 G-99.11 6.37300E+02 8.20520E+05 ' 10.30 6.20120E+02 7.98210E+05 10,69 6.04490E+02 7.77900E+05 11.08 5.90270E+02 7.59420E+05 t 11.47 5.77360E+02 7.42620E+05 11.86 5.65650E+02 7.27370E+05 12.25 5.55040E+02 7.13540E+05 12.64 5.45460E+02 7.01020E+05 13.03 5.36800E+02 6.89700E+05 13.42 5.29000E+02 6.79490E+05- ~s 13.81 5.21990E+02 6.70280E+05-14.20 5.15690E+02 6.61990E+05 14.59 5.10040E+02 6.54530E+05 14.98 5,04980E+02 6.47830E+05 15.37 5.00460E+02 6.41820E+05 15.76 4.96420E+02 6.36430E+05 l 16.15 4.92840E+02 6.31620E+05 16.54 4.89650E+02 6.27310E+05 O\ 16.93 4.86810E+02 6.23450E+05 , 1 Amendment G April 30, 1990
t CESSAR E ben t l TABLE 6.2.1-6 (Cont'd) (Sheet 6 of 14) _ DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MAXIMUM SIS FLOW 2 (9.8175 ft Total Area) I PART A: Mass / Energy Release Data l Break Mass Break Energy l Time Flow Rate- Flow Rate .! (sec) (lba/sec) (Btu /sec) 173.21' 4.84310E+02 6.20010E+05 , 177.21 4.82030E+02 6.16860E+05 181.11 4.80070E+02 6.14110E+05 : 185.01' 4.78340E+02 6.11650E+05 6.09450E+05 l 188.91 4.76800E+02 .; ' 192.81 4.75450E+02 6.07470E+05 196.71 4.74250E+02 6.05690E+05-200.61 4.73200E+02 6.04090E+05 ' 204.51 4.72270E+02 6.02650E+05 - 208.41 4.71460E+02 6.01350E+05 ' 212.31 4.70740E+02 6.00180E+05 216.21 4.70110E+02 5.99110E+05 220.11 4.69560E+02 5.98150E+05
.224.01- 4.69070E+02. 5.97260E+05 l
227.91 4.68650E+02 5.96460E+05 231.81 '4.68280E+02 5.95710E+05 235.71 4.67960E+02- 5.95030E+05 G 239.61 4.67680E+02 5.94400E+05 243.51 4.67430E+02 5.93810E+05 247.41 4.67230E+02 5.93270E+05 251.31 4.67060E+02 5.92770E+05 255.21 4.66920E+02 5.92290E+05 259.11 4.66790E+02 5.91840E+05 263.01 4.66690E+02 5.91420E+05 263.41 4.66550E+02 5.91260E+05 266.91 4.66460E+02 5.90880E+05 270.81 4.66410E+02 5.90520E+05 270.91 4.73590E+02 5.89995E+05' r -271.01 4.56650E+02 5.90552E+05 271.11 4.83350E+02 5.90599E+05 271.21- 4.622952+02 6.09401E+05 271. 1 4.77705E+02 5.82139E+05 271.71 4.66418E+02 6.11194E+0S 272.01 4.73582E+02 5.89820E+05 272.41 4.72774E+02 6.10180E+05 Amendment G April 30, 1990
t CESSARli!Rncuc, o l ,' c? b TABLE 6.2.1-6 (Cont'd) l (Sheet 7 of 14) DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MAXIMUM SIS FIDW 1 2 (9.8175 ft Total Area) i l PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1be/sec) (Btu /sec) 272.81 4.82226E+02 5.96155E+05-273.31 4.74207E+02 6.11845E+05 273.91 4.89126E+02 6.06996E+05 274i51 4.81000E+02 6.16338E+05 275.21 4.90428E+02 6.14482E+05- ' 275.91 4.90286E+02 6.25518E+05 - 7 276.71 4.94714E+02 6.22351E+05 { 277.51- 4.98008E+02 6.32649E+05 . L '( 278.41 5.04214E+02 6.33093E+05 279.31 5.05342E+02 6.42463E+05 280.31 5.10658E+02 6.42844E+05-281.41 5.14797E+02 6.49884E+05 282.51 5.19749E+02 6.55571E+05 283.61 5.22640E+02 6.60793E+05 284.91 5.29668E+02 6.65596E+05 , 286.11- 5.31366E+02 6.71076E+05 G 287.51 5.38634E+02 6.77496E+05 288.81 5.41553E+02 6.82504E+05 ' 290.31 5.46563E+02 6.88201E+05 291.81- 5.52103E+02 6.94465E+05 293.31 5.55978E+02 6.99164E+05 294.91 5.61522E+02 7.05836E+05 ' 296.61 5.65614E+02 7.10767E+05 298.31 5.73210E+02 6.85703E+05 300.11 5.55583E+02 6.56263E+05 301.91 5.24278E+02 6.15503E+05 303.81 4.98464E+02 5.88708E+05 305.71 4.76165E+02 5.59659E+05 307.71 4.57835E+02 5.39341E+05 309.81 4.39173E+02 5.16959E+05 311.91 4.24708E+02 5.00184E+05 314.01 4.10434E+02 4.83461E+05 316.21 3.98657E+02 4.69266E+05 318.51 3.86050E+02 4.54586E+05 320.81 3.75015E+02 4.41418E+05 Amendment G April 30, 1990
- CESSAR ;!!alen.
O TABLE 6.2.1-6 (Cont'd) (Sheet 8 of 14) DOUBLE-ENDED DISCHARGE LBG SIET BREAK - MAXIMUM SIS FIDW 2 (9.8175 ft Total Area)' _ l PART A: Mass / Energy Release Data J l Break Mass r'. Energy Time Flow Rate Pate (sec) (lbe/sec} -c) 323.21- 3.64236E+02 4.2'<76 F05 325.71 3.53542E+02 4.18 +05 328.11 3.44064E+02 4.0i.ia0E+0F 330.71- 3.33914E+02 3.93832E+05 d 333.31 3.24669E+02 3.81996E+05 336.01 3.15471E+02 3.71483E+05. 338.71 3.06861E+02 3.61285E+05 341.41 2.99065E+02 3.51594E+05 344.31 2.90250E+02 3.42199E+05 347.11 2.82858E+02 3.33515E+05 - 350.11 2.75808E+02 3.24164E+05 353.11 2.68336E+02 3.16003E+05 356.11 2.63957E+02 3.11330E+05 1 359.21 2.60559E+02 3. 063 24 E+05 362.41 2.55393E+02 3.00935E+05 365.61 2.51571E+02 2.96565E+05 G 368.81 2.47817E+02 2.91383E+05-372.21 2.43964E+02 2.88029E+05
.375.51 2.40369E+02 2.82707E+05 379.01 2.36857E+02 2.79007E+05 382.51 2.33519E+02 2.74888E+05 386.01' 2.30481E+02 2.71397E+05 l 389.61 2.27364E+02 2.68047E+05
- 393.31 2.24528E+02 2.64067E+05 l
397.01 2.21958E+02 2.60899E+05 400.71. 2.18811E+02 2.58020E+05 ,., 404.51 2.16461E+02 2.54731E+05 408.41 2.13878E+02 2.51936E+05-412.41 2.11622E+02 2.49064E+05 416.31 2.09055E+02 2.46054E+05 420.41 2.07042E+02 2.43702E+05 424.51 2.04733E+02 2.41176E+05 428.61 2.02667E+02 2.38503E+05 432.81 2.01142E+02 2.36735E+05 437.11 1.98349E+02 2.33734E+05 Amendment G April 30, 1990
CESSAR E!Miem:n 1
.)
l TABLE 6.2.1-6 (Cont'd)- ! (Sheet 9 of 14) l DOUBLE-ENDED DISCHARGE LEG SIAT BREAK - MAXIMUM SIS FIDW l (9.8175 ft Total Area). l PART 7.: Mass / Energy Release Data I l Break Mass Break Energy i Time Flow Rate Flow Rate ! (sec) (lbs/sec) (Btu /sec) 441.41 1.97465E+02 2.31848E+05 1 445.81 1.94691E+02 2.29306E+05 ) 450.21 1. 9 3 4.' 0 E+ 0 2 2'27967E+05
. l 454.71 1. 914 4 t'E+ 02 2.24934E+05 459.21 1.89615E+v2 2.23955E+05 l 463.81 1.88211E+02 2.21119E+05. . 468.51 1.86682E+02 2.'19732E+05 e 473.21_ 21.84641E+02 2.17449E+05 g ;
478.01 1.83693E+02 2.15884E+05 L( 482.81 .1.81564E+02 2.13835E+05 L 487.61 1.80520E+02 2.12415E+05 492.61 1.78770E+02 2.10427E+05 l 497.61 1.77630E+02 2.08773E+05 502.61 1.75785E+02 2.06964E+05 507.71 1.74803E+02 2.05585E+05~ i 512.81 1.73140E+02 2.03833E+05 .l 518.01 1.72245E+02 2.02321E+05 . 523.31 1.70280E+02 2.00480E+05 l 528.61 1.69343E+02 1.99142E+05 534.01 1.62879E+02- 1.91599E+05 539.41 1.33266E+02 1.56782E+05 544.91 1.29279E+02 1.52309E+05 I 550.41 1.15660E+02 1.35975E+05-l L o Amendment G April 30, 1990
CESSAR ?!inneur.. 3: O' TABLE 6.2.1-6 (Cont'd) (Sheet 10 of 14)
~
i o DOUBLE-ENDED DISCHARGE LEG SIDT h6. - MAXIMUN - SIS FIANI i SPILLhGE DATA . t i Time, Sec Mass, lha Energy,~10 Btu G End of. Blowdown 18.81 0.0 0.0 End-of Reflood.270.71 332,902 50.960 End'of Post Reflood 550.41 455,561' 82.891 i
'l
{ 1 l 1 O Amendment G April 30, 1990
lCESSARiin%=. ' J f'p , L /_ TABLE 6.2.1-6 (Cont'd) ') (Sheet 11 of:14) DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MAXIMUM SIS - FIDW (9'.8175.ft Total Area)- PART-8: . Reactor' Vessel' Pressure vs. Time l Reactor Vessel f Time (sec) ^ Pressure (psia) i 1 0.000 2275 0.028 2194
.0.053 1896' O.101 1699 1 l
0.155- 1686 0.202' 1681 1 0.257 1670. 0.308. 1650 i 0.504 1617 g O -O.662 1604
')
0.812 1582 i , 0.952~ 1558: . l'E 1.212 1526 -( 1.412 1517. .)
.J 2 . 012 1431 l 3.012 1268
'4.012- 1198 5.012 1161 6.006 1131 1 7.012 1097 I 8.004 1057 -l 9.000 995 10.000 933 11.000 835 1 12.000 686 14.000 377 16.000 192 ,
18'000
. 87 j j
O i lunendment G April 30, 1990 ; i
CESSAR8!NLmw -
-.~
TABLE 6.2.1-6 (Cont'd) (Sheet 12 of 14) DOUBLE-ENDED DISCHARGE LEG SIAT BREAK - MAXIMUM SIS FLOW 2 (9.8175 ft Total Area) PART B:' Reactor Vessel Pressure vs.' Time Reactor Vessel' Time (sec) Annulus Pressure (psia) 19.71- 57.944 20.71 62.385' 21.71' 69.833 22.71 75.374 23.71 75.290 24.71 75.215
.6.71 2 75.062 28.71 74.896 30.71 74.736 32.711 74.580 34.71- 74.471 36.71 74.255 38.71 74.'093 40.71- 73.931 42.71 73.770 44.71 73.609 G 46.71 73.506 48.71 73.405 50.71 73.305 52.71 73.206 54.71 73.107 56.71 73.008 58.71: 72.910 60.71 72.810 70.71 72.333 80.71 70.821 90.71 68.722 99.71 67.214 119.91 65.082 139.91 64.058 159.91 63.643 179.91 63.566 199.91 63.669 219.91 63.868 310.81 66.852 410.81 61.759 450.81 60.684 Amendment G April 30, 1990
- - - .-. _ _ . -. .- .. . . . . _ . . _ . . .~ -
t >
! !h!h k!I b$r$flCATl!Di' I
TABLE 6.2.1-6 (Cont'd) 1 (Sheet 13.of 14) DOUBLE-ENDED DISCHARGE ' LEG SIDT BREAK - MAXIMUM SIS FIDW l 2 (9.8175 ft Total Area) J PART B: Reactor Vessel Pressure vs. Time Reactor Vessel . Time (sec) Annulus Pressure (psia) . 470.81 .60.193: 490.81 59.720 510.81 59.'260-530.81- .58.881 550.81 58.112 -t
, i, a
l l 1 l 1 l Amendment G i April 30, 1990 i
;CESSAR ninne.noo I
el e TABLE 6.2.1-6 (Cont'd) l (Sheet.14 of 14) DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MAXIMUM SIS FIDW (9.8175 ft Total Area) i I
.PART C: Chronology of Events ;
3
. Time (seconds) Event 0.0 Break Occurs .
l t 12.15 Start Safety Injection Tank Injection 1 16.38 Peak Containment Pressure-(Blowdown) G 18.80 Start SIS Injection Phase ; l'
.18.80 End of Blowdown 22.71 Downcomer Full 69.90 Start Spray Injection ~
75.61 Safety Injection Tank Empty 270.71 End of Core Reflood f 323.80 Peak Containment Pressure Subsequent To end of Blowdown l.' 550.41 End of Post Reflood i l-l e Amendment G l April 30, 1990 1
CESSAR !!nhiz TABLE 6.2.1-7 (Sheet 1 of 14) DOUBLE-ENDED DISCHARGE LEG SLOT BREAK - MINIMUN SIS FIDW (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass' Break Energy Time Flow Rate Flow Rate (sec) (1ba/sec) (Btu /sec) 0.000 0.00000E+00- 0.00000E+00 0.028 7.54399E+04 4.16739E+07 0.053 7.49763E+04 4.13519E+07 0.101 7.65384E+04 4.21770E+07 0.155- 1.03933E+05 5.73492E+07 0.202 1.03370E+05 5.70807E+07 0.257 1.02327E+05 5.65480E+07 0.308 1.02142E+05 5.64862E+07 9 0.351 0.504 1.01734E+05 9.96669E+04 5.62937E+07 5.52347E+07-0.662 9.78153E+04 5.42757E+07 0.812 9.61317E+04 5.33980E+07 0.952 9.42525E+04 5.24081E+07 1.212 9.16780E+04 5.10959E+07 G 1.412 9.03275E+04 5.04767E+07 1.612 8.80764E+04 4.93908E+07 1.812 8.53110E+04 4.80292E+07' 2.012 8.19683E+04 4.63246E+07 3.012 7.24783E+04 4.13978E+07 4.012 6.21422E+04 3.56171E+07 5.011 5.82256E+04 3.34663E+07 6.005 5.25307E+04 3.07886E+37 7.012 4.11874E+04 2.60995E+D7 8.004 2.86077E+04 2.13623E+07 9.001 2.34231E+04 1. 89132E4 07 10.00 2.07247E+04- 1.73225Er07 11.00 1.63386E+04 1.50402T.+07 12.00 1.10760E+04 1.18 02 F.E+07 13.00 1.30962E+04 9.43083E+06 14.00 1.35235Ft04 7.75199E+06 15.00 1.18662E+04 6.04711E+06 16.00 8.35923E+03 3.98625E+0C 16.20 7.61847E+03 3.60785E+06 9- 16.40 16.60 7.09407E+03 6.48957E+03 3.31586E+06 3.00631E+06 l Amendment G April 30, 1990
? , .
CESSAR s!nificui: 1 I e! TABLE 6.2.1-7 (Cont'd)- (She.st 2 of 14) DOUBLE-ENDED DISCHARGE LEtt SLOT BREAK - MINIMUM SIS FIDW (9.8175 f6 2 Total Area) PART A: Mass / Energy Release Data j l B%W Mass . Break Energy Tilme Flcv Rate Flow Rate l '- (sec) (lbufsec) (Btu /sec) L 16.80 3.3772:E+03 1.55052E+06 l- 17.00 3.35931<E+03 1.53579E+06 17.20 3.34360E+03 1.52408E+06 ' l 17.40- 3.29544E+03 1,50291E+06 17.60 3.21787E+03 1.47673E+06 l
-17.80 2.85056E+03 1.31226E+06 l 18.00 2.54325E+03 1.17246E+06 ;
18.20 2.24486E+03 1.03981E+06 - 18.40 1.94972E+03 9.11453E+05. i 18.60 1.65285E+03 7.85437E+05 18.80 1.34658E+03 6.58857E+05 18.81 0.00000E+00 0.00000E+00 1 18.91 1.52940E+02 1.97890E+05 19.61 3.08650E402 4.01380E+05 20.21 3.49220E+02 4.54320E+05 20.71 4.79120E+02 G 6.23740E+05 21.31 6.09580E+02 7.93520E+05 L 21.91 7.24510E+02 9.42690E+05 22.41 8.12830E+02 1.05705E+06 22.71- 8.42050E+02 1.09491E+06 22.72 4.88389E+02 6.35048E+05 23.01 4.88372E+02 6.34955E+05 p 23.61. 4.88024E+02 6.34346E+05 i
.24.11 4.87722E+02 6.33894E+05 24.71 4.87310E+02 6.33273E+05 25.31 4.86835E+02 6.32583E+05 25.81 4.86492E+02 6.32194E+05 26.41 4.85982E+02 6.31469E+05 27.01 4.85460E+02 6.30738E+05 27.51 4.85025E+02 6.30129E+05 28.11 4.84503E+02 6.29393E+05 28.71 4.83987E+02 6.28668E+05' l 29.21 4.83581E+02 6.28169E+05
- 29.81 4.83094E+02 6.27479E+05
! 3l.41 4.82554E+02 6.26713E+05 L Amendment G April 30, 1990
CESSAR nnne.m
/
lg s , TABLE 6.2.1-7 (Conc'd) (Steet 3'of 14)- DOUBLE-ENDED DISCHARGE LIG SIDT BREAK - MINIMUM SIS FIDW
-(9.8175 ft Total Area) >>
PART A: Mass / Energy Releasts Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) , (lba/sec) (Btu /sec) 30.91 4.82125E+02 6.26104E+05 > 31.51 4.81620E+02 6.25391E+05 32.01 4.81249E+02 6.24927E+05 32.61 4.80745E+02 6.24213E+05 33.21 4.80240E+02 6.23500E+05 . 33.71 4.79817E+02 6.22903E+05
- l. 34.31 4.79312E+02 6.22189E+05 34.91 4.78807E+02 6.21476E+05
(' i 35.41 4.78390E+02 6.20878E+05
\
L - 36.01 4.77885E+02 6.20165E+05 I 36.61 4.77381E+02 6.19452E+05 37.11 4.76957E+02 6.18854E+05 37.71 4.76453E+02 6.18141E+05 38.31 4.75948E+02 6.17427E+05 i 38.81 4.75530E+02 6.16830E+05 a 39.41 4.75026E+02 6.16117E+05 40.01 4.74521E+02 6.15403E+05 40.51 -4.74098E+02 6.14812E+05 41.11 4.73605E+02 6.14116E+05
- 41.'71 4.73141E+02 6.13449E+05 42.21 4.72752E+02- 6.12898E+05
, 42.81 4.72288E+02 6.12236E+05 L 43.31 4.71900E+02 6.11685E+05 43.91 4.71430E+02 6.11018E+05 44.51 4.70966E+02 6.10357E+05 J 45.01 4.70577E+02 6.09806E+05 1 45.61 4.70107E+02 6.09139E+05 46.21 4.69638E+02 6.08478E+05 46.71 4.69249E+02 6.07927E+05 47.31 4.68785E+02 6.07260E+05 47.91 4.69315E+02 6.06599E+05 l 48.41 4.67927E+02 6.06048E+05 ; 49.01 4.67457E+02 6.05381E+05 l O
%/
49.61 50.11 4.66987E+02 4.66598E+02 6.04720E+05 6.04163E+05 i l l l Amendment G April 30, 1990 l 1
'CESSAR1RWiem g
TABLE 6.2.1-7 (Cont'd) (Sheet 4 of 14) DOUBLR-ENDED DISCHARGE LEG SIAT BREAK MINIMUM SIS FIDW (9.8175 ft Total Area) PART;A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rat.e Flow. Rate (sec) (lbe/sec) (Btu /sec) 50.71 4.66129E+02 6.03502E+05 51.31 4.65659E+02 6.02835E+05 51.81 4.65270E+02 6.02284E+05 52.41 4.64800E+02 6.01622E+05 53.01 4.64331E+02 6.00955E+05 53.51 4.63936E+02 6.00404E+05 54.11 4.63466E+02 5.99737E+05 54.61 4.63078E+02 5.99186E+05 55.21 4.62608E+02 5.98519E+05 55.81 4.62138E+02 5.97858E+05-56.31 4.61744E+02 5.97307E+05 56.91 4.61274E+02 5.96640E+05 57.51 4.60810E+02. 5.953*19E+05
-58.01 4.60421E+02 5. 954 28t;; 05 58.61 4.59957E+02 5.94773E+05 59.21 4.59493E+02 5.94111E+05 59.71 4.59105E+02 5.93566E+05 G 60.31 4.58641E+02 5.92905E+05 60.91 4.58171E+02 5.92244E+05 61.41 4.57788E+02 5.91699E+05 62.01 4.57318E+02 5.91037E+05 62.61 4.56854E+02 5.90382E+05 63.11 4.56466E+02 5.89831E+05 63.71 4.56002E+02 5.89170E+05 64.31 4.55532E+02 5.88514E+05 64.81 4.55143E+02 5.87963E+05 65.41 4.54679E+02 5.87302E+05 65.91 4.54291E+02 5.86757E+05 66.51 4.53821E+02 5.86096E+05 G7.11 '4.53357E+02 5.85435E+05 67.61 4.52963E+02 5.84884E+05 68.21 4.52499E+02 5.84228E+05 68.81 4.52029E+02 5.83567E+05 69.31 4.51640E+02 5.83016E+05 69.91 4.51170E+02 5.82355E+05 Amendment G April 30, 1990
. .~ _ -- - . - . - . .
1 1 1 b 12AT15N 1 I l TABLE 6.2.1-7 (Cont'd) I (Sheet 5 of 14) 1 DOUBLE-ENDED DISCHARGE LEG SLOT BREAK - MINIMUM SIS FIDW q (9.8175 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lbe/sec) (Btu /sec) 70.51 4.50701E+02 5.81690E+05 ; 71.01 4.50312E+02 5.81148E+05 71.61 4.49842E+02 5.80487E+05 -! 72.21 4.49372E+02 . 5.79826E+05 ' .l 72.71- 4.48984E+02 5.79281E+05 73.31 4.48514E+02 5.78620E+05 73.91 4.48044E+02 5.77958E+05
'74.41 4.47656E+02 5.7740" A05 75.01 4.47186E+02 5.76746E+05
~
75.50 4.46791E+02 5.76195E+05 g 75.51 7.70330E+02- 9.93440E+05 > p 84.81 6.51020E+02 8.39790E+05- ~ L 94.01 5.45410E+02 7.03930E+05 L 103.31 4.56190E+02 5.89090E+05
.112 . 51 ~ 3.83630E+02 4.95570E+45 l 121.81 3.24530E+02 4.19300E+05. G L 131.01 2.79010E+02 3.60450E+05' l 140.21 2.45140E+02 3.16610E+05 l'
149.51 2.21120E+02 2.85470E+05 L 158.71 2.0537GE+02 2.65040E+05-168.01 1.95350E+02 2.52020E+05 177.21 1.89350E+02 2.44210E+05 186.51 1.85770E+02 2.39540E+05 195.71 1.83730E+02 2.36860E+05 204.91 1.82580E+02 2.35330E+05 214.21 1.81930E+02 2.34450E+05 223.41 1.81590E+02 2.33970E+05 232.71 1.81420E+02 2.33710E+05 241.91 1.81350E+02 2.33580E+05 251.11 1.81330E+02 2.33510E+05 260.41 1.81340E+02 2.33480E+05 269.61 1.81370E+02 2.33480E+05 278.91 1.81400E+02 2.33480E+05 288.11 1.81450E+02 2.33490E+05 297.41 1.81490E+02 2.33500E+05 Amendment G April 30, 1990
CESSARHELuia
$1I TABLE 6.2.1-7 (Cont'd)
(Sheet 6 of 14) DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MINIMUM SIS FIDit 2 (9.8175 ft Total Area) PART A: Mass / Energy Release Data
-Break Mass Break Energy Time Flow Rate- Flow Rate (sec) (1ba/ sac) (Btu /sec) 306.61 1.81530E+02 2.33500E+05 315.81 1.81560E+02 2.33500E+05 325.11: 1.81600E+02 2.33500E+05 334.31 1.81630E+02 2.33500E+05
'343.61 1.81670E+02 2.33500E+05 352.81 1.81700E+02 2.33490E+05 362.11 1.81730E+02 2.33480E+05 371.31 1.81760E+02 2.33460E+05 380.51 389.81 1.81780E+02 1.81800E+02 2.33440E+05 2.33420E+05 .h 399.01 1.81830E+02 2.33400E+05 408.31 1.81850E+02 2.33370E+05 417.51 1.81870E+02 2.33340E+05 426.71 1.81880E+02 2.33310E+05 436.01 1.81900E+02 2.33280E+05 445.21 1.81920E+02 2.33240E+05 454.51 1.81930E+02 2.33210E+05 G 463.71 1.81950E+02 2.33170E+05 473.01 1.81960E+02 2.33130E+05 482.21- 1.81970E+02 2.33090E+05 491.41 1.81980E+02 2.33050E+05
.500.71 1.82000E+02. 2.33010E+05 509.91- 1.82010E+02 2.32960E+05 519.21 1.82020E+02 2.32920E+05 528.41 1.82030E+02 2.32870E+05 537.61 1.82040E+02 2.32820E+05 537.71 1.80000E+02 2.30000E+05 537 81 1.81175E+02 2.50000E+05 538.01 1.88825' 02 2.22010E+05 538.41 1.8018*,d+02 2.47990E+05 539.01 1.93146E+02 2.31612E+05 539.61 1.83964E+02 2.48388E+05 540.41 1.93536E+02 2.38536E+05 541.31 1.90543E+02 2.52575E+05 542.41 1.98548E+02 2.45805E+05 Amendment G April 30, 1990
1 q l l HEAkit 'c!!R!i,carit. l 1 l TABLE 6.2.1-7 (Cont'd) l
-(Sheet 7 of 14) ]
DOUBLE-ENDED DISCHARGE LEG SIDT' BREAK - MINIMUM SIS - FIDW j 2 (9.8175 ft Total. Area)
- PART A: Mass / Energy Release Data ,
Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1be/sec) (Btu /sec) 543.61 1.96468E+02 2.57528E+05' I 544.91 2.01994E+02 2.55210E+05 546.31 2.03950E+02 2.63361E+05 547.91 2.08550E+02 2.64670E+05 549.61- 2.09889E+02 2.69448E+05
'551.41 2.15667E+02 2.73885E+05 ;
553.41 2.16252E+02 2.77337E+05 555.51 2.21844E+02 2.82663E+05-p' A' 557.71 2.22641E+02 2.85085E+05; ; 560.11 2.27359E+02 2.89915E+05 562.61 2.29516E+02 2.93285E+05 , 565.21 2.32792E+02 2.96792E+05 567.91 2.34878E+02 3.00985E+05 570.81' 2.38226E+02 3.03386E+05 573.81 2.40032E+02 3.06682E+05 577.01 2.43093E+02 '3.10193E+05 G 580.31 2.44873E+02 3.12416E+05 583.71 2.47480E+02 3.15820E+05 587.21 2.48760E+02 3.17482E+05
- 590.91 2.51240E+02 3.19873E+05 594.71 2.52239E+02 3.22232E+05 598.61 2.54941E+02 3.23946E+05 602.71 2.55208E+02 3.25810E+05 606.91 2.57173E+02 3.27523E+05 611.21 2.59176E+02 3.28818E+05 615.71 2.59224E+02 3.30293E+05 620.31 2.60341E+02 3.31448E+05-625.01 2.61361E+02 3.32807E+05 629,81 2.62011E+02 3.33263E+05 634.81 2.62799E+02 3.34737E+05 639.91 2.63478E+02 3.35098E+05 645.21 2.64080E+02 3.35845E+05 650.61 2.64809E+02 3.36384E+05 3.36799E+05 -hN.
656.11 6G1.71 2.65055E+02 2.65660E+02 3.37487E+05 Amendment G April 30, 1990
l' ' . . E :CESSAR ;!ninco. H l -l l h. TABLE 6.2.1-7 (Cont'd) i (Sheet 8 of.14). - , l DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MINIMUM SIS FIXMt (9.8175 ft Total Area). PART A: Mass / Energy Release Data Break Mass Break Energy l Time Flow Rate Flow Rate (sec)- (1ba/sec) (Btu /sec) ;
- 667.51 2 ~. 65 8 05E+0 2 3.37685E+05'
'673.41 2.66140E+02 3.37944E+05 679.41 2.66860E+02 3.38389E+05 685.61 2.66563E+02 3.38246E+05 691.91 2.67088E+02 3.38897E+05 698.31 2.67067E+02 3.38495E+05 704.91 '2.67478E+02 3.39080E+05 711.61 2.67459E+02 3.38681E+05
.18.41 7 2.67571E+02 3.38966E+05 725.31 '2.67792E+02 3.38752E+05 i 732.41 2.67585E+02 3.38994E+05 739.61 2.67971E+02 3.38628E+05.
747.01 2.67975E+02 3.38685E+05 i 754.41 2.67840E+02 3.38883E+05 762.01 2.68212E+02 3.38391E+05 769.81 2.68011E+02 3.38532E+05 777.71 2.68192E+02 3.38430E+05 a
- 785.71 .2.68104E+02 3.38155E+05 793;81 2.68196E+02 3.38388E+05 802.01 2.68389E+02 3.'37S50E+05 810.41 2.68155E+02 3.37865E+05 819.01 2.68357E+02 3.37717E+05
.827.61 2.68387E+02 3.37632E+05 836.41 2.60704E+02 3.14186E+05 845.31 2.39827E+02 2.84027E+05 4 854.41 2.33140E+02 2.72456E+05 863.61 2.17800E+02 2.55917E+05 872.91 2.13168E+02 2.49460E+05 882.31 2.04913E+02 2.40596E+05 '
891.91 2.01545E+02 2.36071E+05 901.61 1.96752E+02 2.30726E+05 911.41 1.94065E+02 2.27846E+05 921.41 1.91238E+02 2.24052E+05 931.51 1.89158E+02 2.21889E+05 941.71 1.87163E+02 2.19389E+05 Amendment-G April 30, 1990
CESSAR Enema ' ($)I , TABLE 6.2.1-7 (Cont'd) (Sheet 9 of 14) f DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MINIMUM SIS FIDW 2 ' (9.8175 ft Total Area) ; PART.A: Mass / Energy Release Data Break Mass- Break Energy ; Time Flow Rate Flow Rate. >
-(sec) (lbm/sec) (Btu /sec)
~
952.11 1.85530E+02 2.17534E+05 962.61 1.84012E+02 2.15799E+05 , 973.21 1.82636E+02 2.14081E+05 983.91 1.81476E+02 2.12835E+05 994.81 1.80228E+02 2.11169E+05 1005.91 1.79232E+02 2.10272E+05 1017.01 1.78103E+02 2.08638E+05 1028.31 1.77296E+02 '2.07822E+05 1039.*,'1 1.76119E+02 2.06362E+05 .cO 1051.21 1.75233E+02 2.05638E+05' 1062.91 1.74511E+02 2.04233E+05 G 1074.71 l'73398E+02
. 2.03564E+05 1 1086.61 1.72821E+02 2.02156E+05 1098.71 1.71736E+02 2.01480E+05 1110.91' 1.71051E+02 2.00352E+05-1123.21 1.70250E+02 1.99362E+05 1135.71 1.69335E+02 1.98558E+05 ,
1148.31 1.68616E+02' 1.97517E+05 l 1161.01 1.67920E+02 1.96658E+05 1173.91 1.67120E+02 1.95757E+05 1186.81 1.66379E+02 1.94940E+05~ 1199.91 1.65682E+02 1.94067E+05 1 l l Amendment G April 30, 1990
.C N SSAR E Lm. O TABLE 6.2.1-7-(Cont'd)-
.(Sheet 10 of 14)
DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MINIMUM SIS FIDW SPILLAGE DATA Time, Sec Mass, iba Energy, 10 Btu g End'of Blowdown 18.81 0.0 0.0 End of Reflood $37.61 309,992 48.933 End of Post Reflood 1199.91 334,234 54.853 9 O Amendment G April 30, 1990
;; , J j l
J LCESSAR ERWrim . i b TABLE 6.2.1-7-(Cont'd)- l (Sheet 11 of 14) I l
.,,n -
DOUBLE-ENDED - DISCHARGE LEG SIDT BREAK -- MINIMUM SIS FIDW i l 2 (9.8175 ft Total Area). PART B: Reactor Vessel Pressure vs. Time Reactor Vessel Time (sec) Pressure (psia) 0.000 2275 0.028: 2194 0.053 1896 0.101. 1699 0.155 1686 0.202 1681 0.257 1670 p 0.308- 1650 1 0.504 1617 4 0.662 1604 i - ()l 0.812 1582 o (( L 0.952 1558 l 1.212 1526. 1.412 1517 ; 2.012 1431 ; 3.012 1268 4.012 1198 5.012 1161 6.006 1131 7.012 1097 8.004 1057 ' 9.000 995 10.00 933
. 11.00 835 12.00 686 L 14.00 377 L ,.
16.00 192 18.00 87 i O Amendment G April 30, 1990
CESSAR !!nha . l 3 TABLE 6.2.1-7 (Cont'd)- (Sheet 12 of 14) ; i DOUBLE-ENDED DISCHARGE LEG SIMF BREAK - MINIMUM SIS FLOW ! 2 (9.8175Lft Total' Area) i PART.B: . Reactor Vessel Pressure vs. Time f i Reactor Vessel l' Time (sec) Pressure,(psia)- 22.71 75.369 23.71 75.286 ! 24.71- 75.210 1 25.71 75.138 26.71 75.054-27.71 74.969 28.71 74.885 30.71 74.646 32.71 74.480 - i 34.71 74.315-36.71 74.149 - 38.71 74.067 40.71 73.902
, 42.71 73.775 44.71 73.650 t 46.71 73.526 G 48.71 73.402 50.71 73.278-52.71 73.155 54.71 73.032 i
56.71 72.910 58.71 72.789 60.71 72.668 70.71 72.070 80.71 69.181 99.71 61.998 119.91 57.925 139.91 56.042 159.91 55.213 179.91 54.812 199.91 54.557 219.91 54.346 239.91 54.148 259.91 53.954 279.91 53.761 299.91 53.567 Amendment G l April 30, 1990
l CESSAR1!nbe,.
. I TABLE 6.2.1-7 (Cont'd)
(Sheet 13 of 14) l DOUBLE-ENDED DISCHARGE LEG SIDT BREAK 'AINIMUN SIS FIDW (9.8175 ft ' Total Area) PART B: Reactor Vessel Pressure vs._ Time-Reactor Vessel Time ~(sec) Pressure (psia) 319.91 53.374 i 339.91 53.180 359.91 '52.986 > 379'.91 52.791
.399.91' 52.597 419.91 52.402 1439.91 52.207 459.91 52.012 l . 479.91 51.816 499.91 51.621 519.91 51.426 557.61- 51.650 617.61 51.672 757.61 50.433 897.61 48.117-917.61; 47.850 G 937.61 47'611 957.61 47.387 !
977.61 47.172 1019.91 46.730 1079.91- 46.116 1139.91 45.507 1179.91 45.104 1199.91 44.903 Amendment G April 30, 1990 l .~ T w -- - _ _ - - - - . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . _ _ _ _ _ m- _ - _ _ _ _ _ _ -
CESSARlinLmt g TABLE 6.2.1~7 (Cont'd) (Sheet-14 of 14) DOUBLE-ENDED DISCHARGE LEG SIDT BREAK - MINIMUM SIS FLOW (9.8175 ft2Total Area) PART C: Chronology of Events Time (seconds). Event
'O.00 Break Occurs 12.15 Start Safety Injection Tank Injection 16.38 Peak Containment' Pressure (Blowdown) 18.80 Start SIS Injection Phase 18.80 End of Blowdown 22.71- Downcomer Full 69.90 Start Spray Injection g 75.71 Safety Injection. Tank Empty 104.80 Peak Containment Pressure Subsequent To end of Blowdown .
537.61 End of Core Reflood 1199.91 End of Post Reflood 9 Amendment G April 30, 1990
CESSAR !!nkrin. N
' (J TABLE 6.2.1-8 (Sheet 1 of 4)
DOUBLE-ENDED HOT LEG SIDT BREAK (19.2423 ft Total Area) y PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lba/sec) (Btu /sec) 0.000 0.00000E+00 0.00000E+00 0.026 1.56946E+05 '9.91029E+07' O.051 1.45846E+05 9.15868E+07 0.102 1.61853E+05 1.01906E+08 0.149 1.69575E+05 1.07099T'.^8 0.205' 1.62257E+05- 1. 02 4 92 E M '3 0.252 1.51330E405 9.54450E+07 0.306 1.40105E+05 8.82183E+07 ; [,_)
\-
0.358- 1.33914E+05 8.42386E+07 0.504 1.28192E+05 8.04605E+07 0.651 1.22435E+05 7.65941E+07 0.815 1.19409E+05 7.45462E+07 0.9S0 1.15364E+05 7.19791E+07 G 1.210 1.09824E+05 6.87925E+07 1.410 1.04354E+05 6.60228E+07 1.61 1.00091E+05 6.39493E+07 1.81 9.61342E+04 6.20413E+07 2.01 9.25379E+04 5.98969E+07 3.01 8.60811E+04 5.42989E+07 4.01 7.11911E+04 4.66043E+07 5.01 5.86461E+04 4.01299E+07. , 6.01 4.57787E+04 3.37364E+07 7.01 3.26309Et04 2.64492E+07 8.01 1.39344E+04 1.44302E+07 9.01 6.85473E+03 7.23996E+06 10.02 8.51957E+03 9. 254 2 6E4 06 11.02 9.36607E+03 9.96197E+06 12.02 9.58873E+03 9.32038E+06 13.02 8.85817E+03 8.06449E+06 14.01 4.19301E+03 4.71867E+06 15.01 5.04376E+02 5.16479E+05 15.20 3.30535E+02 3.45891E+05 15.40 1.22793E+02 1.03569E+05 15.60 0.00000E+00 0.00000E+00 Amendment G April 30, 1990 L
CESSARnnLm O TABLE 6.2.1-8 (Cont'd) (Sheet 2 of 4) DOUBLE-ENDED HOT LEG SIDT BREAK SPILLAGE DATA Time, Sec Mass, Ibn Energy, 10 Btu G End of Blowdown 15.60 0.0 0.0 0
#ll Amendment G April 30, 1990
- \
^ C E S S A R SI M .eu o.
O TABLE 6.2.1-8 (Cont'd) (Sheet 3 of 4) DOUBLE-ENDED HOT TAG SIAT BREAK (19.2423 ft Total Arsa) PART B: Reactor Vessel Pressurs vs. Time s Reactor Vessel Time (sec) Pressure (psia) 0.000 2275 s 0.025 1982 0.506 1700 0.101 1685 0.149 1669 0.204 1654 0.252 1636 0.357 1593 l+
- 0.503 1529 O.650 1482
- s. 0.814 1448 0.950 1394 1.210 1306 1.410 1244 G 2.010 1053 3.010 975 4.010 915 5.010 831 6.010 742 7.010 591 8,009 353 9.007 195 10.018 300 11.018 286 12.018 253 13.018 203 14.007 120 15.005 58 O
Amendment G April 30, 1990 e
C: ! CESSARff8h a - O
"!ABLE 6.2.3-8 (Cont'd)
(Sheet 4 of 4) l DOUBLE-ENDED H&P LEG SI&f BREAK l (19.2423 ft Total Area) 1 PART C: Chronology of Events l Tinc. (seconds) Event G 0.00 Break Occurs l 6.71 Start Safety Injection Tank Injection 14.48 Peak Containment Pressure (Blowdown) 15.60 End of Blowdown O O Amendment G April 30, 1990
CESSAR !!RL . C ; TABLE f.2.1-9 (Sheet 1 of 7) AIN STEAM LINE BREAK, 102% POWER - i IDSS OF ONE CSS TRAIN ) (8.72 ft Total Area) PART A: Mass / Energy Release Data l Break Mass Break Energy Time Flow Rate Flow Rate ! (sec) (Ibe/sec) (Btu /sec) 0.00 9323.395 1.11125E+07 0.50 8743.53G 1.04388E+07 ! 0.99 8446.128 1.00922E+07 l 1.49 8008.505 9.81471E+06 1.99 7994.404 9.56424E+06 2.49- 7800.098 9.33654E+06 2.99 7623.162 9.12887E+06 ( 3.49 7553.123 9.04699E+06 3.99 7814.884 9.35584E+06 4.49 7914.109 9.47278E+06 - 4.99 7916.190 9.47545E+06 5.49 7888.772 9.44344E+06 5.99 7854.208 9.40301E+06 G 6.49 7819.081 9.36189E+06 6.99 7785.357 9.32239E+06 7.49 7753.912 9.28554E+06 7.99 7725.139 9.25181E+06 8.49 2388.295 2.85908E+06 8.99 2400.749 2.87365E+06 9.49 2412.599 2.88751E+06 9.99 2423.798 2.90060E+06 10.49 2434.213 2.91277E+06 10.99 2443.651 2.92380E+06 11.49 2451.880 2.93341E+06 ' 11.99 24.50.630 2.94132E+06 12.99 2466.868 2.95091E+06 13.99 2474.801 2.96017E+06 14.99 2477.95L 2.96385E+06 15.99 2471.426 2.95623E+06 16.99 2456.709 2.93905E+06 17.99 2435.842 2.91468E+06 18.99 2411.076 2.88573E+06 19.99 2384.539 2.85468E+06 Amendment G April 30, 1990
~. - . _ . _ _ _ _ .. _. _ _ __ . _ _.
1
'( !$$h$h.k.hI !!h$kmAYlor, e
TABLE 6.2.1-9 (Cont'd) (Sheet 2 of 7) . MAIN STEAM LINE BREAK, 102% POWER - 1 14S8 OF ONE CSS TRAIN 2 (8.72 ft Total Area) PART At Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec)- __ (Ibm /sec) (Btu /sec) 21.99 2331.730 2.79399E+06 23.99 2288.540 2.74215E+06 . 27.00 2236.221 2.68070E+06 29.00 2205.563 2.64462E+05 31.00 2174.052 2.60752E+06 33.00 2141.110 2.56970E+06 35.00 2108.113 2.52978E+06 37.00 2076.802 2.49291E+06 39.00 2048.515 2.45938E+06 41.00 2022.914 2.42910E+06 43.00 1999.376 2.40124E+06 45.00 _1977.028 2.37478E+06 a , 47.00 1955.603 2.34940E+06
- 49.00 1935.101 '2.32511E+06 54.00 1885.774 2.34207E+06 L 59.00 1840.736 2.28547E+06 L 64.00 1794.233 2.22721E+06 69.00 1737.531 2.15711E+06 ;
! 74.00 1683.010 2.09014E+06 79.00 1632.572 2.02792E+06 84.00 1584.388 1.96819E+06 89.00 1537.343 1.90996E+06 }
- 94.00 1491.409 1.85314E+06 i 99.00 1446.300 1.79735E+06 l- 109.00 1350.948 1.67970E+06 L 119.00 1252.714 1.55894E+06 129.00 1180.127 1.46887E+06 139.00 1128.675 1.40399E+06 149.00 1070.632 1.33161E+06 159.00 1008.543 1.25452E+06 169.00 942.489 1.17296E+06 l 179.00 872.722 1.08703E+06 Amendment G April 30, 1990
CESSAR !BMP.co.. O TABLE 6.2.1-9 (Cont'd) (Sheet 3 of 7)
!!!LIN STFAM LINE BREAK, - 102% POlmR -
IDSS OF ONE CSS TRAIN (8.72 ft 2Total Area) PART At Mass / Energy Release Data i Break Mass Break Energy Time Fliv Rate Flow Rate (sec) (1bs/sec) (Bte/sec) 189.00 799.402 9.96844E+05 199.00 722.693 9.02490E+05 209.00 643.296 8.04706E+05 219.00 560.726 7.02732E+05 229.00 487.049 6.11411E+05 239.00 429.919 5.40346E+05 ! G 249.00 375.605 4.72639E+05 1 259.00 335.175 4.22424E+05 263.90 270.192 3.20550E+05 279.00 111.812 1.41328E+05 ; 359.00 110.758 1.40000E+05 i 399.00 110.000 1.39040E+05 ! 459.00 110.599 1.39800E+05 499.00 110.555 1.39746E+05 O Amendment G l April 30, 1990
hhk !E !I tCATION l l O1 l TABLE 6.2.1-9 (Cont'd) l (Sheet 4 of 7) MAIN STEAN LINE BREAK, 102% PONER - IDSS OF ONE CSS TRAIN 2 (8.72 ft Total Area) ) l PART B: Steam Generator Press,ures 1 Unaffected Steae Affected Steam Time Generator Generator j (sec) Pressure (psia) Pressure (psia) < 0.00 1021.0 1021.0 0.50 1011.6 1004.5 0.99 995.4 986.0 1.49 979.7 968.2 1.99 965.1 951.8 2.49 951.4 936.5 2.99 938.4 922.3 3.49 926.2 909.0
- 3.99 914.8 897.2 l 4.49 907.7 888.8 l
4.99 903.3 J32.4 G 5.49 899.7 877.0 5.99 896.3 872.1 6.49 893.0 867.6 6.99 889.9 863.5 7.49 886.9 859.8 7.99 884.2 856.5 8.49 887.4 855.3 8.99 910.2 859.8 l 9.49 932.5 864.1 9.99 954.2 868.2 1 10.49 975.1 872.0 i 10.99 995.2 875.4 h 11.49 1014.5 878.4 11.99 1032.7 880.9 12.99 1065.9 883.8 13.99 1093.7 886.7 14.99 1115.7 887.8 15.99 1132.1 885.4 16.99 1143.5 880.0 17.99 1150.9 872.4 l 18.99 1155.4 863.4 l O Amendment G April 30, 1990
i CESSAR !HMrien.. : O TABLE 6.2.1-9 (Cont'd) (Sheet 5 of 7) NAIN STEAM LINE BREAK, 102% POWER - IDSS OF ONE CSS TRAIN (8.72 ft 2Total Area) PART B: Steam Generrtor Pressures Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 19.99 1157.9 853.7 21.99 1160.5 834.9 23.99 1162.7 818.8 27.00 1167.0 799.3 29.00 1168.2 788.1 31.00 1166.7 776.6 33.00 1162.5 764.6 O- 35.00 37.00 1156.5 1152.6 752.6
'741.3 39.00 1152.5 731.0 41.00 1152.4 721.7 43.00 1152.3 713.2 G 45.00 1152.2 705.1 47.00 1152.1 697.1 49.00 1152.0 689.3 54.00 1151.8 670.5 59.00 1151.5 653.3-64.00 1151.3 635.5 69.00 1151.0 613.8 74.00 1150.7 593.0 l 79.00 1150.5 573.7 84.00 1150.2 555.7 i 89.00 1149.9 537.3 94.00 1149.6 519.8 )
'99.00 1149.4 502.5 1 109.00 1148.8 467.7 i 119.00 1148.3 431.9 129.00 1147.7 405.6 139.00 1147.2 386.9 149.00 1146.7 365.7 159.00 1146.2 343.1 1
Amendment G April 30, 1990 _. _ _ _ l
;CESSARinW.co . -
g: TABLE 6.2.1-9 (Cont'd) , (Sheet 6 of 7) MAIN STEAM LINE BREAK, 102% PONEk - LOSS OF ONE CSS TRAIN 2 (8.72 ft Total Area) PART B: Steam Generator Pressures Unaffected Steae Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 160.00 1145.6 319.1 179.00 1145.1 293.7 189.00 1144.6 266.9 199.00 1144.1 239.0 209.00 1143.6 210.1 219.00 1143.2 180.9
'229.00 1142.7 155.4 239.00 1142.3 135.6 249.00 1141.9 116.7 <
259.00 1141.6 95.3 263.90 1141.4 85.7 279.00 1140.8 62.5 359.00 1139.0 60.9 g 399.00 1138.2 60.3 459.00 1137.0 59.4 499.00 1136.3 58.9 L Amendment G April 30, 1990 l l 1
4:IEdlEBAkit ;!.Mrication 1 l TABLE 6.2.1-9 (Cont'd) (Sheet 7 of 7) l MAIN STEAM LINE BREAK, 102% PONER - 1 IhSS OF ONE CSS TRAIN 4 2 (8.72 ft Total Area) PART C: Accident Chronology j Time (secs) Event Setpoint 0.00 Break Occurs 2.23 Containment Pressure Reaches 4.0 psig Reactor Trip Setpoint 2.23 Containment Pressure Reaches 4.0 psig Main Steam Isolation Signal
- (MSIS) Analysis Setpoint I \- 3.23 High Containment Pressure Reactor Trip Signal and MSIS Generated 3.38 seactor Trip Breakers Open g 3.38 Turbine Admission Valves Closed l 5.80 Containment Spray Actuation 10.0 psig ,
Signal l 8.38 Main Steam isolation Valves Closed 13.38 Main Feedwater Isolation Valves closed 73.80 Containment Spray Reaches Containment 73.80 Peak Containment Temperature 267.80 Peak Containment Pressure Amendment G ! April 30, 1990
CESSAR EiWico.., . O TABLE 6.2.1-10 (Sheet 1 of 7) MAIN STEAM LINE BREAK, 102% POWER - 1 MSIV FAILURE (%72 ft 2Total Area) PART As Mass / Energy lelei.'ei Data BrMk Mass Break Energy Time F1,w Rate Flow Rate (sec)- (1bm/sec) (Btu /sec) 1 1 0.00 9323.395 1.11125E+07 i 0.50 8743.538 1.04383E+07 I 0.99 8446.128 1.00922E+07 1.49 8208.505 9.81471E+06 1.99 7994.404 9.56424E+06 2.49 7800.098 9.33654E+06 2.99 7623.162 9.12887E+06 i t' 3.49 7553.123 9.04699E+06 3.99 7814.884 9.35584E+06 4.49 7914.109 9.47278E+06
'4.99 7916.190 9.47545E+06 5.49 7888.772 9.44344E+06 5.99 7854.208 9.40301E+06 G 6.49 7819.081 9.36189E+06 6.99 7785.357 9.32239E+06 ;
7.49 7753.912 9.28554E+06 7.99 7725.139 9.25181E+06 i 8.49 7578.083 9.07881E+06 8.99 6990.849 8.38486E+06 9.49 6470.701 7.76663E+06 ; 9.99 6010.419 7.21696E+06 10.49 5603.319 6.72895E+06 10.99 5243.297 6.29607E+06 + 11.49 4924.738 5.91217E+06 11.99 4642.607 5.57161E*06 ' 12.99 4169.872 5.00024E+06 13.99 3796.363 4.54845F+06 14.99 3500.1?1 4.19046E+06 15.99 3259.90L 3.900923+06 16.99 3062.516 3.66377E+06 l 17.99 2898.771 3.46769E+06 18.99 2762.361 3.30486E+06 19.99 2648.965 3.1698!iE+06 21.99 2479.392 2.9685:3E+06 Amendment G l-l April 30, 1990 I-
CESSAR iinincei: O TABLE 6.2.1-10 (Cont'd) (Sheet 2 of 7) MAIM STEAM LINE BREAK, 102% POWER - MSIV FAILURE 2 (8.72 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1be/see) (Btu /sec) 23.99 2369.645 2.83859E+06 27.00 2274.596 2.72628E+06 29.00 2234.656 2.67918E+06 31.00 2200.541 2.63898E+06 33.00 2168.559 2.60128E+06 35.00 2136.741 2.56375E+06 37.00 2104.658 ?.52585E+06 39.00 2074.445 '.49013E+06 41.00 2046.705 2.45731E+06 43.00 2021.168 2.42708E+06 45.00 1997.486 2.39903E+06 47.00 1975.131 2.37254E+06 G 49.00 1953.814 2.34728E+06 54.00 1903.777 2.363J5E+06 59.00 1841.939 2.28689E+06 64.00 1796.953 2.23042E+06 69.C0 1739.892 2.15990E+06 74.00 1685.393 2.'09296E+06 79.00 1646.873 2.04482E+06 84.00 1599.198 1.98569E+06 89.00 1537.636 1.91030E+06 94.00 1493.620 1.85575E+06 99.00 1449 591 1.80124E+06 109.00 1353.860 1.68314E+06 119.00 1254.034 1.56050E+06 129.00 1181.503 1.47049E+06 139.00 1126.669 1.40158E+06 149.00 1068.707 1.32929E+06 159.00 1007.240 1.25295E+06 169.00 940.314 1.17034E+06 179.00 876.790 1.09184E+06 189.00 802.757 1.00080E+06 199.00 727.730 9.08431E+06 Amendment G April 30, 1990
Y CESSAR;3Wco
'I TABLE 6.2.1-10 (Cont'd)
(Sheet 3 of 7) MAIN STEAM LINE BREAK, 102% POWER - MSIV FAIIRRE (8.72 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lba/sec) (Btu /sec) G 209.00 642.590 8.03859E+05 219.00 562.277 7.04561E+05 229.00 487.598 6.12059E+05 239.00 434.278 5.45484E+05 249.00 373.103 4.69661E+05 259.00 320.472 4.03894E+05 264.90 296.037 3.51325E+05 O
- 279.00 113.418 1.43238E+05 299.00 111.734 1,41231E+05 359.00 114.545 1.44517E+05
-399.00- 111.332 1.40729E+05 459.00 111.189 1.40550E+05 499.00 111.112 1.40454E+05 O
Amendment G April 30, 1990
CESSAR aninemo ,
, l TABLE 6.2.1-10 (Cont'd)
(Sheet 4 of.7) MAIN STEAM LINE BREAK, 102% POWER - AdIV FAILURE (8.72 ft 2Total Area) PART B: Steam Generator Pressures , Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) O.00 1021.0 1021.0 0.50 1011.6 1004.5 0.99 995.4 986.0 1.49 979.7 968.2 1.99 965.1 951.8 2.49 951.4 936.5 2.99 938.4 922.3 3.49 926.2 909.0 3.99 914.8 897.2 4.49 907.7 888.8 4.99 903.3 882.4 5.49 899.7 877.0 5.99 '96.3 872.1 6.49 893.0 867.6 6.99 889.9 863.5 7.49 G
- 886.9 859.8 7.99 884.2 856.5 l
8.49 887.4 855.3 i 8.99 910.2 859.8 , 9.49 932.5 864.1 l 9.99 954.2 868.2 10.49 975.1 872.0 10.99 995.2 875,4 11.49 1014.5 878.4 11.99 1032.7 880.9 12.99 1065.9 883.8 13.99 1093.7 886.7 14.99 1115.7 887.8 I 15.99 1132.1 885.4 16.99 1143.5 880.0 17.99 1150.9 872.4 ' 18.99 1155.4 863.4-19.99 1157.9 853.7 21.99 1160.5 834.9 - Amendment G April 30, 1990
CESSAR !!!Mieu.. I TABLF 6.2.1-10 (Cont'd) (Sheet 5 of 7) , MAIN STEAM LINE BREAK, 102% POWER - MSIV FAILURE (8.72 ft 2Totcl Area) PART B: Steam Generator Pressures i Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (pela) Pressure (psia) 23.99 1162.7 818.8 27.00 1167.0 799.3 29.00 1168.2 788.1 31.00 1166.7 776.6 33.00 1162.5 764.6 35.00 1156.5 752.6 37.00 1152.6 741.3 _; 39.00 1152.5 731.0 41.00 1152.4 721.7 i 43.00 1152.3 713.2 l 45.00 1152.2 705.1 1 47.00 1152.1 697.1 , 49.00 1152.0 689.3 l 54.00 1151.8 670.5 G l 59.00 1151.5 653.3 ) 64.00 1151.3 635.5 l 69.00 1151.0 613.8 l 74.00 1150.7 593.0 79.00 1150.5 573.7 84.00 1150.2 555.3' 89.00 1149.9 .S37.3 94.00 1149.6 519.8 99.00 1149.4 502.5 i 109.00 1148.8 467.7 l 119.00 1148.3 431.9 129.00 1147.7 405.6 139.00 1147.2 386.9 149.00 1146.7 365 7 159.00 1146.2 342.1 169.00 1145.6 3T .1 179.00 1145.1 253.7 189.00 1144.6 266.9 f] V. 199.00 1144.1 239.0 l Amendment G April 30, 1990
CESSAR !!nh. ' I I TABLE 6.2.1-10 (Cont'd) ! l (Sheet 6 of 7) l MAIN STEAM LINE 3REAK, 102% POWER - MSIV FAILURE l (8.72 ft Total Area) , PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator- Generator (sec) Pressure (psia) Pressure (psia) 209.00 1143.6 210.1 219.00 1143.2 180.9 229.00 1142.7 155.4 239.00 1142.3 135.6 249.00 1141.9 116.7 259.00 1141.6 98.0 264.90 1141.3 86.5 , 279.00 1140.8 61.7 299.00 1140.3 60.9 359.00 1139.0 58.6 399.00 1138.2 57.2 459.00 1137.0 55.3 499.00 1136.2 54.2 I-O P Amendment G April 30, 1990 l
i CESSAR ERWico... lO TABLE 6.2.1-10 (Cont'd) (Sheet 7 of 7) , MAIN STEAM LINE BREAK, 102% POOFER - MSIV FAILURE 2 (8.72 ft Total Area) PERT C: Accident Chronology Time Setpoint (secs) _, , Event 0.00 Break Occurs 4 2.23 Containment Pressure Reaches 4.0 psig Reactor Trip Setpoint 2.23 Containment Pressure Reaches 4.0 psig Main Steam Isolation Signal (MSIS) Analysis Setpoint 3.23 High containment Pressure Reactor Trip Signal-and MSIS Generated G 3.38 Reactor Trip Breakers Open < 3.38 Turbine Admission Valves Closed 5.80 Containment Spray Actuation Signal 10.0 psig 8.38 Main Steam Isolation Valves Closed 13.38 Main Feedwater Isolation Valves Closed 73.80 Containment Spray Reaches containment 73.80 Peak Containment Temperature 246.80 Peak Containment Pressure O Amendment G April 30, 1990
CESSAR Elen... - a v TABLE 6.2.1-11 (Sheet 1 of 7) MAIN STEAM LINE BREAK,-50% POWER - IDSS OF ONE CSS TRAIN ; (8.72 ft 2Total Area) PART A:- Mass / Energy Release Data Break Masa Break Energy Time Flow Rate Flow Rate (sec) (1be/sec) (Stu/sec) 0.00 9719.073 1.10190E+07 0.50 9222.472 1.09949E+07 0.99 8964.395 1.06958E+07 ' 1.49 8765.113 1.04646E*07 1.99 8591.319 < 1. 02 62 6E4 07 2.49 8432.102 1.00773E;S7 2.99 8282.903 5.90334E+06 L ( 3.49 8229.449 5.84134E+06 3.99 8195.518 9.80206E+06 L 4.19 8116.798 9.71021E+06 4.99 8029.477 9.60819E+06 0 5.49 7942.380 9.50634E+06 5.99 7858.112 9.40770E+06 6.49 7777.374 9.31312E+06 6.99 7700.410 9.22290E+06 7.49 7627.361 9.13720E+06 7.99 7558.226 9.05605E+06 8.49 2322.795 2.78234E+06 8.99 2320.635 2.77981E+06 9.49 2318.550 2.77737E+06 ., 9.99 2316.543 2.77501E+06 7 10.49 2314.563 2.77269E+06-10.99 2312.518 2.77029E+06 11.49 2310.287 2.76767E+06 ( 11.99 2307.723 2.76467E+06 12.99 2300.989 2.75677E+06 13.99 2296.769 2.75181E+06 14.99 2290.012 2.74388E+06 15.99 2278.453 2.73031E+06 16.99 2262.480 2.71155E+06 17.99 2242.905 2.68855E+06 18.99 2220.809 2.66256E+06 19.99 2197.352 2.63496E+06 O. 21.99 2150.424 2.57968E+06 Amendment G April 30, 1990
CESSARin hio. O TABLE 6.2.1-11 (Cont'd) (Sheet 2 of 7) MAIN STEAM LINE BREAK, 50% POWER - l IASS OF ONE CSS TRAIN 2 (8.72 ft Total Area) PART A: Mass / Energy Release Data l Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lbe/sec) JBtu/sec) 23.99 2107.805 2.52941E+06 1 27.00 2653.595 2.46538E+06 l 29.00 2021.735 2.42771E+06 31.00 1990.501 2.39074E+06 33.00 1959.326 2.35381E+06 l 35.00 1929.926 2.31897E+06 37.00 1901.895 2.28573E+06 l l 39.00 1875.916 2.25490E+06 41.00 1852.057 2.22657E+06 l 43.00 1829.892 2.20023E+06 l 45.00 1808.878 2.17524E+06 47.00 1788.643 2.15116E+06 l 49.00 1769.043 2.12783E+06 54.00 1723.094 2.13762E+06 59.00 1681.248 2.08517E+06 64.00 1641.874 2.03558E+06 69.00 1604.441 1.98847E+06 G 74.00 1567.325 1.94188E+06 79.00 1527.382 l'.89201E+06 84.00 1480.416 1.83418E+06 89.00 1437.098 1.78093E+06 94.00 1395.443 1.?2935E+06 99.00 1355.709 2.68001E+06 109.00 1279.992 1.58610E+06 119.00 1208.072 1.49694E+06 129.00 1136.571 1.40855E+06 139.00- 1064.651 1.31994E+06 149.00 993.335 1.23220E+06 159.00 929.243 1.15316E+06 169.00 900.563 1.11680E+06 179.00 868.403 1.07595E+06 389.00 832.651 1.03115E+06 199.00 793.439 9.82406E+05 Amendment G April 30, 1990
CESSAR ine'c n. O TABLE 6.2.1-11 (Cont'd) (Sheet 3 of 7) MAIN STEAM LINE BREAK, 50% Potter - LOSS OF ONE CSS TRAIN (8.72 ft 2Total Area) PART At Mass / Energy Release Data Break Mass Break Energy i Time Flow Rate Flow Rate (sec) (Ibe/sec) (Btu /sec) L 209.00 751.900 9.30969E+05 I 219.00 708.482 8.77393E+05 3 229.00 663.514 8.22028E+05 239.00 617.286 7.65183E+05 249.00 568.265 7.04907E+05 259.00 518.860 6.44158E+05 269.00 469.631 5.83565E+05 ' 5.24002E+05 O-L 279.00 289.00 421.286 381.105 4.74393E+05 , 299.00 391.149 4.87322E+05 1 309.00 322.334 4.01755E+05 319.00 265.553 3.31066E+05 2.64757E+05 G 329.00 212.316 339.00 161.792 2.01813E+05 349.00 118.102 1.47350E+05 359.00 116.108 1.44852E+05 369.00 133.040 1.65914E+05 379.00 116.289 1.45020E+05 389.00 93.605 1.16756E+05 399.00 94.726 1.18163E+05 j 419.00 129.916 1.61912E+05 439.00 100.751 1.25563E+05 459.00 95.960 1.19600E+05 479.00 134.085 1.66947E+05 499.00 104.290 1.29856E+05 j i i i O Amendment G April 30, 1990
CESSAR iinkm O TABLE 6.2.1-11 (Cont'd) (Sheet 4 of 7) MAIN STEAM LINE BREAK, 50% POWER - IDSS OF ONE CSS TRAIN 2 (8.72 ft Total Area) PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 0.00 1052.0 1052.0 0.50 1040.5 1032.3 0.99 1023.3 1011.3 1.49 1005.8 991.5 1.99 988.7 972.8 2.49 972.3 955.3 2.99 956.6 938.8 3.49 942.0 923.4 3.99 930.7 910.5 4.49 920.8 899.0 d.00 911.3 888.2 5.49 902.2 878.1 5.99 893.4 868.6 6.49 884.9 859.5 6.99 876.7 851.0 G 7.49 868.9 843.0 7.99 861.5 835.4 8.49 864.1 831.3 8.99 876.9 830.5 9.49 889.5 829.8 9.99 901.9 829,1 10.49 914.0 828.3 10.99 925.8 827.6 11.49 937.3 826.8 11.99 948.5 825.8 12.99 969.2 823.4 13.99 987.6 821.9 14.99 1003.1 819.4 15.99 1015.7 815.2 16.99 1025.3 809.3 17.99 1032.4 802.2 18.99 1037.5 794.1 19.99 1040.9 785.6 21.99 1045.1 768.5 Amendment G April 30, 1990
C E S S A R EI! #icu e l O l TABLE 6.2.1-11 (Cont'd) (Sheet 5 of 7) MAIN STEAM LINE BREAK, 50% PONER - IDSS OF ONE CSS TRAIN 2 (8.72 ft Total Area) PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 23.99 1048.0 753.0 27.00 1051.5 732.8 29.00 1052.4 721.2 31.00 1051.4 709.8 33.00 1048.3 698.4 - 35.00 1043.8 687.2 37.00 1038.4 676.5 O 39.00 41.00 1034.4 1034.3 666.6 657.5 43.00 1034.3 649.1 45.00 1034.3 641.1 47.00 1034.3 633.4 49.00 1034.2 625.9 i 54.00 1034.1 608.3 G l 59.00 1034.0 592.4 i 64.00 1033.9 577.3 l 69.00 1033.8 563.0 74.00 1033.7 548.8 l 79.00 1033.6 533.5 ' 84.00 1033.5 515.6 89.00 1033.3 499.1 l 94.00 1033.2 483.9 i 99.00 1033.1 469.5 109.00 1032.8 441.9 119.00 1032.5 415.7 129.00 1032.2 389.7 l 139.00 1031.9 363.5 l 149.00 1031.5 337.6 159.00 1031.2 314.4 j 169.00 1030.3 303.9 179.00 1030.5 292.2 189.00 1030.1 279.2 l 199.00 1029.7 264.9 l l Amendment G April 30, 1990
CESSAR anincui:. O TABLE 6.2.1-11 (Cont'd) (Sheet 6 of 7) MAIN STEAM LINE BREAK, 50% POWER - IDSS OF ONE CSS TRAIN (8.72 ft Total Area) PART B: Steam Generator Pressures l l Unaffected Steam Affected Steam j Time Generator Generator I (sec)- Pressure (psia) })ressure (psia) ! l 209.00 1029.3 249.8
)
219.00 1028.9 234.0 l 229.00 1028.5 217.6 j 239.00 S.028.1 200.7 249.00 *iO27.7 183.7 259.00 1027.3 166.5 269.00 1026.9 149.4 279.00 1026.4 132.6 289.00 1026.0 118.7 299.00 1025.6 103.8 309.00 1025.2 93.7 319.00 1024.7 85.7 ' 329.00 1024.3 78.7 - 339.00 1023.9 72.7 349.00 1023.4 68.3 i 359.00 1023.0 68.0 369.00 1022.6 69.3 l 379.00 1022.2 67.8 ! 389.00 1021.8 65.6 l 399.00 1021.3 65.5 '~ 419.00 1020.5 68.1 l '439.00 1019.6 65.4 l 459.00 1018.8 64.7 l: 479.00 1017.9 67.7 499.00 1017.1 64.9 l-l l O Amendment G l 14pril 30, 1990
CESSAR !!nhua _.
- ('^
v-TABLE 6.2.1-11 (Cont'd) (Sheet 7 of 7) MAIN STEAM LINE BREAK, 50% PONER - IDSS OF ONE CSS TRAIN (8.72 ft Total Area) PART C: Accident Chronology Time (secs) Event Setpoint 0.00 Break occurs 2.10 Containment Pressure Reaches 4.0 psig Reactor Trip Setpoint 1 2.10 Containment Pressure Reaches 4.0 psig Main Steam Isolation Signal (MSIS) Analysis Setpoint 3.10 High containment Pressure Reactor Trip Signal and MSIS Cenerated 3.25 Reactor Trip Breakers Open G 3.25 Turbine Admission Valves closed t 5.80 Containment Spray Actuation 10.0 psig Signal 8.25 Main Steam Isolation Valves ' closed 13.25 Main Feedwater Isolation Valves Closed 73.80 Containment Spray Reaches Containment 73.80 Peak Containment Temperature 321.80 Peak Containment Pressure O Amendment G April 30, 1990
.i-
CESSAR autricui:,, ; 1 1 O ! TABLE 6.2.1-12 (Sheet 1 of 7) l MAIN STEAM LINE BREAK, 504 POEMR - l nsrv FuwR 2 (8.72 ft Total Ars.a) PART A: Mass / Energy Release Data Break mass Break Energy Time Flow Rate Flcnt Rate (sec) (1be/sec) (Btu /sec) 0.00 9719.073 1.15690E+07 1 0.50 9222.472 1.09949E+07 0.99 8964.395 1.06958E+07 1.49 8765.113 1.04646E+07. I 1.99 8591.319 1.02626E+07 l 2.49 8432.102 1.00773E+07 2.99 8282.903 9.90334E+06 e . 3.49 8229.449 9.84134E+06 3.99 8195.518 9.80206E+06 4.49 8116.798 9.71021E+06
-4.99 8029.477 9.60819E+06 5.49 7942.380 9.50634E+06 5.99 7858.112 9.40770E+06 6.49 7777.374 9.31312E+06 6.99 7700.410 9.22290E+06 G 7.49 7627.361 9.13720E+06 7.99 7558.226 9.05605E+06 8.49 7238.303 8.67890E+06 8.99 6671.346 8.00729E+06 9.49 6168.335 7.40837E+06 9.99 5722.438 6.87524E+06 10.49 5327.377 6.40132E+06 10.99 4977.432 5.98043E+06 11.49: 4667.403 5.60682E+06 11.99 4392.573 5.27517E+06 12.99 3929.958 4.71634E+06 15.99 3565.094 4.27539E+06 14.99 3275.782 3.92611E+06 15.99 3042.762 3.64540E+06 16.99 2852.861 3.41724E+06 17.99 2696.449 3.22986E+06 18.99 2F66.698 3.07484E+06 O 19.99 21.99 c.D.000 2096.841 2.94648E+06 2.75372E+06 l
Amendment G April 30, 1990
C E S S A R T!R % m.. i e1 l TABLE 6.2.1-12 (Cont'd) (Sheet 2 of 7) MAIN STEAM LINE BREAK, 50% POWER - MSIV FAILURE 2 (8.72 ft Total Area) l PART A: Mass / Energy Release Data
]
l Break Mass Break Energy ) Time Flow Rate Flow. Rate l (sec) (1be/sec) (Btufsec) 23.99 '2189.989 2.62700E+06 ; 27.00 2097.706 2.51771E+06 29.00 2056.000 2.46834E+06 31.00 2020.720 2.42656E+06 33.00 1988.140 2.38795E+06 35.00 1957.144' 2.35122E+06 , 37.'00 1928.103 2.31677E+06 ! 39.00 15)0.131 2.28357E+06 41.00 1874.980 2.25370E+06 - 43.00 1850.674 2.22482E+06 45.00 1828.902 2.19893E+06 47.00 1852.491 2.22662E+06 49.00 1778.208 2.13866E+06 54.00 1721.858 2.13614E+06 59 00 1683.119 G 2.08738E+06 64.00 1645.275 2.03960E+06 69.00 1603.079 1.98683E+06-74.00 1563.933 1.93779E+06 79.00 1541.591 1.90879E+06
- 84.00 1494.987 1.85138E+06 89.00 1439.882 1.78421E+06 94.00 1398.732 1.73323E+06 99.00 1360.304 1.68543E+06 109.00 1282.444 1.58899E+06 119.00 1209.613 1.49876E+06 129.00 1134.598 1.40618E+06 139.00 1062.887 1.31782E+06 149.00 996.389 1.23580E+06 159.00 926.510 1.14987E+06 169.00 902.824 1.11947E+06
,179.00 864.572 1.07134E+06 189.00 835.283 1.03425E+06 199.00 794.272 9.83388E+06 Amendment G April 30, 1990 -,. , .- . ~
CESSAR E!N?,cui0.
.# TABLE 6.2.1-12 (Cont'd)
(Sheet 3 of 7) MAIN STEAM LINE BREAK, 50% POWER - MSIV FAILURE 2 (8.72 ft Total Area)
-PART A: Mass / Energy Release Data ,
Break Mass th+%t Row gy
- Time Flow Rate 1%~ hte (sec) (1befsec) (Btu /sec) 209.00 750.081 9.28784E+05 219.00 711.852 8.81366E+05 229.00 666.738 8.25829E+05 239.00 620.641 7.59137E+05' 249.00 565.725 7.01867E+05 259.00 520.964 6.46629E+05 269.00 471.749 5.86062E+05 <
279.00. 422.711 5.25682E+05 , 4.76768E+05 l -, . 289.00 383.120
- 299.00 350.538 4.36465E+05 309.00 318.234 3.96612E+05 319.00 289.682 3.61331E+05 329.00 269.449 3.35909E+05 339.00 240.446 3.00146E+05 341.90 235.662 2.78027E+05 349.00 110.273 1.37965E+05 G 399.00 111.539 1.39313E+05 459.00 114.410 1.42677E+05 499.00 111.269 1.38977E+05 i
l l. L i I Amendment G l April 30, 1990
l)! !h h k!k b0N!lCATCM O l TABLE 6.2.1-12 (Cont'd) (Sheet 4 of 7) 1 l NAIN STEAM LINE BREAK, 50% POWER - MSIV FAILURE (8.72 ft 2Total Area) , PART B: Steam Generator Pressures I l l Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 1 0.00 1052.0 1052.0 0.50 1040.5 1032.3 0.99 1023.3 1011.3 1.49 1005.8 991.5 1.99 988.7 972.8 2.49 - 972.3 955.3 2.99 956.6 938.8 3.49 942.0 923.4 3.99 930.7 910.5 4.49 920.8 899.0 4.99- 911.3 888.2 5.49 902.2 878.1 5.99 893.4 868.6 G 6.49 884.9 859.5 6.99 876.7 851.0 ' 7.49 868.9 843.0 7.99 861.5 835.4 8.49 864.1 831.3 8.99 876.9 830.5 9.49 889.5 829.8 9.99 901.9 829.1 10.49 914.0 828.3 10.99 925.8 827.6 11.49 937.3 826.8 11.99 948.5 825.8 12.99 969.2 823.4 13.59 987.6 821.9 14.99 1003.1 819.4 15.99 1015.7 815.2 16.99 1025.3 809.3 17.99 1032.4 802.2 , 18.99 1037.5 794.1 19.99 1040.9 785.6 21.99 1045.1 768.5 Amendment G April 30, 1990
C E S S A R En Wicu e. O TABLE 6.2.1-12 (Cont'd) (Sheet 5 of 7) MAIN STEAM LINE BREAK, 50% PoemR - MSIV FAILURE 2 (8.72 ft Total Area) PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Goerator (sec) Pressure (psia) Pressure (psial 23.99 1048.0 753.0 27.00 1051.5 732.8 29.00 1052.4 721.2 31.00 1051.4 709.8 33.00 1040.3 . 698.4 35.00 1043.8 687.2 l 37.00 1038.4 676.5 39.00 1034.4 666.6 l 41.00 1034.3 657.5 43.00 1034.3 649.1 45.00 1034.3 641.1 47.00 1034.3 633.4 49.00 1034.2 625.9 G 54.00 1034.1 608.3 59.00 1034.0 592.4 64.00 1033.9 577.3 69.00 1033.8 563.0 74.00 1033.7 548.8 79.00 1033.6 533.5 84.00 1033.5 515.6 89.00 1033.3 '499.1 94.00 1033.2 483.9 99.00 1033.1 469.5 l 109.00 1032.8 441.9 119.00 1032.5 415.7 129.00 1032.2 389.7 139.00 1031.9 363.5
- 149.00 1031.5 337.6 L 159.00 1031.2 314.4 169.00 1030.8 303.9 179.00 1030.5 292.2 189.00 1030.1 279.2 199.00 1029.7 264.9 Amendment G April 30, 1990
CESSAR %iMnean* TABLE 6.2.1-12 (Cont'd) e (Sheet 6 of 7) NAIN STEAM LINE BREAK, 50% PotFER - MSIV PAIIRRE - (8.72 ft 2Total Area) PART B: Steam Generator Pressures Unaffected Steam Affected Steam -_ Time Generator Generator (sec) Pressure (psia) Pressure (psia) 7
.. L 209.00 1029.3 249.8-219.00 1028.9 234.0'
~229.00 1028.5 217.6' '
239.00 1028.1 200.7 249.00- 1027.7 183.7 259.00 1027.3 166.5 269.00 1026.9 149.4~ 279.00 1026.4 132.6- 'i 239.00 1026.0 '118.7 299.00 1025.6 107.6 309.00 1025.2- 9 7 .1= g 319.00 1024.7 86.8 329.00 1024.3 76.9 4 339.00 1G23.9 67.2-341.90 1023.7 63.8 349.00 1023.4 60.7
-399.00 1021.4 58.8-459.00 1019.1 56.7 499.00 1017.6 55.5 e
Amer.dment C ' April 30, 1990 l - Md
CESSAR1!Encuiu TABLE 6.2 1-12 (Cont'd)- s (Sheet 7 of 7) { MAIN STEAM LIND BREAK, 50% POWER -- j MSIV FAIIRRE . 5 (8.72 ft 2 Total Area) 1. PART C: Accident Chronology l
= Time -
(secs) Event Setpoint
)
0.00 Break Occurs j 2.10 Containment Pressure Roaches 4.0 psig Reactor Trip Setpoint 2.10 Containment Pressure Reaches 4.0 psig Main. Steam Isolation Signal (MSIS) Analysis Setpoint 3.10 High Containment Pressure Reactor Trip Signal and.MSIS Generated ; 3.'25 Reactor Trip' Breakers Open 3.25 Turbine Admission Valves: Closed 6 5.80 Containment Spray Actuation 10.0 psig- ! Signal-8.25 Main Steam Isolation Valves Closed
-13.25 Main Feedwater Isolation Valves closed 73.80 Containment Spray Reaches Containacnt
-73.80 Peak Containment Temperature 289.80 Peak Contai.nment Pressure Amendment G April 30, 1990
CESSAR !!&ma .
-l TABLE 6.2.1-13 i (Sheet 1 of 9) i MAIN STEAM LINE BREAK, 20% POWER - j LOSS OF ONE CSS TRAIN 2 k (8.72 ft Total Area)
PART A: -Hass/ Energy Release Data' Break Mass Break-Energy Time- Flow Rate Flow Rate i (se":) (lba/sec) (Btu /sec) O.00 9671.991 1.15091E+07-
- 0. 5 0- 9391.411 1.11905E+07 0.99 9160,713 1.09238E+07 1.49 8971.808 1.07051E+07
=1.99 8799.940 1.05057E+07 2.49 8639.088 1.03189E+07 "
.2.99 8486.550 1.01414E+07 9' 3.49 3.99 4.49 8375.439 8266.392 8147.882 1.00121E+07 9.88498E+06 9.74664E+06. i 4.99 8031.627 9.61080E+06 '}
5.49 7916.824 9.47885E+06 5.99 7809.991 9.35141E+06
-6.49 7705.186 9.22856E+06 0 6.99 7604.371 9.11029E+06 7'.49 7507.466 8.99649E+06 ,
7.99 7414.455 8.88718E+06 8.49 2265.754 2.71540E+06 ' 8.99 2253.579 2.70109E+06 9.49 2241.7E2 2.68723E+06 9.99 8230.392 2.67383E+06 10.49 2219.403 2.66091E+06 30.99 2208.790 2.64842E+06 11.49 2198.505 2.63632E+06 11.99 2188.484 2.62452E+06 12.99 2168.931 2.60149E+06 13.99 2153.602 2.58343E+06 14,99 2138.,577 2.56571E+06 15.99 2122.240 2.54645E+06 16.9C 2104.469 2.52548E+06 17.99 2085.376 2.50293E+06 18.99 2065.275 2.47919E+06
~
19 99 2044.601 2.45475E+06 21.99 2003.403 2.40601E+06 23.99 1964.763 2.36027E+06 Amendment G April 30, 1990 l
~ ~ ' '
W.W i y ji
? ,{( f 4 *
\c,' ,-
-' 'e
1 CESSAR E!Encu0. i TABLE 6.2.1-13
'(Sheet 2 of 9)-
MAIN STEAM LINE BREAK, 20% POWER - IDSS OF ONE CSS TRAIN (8.72 ft 2Total Area) PART A: Mass / Energy Release Data #
. a Break Mass Break Energy i Time ~ f Flow Rate Flow Rate (sec) (1bs/sec) (Btu /sec) 27.00 1915.230 2.30155E+06-29.00 1885.294 2.26603E+06 31.00 1856.019 2.23127E+06 33.00" 1826.950 2.'19673E+06 L 35.00 1798.475 2.16286E+06' 37.00 1771.313 2-13053E+06' 39.00 1745.958 2.10033E+06 '
41.00 1722.417 2.07227E+06 - - 43.00 1700.328 2.04593E+06 ' 45.00 1679.243 2.02077E+06 47.00 1658.885 1.99646E+06 L 49.00 1639.222 1.97297E+06 p 54.00 1593.814 1.91868E+06 59.00 1553.496 1.87043E+06 64'.00 1516.350 1.82593E+06 G 69.00 1481.902 1.78463E+06-74.00 6449.827' 1.80073E+06 79.00 1418.520 1.76109E+06 .;
=84.00 1388.141 1.72266E+06 89.00 1359.728 1.68660E+06 L 94.00 1330.973 1.65032E+06 !
L 90.00 1299.721 1.61101E+0C' ! 109.00 1227.277 1.52136E+06 119.00 1160.124 1.43822E+06 129.00 1100.084 1.36357E+06 139.00 1043.936 1.29374E+06 149.00 990.583 1.22744E+36 159.00 938.551 1.16290E+06 : o 169.00 886.477 1.09846E+06 ! l . 179.00 835.586 1.03558E+06 l 189.00 785.955 9.74312E+05 199.00 737.475 9.14506E405 209.00 689.983 8.55949E+05 j
'219.00 643.362 7.98487E+05 L 229.00 596.808 7.41104E+05 Amendment G April 30, 1990
C E S S A R EM M e m .., O TABLE 6.2.1-13 (Sheet 3 of 9) MAIN STEAM LINE BREAK, 20% POWER - IDSS OF ONE CSS TRAIN (8.72 ft Total Area) PART A: hass/ Energy Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1bs/sec) (Bto/sec) 239.00 550.279 6.83722Er05 249.00 505.084 6.27961E+05 259.00 461.318 5.73928E+05 269.00 419.288 5.21995E+05 279.00 379.431 4.72697E+05 289.00 399.112 4.97653E+05 299.00 348.949 4.35207E+05 309.00 345.981 4.31141E+05 319.00 342.613 4.26676E+05 329.00 339.049 4.21935E+05 339.00 335.148 4.16782E+05 349.00 330.722 4.11008E+05 359.00 325.950 4.04814E+05 369.00 320.523 3.97818E+05 G 379.00 314.499 3.90105E+05 389.00 307.806 3.81579E+05 399.00 300.385 3.72172E+05 419.00 283.23b 3.50566E+05 439.00 263.035 3.25275E+05 459.00 241.624 2.98568E+05 479.00 221.046 2.72950E+05 499.00 201.272 2.48378E+05 519.00 183.028 2.25736E+05 539.00 167.054 2.05926E+05 559.00 153.457 1.89072E+05 579.00 142.317 1.75262E+05 599.00 133.669 1.64534E+05 619.00 126.968 1.56214E+05 639.00 121.949 1.49971E+05 659.00 118.240 1.45344E+05 679.00 115.518 1.41935E+05 699.00 113.523 1.39421E+05 9 719.00 739.00 759.00 112.053 110.960 110.133 1.37555E+05 1.36151E+05 1.35077E+05 Amendment G April 30, 1990
.CESSAR n'nne.m.
g TABLE 6.2.1-13 (Sheet 4 of 9) MAIN STEAM LINE BREAK, 20% POWER - IDSS OF ONE CSS TRAIN 2 (8.72 ft Total Area)- PART A: Mass / Energy Release Data l Break Mass Break Energy Ties Flcnr Rate Flow Rate (sec) (lbe/sec) (Btu /sec) 779.00 109.496 1.34236E+05 799.00 108.992 1.'33559E+05' -2 819.00 108.583 1.32998E+05= e 839.00 108.241 1.32520E+05 ,
'859.00 107.946 1.32100E+05 879.00 107.685 1.21721E+05 899.00 107.447 J.31173E+05 919.00- 107.227 1.31044E+05 .
! 939.00 107.017- 1.30730E+05 l 959.00 106.816. 1.30426E+05 979.00 106.619- 1.30128E+05 999.00 106.426 1.29834E+05 3 i O
- . Amendment G April 30, 1990
W CESSAR E!!Gcam,. a' i 4:' . TABLE 6.2.1-13 (Sheet 5 of 9) NAIN STEAN LINE BREAK, 20% POWER - LOSS OF ONE CSS TRAIN (a.72'ft 2Total Area). ! PART B: Steam Generator Pressures Unaffected Steam Affacted Steam Time Generator Generator l (sec) PresGure (psia) Pressure (psia) 0.00 1060.0 1060.0 i 0.50 1047.6 1039,2 } 0.99 1030.5 1017.7 i 1.49 1013.2 997.4 1.99 996.4 978.3 2.49 980.2 960.3 2.99 964.5 943.2 3.49 949.6 927.2 6O 3.99 936.0 912.5 4.49 923.1 898.7 4.99 .910.7 885.5 5.49 898~.6 872.9 G 5.99 886.9 860.9
~6.49 875.5 849.3 6.s9 864.5 .838.2
'7.49 853.9 827.5 7.99 843.7 817.3 8.49 843.4 810.5 8.99 850.3 806.1 9.49 857.2 801.8 9.99 864.0 797.6 10.49 870.8 793.6 10.99 877.5 789.8 11.49 884.1 786.0' 11.99 890.5 782.4 12.99 903.1 775.3 l 13.99 914.8 769.7 14.99 925.4 764.2 15.99 934.8 758.3 16.99 9 /- 2 . 9 751.8 17.99 ? 49.7 744.8 18.99 955.4 737.5 ,
19.99 C60,1 730.0 O' 21.99 'e 6 7 . 4 715.0 i I
.23.99 973.2 700.9 Amendment G l April 30, 1990 l l
C: LCESSAR E!Micue T A B L E 6 .~2 . 1 - ?.3 - (Sheet 6 of 9) MAIN STEAM LINE BREAK, 20% PONER - IDSS OF ONE CSS TRAIN (8.72.ft 2Total Area) PART B: Steam Generator Pressures
' Unaffected Steam Affected Steam Time Generator Generator l (sec) Pressure (psia) Pressure (psia) 27-.00 900.6- 681.5
-29.00 983.9 670.1 31.00 985.6 658.9 33.00 .985.5- 647.8 35.00 984.1 637.0.
37.00 981.8 626.6 39'.'00 978.9 616.9 141.00 975.6 608.0 . 43.00- 974.8 599.6
'45.-00 974.8 591.5 ,
47.00 974.8 583.7 49.00 974.8 576.2 .o ! 54.00 974.7 558.9 59.00 974.7 543.5 .; 64.00 974.7 529.4 !
'69.00 974.6 516.2 74.00- 974.6 504.0 m 79.00 974.5 492.4
'84.00 974.4 481.3 89.00 974.3 471.0 94.00' 974.2 460.5 99.00 974.1 449.1 109.00 973.9 422.7 119.00 973.6 398.3 129.00 973.2 376.5 ;
139.00 972.8 356.0 149.00 972.4 336.6 159.00 972.0 317.7. 169.00 971.5 298.7 179.00 970.9 280.2 189.00 970.4 262.1 199.00 969.8 244.5 209.00 969.2 227.2 219.00 968.5 210.2 229.00 967.9 193.6 ,. Amename'.st G 3 April 30, 1990
- i. _
CESSAR tin %uiu ((() ! TABLE 6.2.1-13 (Sheet 7 of 9) l MAIN STEAM LINE BREAK, 20% POWER - LDSS OF ONE CSS TF1'M ; 2 (8.72 ft Total Area) ! i PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Generator * (sec) Pressure (psia) Pressure:(psia) 239.00- 967.2 177.4 249.00 966.4 161.7 259.00 965.'; 146.5 269.00 964.9 131.9 279.00 964.1 -118.1 289.00 .963,7 103.0. 299.00 962.5 95.6 309.00 961.7 95.3 1 - 319.00 960.9 94.9 329.00 960.0 94.5 , 339.00 959.2 94.0 349.00 958.3 93.5 359.00 957.4 92.9 369.00 956.5 92.2 379.00 955.6 91.4 G 389.00 954.6 90.6 399.00 953.7 89.6 419.00 951.7 87.3 439.00 949.8' 84.6 459.00 947.8 81.8 479.00 945.7 79.1 499.00 943.6 76.6 519.00 941.5 74.4
- 39.00 939.3 72.5 359.00 937.2 70.9 579.00 935.0 69.6 399.00 932.7 68.6 619.00 930.5 67.7 639.00 928.3 67.1 659.00 926.0 66.5 e a 679.00 923.7 66.1 699.00 921.4 65.7 719.00 919.1 65.3 739.00 916.8 65.0 759.00 914.4 64.7 Amendment G April 30, 1990
CESSAR1!nhin TABLE 6.2.1-13
-(Sheet 8 of 9)
MAIN STEAM LINE BREAK, 20% POWER - LOSS OF ONE CSS TRAIN (8.72-ft 2Total-Area) PART B:: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Generator
'(sec) Pressure (psia) _ Pressure (psia) 779.00 912.1 64.4 799.00- 909.7 64.2 819.00 907.3 63.9 839.00 904.9 63.7 859.00 902.5 63.5 879.00 900.1 63.2 G 899.00 897.7 63.0 919.00 895.3 62.8 939.00 892.8 62.6 959.00 890.3 62.4 979.00 887.7 '62.2 999.00 885.4 62.0 O
Amendn.ent G , April 30, 1990 1
' iCESSARiinificam OJ TABLE 6.2.1-13 (Cont'd) >
(Sheet 9 of 9) j MAIN STEAN LINE BREAK, 20% POWER - 1 IDSS OF ONE CSS TRAIN (8.72 ft Total Area) PART C: Accident Chronology Time l' (secs)- Event Sotpoint2 0.00 Break Occurs 11 4 2.09 ' Containment Pressure Reaches -4.0 psig Reactor Trip Setpoint I. 2.09- Containment Pressure Reaches 4.0 psig : Main-Steam Isolation. Signal
; . (MSTS) Analysis Setpoint 3.09- High Containment Pressure Reactor Trip Signal and-MSIS -G Generated ~
3.24 Reactor. Trip Breakers Open 3.24 Turbine Admission Valves Closed 5.80 Containment Spray Actuation 10.0 psig
' Signal '
8.24 Main Steam Isolation Valves Closed 13.24 Main Feedwater Isolation Valves Closed 73.80 Containment Spray Reaches Containment 73.80 Peak Containment Temperature
.485.80 Peak Containment Pressure O
Amendment G April 30, 1990
. . - - . . . . . . . . , , . - . . .~ .. .
-CESSARanLmt y TABLE 6.2.1-14 1 (Sheet 1 of 9)
MAIN STEAM LINE BREAK, 20% A M - -; MSIV FAILURE (8.72 ft 2Total Area) b PART A: Mass / Energy Release Data _ l i Break Mass Break Energy '} - Time Flow Rate Flow Rate (sec) (1ba/sec) (Btu /sec) [ 0.00 9671.991 1.15091E+07 i 0.50 9391.411- 1.11905E+07 ' 0.99 9160.713 1.09238E+07 1.49 8971.808 1.07051E+07 1.99 8799.940 1.05057E+07 ! 2.49 8639.088 1.03169E+07 l 2.99 8486.550 1.01414E+07 i
-3.49 8375.439 1.00121E+07 ;
3.99 8266.392 9.88498E+06 4.49 '8147.882 9.74664E+06- , 4.99 8031.627 9.61080E+06 5.49 7918.824 9.47885E+06- ! 5.99 7809.991 9.35141E+06-6.49 7705.186- 9.22856E+06 6.99 7604.371 9.11029E+06 7.49 7507.466 8.99649E+06 G' d 7.99 7414.455 8.88718E+06 , 8.49 7042.827 8.44855E+06 8.99' 6494.009 7.78581E+06 9.49 5587.473 7.19411E+06 9.99 5546.586 6.66668E+06 10.49 5155.313 6.19717E+06 10.99 4808.124 5.77956E+06 11.49 4500.059 5.40836E+06 11.99 4226.591 5.07844E+06 12.99 3763.820 4.51963E+06 13.99 3399.607 4.07971E+06 14.99 3110.934 3.73137E+06 15.99 2879.588 3.45274E+06 16.99 1692.361 3.22777E+06 17.99 2539.285 3.04430E+06 18.99 2413.160 2.89352E+06 19.99 2308.852 2.76909E+06 21.99 2153.302 2.58400E+06 23.99 2053.015 2.46492E+06 Amendment G April 30, 1990
C E S S A R Hi D. cam . - t i TABLE 6.2.1-14 " (Sheet 2 of 9) h MAIN STEAM LINE BREAK, 20% POWER - s MSIV FAILURE ., 1 2 (8.72'ft ' Total Area) PART A: Mass / Energy Release Data- f Break Mass Break Energy Time Flow Rate. Flow Rate (sec) (Ibe/sec) (Btu /sec) U7.00 1960.141 2.35476E+06 29.00 1919.153 2.30613E+06 l
.31'00
. 1885.430 2.26609E+06 33.00' 1854.320 2.22912E+06 - t 35.00 1825.055 2.19431E+06 ('
37.00 1796.643 2.16049E+06' 39.00 1769.837 2.12856E+06- . 41.00 1770.745 2.12935E+06 ' 43.00 1708.059 2.05506E+06 45.00 1682.601 2.02473E+06-47.00 1663.721 2.00217E+06 49.00 1643.013- 1.97745E+06
'54.00 1596.969 1.92241E+06-59.00 1556.924 1.87448E+06
-64.00 1519.206 1.82930E+06' G 69.00 1480.008 1.78234E+06 74.00- 1451.772 1.80302E+06 79.00 1431.852 1.77683E+06 ;
84.00 1395.447 1.73127E+06 89.00 1359.310 1.68610E+06 94.00 1332.467 1.65208E+06-99.00 1301.965 1.61365E+06 109.00 1225.730 1.51949E+06 119.00 1158.083 1.43576E+06 4 129.00 1098.449 1.36160E+06 j 139.00 1042.078 1.29151E+06 149.00 992.406 1.22959E+06 -
- 159.00 941.000 1.16578E+06 169.00 883.44? 1.09482E+06 .!
179.00 838.657 1.03920E+06-189.00 788.194 9.76951E+05 199.00 738.172' 9.15327E+05'
.209.00 692.358 8.58747E+05- .
219.00- 644.828 8.00214E+05 229.00 600.124 7.45011E+05 Amendment G April 30, 1990 l e
CESSARiinh a O L TABLE 6.2.1 , (Sheet _3 of 9) MAIN STEAM LINE-BREAK, 20% POWER - MSIV FAILURE (8.72 ft Total Area) PART A: Mass / Energy Release Data Break Mass Break Energy
. Time' Flow Rate Flcnr Rate-(sec)' (lba/sec) (Btu /sec)_
239.00 548.014 6.81012E+05 249.00- 509.778 6.33491E+05 259.00 462.827 5.75706E+05 , 269.00 417.620 5.20008E+05 279.00- 381.067 . 4.74623E+05 289.00 344.173 4.28915E+05 299.00 318.549 3.97061E+05 309.00 313.322 3.90649E+05
\ .319.00 316.347 3.93911E+05 329.00' 312.435 3.88959E+05 339.00 309.795- 3.85530E+05 349.00 310.191- 3.85690E+05 359'.00. 312.804 3.88448E+05 369.00 308.120 3.82598E+05 ;
379.00 308.533 1 3.82752E+05 389.00 304.511 3.77665E+05 G l
=
399.00 303~.940 3.76642E+05 419'.00 295.814 3.66320E+05
'439.00 290.183 3.58891E+05 ;
459.00 280.424 3.46513E+0b ! 479.00 270.386 3.33777E+05 , ! 499.00 250.887 3.09884E+05 519.00 236.847 .2.92428E+05 539.00 228.043 2.81157E+05-559.00 176.357 2.16942E+05 579.00 140.015 1.71924E+05 599.00 122.887 1.50708E+05 619.00f .115.689 1.41782E+05 639.00 111.613 1.36699E+05 659.00 112.954 1.3832SE+05 679.00 109.495 1.33981E+05 699.00 110.484 1.35193E+05 719.00 109.506 1.33934E+05 O 739.00 759.00 111.239 109.160 1.36011E+05 1.33461E+05 i Amendment G April 30, s990
t CESSAR1!ni?,c.m p t- ' h .
' TABLE 6.2.1-14 (Sheet 4 of19)
MAIN STEAM LINE BREAK, 20% POWER - , MSIV FAILURE l (8'72.ft
. Total Area)
PART_A: . Mass / Energy. Release Data Break Mass Break Energy l Time Flow Rate Flow. Rate
-(sec) (lbm/sec) (Btu /sec) .
G ! 779.00 110.779 1.35398E+C5
'799.00 112,550 1.37506E+05 .!
W 819.00 111.131 1.35733Ff'r05 L 839.00 111.349 1.35883E+05 ,
, 859.00 113.333 1.38359E+05 ~'
879.00 111.995 1.36684E+05 899.00 111.717 1.36254E+05 919.00 111.912' 1.36382E+05 939.00. 108.888 1.32736E+05 1 L 959.00 112.581 1.37191E+05 l: 979.00 107.608 1.31045E+05- ; 999~.00 109.852 1.33796E+05 f l l l L 1 O j' Yg l\ , Amendment C W April 30, 1990 , te
LCESSAR2!Knema l
?
i TABLE 6.2.1-14 (Cont'd) , (Sheet 5 of 9) MAIN: STEAM D %d BREAK, 20% POWER - MSIV FAILURE ) (8.72 ft 2Total Area) . PART B:- Steam' Generator Pressures i Unaffected steam Af cted Steam Time Generator snarator (sec)- Pressure (psia) Pressure'(peia). 0.00 1060.0 1060.0 J0.50 1047.6 1039.2-0.99 1030.5 1017.7 1.49 1013.2 997.4 1.99 996.4 978.3 2.49 980.2 960.3 2.99 964.5 943.2 3.49 949.6 927.2 O- 3.99 936.0 912.5 4.49- 923.1 898.7_ ,1 4.99 910.7 885.5 J5.49 898.6 -872.9 5.99 886.9 860.9 G 6.49 875.5 849.3 6.99 864.5 838.2 7.49 853.9 827.5 7.99' 843.7 817.3 8.49 843.4 '810.5 8.99 850.3 806.1 9.49 857.2 801.8 9.99 364.0 797.6 10.49 870.8 793.6 10.99 877.5 789.8 11.49 884.1 786.0 11.99 890.5 782.4 l 12.99 903.1 775.3 13.99 914.8 769.7 14.99 925.4 764.2 15.99 934.8 758.3 16.99 942.9 751.8 l 17.99 949.7 744.8 i
,, 18.99 955.4 737.5 19.99 960.1 730.0 i (m, 21.99 967.4 715.0 l 23.99 973.2 700.9 )
l Amendment G i April 30, 1990 1
- id
. _ _ . . - . . - ~ - _ . . , .
3 TCESSAR E!!fi?icamn O TABLE 6.2.1-14-(Cont'd) t J 4 (Sheet 6 of 9). MAIN STEAM LINE' BREAK, 20%~ POWER -
, MSIV FAILURE (8.72'ft 2Total' Area) '
PART B:- Steam Generator Pressures "i ' Unaffected Steam. Affected Stream Time Generator- Generator
;, (sec) Pressure (psia) Pressure (psia) -'
~27.00 980.6 681.5 H 29.00~ 983.9 670.1 1 31.00 985.6 658.9 33.00 985.5 647.8 35.00 984.1 637.0 37.00- 981.8 626.6 j 39.00 978.9 616.9- -- d
- u. <41.00 975.6 608.0 >
L 43.00 974.8 599., 45.00 974.8 591.5 47.00 974.8 583.7' 49.00~ 974.8 576.2 j 54.00. 974.7 -558.9 9 59.00 974.7 543.5 a
- b. 64.00 974.7 529.4 g 69.00- 974.6 516.2 i 74.00 974.6 504.0 79.00 974.5 492.4 84.00 974.4 481.3 89.00 974.3 471.0 94.00 974.2 460.5 !
99.00 974.1' 449.1 109.03 973.9 422.7 119.00 973.6 398.3 L 129.00 973.2 376.5 139.00 972.8 356.0 149.00 972.4 336.6 l 159.00 972.0' 317.7 l; 169.00 971.5 298.7 L
-4 179.00 970.9 280.2 4 l ' 189.00 970.4 262.1 ,
199.00 969.8 244.5 > 209'.00- 969.2 227.2 219.00 968.5 210.2 , 229.00 967.9 193.6 Amendment G April 30, 1990 n
, , , , _ .w
m LCESSAR;!nL m. J () t
, TABLE 6.2.1-14 (Cont'd)_
(Sheet 7 of 9) MAIN STEAM LINE BREAK, 20% POWER - ' MSIV FAILURE 2 (8.'72 ft Total Area)' PART B: Steam Generator Pressures . P
^
Unaffacted Steam Affacted Steam: Time Generator Generator (sec) Pressure (psia) Pressure (psia) i 239.00 967.2 '177.4 249.00 966.4 161.7 259.00 965.7 146.5 269.00 964.9 131.9 279.00 964.1 118.1 289.00 963.3 105.1 299.00 962.5 95.8 5
, 309.00 961.7 95.5 319.00 960.9 95.2 329.00 960.0 -94.9 339.00 .959.2 94.5-349.00 958.3 94.1 i
'359.00 957.4 93.6 G 369.00 956.5 93.0 379.00- 955.6 92.4 4 389.00 954.7 91.6 399.00 953.7 90.7 419.00 951.8 8 18 . 5 439.00 949.9 85.0
-459.01 947.9 81.5 479.VJ 945.8 76.9 i 499.00 943.7 71.8' 519.00 941.6 66.4 539.00 939.5 60,9 559;00 937.3 57 4-579.00 935.0 56.7 599.00 932.8 56.1 619.00 930.5 55.5 639.00 928.2 54.9-659.00 925.9 5 <4 . 3 679.00 923.6 53.7 699.00 921.3 53.2
/* 719.00. 918.9 52.7 l s,) 739.00 916.6 52.1 759.00 914.2 51.6 x
Amendment G April 30, 1990
.l 1
CESS 6hFPL m<
. u. ;
TABLE 6.2.1-14 (Cont'd) . (Sheet 8 of 9)- f NAIN STEAM LINE BREAK, 20% POWER - MSIV FAILURE ! 2 (8.72 ft Total Area) PART B: -Steam Generator Pressures Unaffected Steam Affected Steam Time Generator =Generatur (sec) Pressure (psia) Pressure (psia) G 779.00 911.9' -51.1 > 799.00 909.5 50.7 819.00 907.1 50.2' 839.00 904.7 49.7 859.00 902.2 49.3 879.00 899.8 .48.9 899.00 897.3 48.5 - 919.00 894.9 48.1 939.00 892.4 47.7-959.00 889.9 47.3 979.00 887.4 47.0 999.00 884.9 46.6 l: O Amendment G
. April 30, 1990 l
CESSAR kHHncoisa i TABLE 6.2.1-14 (Cont'd) ! (Sheet 9 of 9) NAIN STEAM LINE BREAK, 20% POWER:- MSIV FAILURE (8.72 ft Total Area) PART C: Accident Chronolog! I Time (secal Event , Setpoint i 0.00. Break Occurs 2.09 Containment Pressure Reaches 4.0 psig Reactor Trip Setpoint' 2.09 Containment Pressure Reaches 4.0 psig-Main Steam Isolation Signal (MSIS) Analysis Setpoint-O- 3.09 High Containment Pressure Reactor Trip Signal and MSIS Generated-3.24 Reactor Trip Breakers Open 3.24 Turbine Admission Valves Closed G 5.80 Containment Spray Actuation 10.0 psig ( Signal 8.24 Main Steam Isolation Valves Closed 13.24 Main Feedwater Isolation Valves ( Closed 73.80 Containment Spray Reaches Containment 73.80 Peak Containment Temperature 282.80 Peak Containment Pressure IO Amendment G April 30, 1990
.s-
i
- CESSAR iininc== l a
TABLE 6.2.1-15 (Sheet 1 of.7) MAIN STEAM LINE LREAK, 0% POWER -- IDSS OF ONE CSS TRAIN 2 (4.50 ft Total Area) ,
=PART A: Mass / Energy Release Data ,
f 1 Break Mass Bruak Eriergy Time Flow Rate Flow Rate : (sec) {lba/sec) (Btu /sec) _ 1 0.00- 7987.046 9.49130E+06 0.50 7784.664 9.25871E+06 0.99 7664.886 9.12127E+06 1.49 7556.181 8.99637E+06 1.99 7453.672 8.87839E+06 2.49 7355.946 8.76574E+06
- 2.99- 7262.413 8.65777E+06 8.55351E+06 O' 3.49 3.99-7172.224 7085.553 8.45319E+06 4.49 7002.546 8.35700E+06
'4.99 6923.242 8.26500E+06 ;
5.49 6847.639 8.17721E+06 5.99 6775.672 8.09357E+06 6.49 6708.401 8.01534E+06 6.99 6646.164 7.94176E+06 7.87211E+06 G 7.49 6585.344 7.99 6528.788 7.80620E+06 8.49 6475.588 7.7441/E+06
- 8.99 2486.470 2.97379E+06 'i 9.49 2474.712 2.96007E+C6 .;
9.99 2463.511 2'.94699E+C6 10.49 2452.870 2.93457E+06 10.99 2442.756 2.92275E+06 11.49 2433.097 2.91147E+06
'11.99 2423.788 2.90059E+06 12.99 2405.682 2.87942E+06
-13.99 2388.044 2.85878E+06 14.99 2370.374 2.93810E+06 15.99 2350.450 2.81476E+06 l
16.99- 2328.032 2.78849E+06 17.99 2303.252 a.75942E+06 18.99 2276.553 2.72808E+06 -
, . 19.99 2248.588 2.69523E+06 l - % . 21'.99 .!191.823 2.62845E+06 j .8 Amendment G April'30, 1990 1 _
y. u CESSAR !!nLmr 1
;e;
't TABLE 6.2.1-15 (Cont'd)
(Sheet 2 of 7) MAIN STEAN LINE BREAK, 0% POWER - ILSS OF ONE CSS TRAIN 2 (4.50 ft Total Area) PART A: Mass / Energy Release Data-i Break Mass Ereak Energy Time Flow Rate Flow Rate
-(sec) - (1be/sec) (Btu /nec) m ~
23.99 '2138.32d 2.56541E+06 27.00 2070.131 2.48493E+06 y .29.00 2033.965 2.44217E+06 t 31.00 2003.034 2.40558E+06 33.00 1975.205 J.37262E+06-35.00 1949.348 -2.34199E+06-37.00 1924.462 2,31250E+06- ' 39.00 1899.778 '2 . 2 83 2 2 E+0 6 - 2.25453E+06-41.00- 1875.SO4 -
-43.00 1852.3LJ 2.226920+06
- 45.00 1830.309 2.20072E+06-47.00 1809.529 2.17601E+06 49.00 1789.866'- 2 '.152 62 E+06 G 54.00 1743.920 2.09790E+06 59.00 1700.400 2.04602E+06 64.00 1658.988 1.99,659E+06 69.00 1619.930 1.94992E+06 74.00 1582.755 1.90545E+06 79.00 1547.121 '
1.86280E+06 84.00 s 1512.851 1.8586EE+06 89.00 1479.775 1.81747E+06 94.00 1448.068 1.77792E+06 i 99.00 1415.729 1.73769E+06 109.00 1348.723 1.65472E+06 119.00 1280.936 1.57139E+06 129.00 1221.280 1.49736E+06 139.00 1165.418 1.42821E+06 149.00 1112.931 1.36320E+06 159.00 1062.691 1.30110E+06 169.00 1013.930 1.24094E+06 179.00 966.837 1.18296E+06 189.00 ' 920.783 '1.12637E+06 199.00 875.093 1.07039E+06
; - Amendment G
.I April 30, 1990 i x
s 4CkSSARBiEc.m " 3 l :. 1 f
.,. f TABLE 6.2.1-15l(Cont'd) 1
-p (Sheet 3lof 7)- q MAIN STEAM LINE BREAK, 0% POWER -
IDSS OF ONE CSS TRAIN l 2 3 (4.50 ft Total Area) PART A:; Mass, Energy Release Data
+ Break Mass: Break Energy -]
Time. Flcw Rate Flow Rate l
! (sec} 41L /sec) (Btu /sec) !
l 209.00' 828.744 1.01378E+05 4 l 219.00 780.286. 9.54806E+05 229.00 727.490 8.90802E+05 ,
../1 239.00 657.032 8.17745E+05 i
; ;249.00 594.198 7.29823E+05
.l a =259.00 528.239 6.49906E+05 l- ~
269.00 452,.056 5.39475E+05 274.90j 375.491 4.48103E+05 l l- ' 289.00' 111.325 1.38487E+05-. I
,359.00 110.508- s 1.37472E+05
- 4 .
$399.00 110.532 1.37501E+05 g
~
459.00 110.495 1.36452E+05 'l 499.00 110.473 1.37429E+05 G l
* , j a,
.l H
\ +
l
,y, 1 M
M , i t l ll ' ' 1 i s'; Anendment G. )
/ ,- April 30,'1990 l
.\, e .
- ,a.; .d'
CESSAR !!nificma O TABLE 6.2.1-15 (Cont'd) (Sheet 4 of 7) MAIN STEAM LINE BREAK, 0% POWER - IDSS OF ONE CSS TRAIN (4.50 ft Total Area) PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 0.00 1100.0 1100.0 0.50 1090.4 1081.3 0.99 1077.9 1063.4 1.49 1065.3 1047.1 1.99 1052.9 1032.0 2.49 1040.7 1017.9 2.99 1028.7 1004.6 3.49 1017.0 991.9 3.99 1005.7 979.9 4.49 994.7 968.5 .. 4.99 984.1 957.7 [ 5.49 973.9 947.? 5.99 964.2 r
.o 6.49 955.0 9a8.3 g 6.99 946.2 919.6 7.49 937.9 911.3 7.99 930.0 903.5 8.49 922.5 896.1 8.99 924.7 890.9 9.49 931.0 886.6 9.99 937.0 882.5 10.49 942.9 878.6 10.99 948.7 875.0 11.49 954.4 871.4 11.99 960.0 868.1 12.99 970.7 861.5 13.99 980.3 855.0 l 14.99 988.4 848.6 15.99 994.6 841.3 16.99 998.9 833.2
, : 17.99 1001.3 824.1 18.99 1002.0 814.4 19.99 1001.3 804.2 4 I f ; Amendmemt G l -
- April 30, 1590 1 ..
3 WW M
CESSAR1!nL m. 1 ((() 1 TABLE 6.2.1-15 (Cont'd) (Sheet 5 of 7) i NAIN STEAM LINE BREAK, 0% POWER - LOSS OF ONE CSS TRAIN 2 (4.50 ft Total Area)' ! PART B: Steam Generator Pressures I t'Dffacted Steam Affacted Steaa ; Time Generator Generator (sec) Pressure (psia) Pressure (psia) i 21.99 997.0 783.6 23.99 991.1 764.1 j 27.00 982.3 738.7 e 29.00 977.7 725.6 31.00 973.4 714.4 i 33.00 973.4 704.3 , 35.00 973.4 694.7 l O 37.00 39.00 41.00 973.4 973.4 973.3 685.2 675.7 666.5 43.00 973.3 657.7 45.00 973.3 649.3 47.00 973.3 641.3 i 49.00 973.3 633.8 G 54.00 973.2 616.3 59.00 973.2 599.7 ! 64.00 973.1 583.9 69.00 973.0 568.9- ! 74.00 972.9 554.8 79.00 972.8 541.1 84.00 972.7 528.0 89.00 972.5 515.4 94.00 972.4 503.3 y 99.00 972.2 49,1.4 109.00 971.8 466.9 119.00 971.4 442.3 129.00 970.9 420.6 139.00 970.3 400.3 149.00 969.7 381.2 159.00 969.1 362.9 169.00 968.4 345.1 179.00 967.7 328.0 Amendment G April 30, 1990 n,
{
'CESSAR1!L"ic m. -
TABLE 6.2.1-15 (Cont'd) (Sheet 6 of 7) ,. MAIN STEAM LINE BREAK, 0% POWER - i LOSS OF ONE CSS TRAIN 2 (4.50 ft Total-Area) PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Generator Pressure (psia') (sec) Pressure (psia) 189.00 967.0 311.2 199.00 966.2 294.6 209.00 965.3 277.7 219.00 964.4 260.1 1 229.00 963.5 240.8 I 239.00 962.6- 218.8 j 249.00 961.6 192.6 ; 259.00 960.6 169.7 269.00 959.6 143.1 - 274.90- 959.0 116.5 289.00 957.5 62.8 359.00 949.9 61.3 399.00 945.6 60.6 459.00 939.2 59.7 499.00 935.0 59.1 o l I e I L l Amendment G April-30, 1990
1 CESSARsi!b m. i i TABLE 6.2.1-15 (Cont'd) ! (Sheet 7 of 7)' i MAIN STEAM LINE BREAK, 04 POWER - LOSS OF ONE CSS TRAIN (4'.50 ft2Total Area) i PART C: Accident Chronology Time (secs) Event Setpoint. P 0.00 Break Occurs 2.50 Containment Pressure Reaches 4.0 psig Reactor Trip Setpoint-2.50 Containment Pressure Reaches 4.0 psig ; Main Steam Isolation Signal (MSIS) Analysis Setpoint 3.50 High Containment Pressure Reactor Trip Signal and MSIS Generated ., 1 G 3.65 Reactor Trip Breakers Open 3.65 Turbine Admission Valves Closed ! 6.80 Containment Spray Actuation 10.0 psig Signal 8.65 Main Steam Isolation Valves Closed l L 13.65 Main Feedwater Isolation Valves Closed l-74.80 Containment Spray Reaches Containment 74.80 Peak Containment Temperature 279.80 Peak Containment Pressure (~) l
'Q Amendment G April 30, 1990 l
m LCESSAR inW'icuiw I i%/-
' TABLE 6.2.1-16 (Sheet 1 of 7) l MAIN STEAM LINE BREAK, 04 POWER - l MSIV FAILURE 2
(4.50 ft Total Area) j PART A: Mass / Energy Release Data 1 Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1bs/sec) (Btu /sec) 0.00 7987.046 9.49130E+06 ! O.50 7784.664 9.25871E+06 0.99 7664.886 9.12127E+06 1.49 7556.181 8.99637E+06 1.99 7453.672 8.87839E+06 2.49 7355.946 8.76574E+06 2.99 7262.413 8.65777E+06
, 3.49 7172.224 8.55351E+06
\m- 3.99 7085.553 8.45319E+06 4.49 7002.546 8.35700E+06 4.99 6923.242 8.26500E+06 5.49 6847.639 8.17721E+06 5.99 6775.672 8.09357E+06 6.49 6708.401 8.01534E+06 6.99 6645.164 7.94176E+06 G 7.49 6585.344 7.87211E+06 7.99 6528.788 7.80620E+06 8.49 6475.588 7.74417E+06 8.99 6211.672 7.43426E+06 9.49 5864.109 -7.02438E+06 9.99 5545.891 6.64759E+06 10.49 5254.751 6.30167E+06 ,
10.99 4988.562 5.98447E+06 11.49 4740.247 5.68781E+06 11.99 4512.834 5.41556E+06 12.99 4116.068 4.93953E+06 s 13.99 3785.635 4.54231E+06 14.99 3510.053 4.21071E+06 15.99 3278.870 3.93256E+06 , 16.99 3084.221 3.69853E+06 l 17.99 2919.758 3.50101E+06 L 18.99 2780.404 3.33387E+06 L 19.99 2662.017 3.19207E+06 21.99 2475.300 2.96884E+06 Amendment G April 30, 1990
.CESSAR ;! Enc m.,
I TABLE 6.2.1-16 (Cont'd)- (Sheet 2 of 7) l NAIN STEAM LINE BREAK, 0% POWER - ! MSIV FAILURE (4.50 ft Total Area) PART.A: Mass / Energy Release Data Break Mass Break Energy Time. Flow Rate Flow Rate (sec) (lbm/sec) (Btu /sec) 23.99 2338.521 2.80564E+06 ' 27.00 2195.215 2.63490E+06 29.00 2128.254 2.55518E+06 31.00 2074.889 2.49166E+06 33.00' 2032.564 2.44132E+06 35.00 1997.710 2.39990E+06 37.00 1967.568 2.36411E+06 39.00. 1939.600 2.33089E+0C ' 41.00 1913.512 2.29990E+06 43.00 1888.545 2.27023E+06 45.00 1865.575 2.24292E+06 47.00 1842.965 2.21601E+06 I 49.00 1820.713- 2.18951E+06 54.00 1772.543 2.13213E+06 G 59.00 1726.553 2.07727E+06 64.00 1683.775 2.02620E+06 69.00 1643.376 1.97792E+06-74.00 1605.346 1.93243E+06 ! 79.00 1568.666 1.88851E+06 84.00 1533.467 1.88328E+06 89.00 1500.474- 1.84215E+06 94.00 1467.132 1.80065E+06 99.00 1435.135 1.76082E+06 109.00 1369.927 1.67997E+06 119.00 1301.065 1.59535E+06 129.00 1239.519 1.51905E+06 139.00 1182.943 1.44903E+06 149.00 1159.393 1.41829E+06 159.00 1068.305 1.30775E+06 169.00 1018.295 1.24611E+06 179.00 971.993 1.18906E+06 189.00 925.817 1.13233E+06 199.00 880.184 1.07641E+06 Amendment G April 30, 1990
CESSARJ!nhn. O TABLE 6.2.1-16 (Cont'd) (Sheet 3 of 7) l MAIN STEAM LINE BREAK, 0% POWER - .i MSIV FAILURE ; (4.50 ft 2Total Area) PART A: Mass / Energy Release Data ; Break Mass Break Energy , Time Flow Rate Flow Rate I (sec) (lba/sec) (Btu /sec) 209.00 834.029 1.02003E+06 G (' 219.00 785.824 9.61357E+05 229.00 733.393 8.97783E+05 239.00 673.969 8.25946E+05~ 249.00 601.703 7.38691E+05 } 259.00 535.056 6.57958E+05 l 269.00 459.430 5.66033E+05 O 274.90 289.00 359.00 382.191 112.175 111.141 4.56006E+05 1.39542E+05' 1.38259E+05 399.00 111.007 1.38092E+05 459.00 110.868 1.37920E+05 499.00 110.820 1.37860E+05-O Amendment G April 30, 1990
CESSARiEricue e; TABLE 6.2.1-16 (Cont'd) l (Sheet 4 of 7) MAIN STEAM LINE BREAK, 04 POWER - MSIV. FAILURE (4.50 ft Total Area)_ PART B: Steam Generator Pressures Unaffected Sty w Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 0.00 1100.00 1100.0 . 0.50 1090.4 1081.3 0.99 1077.9 1063.4 1.49 1065.3 1047.1 1.99 1052.9 1032.0 2.49 1040.7 1017.9 2.99 1028.7 1004.6 3.49 1017.0 991.9 3.99 1005.7 979.9 4.49 994.7 968.5 4.99' 984.1 957.7 5.49 973.9 947.3 5.99 964.2 937.6 6.49 955.0 928.3 G 6.99 946.2 919.6 7.49 937.9 911.3 7.99 930.0 903.5 8.49 922.5 896.1 8.99 924.7 890.9 9.49 931.0 886.6 9.99- 937.0 882.5 10.49 942.9 878.6 10.99 948.7 875.0 11.49- 954.4 871.4 11.99 960.0 868.1 12.99 970.7 861.5 13.99 980.3 855.0 14.99 988.4 848.6 15.99 994.6 841.3 16.99 998.9 833.2 17.99 1001.3 824.1 18.99 1002.0 814.4 19.99 1003.3 804.2 21.99 997.0 783.6 Amendment G April 30, 1990
CESSAR Enc.m., TABLE 6.2.1-16 (Cont'd) (Sneet 5 of 7). ; MAIN STEAN LINE BREAK, 0% POWER - MSIV FAILURE (4.50 ft Total Area) PART B: Steam Generator Pressures Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 23.99 991.1 764.1 27.00 982.3 738.7 29.00 977.7 725.6 31.00 973.4 714.4 33.00 973.4 704.3 35.00 973.4 694.7 L () 37.00 973.4 685.2 39.00 973.4 675.7 41.00 973.3 666.5 L 43.00 973.3 -657.7 45.00 973.3 649.3 47.00 973.3 641.3 ; 49.00 973.3 633.8 54.00 973.2 616.3 G 59.00 973.2 599.7 64.00 973.1 583.9 69.00 973.0 568.9 74.00 972.9 554.8 79.00 972.8 b41.1 84.00 972.7 528.0 89.00 972.5 515.4 94.00 972.4 503.3 99.00 972.2 491.4
- 109.00 971.8 466.9
! 119.00 971.4 442.3
- 129.00' 970.9 420.6 L 139.00 970.3 400.3 149.00 969.7 381.2 159.00 969.1 362.9 169.00 968.4 345.1 179.00 967.7 328.0 189.00 967.0 311.2 966.2 294.6
] 199.00 Amendment G April 30, 1990
CESSAR;IMb c. TABLE-6.2.1-16 (Cont'd) ; (Sheet 6 of 7) , MAIN STEAM LINE BREAK, 0% POWER - MSIV FAILURE (4.50 ft 2Total Area) PART B: Steam Generator Pressures i Unaffected Steam Affected Steam Time Generator Generator (sec) Pressure (psia) Pressure (psia) 209.00 965.3 277.7 219.00' 964.4 260.1 229.00 963.5 240.8 239.00 962.6 218.8 249.00 961.6 192.6 g , 259.00 960.6 169.7 269.00 959.6 14 3 .1~ f 274.90 959.0 116.5 289.00 957.5 62.2 359.00 949.9 59.3 399.00 945.6 57.8 459.00 939.2 55.9 499.00 935.0 54.7 1 i O Amendment G April 30, 1990
.?
CESSARcEnca. L([3) TABLE 6.2.1-16'(Cont'd) i
-(Sheet 7 of 7)
MAIN liTEAM LINE BREAK,-0% POWER - MSIV FAILURE 2 (8.72 ft Total Area) ; PART C: Accident Chronology l Time (secs) Event Setpoint 0.00 Break Occurs
-2.50 Containment Pressure Reaches 4.0 psig Reactor Trip Setpoint 2.50 Containment Pressure Reaches- 4.0 psig Main Steam Isolation Signal f
(MSIS) Analysis Setpoint 3.50- High Containment Pressure 4 Reactor Trip Signal and MSIS I Generated 3.65 Reactor Trip Breakers Open i 3.65 Turbine Admission Valves Closed 1 6.80 Containment Spray Actuation 10.0 psig l Signal l 8.65 Main Steam Isolation Valves Closed . 13.65 Main Feedwater Isolation Valves Closed l 74.80 Containment Spray Reaches Containment l l 74.00 Peak Containment Temperature ' 279.80 Peak Containment Pressure Amendment G April 30, 1990
O O O TABLE 6.2.1-17
- (Sheet 1 of 2)
SIM WtY RESULTS OF POSTULATED PIPE IluPTURE ANALYSIS A. PEAK CONTAllSIENT PRESSURE AfE TEMPERATURE FOR LOSS-OF-000UWIT ACCIDEllTS DESLSI *) DESLS DEDLS(b) g Break Location -Max. SIS - -Min. SIS -Max. SIS -Min. SIS DEHLS ICI 2 Break area (total), ft 9.82 9.82 9.82 9.82 19.2 G . Peak pressure, psig 44.54 44.96 44.15 39.97 45.60 Peak temperature, "F 267 268 267 260 269 Time to peak pressure, sec. 109 110 324 17 14 Energy release to containment 508 512 603 405 443 i atmosphereuptoghetimeof peak pressure, 10 Btu l l l l i i i Amendment G April 30, 1990 ( -
'd w TABLE 6.2.1-17 (Cont'd) (Sheet 2 of 2) SUMARY RESULTS OF POSTULATED PIPE RUPTURE ANALYSIS B. PEAK CONTAIMENT PRESSURE AM TEMPERATURE FOR MAIN STEM LINE BREAKS At 102% At 102% At 50% At 505 At 205 At 205 At OE ' At 95 : Power - Power - Power - Power - Power - Power'- Power - Peuer - Break Location CSS Fail MSIV Fail CSS Fail MSIV Fail CSS Fall MSIV Fall CSS Fall MSIV Fall 2 4.50 Break area (total), ft 8.72 8.72 8.72 8.72 8.72 8.72 4.50 Peak pressure, psig 48.04 47.59 48.30 46.92 47.42 44.35 48.34 47.89 - Time to peak pressure, 268 247 322 290 486 283 280 280 sec. Peak temperature, *F 396.85 405.71 385.10 394.74 372.90 383.37 372.06 382.97 Time to peak temperature, 74 74 74 74 74 74 75 75 sec.
, Energy release to 467 485 485 498 521 471 472 501 containment up to the tigeofpeakpressure, 10 Btu NOTES: a. DESLS: Double-ended suction line slot break
- b. DEDLS: Double-ended discharge line slot break
- c. DEHLS: Double-ended hot leg _ slot break Amendment..G G G_ April 30,.1
- - - - - - . - - - --. = - - -- - -- , --
CESSAR HE"icam. ,
'l l '/ ,
l hJ- ' TABLE 6.2.1-18 INITIAL ColeITIONS FOR C0KfAllMENT PEAK PRESSURE ANALYSIS Parameter Value _ Reactor Coolant System l Reactor power level, MWt* 3876 j Average coolant temperature, *F 586 ' Mass of reactor coolant system-liquid, lbe 647,151 Mass of reactor coolant system steam, lbm 8263-Reactor coolant system liquid plus steam,106 Btu ** 391.194 Steam generatog energy from feed nozzle to MSIV** 271.700 ; (perunit),10 Btu Containment
- Pressure (maximum), psia 15.1 Temperature, 'F 110 Relative humidity, %- 10 o Component cooling water temperature, *F 120
/ IRWST water temperature, "F 110 l i Secondary containment temperature, 'F N/A Outside air temperature, 'F 6 3 NA Net free volume (minimum), 10 ft 3.377 Stored Water (as applicable)
IRWST, gal 495,000 3 Safety injection tanks, ft 7708 ' t u
- At full power plus 2% uncertainty L ** Energy is relative to 32*F -
t 1 l O l l Amendment G April 30, 1990
CESSAR Sinacma
;f%.
V 4 TABLE 6.2.1-19 (Sheet 1 of 2) l ESF SYSTEMS PARAMETERS FOR CONTATIMENT PEAK PRE 550RE ANKEYSIS l Value Used for Peak Pressure System / Item Full Capacity Analyses
- Passive Safety Injection System
[ Number of accumulators (safety 4 4 I injection tanks) i Pressuresgtpoint,'psig 570-632 600~ l Volume, ft / accumulator i Maximum- 1927 1927 Minimum 1600 Active Safety Injection System Number of trains 1/2* Number of pumps / train 2 2 Flowrate, gpm/ train 0 0 psia -1219 1219 Maximum 1232 1232 Minimum 980 C 980 Containment Spray System Number of lines 2 1 Number of pumps 2 1 Number of headers 2 1 Flowrate, gpm/ pump Minimum 5000 5000 Maximum 6500
- 1 - minimum SIS, 2 - maximum SIS l'
O Amendment G April 30, 1990
C E S S A R in 0,c. m . l TABLE 6.2.1-19 (Cont'd) (Sheet 2 of 2)- ESF SYSTEMS PAIUMETERS FOR C0KTKIMENT PDCPRE350RE ANEYSIS Value Used for j Peak Pressure System / Item Full Capacity Analyses
- l Containment Spray Heat Exchanger i Type Shell and Shell and U-tube U-tube '
Number 2 1 Heat transfer area,-ft /2 unit 5100 5100 ,
- Overallhegttransfercoefficient, 356 356 l Btu /hr-ft *F !
Flowrates Spray (tube), (minimum), gpm 5000 5000 CCW (shell), gpm 8000 8000 a ; i Source of cooling water. Component Component Cooling Water Cooling Water ' (CCW) (CCW) i l-i
?
l O Amendment G April 30, 1990
1 CESSAR unkm 0 TABLE 6.2.1-20 1; l CONTAllMDIT SPRAY PLMP ACTIVATI0lt CHARACTERISTICS . j i Maximum Spray i Actuation Delay Time (sec)- -j Loss'of Offsite ! Description Offsite Power- _Poey Available ) 1 Loss of Coolant Accident (LOCA) 68 N/A 1 Main Steam Line Break (MSLB) N/A* 68** ! G
- Containment pressurization effect of MSLB is more severe when the offsite O power-is available,
** Value is used in analysis. Actual time is less.
i 1
-l 4
O Amendment G April 30, 1990
. CESSAR EiEncamo I
TABLE 6.2.1-21 l (Sheet 1 of 7) TYPICAL PASSIVE HEAT SIK DATA Part A:. Detailed Listing for a Typical . System 80+ Containment
)
1 CONCRETE l l Thickness (1) (2) Remarks- l (ft) Surface Surface (All areas given are for -l concrete only. Embedded Description / (exposed . Location one side) Areg) (ft Areg) (ft plate has been deducted). Reactor Cavity 1.00 480 528 Area includes internal 1.50 537 591 surface of.the reactor 2.00 436 480 cavity from MAVEC 2.50 1,020 1,120 opening to the pool { 3.00 1,840 2,020 . seal. 3.25 3,610 3,970 5.00 501 551 o El 62+0 to 91+9 0.50 187 206 Areas from El 62+0 l (including top and 1.00 147 162 exclusive of Reactor a bottom of slab) 1.50 13,200 24,700 Cavity areas. 2.00 457 503 1 3.00 1,890 2,080 1 5.00 148 163 El. 91+9 to 112+6 1.50 16,300 17,900 l (excluding top 2.00 13,800 15,100
- of slab) 2.50 1,590 1,750 3.00 7,470 8,210 )
7.00 144 158 l El. 115+6 to 143+0 0.50 2,540 2,790 l'.00 6,030 6,630 1.50 25,800 28,400 2.00 17,900 19,600 2.50 607 668 3.00 6,320 6,950 11 El. 146+0 1.00 6,480 7,120 and above 1.50 25,100 27,600 2.00 40,800 44,900 Amendment G April 30, 1990
. . ~ . .- . ;CESSAR1EL". cam.
9: TABLE 6.2.1-21 (Cont'd) (Sheet 2 of 7) TYPICAL PASSIVE HEAT SIK DATA , Part A: Detailed Listing for a Typical System 80+ Containment t STEEL t Thickness (1) (2)- (in) Surface Surface Remarks: Description / (exposed (All carbon steel Location one side) Areg) (ft Areg) (ft except as noted). Liner Plate (3) 0.1875 13,000 93,500- Stainless Steel
- includes -IRWST Grating. 0.0938 112,000 168,000 - Galvanized Support Steel 0.2500 79,800 120,000 '!
Cable Tray 0.0469 22,800 34,200 Galvanized 0.0525- 11,000- 11,000 0.1250 1,000 1,000 Polar Crane 0.1250 3,950 5,930 0.1875 300 -450 0.3750 1,250 1,880 ' O.5000 3,370 5,060 l 0.6250 5,360 8,040 L- l'.0000 2,600 3,900 l 3.0000 2,410 3,620' l-Containment Plate '1.7500 104,000- 114,000 i Embedded Plate 0.5000 37,500 56,300 Plate on ceilings & walls of > 1 ft concrete. Safety Injection 2.0000 5,170 5,170 Tank Miscellaneous 0.5000 0 64,000 Equipment O 1 Amendment.G April 30, 1990 l l
k L CESSAR nnincui:. n : V TABLE 6.2.1-2.1 (Cont'd) (Sheet 3 of 7) TYPICAL PASSIVE HEAT SI E DATA 3 Pa:-t A:- Detailed Listing for a Typical System 80+' Containment . NOTES: l (1) This column of information is intended to be used for analyses where a small free volume or heat sink is conservative. G . (2) This column.of information is intended to be used for analyses where a large free volume or heat sink is conservt.tive. ; (3) Minimum' surface area does not reflect IRWST surface while the maximum surface area does. O
-r i
O . Amendment G April 30, 1990
CESSARinL",cui:n J O
. TABLE 6.2.1-21 (Cont'd) l (Sheet 4 of 7)
TYPICAL PASSIVE HEAT SINK DATA 1 Part B: Modeling of Heat Sinks for' Computer Input for a Typical System 80+ , Containment : (4) Surfacg Mass Passive Heat Sink _ Material Thickness (ft) Area (ft ) (1ba) Containment Plate Organic Paint 0.0004167 104,000 2,600 Inorganic Paint 0.0001667 2,947 Carbon Steel 0.1458- 7,430,000 0.5 Foot Concrete. Organic Paint 0.000667 2,727 109 Concrete 0.5 196,344 1.0 Foot Concrete Organic Paint 0.000667 13,137 526 , Concrete 1.0 1,892,000 1.5 Foot Concrete Organic Paint 0.000667 80,937 3,239 Concrete 1.5 1,748,000 2.0 Foot Concrete Organic Paint 0.000667 73,393 2,937 Concrete 2.0 21,137,000 2.5 Foot Concrete Organic Paint 0.000667 3,217 129 Concrete 2.5 1,158,000 3.0 Foot Concrete Organic Paint 0.000667 17,520 701 Concrete 3.0 7,569,000 3.25 Foot Concrete Organic Paint 0.000667 3,610 144 Concrete 3.25 1,689,000 5.0 Foot Concrete Organic Paint 0.000667 649 26 Concrete 5.0 467,300 ll 7.0 Foot Concrete Organic Paint 0.000667 144 6 j Concrete 7.0 145,000 Liner Plate Stainless Steel 0.015625 13,000 99,531 Grating Zinc Galvanizing 0.0002833 112,000 14,120 Carbon Steel 0.007817 429,000 Amendment G l 1 April 30, 1990 l
.-1
i
;CESSAR innnem:n .n,
( TABLE 6.2.1-21 (Cont'd) i (Sheet 5 of 7) TYPICAL PASSIVE HEAT SIE DATA
-Part 8: Modeling of Heat Sinks for Computer Input for a Typical System 80+ r Containment (4)
Surfacg Mass Passive Heat Sink Material Thickness (ft) Area (ft ) (lbs) Cable Tray Zinc Galvanizing 0.0002167 22,800 2,199 Carbon Steel 0.003908 43,660 Cable Tray Organic Paint 0.0004167 11,000 275 ' Inoraanic Paint 0.0001667 312-Carbon Steel 0.004375 23,581
- Cable Tray Organic Paint 0.0004167 1,000 25 I
-O V
Inorganic Paint Carbon Steel 0.0001667 0.01041 5,100 28 Support Steel Organic Paint 0.0004167 79,800 l',995 G ; Inorganic Paint 0.0001667 2,261 Carbon Steel 0.02083 814,500 Polar Crane Organic Paint 0.0004167 3,950 99 , (0.1250 In.) Inorganic Paint Carbon Steel 0.0001667 0.01041 20,150 112 l Polar Crane Organic Paint 0.0004167 300 7 (0.1875 In.) Inorganic Paint 0.0001667 8 Carbon Steel 0.015625 2,297 Polar Crane Organic Paint 0.0004167 1,250 31 (0.3750 In.) Inorganic Paint 0.0001667 35 Carbon Steel 0.03125 19,140 Polar Crane Organic Paint 0.0004167 3,370 84 (0.50 In.) Inorganic Paint 0.0001667 95 Carbon Steel 0.041667 68,804 Polar Crane Organic Paint 0.0004167 5,360 134 (0.625 In.) Inorganic Paint 0.0001667 152 Carbon Steel 0.052083 136,800 Amendment G April 30, 1990
'CESSAR !!Wicm:n
.N TABLE 6.2.1-21 (Cont'd)
(Sheet 6 of 7) TYPICAL PASSIVE HEAT SIR MTA Part B: Modeling of Heat Sinks for Computer Input for a Typical System 80+ Containment (4) Surfacg Mass Passive Heat Sink Material Thickness (ft) Area (ft ) (1ba) Polar Crane Orgar.tc Paint 0.0004167 2,600- 65 (1.0 In.') Ir.arganic Paint 0.0001667 74 Carbon Steel 0.08333 106,162 Polar Crane- Organic Paint 0.0004167 2,410 60 (3.0 In.) Inorganic Paint 0.0001667 68-Carbon Steel 0.25 295,225 L Safety Injection Organic Paint 0.0004167 5,170 129 [ ~ Tanks Inorgaaic Paint -0.0001667 146 i CarboroSteel 0.1667' 422,300 - G NOTE: (4) Total surface area exposed to containment atmosphere. All walls assumed I to be insulated on back side. ii ; O Amendment G April 30, 1990
CESSAR sininum l
-f TABLE 6.2.1-21 (Cont'd)
(Sheet 7 of 7) TYPICAL PASSIVE HEAT SlK DATA i T>pical Material Properties used in containment pressure and temperature ant. lyses in Sections 6.2.1.l* and 6.2.1.5**. i Thermal Volumetric Conductivity Density HeatCagacity
- Material (Btu /hr-ft *F) (lbm/ft ) (Btu /ft *F)
Carbon Steel. 24.000 490 54.5 , Zinc Galvanizing 65.300 445 40.7 Concrete 0.800 144 28.8 Stainless Steel 8.800 490 57.5 Organic Paint 0.167 60 28.8 1 Inorganic Paint 1.167 170 18.2
- Thermal Volumetric i
Conductivity HeatCagacity
** Material (Btu /hr-ft *F) (Btu /ft *F)
Carbon Steel 26.4 59.95 o Zinc Galvanizing 71.83 44.77-Concrete 1.5 31.68 Stainless Steel 9.68 63.25= Organic Paint 0.184 31.68 Inorganic Paint 1.284 20.02' s l O t L Amendment G April 30, 1990 1
CESSARIEnc.m. fQ TABLE 6.2.1-22 INITIAL C0fE)ITIONS FOR CONTAllMEKT MINIltM PRESSURE ANALYSIS , Parameters ' Assumed Value Initial temperature, *F (max) 110 Initial pressure, psia (min) -14.3-Relative humidity, % (max) 100 j U Refueling water temperature, 'F (min) 80 No heat input to containment from structures or -- primary and secondary system components Ideal gas behavior of air in containment -- O t i O u Amendment G April 30, 1990 l c
! hhh k!I b k ICATl3N' O.
J
. TABLE 6.2.1-23 (Sheet 1 of 12)
IDNG-TERN NASS AND ENERGY RELEASE 1
^
PART A: Blowdown, Reflood, Post Reflood Release-Data i Break Mass Break Energy Time Flow Rate Flow Rate (sec)~ (lbe/sec) (Btu /sec) 0.000L 0.00000E+00 0.00000E+00 0.028 7.54399E+04 4.16739E+07-0.053 7.49763E+04 4.13519E+07 0.101 7.65384E+04 4.21770E+07 i 0.155 1.03933E+05 5.73492E+07' 0.202 1.03370E+05 5.70807E+07 0.257 1.02327E+05 5.65480E+07 0.308 1.02142E+05 5.64862E+07' O.351 1.01734E+05 5.62937E+07 [ 0.504 9.96669E+04 5.52347E+07 0.662 9.78153E+04 5.42757E+07 , 0.812 9.61317E+04 5.33980E+07' ' O.952- 9.42525E+04 5.24081E+07 1 1.212- 9.16780E+04 5.10959E+07 1.412 9.03275E+04 5.04767E+07 1.612 8.80764E+04 4.93908E+07 1.812 8.53110E+04 4.80292E+07 G 2.012 8.19683E+04 4.63246E+07 3.012 7.24783E+04 4.13978E+07. 4.012 6.21422E+04 3.56171E+07 5.012- 5.82256E+04 3.34663E+07-6.006 5.25307E+04 3.07886E+07 7.012 4.11574E+04 2.60995E+07-8.004 2.86077E+04 2.13623E+07 .; 9.001 2.34231E+04 1.89132E+07 10.00 2.07247E+04 1.73225E+07 3 11.00 1.63386E+04 1.50402E+07 12.00 1.10760E+04 1.18028E+07 - 13.00 1.30962E+04 9.43083E+06 14.00 1.35235E+04 7.75199E+06 15.00 1.18662E+04 6.04711E+06 16.00 8.35923E+03 3.98625E+06 16.20 7.61847E+03 3.60785E+06 16.40 7.09407E+03 3.31586E+06 16.60 6.48957E+03 3.00631E+06 Amendment G April 30, 1990
hk R$ICAftN
' TABLE 6.~2.1-23 (Cont'd) :!
(Sheet 2 of 12) IDNG-TERN MASS AND ENERGY RELEASE PART A: Blowdown,-Reflood, Post Reflood Release Data Break Mass Break Energy i' Time Flow Rate Flow Rate (sec) (lba/sec) (Btu /sec) 16.80 3.37722E+03 1.55052E+06 17.00 3.35939E+03 1.53579E+06 17.20 3.34360E+03 1.52408E+06 17.40 3.29544E+03 1.50291E+06 17.60 3.21787E+03 1.47673E+06 17.80 2.85056E+03 1.31226E+06 1.8 00 2.54325E+03 1.17246E+06 ' 18.20 2.24486E+03 1.03981E+06' 18.40 1.94972E+03 9.11453E+05 18.60 1.65285E+03: 7.85437E+05 - 18.80 1.34658E+03 6.58857E+05 . - 18.81 0.00000E+00 0.00000E+00 18.91 1.54910E+02 2.00460E+05 19.61 3.14500E+02 4.09020E+05-20.21 3.61050E+02 4.69770E+05 G 20.71 4.92570E+02 6.41270E+05 21.31 6.25290E+02 8.13960E+05 21.91 7.42770E+02' 9.66380E+05 L 22.41 8.33360E+02 1.08380E+06 22.71 8.42330E+02 1.09522E+06 22.72 4.88551E+02 6.35228E+05 L 23.01 4.88389E+02 6.34955E+05 l= 23.61 4.88029E+02 6.34334E+05 24.11 4.87728E+02 6.33888E+05 24.71 4.87299E+0; 6.33250E+05 25.31 4.86823E+02 6.32565E+05 t 25.81 4.86481E+02 6.32171E+05 26.41 4.85970E+02 6.31452E+05 27.01 4.85454E+02 6.30727E+05 27.51 4.85019E+02 6.30118E+05 28.11 4.84503E+02 6.29387E+05 28.71 4.83993E+02 6.28662E+05 29.21 4.83610E+02 6.28198E+05 29.81 4.83105E+02 6.27485E+05 30.41 4.82566E+02 6.26713E+05 Amendment G April 30, 1990
CESSAR1=L m w , i l TABLE 6.2.1-23 (Cont'd) (Sheet 3 of 12) IDNG-TERN MASS AND ENERGY RELEASE l l PART A: Blowdown, Reflood,-Post Reflood Release Data 1 Break Mass Break Energy Time- Flow Rate Flow Rate (sec) (1ba/sec) (Btu /sec) 30.91 4.82142E+02 6.26122E+05 31.51 4.81638E+02 6.25484E+05 32.11 4.81185E+02 6.24828E+05 32.61 4.80768E+02 6.24237E+05 33.21 4.80263E+02 6.23523E+05 33.81 4.79764E+02 6.22816E+05 34.31 4.79347E+02 6.22224E+05 34.91 4.78848E+02 6.21511E+05
.n 35.51 4.78343E+02 6.20803E+05 l 36.01' 4.77926E+02 6.20211E+05 >
36.61 4.77427E+02 6.19498E+05 37.21 4.76922E+02 6.18790E+05 , 37.71 4.76505E+02 6.18199E+05 ., 38.31 4.76006E+02 6.17485E+05 G 38.91 4.75501E+02 6.16778E+05 39.41 4.75084E+02 6.16186E+05 40.01 4.74579E+02 6.15479E+05 40.61 4.74080E+02 6.14765E+05 41.11 4.73657E+02 6.14174E+05 41.71 4.73158E+02 6.13466E+05 42.31 4.72654E+02 6.12753E+05 42.81 4.72236E+02 6.12161E+05' 43.41 4.71731E+02 6.11453E+05 r 44.01 4.71227E+02 6.10740E+05 44.51' 4.70809E+02 6.10148E+05 45.11 4.70328E+02 6.09470E+05 45.71 4.69881E+02 6.08838E+05 46.21 4.69516E+02 6.08310E+05 46.81 4.69069E+02 6.07678E+05 47.41 4.68628E+02 6.07045E+05 47.91 4.68257E+02 6.06518E+05 48.51 4.67811E+02 6.05885E+05 40.01 4.67445E+02 6.05358E+05 49.61 4.66999E+02 6.04725E+05 50.21 4.66552E+02 6.04093E+05 L Amendment G April 30, 1990
> ?
CEDSARTEncmo TABLE 6.2.1-23 (Cont'd)_ (Sheet 4'of.12) IDNG-TERM MASS AND ENERGY RELEASE PART A: Blowdown, Reflood, Post Reflood Release Data ' Break Mass Break Energy o Time Flow Rate Flow Rate 3* (sec) (lbm/sec) (Btu /sec) 50.71 4.66181E+02 6.03565E+05 51.31 4.65740E+02 6.02927E+05 51.91 4.65293E+02 6.02295E+05 52.41 4.64922E+02 6.01767E+05
'53.01 4.64476E+02 6.01135E+05 j 53.61 4.64029E+02 6.00503E+05 1 54.11. 4.63658E+02 5.99975E+05-54.71 4 . 63 2 :.1E+02 5.99337E+05:
i 55.31 4.62705E+02 5.98705E+05 55.81 4.62393E+02 5.98177E+05 56.41 4.61947E+02 5.97545E+05 57.01 4.61500E+02 5.96913E+05 57.51 4.61135E+02 5.96385E+05 58.11 4.60694E+02 5.95753E+05 58.71 4.60247E+02 5.95126E+05 59.21 4.59882E+02 5.94599E+05 G 59.81 4.59441E+02 5.93972E+05
-60.41 4.59000E+02 5.93346E+05 60.91 4.58629E+02 5.92818E+05 61.51 4.58188E+02- 5.92192E+05 62.11 4.57748E+02 S.91559E+05 62.61 4.57376E+02 5.91037E+05 63.21 4.56936E+02 5.90405E+05 s 63.81 4.56489E+02 5.89773E+05 l
64.31 4.56124E+02 5~.89251E+05 64.91 4.55677E+02 5.88619E+05 65.51 4.55236E+02 5.87987E+05 66.01 4.54865P+02 5.87465E+05 66.61 4.54418E+02 5.86832E+05 67.21 4.53972E+02 5.86200E+05 L 67.71 4.53606E+02 5.85678E+05 68.31 4.53160E+02 5.85046E+05 68.91 4.52713E+02 5.84414E+05 69.41 4.52342E+02 5.83886E+05 70.01 4.51895E+02 5.83260E+05 l Amendment G L April 30, 1990 i
+
@ESSARIBM.co.
O . TABLE 6.2.1-23 (Cont'd) (Sheet 5 of 12) . IDNG-TERN NASS AND ENERGY RELEASE PART At Blowdown, Reflood, Post Reflood Release Data 1 Break Mass Break Energy Time Flow Rate Flow Rate (sec) (lba/sec) (Btu /sec) 70.61 4.51449E+02 5.82627E+05 ' 71.11 4.51078E+02 5.82100E+05 71.71 4.50631E+02 5.81467E+05 i 72.31 4.50184E+02 5.80835E+05 , 72.81 4.49813E+02 5.80307E+05 73.41 4.49367E+02 5.79675E+05 74.01 4.48920E+02 5.79043E+05 i 74.51 4.48549E+02 5.78515E+05 l 75.11 4.48096E+02 5.77883E+05 0 75.60 75.61 79.61 4.47725E+02 7.71940E+02 7.45540E+02 5.77355E+05 9.95440E+05 9.61080E+05 83.51 7.21020E+02 9.29210E+05 87.41 6.98280E+02 8.99660E+05 G 91.31 6.76770E+02 8.71740E+05 95.21 6.56150E+02 8.44970E+05 . 99.11 6.37300E+02 8.20520E+05 I 10.30 6.20120E+02 7.98210E+05 10.69 6.04490E+02 7.77900E+05 11.08 5.90270E+02 7.59420E+05 11.47 5.77360E+02 7.42620E+05 11.86 5.65650E+02 7.27370E+05 12.25 5.55040E+02 7.13540E+05 12.64 5.45460E+02 7.01020E+05 13.03 5.36800E+02 6.89700E+05 13.42 5.29000E+02 6.79490E+05 i 13.81 5.21990E+02 6.70280E+05 j 14.20 5.15690E+02 6.61990E+05 " 14.59 5.10040E+02 6.54530E+05 14.98 5.04980E+02 6.47830E+05 15.37 5.00460E+02 6.41820E+05 15.76 4.96420E+02 6.36430E+05 16.15 4.92840E+02 6.31620E+05 16.54 4.89650E+02 6.27310E+05 16.93 4.86810E+02 6.23450E+05 Amendment G April 30, 19 9 'J
C E S S A R tl W .c m . O TABLE 6.2.1-23 (Cont'd) (Sheet 6 of 12) IANG-TERN MASS AND ENERGY RELEASE PART A: Blowdown, Reflood, Post Reflood Release Data 1 Break Mass Break Energy Time Flow Rate Flow Rate (sec) (Ibe/ rec) (Btu /sec) 173.21 4.84310E+02 6.20010E+05 177.21 4.82030E+02 6.16860E+05 181.11 4.80070E+02 6.14110E+05 185.01 4.78340E+02 6.11650E+05 188.91 4.76800E+02 6.09450E+05 192.81 4.75450E+02 6.07470E+05 196.71 4.74250E+02 6.05690E+05 200.61 /.73200E+02 6.04090E+05 204.51 4.72270E+02 6.02650E+05 208.41 4.71460E+02 6.01350E+05 212.31 4.70740E+02 6.00180E+05 216.21 4.70110E+02 5.99110E+05 220.11 4.69560E+02 5.98150E+05 G ! 224.01 4.69070E+02 5.97260E+05 L 227.91 4.68650E+02 5.96460E+05 231.81 4.68280E+02 5.95710E+05 235.71 4.67960E+02 5.95030E+05 239.61 4.67680E+02 5.94400E+05 , 243.51 4.67430E+02 5.93810E+05 247.41 4.67230E+02 5.93270E+05 251.31 4.67060E+02 5.92770E+05 255.21 4.66920E+02 5.92290E+05 259.11 4.66790E+02 5.91840E+05 263.01 4.66690E+02 5.91420E+05 263.41 4.66550E+02 5.91260E+05 266.91 4.66460E+02 5.90880E+05 270.81 4.66410E+02 5.90520E+05 270.91 4.73590E+02 5.89995E+05 271.01 4.56650E+02 5.90552E+05 271.11 4.83350E+02 5.90599E+05 271.21 4.62295E+02 6.09401E+05 271.41 4.77705E+02 5.82139E+05 271.71 4.66418E+02 6.11194E+05 272.01 4.73582E+02 5.89820E+05 272.41 4.72774E+02 6.10180E+05 Amendment G April 30, 1990 l
CESSAR W L m. i-TABLE 6.2.1-23 (Cont'd) (Sheet 7 of 12) IONG-TERN MASS AND ENERGY RELEASE PART A: Blowdown, Reflood, Post Reflood Release Data i Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1bs/sec) _1 Btu /sec) 272.81 4.82226E+02 5.96155E+0C 273.31 4.74207E+02 6.11845E+05 273.91 4.89126E+02 6.06996E+0L 274.51 4.81000E+02 6.16338E+05 , 275.21 4.90428E+02 6.14482E+05 275.91 4.90286E+02 6.25518E+05 ' 276.71 4.94714E+02 6.22351E+05 , 277.51 4.98008E+02 6.32649E+05 , 278.41 5.04214E+02 6.33093E+05 l (' 279.31 5.05342E+02 6.42463E+05 l 280.31 5.10658E+02 6.42844E+05 l 281.41 5.14797E+02 6.49884E+05 G 282.51 5.1974 9 E+02 6.55571E+05 283.61 5.22640E+02 6.60793E+05 284.91 5.29668E+02 6.65596E+05 286.11 5.31366E+02 6.71076E+05 287.51 5.38634E+02 6.77496E+05 288.81 5.41553E+02 6.82504E+05 290.31 5.46563E+02 6.88201E+05 291.81 5.52103E+02 6.94465E+05 293.31 5.55978E+02 6.99164E+05 294.91 5.61522E+02 7.05836E+05 296.61 5.65614E+02 7.10767E+05 298.31 5.73210E+02 6.85703E+05 300.11 5.55583E+02 6.56263E+05 , 301.91 5.24278E+02 6.15503E+05 l 303.81 4.98464E+02 5.88708E+05 t 305.71 4.76165E+02 5.59659E+05 307.71 4.57835E+02 5.39341E+05 309.81 4.39173E+02 5.16959E+05 311.91 4.24708E+02 5.00184E+05 314.01 4.10434E+02 4.83461E+05 316.21 3.98657E+02 4.69266E+05 318.51 3.86050E+02 4.54586E+05 320.81 3.75015E+02 4.41418E+05 l Amendment G i April 30, 1990 1
CESSAR nfa.co.. ! i d TABLE 6.2.1-23 (Cont'd) I (Sheet 8 of 12) IDDIG-TERN MASS AND ENERGY RELEASE FART A: Blowdown, Reflood, Post Reflood Release Data Break Mass Break Energy Tlas Flow Rate Flow Rate (sec) (Ibe/sec) (Btu /sec) 323.21 3.64236E+02 4.28776E+05 325.71 3.53542E+02 4.16824E+05 328.11 3.44064E+02 4.04630E+05 330.71 3.33914E+02 3.93032E+05 333.31 3.24669E+02 3.81996E+05 'i 336.01 3.15471E+02 3.71483E+05 238.71 3.06861E+02 3.61285E+05 341.41 2.99065E+02 3.51594E+05 344.31. 2.90250E+02 3.42199E+05 347.11 2.82858E+02 3.33515E+05 350.11 2.75808E+02 3.24164E+05 353.11 2.68336E+02 3.16003E+05 356.11 2.63957E+02 3.11330E+05 359.21 2.60559E+02 3.06124E+05 g 362.41 2.55393E+02 3.00935E+05 . 365.61 2.51571E+02 2.96565E+05 368.81 2.47817E+02 2.91383E+05 372.21 2.43964E+02 2.88029E+05 l 375 51 2.40369E+02 2.82707E+05 2.36857E+02 2.79007E+05 379.01 382.51 2.33519E+02 2.74888E+05
- 386.01 2.30481E+02 2.71397E+05 l 389.61 2.27364E+02 2.68047E+05 .
l 393.31 2.24528E+02 2.64067E+05 397.01 2.21958E+02 2.60899E+05 400.71 2.18811E+02 2.58020E+05 404.51 2.16461E+02 2.54731E+05 l 408.41 2.13878E+02 2.51936E+05 412.41 2.11622E+02 2.49064E+05 416.31 2.09055E+02 2.46054E+05 420.41 2.07042E+02 2.43702E+05 424.51 2.04733E+02 2.41176E+05 l 428.61 2.02667E+02 2.38503E+05 432.81 2.01142E+02 2.36735E+05 437.11 1.98349E+02 2.33734E+05 l Amendment G I L April 30, 1990 ) l l
CESSAR W=iew.a O TABLE 6.2.1-23 (Cont'd) - (Sheet 9 of 12) j IANG-TERN MASS AND ENERGY RELEASE 1 PART A: Blowdown, Reflood, Post Reflood Release Data Break Mass Break Energy Time Flow Rate Flow Rate (sec) (1bs/sec) (Btu /sec) 441.41 1.97465E+02 2.31848E+05 445.81 1.94691E+02 2.29306E+05 450.21 1.93490E+02 2.27967E+05 l 454.71 1.91449E+02 2.24934E+05 459.21 1.89615E+02 2.23955E+05 0 463.81 1.88211E+02 2.21119E+05 468.51 1.86682E+02 2.19732E+05 473.21 1.84641E+02 2.17449E+05 478.01 1.83693E+02 2.15884E+05 O 482.81 1.81564E+02 2.13835E+05 487.61 1.80520E+02 2.12415E+05 i 492.61 1.78770E+02 2.10427E+05 i 497.61 1.77630E+02 2.08773E+05 502.61 1.75785E+02 2.06964E+05 507.71 1.74803E+02 2.05585E+05 512.81 1.73140E+02 2.03833E+05 518.01 1.72245E+02 2.02321E+05 523.31 1.70280E+02 2.00480E+05 528.61 - 1.69343E+02 1.99142E+05 534.01 1.62879E+02 1.91599E+05 539.41 1.33266E+02 1.56782E+05 544.91 1.29279E+02 1.52309E+05 550.41 1.15660E+02 1.35975E+05 O Amendment G April 30, 1990
i CESSAR ElinPco..a O. TABLE 6.2.1-23 (Cont'd) (Sheet 10 of 12) - IDNG-TERN MASS AND ENERGY RELEASE SPILIAGE DATA Time, Sec Mass, lha Energy, 10 Btu End of Blowdown 18.81 0.0 0.0 G End of Reflood 270.71 332,902 50.960 End of Post Reflood 550.41 455,561 82.891 0 1 1 0' Amendment G April 30, 1990
CESSAR !!W.co... O TABLE 6.2.1-23 (Cont'd) (Sheet 11 of 12) IDNG-TERN MASS AND ENERGY RELEASE PART B: Long Term Steam Release Data Break Mass Break Energy Time Flow Rate Flow Hate (sec) (Ibm /sec) (Stu/sec) 550.40 0.000 0 550.41 147.782 174182 i 799.9 140.120 165028 > 1297.5 122.534 144531 1797.5 115.740 136495 2190.0 76.232 89963 3044.0 72.292 85333 . 3044.1 72.292 85333 g 3190.0 72.735 85856 . h 4190.0 5190.0 68.756 55.389 81123 65356 9190.0 46.336 54603 12975.0 42.290 49868 22975.0 36.772 43301 27975.0 35.008 41184 42975.0 31.097 36533 62975.0 27.850 32702 97975.0 24.466 28669 191900.0 19.595 22896 l 241900.0 18.032 21020 291900.0 16.836 19617 391900.0 15.120 17587 491900.0 13.919 16183 591900.0 12.943 15042 691900.0 12.169 14138 791900.0 11.542 13387 891900.0 11.015 12772 1000000.0 10.526 12203 O Amendment G April 30, 1990
CESSAR Ennen... O TABLE 6.2.1-23 (Cont'd) (Sheet 12 of 12) IONG-TERN MASS AND ENERGY RELEASE PART C: Long Term Spill Release Data Spill Mass Spill Energy Time Flow Rate Flow Rate (sec) (1ba/sec) (Btu /sec) 550.40 0.000 0 550.41 525.817 40987 799.9 533.479 41584 . 1297.5 551.065 42955 ; 1797.5 557.859 43485 l 2190.0 597.367 46564 l 3044.0 600.362 46797 1 3044.1 600.362 101158 l 3190.0 600.864 101666 4190.0 604.843 105222 5190.0 618.210 110301 9190.0 627.263 118907 12975.0 631.309 122665 22975.0 636.827 125628 27975.0 638.591 125638 42975.0 642.502 123723 G 62975.0 645.749 120046 97975.0 649.133 114715 191900.0 654.004 106562 241900.0 655.567 103904 291900.0 656.763 101739 391900.0 658.479 98459 491900.0 659.680 96049 591900.0 660.656 94112 691900.0 661.430 92518 791900.0 662.057 91215 891900.0 662.589 90114 1000000.0 663.073 89082 O Amendment G April 30, 1990
i O O O TABLE 6.2.1-24 (Sheet 1 of 15) i ; I ENERGY BAUWICES [ 1 ! Part A: Double-Ended Suction Leg Slot with Minimum SIS Flou l Energy (106 8tu) Prior to End At Peak Pressum Of Bloudoun At End of After End of At End of At End of Prior Peak Pressum Blowdoun 81mudoun Reflood Pest Reflood Energy Description to LOCA (24.80 sec) (24.80 sec) (199.80 sec) (114.21 sec) (166.71 sec) ! Re:ctor Coolant System Water 391.194 30.400 30.400 48.980 49.001 46.776 Internal Energy
- Safety Injection Tank Water 42.716 31.658 31.658 5.758 4.535 0.0
! Internal Energy Energy Stored in Core 29.268 11.617 11.617 3.777 3.593 6.016 . Energy Stored in RV Internals 40.742 35.819 35.819 32.458 32.279 30.722 g Energy Stored in RV Metal 75.257 73.986 73.986 73.719 73.703 73.997 i Energy Stored in Pressurizer, 183.635 168.470 168.470 170.214 170.239 170.462 Primary Piping, Valves, anc i Pumps Energy Stored in Steam 34.184 30.573 30.573 21.558 21.454 20.655 Gen:rator Tubes Energy Stored in Steam 153.046 152.934 152.934 149.650 149.416 146.855 Gen:rator Secondary Walls Amendment G April 30, 1990
i I i TABLE 6.2.1-24 (Cont'd) l (Sleeet 2 of 15) l Euracy BAUUICES l l Pcrt A: Double-Ended Suction Leg Slot with Minimum SIS Flow Energy (106 Btu) , Prior to End At Peak Pressure j Of 81W At End of After End of At End of At End of Prior Peak Pressure Bloudoun Bleudoun Reflood Fest Refleed l Energy Description to LOCA (24.80 sec) (24.80 sec) (199.80 sec) 1114.21 sec) (166.T1 sec) Secondary Coolant Internal 135.850 138.610 138.610 120.068 119.248 114.985 , Energy in Steam Generator 1 Secondary Coolant Internal 135.850 128.613 128.613 77.612 75.974 67.829 Energy in Steam Generatcr 2 g Secondary Coolant Internal 35.996 35.996 35.996 35.996 35.996 35.996 Energy in Steam Line Total NSSS Stored Energy 1257.738 838.680 838.680 739.790 735.438 713.597 , Feedwater to Steam Generator 1 0.0 12.620 12.620 12.620 12.620 12.620 4 Feedwater to Steam Generator 2 0.0 12.620 12.620 12.620 12.620 12.620 Steam Flow to Turbine 0.0 0.058 0.058 0.058 0.058 0.058 Energy Generated During Shutdown 0.0 7.150 7.150 22.032 22.685 30.290 from Decay Heat Amendment G April 30, 1
O O O TABLE 6.2.1-24 (Cont'd) (Sheet 3 of 15) ENERGY BAUWICES Pcrt A: Double-Ended Suction Leg Slot with Minimum SIS Flow Energy (106 Btu) Prior to End At Peak Pressure Of Bleudoun At End of After End of At End of At End of 4 Prior Peak Pressure Bi h 81eudoun Reflood Post Re*1eed i Energy Description to LOCA (24.80 sec) (24.80 sec) (199.80 sec) (114.21 sec) (166.71 sec) Break 0.0 426.150 426.150 518.790 521.200 526.440 Spillage 0.0 0.0 0.0 61.632 64.111 80.451 Total 0.0 426.150 426.150 572.339 577.445 606.891 g Energy Content of RC'> 5.281 337.486 337.486 364.682 364.239 345.982 Atmosphere , I Energy Content of RCB Internal 0.0 28.237 28.237 70.933 72.890 90.149 Structures (Relative to 110*F) Energy Content of IRWST Water 318.696 385.360 385.360 458.683 462.175 492.282 ; Energy Removed by Containment 0.0 0.0 0.0 0.0 0.0 0.0 !
- Spray Heat Exchangers !
i 4 Amendment G April 30, 1990 L v -_ _ _ ,ie- ,.e, -. , , , . _ _ _ _ ___ _. ,__m._m_
TABLE 6.2.1-24 (Cont'd) (Sheet 4 of 15) ERERGY BAUNCES
..Pcrt 8: Double-Ended Discharge Leg Slot with Maximum SIS Flow Energy (106 Stu) Gl Prior to End At Peak Pressure of Bloudoun At End of At End of After End of At End of i Prior Peak Pressure 81oudoun Reflood 81oudoun Post Reflood Energy Description to LOCA (16.38 sec) (18.80 sec) (270.71 sec) (323.80 sec) (558.41 sec) At I Day Reactor Coolant System Water 391.194 32.630 32.329 64.605 63.787 57.974 19.425 >
l Internal Energy , Safety Injection Tank Water 42.716 37.267 33.898 0.0 0.0 0.0 0.0 i Internal Energy Energy Stored in Core 29.268 12.410 11.609 6.152 5.794 5.648 5.602 i Energy Stored in RV Internals 40.742 34.583 34.636 29.865 29.277 27.083 16.686 Energy Stored in RV Metals 75.257 74.563 74.319 73.5% 73.452 72.780 31.442 Energy Stored in Pressurizer, 183.635 177.810 175. 9 9 169.975 168.121 160.472 103.609 Pricary Piping, Valves and Pumps ;
- En
- rgyStoredinSteam 34.184 32.220 32.249 19.420 18.347 15.936 14.152
, Gen:rator Tubes ! Energy Stored in Steam 153.046 153.909 153.780 146.02Z 144.525 138.173 66.178 Gen:rator Secondary Walls ! Amendment G April 30,
+,e..q-e ,_.-a - .
vm. g y 4 - ~ ,pys w - sp*. g .,
O O O TABLE 6.2.1-24 (Cont'd) (Sheet 5 of 15) ENEREY RAIJINCES , Part B: Double-Ended Discharge Leg Slot with Maximum SIS Flow Energy (106 Btu)- l Prior to End At Peak Pressere Of Blowdown At End of At End of After End of At End of Prior Peak Pressure Bloudoun Reflood Bloudun Ptst Reflood Enerviy Description to LOCA (16.38 sec) (18.80 sec) (270.71 sec) (323.80 ser._ (550.41 sec) At 1 Day Secondary Coolant Internal 135.850 137.030 151.724 97.738 89.254 68.244 58.434 gl l Energy in Steam Generator 1 t Secondary Coolant Internal 135.850 133.910 148.517 79.191 70.914 68.023 58.434 Energy in Steam Generator 2 ! Secondary Coolant Internal 35.996 35.9 % 35.9 % 35.996 35.996 35.996 35.996 Energy in Steam Line Total NSSS Stored Energy 1257.738 861.930 854. % 5 722.560 699.467 650.284 409.958 . r Feedwater to Steam Generator 1 0.0 0.0 12.620 12.620 12.620 12.620 12.620 , l Feedwater to Steam Generator 2 0.0 0.0 12.620 12.620 12.620 12.620 12.620 i Steaa Flow to Turbine 0.0 0.058 0.058 0.058 0.058 0.058 0.058 ' Energy Generated During 0.0 4.84 5.375 43.137 49.313 75.579 3012.579 :
- Shutdown from Decay Heat l
Amendment G-April 30, 1990 ,
4 4 TABLE 6.2.1-24 (Cont'd) (Sheet 6 of 15) J ENERGY BAUUICES Pcrt B: Double-Ended Discharge leg Slot with Maximum SIS Flow i Energy (106 Bte) Prior to End At Peak Pressure Of Bleudoun At End of At End of After End of At End of ' Prior Peak Pressure Bleudoun Reflood Bleudoun Pest Reflood Energy Description to LOCA (16.38 sec) (18.80 sec) (270.71 sec) (323.80 sec) (550.41 sec) At 1 Der Break 0.0 404.030 408.090 573.605 603.497 661.505 985.697 - G Spillage 0.0 0.0 0.0 50. % 0 53.057 82.891 208.025 Total 0.0 404.030 408.090 624.565 656.554 744.3 % 1193.722
' Energy Content of RCB 5.281 324.762 322.358 349.570 359.661 337.916 139.047 Atmosphere Energy Content of RC8 Internal 0.0 19.620 23.061 114.471 126.837 168.959 688.897 i Structures (Relative to 110*F)
Energy Content of IRWST Water 318.696 383.625 386.598 469.557 470.418 529.865 841.659 Energy Removed by Containment 0.0 0.0 0.0 0.0 0.0 0.0 2404.049 Spray Heat Exchangers Amendment G April 30, 1
- O O O l TABLE 6.2.1-24 (Cent'd)
(Sheet 7 of 15) ENERGY BAUWICES Part C: Double-Ended Hot Leg Slot Energy (100 Btu) Prior *,e End , Of Blaudoun At End of l Prior Peak Pressure Blaudoun l Energy Description to LOCA (14.48 sec) (15.68 sec) , Reactor Coolant System Water 391.194 24.465 24.900 Internal Energy Safety Injection Tank Water 42.716 33.637 32.210
- Internal Energy !
Energy Stored in Core 29.268 11.331 11.496 Energy Stored in RV Internals 40.742 37.310 37.080 Energy Stored in RV Metal 75.257 69.391 68.963 Energy Stored in Pressurizer, 183.635 173.735 173.284 Primary Piping, Valves, and Pumps ! Energy Stored in Steam Generator 34.184 29.610 29.810 Tu%s Energy Stored in Steam Generator 153.046 150.158 150.513 Secondary Walls Amendment G April 30, 1990
TABLE 6.2.1-24 (Cont'd) (Sheet 8 of 15)
- 1 ENERGY BAUNCES, Part C: Double-Ended Hot Leg Slot Energy (106 Btu)
Prior to End . Of Bleudoun At End of Prior Peak Pressure Bleudoun Energy Description to LOCA (14.48 sec) (15.60 sec) Secondary Coolant Internal 135.850 130.955 131.00 Energy in Steam Generator 1 Secondary Coolant Internal 135.850 117.605 117.550 Energy in Steam Generator 2 : Secondary Coolant Internal 35.996 35.996 35.996 :
- Energy in Steam Line o Total NSSS Stored Energy 1257.738 814.206 812.802 ,
f Feedwater to Steam Generator 1 0.0 0.0 12.62 ; i ; ! Feedwater to Steam Generator 2 0.0 0.0 12.62 ! Steam Flow to Turbine 0.0 0.058 0.058 . Energy Generated During Shutdown 0.0 5.175 5.423 from Decay Heat Amendment G April 30, 1
O O O l l l TABLE 6.2.1-24 (Cont'd) (Sheet 9 of 15) i ENEREY RAUNCES l l Part C: Double-Ended Hot Leg Slot .i l Energy (106 Bte) l l Prior to End Of Bleudoun At End of I Prior Peak Pressere Bleudoun . ! Energy Description. to LOCA (14.48 sec) (15.60 sec) Break 0.0 441.735 443.150 Spillage 0.0 0.0 0.0 Total 0.0 441.735 443.150 Energy Content of RCB 5.281 431.191 426.588
- Atmosphere U
Energy Content of RCB Internal 0.0 21.523 27.110 Structures (Relative to 110*F) Energy Content of IRWST Water 318.6 % 310.000 348.910 Energy Removed by Containment 0.0 0.0 0.0 Spray Heat Exchangers f Amerwheent G April 30, 1990
.. . _. _ _~ _. __ _- . . . _ . _ _ _ _ . . _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . - _ _ _ _ _
i l' : l table 6.2.1-24 (Cont'd) l (Sleeet 10 of IS) DIERGY BAUllICES Part D: 0% Power MSLB with Loss of One CSS Train , Energy (106 8tu) At Peak Prior Pressere Energy Description to H5tB (277.80 sec) Reactor Coolant System Water Internal Energy 373.680 265.839 Safety Injection Tank Water Internal Energy 42.716 42.716 Energy Stored in Core 10.075 7.391 Energy Stored in RV Internals 36.006 26.540' Energy Stored in RV Metal U 1 66.509 49.023 ; i ' Energy Stored in Pressurizer, Primary Pipir.g, 166.078 128.837 Valves, and Pumps , ! Energy Stored in Steam Generator Tubes 32.733 24.179 Energy Stored in Steam Generator Secondary 155.694 132.740 Walls Amendament'G April 30,
O O O TABLE 6.2.1-24 (Cont'd) (Sheet 11 of 15) l UERGY BALE % . Part D: 0% Power MSLB with Loss of One CSS Train ; Energy (106 Stu) At Peak Prior Pressure Energy Description to MSLB (277.80 sec) Secondary Coolant Internal Energy in Steam 1%.387 135.380 Generator 1 Secondary Coolant Internal Energy in Steam 1%.387 2.085 Generator 2 Secondary Coolant Internal Energy in Steam 39.867 32.253 Line l Total NSSS Stored Energy 1316.132 846.983 Feedwater to Steam Generator 1 0.0 0.0 I Feedwater to Stean Generator 2 0.0 2.640 i Steam Flow to Turbine 0.0 -0.004 Energy Generated During Shut <10wn from 0.0 0.016 ] Decay Heat 1 Amendment G April 30, 1990
--,e, . - - - , .o - . , ~ -e s . ,
k 4 TABLE 6.2.1-24 (Cont'd) (Sheet 12 of 15) olosv aAUUICES Part D: 0% Power MSLB with loss of One CSS Train Energy (106 Btu) i At Peak Prior Pressure Energy Description to ftSLB (277.80 sec) Break 0.0 472.572 Spillage 0.0 0.0 Total 0.0 472.572 5.281 378.515 G Energy Content of RCB Atmosphere 4 Energy Centent of RCB Internal Structures 0.0 62.308
- (Relative to 110*F)
! Energy Content of IRWST Water 318.6 % 355.726 Energy Removed by Containment Spray Heat 0.0 0.0 Exchangers I i Ameruheent G April 30, 1 f
O O O . a TAti:5. 6.2.1-24 (Cont'd) (Sheet 13 of 15) , ElIENCY BAUWICES t Part E: 102% Power MSLB with MSIV Failure 1 Energy (106 gg,) At Peak Prior Pressure Energy Description to MSLB 1242.80sec) Reactor Coolant System Water Internal Energy 391.194 291.916 Safety Injection Tank Water Internal Energy 42.716 42.716 i Energy Stored in Core 29.268 8.728 1 Energy Stored in RV Internals 40.742 32.954 l Energy Stored in RV Metal 75.257 60.870 g Energy Stored in Pressurizer, Primary Piping, 183.635 147.600 ' Valves, and Pumps Energy Stored in Steam Generator 1:bes 34.184 26.603 l i Energy Stored in Steam Generator 153.046 142.307 Secondary Walls Amendment G April 30, 1990 4
- ' TABLE 6.2.1-24 (Cont'd)
(Sheet 14 of 15) ENENEY BAUWCES Part E: 102% Power MSLB with MSIV Failure Energy (106 Btu) r I At Peak Prior Pressure -i Energy Description to MSLB (242.80 sec) Secondary Coolant Internal Energy in 135.850 101.502 l Steam Generator 1 Secondary. Coolant Internal Energy in '135.850 3.537 i Steam Generator 2 i Secondary Coolant' Internal Energy in 35.9 % 3.142 i Steam Line g Total NSSS Stored Energy 1257.738 861.875 Feedwater to Steam Generator 1 0.0 0.0 feedwater to Steam Generator 2 0.0 45.345 ! Steam Flow to Turbine 0.0 -16.980 : Energy Generated During Shutdown from 0.0 61.726 _ Decay Heat Amendment G April 30, 1
O O O TABLE 6.2.1-24 (Cont'd) (Sheet 15 of 15) DIERGY BAUWICES Part E: 102% Power MSLB with MSIV Failure Energy (106 Stu) At Peak Prior Pressure Energy Description to ftSLB (242.80 sec). Break 0.0 486.000 1 Spillage 0.0 0.000 Total 0.0 486.000 Energy Content of RCB Atmosphere 5.281 372.110 U Energy Content of RCB Internal Structures 0.0 60.848 (Relative to 110*F) Energy Content of IRWST Water 318.6 % 377.619 Energy Removed by Containment Spray 0.0 0.0 Heat Exchangers l l Amendment G t April 30, 1990 I l
.~ . -. -- .
C E S S A R ti! W.cu ;
.O TABLE 6.2.1-25 ;
PRIMARY-SIDE RESISTANCE FACTORS FLOOD N002 CODE Resistance Factor, R' psi 2 3 x 100 lbm ft :' Path see sec Core i Lower Core 0.3719 Upper Core 0.3260 , Upper Plenum to Steam Generator, Broken Side l Upper Plenum to Tubes 1.996 Tubes to Steam Generator Outlet 1.913 Steam Generator Outlet in Broken Side of Annulus Forward Flow 14.82 O\ Reverse Flow 76.48 4 Annulus to Break Suction Leg Break 73.87 j Discharge Leg 1.299 g J Steam Generator Outlet is Broken Side to Break Suction Leg Break 2.612 Discharge Leg Break 13.54 Upper Plenum to Annulus, Intact Side Upper Plenum to Tubes 1.996 Tubes to Annulus 5.619 Break Resistances 1.0 Break 4.479 O Amendment G April 30, 1990 l
CESSAR Kncoc. (ABLE 6.2.1-26 l (Sheet 1 of 3) BLOWD0tm Als REFLOOD MASS Als ENERGY RELEASE FOR THE NINIIRM CONTAllMEKt PRESSURE ANALYSIS Integral Integral of of Time Mass Flow Energy Release Mass Flow Energy Release (sec) (1bs/sec) _ (8tu/sec) (1ba) (Btu) 0.00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.05 7.3924E+04 4.0122E+07 3.7078E+03 2.0057E+06 I 0.10 1.0157E+05 4.7505E+06 8.7514E+03 4.7505E+06 0.15 1.0084E+05 5.5441E+07 1.3769E+04 7.5020E+06 , 0.20 9.8853E+04 5.4536E+07 1.8746E+04 1.0243E+07 O.25 9.7259E+04 5.3762E+07 2.3621E+04 1.2936E+07 0.35 9.4620E+04 5.2399E+07 3.3260E+04 1.8270E+07 1 0.45 9.2814E+04 5.1430E+07 4.2624E+04 2.3457E+07 1.0 8.7613E+04 4.8608E+07 9.2246E+04 5.0971E+07 1.4 8.1883E+04 4.5506E+07 1.2647E+05 6.9977E+07 1.8 O' 2.2 7.1167E+04 6.5254E+04 3.9592E+07 3.6383E+07 1.5704E+05 1,8420E+05 8.6970E+07 1.0210E+08 i l 2.6 6.0573E+04 3.3868E+07 2.0931E+05 1.1612E+08 1 3.0 5.8184E+04 3.2640E+07 2.3292E+05 1.2934E+08 G 3.4 5.6180E+04 3.1622E+07 2.5579E+05 1.421?E+08 3.8 5.4704E404 3.0933E+07 2.7796E+05 1.5470E+08 4.4 5.1784E+04 2.9544E+07 3.1060E+05 1.7323E+08 5.2 4.7547E+04 2.7625E+07 3.5026E+05 1.9605E408 6.0 4.2624E+04 2.5352E+07 3.8618E+05 2.1717E+08 l 6.8 3.8966E+04 2.3628E+07 4.1875E+05 2.3673E+08 7.6 3.5269E+04 2.1765E+07 4.4841E+05 2.5488E+08 8.4 3.0855E+04 1.9673E+07 4.7493E+05 2.7146E+08 9.2 2.2337E+04 1.6374E+07 4.9622E+05 2.8579E+08 10.0 1.8031E+04 1.4281E+07 5.1173E+05 2.9791E+08 11.0 1.6591E+04 1.3144E+07 5.2923E+05 3.1165E+08 12.0 1.4318E+04 1.1918E+07 5.4467E+05 3.2416E+08 13.0 1.2104E+04 1.0808E+07 5.5785E+05 3.3441E+08 14.0 1.0146E+04 9.6475E+06 5.6896E+05 3.4577E+08 l o Amendment G I April 30, 1990 1
CESSAR !!ninc no. G-TABLE 6.2.1-26 (Cont'd) (Sheet 2 of 3) BL0llD0lfN Als REFLOOD DESS Alm ENERGY RELEASE S Integral Integral of of Time Mass Flow Energy Release Mass Flow Energy Release (sec) (1bm/sec) (Btu /sec) (1bs) (Btu) 15.0 1.0897E+04 8.8689E+06 5.7904E+05 3.5494E+08 16.0 1.1092E+04 7.7999E+06 5.9004E+05 3.6325E+08 17.0 1 ll35E+04 7.0412E+06 6.0122E+05 3.7067E+08 18.0 1.0577E+04 6.2292E+06 6.1208E+05 3.7729E+08 19.0 1.0014E+04 5,5094E+06 6.2238E+05 3.8315E+08 20.0 9.4133E+03 4.8418E+06 6.3213E+05 3.8833E+08 21.0 8.4408E+03 4.1023E+06 6.4106E+05- 3.9280E408 22.0 7.7334E+03 3.5410E+06 6.4910E+05 3.9660E+08 23.0 6.9631E+03 2.9982E+06 6.5649E+05 3.9988E+08 24.0 6.1905E+03 2.5142E+06 6.6284E+05 4.0255E+08 25.0 26.0 27.0 2.8509E+03 2.5439E+03 1.8511E+03 1.2547E+06 1.0635E+06 7.7339E+05 6.6767E+05 6.7094E+05 6.7314E+05 4.0452E+08 4.0586E+08 4.0678E+08 h , a 28.0 1.0931E+03 4.7983E+05 6.7461E+05 4.0741E+08 28.5 6.8173E+02 3.2866E+05 6.7502E+05 4.0759E+08
-Time of Annulus Downflow
,- Start of Reflood (values below are for steam only) l l 38.3 0.0000E+00 0.0000E+00 6.7502E+05 4.0759E+08 48.3 0.0000E+00 0.0000E+00 6.7502E+05 4.0759E+08 58.3 0.0000E+00 0.0000E+00 6.7502E+05 4.0759E+08 68.3 0.0000E+00 0.0000E+00 6.7502E+05 4.0759E+08 l 78.3 0.0000E+00 0.0000E+00 6.7502E+05 4.0759E+08 88.3 0.0000E+00 0.0000E+00 6.7502E+05 4.0759E+08 98.3 1.8090E+02 2.3788E+05 6.7561E+05 4.0837E+08 108.3 1.9152E+02 2.5185E+05 6.7748E+05 4.1082E+08 l 118.3 1.9818E+02 2.6060E+05 6.7943E+05 4.1339E+08 128.3 2.0306E+02 2.6702E+05 6.8144E+05 4.1603E+08 l' l O' l l. f-
CESSAR finincano. O TABLE 6.2.1-26 (Cont'd) (Sheet 3 of 3) BLOWDOWN Als REFLOOD MASS M ENERGY RELEASE FOR THE MINURM CONTAHOENT PRE 55URE ANALYSIS Integral Integral of of Time Mass Flow Energy Release Mass Flow Energy Release M (lbs/sec) (Btu /sec) (lba) (8tu) 138.3 2.0656E+02 2.7163E+05 6.8349E+05 4.1872E+08 148.3 2.1857E+02 2.8741E+05 6.8563E+05 4.2155E+08 158.3 2.1902E+02 2.8801E+05 6.8782E+05 4.2442E+08 168.3 2.1842E+02 2.8723E+05 6.9001E+05 4.2730E+08 178.3 2.1802E+02 2.8669E+05 6.9219E+05 4.3017E+08 , 188.3 2.1769E+02 2.8627E+05 6.9437E+05 4.3303E+08 198.3 2.1739E+02 2.8587E+05 6.9654E+05 4.3589E+08 208.3 2.1708E+02 2.8546E+05 6.9872E+05 4.3875E+08 228.3 2.1695E+02 2.8528E+05 7.0305E+05 4.4445E+08 248.3 2.1605E+02 2.8411E+05 7.0738E+05 4.5015E+08 268.3 2.1581E+02 2.8380E+05 7.1170E+05 4.5583E+08 O' 288.3 308.3 2.1556E+02 2.1527E+02 2.8346E+05 2.8308E+05 7.1602E+05 7.2032E+05 4.6150E+08 4.6716E+08 328.3 2.1496E+02 2.8267E+05 7.2463E+05 4.7282E+08 348.3 2.1463E+02 2.8224E+05 7.2892E+05 4.7847E+08 368.3 2.1460E+02 2.8220E+05 7.3321E+05 4.8411E+08 388.3 2.1416E+02 2.8162E+05 7.3750E+05 4.8975E+08 G 408.3 2.1363E+02 2.8092E+05 7.4178E+05' 4.9538E+08 428.3 2.1306E+02 2.8017E+05 7.4604E+05 5.0099E+08 448.3 2.1275E+02 2.7977E+05 7.5030E+05 5.0659E+08 468.3 2.1240E+02 2.7931E+05 7.5455E+05 5.1217E+08 488.3 2.1204E+02 2.7883E+05 7.5880E+05 5.1776E+08 508.3 2.ll66E+02 2.7833E+05 7.6304E+05 5.2333E+08 528.3 2.1128E+02 2.7783E+05 7.6727E+05 5.2889E+08 538.3 2.1110E+02 2.7759E+05 7.6938E+05 5.3167E+08 Amendment G l' April 30, 1990
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O 4000 - - O i l 3000 - - a uf 4 a: B 3 u. 5 2000 - - e .. h i 1000 - 169.90 LBM/SEC - AT 13.65 SECONDS I I I ! 0 O 5 10 15 20 TIME, SECONDS FEEDWATER ADDITION: PUMPED FEEDWATER FLOW BEFORE MSIV'S CLOSE Amendment G April 30,1990 O' MAIN STEAM LINE BREAK - 0% POWER gg LOSS OF ONE CSS TRAIN 6.2.1 29 Sheet 1 of 2 r
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$7d hu HEAT REMOVAL CAPACITY OF THE FAN COOLERS vs. CONTAINMENT ATMOSPHERIC TEMPERATURE 6.2.1 31
v 25000 i i i i Effective Annulus Spillage of SIT Injection
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-gyg7g COMBINED SPILLAGE AND SPRAY INTO CONTAINMENT
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I TIME OF ANNULUS l l DOWNFLOW l l Amendment G s, April 30,1990 tu Figure dfd hw CONDENSING HEAT TRANSFER COEFFICIENTS FOR STATIC HEAT SINKS 6.2.1 33 i ___--_:___-_-____-_---_----_______--______.. 1
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= MIN, CONTAINMENT PRESS. FOR ECCS PERF. ANAL. 6.2.1 34
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e 2 i CESSAR ;!!#,cuen
-l v) rs .
- i 6.3.2.6.3 Seismic Design Since operation of- the SIS 's i essential following a !
Loss-of-Coolant Accident, it is considered Category I for' seismic design. The general design basis for Category I equipment is that it must be able to withstand the appropriate seismic loads plus other applicable loads without loss of design functions which are required to protect the public. For the SIS, this means that the components must be able to 4 withstand the stresses resulting from emergency _ operation following a LOCA,_ simultaneous with the stresses resulting from the Safe Shutdown Earthquake (SSE) without loss of function. Refer to Section 3.7 for details on seismic design and analysis methods. 6.3.2.7 Reauired Manual Actions The short-term injection. mode of operation is automatically C initiated by a Safety Injection Actuation Signal.(SIAS). Long _ term core cooling is manually initiated at approximately 2
- hours post-LOCA at which - time the hot leg injection valves are opened to provide simultaneous hot leg and direct 'essel- l C injection, For whichbreaks, small pipe resultsthe in a SIcirculation flowmakeup Pumps provide throughforthe core. l spillage, C while the RCS is cooled down and depressurized to shutdown cooling initiation conditions utilizing the steam generator Atmospheric Dump Valves and Emergency Feedwater System. For ,
small LOCA's, the SITS must be vented to allow- RCS depressur-ization. This is followed by manual shutdown cooling operation. I Amendment C 6.3-21 June 30, 1988
l 2 CESSAR ;Encoc., ; 1 t i J L-I l - s l -. l THIS PAGE INTENTIONALLY BLANK i O e p , l-- l' i e i O 6.3-22
~ CESSAREin% m .
5 g 6.3.3 PERFORMANCE EVALUATION 6.3.3.1 Introduction and Summary r 10 CFR 50.46 provides.the Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Reactors (Reference 1). The Safety Injection System (SIS) performance analysis presented in this section demonstrates that the- System 80+ SIS design satisfies these criteria. , The analysis is performed for a complete spectrum of break sizes .! and locations. The limiting break, that which limits the - peak ; linear heat. generation rate (PLHGR), is identified as the 1.0 x DEG/PD*. The results of the analysis demonstrate, that for a PLHGR of. 13.7 kw/ft, the SIS design acets the 10 CFR 50.46 Acceptance-Criteria. Conformance is as follows: > Criterion (1) Peak Cladding Temperature. "The calculated maximum fuel element cladding temperature shall not exceed 2200*F". The spectrum analysis yields a peak cladding temperature of 2147'F for the 1.0 x DEG/PD' break. l ', l ( l Criterion (2) Maximum Cladding Oxidation. "The calculated total oxidation of the cladd'ng shall nowhere exceed l 1 0.17 times the total cladding thickness before oxidation". g The spectrum analysis yields a maximum cladding oxidation of 7.51% for the 1.0 x DEG/PD break. Criterion (3) Maximum Hydrogen Generation. "The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed'O.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react". The spectrum analysis yields a peak core-wide oxidation less than 0.843% for the 1.0 x DEG/PD break.
- DEG/PD = Double-Ended Guillotine at the Pump Discharge O
V Amendment G 6.3-23 April 30, 1990
.CESSAR1nhen l
l Criterion (4) Coolable Geometry. " Calculated changes in core :! geometry shall be such- that the core -remains- , amenable.to cooling." i The cladding- swelling and rupture model which -is part of the Evaluation Model-(References 2 and 3) i accounts for the effects of changes in core > geometry if - such changes are. predicted to- occur. Adequate core cooling is demonstrated, with the l predicted core geometry changes. The analysis.is ' performed to the point where cladding temperatures ! are decreasing precluding any further cladding ! deformation. Therefore, a coolable geometry is r demonstrated. Criterion (5) Long Term Cooling. "After any calculated successful initial operation of the ECCS, the ; calculated core temperature shall be maintained at i an acceptably low value and decay heat shall be removed for-the extended period of time required by the long-lived radioactivity remaining in the Core". I The spectrum analysis shows that the ' rapid ' insertion of borated water from the SITS will
-suitably limit the peak clad temperature and cool the core within a short period of time.
l: Subsequently, .the safety in-)ection pumps will supply cooling water from the refueling water tank-to remove heat from the long lived radioactivity l l remaining in the core. A detailed analysis and G description of the long-term cooling performance is given.in Section 6.3.3.4. 6.3.3.2 Large Break Analysis 6.3.3.2.1 Mathematical Model The SIS . performance analysis reported in this section used the
'large break LOCA evaluation model described in Reference 2 and approved in Reference 12. The CEFLASH-4A (Reference 4) computer program determines the primary system flow parameters during tne blowdown phase, and the COMPERC-II (Reference 5) computer program determines the system behavior during the refill and reficod phases. The core flow and thermodynamic parameters from these two codes are input to the STRIKIN-II (Reference 6) computer program, which calculates the hot rod cladding temperature transient. The peak cladding temperature ar.d maximum cladding oxidation are obtained from the STRIKIN-II calculation. Also, input into STRIKIN-II are steam cooling heat transfer coefficients which are calculated using the HCROSS (Reference 14)
Amendment G 6.3-24 April 30, 1990
- C E S S A R a n 7ei m ,.
l w 1 and PARCH (References 11 and- 14) computer programs. The core-wide cladding oxidation is obtained from the results of both the STRIKIN-II and COMZIRC (Reference 5) computer programs, i 6.3.3.2.2 -Safety Injection System Assumptions The SIS consists of four Safety Injection (SI) pumps and four Safety Injection Tanks (SITS). Automatic operation of the -SI , pumps is actuated by either a low pressurizer pressure signal or l a high containment pressure signal. Flow is initiated from the i SITS by the opening of a check valve when the reactor vessel . downcomer pressure drops below the SIT pressure. SI flow is l delivered by Direct Vessel Injection (DVI) connections. l l An evaluation of possible single failures shows that no single l failure in the SIS or the diesel-generator system is the worst I condition for the large break analysis. For the limiting break l location (discharge leg), there is no single failure that results- l in an injection flow rate which cannot keep the downcomer filled j to the elevation of the discharge leg. Consequently, the ' assumption of no failure in the SIS is the worst condition, because it maximizes the safety injection spilling to containment l which minimizes the containment pressure. This, in turn, minimizes the core reflooding rate. Therefore, the - following SIS conditions are used in the large i break analysis: l A. Flow from all four SITS assuming maximum initial liquid i inventory. G B. Maximum flow rate from all four SI pumps. The SI pumps are assumed to begin injecting after the lower plenum and downcomer have been filled by the SITS. This maximizes the time to fill the lower plenum and downcomer. > l 6.3.3.2.3 Core and System Parameters The significant core and system parameters of the large break analysis are presented in Table 6.3.3.2-2. The PLHGR is selected 4 to occur in the top of the core, the limiting location as I identified in Appendix A of Supplement 3-P-A of Reference 2. A conservative4 beginning-of-life moderator temperature coefficient (+ 0.5 x 10 Ap/'F) is used in all large break cases. i The large break analysis accounts for 10% steam generator tube plugging which may occur during the plant's lifetime. Amendment G 6.3-25 April 30, 1990
CESSAR !!nificma l
~
The initial _ steady state fuel rod. conditions are determined as a e? . l I function of burnup using the FATES 3 computer program (Reference. L 7). .The limiting condition for SIS performance is determined to occur for a hot rod average burnup of - 34,000 MWD /MTU. A ) parameter study demonstrates that cladding- temperature and l oxidation are maximized at this burnup. The results of this- 1 study are presented in Figure 6.3.3.2-11. J 6.3.3.2.4 Containment Parameters Section 6.2.1.5 ' presents the containment parameters used in the l SIS performance ' analysis. The values for these parameters are chosen to minimize containment pressure such that a conservative j determination of the core reflood rate is made. 6.3.3.2.5 Break Spectrum The large break analysis is performed for nine . breaks. These breaks include slot and guillotine breaks,. ranging in size from 2 0.5 ft to full double-ended break area. Break locations-include the reactor coolant pump suction and discharge legs and the hotz leg. Table 6. 3. 3. 2-3. lists the various break sizes, types and locations analyzed. As previously demonstrated in Reference 2, the limiting break location is the pump discharge leg. Pump discharge leg breaks r are limiting because both the blowdown core flow rate and reflood rate are minimized for this location. 6.3.3.2.6 Results and Conclusions a The important results of this SIS performance analysis are i summarized in Table 6.3.3.2-4 and the figures listed in Tables 6.3.3.2-5 and 6.3.3.2-6. Times of interest for the various breaks analyzed are presented in Table 6.3.3.2-1. Peak cladding temperature vs. break size is shown in Figure 6.3.3.2-10. > These results demonstrate that the SIS for the System 80+ Standard Design is in compliance with the Acceptance Criteria of 10 CFR 50.46 (Reference 1), and is adequate to perform its intended function of maintaining the integrity of the core, thereby limiting radiation release to the environment. O Amendment G 6.3-26 April 30, 1990
.__ __ _ ._. .~
L l lCESSARin# enc. - (
.q) .
TABLE 6.3.3.2-1 TIME SEQUENCE OF IMPORTANT EVENTS FOR A SPECTRt31 0F LARGE BREAK LOCAs * (SECusqu5 AtIt.R BREAK) SI Tanks End of Start of SI Tanks. Hot Rod Break ON Bypass Reflood Empty Rupture , 1.0 DES /PD* 13.2 27.9 37.76 93.80 52.8- 3 0.8 DES /PD 14.5 28.7 38.48 95.04 52.6 0.6 DES /PD 16.9 30.2 39.95 97.37 66.2 2 0.5 FT S/PD 183.5- 190.8 199.31 264.11 190.9 G 1.0 DEG/PD 14.3 28.5 38.20 94.96 47.3 0.8 DEG/PD 15.8 29.3 39.07 96.47 64.5 0.6 DEG/PD 18.7 31.0 40.59 99.20 67.9
] 1.0 DEG/PS 16.1 23.9' 32.88 96.57 63.4 1.0 DEG/HL 7.0 10.6 19.13 85.71 46.0-l L
l- - l '.
- See Table 6.3.3.2-3 for an explanation of these abbreviations.
1 O Amendment G April 30, 1990
7 ,
't
- l. '
L CESSAR ;!Micam,. i
- V i i- TABLE 6.3.3.2-2 l
GENERAL SYSTEM PARAMETERS Am INITIAL COMITIONS LARGE BREAK SIS PERFORMMCE i Quantity Value Units a 3876 ! Reactor Power Level (102% of Nominal) MWt Average Linear Heat Generation Rate 5.7 kw/ft (102% of Nominal) Peak Linear Heat Generation Rate 13.7 kw/ft l Gap Conductance at Peak Linear Heat 2697 Btu /hr-ft2_.y , Generation Rate
- Fuel Centerline Temperature at Peak Linear 3158 'F Heat Generation Rate
- Fuel Average Temperature at Peak Linear 1934 'F Heat Generation Rate *
. Hot Rod Gas Pressure 2224 psia v Moderator Temperature Coefficient at + 0.5 x 10 ap/*F Initial Density 6
l System Flow Rate-(Total) 165.6 x 10 lbs/hr 6 Core F. low Rate 160.7 x 10 - lbs/hr a Initial System Pressure 2250 psia Core Inlet Temperature 558 'F Core Outlet Temperature 618 'F Active Core Height 12.5 ft. Fuel Rod OD 0.382 in. Number of Cold Legs 4 Number of Hot Legs 2
. Cold Leg Diameter 30 in.
Hot Leg Diameter 42 in. Safety Injection Tank Pressure 584.7 psia 3 Safety Injection Tank Gas / Water Volume 806/1600 ft r These quantities correspond to the burnup (34,000 MWD /MTV, hot rod average) Oe
' yielding the highest peak cladding temperature.
Amendment G April 30, 1990
h CESSAR%"icuen t . TABLE 6.3.3.2-3 LARGE BREAK SPECTRLM Break. Size Type, and Location Abbreviation Figure No. 1.0 x Double-Ended Slot Break 1.0 DES /PD 6.3.3.2-1 In Pump Discharge Leg 0.8 x Double-Ended blot Break 0.8 DES /PD 6.3.3.2-2 In Pump Discharge Leg 0.6 x Double-Ended Slot Break 0.6 DES /PD 6.3.3.2-3 In Pump Discharge Leg 2 2 0.5 ft Slot Break in Pump 0.5 ft S/PD 6.3.3.2-4 Discharge Leg 1.0 x Double-Ended Guillotine 1.0 DEG/PD 6.3.3.2-5 ' Break in Pump Discharge leg O.8 x Double-Ended Guillotine 0.8 DEG/PD 6.3.3.2-6 Break in Pump. Discharge Leg 0.6 x Double-Ended Guillotine 0.6 DEG/PC 6.3.3.2-7 Break in Pump Discharge Leg 1.0 x Double-Ended Guillotine 1.0 DEG/PS 6.3.3.2-8 Break in Pump Suction Leg 1.0 x Double-Ended Guillotine 1.0 DEG/HL 6.3.3.2-9 Break in Hot Leg l l l O
CESSAR !!!Enemt. I
.b TABLE 6.3.3.2-4 1
PEAK CLAD TEMPERATURE Als OXIDATION PERCENAGE FORME BREAK SPECImm t i Peak Cladding Maximum Cladding Maximum Core-Wide Break Temperature (*F) Oxidation (5) Oxidation (5) i 1.0 DES /PD 2135 7.34 < 0.816 0.8 DES /PD 2135 7.18 < 0.819 0.6 DES /PD 2051 5.80 < 0.787 2 0.5 FT S/PD 1960 4.52 < 0.522 1.0 DEG/PD 2147 7.51 < 0.843 0.8 DEG/PD 2065 6.03 < 0.816' O.6 DEG/PD 2052 5.81 < 0.775
. 1.0 DEG/PS 1756 1.71 < 0.213 0
1.0 DEG/HL 1705 1.49 < 0.076 l l' l l O Amendment G April 30, 1990
- s. .
LCESSARiinLua j 1 r i ift TABLE 6.3.3.2-5 6 VARIABLES PLOTTED AS A FUNCTION OF TIME. FOR EACH LARGE BREAK IN THE S N Variables Designation ! Core Power 'A Pressure in Center Hot- Assembly Node B
- Leak Flow C-Hot Assembly Flow (below hot spot) D.1- [
Hot Assembly Flow (above hot spot) D,2-Hot Assembly Quality E Containment Pressure F l Mass Added to Core During Reflood G Peak Clad Temperature H*' ; i i s G
- For the limiting break, the temperature of the rupture node is also shown.
O Amendment G April 30, 1990
r E TCESSARiinL mr q i TABLE 6.3.3.2-6 O ADDITIONAL VARIABLES PLOTTED AS A FUNCTION OF TIME FOR THE LIMITING LARGE BREAK Figure Variables Designation , Mid Annulus Flow I j Qualities Above and Below the Core -J Core Pressure Drop K U Safety Injection Flow into Reactor Vessel L Water Level in Downcomer During Reflood M ! Hot Spot Gap Conductance- N l Local Clad Oxidation 0 Clad Temperature, Centerline Fuel Temperature, P
.\ Average Fuel Temperature and Coolant Temperature 1 for Hottest Node ';
l Hot ~ Spot Heat Transfer Coefficient Q l l Hot Pin Pressure R Core Bulk Channel Flow Rate S l l l 0 - Amendment G April 30, 1990
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p~ , 6.3.3.6 ~ Radiological Consequences Following a postulated double-ended ' rupture of a' reactor' coolant pipe with i ' subsequent blowdown,'the ECCS limits the clad temperature to well below the melting point and ensures that the reactor; core remains intact and in a
, coolable geometry.. minimizing the~ release of fission products to the contain-7:;
mont. However, a hypothetical accident involving a significant release of J fission products to the containment is. evaluated.- It-is assumed that"100% of the' nobel, gas and 50% of the iodine equilibrium core saturation fission product inventory is inmediately released to the
$
- containment atmosphere. 0f the: iodine released to the containment, 50% is T '
. assumed to plate out'onto the internal surfaces of the-containment or.
adhere'to-internal components. The remaining iodine and. the noble gas activity is assumed to be immediately available for leakage from the p' containment. The source terms and associated assumptions are itemized in Table 6.3.3.6-1. The following specific assumption: were used in the analysis.
- 1. The reactor core equilibrium noble gas and iodine inventories are based on long-term operation at the ultimate core power level.
.2. One hundred percent of the core equilibrium radioactive noble gas inventory is .innediately available for leakage from the containment.
t.
- 3. Fifty percent of the core equilibrium radioactive iodine inventorytis O innediately released to the containment atmosphere. Half is plated out onto the internal surfaces of the containment and the other half is
-available for leakage'from the containment, d i
- 4. 'Of the iodine fission product inventory releas'ed to the containment, 91% is in the form of particulate iodine, and 4% is in the form of organic iodine.
The doses associated with these source terms.will be provided in the Appli-cant's SAR. I l s I 6.3-39
CESSAR EMMncui: l l O 1 I I l 1 I T 1 THIS PAGE INTENTIONALLY BLANK i
. O.
i I < l. l 9 l Amendment G 6.3-40a April 30, 1990
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- 6.3.3.79 Cha'pter..- 15 Acciden't Analysis
~
For the' limit.ing' event in Chapter 15, the safety. systems. actuated are listed i in . tables designated 15.X.X.X-3, " Utilization of. Safety. Systems". , The events - 3
, s;3 .which result in; safety injection actuation-are identified in Table 6.3.3.7-1. .
t b
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6.3-40b July 16,1981 l4 is
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C E S S A R in e.c.m . 1 O REFERENCES FOR SECTION 6.3.3 l
- 1. " Acceptance Criteria for Emergency Core Cooling Systems for LightWater Cooled Nuclear Power Reactors," Federal Register, Vol. 39, No. 3, Friday, January 4, 1974.
l
- 2. " Calculative Methods for the C-E Large Break LOCA Evaluation ]
Model," CENPD-132, August 1974 (Proprietary). I
" Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132, Supplement 1, February 1975 (Proprietary).-
" Calculational Methods for the C-E Large Break LOCA Evaluation Model," CENPD-132, Supplement 2, July 1975 (Proprietary).
" Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS," G CENPD-132, Supplement 3-P-A, June 1985 (Proprietary).
- 3. " Calculative Methods for the C-E.Small Break LOCA Evaluation Model," CENPD-137, August 1974 (Proprietary).
" Calculative Methods for the C-E Small Break LOCA Evalut. tion Model," CENPD-137, Supplement 1, January 1977, (Proprietary).
r
- 4. "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis," CENPD-133, April 1974 (Proprietary). ,
"CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis (Modification) , " CENPD-133, Supplement 2, February 1975 (Proprietary).
"CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis," CENPD-133, Supplement 4-P, April 1977 (Proprietary). "
"CEFLASH-4A, A FORTRAN 77 Digital Computer Program for Reactor Blowdown Analysis," CENPD-133, Supplement 5-P, June 1985 (Proprietary).
- 5. "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," CENPD-134, August 1974 (Proprietary).
O i Amendment G
- 6. 3-40c April 30, 1990
CESSAR !!!#,co,on O
"COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modifications)," CENPD-134, Supplement 1, February 1975 (Proprietary).
"COMPERC-II, A Program for Emergency Refill-Reflood of the G Core," CENPD-134, Supplement 2, June 1985.
- 6. "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," CENPD-135, August 1974 (Proprietary).
"STRIK1H-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modification)," CENPD-135, Supplement 2, February 1975 (Proprietary). -
l "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer g ) Program," CENPD-135, Supplement 4, August 1976 ! (Proprietary).
'"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer !
Program," CENPD-135, Supplement 5-P, April 1977 (Proprietary).
- 7. "C-E Fuel Evaluation Model," -CENPD-139, July 1974 (Proprietary).
" Improvements to Fuel Evaluation Model," CEN-161 (B)-P, July 1981 (Proprietary). G
- 8. System 80 CESSAR FSAR, Docket No. STN 50-470.
- 9. " Post-LOCA Long Term Cooling Evaluation Model," CEl'PD-2 5 4 ,
June 1977 (Proprietary).
- 10. "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the small Break Loss-of-Coolant Accident," ,
CENPD-133, Supplement 1, August 1974 (Proprietary). t "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident," CENPD-133, Supplement 3, January 1977 (Proprietary).
- 11. " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," CENPD-138, August 1974 (Proprietary).
" PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup (Modifications) , "
CENPD-138, Supplement 1, February 1975 (Proprietary). Amendment G 6.3-40d April 30, 1990 l
m ?):;f
- 4 LCESSAR M ncamn
, [- .
" PARCH, A . FORTRAN-IV Digital Program to Evaluate Pool Boiling,. -Axial Rod and' Coolant Heatup," CENPD-138,,
Supplement 2, January 1977 (Proprietary). 12 . ' Letter,_ 0. D. P3rr (NRC) to F. M. . Stern (C-E), June 13, 1975. Letter, O. D..Parr (NRC) to A. E. Scherer- (CE) , December 9,_ 1975. Letter,- Karl Kniel (NRC) to A. E. Scherer. (C-E), September 27,.1977.
.Lettar, D. M.- Crutchfield (NRC) to A. E. Scherer (CE), July G 31, 1986.
- 13. Letter, Robert L. Baer (NRC) to A. E. .Scherer (C-E),
July 30, 1979.
- 14. LD-81-095, "C-E_ ECCS Evaluation Model -Flow Blockage' G
Analysis," 'Decen%ber 1981. h 9 Amendment G 6.3-40e April 30, 1990
CESSAR !!8har 1 EFFECTIVE PAGE LISTING i APPENDIX A
-Table of Contents Page Amendment Overview. F- j 1 F ';
.11 F Text !
Page Amendment j
,i l
A-1 F A-2 F. A-3 G i A-4 F-A F ,i A-6 F
-A.7 G O A A-9 A-10 F
F F =l A-11 F 1
'A-11a G A-11b G !
A-11c' G l
.A-11d' G .;
A-12 F A-13 F i A-14! F. .] A-15 F.
- A-16 F A-17 F i A-18 F A-19 F A-20 F
, A-21 F A-22 F A-23 F A-24 F -;
A-25 F A-26 F A-27 F ( :- A-28 F A F
- A-30 F Amendment G. l l
April 30, 1990
~ ,
;CESSAREinLme O
EFFECTIVE PAGE LISTING'(Cont'd)
-c ,
APPENDIX A Tert (Cont'd) Page ' Amendment A-31 F A-32' F A-33 F A-34= F A-35. F A-36' F A F A-38 F. A-39 F A-40 F A F A-42. F
', .A-43 p A-4 4 - F A-45 F A-4L<
~
G A-45b. G A-45c G
-A-45d G
- A-46 .F
.A' F A-47a G A-47b G
-A-47c'. G A-47d G A-48 F
~A-49: F A-50 F A F A-52 F A-53 F A F i A-55 F A-56 F A F A-58 P-A-59 F A-60 F A-61 F
.A-62 F A-63 F -
A-63a G A-63b G Amendment G April 30, 1990 4
_(ShSot-3_'ef 7). 1;CESSARHuhi:=
-i O
EFFECTIVE PAGE LISTING-(Cont'd) APPENDIX A ! Text (Cont'd)
~
Page' ,4aendment. f A-63c G
, A-63d G 'n 'A-64 F 1 A-65, F- !
A-M F A-CM G
. A- 6 b r;- G" .
A-66c G i A-66d- G' '} A-67 F l A-68 F i
'A-69 F l
.A F :l
. A F '
A-72. F i O -A-73 A-74
'A F F'
F i
=A-76 F '!'
A-77.' F A-78 F !
, 'A-79 F
-A-80' F 1 A-81 F o A-82' F
'A-83 F A-84 F
~A-85 F A-86 F A-87 F A-88 F A-89 F ,
A F ! A-91'- F A-92 F 1 A-93 F A-94 F A F A-96 F A-97~ F A-98 F O- A-99 A-100 F F Amendment G April 30, 1990 l
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EFFECTIVE PAGE LISTING-(Cont'd)- [ APPENDIX A Text'(Cont'd). -
- Page Ainendment- !
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A-100a G
-A-100b. G i A-100c G !
A-100d G j A-1011- F
.A-102' F r A-102a G I
A-102b- 'G
.; A-102c' G
+A-102d G A-103 F i A-104- F ~
[ A-105: F ! A-105a G .l A-105b G .
- A-105c G
.A-105d G l A-105e G cA-105f G I A-106- F
~A-107- F A-108 ' F
.A-108a G !
.A-108b G' g A-109 F
- , .A-110 F
? A-111 F l
. A-112 F s A-113 F A-114 F A-115 F- .
r.A-116- F
'A-117 F F
A-118
; -A-119. F L A-120 F A-121 F >
'A-122 F l A-123 F A-123a G l
.A-123b o '
A-123c G A-123d G Amendment G April 30, 1990 L .
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- CESSARMnem:a: >
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~ EFFECTIVE PAGE LISTING (Cont'd)
APPENDIX A ! i Text (Cont'd) ] , Page Amendment .l
'A-123e G'
'. A-12 3 f .G- 4
.A-123g G A-123h- G ,
-A-124 F
.A-125 F A-126 y .
A-127 F , A-128 F , A-129 F 1
.A-130 F i
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'A-132 F A-133 F
'A-134' F A-135 P'
, M A'-13 6 ' F. l-
- A 137" F A-138 .F A-139 F I
-A-140 F
- i 141
- A y_
-A-142' 'F A-143. F A-144 F A-145 F .;
A-146 -F A-147 F
- A-14 8 - F
* ;A-149 F R -A-150. F l= . A-150a 'G
. A-150b G
,A-150c G
-A-150d. G l[ A-150e- G L A-150f G H A-150g G A-150h G A-151 F
'2 r 'd O A-152
'A-153.
F F A-154 F Amendment G April 30, 1990
- u. .
(Shast 6 of 7).
. ,CESSAR E!NPicam, 1
.O1 i
EFFECTIVE PAGE LISTING'-(Cont'd) . a APPENDIX A .
. Text.(Cont'd)
Page ' Amendment-
,, 1 0 A-155 F.
.A-156-F ,
, 1 A-157 'F A-158. F' I iA-159- G ,I A-160 G A-161- G A-162 G A-163- G 1 A-164 G
' Tables' Amendment Al-1L(Sheet 1)- F- .;- >
Al-1L(Sheet 2) F- - I
'Al-1~(Sheet 3). F- 1 j
'Al (Sheet , 4 )- F Al-1-(Sheet 5) F lAl-1;(Sheet 6) F Al-1~(Sheet 7)-
-- F Al-1'.(Sheet 8) F Al-1;- (Sheet' 9) ' F
- Al-1-1(Sheet.10) F Al-1 (Sheet-11) F Al-1-(Sheet 12) F ,
t,: -Al-11.(Sheet 13)~ F
-Al-1 (Sheet 14)
F Al-1.(Sheet 15) F Al-1 (Sheet 16) F Al-1 (Sheet 17) F Al-1.(Sheet 18) F
-Al-1 (Sheet 19) F ,
Al-1 (Sheet'20) G Al-1-(Sheet 21) F Al-1 (Sheet 22) F Al-1 (Sheet 23) F
;Al-1 (Sheet 24) F Al-1 (Sheet 25) F Al-1 (Sheet 26) G Al-1-(Sheet 27) F Al-1 (Sheet 28) F Al-1'(Sheet 29) G Amendment G April 30, 1990
- l. (Sh00t 7'Cf',7)-
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t CESSARTinhen
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9 > EFFECTIVE PAGE LIST 3M (Corit'd)'
' APPENDIX 1g-Tables;(cont'd) -Amandaant
- .Al (Sheet 30)- G Al-1;(Sheet 31)- F
'Al-1 (Sheet-32)' F. 'i~
'Al-1 (Sheet:33) F Al-1'(Sheet 34)' G-m_
. Al (Sheet3 5) ^ F
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2 Al-1, (Sheet ~'3 6) F Al-1.(Sheet-37)' F-
'Al-l'(Sheet 38)
F
, Al-1-(Sheet.39) G- 1 Al-l'r(Sheet 40)
F i
-Al (Sheet'41) G d Al-1'(Sheet.42) F-Al-1 (Sheet 43) F
-Al-1 (Sheet 44) F Al-1:(Sheet'45) F Al-1 (Sheet _46) F
.- Al-1 (Sheet 47) F Al-1 (Sheet _48) F Al-1 (Sheet.49) F Al-1'(Sheet 50)' F
'Al-1 (Sheet'51) -
F Al-11(Sheet 52) =F l 4 Al-1-(Sheet-52) F Al-1 -(Sheet 54) 'F Al-1 (Sheet 55) F 4 A2-1-(Sheet 1). F
.A2-1-(Sheet 2) F A2'-1.(Sheet 3) F A2-1"(Sheet 4) G l'
A2-l' (Shetzt 5) G A2-1 -(Sheet 6) G A2-1 (Sheet 7) G A2-1-(Sheet 8) G A2-1-(Sheet 9) F A2-1 (Sheet 10) F
< A3-1 (Sheet 1) G A3-1 (Sheet 2) F A4-1 (Sheet 1) G A4-1 (Sheet 2) G
[m^ A4-1.(Sheet 3)_ G A4-1 (Sheet 4) G A4-1 (Sheet 5) G A4-1-(Sheet 6) G Amendment G April 30, 1990
1 I iCESSARlin%u... J
.?
,/ ~ '. -
i 1 ( ? ! TABLE Al-l'(Cont'd) (Sheet 19 of 55) ;
-LISTING OF .
UNRESOLVED SAFETY ISSUES AND <' i GENERIC SAFETY ISSUES-ISSUE ISSUE NUMBER- ISSUE TITLE TYPE CATEGORY-i
'141 LBLOCA WITH CONSEQUENTIAL GSI' la SGTR (
142' LEAKAGE THROUGH GSI la a ELECTRICAL ISOLATORS 143 AVAILABILITY OF CHILLED GSI la WATER SYSTEMS 144 SCRAM WITHOUT.A GSI la TURBINE / GENERATOR TRIP i l
. N-{} ~ 145 IMPROVE SURVE!f, LANCE AND GSI lb STARTUP TESTING PROGRAMS '
A-1 WATER-HAMMER USI 2 1 A-2 ASYMMETRIC BLOWDOWN LOADS USI 2 1 ON RCS A-3 WESTINGHOUSE STEAM USI lb GENERATOR TUBE INTEGRITY A-4 C-E STEAM GENERATOR TUBE USI 2 INTEGRITY- i A-5 B&W' STEAM GENERATOR TUBE USI lb INTEGRITY A-6 MARK I SHORT-TERM PROGRAM USI lb A-7 MARK I LONG-TERM PROGRAM USI lb A-8 MARK II' CONTAINMENT POOL USI lb DYNAMIC LOADS--LONG TERM l ,_ PROGRAM
- 1. [8-)\
c l Amendment F December 15, 1989
CESSARilnincamn
-1 Y
'0 TABLE-Al-1 (Cont'd)-
(Sheet20 of 55) LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE = ISSUE-NUMBER: ISSUE TITLE ' TYPE CATEGORY-A-9 ANTICIPATED' TRANSIENTS WITHOUT USI 2 SCRAM'(ATWS) A-10 BWR FEEDWATER NOZZLE USI lb. CRACKING A-11 . REACTOR VESSEL. MATERIAL USI. lb TOUGHNESS A-12 -FRACTURE TOUGHNESS OF USI' 2 STEAM GENERATOR'& RCP SUPPORTS x A SNUBBER OPERABILITY -GSI 2 ASSURANCE A FLAW-' DETECTION GSI lf A-15 PRIMARY' COOLANT SYSTEM GSI le lG DECONTAMINATION AND STEAM GENERATOR CHEMICAL
' CLEANING A-16 STEAM EFFECTS-ON BWR CORE GSI lb SPRAY' DISTRIBUTION A-17 SYSTEMS INTERACTION USI- 2 A-18 PIPE RUPTURE DESIGN GSI lf~
CRITERIA A-19 DIGITAL COMPUTER GSI id PROTECTION SYSTEM A-20 IMPACTS OF THE COAL FUEL GSI 1d CYCLE O Amendment G l April 30, 1990 { l
- CESSAR !!nbio,. 1 J/^Y N i _,/.
TABLE Al-1 (Cont'd) q (Sheet 25 of 55) LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ', s ISSUE ISSUE NUMBER ISSUE TITLE . TYPE CATEGORY <
'l
.B-20 STANDARD PROBLEM ANALYSIS GSI/LI ld i
B-21 CORE PHYSICS GSI/LI 1d ; B-22 IMR' FUEL GSI la
^
B-23 UHFBR FUEL- GSI/LI 1d e-B SEISMIC QUALIFICATION OF GSI le , ELECTRICAL AND MECHANICAL COMPONENTS
/-
\# B-25. . PIPING BENCHMARK PROBLEMS GSI/LI ld
- B-26 STRUCTURAL INTEGRITY OF =GSI '1c
. CONTAINMENT PENETRATIONS ;
B-27 . IMPLEMENTATION AND USE OF GSI/LI ld SUBSECTION NF-4 B-28 RADIONUCLIDE/ SEDIMENT GSI ig-TRANSPORT PROGRAM B-29 EFFECTIVENESS OF ULTIMATE GSI/LI- la i HEAT SINKS l B-30 DESIGN - BASIS FLOODS AND GSI/LI ld L PROBABILITY { c B-31 DAM FAILURE MODEL GSI/RI ld ,..e, B-32 ICE EFFECTS ON SAFETY- GSI la RELATED WATER SUPPLIES l B-33 DOSE ASSESSMENT GSI/LI id METHODOLOGY
. ['y
% ,/ -
Amendment F December 15, 1989
CESSAR innnenia O TABLE Al-1 (Cont'd) (Sheet 26 of 55) LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE ISSUE NUMBER ISSUE TITLE TYPE CATEGORY B-34 OCCUPATIONAL RADIATION GSI le EXPOSURE REDUCTION B-35 CONFIRMATION OF APPENDIX GSI/LI ld I MODELS FOR CALCULATIONS OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS AND LIQUID EFFLUENTS FROM LWRs B-36 DEVELOP DESIGN, TEST, GSI ld lG MAINTENANCE CRITERIA FOR ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ABSORPTION UNITS FOR ESF SYSTEMS AND FOR NORMAL VENTILATION SYSTEMS B-37 CHEMICAL DISCHARGE TO GSI/RI ld RECEIVING WATERS B-38 RECONNAISSANCE LEVEL GSI lg INVESTIGATIONS B-39 TRANSMISSION LINES GSI lf B-40 EFFECTS OF POWER PLANT GSI lf ENTRAINMENT ON PLANKTON B-41 IMPACT ON FISHERIES GSI lf B-42 SOCIOECONOMIC GSI ld ENVIRONMENTAL IMPACTS B-43 VALUE OF AERIAL GSI lg PHOTOGRAPHS FOR SITE EVALUATION B-44 FORECASTS OF GENERATING GSI lg COSTS OF COAL AND NUCLEAR PLANTS Araendme nt G April 30, 1990
e < g .CESSAR T Encim,.
?~ v !
a ) s NJ . i TABLE Al-1-(Cont'd)
-(Sheet 29 of 55) ,
LISTING OF l UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ; i
-ISSUE lBSUE ISSUE TITLE- $YPE CATEGORY EMMBER B-65 ' IODINE SPIKING GSI lf i
'i B-66 CONTROL ROOM INFILTRATION GSI 2 MEASUREMENTS B-67 : EFFLUENT AND PROCESS GSI le MONITORING INSTRUMENTATION <
B-68 PUMP OVERSPEED DURING GSI lf , LOCA
,(
N B-69 ECC LEAKAGE GSI le i EX-CONTAINMENT B-70 POWER GRID FREQUENCY GSI le l DEGRADATION AND EFFECT ON l PRIMARY COOLANT PUMPS INCIDENT. RESPONSE GSI le -j B-71 B-72 DEVELOPMENT OF MODELS FOR GSI/LI ld ASSESSING RISK OF HEALTH AND LIFE SHORTENING FROM URANIUM AND COAL FUEL CYCLE
'B-73 MONITORING FOR EXCESSIVE GSI le VIBRATION INSIDE THE REACTOR VESSEL C-1 ASSURANCE OF CONTINUOUS GSI le lG LONG-TERM CAPABILITY OF HERMETIC SEALS ON INSTRUMENTATION AND ELECTRICAL EQUIPMENT O
Amendment G April 30, 1990
e a 1. T LC E S S A R iin k m.c 4
. TABLE Al-1 (Cont'd)
(Sheet 30 of 55) j 1 l LISTING OF l UNRESOLVED SAFETY ISSUE 8'AND .j GENERIC SAFETY ISSUES j l 3 ISSUE ISSUE i NUMBER ISSUE TITLE TYPE CATEGORY j C-2 STUDY OF CONTAINMENT GSI- 2 l DEPRESSURIZATION BY j INADVERTENT SPRAY OPERATION l,
]
C-3 INSULATION USAGE WITHIN GSI le CONTAINMENT t C STATISTICAL METHODS FOR- GSI/RI 2 i ECCS ANALYSIS > C-S' DECAY HEAT UPDATE GSI/RI 2 lh C-6= LOCA HEAT SOURCES 1 GSI/RI- ld lg C-7 PWR' SYSTEM' PIPING- GSI 1c ! C-8 MAIN STEAM LINE LEAKAGE GSI lb ;- CONTROL SYSTEMS (IN BWRs) l l C-9 RHR HEAT EXCHANGER TUBE- GSI- If FAILURES C-10 EFFECTIVE OPERATION OF GSI 2 CONTAINMENT SPRAYS IN A , LOCA
-C-11 ASSESSMEU F OF FAILURE AND GSI 1c RELIABILITY OF PUMPS AND VALVE
'C PRIMARY SYSTEM VIBRATION GSI 2 ASSESSMENT C-13 NON-RANDOM FAILURES GSI le O
Amendment G April 30, 1990 J
y . 3 @ESSAR 834%u.. . I [ \: ; 5,,,N TABLE Al-1 (Cont'd) (Sheet 31 of 55) LISTING OF UNRESOLVFC SAFETY ISSUES AND r GFt!dRIC SAFETY ISSUES i i ISSUE ISSUE !
' NUMBER ISSUE TITLE TYPE CATEGORY. ,;
' ~
C-14 STORM SURGE MODEL-FOR GSI/LI lf. COASTAL SITES C-15 NUREG REPORT.FOR LIQUID GSI/LI ld l TANK. FAILURE ANALYSIS C-16 ASSESSMENT OF GSI- lg AGRICULTURAL LAND IN. L RELATION TO POWER PLANT SITING AND COOLING SYSTEM .i
;7 s SELECTION 0 !\
- ' =
~C-17 INTERIM ACCEPTANCE GSI ig l CRITERIA FOR SOLIDIFICATION AGENTS-FOR RADIOACTIVE SOLID WASTES D-1 ADVISABILITY OF A SEISMIC GSI la ;
SCRAM-D-2 EMERGENCY CORE COOLING -GSI la SYSTEM CAPABILITY FOR ! FUTURE PLANTS i i D-3 CONTROL ROD DROP ACCIDENT GSI 1c l- HF 1.1 SHIFT STAFFING GSI 1di K .HF 1.2 ENGINEERING EXPERTISE ON GSI ld SHIFT i HF.1.3 GUIDANCE ON LIMITS AND GSI ld CONDITIONS OF SHIFT WORK l HF 1.3.1 HUMAN FACTOR PROGRAM GSI ld x PLAN--TRAINING Arendment F December 15, 1989
ae e - 1 s CESSAR Unanio,. - 1 , t'. f TABLE Al-1 (Cont'd) 41 { r (Sheet 32 of 55) LISTING OF .
. UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES I
P , ISSUE ISSUE NUMBER ISSUE TITLE TYPE CATEGORY HF 1.3'.2' HUMAN FACTORS PROGRAM GSI- Id PLAN--LICENSING EXAMINATIONS t i' EHF 1.3.3- ' HUMAN FACTORS PROGRAM GSI 1d PLAN--PROCEDURES-OPERATING
.AND' MAINTENANCE.
.HF 1.3.4a HUMAN FACTORS PROGRAM GSI 2 PLAN - MAN MACHINE '
INTERFACE.- LOCAL CONTROL -
' STATIONS .
HF.1.3.4b HUMAN FACTORS PROGRAM GSI 2 , PLAN - MAN = MACHINE *
' INTERFACE - ANNUNCIATORS ,
.1 HF.1.3;4c- HUMAN FACTORS PROGRAM GSI 2 PLAN - MAN MACHINE L
INTERFACE - OPERATIONAL AIDS-HF 1.3.4d HUMAN FACTORS PROGRAM GSI 2 PLAN - MAN MACHINE ' INTERFACE - AUTOMATION I AND ARTIFICIAL l' INTELLIGENCE HFE1.3.4e HUMAN FACTORS PROGRAM GSI 2 PLAN - MAN MACHINE INTERFACE - COMPUTERS AND L COMPUTER DISPLAYS HF 1.3.5 HUMAN FACTORS PROGRAM GSI id PLAN--STAFFING AND L QUALIFICATIONS , lL i !- Amendment F December 15, 1989 l I l 1
f ,, CESSAR1!NLm I TABLE Al-1 (Cont'd) { t (Sheet 33 of 55) l LISTING OF j UNRESOLVED SAFETY ISSUES AND l GENERIC-SAFETY-ISSUES ': I
. ISSUE ISSUE .
NUMBE'R ISSUE TITLE TYPE CATEGORY ! HF 1.3.6 HUMAN FACTORS PROGRAM GSI id l PLAN--MANAGEMENT AND ORGANIZATION HF 1.3.7 lH'UMAN FACTORS PROGRAM GSI 1d PLAN--HUMAN' PERFORMANCE !
- HP 2.1 EVALUATE INDUSTRY GSI id- f
' TRAINING i
1 HF.2.2 EVALUATE'INPO GSI id o ACCREDITATION l HF - 2.'3 REVISE SRP SECTION 13.2 'GSI id HP 3.1 DEVELOP ' JOB KNOWLEDGE GSI id l CATALOG I HF 3.2 . DEVELOP LICENSE GSI id EXAMINATION HANDBOOK i HF 3.3 DEVELOP CRITERIA FOR GSI ld i NUCLEAR POWER PLANT SIMULATORS HF 3.4 EXAMINATION REQUIREMENTS GSI id HF 3.5 DEVELOP COMPUTERIZED EXAM GSI ld SYSTEM HF 4.1 INSPECTION PROCEDURE FOR GSI ld UPGRADED EMERGENCY OPERATING PROCEDURES HF 4.2 PROCEDURES GENERATION GSI ld
. PACKAGE EFFECTIVENESS EVALUATION Amendment.F December 15, 1989
, x CESSARE"icur., l l
9; TABLE A1-1 (Cont'd)
.(Sheet 34 of 55)
LISTING OF l HERESOLVED SAFETY ISSUES AND , EENERIC SAFETY ISSUES l i ISSUE ISSUE NUMBER ISSUE TITLE TYPE CATEGORY HF 4. 3 CRITERIA FOR GSI ld SAFETY-REuATED OPERATOR ACTIONS , i HF 4. 4' GUIDELINES FOR UPGRADING GSI 1d OTHER PROCEDURES , t HF 4.5 APPLICATION OF AUTOMATION AND GS1 le G ARTIFICIAL INTELLIGENCE HF 5.1 LOCAL ^ONTROL STATIONS GSI 2 , HF 5.2 REVIEW CRITERIA'FOR= HUMAN GSI 2 - FACTORS ASPECTS OF ADVANCED-CONTROLS AND~ INSTRUMENTATION HF 5.3 EVALUATION OF OPERATIONAL GSI id AID SYSTEMS HF 5.4 COMPUTERS AND COMPUTER GSI ld ll DISPLAYS , HF 6.1 DEVELOP. REGULATORY GSI 1d POSITION ON-MANAGEMENT AND ORGANIZATION HF;6.2 REGULATORY POSITION ON GSI ld MANAGEMENT AND ; ORGANIZATION'AT OPERATING REACTORS HF 7.1 HUMAN ERR 0; DATA GSI 1d ACQUISITION HF 7.2 HUMAN ERROR DATA STORAGE GSI ld i AND RETRIEVAL ,
.HF 7.3 RELIABILITY EVALUATION GSI 1d SPECIALIST AIDS Amendment G ,
April 30, 1990
l CESSARiHWieu. "([]) ! TABLE A1-1 (Cont'd) ' (8heet 37 of 55) '! LISTING OF HERE93dED_.EhllTl_lREP.RS_.AHQ . GENERIC SAFETY ISSUES ISSUE ISSUE MUMBER ISSUE TITLE TYPE CATEGORY I.A.4.2 (3) TRAINING SIMULATOR GSI/TMI id , IMPROVEMENTS -- LONG TERM
-l I.A.4.2 (4) TRAINING SIMULATOR GSI/TMI le IMPROVEMENTS -- LONG TERM !
i
-I.A.4.3 FEASIBILITY STUDY FOR GSI/LI- 1d PROCUREMENT OF TRAINING SIMULATOR I.A.4.4 FEASIBILITY STUDY TO GSI/LI id EVALUATE POTENTIAL'VALUE <
OF NRC ENGINEERING COMPUTER I . :B .1.1 - (1-4) ORGANIZATION'AND GUI/TMI le l MANAGEMENT - LONG TERM IMPROVEMENTS I.~B.1.1 (5)' MANAGEMENT FOR GSI ig OPERATION--LONG-TERM IMPROVEMENTS I.B.1.1 (6&7) ORGANIZATION AND GSI/TMI le MANAGEMENT - LONG TERM IMPROVEMENTS I.B.1.2 (1-3) MANAGEMENT FOR GSI/LI .g 1 OPERATIONS--EVALUATION OF NTOL APPLICANTS I.B.1.3 (1-3) MANAGEMENT FOR GSI/LI id OPERATIONS--LOSS OF SAFETY FUNCTION I.B.2.1 (1-7) REVISION OF IE GSI/LI id INSPECTION PROGRAM O Amendment F December 15, 1989 1
.t-CESSARiinL mr - 9 TABLE Al-1 (Cont'd) (Sheet 38 of.55) LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES ISSUE ISSUE NU'.iBER ISSUE TITLE TYPE CATEGORY
- 1. B . 2 . 2 - RESIDENT INSPECTORS AT GSI/LI ig OPERATING REACTORS I.B 2.3 INSPECTIONS 10D OPERATING GSI/LI 1d REACTORS--REGIONAL EVALUATIONS I.B.2.4 OVERVIEW OF LICENSEE GSI/LI id PERFORMANCE I.C.1 (1-4) SHORT TERM ACCIDENT GSI 2 A.iALYSIS AND PROCEDURES- -
MEVISION I.C.2 SHIFT AND RELIEF TURNOVER GSI ig PROCEDURES I.C.3 SHIFT SUPERVISOR GSI lg RESPONSIBILITIES I.C.4 OPERATING GSI ig PROCEDURES--CONTROL ROOM ACCESS I.C.5 PROCEDURES FOR FEEDBACK GSI/TMI 19 OF OPERATING EXPERIENCE I.C.6 PROCEDURE FOR GSI 1d VERIFICATION OF CORRECT PERFORMANCE OF OPERATING ACTIVITIES. I.C.7 NSSS VENDOR REVIEW OF GSI ig OPERATING PROCEDURES 9 Alendment F December 15, 1989
l
-- ! hh!h k!k !R$ ICATION. 'f 4
f I Y,}f ' TABLE Al-1 (Cont'd) (Sheet 39 of 55) , LISTING OF -
.t UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES (
. ISSUE ISSUE NUMBER ISSUE TITLE TYPE CATEGORY .
1 I.C.8 PILOT MONITORING OF GSI ig SELECTED EMERGENCY ; PROCEDURES'FOR NTOL APPLICANTS I.C.9 LONG-TERM PROGRAM PLAN FOR GSI le UPGRADING OF PROCEDURES
.1 l I.D.1 CONTROL ROOM DESIGN GSI/TMI 1d >
REVIEWS
'h I.D.2 ' CONTROL ROOM DESIGN -- GSI/TMI 2
' \d PLANT SAFETY PARAMETER DISPLAY CONSOLE b I.D.3 CONTROL ROOM DESIGN -- GSI/TMI ic lG SAFETY SYSTEM STATUS
; MONITORING I.D.4 CONTROL ROOM DESIGN -GSI 2 -i STANDARD I.D.5 '(1) CONTROL ROOM DESIGN -- GSI 2 IMPROVED INSTRUMENTATION RESEARCH - OPERATOR -
PROCESS COMMUNICATION I.D.5 (2) CONTROL ROOM DESIGN -- GSI 2 IMPROVED INSTRUMENTATION RESEARCH - PLANT STATUS AND POST-ACCIDENT MONITORING I.D.5 (3) CONTROL ROOM DESIGN GSI/LI 2
)
-- ON-LINE REACTOR SURVEILLANCE SYSTEM 1
eg i V l Amendment G April 30, 1990 , l l
n :c
?@ESSARiinincurs
-{
TABLE Al-1:(Cont'd) (Sheet 40 of 55) l LISTING OF UNRESOLVED SAFETY ISSUES AND GENERIC SAFETY ISSUES i ISSUE ISSUE j
. NUMBER ISSUE TITLE TYPE CATEGORY j I.D.5 (4) CONTROL ROOM DESIGN GSI 2 :
-- IMPROVED INSTRUMENTATION V RESEARCH - PROCESS MONITORING INSTRUMENTATION 1.D.5 (5) DISTURBANCE ANALYSIS GSI/LI le SYSTEMS I.D.6 CONTROL ROOM- GSI/LI 1d i DESIGN--TECHNOLOGY-TRANSFER CONFERENCE I.E.1 ESTABLISH OFFICE FOR ANALYSIS AND EVALUATION GSI/LI ld 9'i .
l OF OPERATIONAL DATA I.E.2 PROGRAM GSI/LI ld OFFICE--OPERATIONAL DATA EVALUATION I.E.3 OPERATIONAL SAFETY DATA CSI/LI 1d ANALYSIS I.E.4 COORDINATION OF LICENSEE, GSI/LI 1d . INDUSTRY, AND REGULATORY ' PROGRAMS I.E.5 NUCLEAR PLANT RELIABILITY GSI/LI 1d DATA. SYSTEM I.E.6 REPORTING REQUIREMENTS GSI/LI 1d FOR REACTOR OPERATING EXPERIENCE I.E.7 INFORMATION FOR ANALYSIS GSI/LI ld AND DISSEMINATION OF OPERATING EXPERIENCE-- FOREIGN SOURCES Amendment F l December 15, 1989
c CESSARinih m. ! I $)' TABLE Al-1'(Cont'd) (Sheet-41 of 55) , LISTING OF ! UNRESOLVED' SAFETY ISSUES AND GENERIC SAFETY ISSUES .; ISSUE ISSUE- ! NUMBER ISSUE TITLE TYPE CATEGORY i
'I.E.8 HUMAN. ERROR-RATE ANALYSIS GSI/LI id i e
I.F.1 OTTALITY-ASSURANCE - GSI 2 EXPAND QUALITY ASSURANCE LIST FOR EQUIPMENT IMPORTANT .9N) SAFETY ! I.F.2 (1,4,5,7,8,10,11) QUALITY GSI/TMI la ASSURANCE--DEVELOP MORE l
. DETAILED QA CRITERIA i
(\- /
) I.F.2 (2,3) QUALITY ASSURANCE -- GSI/TMI id DEVELOP MORE DETAILED QA CRITERIA I.F.2 -(6,9) QUALITY ASSURANCE -- GSI ld g i DEVELOP MORE DETAILED QA CRITERIA ,
I . G . '1 SCOPE OF TEST GSI lg PROGRAM--PREOPERATIONAL : AND LOW POWER TESTING I.G.2 SCOPE OF TEST PROGRAM GSI lg l II.A.1 SITING POLICY- GSI 1c l- REFORMULATION 1 ? II.A.2 SITE EVALUATIONS OF GSI le
? EXISTING FACILITIES L'
l !O l-Amendment G April 30, 1990 V
CESSAR UnineuiO - O TABLE A1-1 (Cont'd) (8heet 42 of 55) LISTING OF UNRESOLVED BAFETY ISSUES AND GENERIC 8AFETY ISSUES
- ISSUE ISSUE NUMBER ISBUE TITLE TYPE CATEGORY II.B.1 . SAFETY REVIEW GSI/TMI 2 CONSIDERATION -- REACTOR COOLANT SYSTEM' VENTS II.B.2 SAFETY REVIEW GSI/TMI 2 CONSIDERATION -- PLANT SHIELDING TO PROVIDE POST
. ACCIDENT ACCESS TO VITAL AREAS HII.B.3 SAFETY REVIEW GSI/TMI 2 -
CONSIDERATION -- POST
~ ACCIDENT SAMPLING SYSTEM h-s.
'II.B.4 SAFETY' REVIEW .GSI 1g.
CONSIDERATION--TRAINING TO MITIGATE CORE DAMAGE II.B.5 (1&?) BEHAVIOR OF GSI/LI id SEVERELY DAMAGED FUEL & CORE MELT II.B.5 (3) EFFECT OF H2 BURNING GSI/LI. ld AND EXPLOSIONS ON CONTAINMENT STRUCTURE II.B.6 RISK REDUCTION FOR GSI id OPERATING REACTORS WITH SITES WITH HIGH POPULATION DENSITIES II.B.7 SAFETY REVIEW GSI id CONSIDERATION -- ANALYSIS OF HYDROGEN CONTROL O Amendment F December 15, 1989
t 6 LCESSAR Mincuc. s)
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2.0 LIST OP' UNRESOLVED SAFETY ISSUES AND HIGH/ MEDIUM PRIORITY GENERIC SAFETY ISSUES APPLICABLE TO THE SYSTEM 80+ STANC?.RD DESIGN l Table A2-1 ~ of this - section identifies the USIs, land Medium- and l High-priority.GSIs which are technically relevant to the System ;l 80+ Standard Design, consistent with 10 CFR Part 52.47. The- ' process ; for identification ~ of these issues is provided in the overview and in ~ Section 1.0 of this appendix,' along with a ' definition of'the " Issue Types" indicatec. within.the list given I in this section. l
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'%-l' TABLE A2-1 (Cont'd) !
(Sheet 3 of 10). LIST OP-UNRESOLVED SAFETY ISSUES AND HIGH/ MEDIUM PRIORITY GENERIC ISSUES APPLICABLE TO THE ; SYSTEM 80+ STANDARD DESIGN ISSUE ISSUE-NUMBER ISSUE TITLE TYPE 119.1 PIPE RUPTURE REQUIREMENTS GSI/RI , 119.2 PIPE DAMPING VALUES GSI/RI 119.3 DECOUPLING OBE FROM SSE GSI/RI , 119.5 LEAK DETECTION REQUIREMENTS GSI/RI 121 HYDROGEN CONTROL FOR LARGE, GSI l DRY PWR CONTAINMENT-ft ~~ INITIATING FEED AND BLEED GSI A 122.2- ; i 124 AUXILIARY-FEEDWATER SYSTEM GSI RELIABILITY 125.I.3 SPDS AVAILABILITY GSI -t 125.II.7. REEVALUATE PROVISION TO GSI AUTOMATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR DURING LINE BREAK ; 128- ELECTRICAL POWER RELIABILITY GSI , 130 ESSENTIAL SERVICE WATER PUMP GSI FAILURES AT MULTIPLANT SITES ; 135 INTEGRATED STEAM GENERATOR GSI ISSUE A-1 WATER HAMMER USI i A-2 ASYMMETRIC BLOWDOWN LOADS ON USI RCS C-E STEAM GENERATOR TUBE USI ( N/
) A-4 INTEGRITY l_
l' l Amendment F j December 15, 1989
CESSARBH L mN
. O TABLE A2-1 (Cont'd)
(Sheet 4 of 10) LIST OF UNRESOLVED SAFETY ISSUES AND H1GH/ MEDIUM PRIORITY GENERIC 1SSUES APPLI'.'.ABLE TO THE SYSTEM 80+ STANDARD DESIGN.
-ISSUE ISSUE-NUMBER ISSUE TITLE TYPE =
A-9 ANTICIPATED TRANSIENTS WITHOUT -USI SCRAM (ATWS)_ A-12 FRACTURE TOUGHNESS OF STEAM USI GENERATOR AND RCP SUPPORTS-A-13 SNUBBER OPERABILITY ASSURANCE GSI G
-A SYSTEMS INTERACTION USI A-24 ' QUALIFICATION OF CLASS lE USI O-SAFETY RELATED EQUIPMENT A-25 NON-SAFETY LOADS ON CLASS lE GSI
-POWER. SOURCES A -REACTOR VESSEL. PRESSURE USI TRANSIENT PROTECTION A-29 PLANT DESIGN FOR REDUCTION OF GSI VULNERABILITY TO SABOTAGE A-30 ADEQUACY OF SAFETY RELATED DC GSI POWER SUPPLIES
.A-31 RHR SHUTDOWN REQUIREMENTS USI A-35 ADEQUACY OF OFFSITE POWER GSI SYSTEMS A-36 CONTROL OF HEAVY LOADS NEAR USI SPENT FUEL A-40 SEISMIC DESIGN--SHORT TERM USI PROGRAM Amendment G '
April 30, 1990
CESSARtRWeuN t O TABLE A2-1 (Cont'd) (Sheet 5 of 10) LIST OF UNRESOLVED SAFETY ISSUES AND HIGH/ MEDIUM PRIORITY GENERIC ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN ISSUE ISSUE i NUMBER ISSUE TITLE TYPE sE A-43 CONTAINMENT EMERGENCY SUMP USI PERFORMANCE A-44 STATION BLACKOUT USI A-45 SHUTDOWN DECAY HEAT REMOVAL USI REQUIREMENTS A-47 SAFETY IMPLICATIONS OF CONTROL USI ; SYSTEMS j O A-48 HYDROGEN CONTROL, MEASURES & EFFECT OF HYDROGEN BURNS USI ! A-49 PRESSURIZED THERMAL SHOCK USI B-5 DUCTILITY OF TWO-WAY SLABS & GSI , SHELLS -- STEEL CONTAINMENTS i G l B-53 LOAD BREAK SWITCH GSI/RI B-56 DIESEL GENERATOR RELIABILITY GSI j B-58 PASSIVE MECHANICAL FAILURES GSI B-60 LOOSE PARTS MONITORING SYSTEM GSI Ts-61 ALLOWABLE ECCS EQUIPMENT GSI OUTAGE PERIODS J Amendment G April 30, 1990
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TABLE A2-1 (Cont'd) (Sheet 6 of 10) j L ST OP UNRESOLVED SAFETY ISSUES AND HIGH/ MEDIUM PRIORITY GENERIC ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN ! ISSUE ISSUE NUMBER ISSUE TITLE TYPE B-63 ISOLATION OF LOW PRESSURE GSI l SYSTEMS CONNECTED TO THE ! REACTOR COOLANT PRESSURE BOUNDARY B-66 CONTROL ROOM INFILTRATION GSI MEASUREMENTS G r C-2 STUDY OF CONTAINMENT GSI l DEPRESSURIZATION BY INADVERTENT SPRAY OPERATION C-4 brATISTICAL METHOD FOR ECCS GSI/RI ANALYSIS C-5 DECAY HEAT UPDATE GSI/RI G C-10 EFFECTIVE OPERATION OF GSI CONTAINMENT SPRAYS IN A LOCA C-12 PRIMARY SYSTEM VIBRATION GSI ASSESSMENT [ HF 1.3.4a HUMAN FACTORS PROGRAM PLAN - GSI MAN MACHINE INTERFACE - LOCAL CONTROL STATIONS HF 1.3.4b HUMAN FACTORS PROGRAM PLAN - GSI MAN MACHINE INTERFACE - ANNUNCIATORS { ' l Amendment G u April 30, 1990 1:
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TABLE A2-1 (Cont'd) (Sheet 7 of 10) LTST OF UNRESOLVED SAFETY ISSUES AND HIGH/ MEDIUM ! PRIORITY GENERIC ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN ISSUE ISSUE NUMBER ISSUE TITLE TYPE
- HF 1.3.4c HUMAN FACTOR PROGRAM PLAN - GSI MAN MACHINE INTERFACE -
OPERATIONAL AIDS HF 1.3.4d HUMAN FACTORS PROGRAM PLAN - GSI MAN MACHINE INTERFACE - AUTOMATION AND ARTIFICIAL INTELLIGENCE , r HF 1.3.4e HUMAN FACTORS PROGRAM PLAN - GSI , e~s MAN MACHINE INTERFACE - (') COMPUTERS AND COMPUTER DISPLAYS HF 5.1 LOCAL CONTROL STATIONS GSI HF 5.2 REVIEW CRITERIA FOR HUMAN GSI FACTORS ASPECTS OF ADVANCED CONTROLS AND INSTRUMENTATION HF 8.0 MAINTENANCE AND SURVEILLANCE GSI PROGRAM I.C.1 (1-4) SHORT TERM ACCIDENT GSI ANALYSIS AND PROCEDURES REVISION I . D. 2 CONTROL ROOM DESIGN REVIEWS -- GSI/TMI PLANT SAFETY PARAMETER DISPLAY CONSOLE O u I.D.4 CONTROL ROOM DESIGN STANDARD GSI h. O Amendment G April 30, 1990
CESSAR !!mr.co.N O TABIE A2-1 (Cont'd) (Sheet 8 of 10) LIST OF UNRESOLVED SAFETY ISSUES AND HIGH/ MEDIUM PRIORITY GENERIC _ ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN ISSUE ISSUE NUMBER ISSUE TITLE TYPE I.D.5 (1) CONTROL ROOM DESIGN -- GSI IMPROVED INSTRUMENTATION RESEARCH ALARMS AND DISPLAY I.D.5 (2) CONTROL ROOM DESIGN -- GSI IMPROVED INSTRUMENTATION RESEARCH I.D.5 (3) CONTROL ROOM DESIGN -- GSI/LI ON-LINE REACTOR SURVEILLANCE SYSTEMS I.D.5 (4) CONTRfi ROOM DESIGN -- GSI IMPROVED INSTRUMENTATION RESEARCH I.F.1 QUALITY ASSURANCE - EXPAND GSI QUALITY ASSURANCE LIST FOR EQUIPMENT IMPORTANT TO SAFETY G O Amendment G April 30, 1990
C E S S A R RR n?,cui.. O TABLE A3-1 (Sheet 1 of 2) NRC USIs/GSIs CROSS-REFERENCED IN NUREG-0933 AND APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN NRC ISSUE CROSS-REFERENCED NUMBER ISSUE TITLE ISSUES 36 LOSS OF SERVICE WATER A-45 48 LCO FOR CLASS 1E VITAL 128 INSTRUMENT BUSES IN OPERATING REACTORS 49 INTERLOCKS AND LCO'S FOR 128 REDUNDANT CLASS 1E TIE BREAKER 66 STEAM GENERATOR REQUIREMENTS A-4 79 UNANALYZED REACTOR VESSEL A-44 THERMAL STRESS-COOLDOWN G 122.2 INITIATING FEED AND BLEED A-45 125.I.3 SPDS AVAILABILITY I.D.2 128 ELECTRICAL POWER RELIABILITY 48, 49 130 ESSENTIAL SERVICE WATER PUMP A-45 FAILURES AT MULTI-PLANT SITES A-30 ADEQUACY OF SAFETY RELATED DC II.E.1.1 POWER SUPPLIES A-45 SHUTDOWN DECAY HEAT REMOVAL A-44 REQUIREMENTS I.D.5 (2) CONTROL ROOM DESIGN -- II.F.3 IMPROVED INSTRUMENTATION RESEARCH I.D.5 (4) CONTROL ROOM DESIGN -- II.F.2 IMPROVED INSTRUMENTATION O RESEARCH Amendment G April 30, 1990
C E S S A R 88 #.cui. O TABLE A3-1 (Cont'd) (Sheet 2 of 2) MRC USIs/08Is CROSS-REFERENCED IN NUREG-0933 AND. APPLICABLE..TO.TER.8YSTEM 80+ STANDARD DESIGN NRC ISSUE CROBS-REFERENCED MHEREE._ ISSUE TITLE ISSUES II.C.4 RELIABILITY ENGINEERING B-56 Oli l i O Amendment F December 15, 1989
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CESSARHi#co. O 4.0 TECHNICAL RESOLUTIONS FOR UNRESOLVED AND GENERIC SAFETY ISSUES APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN This section presents the technical resolutions for 61 safety issues applicable to the System 80+ Standard Design. The G resolutions for these safety issues are listed in Table A4-1. Resolutions for the remaining applicable issues are scheduled for subsequent submittal. Each issue is structured so as to bo independent of other safety , issues. However, there are some instances where issues do ' overlap one another. Where overlap occurs, as identified in-NUREG-0933, it .is so indicated (see also Section 3.0,
" Cross-references").
As discussed in the Appendix overview, each issue is composed of four parts: (1) a ISSUE statement section, which describes the safety _ concern, (2) a ACCEPTANCE CRITERIA section which discusses ! the applicable NRC guidance and regulations and industry codes, I standards and/or other relevant requirements, (3) a RESOLUTION- j section which describes the technical cases for the resolution of i the issue considering the System 80+ Standard Design, as O described within CESSAR-DC or other relevant documentation (e.g., special technical reports) and finally, (4) a REFERENCES section which lists the references used in the formulation of the i Issue Statement, Acceptance Criteria, and Resolution sections of f the issue. . l O Amendment G A-7 April 30, 1990
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c.:) TABLE A4-1 (Sheet 1 of 6) LIST OF TECHNICAL RESOLUTIONS FOR USIs AND GS"s APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN INCLUDED IN SECTION 4.0 , NRC ISSUE NUMBER ISSUE TITLE PAGE NO. 3 SETPOINT DRIFT IN INSTRUMENTATION A-9 15 RADIATION EFFECTS ON REACTOR A-11b G VESSEL SUPPORTS 22 INADVERTENT BORON DILUTION EVENTS A-12 23 REACTOR COOLANT PUMP SEAL FAILURES A-14 , g-x 29 BOLTING DEGRADATION OR FAILURES IN A-17 () NUCLEAR PLANTS 36 LOSS OF SERVICE WATER A-19 45 INOPERABILITY OF INSTRUMENTS A-20 DUE TO EXTREME COLD WEATHER 48 LCO FOR CLASS 1E VITAL INSTRUMENT A-24 BUSES IN OPERATING REACTORS 49 INTERLOCKS AND LCOs FOR CLASS 1E A-26 , TIE BREAKERS 51 PROPOSED REQUIREMENTS FOR IMPROVING A-28 RELIABILITY OF OPEN CYCLE SERVICE WATER SYSTEMS r 57 EFFECTS OF FIRE PROTECTION SYSTEM A-31 ACTUATION ON SAFETY-RELATED EQUIPMENT 64 IDENTIFICATION OF PROTECTION A-33 SYSTEM INSTRUMENT SENSING LINES 66 STEAM GENERATOR REQUIREMENTS A-35
/ 79- UNANALYZED REACTOR VESSEL THERMAL A-37
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\/ STRESS-COOLDOWN l
Amendment G April 30, 1990 l
CESSAO !!!#,cuio. O I TABLE A4-1 (Cont'd) (Sheet 2 of 6) LIST OF TECHNICAL RESOLUTIONS FOR USIs AND GSIs A?PLICABLE TO THE SYSTEM 80+ STANDARD DESIGN INCLUDED IN SECTION 4.0 J l NRC . ISSUE NUMBER ISSUE TITLE PAGE NO. l l 82 BEYOND DESIGN BASES ACCIDENTS IN A-39 SPENT FUEL POOLS 83 CONTROL ROOM HA)'.TABILITY A-41 93 STEAM BINDING OF AUXILIARY A-43 FEEDWATER PUMPS 94 ADDITIONAL LOW TEMPERATURE A-45b OVERPRESSURE PROTECTION ISSUES G FOR LIGHT WATER REACTORS 103 DESIGN FOR PROBABLE MAXIMUM A-46 PRECIPITATION r 105 INTERFACING SYSTEMS LOCA AT LWRs A-47b G 106 PIPING AND USE OF HIGHLY COMBUSTIBLE A-48 GASFS IN VITAL AREAS -- FIRE PROTECTION 119.1 PIPE RUPTURE REQUIREMENTS A-f0 119.2 PIPE DAMPING VALUES A-52 119.3 DECOUPLII:G OBE FROM SSE A-55 122.2 INITIATING FEED AND BLEED A-58 124 AUXILIARY FEEDWATER SYSTEM A-60 RELIABILITY 125.I.3 SPDS AVAILABILITY A-62 125.II.7 REEVALUATE PROVISIONS TO A-63b 6 AUTOMATICALLY ISOLATE FEEDWATER FROM' STEAM GENERATOR DURING A LINE BREAK Amendment G
GESSAR Esiiricar..N ; _d'~'% s_- ) ( TABLE A4-1 (Cont'd) (Sheet 3 of 6) LIST OF TECHNICAL RESOLUTIONS FOR USIs AND GSIs APPLICABLE TO THE SYSTEM 80+ 1 STANDARD DESIGN INCUUDED IN SECTION 4.0 NRC ISSUE NUMBER ISSUE TITLE PAGE NO. 128 ELECTRICAL POWER RELIABILITY A-64 130 ESSENTIAL SERVICE WATER PUMP A-65 FAILURES AT MULTI-PLANT SITES A-1 WATER HAMMER A-66a A-2 ASYMMETRIC BLOWDOWN LOADS ON RCS A-67 A-9 ANTICIPATED TRANSIENTS WITHOUT A-69
~ ('] SCRAM (ATWS) v A-12 FRACTURE TOUGHNESS OF STEAM A-72 GENERATOR AND RCP SUPPORTS A-13 SNUBBER OPERABILITY ASSURANCE A-74 A-25 NON-SAFETY LOADS ON CLASS 1E A-77 POWER SOURCES A-26 REACTOR VESSEL PRESSURE TRANSIENT A PROTECTION
! A-29 PLANT LBSIGN FOR REDUCTION OF A-84 L VULNERABILITY TO SABOTAGE A-30 ADEQUACY OF SAFETY-RELATED DC A-88 POWER SUPPLIES A-31 RHR SHUTDOWN REQUIREMENTS A-90 A-36 CONTROL OF HEAVY LOADS NEAR A-94 l SPENT FUEL l A-43 CONTAINMENT EMERGENCY SUMP A-97 p PERFORMANCE (/ A-44 STATION BLACKOUT A-100a G Amendment G April 30, 1990
CESSA0 !!nh.N l 9: TABLE A4-1 (Cont'd) (Sheet 4 of 6) LIST OF TECHNICAL RESOLUTIONS FOR USIs AND GSIs APPLICABLE TO THE SYSTEM 80+ i STANDARD DESIGN INCLUDED IN SECTION 4.0 1 NRC l ISSUE j NUMBER I.] SUE TITLE PAGE NO. A-45 SHUTDOWN DECAY HEAT REMOVAL A-101 l REQUIREMENTS G A-47 SAFETY IMPLICATIONS OF CONTROL A-102a SYSTEMS , A-49 PRESSURIZED THERMAL SHOCK A-103 B-53 LOAD BREAK SWITCH A-105a g B-56 , DIESEL RELIABILITY A-105d B-60 LOOSE PARTS MONITORIF.G SYSTEM A-106 0-B-63 ISOLATION OF LOW PF. ESSURE A-10Ba SYSTEMS CONNECTED f0 THE 6 REACTOR COOLANT PLESSURE BOUNDARY C-4 STATISTICAL METHODS FOR ECCS A-109 ANALYSIS C-5 DECAY HEAT UPDATE A-111 C-12 PRIMARY SYSTEM VIBRATION ASSESSMENT A-113 HF 1.3.4a HUMAN FACTORS PROGRAM PLAN - A-115 LOCAL CONTRCL STATIONS HF l'.3.4b HUMAN FACTORS PROGRAM PLAN - A-116 ANNUNCIATOR SYSTEMS HF 1.3.4c HUMAN FACTORS PROGRAM PLAN - A-117 OPERATIONAL AIDS HF 1.3.4d HUMAN FACTORS PROGRAM PLAN - A-117 AUTOMATION AND/OR ARTIFICIAL INTEuLIGENCE SYSTEMS Amendment G April 30, 1990
r CESSARHubio. ; i f~N 1 N/' l TABLE A4-1 (Cont'd) (Sheet 5 of 6) LIST OF TECHNICAL RESOLUTIONS FOR .l Mils AND GSIs APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN INCLUDED IN SECTION 4.0 1 NRC ISSUE NUMBER ISSUE TITLE PAGE NO. HF 1.3.4e HUMAN FACTORS PROGRAM PLAN - A-117 COMPUTERS AND COMPUTER DISPLAY TECHNOLOGY HF 5.1 LOCAL CONTROL STATIONS A-120 P HF 5.2 ' REVIEW Ol' CRITERIA FOR HUMAN A-121 FACTORS ASPECTS OF ADVANCED . INSTRUMENTATION AND CONTROLS (ANNUNCIATORS) (
\- I.C.1 SHORT TERM ACCIDENT ANALYSIS A-122 AND PROCEDURES REVISION I.D.2 CONTROL ROOM DESIGN -- PLANT A-123a SAFETY PARAMETER DISPLAY CONSOLE G I.D.4 CONTROL ROOM DESIGN STANDARD A-123d I.D.5 (1) CONTROL ROOM DESIGN -- A-123h OPERATOR - PROCESS COMMUNICATION I.D.5 (2) CONTROL ROOM DESIGN -- A-124 IMPROVED INSTRUMENTATION RESEARCH - PLANT STATUS AND POST-ACCIDENT MONITORING I.D.5 (3) CONTROL ROOM DESIGN -- A-127 ON-LINE REACTOR SURVEILLANCE
- SYSTEMS I.D.5 (4) CONTROL ROOM DESIGN -- A-130 l PROCESS MONITORING
, INSTRUMENTATION L E
.Q k/ SAFETY REVIEW CONSIDERATION -- A-133 m II.B.1 REACTOR COOLANT SYSTEM VENTS Amendment G L
April 30, 1990 L _ _
(CESSAR !!!hno. , 9 TABLE A4-1 (Cont'd) (Sheet 6 of 6)
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LIST OF TECHNICAL RESOLUTIONS FOR USIs AND GSIs APPLICABLE TO THE SYSTEM 80+ STANDARD. DESIGN INCLUDED IN SECTION 4.0 NRC ISSUE NUMBER ISSUE TITLE PAGE NO. II.B.3 SAFETY REVIEW CONSIDERATION -- A-135 POST ACCIDENT SAMPLING SYSTEM II.C.4 RELIABILITY ENGINEERING A-138 II.D.1 COOLANT SYSTEM VALVES -- A-141 PERFORMANCE TESTING REQUIREMENTS II.D.3 COOLANT SYSTEM VALVES -- VALVE A-143 POSITION INDICATION II.E.1.1 AUXILIARY FEEDWATER SYSTEM A-146 EVALUATION II.E.1.2 AUXILIARY FEEDWATER SYSTEM A-149 AUTOMATIC INITIATION AND FLOW INDICATION II.E.4.2 CONTAINMENT DESIGN -- ISOLATION A-150a DEPENDABILITY g II.E.4.4 CONTAINMENT DESIGN -- PURGING A-150e II.F.2 INSTRUMENTATION FOR DETECTION A-151 OF INADEQUATE CORE COOLING II.F.3 INSTRUMENTATION FOR MONITORING A-154 ACCIDENT CONDITIONS II.G.1 POWER SUPPLIES FOR PRESSURIZER A-157 RELIEF VALVES, BLOCK VALVES, AND LEVEL INDICATORS II.K.1 MEASURES TO MITIGATE SMALL BREAK A-159 IN LOSS-OF-COOLANT ACCIDENTS AND 6 LOSS-OF-FEEDWATER ACCIDENTS--IE BULLETINS II.K.3 FINAL RECOMMENDATIONS OF BULLETINS A-161 AND ORDERS TASK FORCE Amendment G April 30, 1990
CESSAR !!!nnenic ! n - O REFERENCES ,
- 1. NUREG-0933, "A Status Report en Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
- 2. Regulatory Guide 1.105, Revision 2, " Instrument Setpoints For Safety-Related Systems", U.S. Nuclear Regulatory Commission, February 1986.
- 3. Standard ISA-S67.04-1987, "Setpoints for Nuclear Safety-Related Instrumentation Used In Nu: lear Power ,
Plants", Instrument Society of America. L l l O Amendment F A-11 December 15, 1989
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1 L I CESSAR !!nificama r. O THIS PAGE INTENTIONALLY BLANK O O Amendment G A-11a April 30, 1990
CESSAR Minem, s i ^F GJ 015: RADIATION EFFECTS ON REACTOR VLSSEL SUPPORTS ISSUE , Unresolved Safety Issue (USI) 015 in NUREG-0933 (Reference 1), addresses the potential for failure of the reactor vessel support structure (RVSS) due to a combination of low temperature and low neutron flux irradiation embrittlement. Neutron irradiation of structural materials used in the RVSS causes embrittlement that may increase the potential for propagation of pre-existing cracks or flaws within these materials. The potential for brittle fracture of these materials is typically measured in terms of their nil ductility transition temperature (NDTT) . As long as the operating environment in which a material is used has a temperature that is significantly higher than the NDTT of the material, no failure by brittle fracture would be expected. Many materials, when subjected to neutron irradiation, experience an upward shift in the NDTT, i.e., they become more susceptible to brittle fracture. This effect must be accounted for in the design and fabrication of n RVSS. b] During 1988, new data was developed for the RVSS materials at Oak G Ridge National Laboratory (ORNL) (References 2 and 3). This data indicated that low neutron flux at low temperatures caused greater embrittlement of the materials used in RVSS than previously anticipated. This increased material embrittlement or
" upward shift" in NDTT reduces the fracture toughness of these materials and, under certain specific and conservative transient conditions such as an earthquake or Large-Break Loss-Of-Coolant Accident (LBLOCA), could conceivably result in the failure of the supports thus permitting the reactor vessel to move.
As a result of the ORNL work, the NRC re-prioritized this 1. c' e and is reviewing its regulatory position relative to low temperature and low neutron flux radiation embrittlement. ACCEPTANCE CRITERIA The acceptance criterion for the resolution of USI 015 is that the material integrity for the reactor vessel support structure shall be maintained for the design lifetime of the plant. Specifically, the design of the reactor vessel supports shall address irradiation effects (including low temperature and low neutron flux) and the attendant material embrittlement, and be
/^) designed to restrain the reactor vessel under the combined Safe V Shutdown Earthquake (SSE) and branch line pipe break loadings in accordance with the stress and deflection limits established in Section III of the ASME B&PV Code (Reference 4). 3 Amendment G A-11b April 30, 1990
L i SESSAD n#cu... O RESOLUTION The RVSS for the System 80+ Standard Design is described in CESSAR-DC, Section 5.4.14.2 and Figure 5.4.14-2. It consists of four vertical columns (this configuration is defined by ORNL as the "long column type") which are located under the vessel inlet nozzles. These columns are designed to flex in the direction of horizontal thermal expansion and thus allow unrestrained heatup and cooldown. In addition, they also act as holdown devices for the reactor vessel. The supports are designed to accept normal, seismic, and branch line pipe break loads. Irradiation effects are addressed in the fracture mechanics analysis of the columns which is performed using the philosophy of ASME Section III Appendix G to ensure that structural-integrity is maintained. This fracture mechanics j analysis addresses potential embrittlement and accident loads including SSE and LBLOCA. This analysis demonstrates that the structural integrity of the columns would be maintained, even if' G j large cracks existed in the columns and they were subjected to the lowest possible temperatures and the maximum normal and SSE j loadings. The sensitivity to uncertainty in the extent of the i embrittlement is also addressed. The conservatism of this l analysis is further enhanced by the adoption of the l Leak-Before-Break (LBB) method in the System 80+ Design Basis. ~l The concrete pedestal and anchorage embedments for the Reactor Vessel Support columns are not subjected to significant lieutron ! flux. These items are not, therefore, addressed in the above j analysis. l t In conclusion, the irradiation effects have been addressed in a fracture mechanics analysis of the System 80+ reactor vessel supports. This analysis indicates thr.t the structural integrity of the columns will be maintained when the conservative h combination of irradiation embrittlement, low temperature, and l the design basis loads is imposed on the columns. Therefore, this issue is resolved for the System 80v Standard Design. BEFERENCEB
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues, i U.S. Nuclear Regulatory Commission, January 1989.
- 2. ORNL/TM-10444, " Evaluation of HFIR Pressure Vessel Integrity Considering Radiation Embrittlement", Oak Ridge National Laboratory, 1988.
- 3. ORNL/TM-10966, " Impact of Radiation Embrittlement on the ;
Integrity of Pressure Vessel Supports for Two PWR Plants", j f Oak Ridge National Laboratory, 1988. Amendment G A-11c April 30, 1990 l
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CESSAR RHrTco... I i O ;
- 4. American Society of Mechanical Engineers, Boiler &. Pressure Vessel Code, Section III (Nuclear), American Society of G Mechanical Engineers.
t e O i h s l i I' O 2 Amendment G A-11d April 30, 1990
CESSAR c*t!#cm.. l O 022: INADVERTENT BORON DILUTION EVENTS 198N% Generic Safety Issue (GSI) 022 in NUREG-0933 (Reference 1), addresses the possibility of core criticality during cold shutdown conditions because of an inadvertent boron dilution event. Inadvertent boron dilution events have occurred at PWR's during ' maintenance and refueling periods. If the boron in the RCS is sufficiently diluted and the reactor core is near the beginning of life, there is the potential for core criticality with all rods inserted (i.e., during cold shutdown conditions). The NRC and others performed a variety of studies of the consequences of an inadvertent boron dilution event. The conclusions of the NRC assessment along with other studies were (1) that the consequences of an unmitigated boron dilution event, although undesirable, are not severe enough to warrant backfit of additional protective features at operating plants and (2) Standard Review plan (SRP) Section 15.4.6 (Reference 2) is adequate for plants presently undergoing license review. ACCEPTANCE CRITERIA The acceptance criterion for GSI 022 is that new plants shall minimize the consequences of inadvertent boron dilution events by meeting the intent of SRP Section 15.4.6. Specifically, when performing a safety analysis to evaluate the consequences of an inadvertent boron dilution, plant designers should consider: (1) design limits for maximum RCS pressure and minimum DNBR, (2) moderate frequency events in conjunction with a single failure or operator error and their possible effects on fuel integrity and radiological dose calculations, (3) and time limits specified for each mode of plant operation, if operator action is required to terminate an inadvertent boron dilution. I B180LUTION As part of the design process for the System 80+ Standard Design, Safety Analyses are performed. These analyses address a variety of design bases events including inadvertent boron dilution (see CESSAR-DC, Section 15.4. 6) . Furthermore, these analyses consider , SRP Section 15.4.6 criteria including, design limits (e.g., l maximum RCS pressure and minimum DNBR), a single failure in conjunction with moderate frequency events, and the impact of a single failure or operato'/ error on fuel integrity and , l radiological dose calculatiors. ' i i Amendment F o L A-12 December 15, 1989 l
1 O E S S A R !!n % uion 7 N, In summary, the System 80+ Standard Design addresses steam binding of the EFW pumps in four ways. First, each train is equipped with two normally-closed isolation valves namely a MOV and a check valve. Thus redundant isolation of the EFW system ' from the main feedwater system and associated steam generator is achieved. Second, each subtrain is separated from the other such that a leak of a single set of valves does not affect all of the , pumps. Third, TI's . in each EFW pump discharge line alert the ! plant operator should valve leakage be present. Finally, open , lines permit valve leakage to be vented through the EFWST vents. Since the EFW System in the System 80+ Standard Design meets and exceeds the intent of GL 88-03, this issue is resolved for the System 80+ Standard Design. REFERENCES
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
- 2. Generic Letter 88-03, " Steam Binding of Auxiliary Feedwater Pumps", U.S. Nuclear Regulatory Commission.
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l Amendment F A-45 December 15, 1989 L
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L CESSAR !!Nacam. i h 094: ADDITIONAL LOW TEMPERATURE OVERPRESSURE PROTECTION ISSUES FOR LIGHT WATER REACTORS ISSUE Generic Safety Issue (GSI) 094 in NUREG-0933 (Reference 1), addresses the establishment of additional guidance for Reactor , Coolant System (RCS) low temperature overpressure (UTOP) , protection to assure reactor vessel and RCS integrity, beyond the ' guidance for the provision of LTOP protection identified in the resolution of USI A-26. (USI A-26 is addressed in this appendix and is considered resolved for the System 80+ Standard Design.) Low temperature overpressurization was originally identiff.ed as a
~s afety issue in the early 1970's because of numerous incidents of plants - exceeding pressure-temperature limits. The majority of these events occurred while in a water-solid condition, during !
startup or shutdown operations, and at relatively low reactor vessel temperatures. Additional RCS overpressurization incidents have occurred since the implementation of-USI A-26 guidelines by operating plants. Two events in particular, were severe enough to be identified as
/ s abnormal occurrences. After a further study, the NRC concluded d in NUREG-1326 (Reference 2) that LTOP protection system unavailability is the dominant contributor to risk from low-temperature overpressure transients, and that a_ substantial improvement in reliability can be achieved through improved O
administrative restrictions on the LTOP protection system. Consideration was also given to requiring the system to be safety-grade, but' this was not found to be cost-effective for existing plants. The recommended NRC resolution for GSI 094 is, therefore, to request a Technical Specification to ensure that both LTOP protection system channels be operable to provide additional protection against brittle vessel failure. ACCEPTANCE CRITERIA L The acceptance criterion for the resolution of GSI 094 is to increase the protection of the reactor vessel from brittle fracture resulting from low temperature overpressure operation of the RCS (e.g., during such plant evolutions as heatup, cooldown, water solid, and maintenance operations). Specifically, Limiting Conditions for Operation (LCOs) shall be identified to ensure the operability of both LTOP protection O system channels during plant evolutions requiring protection V against reactor vessel brittle failure. Amendment G A-45b April 30, 1990
C E S S A 9 in Wican. O Furthermore, protection of the reactor vessel from brittle fracture.would also consist of: (1) establishing a conservative design basis for the relief valves for LTOP protection based upon the worst case mass and energy addition, (2) assuring that the flow paths to the LTOP relief valves are open when protection is required, and (3) the'use of improved materials and fabrication techniques for the reactor vessel. RESOLUTION The System 80+ Standard Design addresses the integrity of the RCS under low temperature and pressure conditions by focusing on the issues that impact RCS and reactor. vessel integrity (i.e., relief valves, . limiting conditions for operation, reactor vessel materials, and vessel manufacturing techniques). The System 80+ Standard Design includes a safety-grade Shutdown Cooling System (SCS). In addition to providing a method of removing core decay heat, this system provides for overpressure protection of the RCS at reduced temperatures (see CESSAR-DC, Section 5.2.2.10) by providing a relief path during heatup and cooldown through relief valves included in the SCS suction lines connected to the RCS, A design basis for these relief valves is the anticipated worst case mass and energy addition to the RCS and the valves are sized and adjusted to the appropriate setpoint(s). This practice ensures adequate overpressure protection for the RCS at reduced temperatures. Also, a Technical Specification establishes an LCO for the SCS, which ensures SCS operability, and LTOP protection system availability in accordance with the recommendation of NUREG-1326. G In addition, to ensure that the flow paths to the LTOP relief valves are open to thu RCS when required, the SCS contains alarms which warn the operator in the control ro of a requirement for LTOP protection no as to prevent inadvertent closure of the SCS isolation valves. The reactor vessel itself is designed to be less sensitive to LTOP events. Since vessel material embrittlement is of primary concern 'in assuring vessel integrity, the System 80+ vessel design incorporates improved materials (see CESSAR-DC, Section 5.2.3) to limit the need for LTOP protection. Furthermore, the reactor vessel is designed to assure that the welds employed in the fabrication process are controlled to avoid cracking and embrittlement from irradiation and reactor coolant water chemistry (see CESSAR-DC, Section 5.2.3.3), thus enhancing its resistance to embrittlement and improving its tolerance to low temperature and pressure overpressurization events. Also, a O Amendment G A-45c April 30, 1990
CESSAR !!!Mncm 7 1 ) u/ ring-forged manufacturing technique is specified to eliminate vertical welds in the reactor core region (see CESSAR-DC, Section 5.3). In summary, prevention of vessel failure during low-temperature overpressure modes of operation is a combination of design, manufacture and operating features. The System 80+ Standard Design employs features like SCS relief valves, limiting conditions for operation, and advanced reactor vessel materials and fabrication techniques to assure continuous vessel integrity over the 60-year plant design lifetime. Since reliable LTOP g protection for the RCS and reactor vessel is assured for the reasons described above, this issue is resolved for the System 80+ Standard Design. REFERENCES I
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
- 2. NUREG-1326, " Regulatory Analysis for the Resolution of '
Generic Issue 94 - Additional Low-Temperature Overpressure n Protection for Light Water Reactors", U.S. Nuclear (j Regulatory Commission, December 1989. l l l l l l l l l O A Amendment G A-45d April 30, 1990 l 1
l CESSAR !!nincuio ; l O' ! 103: DESIGN FOR PROBABLE MAXINUM PRECIPITATION ISSUE Generic Safety Issue (GSI) 103 in NUREG-0933 (Reference 1), addresses the accepted methodology used for determining the - design flood. level for a particular reactor plant site. Accurate i determination of the design flood level for - a specific reactor site is necessary in order to ensure adequate protection of safety-related equipment against possible site flooding. Reactor plant sites are designed to accommodate maximum flood level because flooding could disabic safety-related equipment. Historically,' estimating design flood levels for specific reactor plant sites has been based upon input data for probable maximum flood (PMF) provided by the U.S. Army Corp. of Engineers for the specific site. The -guidance identified in the Standard Review Plan (SRP) Sections 2.4.2,_Rev. 3, and 2.4.3, Rev. 3 (Reference
- 2) is used in predicting design flood levels. Furthermore, general requirements are defined in General Design Criteria (GDC) 2 (Reference 3). The SRP's state that " design basis flood levels" incorporate the most severe historical environmental data with -
" sufficient margin". What is considered to be " sufficient margin" and procedures for estimating PMF's are identified in Regulatory Guides 1.59 and 1.102, and ANSI /ANS 2.8 (References 4, 5, and 6).
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI 103 is that the site chosen for a commercial nuclear generating faci'lity shall be designed to acccvmodate a maximum expected flood . from precipitation without jeopardizing the safe operation of the facility, in accordance with the guidance given in SRP 2.4.2, Rev. 3 and- SRP 2.4.3, Rev. 3. Also, the facility design, including structures, systems, and components important to safety, shall meet the criteria specified in 10 CFR 50 Appendix A (GDC 2). RESOLUTION The System 80+ Standard Design is designed to meet the requirements of GDC 2 as described in CESSAR-DC, Section 3.1.2. l The System 80+ Standard Design is based upon a set of assumed site-related parameters. These parameters were selected to envelope most potential nuclear power plant sites in the United L States. A summary of the assumed site design parameters, including maximum flood level, is given in CESSAR-DC, Section t 2.0, Table 2.0-1. Amendment F A-46 December 15, 1989
CESSARHEL.n . l^* Q. Detailed-site characteristics based upon historical site specific environmental' data will be provided by the site owner-operator for any specific application, the site owner-operator will review and evaluate these characteristics to ensure compliance with-the enveloping assumptions of Table 2.0-1. Since the System 80+ Standard Design is designed in accordance with GDC 2 for the most severe expected environmental conditions, including flooding, tornado, hurricane etc., and meets the intent of SRP Section 2.4.2, Rev. 3, and SRP Section 2.4.3, Rev. 3 with respect to plant design, therefore this issue is resolved for the System 80+ Standard Design. , EEFERENCES
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
- 2. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants-- LWR Edition",
U.S. Nuclear Regulatory Commission.
- 3. 10 CFR 50 Appendix A, " General Design Criteria for Nuclear
[-3 ' Fower Plants", Office of The Federal Register, National Archives and Records Administration.
- 4. Regulatory Guide 1.59, " Design Basis Floods for Nuclear Power Plants", U.S. Nuclear Regulatory Commission, August 1977.
- 5. Regulatory Guide 1.102, " Flood Protection for Nuclear Power Plants", U.S. Nuclear Regulatory Commission, September 1976.
- 6. ANSI /ANS 2.8, " Standard for Determining Design Basis Flooding at Power Reactol ites", American Nuclear Society.
t ,/ ~
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,s Amendment F A-47 December 15, 1989
CESSARMWcm., !i 1 O l I 1 l 1 1
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CESSAR nutrico... _ O' 105: INTERFACING SYSTEMS LOCA AT LWB.E 1R&B Generic Sefety Issue (GSI) 105 in NUREG-0933 (Reference 1), addresses reducing the risk of loss of primary coolant outside
-the containment via a system - which connects with the reactor coolant system (RCS).
The interfacing system LOCA (ISL) is presumed to result ~ from-exposing low pressure piping (design pressure 400-700 psi) of the interfacing system to high primary system pressure - (about 2250 psi). The initial plant response to a ISL is the same as the , response to an equivalent sized LOCA inside containment. containment and is However, not -returned RCS inventory is discharged to the containment outside sump for - recirculation. In addition, an ISL will provide a path, which bypasses the containment, for release of radioactive materialA The NRC used probabilistic risk assessment (PRA) analysis in NUREG/CR-5102 (Reference 2) to evaluate three older operating
- - pressurized water reactors ' for the effectiveness of. proposed
; -- requirement changes in reducing the risk of an interfacing system
=
LOCA, and for calculating the contribution of the ISL to the overall core damage frequency estimate. ACCEPTANCE CRITERIA G The acceptance criterion for the resolution of GSI 105 is that the-arrangement and design of systems that interconnect with the RCS-and also extend beyond the containment shall be such as to
. minimize the probability of an interfacing system LOCA.
Identification of the potential pathways for an ISL and estimation of the probable ISL frequency of occurrence shall be includod in the overall plant PRA required by 10 CFR, Part 52 (Ref m nce 3). REBOLUTION ISLs are addressed in the System 80+ Standard Design via design features and by evaluation of the potential leakage paths. Analyses of previous designs have identified several potential paths (interfacing piping of all sizes) for an ISL, with the most significant being the suction lines of the Shutdown Cooling Syatem (SCS) and the injection lines of the Low Pressure Safety Injection System. To minimize the possibility of an ISL outside the containment building, changes have been incorporateo in the System 80+ Standard Design. These changes are summarized as follows: 4 Amendment G A-47b April 30, 1990
y CESSARin h ,.., The SCS (CESSAR-DC, Section 5.4.7) for the System 80+ Standard Design contains two suction lines which are L similar to those of ; previous designs. However, the design pressure for the SCS has been increased from 650 psia to 900 psia (see CESSAR-DC, Section-5.4.7, Table 5.4.7-1), and the ultimate strength of the piping- i material will not be exceeded even if the system is accidentally i subjected to a pressure of more than 2000 psi. Also, each suction line of the shutdown cooling system is isolated frem normal operating pressure of the reactor coolant system by three motor operated valves in series. The two SCS return lines connect into the Safety Injection System (SIS) lines to twoof. the four direct injection nozzles on the reactor vessel. Each SCS return line is icolated from the normal RCS. operating pressure by two check valves in the corresponding SIS line, and by a third cneck valve-in the SCS. The SIS' (CESSAR-DC, Section 6.3.2) for the System 80+ Standard Design doe 9 not have a subsystem for low pressure injection. The SIS consir,ts of four redundant trains each of which injects directly into the reactor vessel. Each train consists of high-pressure pi; Lng which is capable of withstanding the normal ' operating pre u re of the RCS, Also, each train is isolated from the RCS during normal operation by three valves in series (two check valves and one motor operated valve) . Because the design-pressures of the SIS and RCS are approximately the same, the direct vessel injection paths are not considered to be potential paths for an interfacir.g system LOCA. In addition to the modifications to the SCS and SIS, changes have g also been incorporated in the letdown portion of the Chemical and Volume Control System '(CVCS ) (CESSAR-DC Section 9.3.6). The letdown heat exchanger is located inside the containment along-with additional isolation valves, and is designed to withstand the full RCS operating pressure. Several valves in series are ! located upstream of the letdown and regenerative heat exchangers. These valves are normally open during normal plant operation and any one of the valves may be closed - if required to isolate ! letdown flow. As a result of these system design changes, the PRA of the System 80+ Standard Design in CESSAR-DC Appendix B identifies the four most significant potential paths for interfacing system LOCA's as ; u L the suction and -return lines of the SCS. Other systems which L interface with the RCS include the SIS, the CVCS and the Sampling System. Potential paths for ISLs are listed in CESSAR-DC, Appendix B, Table B3.1.13-1. Except for the four SCS paths, other potential paths are considered to be non-credible from the standpoint of assessing core damage frequency for one or more of the following reasons: (1) The system piping is designed to accommodate pressures l greater than 2000 psig, Amendment G A-47c April 30, 1990
. y :
L LCESSAR !!n%mo, 1 l '- 1 Qf 4 (2) Charging pumps can make up lost inventory and allow plant l" cooldown,
-(3) -The break can only credibly occur inside containment, or >
(4) Tha flow ; path contains normally open valves which can be closed to isolate the break. The System 80+ PRA results show that ISLs provide only a minor ! contribution to core damage frequency (i.e., a contribution of. approximately 3.0E-9 relative to the core damage frequency goal . of 1.0E-5). ?
-In summary, the System' 80+ Standard Design minimizes the likelihood of an interfacing system LOCA by (1) eliminating low pressure safety injection, one of the most likely leakage paths of previoos designo, (2) increasing the over-pressure capability I of the SCS, the next most likely leakage path, (3) increasing the r design pressure of other components (e.g., the letdown heat i exchanger), 'and (4) improving the location of components to reduce ISL probability. This conclusion is supported by the overall plant PRA, which shows that the contribution of ISL to-
- q the overall core damage frequency is insignificant. This issue.
is therefore resolved for the System 80+ Standard Design.
. 4] {
j BE1 RENCES G !
- 1. NUREG-0933, "A Status Report on Unresolved Safety . Issues",
1
'U.S.
Nuclear Regulatory Commission, April 1989.
- 2. NUREG/CR-5102, " Interfacing Systems -LOCA: Pressurized Water ,
Reactors", U.S. Nuclear Regulatory Commission. ;
- 3. 10 - CFR, Part 52, "Early Site Permits; Standard ' Design r Certifications", Code Of Federal Regulations, Office of the Federal Register,, National Archives and Records
'I Administration.
i i () Amendment G A-47d April 30, 1990
;CESSAR neificuia i
O 106: PIPING AND THE USE OF COMBUSTIBLE GASES IN VITAL AREAS-
'IBSUE Generic Safety Issue (GSI) 106 in NUREG-0933 (Reference 1),
addresses the issue of combustible gases accumulating in buildings containing safety-related equipment. Except for hydrogen, most combustible gases are used in limited quantities and for relatively short periods of time. -Hydrogen is stored in high pressure storage vessels and supplied to various systems in the Auxiliary Systems Building through small diameter pipe. A leak or break in this pipe could result in a combustible or explosive mixture of air and hydrogen posing a potential loss of safety-related equipment. SRP Section 9.5.1, (Reference 2) addresses this concern for plants under construction and new plant designs. ACCEPTANCE CEITERIA
-The acceptance criterion for the resolution of USI 106, is that the hydrogen and other combustible piping be designed to pteclude large releases and accumulation of combustible or explosive gases
- in buildings which-enclose safety-related equipment. This can be
; - accomplished either by designing piping to preclude failure or
.providing means to limit the amount of hydrogen leakage in the event of a pipe rupture. Furthermore, in consideration of the above, the designer shall consider the guidance described in SRP Section 9.5.1.
RESOLUTION The System 80+ Standard Design incorporates various Compressed Gas Systems as desc11 bed in CESSAR-DC, Section 9.5.10. The compressed gas systems provide a variety of gases (e.g., hydrogen and nitrogen) under pressure, for numerous plant operating applications including welding, equipment, instrumentation, system purging, inerting and diluting. The systems typically consist of high pressure gas cylinders, pressure regulators and piping to distribute the gases throughout the plant. These non-safety-related compressed gas systems are designed to assure that their failure does not jeopardize the operati n of any safety-rclated s/ stem and/or component (see CESSAR- C, Section 9. 5.10.1) . Furthermore, with respect to the hydrogen compressed gas system, the system is designed to be isolable and a leak detection system is included. Also, Amendment F A-48 December 15, 1989 l i
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; CESSARHub -
O The IPSO panel = receives data from both the DIAS and. DPS via different data links. The IPSO keeps operations personnel informed about the status of the plant's critical. safety functions and success paths as described in CESSAR-DC Section i 18.7.1.2. It also provides a limited set of key plant I parameters. Implementation of the IPSO panel hardware considers- [ I redundancy for enhanced reliability. i. The DPS is - configured redundantly for improved reliability. It acquires plant data (e.g., process variable and component status) j validates.it, and executes applications programs for its display . page hierarchy. The portion which. addresses SPDS requirements 1 includes: IPSO , . critical- safety- functions, and -success path ; monitoring to aid the operator in gathering supporting _l information and problem diagnosis (see CESSAR-DC, Section 18.7.1.8.2). This is the primary means of implementing the SPDS functions in the ACC. Finurec 7.7-16 and 7.7-17 in CESSAR-DC show the basic cobfiguration of the DIAS design. The DIAS employs discrete l indicators that are used to display validated safety and non-safety-related plant process parameters including chose 3 required by - the SDDS functions. It uses a segmented design to . provide a degree of hardware independence and fault resistance ' between various segments. The DIAS channel P (DIAS-P) segment is designed to be physically separate -from and electrically , independent of the remaining DIAS channel N (DIAS-N) segment and j the DPS such that a single failure.will not cause a loss of more than one of the three display methods (DIAS-P, DIAS-N or DPS). .i In summary, the SPDS functions identified in NUREG-0737, I.D.2, are performed by IPSO, DPS, and DIAS in the Advanced Control ; Complex.. Each system incorporates improved design features such as separate and redundant hardware, power supplies (including battery backup), and system self-test features. These design features assure that the IPSO, DIAS, and DPS are very reliable thus minimizing the availability concern associated with the SPDS. Therefore, this issue is resolved for the- System 80+ Standard Design. REFERENCES ,
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
- 2. NUREG-0737, " Clarification Of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission.
- 3. Generic Letter No. 82-33, Supplement 1 to NUREG-0737, U.S.
O- - Nuclear Regulatory Commission. Amendment F A-63 December 15, 1989 l
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O Amendment G A-63a April 30, 1990
CESSAR ?!nLm i g i 125.II.07: REEVALUATE PROVISION TO AUTOMATICALLY ISOLATE FEEDWATER FROM STEAM GENERATOR DURING A LINE BREAK i 1 ISSUE ! Generic Safety Issue (GSI) 125.II.0*? in NUREG-0933 (Reference 1), o addresses the need for owner-ope c ators ' and p '.nt designers to i
- eassest the benefits of automatically isolating the ' emergency
-l feedwater.(EFW) system after a main steam line or main feedwater l line break. l Automatic isolation of EFW from a steam generator (SG) can help to mitigate the consequences of the break. Typically, upon a low SG pressure - signal, main steam isolation valves are closed and EFW io isolated from the depressurizing or faulted SG. This minimizes blowdown from the line break, and limits primary system overcooling and-the potential for a return to criticality. If-the EFW . were not isolated the peak pressure in the' containment' for secondary side breaks could exceca that due to a large break LOCA, the usual basis for containment design. The . automatic isolation logic . also diverts EFW flow from the faulted to the intact SG.
In contrast,. there are disadvantages to automatic isolation of EFW. If both channels of the controlling isolation logic systems were to spontaneously actuate either during normal operation or in the course of a transient, the availability of EFW would be lost and the main steam. isolation valves would-close. Most newer 6 plants use turbine-driven main feedwater pumps. -Thus, main feedwater would also be lost, resulting in complete loss of the. secondary heat sink. Capabilit'y to lock-out theLisolation logic is necessary'to preclude such scenarios.
- Proper and timely operator action following the various loss of cooling events is essential to attaining cold shutdown with minimum adverse consequences. If follows that the time available
'for accurate diagnosis of a problem by the operator before having to take action becomes an important factor. In addition, beca -;e s steam or feedwater line breaks make only a small contribution co the probability of core damage, the NRC concluded that plant safety would not be significantly improved or degraded by either the exclusion or inclusion of the automatic EFW isolation feature.
Therefore, the choice to include automatic EFW isolation in the design is dependent upon the containment design and the time that can be made available for operator action. Amendment G A-63b April 30, 1990
" " " ' " " " " " " " " " ' " " - - - ' - - - - - -----imm-a m- , , -
CESSAR ;!nincuoi j' 9: ACCE?TANCE' CRITERIA The acceptance criterion for the ' resolution of GSI 125.II.7 is that ; the design of a new plant need not incorporate automatic isolation of emergency feedwater afte: a main steam line'or main
-feedwater line break . provided that: (1) the containment _ design -
can accommodate _ the peak containment pressure, taking into account'the-e.ffects of EFW flow to the faulted: steam generator,- and (2) there is an adequate period of time following the .- break
.for the operators to diagnose the event and regulate or terminate EFW flow to the faulted steam generator to prevent primary system overcooling or containment overpressurization.
The plant' designer shall therefore define the mass and energy input to containment to include flow of emergency feedwater to-the affected steam generator following a main steam line break. It should be assumed that the operators will- not take action to terminate the flow of emergency feedwater to the affected steam generator within 30 minutes of the break, based upon current industry recommendations. RESOLUTION The System 80+ Standard Design does not include automatic steam generator. isolation logic. The calculated mass and energy- G. , release to the containment building as the result of a main steam line break- (MSLB) includes the additional mass and energy introduced ' from the MSLB because of emergency feedwater flow.
'This: additional mass-and energy addition is assumed to continue g for at'least 30 minutes after a MSLB.
The plant design incorporates an Emergency Feedwater System which-provides an independent safety-related means of supplying quality feedwater to the steam generator (s) for removal of heat and prevention of reactor core uncovery during emergency phases of
' plant operation. The EFW System is a dedicated safety-related system which has no functions for normal plant operation (See CESSAR-DC, Section 10.4.9).
The EFW System is designed to be automatically or manually initiated, supplying feedwater to the steam generators for any event that results in the loss of normal feedwater and requires heat removal t.hrough the steam generators, including the loss of normal onsite and normal offsite AC power. ! Four-channel c etrol logic is provided, so that a single failure neither opuriously actuates nor prevents EFW supply. In addition, manually reset vexiable setpoints are used, to enable cooldown to be achieved without actuating the main steam isolation signal. l I Amendment G A-63c April 30, 1990 l l
CESSART!n%.m ,
.E r
j
. (N!-
.i v.
The design' criteria for the EFW System include a~ requirement that i .the maximum EFW flow to 'each steam generator be restricted by a , cavitating venturi : to protect the EFW pumps from damage.due to excessive runout flow. This flow restriction' also permits the P ' operator 30 ' minutes to regulate or terminate EFW flow to prevent primary. system overcooling, steam generator overfill,. or containment overpressurization (See CESSAR-DC, Section ;
- 10. 4 . 9.1. 2 )'. _ In addition, the EFW System has a feedwater storage volumeL of. 350,000 gallons in two safety-related emergency ;
feedwater storage tanks (EFWSTs) to achieve safe cold shutdown. This-volume allows for a main feedline break without isolation of. L EFW flow to the affected steam generator for 30 minutes. 1: Thus, . sufficient emergency feedwater can be provided at the \ required temperature and pressure even if a steam line or feed [' line pipe break is the 4.nitiating event, if any one EFW pump-subtrain fails to deliver flow (the EFW train for each steam generator has two full capacity pump subtrains), and if no operator action is taken for up to 30 minutes following the
- event.
Also, an adequate emergency feedwater supply is available to p allow the plant to remain at hot standby for 8 hours followed by. f,'j an' orderly. cooldown- to the primary system. pressure and temperature at which the Shutdown Cooling System (SCS) ' can be _ initiated to continue cooldown to cold' shutdown conditions. Level G ;
. instrumentation and a low level alarm are provided on each EFWST to help the operator align the . EFWST from the other train to preclude the tank from being emptied before the changeover to SCS cooling can be effected.
In the Main Feedwater System, three . of the pumps are steam turbine driven. The. fourth pump'is motor driven and is normally. only used for startup and shutdown. This pump starts-K automatically on loss of main feedwater and reactor trip (see , CESSAR-DC, Section 10.4.7.2.3.D). , L t Since the System 80+ Standard Design meets the criteria stated l
.above, this issue is resolved.
I REFERENCES
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989, o 7 (V Amendment G A-63d April 30, 1990
lCESSARi!!nnenc,r w 128: ELECTRICAL POWER' RELIABILITY I ISSUE i i Generic' Safety Issue (GSI). 128 in NUREG-0933- (Reference ~ 1), addresses- the reliability of- onsite electrical systems.: NUREG-0933- combined three GSI's previcusly individually listed ;
.under' NUREG-0737 (Reference 2) in order to provide- a more integrated approach to resolving.these interrelated issues.
ACCEPTANCE CRITERIA ! The acceptance criteria for the resolution of GSI 128 are' encompassed in ' other GSI's namely 48, 49, and A-30, which are , given in NUREG-0933. l RESOLUTION The resolution for GSI 128 is identified in the responses to GSI's 48, 4 9, - and ~ A-30 which are addressed and resolved' in-CESSAR-DC, Appendix A. Since GSI 128 is-subsumed by these three l GSI's, this issue is resolved for the System 80+ Standard Design. REFERENCE _S 1.- NUREG-0933, " A Status Repor'; On Unresolved Safety Issues", U.S. Nuclear Regulatory Commission,-April 1989.
- 2. NUREG-0737, " Clarification of TMI Action Plan Requirements"',
U.S. . Nuclear Regulatory Commission, October 1980. i 1 I l' l l ll O Amendment F A-64 December 15, 1989 l .
+ .
3 iCESSARiinincui. j.- 1 J A-01: WATER HAMMER ISSUB q
. Unresolved Safety Issue -(USI) A-01 in NUREG-0933 (Reference 1) addresses identifying the - probable causes of water hammer and minimizing the susceptibility of-fluid systems and components to water hammer by correcting design and operational deficiencies.
Water hammer is defined as a rapid deviation in. pressure caused by a change in the velocity of a fluid in a closed volume.- There are various types of water hammer, including steam condensation-induced water hammer, which occurs in the secondary side.of a-PWR steam generator at the connection to the'feedwater ( line. This type of water hammer involves steam generator feedrings end piping. Water hammer has been observed in many fluid systems including residual heat removal, containment spray,
-service water, feedwater systems, and main steam lines. In addition. to condensation-induced' water hammer, other forms of initiating events which cause water hammer can occur, such as steam driven slugs of water, pump startup with partially ' empty ,
l >> j)- b lines, and rapid valve cycling. Regardless' of the initiating svent, water hammer and the 1. resulting fluid accelerations can cause damage to.the affected fluid system.-1The level of severity of damage depends upon the event, and can range from minor damage such as overstressed pipe J hangers to major damage to restraints, piping and components. g According to NUREG-0927 (Reference 2), water hammer can be i induced- by operator / maintenance actions and by design inadequacies. Experience has shown that water. hammer. events reported on LERs are about equally divided between - operator or E maintenanca actions and design deficiencies. The NRC implemented _' [ SRP changes relative to the design, operation, and maintenance of ! new plants to minimize the probability and effects of water l hammer, and issued a Branch Technical Position (BTP) for pre-operational tests. 1 BCCEPTANCE CRITERIA 1. l The acceptance criterion for the resolution of USI A-01 is that safety-related fluid systems shall be designed to meet the requirements of 10 CFR 50 Appendix A, GDC 4 (Reference 3), by implementing the guidance identified in the following SRP (Reference 4) Sections: 5.4.7, 6.3, 9.2.1, 9.2.2, 10.3, and 10.4.7 (including BTP ASB 10-2).
' (O) Specifically, the Feed and Condensate, Emergency Feedwater, Main Steam, Safety Injection, Containment Spray, Shutdown Cooling, Amendment G A-66a April 30, 1990
e CESSARina mw and Service Water fluid systems shall, e Component- Cooling, in order to- assure that .the system- safety functions can be accomplished, be designed to withstand the adverse dynamic loads imposed by condensation-induced and other water hammer events, and include features to minimize the probability of water hammer occurrences. BTP ASB 10-2 requires pre-operational tests to be performed on the feedwater system -to demonstrate the effectiveness of the design and operating procedures to inhibit steam generator water hammer. In addition, operating and maintenance procedures shall include adequate precautions to minimize the potential for the occurrence of water' hammer. RESOLUTION The System- 80+ Standard Design adequately addresses system
. dynamic loads -such as may result from water hammer. -Each feedwater nozzle of- the System 80+ steam generators has a
. ectly attached 90-degree elbow turning down into a vertical
- ion of feedwater piping, as described in CESSAR-DC Sections it, .7.2.2.E and 10.4.9.1.2.M. In addition, the main and emergency feedwater piping is continuously sloped away from the steam generators inside containment to prevent draining into the steam generator, minimizing the formation of steam voids in the -
feedwater piping during periods of low flow. Main feedwater flow during plant startup is only delivered to the economizer inlet feedwater nozzles of the stean generators at temperatures at or above 200*F, which minimizes the probability G-of condensation-induced water hammer in the economizer sections of the generators. Below a predetermined power level,- main feedwater is delivered to the downcomer feedwater inlet nozzles. Changeover to the economizer nozzles at this power level is effected by the Main Feedwater Control System. Emergency
-feedwater is always delivered to the downcomer inlet nozzles.
(See CESSAR-DC, Sections 10.4.7.2.1D and 7.7.1.1.4.) The. design of the System 80+ Main Steam System adequately addresses dynamic loads caused by condensation-induced water hammer and has piping arrangement and drainage provisions to protect against water entrainment. These are described in CESSAR-DC, Sections 10.3.2.2D, M, N, and P. Dynamic loads, and the provision of vents and drains where appropriate, are also addressed in the design of the Shutdown Cooling, Safety Injection, Containment Spray, Component Cooling and Service Water Systems. These systems are described in CESSAR-DC, Sections 5.4.7, 6.3.1, 6.5.1, 9.2.1, and 9.2.2 respectively. Amendment G A-66b April 30, 1990
LCESSARfd!Oicus
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1 % v.._/ Plant operating and maintenance procedures are prepared by the owner-operator, in accordance with guidelines established by C-E, l
' and require proper precautions to minimize the potential for water hammer.
. . . I Guidelines are also established for hot functional tests to-be performed by the owner-operator in accordance with BTP ASB 10-2.
These' tests are to verify that unacceptable feedwater system water hammer'does not occur when,(a)~using normal plant operating-procedures-for normal and emergency restoration of SG water level following loss of main feedwater, and. (b) transferring main feedwater d'1 ring normal operation from the'SG downcomer feedwater inlet nozzlee to the , . economizer inlet nozzles. .(See CESSAR-DC, Sections ^ 14.2.12.1.63 and-14.2.12.4.13, respectively.) i Since the design'and testing of the safety systems potentially subject to water hammer meets the' intent of the acceptance criteria above, this issue is resolved for the System 80+ Standard Design.
-REFERENCES
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
L 4 U.S. Nuclear Regulatory Commission, April 1989,
- 2. NUREG-0927, Revision 1, " Evaluation of Water Hammer Occurrences in Nuclear Power Plants", U.S. Nuclear Regulatory Commission, March 1984. >
- 3. 10 CFR 50 Appendix A, " General Design Criteria for Nuclear G Power Plants", Code of Federal Regulations, Office of the Federal Register, National Archives and Records l Administration.
- 4. NUREG-0800, Standard Review Plan for the Review. of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
Nuclear Regulatory Comnission. l
.O Amendment G A-66c April 30, 1990
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O Amendment G A-66d April 30, 1990 1
;CESSAR 'Jnisma, R, )
A-44: STATION BLACDUT ISSUE Unresolved' Safety Issue (USI) A-44 in NUREG-0933 (Reference 1l, addresses,the concern:that the complete loss of all alternating-current (AC) electrical power *:o the essential and non-essential switchgear buses -in a nuclea power plant, referred to as a
" Station Blackout", could lead to'a severe core damage accident.
10 CFR 50.2, " Definitions", dafines station blackout as i.he loss of the of fsite' electric power system concurrent with a turbine .
. trip and unavailability of the onsite emergency AC power system. '
l It does no_t include the Icss of available AC power to buses fed by station batteries through inverters .o r by alternate AC sources. -Since many systems required for core decay heat removal and containment heat removal depend on AC power, a station blackout can result in unacceptable consequences unless AC power is restored.in'a timely manner.-or AC power is supplied from an alternate source. The issue involves the likelihood and duration of station ' blackout and the potential for core damage as a g result. The issue - of station blackout arose because of the historical experience regarding the reliability of AC power supplies. There had been-numerous reports of emergency diesel generators failing to start and run in operating plants. In addition, a number of G operating plants experienced a total loss of offsite electrical , power. After performing an evaluation of station blackout accidents at' nuclear power plants [NUREG-1032 (Reference 2)] and a Regulatory Impact Analysis (Reference 3), the NRC published a j new rule [10 CFR 50.63 (Reference 4)] and Regulatory Guide 1.155,
" Station Blackout" (Reference 5).
Specifically, Regulatory Guide 1.155 specifies that a coping analysis should be performed for a Station Blackout to establish that a. plant can tolerate a total loss of AC power. However, Regulatory Guide 1.155 also states that if a plant is provided I with an alternate AC power (AAC) source which is av;ilable within 10 minutes of a station blackout then a coping cnalysis for the event need not be performed. In addition, Regulatory Guide 1.155 identifies reliability criteria for the emergency diesel L- generator. Tne reliability of the emergency diesel generator is discussed.in a related safety issue, USI B-56. f'\
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Amendment G A-100a April 30, 1990
CESSARHahm . O ACCEPTANCE CRITERIA
-The acceptance criteria for USI A-44-are that the plant- design . meet the intetit of Regulatory Guide 1.155 and be capable of ' maintaining core cooling and containment integrity during' a station blackout.- According to 10 CFR 50.63, onsite alternate AC power sources that are independent and diverse from the normal Cl_ss a 1E emergency AC . power sources constitute ~ an acceptable Station Blackout event coping capability. (The acceptance criteria for the reliability of the emergency diesel generator.is identified in USI B-56).
EpBOLUTION: The System 80+ Standard Design includes improved design features and electrical systems to ensure a safe shutdown of the reactor in -the event of a station blackout. These improvements are summarized below: (1) One -turbine-driven emergency feedwater pump is included for. each steam generator. (These are in addition to the two-motor-driven emergency feedwater pumps.) In previous designs one turbine-driven pump was shared by both steam generators. (2) Each of the four safety-related instrument channe3s has a battery backup. In addition, Class 1E Electrical Divisions g I and II, which include the emergency diesel generaters, have their own batteries. (3) The design has full load rejection capability and the capability to subsequently provide electrical power from the turbine generator. (4) An alternate source of AC power which is diverse from the safety-grade emergency diesels is included (See CESSAR-DC, Section 8.3.1.1.5). This alternate AC source is a control-grade gas turbine and has its own battery. As described in CESSAR-DC Sections 8.1.4.2 and 8.3.1.1.5, the installation and design of the AAC source is in compliance with the intent of Regulatory Guide 1.155. The AAC power source is designed to be made available to power one safety-related load division and its corresponding essential non-safety-related load bus within 10 minutes of the onset of the station blackout, such that the plant is capable of maintaining core cooling and containment integrity. In addition, sufficient fuel is stored on site to support 24 hours of AAC operation at rated load. O Amendment G A-100b April 30, 1990
4l
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j' -A) 6 y -The AAC source is-not normally directly connected to the plant's main or standby _offsite power sources or to the Class 1E' power distribution system, thus minimizing the potential for common cause failure. Quality assurance guidelines for the AAC are glven .in CESSAR-DC Section 8.3.1.1.5.5. (The ' emergency diesel l: ger.erator reliability aspect of Station Blackout as described in Regulatory. Guide 1.155 is. covered in USI B-56.)' .;. In suamary, 'the System 80+ Standard Design accommodates the l
' Station Blackout event by means of an' AAC power source- and reliable emergency diesel generators and therefore, meets 10 CFR s 50.63 and the intent of Regulatory Guide'1.155. Thus, this issue L is resolved for the. System 80+ Standard Design.
REFERENCES
- 1. NUREG-0933, "A Status Report on. Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989. )
- 2. NUREG-1032,- " Evaluation of . Station Blackout ' Accidents at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, June 1988. :
- 3. NUREG-1109, " Regulatory /Backfit Analysis for the Resolut.lon
'-' -of Unresolved Safety Issue A-44, Station Blackout," U.S. 1 Nuclear Regulatory Commission, June 1988.
l
.4. Federal Register Notice 53 FR 23203, "10 CFR 50, . Station g' Blackout", June 21, 1988.
- 5. Regulatory Guide 1.155, " Station Blackout", U.S. Nuclear Regulatory Commission, August 1968.
i l l' . l ' Amendment G l A-100c April 30, 1990
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f % ( LJ A-47 SAFETY IMPLICATIONS OF CONTROL SYSTEMS h ISSUE Unresolved. Safety Issue (USI) A-47 in NUREG - 0933 (Reference 1). concerns the potential for accidents or transients being made more severe as a result of control system-failures.
- )
Non-safety-grado control systems are not relied on to perform any_ t safety functions, but they are used to control plant processes that could have a significant impact on plant dynamics. For the resolution of USI A-47 tne NRC evaluated the effects of-control plants, i system- failures- on three operating PWR reference including ~ a representative C-E design, subjected to single >and I multiple control system ' failures durjng automatic- and manual modes of operation. The evaluation was documented in NUREG-1217
-(Reference 2), and identified two concerns related- to the C-E q '
-design: (1) steam generator (SG) overfill, and (2) reactor core heat. removal to cold shutdown after a small-break loss-of-coolant accident (SBLOCA), without overcooling the reactor vessel.
NUREG-1217 conclusions were that: (1) PWR plant designs having a fq redundant , control grade method of overfill protection are L 1 considered acceptable, and (2) PWR plant' designs that have the 1 capability of high pressure safety injection at more than 1275~' g psia, and that provide automatic initiation of- emergency feedwater flow on low SG level, are considered acceptable ' for small-break LOCA core heat removal. Manually controlled methods ' of heat: rejection and reactor coolant system depressurization without' compromising the reactor vessel nil ductility transition reference temperature (RT limits are also considered acceptableifdetailedin-emke)ncyprocedures. In addition, the NRC issued GL 89-19 (Reference 3), which pursuant to 10 CFR 50.54 (f) requires all operating PWR plants and: plants under construction to provide automatic protection from SG overfill by main feedwater. t
/
ACCEPTANCE CRITERIA The first acceptance criterion for the resolution of USI A-47 is ; that the plant shall have, as a minimum, control-grade protection against SG overfill by main feedwater, consistent with the requirements and guidance of GL 89-19. Also, in accordance with GL 89-19, technical specifications and plant operating procedures shall ensure in-service verification of the availability of the overfill protection.
/ }
Amendment G A-102a April 30, 1990 l
b CESSAR W h c. O' A second criterion based.on the conclusion of NUREG-1217 'is that 1 the safety. injection ~ pressure capability' should be greater than 1275 psiaand emergency feedwater should- be automatically-initiated on a low SG water level signal. RESOLUTION The System 80+- Standard Design has a Main Feedwater Isolation System;(see CESSAR-DC Section 10.4.7.2.2) to protect the SGs from-overfill. The system' includes redundant- remotely operated isolation valves in each main feedwater line to each SG. The valve actuation . system (see CESSAR-DC Section 7.3.1.1.10.3)- is , composed- of- . redundant trains A and B, and- each train's
-instrumentation and controls are physically and electrically ;
separate from and independent of those of the other train. A failure of one - train will not impair the action - of the other. The main feedwater isolation valves are automatically actuated by a Main Steam Isolation Signal (MSIS) from the Engineered Safety i Featurns Actuation System (ESFAS, see CESSAR-DC Section 7.3.1). High SG water level, in a 2-out-of-4 logic, is one of the initiators for the MSIS. The main feedwater isolation valves can be in-service tested in accordance with ASME Code Section XI, subsection IWV. A_ Technical Specification (CESSAR-DC Chapter 16) will establish testing requirements for the valve- actuation system. These ' requirements will also be incorporated into the plant maintenance procedures. g i s In a S BLOCA , high pressure safety injection in the System 80+ , Standard Design is delivered at. a pressure considerably above 1275 psia. The System 80+ Standard Design also incorporates a safety-grade Emergency Feedwater System (EFWS, see CESSAR-DC Section 10.4.9) : which is automatically actuated by an Emergency Feedwater Actuation Signal (EFAS) from the ESFAS. There is one EFAS for each SG, initiated by low SG water level in a 2-out-of-4 logic -(see CESSAR-DC Section 7.3.2.2.5). . The EFWS, in conjunction with safety-grade atmospheric steam dump valves, provides an independent means of residual heat removal from the
' Reactor Coolant System (RCS) via the secondary system until the RCS pressure and temperature permit actuation of the Shutdown Cooling System. An ESFAS high SG water level interlock will isolate EFW to preclude SG overfill. The RCS depressurization rate is manually controlled by the operator from the control room to prevent overcooling of the reactor vessel, by throttling the EFW and/or using the pressurizer auxiliary sprays.
O Amendment G A-102b April 30, 1990
CESSAR1HFificui. i v-In addition to the above design features, guidelines are provided- ,
.to' assist the owner-operator- in preparing emergency operating procedures detailing the actions to be taken- by the plant operators in the event of 'a SBLOCA, as contained in report CEN-152 (Reference 4).
^
In summary, consistent with the requirements and guidance of
- GL 89-19, the System 80+ Standard Design incorporates: (1) SG ,-
overfill. protection, and (2) an automatically initiated safety-grade Emergency Feedwater System. Furthermore, a Technical -Specification for verifying overfill protection availability, and emergency procedure guidelines for a SBLOCA, are established. This issue, is therefore, resolved for the System 80+ Standard Design. BXFERENCES ,
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues", .,
U.S. Nuclear Regulatory Commission, April 1989.
- 2. NUREG-1217, " Evaluation of Safety Implications of Control Systems in LWR Nuc, lear Power Plants", - Technical Findings p Rolated to USI A-47 (Draft Report), U.S. Nuclear Regulatory Ccmmission, April 1988. ,
- 3. Generic Letter 89-19, " Request .for Action Related to 6
Resolution of USI A-47,. Pursuant to 10 CFR 50. 54 ( f) " , U.S. Nuclear Regulatory Commission, September 20, 1989. j
- 4. CEN-152, Revision 3, " Emergency Procedure Guidelines",
Combustion Engineering, Inc. 3 i i i v Amendment G A-102c April 30, 1990 a
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CESSNR ;!nifieni:,, l f% ': j' , 1 :i p Y 2. [101 CFR 50.61, " Fracture Toughness Requirements for
- !- Protection .Against Pressurized Thermal Shock Events",
' Federal Register, July 23,1985.
't 7
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3.. American Society of. Mechanical' Engineers,. Boiler.& Pressure- . "U Vessel Code,--Section III (Nuclear), American- Society of J Mechanical Engineers. J Ll o
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Amendment F A-105 December 15, 1989
.CESSAR1Enemo,, ;
B-53: LOAD BREAK SWITCH ISSUE Generic Safety Issue (GSI) B-53 in NUREG-0933 (Reference 1) addresses the use of load break switches or circuit breakers in some onsite/ offsite power system designs and_their reliability with respect'to the requirement identified in GDC 17 (Reference ;
- 2) to provide power from the offsite transmission grid to all station vital loads on a loss of all onsite power via at least two seperate circuits,one of which must be immediately _;'
acce v:ble. Most nuclear power plants meet the requirements identified in GDC 17 by employing designs which involve startup transformers and auxiliary transformers'in addition to the station transformer. A few operating' plant designs employ a load break switch or circuit breaker to remove the turbine generator from the station stepup transformer after a turbine trip. This permits offsite power to be provided in the reverse direction through the station transformer to the onsite power system and its vital loads. The NRC expressed a concern about the reliability of these breakers' and switches because, should they fail and not isolate the turbine generator, the offsite power system would be incapable of . G supplying the onsite power system as required by GDC 17. New requirements have, therefore, been established by the NRC in Appendix A to SRP 8.2 (Reference 3) for plants whose onsite/offsite power systems propose to rely on load break switches . or breakers. These requirements entail additional _ and more rigorous testing, similar to that required for Class 1E (safety-related) equipment. Only circuit breakers, which have a ; 100%- fault -current interuption capability, are considered acceptable for generator isolation to assure an immediately accessible offsite power source. ACCEPTANCE-CRITERIA The acceptance criterion for the resolution of Generic Safety Issue B-53 (Reference 1), is that plants employing onsite and offsite power systems incorporating turbine generator circuit
- - breakers, shall meet the intent of guidance provided in SRP 8.2, i- Rev. 3,-Appendix A.
L Specifically, only devices which have maximum fault current l interrupting capability, i.e. circuit breakers can be used to L isolate the unit generator from the onsite-offsite power systems. O Amendment G l l A-105a April 30, 1990 1 1 I l
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In addition, - generator circuit b r e a k e r s , s h o u l d .- b e designed to-perform - their intended function during steady-state operation, power system transients and major faultu; tests should be
. performed.on the circuit breaker to verify.these capabilities. As l a minimum, the performance tests and capabilities should be demonstrated including, dielectric tests, load switching, fault current interrupting capability,_ maximum rise of recovery voltage,- short-time current carrying capability, momentary, current carrying capability, transformer magnetizing current!
interruption, thermal capability, and mechanical operation. These . tests should be in accordance with the applicable requirements of ANSI standards C37.04, C37.06, and C37.09 (References 4, 5, and 6). RESOLUTION The System 80+ Standard Design utilizes offsite and onsite power systems.to supply the unit auxiliaries _uring normal operation, and these plus the Reactor Protection System and Engineered Safety Features Systems during abnormal and accident conditions. In addition, the onsite and offsite power systems are designed in accordance 'with accepted industry codes and standards (see-CESSAR-DC Section 8.0), and do not employ load break uwitches.
~
The System 80+ unit is connected to a switchyard and the G transmission grid system via two separate and independent transmission lines. The power circuit breakers, along with the
-unit main transformers and standby auxiliary transformer, allow one of these lines not only to supply power to the transmission grid system during normal operation, but also to serve as an immediately accessible source of preferred onsite power. The other separate transmission line is connected, via the switchyard and standby auxiliary transformer, to provide a second independent immediate source of offsite power to the onsite power distribution system for safety-related and permanent non-safety-related loads (See CESSAR-DC, Figure 8.2-1). The power _ circuit breakers are designed, installed, inspected, and periodically tested as described in the acceptance criteria in accordance with ANSI Standards C37.04, C37.06 and C37.09
-(References 4, 5, and 6) and other accepted industry codes and standards, and therefore meet the intent of SRP 8.2, Appendix A (See'CESSAR-DC, Section 8.1.5). Descriptions of the offsite and onsite power systems can be found in CESSAR-DC Section 8.2 and
-8.3, respectively.
Since the intent of the acceptance criteria above is mat, this issue is resolved for the Syrtem 80+ Standard Design. O Amendment G A-105b April 30, 1990
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REFBERNGM -
- 1. 'NUREG-0933, "A Status Report on Unresolved Safety Issues", .
-U.S. Nuclear Regulatory Commission,. April 1989. - 1 2., 10CFR50' Appendix ~ A, " General Design- Criteria for Nuclear f
Power. . Plants", Office of the Federal. Register, National i Archives and Records Administration. ..
. s I
3.- NUREG-0800, Standard Review Plan for the Review of Safety' Analysis Reports' for Nuclear Power Plants - - LWR Edition", ; Nuclear Regulatory Commission. ]. 4.- : ANSIf Standard C37.04, " Rating Structure for AC High-Voltage. g Circ.uit 1 -Breakers Rated on a Symmetrical Current Basis",_ American National-Standards' Institute.
- 5. ANSI Standard C37.06, " Preferred Ratings and Related Required Capabilities for AC High-Voltage Circuit Breakers Rated on a Symmetrical Current Basis", American National. l j: Standards Institute.-
- 6. ' ANSI Standard C37.09, " Test Procedure for AC High-Voltage '
Circuit Breakers Rated on a Symmetrical Current Basis", - American National Standards Institute. - l i r O Amendment G ' A-10Sc April 30, 1990
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B-56: DIESEL RELIABILITY l ISSUE Generic: Safety Issue (GSI) B-56 in NUREG-0933 (Reference 1), ! addresses emergency diesel generator reliability. The reliability j goal identified in NSAC-108, (Reference 2) for emergency diesel generator startup, is between .95:and .975 per demand.. !
.i Typical onsite electrical distribution systems for plants use !
-diesel generators as an emergency . source of power. These ;
emergency power sources supply safety-related equipment, which is j
-used to prevent or mitigate accidents, in the event of a. loss of offsite power. j Because of the safety significance of the emergency diesel generators, limiting conditions for operation (LCOs) were developed and placed in the plant technical specifications. These LCOs require periodic testing. Licensee Event Reports (LERs) sent to the NRC document problems encountered during periodic testir - ,' !
i of the emergency diesel generators- (to demonstrate operability) . As discussed in NUREG-0933, a review of the LERs conducted by the g -l NRC. revealed that a diesel generators starting reliability is, on
.the average, about .94 per demand. Thus, the NRC determined that there was a need to upgrade the reliability of emergency diesel generators. A new reliability of between .95 and .975 per demand.
for emergency diesel generator design, operation and periodic testing, was established in Regulatory Guide 1.9, Rev. 3 (DRAFT) (Reference 3). The specific emergency diesel generator starting reliability identified in Regulatory Guide 1.155 (Reference-4) is the same as in~ Regulatory Guide 1.9, Rev. 3 (DRAFT) (i.e., it. ranges from .95 to .975 per. demand) . The resolution ' of a related Unresolved Safety Issue (USI) A-44, Station' Blackout, addresses the plant response to station blackout conditions. ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI B-56, is that emergency diesel generator design, operation, and periodic testing shall ensure, as a minimum, a starting reliability of .95 per demand, as identified in Regulatory Guides 1.9, Rev. 3 (DRAFT) and 1.155. O Amendment G A-105d April 30, 1990
- CESSART!nh w
' i RESOLUTION ,
The System ' 80+ Standard Design includes an onsite . electrical distribution system which employs two redundant and independent . Class 11E load group - divisions. The Class 1E safety loads, are '
' capable of being supplied, in decreasing priority, from the unit main turbine generator, unit ' main transformers, the emergency '
diesel generators and the alternate AC source _ (See CESSAR-DC, Section 8.1). 'i .
-As described above, each Class 1E division can be supplied with emergency. standby power from an independent diesel generator. The emergency diesel generator is designed and sized with sufficient capacity to ' operate all the . needed safety-related loads powered from its respective Class 1E Safety Division Bus. Furthermore,
~
each Division can be supplied from the alternate AC source (gas turbine) which!is diverse from the diesel generators. Also, the < alternate AC source is designed to the same reliability criteria as the emergency diesel generator. Each diesel generator is specified to start reliably. and, with present technology, industry exper.ience has shown that a starting reliability of '.986 per demand may be achieved as identified : -in the EPRI ALWR Utility Requirements Document, (Reference 5). The : emergency diesel generators' recul mi start and loading response times have been eased and the d. el generators are now required ; to . attain rated voltage and ' aquency and to begin accepting
. load within 20 seconds after rece'ipt of a start signal. This G
reduces their- starting stress .and contributes to improved reliability over the life of thJ units. .These response times are necessary to meet the times assumed in Chapter 15 Safety Analyses (See CESSAR-DC, Section 8.3.1.1.4). Improvements in the System 80+ Standard Design loading sequence logic prevent unnecessary load shedding and reloading due to subsequent emergency safeguards actuation signal (ESFAS)
- actuations. This provides additional overall reliability in response to changing plant conditions by reducing unnecessary transient demands on the diesel generators. 1 L A variety of tests are performed to assure emergency diesel
- generator reliability and operability. In addition to factory l tests, a number of preoperational and onsite acceptance tests and
! periodic tests are conducted on each diesel generator and auxiliaries. These tests are identified in CESSAR-DC, Section 8.3.1.1.4.11. Also, conditions for operation are imposed to ensure continual reliability. The periodic testing of the diesel generator meets the intent of Regulatory Guide 1.9, Rev. 3 (DRAFT). Amendment G A-105e April 30, 1990 l
CESSAR !!n%m.. = O In summary, the System 80+ Standard Design useo diesel generators as emergency power sources which are incorporated in the onrite electrical distribution syctem and which have a diverse backup (i.e., the alternate AC source). The onsite electrical distribution system meets the intent of the guittnce given in Regulatory Guides 1.9, Rev. 3 (DRAFT) and 1.155. Therefore, this issue is resolved for the System 80+ Standard Design. j REFERENCES k
- 1. NUREG-0933, "A Status Report on Unrcsolved Safety Issues",
U.S. Nuclear Regulatory Commission, January 1989.
- 2. NSAC-102 " Reliability of Emergency Diesel Generttors at G U.S. Nuclear Plants", Electric Power Research Institute, September isB6. ;
- 3. Regulatory Guide 1.9, Rev. 3 (DRAFT), " Selection, Design, Qualification, Testing, and Reliability of Diesel Generator Units .Used as Onsite Electrical Power Systems at Nuclear !
Power Plants", U.S. Nuclear Regulatory Commission, j November 1988.
- 4. Regulatory Guide 1.155, " Station Blackout", U.S. Nuclear Regulatory Commission, August 1988.
' 5. EPRI, "Advane1d Light Water Reactor Utility Requirements Document", Electric Power Research Institute, Chapter 11, ; April 1989. s i O i Amendment G A-105f April 30, 1990
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CESSAD !!Nncu.. O B-60: LOOSE PARTS MQ)TITORING SYS'tM : ISSUE - Generic Safety Issue (GSI) B-60 in NUREG-0933 (Reference 1), addresses the use of a loose parts monitoring system to detect debris in the reactor coolant system (RCS) which could damage RCS componants and/or fuel. A loose part - whether it be from an item inadvertently left in the primary system during construction, refueling, or maintenance, or from component failures - can contribute to further component damage and material wear by frequent impacting with other parts in the system. A loose part can potentially create a partial core flow blockage which could result in failure of fuel cladding. In aridition, a loose part may increase l the potential for control rod jamming and for accumulation of n increased levels of radioactive crud in the primary system. The primary purpose of the loose part detection program is the early detection of loose metallic parts in the primary system. Early detection can provide the time required to avoid or mitigate safety-related damage to, or malfunction of, primary system componenta. Therefore, the NRC established the guidance in Regulatory Guide 1.133, Rev. 1 (Reference 2). A9_QEPTANCE CRITERIA The accept 4nce criterion for the resolution of GSI B-60 is that a plant sSa]l have a loose _part monitoring system which is capable of early <t:.tection of loose metallic parts in the primary system. The system should have design features which are identified in Regulatory Guide 1.133 and include at least two acoustic sensors, an appropriate minimum system sensitivity and physical separation of each channel. In addition, the system should be designed with a data acquisition system with both manual and automatic start-up capability, an established " alert level" for loose parts, and capability for sensor channel operability testing. Finally, the system should be designed for expected environmental and seismic conditions, contain quality components, and provide 1, for enhanced maintainability. O Amendment F A-106 December 15, 1989
JESSAO!!nLui. ()m, i B-63: ISOLATION OF LOW PRESSURF SYSTEMS CONNECTED TO THE REACTOR COOLANT PRESSURE BOUNDARY ISSUE Generic Safety (GSI) B-63 in NUREG-0933 (Reference 11 addresses the _need to ensure the integrity (i.e., leak-tichtness) of boundary valves installed between high pressure (HP) (i.e., the Reactor Coolant System pressure boundary) and low pressure (LP) i safety-related systems, during plant operation by performing periodic inservice testing. The AME B&PV Code, Section III controls the design, fabrication, and initial testing of boundary and relief valves. During operation, the ASME B&PV Code, Section XI, specifies boundary and relief valve testing requirements to assure continued valve integrity. Because of the importance of the HP to LP interface for safety-related systems, the NRC reviewed and updated SRP Section 3.9.6 by issuing Revision 2 (Reference 2). 'l his SRP references and' endorses the ASME B&PV Code, Section XI (for the in-service
.-s testing of the boundary valves).
( ) G V' (A related issue, which also discusses the integrity of the HP to LP interface between safety-related systems is GSI 105,
" Interfacing Systems LOCA".)
ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI B-63 is that the periodic inservice testing of the HP to LP system boundary valv(s shall meet the intent of SRP 3.9.6, Revision 2. Because SRP 3.9.6, Revision 2, endorses the requirements of the ASME B&PV Code Section XI, periodic testing of these valves shall be performed in accordance with the code. Specifically, these boundary valves shall comply with the , requirements of the applicable IUV subarticles identified within Section XI of the ASME B&PV Code. This compliance shall include the appropriate classification and/or categorization of safety-related valves and the development of the proper test l procedures for pre-operational and periodic inservice valve testing. RESOLUTION
~'
Because of the importance of the interface between HP and LP ( safety-related systems, all pressure containing components used (_)\ in the System 80+ Standard Design identified as Safety Class 1, Amendment G A-10Ba April 30, 1990
CESSAR FRiif,em. t O 2, or 3 (including all HP to LP safety-related system boundary valves, e.g., SCS isolation valves) are designed, manufactured, and tested in accordance with the ASME B&PV Code, Section III ' (?,ee CESSAR-DC, Section 3.2.2. Also, see CESSAR-DC, Table 3.2-2, which provides a cross-reference between safety class and code e class.) I Furthermore, with respect to assuring the integrity of boundary valves during operation, these valves will be periodically , inservice tested in accordance with Section XI of the ASME B&PV Code. Specifically, code Class 1, 2 and 3 valves will be categorized in accordance with the provisions of subarticle ' IWV-2100 of the ASME B&PV Code Section XI. Valves will be tested to the requirements of subsection IWV for each valve category. ; Finally, the valve test requirements are included in Section 3.9.6.2 and the frequency of valve performance testing is , included in CESSAR-DC, Chapter 16. , In summary, the design, manufacture, pre-operational testing, and in-service testing of the boundary var'es used in the interface between safety-related HP and LP systems is controlled in accordance with the ASME Code and, thus, satisfies the intut of SRP Section'3.9.6, Revision 2. Therefore, this issue is resolved for the System 80+ Standard Design. REIIBENCES G i
- 1. NURE3-0933, "A Status Report on Unresolved Safety Issues, U.S. Nuclear Regulatory Commissie n, April, 1989.
- 2. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Peports for Nuclear Power Plants -- LWR Edition",
U.S. Nuclear Regulatory Commission.
- 3. ASME Boiler & Pressure Vessel Code, Sections III and XI.
O Amendment G A-108b April 30, 1990
CESSAR Minco.. i O Specifically, parts 1, 2, & 3 listed in the ACCEPTANCE CRITERIA sevtion above are to be met. The ultimate responsibility for meeting NUREG-0737, Supplement 1 and Generic Letter GL 82-33, remains with the utility cwner-onerator. Combustion Engineering, however, assists the owner-operator in establishing these procedures and training the plant operators and staff by providing Emergency Frecedure Guidelines as contained in report CEN-152 (See Reference 4). Specifically, Section 1.3 of CEN-152 addresses the guidance and responses to NUREG-0737, including loss of instrumentation, ^ multiple and consequential failures, adequacy of core cooling, operator errors during long-term Jooling, and optimal recovery guidelines for other plant accidents. Combustion Engineering provides analyses and guidance (CEN-152) to assist the owner-operator in meeting the guidance of NUREG-0737, Supplement 1 and Generic Letter GL 82-33. Therefore, this-issue is resolved for the System 87+ Standard Design. BEJERENCES
- 1. NUREG-0933, "A Status Reptsrt on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989. , 1
- 2. NUREG-0737, " Clarification of TMI Action Plan Requirements,"
U.S. Nuclear Regulatory Commission, February 1983.
- 3. Generic Letter 82-33, " Supplement 1 to NUREG-0737 - Require-ments for Emergency Responso Capability". U.S. Nuclear Regulatory Commission, December 1982.
- 4. CEN-152, " Emergency Procedure Guidelines", Combustion Engineering, Inc.
i i O Amendment F j A-123 December 15, 1989 :
OESSAO Ennnen. , 5 O I.D.2: PLANT SAFETY PARAMETER QISPLAY CONSOLE ID_H Generic Safety Issue (GSI) I.D.2 in NUREG-0933 (Reference 1), identifies the need for the provision of a safety parameter display system (SPDS) that displays a minimum set of parameters which define the safety status of the plant. Provision of a plant safety parameter display console was made a requirement by 10 CFR 50. 3 4 (f) , (Reference 2). The functional criteria and design requirements for the SPDS were defined in NUREG-0737 (Supplement 1) (Reference 3). The primary function of the SPDS and the display console is to help operating personnel in the control room make quick assessments of plant safety status , during abnormal and emergency conditions. ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI I.D.2 are that the SPDS shall meet the intent of the applicable requirements of 10 CFR 50. 34 (f) and NUREG-0737 (Supplement 1); and shall be human factors engineered consistent with the guidance of SRP Sections 18.0 and 18.2 (Reference 4) and the supplemental guidance of NUREG-0737 (Supplement 1). NUREG 0700 (Reference 5) provides guidelines for verifying the application of acceptable- human g factors principles and criteria. The SPDS functions shall be integrated into the overall control room design. The design of the safety parameter display shall incorporate accepted human factors principles so that the information can readily be perceived and comprehended by the users. As a minimum, information sufficient for determining the safety status of the plant and assessing whether abnormal conditions warrant corrective action to avoid core damage shall be provided. In addition to the control room displays, SPDS information should be provided to the Technical support Center (TSC) and the Emergency Operations Facility (EOF) to aid in exchange of information between these facilities and the control room, and to assist corporate and plant management in making decisions during emergency conditions. FJSOLUTION The System 80+ Standard Design employs the Nuplex 80+ Advanced Control Complex (ACC) in which the functions of the SPDS have been integrated into the control room design consistent with the guidance in SRP Section 18.0. The information is provided on separate display formats that specifically address plant safety status. Amendment G l A-123a April 30, 1990 1
CESSAR !!n%=,, O Human factors engineering principles such as those presented in NUREG-0700 have been applied during all stages of the design process to ensure a design that provides information structure, case of use, and reduced human error. Accomplishment of the SPDS functions meets the intent of the requirements of NUREG-0737 (Supplement 1) and the guidance of SRP Section 18.2.
- SPDS information is provided in all levels of the overall Nuplex 80+ information hierarchy as described in CESSAR-DC Section 18.7.1. The highest icvel of the information hierarchy is a large overview status board, known as the Integrated Process Status Overview (IPSO), that_ allows for a quick assessment of overall plant process performance and helps guide the operator to more detailed information. IPSO (see CESSAR-DC Section 18.7.1.2) is a continuous display format centrally located above the control room panels and contains information sized to be readable from anywhere within the control room.
- critical function matrix, similar to that provided on the Critical Function Monitoring System for San Onofre Generating Station, Units 2 & 3, alerts the operator to problems impacting twelve safety functions. These problems include:
- Safety related parameters outside of anticipated range (s) G
- Unavailability of a safety system that can be used to support the critical functions
- Poor performance of a safety system that is being used to support the critical functions.
The operator obtains the specific cause of the critical function violation by responding to the corresponding critical function alarm tile, located on the safety monitoring panel, or by accessing CRT display page formats within the Critical Function section of the display page hierarchy, see CESSAR-DC Sections 18.7.1.5 and 18.7.1.3, respectively. The Critical Function section of the display page hierarchy is broken down by critical function and the success paths that can be used to support the critical functions. The display pages have been designed to support the Emergency Procedure Guidelines provided by Combustion Engineering in Report CEN-152 (Reference 6) to the owner-operator. Success path information relating to the performance and availability of the plant safety system is presented on these display pages. CESSAR-DC Section 18.7.1.8 provides further details of the safety related information in Nuplex 80+, O Amendment G A-123b April 30, 1990
k y CESSA0 !!!!%ui. ; O' The Nuplex 80+ CRTs are driven by the Data Processing System (DPS) described in CESSAR-DC Section 7.7.1.7. The DPS is a computer-based system that provides plant data and status information to the operator, derived or procesced from plant sensors including the Post Accident Monitoring Instrumentation sensors. The information is available on both a real-Lime and > historical basis. SPDS and other information necessary for the handling of emergency plant conditions and assessment of their consequences.is also provided by the DPS to the TSC and EOF when they are activated and manned, as described in CESSAR-DC Sections
- 1. 3.3.1 and 13.3.3.2 respectively. Key. parameter values processed by the DPS are also indicateo directly via the Discrete Indicition and Alarm System (CESSAR-DC Section 7.7.1.4) on discrete indicators located on the main control room and remote shutdow.) panels. j Since the Nuplex 80+ ACC integrates the SPDS functions into the control room design and meets the intent of the various acceptance criteria, this safety issue is. resolved for the System 80+ Standard Design.
REFERENCES
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues", G U.S. Nuclear Regulatory Commission, April 1989.
- 2. 10 CFR 50, Code of Federal Regulations, Office of the Federal Register, National Archives and Records Administration.
- 3. NUREG-0737 Supplement 1, " Emergency Response Capability Requirements", U.S. Nuclear Regulatory Commission, December 1982.
- 4. NUREG-0800 " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - LWR Edition",
U.S. Noclear Regulatory Commission.
- 5. NUREG-0700, " Guidelines for Control Room Design Reviews",
U.S. Nuclear Regulatory Commission, September 1981.
- 6. Combustion Engineering Emergency Procedure Guidelines, C-E Document CEN 152 Rev. 3, May 1987.
O Amendment G A-123c April 30, 1990
m L CESSAR !!Mienion O I.D.4 CONTROL ROOM DESIGN STANDARD ISSUE Generic Safety Issue- (GSI) I.D.4 in NUREG-0933 (Reference 1), addresses the neea for guidance on the design of control rooms to incorporate human factor considerations, and the desirability of i endorsing an industry standard for future control room designs. Under GSI I.D.1, the NRC issued NUREG-0700 (Reference 2) for the guidance of the detailed control room design reviews required of operating plant control rooms for conformance to accepted human factors principles. SRP Section 18.1 (Reference 3) was issued to document the NRC review process for both existing and advanced control room designs. Appendix B of NUREG-700 contains guidance of the design of new control rooms. Additional human factors engineering guidance on various aspects of control room design has been developed in NUREG/CR reports 3217, 3987, 4221 (References 4 to 6, respectively) and various industry publications. In assessing whether to develop a Regulatory Guide to endorse an industry standard or standards for the human factors engineering G O of advanced control rooms, the- NRC concluded that the combined guidance of SRP Section 18.1, the NUREG reports noted above, and acceptable industry publications is adequate, and that development.of a new Regulatory Guide is not needed. . ACCEPTANCE CRITLx/A The acceptance criterion for the resolution of GSI I.D.4 is that the advanced control room design meet the applicable requirements of 10 CFR 50 Appendix A, GDC 19 (Reference 7) as it relates to the control room being designed with appropriate human factors l engineering (HFE) principles to assure that the operator-machine interfaces are adequate to support safe operation of the plant. l This shall be accomplished by conforming to the guidance of SRP Section 18.1 and NUREG-0700 Appendix B, supplemented as applicable by the guidance in References 4 to 6 and 8 to 13. The control room should utilize CRT displays and other advanced display technologies. It should be designed only after a full analysis has been made of the control tasks to be performed, and should provide means for data gathering and processing which support operator tasks and decision making. Human factors i princples and criteria should be applied to work spa ca. , work I environment, annunciator warning systems, panel layout and control-display integration. Amendment G A-123d April 30, 1990
h GESSAR E%uio. O.
,tESOLUTION 'Ihe System 80+ Standard Design employs the Nuplex 80+ Advanced Control Complex ( ACC) . HFE methods, principles and criteria such as those presented in NUREG-f100, Appendix B, and the other references as appropriate, P ave been applied throughout the design of the ACC, as documen+ ed in CESSAR-DC, Chapter 18.
The process used in the des.gn M the ACC man-machine interface and the applications of HFF are s.4mmarized in CESSAR-DC Section 18.4. A dedicated, multi .fisciplit ary team was established ~ for the design of the man-m4 chine in;erface, consisting of human
' factors specialists, sy stems eng.neers, instrumentation and controls engineers, and senior reactor operators. A separate Nuplex 80+ design revirw team, including Duke Power Company personnel to provide the perspective of a nuclear plant utility end constructor, pertorms regular reviews of the ACC design as it is developed. These teams are described in CESSAR-DC Secticn 18.2. Consistent with SRP Section 18.0, a detailed functional task analysis was performed to provide a basis for the ACC design. As part of the functional task analysis, information and control characteristics requirements were developed and man-machine functions allocated. The analysis detailed the '
operator's tasks involved in decision crocessing to ensure that: (a) only needed information is presented to the operator, (b) the G amount of information does not exceed human cognitive limitations, and (c) information is presented in usable form. The functional task analysis and its results are described in CESSAR-DC Section 18.5. A control room configuration was established (see CESSAR-DC Section 18.6) based on accepted HFE principles and on analysis of staffing requirements. Environmental and commreications criteria for the ACC were also developed at this time. Standard panel design criteria and algorithms for alarm and parameter validation processing were then developed. Information presentation methods and panel layouts in the Nuplex 80+ ACC are detailed in CESSAR-DC Section 18.7. The main control panels are designed as compact workstations. Each workstation integrates miniaturized component control switches, discrete indicators, alarm tiles, message windows, and video display units l such that both safety and non-safety display devices are j routinely used by the operator. 1 formation is presented i hierarchically on four levels: (a) high level plant overview, (b) l- for general system monitoring, (c) for component / system control, and, (d) for problem diagnosis. Two independent and diverse (yet integrated) computer based systems are utilized to process data and drive the display devices. Amendment G A-123e April 30, 1990
CESSAR !!nincam. The final steps in the HFE design of the Nuplex 80+ ACC are verification and validation (see CESSAR-DC Section 18.9) consistent with the guidance of NUREG-0700. Verifice. tion consists of a detailed evaluation of the control room design and prototypes of the equipment based upon the functional task , analysis to ensure that they are suitably designed for their intended use. Validation uses a ' partially dyna'nic mockup of a Nuplex 80+ panel to evaluate the integration of display and control components, followed by a demonstration of one-man normal operation, using a full scale partially dynamic mockup of the main control panel. Since the human factors engineering of the Nuplex 80+ ACC complies with the acceptance criteria as described above, this issue is resolved for the System 80+ Standard Design. BEJIEENCES
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989.
,q 2. NUREG-0700, " Guidelines for Control Room Desigr 'aviews",
U.S. Nuclear Regulatory Commission, September.1981. Q- g
- 3. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Rcports for Nuclear Power Plants - LWR Edition",
U.S. Nuclear Regulatory Commission. ,
- 4. NUREG/CR-3217, "Near-Term Improvements for Nuclear Power Plant Control Room Annunciator Systems", U.S. Nuclear Re1dlatory Commission, April 1983.
- 5. NUREG/CR-3987, " Computerized Alarm Systems", U.S. Nuclear Regulatory Commission, June 1985.
- 6. NUREG/CR-4221, " Human Engineering Guidelines for the Evaluation and Assessment of Video Display Units", U.S.
Nuclear Regulatory Commission, July 1985. 7, 10 CFR 50 Appendix A " General Design Criteria", Office of the Federal Register, National Archives and Records Administration.
- 8. EPRI-3659, " Human Factors Guide for Nuclear Power Plant Control Room Development", Electric Power Research Institute, August 1984.
? (~'\ k.. Amendment G A-123f April 30, 1990 L 1'
CESSAR !!!iirico. I I
- 9. EPRI-3448, "A Procedure for Reviewing and Improving Power l Plant Alarm Systems", Electric Power Research Institute, April 1984.
- 10. EPRI-3701, " Computer Generated Display System Guidelines", !
Electric Power Research Institute, September 1984.
- 11. IEEE Standard 1023-1988, "IEEE Guide for the Application of Human Factors Engineering to . Systems, Equipment and '
Facilities of Nuclear Power Generating Stations", October 1988. G
- 12. MIL-STD-1472C, " Human Engineering Design Criteria for Military Systems, Equipment and Facilities", December 1974.
- 13. VanCott and Kincade, " Human Engineering Design for Equipment Design", 1972.
O l l l .- l-O Amendment G A-123g April 30, 1990
CESSARin#,cuin
,rm k ,'
I.D.5(1): OPERATOR-PROCESS COMMUNICATION ISSUE Generic Safety Issue (GSI) I.D.R(1) in NUREG-0933 (Reference 1), addresses the man-machine interface in the control room with ' reference to the use of lights, alarms, and annunciators. Operator error may result from information overload, unnecessary distractions, and lack of information organization. ,
]LQQJP_TANCE CRITERIA The acceptance criteria for the resolution of GSI I.D.5(1) are included in GSI HF1.3.4b.
RESOLUTION The resolution for GSI I.D.5(1) is included in the resolution to GSI HF1.3.4b. Since GSI I.D.5(1) is subsumed by the above GSI, this issue is resolved for the System 80+ Standard Design. g REFERENCEE
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory Commission, April 1989. , P Amendment-G A-123h April 30, 1990
CESSAR EHWico... O I.D.6 (2): IMPROVED _CnNTROL ROOM INSTRUMENTATION - PLANT STATUS AND POST-ACCIDENT MONITORING IMP 3 Generic Safety Issue (GSI) I.D.5 (2) in NUREG-0933 (Reference 1), addresses the need to improve the operators' ability to prevent, diagnose and properly respond to accidents. This issue was originally identified in the TMI A'ction Plan (Reference 2) and resulted in the establishment of new NRC requirements. Guidance for addressing the issue is provided in Regulatcry Guide 1.47 (Reference 3) which provides an acceptable method for implementing the requirements of IEEE 279-1971 and i 10 CFR 50, Appendix B (Criterion XIV) with respect to the bypass or inoperable status of safety systems, and Regulatory Guide 1.97 (Reference 4) which defines an acceptable method for implementing NRC requirements to provide instrumentation and to monitor plant variables and systems during and following an accident. ACCEPTANCE CRITERIA The acceptance criteria for the resolution of GSI I.D.5 (2) are 4 contained in:
- 1. Regulatory Guide 1.47 for emergency safeguards features (ESP) status monitoring. Automatic bypassed or inoperable status indication at the system level is recommended for the plant protection system, safety systems actuated or controlled by the protection system and their auxiliary and supporting systems. These indications should be provided in the control room and should have manual input capability.
- 2. Regulatory Guide 1.97 for post-accident monitoring instrumentation. This Regulatory Guide identifibs criteria for design and qualification of the instrumentation divided into three categories, designated 1, 2 and 3, which provide a graded approach to requirements based on the importance to safety of the variable being monitored. Criteria exist for equipment qualification, redundancy, power sources, channel availability, quality assurance, display and recording, range, equipment identification, interfaces, servic;.g, testing and calibration, human factors and di: act measurement. The actual variables to be monitored for a pressurized water reactor are tabulated in the guide by type and the instrumentation design and qualification i
requirement category (1, 2 or 3) is identified for each variable. Amendment F A-124 December 15, 1989
h 4ESSAO !!!Mem.
-\
II.E.4.2i CONTAINMENT DES IGN--ISOLATION DEPENDABILITY ISSUE - Generic Safety Issue (GSI) II.E.4.1, in NUREG-0933 (Reference 1), addresses the TMI requirements identified in NUREG-0737 (Reference 2), which references the guidance given in SRP 6.2.4, Rev. 2 (Reference 3), regarding the need to maintain containment building isolation integrity throughout the plant design lifetime ' for both normal (especially refueling and extended maintenance activities) and accident conditions. The TMI accident demonstrated the continued need for dependable containment isolation after an accident as well as during startup, shutdown, and normal operation (when systems such as the containment purge and vent systems are used). There are three areas of concern for isolation of containment building penetrations, which are: (1) fluid systems piping, (2) instrumentation and controls sensing lines, and (3) containment building purge and vent system ventilation " ducts". The majority of containment penetrations are from fluid systems . piping. However, there are typically a small number of instrumentation
~s and control sensing lines and several vent and purge ducts which f i also penetrate the containment building. ;
V ACCEPTANCE CRITERIA g The acceptance criterion for the resolution of GSI II.E.4.2 is that the plant design shall conform to the requirements of NUREG-0737 for the containment isolation system and meet the intent of the overall guidance identified in SRP Section 6.2.4, Rev. 2. This guidance establishes criteria for the containment isolation system which encompasses, the fluid systems piping and instrumentation and controls sensing lines and the containment purge and vent systems. SRP 6.2.4 references' specific guidance identified in Regulatory Guide 1.141 (Reference 4) for fluid system piping, Regulatory Guide 1.11 (Reference 5) for instrument sensing lines, and BTP CSB 6-4 (Reference 2) for containment purge and vent system ducts, all of which taken together satisfy GDCs 1, 2, 4, 16, and 54 through 57 describnd in 10 CFR 50 Appendix A (Reference 6). Regulatory Guidc 1.141 further endorses the industry requirements established in ANSI N271-1976 (Reference 7). Specifically, implementation of the SRP and other references shall consist of the following: (1) meeting the requirements of GDCs 1, 2, and 4 for the design, CN fabrication and testing of the containment isolation system
-() with respect to quality assurance, qualification, ceismic and dynamic effects; environmental Amendment G A-150a April 30, 1990
GESSAR nninmi. t (2) F4eeting the requirements of GDC 16 with respect to l maintaining an essantially leak-tight containment boundsry; (3) meeting the requirements of ANSI N271-1976 which satisfies the guidance icentified in Regulatory Guide 1.141 and the requirements of GDCs 54 through S7 for the capability of isolating fluid systems which penetrate the containment boundary; (4) meeting the intent of the guidance in Regulatory Guide 1.11 i for.the isolation of instrumentation and controls sensing lines which penetrate the containment building. However, instrument sensing lines which provide input to pressure transmitters that monitor containment building pressure are specifichlly exempt from this guidance; and (5) the containment purge and vent systems shall be designed to meet the requirements of 10 CFR 50 Appendix A, GDCs 54 and , 56 identified in BTP CSB 6-4 with respect to maintaining containment integrity during a loss-of-coolant accident (LOCA). RES0WTJ_QE The System 80+ Standard Design incorporates a Containment Isolation System for fluid systems piping and for the Containment 9t - g Purge Ventilation System. Instrumentation and control sensing.
-lines which penetrate the containment building are provided with containment isolation provisions which meet the intent of Regulatory Guide 1.11, except for lines such as the four containment building pressure instrument sensing lines, which are exempt (see CESSAR-DC Section 6.2.4.1.1). ,
The containment isolation system is designed to prevent or limit the release of radioactivity to the environment during and af ter an accident while ensuring continued operability of safety-related systems which might be needed to limit or prevent - the consequences of an accident. This " system" is, in fact, not a single system but is composed of a variety of containment penetrations whose isolation valve arrangement is uniformly designed, fabricated, and tested according to the criteria , specified above. A more detailed description for the fluid systems piping containment isolation system is presented in CESSAR-DC, Section 6.2.4. CESSAR-DC, Section 9.4.5, describes that part of the containment isolation system which addresses the i containment purge ventilation system. The. System 80+ Standard Design meets the acceptance criteria in i _the following ways: l Amendment G A-150b April 30, 1990 l-
i i CESSAD!!n%.m I s
]
(1) The ^ Nainment isolation system for the fluid systems pipin 'd containment purge ventilation system ducting meets me intent of the overall guidance described in SRP 6.2.4, Rev. 2 and the supplemental guidance identified in the BTP CSB 6-4, including the requirements in GDCs 1, 2, and 4. See CESSAR-DC, Sections 3.1.1, .3.1.2, and 3.1.4, respectively. (2) With regard to GDC 16, which addresses maintaining the leak-tightness of the containment building, the containment building is designed to protect the public from the consequences of an accident (i.e., minimize the release of radioactivity) and to safely withstand all internal and external environmental conditions that may be reasonably expected to occur during the . plant's lifetime (see CESSAR-DC, Sections 3.1.12 and 6.2.4 respectively). (3) The containment isolation system for the System 80+ Standard Design conforms to the requirements of ANSI N271-1976 and thus meets the intent of Regulatory Guide 1.141 and the requirements identified in GDCs 54 through 57, for the isolation of fluid systems. The system's desi.jn ht:4s O. addresses such requirements as leak detection, isolation, J (j ~ and leakage containment capabilities. It also establishes such design features as redundant and reliable isolation G valves, and defines the system's performance requirements (see CESSAR-DC, Section 6.2.4.2). t (4) In accordance with the requirement of NUREG-0737, the design ! of instrumentation and control systems for the automatic containment isolation valves is such that resetting the isolation signal does not result in the automatic reopening of the valves. Reopening of containment isolation valves requires deliberate operator action to open valves on an individual containment penetration basis (see CESSAR-DC Section 6.2.4.5). (5) In addition to fluid systems piping, the containment isolation system also includes the Containment Purge Ventilation System which must be isolated during a LOCA. This system is designed to provide a means of purging and venting the containment building whenever the containment is or will be occupied by plant personnel such as for plant refueling and extended maintenance activities. The system is, therefore, designed to meet the intent of BTP CSB 6-4 (which references GDCs 54 and 56) with respect to l maintaining containment integrity during and after a LOCA
,- (see CESSAR-DC, Section 9.4.5).
Yl l Amendment G A-150c April 30, 1990
i
@ESSAO !!!nnem.. I 1
l In summary, the- containment isolation system is designed to conform. to the requirements of NUREG-0737 and meet the overall guidance identified in SRP 6.2.4, Rev. 2. These documents , encompass the requirements of 10 CPR 50 Appendix A, the guidance l given within BTP CSB 6-4, Regulatory Guides 1.141 and 1.11, and the requirements of industry standard ANSI N271-1976. Therefore, all of the requirements and guidance with respect to maintaining containment building isolation integrity throughout the ' plant design life have been met and this issue is resolved for the System 80+ Standard Design. EHERREH
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory' Commission, April, 1989.
- 2. NUREG-0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission, June, 1985.
- 3. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -- LWR Edition",
Nuclear Regulatory Commission.
- 4. Regulatory Guide 1.141, " Containment Isolation Provisions for . Fluid Systems", Nuclear Regulatory Commission, April 1978.
- 5. Regulatory Guide 1.11, " Instrument Lines Penetrating Primary G Reactor Containment", Nuclear Regulatory Commission, 1971.
- 6. 10 CFR 50 Appendix A, " General Design Criteria for Nuclear Power Plants", Code Of Federal Regulations, Office of the Federal Register, National Archives and Records Administration.
- 7. ANSI N271-1976, " Containment Isolation Provisions for Fluid Systems", American National Standards Institute, 1976.
I O Amendment G A-150d April 30, 1990
CESSAR Eniincuio. I ) J II.E.4.48 CONTAINMENT RESIGN -- PURGI}M ISSUR Generic Safety Issue (GSI) II.E.4.4 in NUREG-0933 (Reference
- 1) addresses the acceptability of the arrangements for purging / venting the reactor containment building.
Parts (1), (2), and (3) of thin issue required licensees of operating plants to: (1) minimize purging of the containment building cluring power operation and to justify additional purging; (2) provide the NRC vith information on the operability of ' the . containment isolation valves; (3) ensure operability of the containment purge and vent valves. Parts (4) and (5) of the issue were consolidated in NUREG-0933 and addressed the radiological consequences of containment purging / venting during the power operation mode. The results of the NRC studies were then to be factored into new plant designs. Nuclear power plant containment building vent / purge' systems are designed to provide a method of introducing
" conditioned" frerb air into the containment building in -
o preparation for and during plant shutdowns and to control ( ) the release of contamination to the environment. These V systems must also close upon occurrence of a containment isolation signal from the Engineered Safety' Features System. 6 The safety concern with respect to the containment building purge and vent system is that if a loss-of-coolant accident (LOCA) were to occur while the building is being purged or vented, radiation releases to the environment would likely ( result. This concern is addressed by assuring that the containment isolation valves meet the closure requirements, including time to close, thus assuring that any radiological release would be small. Guidance on these requirements is given in SRP Section 6.2.4 (Reference 2). ACCEPTANCE CRITERIA The acceptance criterion for the resolution of GSI II.E.4.4 Parts (1) through (5) is that the design of the containment vent and purge system shall be such that the potentie.1 for the loss of its safety function be minimized. Specifically, the containment vent / purge- system shall be designed to meet the intent of the guidance given in SRP Section 6.2.4. t ;Q Q ,) Amendment G A-150e April 30, 1990
CESSAR !!nincuio. O RESOLUTION The Containment Purge Ventilation System in the System 80+ Standard Design is designed to provide clean, fresh air whenever the containment and/or incore instrumentation room is or will be occupied. Containment air is exhausted to the environment through the purge filter trains. The system is described in CESSAR-DC Section 9.4.5, and consists of two sub-systems: high volume purge and low volume purge. The containment High Volume Purge sub-system is designed to maintain the average containment air temperature between 60 and 90 degrees F during inspection, testing, maintenance, and refueling operations and to limit the release of any contamination to the environment. This sub-system will not be used during power operation. The containment Low Volume Purge sub-system is designed to provide air circulation and reduce airborne radioactivity for access during normal operation or after reactor shutdown. This sub-system will be used only on an as-needed basis during power operation. G Each containment penetration for the two sub-systems is provided with two isolation valves, one on each side of the containment pressure ' boundary. The containment purge isolation i val w meintain primary containment integrity during a postulated
- LOCA and acet the intent of the guidance given in SRP Section
! 6.2.4. l l The containment Low and High Volume Purge sub-systems for .the System 80+ Standard Design are designed to be periodically inspected, tested and maintained (see CESSAR-DC, Section 9.4.5.4). Furthermore, in order to assure system operability during normal and accident conditions, limiting conditions for operation (LCO's) are specified (see CESSAR-DC, Section 16). The usage of these systems during power operation will also be minimized to reduce the probability of radiation releases to the public environment. l l The required closure time for the Low Volume Purge isolation
- valves is justified (see CESSAR-DC, Section 6.2.4) by an analysis I of
l
- a. the radiological consequences of a LOCA, assuming appropriate source terms for the loss of inventory, and O
Amendment G A-150f April 30, 1990
CESSAR !!!Nico. ! l f ] (
- b. the allowable loss of containment atmosphere while the isolation valves are closing, which affects the value for minimum containment pressure used in evaluating the emergency. core cooling system effectiveness.
Since the intent of the acceptance criterion'is met as described above, this issue is resolved for the System 80' Standard Design. REFERENCES
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues", U.S. Nuclear Regulatory Commission, April g 1989.
- 2. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - -
LWR Edition", U.S. Nuclear Regulatory Commission. O I Amendment G A-150g April 30, 1990
4 r CESSAR ?!!G?,co... _ i O THIS PAGE INTENTIONALLY DIANK . O-l l. i f. P Ol Amendment G A-150h April 30, 1990
GESSAD WMnemon f} , Q) H.K.1: MEASURES TO MITIGATE SMALL-BREAE # LOSB-OF-COOLANT ACCIDENTS AND LOSS-OF-FEEDWATER ACCIDENTS ISSUB Generic Safety Issue (GSI) II.K.1 in NUREG-0933 (Reference 1), has twenty-eight (28) sub-issues which address the need to improve emergency operating procedures, operator training, and hardware to mitigate the consequences of the small-break LOCAs and loss-of-feedwater events based on the Bulletin and Orders Task Force review of the I&E Bulletins. These sub-issues are also directly relatable to other USIs and GSIs as cited in NUREG-0933. Guidance for design and operating improvements for these sub-issues is given in NUREG-0737 (Reference 2). For example, the NRC requires (a) operating procedures which can recognize, prevent and mitigate the formation of voids in the reactor vessel head during transients; (b) a review of safety-related emergency feedwater valve positioning after maintenance, inspection, and G
,q operation; and (c) a design and procedures which will assure automatic tripping of the reactor coolant pumps for all required i
V) circumstances. ACCEPTANCE CRITERIh < Table 1.0, provided in the RESOLUTION section, lists only the sub-issues and the corresponding GSI/USIs which are applicable to the System 80+ Standard Design. The other sub-issues of GSI II.K.1 are not applicable. The acceptance criteria for the resolution of these sub-issues identified in GSI II.K.1 are encompassed by the other Generic and Unresolved Safety Issues which are given in NUREG-0933. In general, these criteria provide that plant design and operation adequately address both small-break LOCAs and loss-of-feedwater events in accordance with the guidance given in NUREG-0737. RESOLUTION L L GSI II.K.1 is a comprehensive issue covering a broad range of l safety aspects of both plant design and emergency procedures. L Each sub-issue which is applicable to the System 80+ Standard l Design is identified in the following Table and cross-referenced l -(] to the particular USI and/or GSI which specifically addresses the (/ sub-issue. Amendment G A-159 April 30, 1990
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ICIE!BIEJLit !!!Mnc.Ic. i O Resolutions for the applicable sub-issues of GSI II.K.1 are subsumed by the individual USIs and/or GSIs, each of which is separately resolved and included in C-E's USI and GSI submittals. Therefore, Issue II.K.1 is resolved for.the System 80+ Standard ' Design. ThBLE 1.0 TABLE OF GSI II.K.1 SUB-ISSUES j AND OTHER GSI/USIs WHICH ARE APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN GSI II.K.1 GSI/UBI CROSS-REFERENCES SUB-ISSUE APPLICABLE TO SYSTEM 80+ 3 I.C.1 4d II.F.2 6 II.E.4.2 9 II.E.4.2 14 II.E.4.1, II.F.1 15 II.E.1.2 16 I.C.1, II.D.3 24, 25- I.C.1 G 26 I.C.1 27 I.C.1, II.F.2 28 II . K. 3 (5) REFERENCRg
- 1. NUREG-0933, "A Status Report on Unresolved Safety Issues",
U.S. Nuclear Regulatory commission, January 1989.
- 2. NUREG-0737, " Clarification of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission, October 1980. Amendment G A-160 April 30, 1990
l C E S S A R in #,cui. 1p
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II.K.3 FINAL RECOMMENDATIONS OF ' BULLETINS AND ORDERS (B&O) TASK FORCE 1&&llK Generic Safety Issue (GSI) II.K.3 in NUREG-0933 (Reference 1), has fifty-seven (57) sub-issues which address the need to improve emergency operating procedures, operator training, and hardware particularly with respect to accident prevention, monitoring, and mitigation based upon . the recommendations of.the Bulletins and ' Orders Task Force. These sub-issue are also directly relatable to other USIs and GSIs as cited in NUREG-0933. In addition, guidance for improved emergency operating procedures, operator training and hardware with respect to accident prevention monitoring and mitigation is given in NUREG-0737 (Reference 2). G ACCEPTANCE CRITERIA Seven sub-issues apply to the System 80+ Standard Design of which 6, 8, and 55 are each covered separately under their own USI g The fou' remaining sub-issues are 5, 25, and/or GSI resolution. V) 30, and 31. The acceptance criteria for the resolution of sub-issues 6, 8,
' and 55, are encompassed in other Generic and Unresolved Safety
./. Issues as shown in Table 1.0 and are given in NUREG-0933. The acceptance criteria for resolution of sub-issues 5, 25, 30 and 31 are given in NUREG-0737 and are summarized below. Sub-Issue (5): Automatic Trin of Reactor Coolant Pumns.(RCPs) on Smidl Break LOCAs Licensee should . consider the effects of automatically trfpping the RCPs upon the occurrence of a small break LOCA to assist in accident mitigation. Sub-Issue (25); Ef%gst of ' Loss of Alternatina-Current (AC) Pouqr on Pumo Seals I: Licensee should determine the consequences of a j loss of cooling water to the RCP seal coolers i fc11owing the loss of offsite AC power. The
' seals should be designed to withstand a complete
=i loss of AC power for two (2) hours.
. ,m i.
Amendment G - ( A-161 April 30, 1990
G CESSAOEmincm. a O Sub-issue (30): Revised Small-Break IOCA Methods to Show Comoliance with 10 CFR 50. Accendix K In accordance with the revised guidance given in this sub-issue in NUREG-0737, the Licensee is allowed to demonstrate compliance by the justification of the acceptability 'ts current small-break LOCA model. . Sub-issue (31): Plant-Soecific Calculations to She f 3 . ;;; > 7 with 10 QFR 50.46 , Licensee shall submit plan s A tt . calculations to show compliance by un on NRC-approved model for small-break LOCAs. RESOLUTION GSI II.K.3 is a comprehensive issue covering a broad range of safety aspects of both plant design and emergency procedures. Three of the sub-issues which are applicable to the System 80+ Standard Design, 6, 8, and 55, are covered separately by other USIs and GSIs, and each sub-issue is cross-referenced in Table 1.0 to the particular USI and/or GSI which specifically addresses g that sub-issue. With respect to sub-issues 5, 25, 30 and 31, the System 80 and the System 80+ Standard Designs are essentially the same. , Therefore the resolutions, which have been approved by the NRC in NUREG-0852'(Reference 3) for System 80, apply also to System 80+- and'are given below for these sub-issues. Sub-issue (5): Automatic Trio of RCPs on Small-Break LOCAs The effects of automatic tripping of the RCPs on Small-Break LOCAs were reported by Combustion Engineering, Inc. (C-E) in CEN-268 (Reference 4), which identifies the RCP trip methodology. Sub-issue (25): Effect of Loss of AC Power on Pumo Seals The seals are cooled by redundant systems, i.e., seal injection water (SIW) and component cooling ' water (CCW). In the event of a loss of AC power, the SIW can be restored by manually furnishing essential power to the charging pumps. In addition, a series of tests which were performed on a System 80 RCP demonstrated the ability of the RCP seals to= withstand a loss Amendment G A-162 April 30, 1990
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of AC power. The first test involved a s.'3ultaneous loss of CCW and SIW for 35 minutes to a RCP. The second test was a loss of CCW with SIW I available for two hours before CCW was restored. The results of the above tests demonstrated that I the RCP seal temperatures do not exceed the pump ! manufacturer's maximum allowable operating temperatures under a postulated loss of AC power ' condition. Sub-issue (30): Revised Small-Break LOCA Methods To Show Compliance with 10 CFR 50. Anoendix K C-E developed a topical report, CEN-203, (Reference 5) that demonstrated the continued ' acceptability of C-E's approved small break LOCA evaluation model. Sub-issue (31): Plant-Snecific Calculations to Show Compliance With 10 CFR 50.46 ( CEN-203 demonstrates- continued acceptability of
'- C-E's approved small-break LOCA model and G l CESSAR-DC, Section 6.3.3 demonstrates compliance l with 10 CFR 50.46.
j Resolutions for the applicable sub-issues of GSI II.K.3 except 1 for the four resolved above, are subsumed by the individual USIs '! and/or GSIs, each of which is separately resolved and included in ! C-E's USI and GSI submittals. Since the above four sub-issues- j are resolved and the remaining sub-issues which have been i identified as applicable to System 80+ are resolved separately by other USIs and/or GSIs as shown Table 1.0, Issue GSI II.K.3 is, {' therefore resolved for the System 80+ Standard Design. l TABLE 1.0 TABLE OF REMAINING II.K.3 SUB-ISSUES AND OTHER GSI/USIs WHICH ARE APPLICABLE TO THE SYSTEM 80+ STANDARD DESIGN GSI II.K.3 SUB-ISSUE GSI/USI CROSS-REFERENCES l 6 II.F.2, II.F.3 l 8 A-45 I
#')
Lj 55 I.D.2 f l Amendment G A-163 April 30, 1990 l-L l
CESSAR !!Gicm:. O ARKE'EH
- 1. NUREG-0933, "A Status Report . on Unresolved Safety Issues", ;
i U.S. Nuclear Regulatory Commission, January 1989. ]
- 2. NUREG-0737, " Clarification Of TMI Action Plan Requirements",
U.S. Nuclear Regulatory Commission, October 1980.
- 3. NUREG-0852, " Safety Evaluttion Report related to the final design of the Otandard Nuclear Steam Supply neference System CESSAR Syr, tem E0", U.S. Nuclear Regulatory Commission, November 1981; and Supplements 1, 2, and 3 dated March 1983, September 1983 and December 1987, respectively.
- 4. CEN-268, Revision 1, " Justification of Trip Two/ Leave Two Reactor Coolant Pump Trip Strategy", Combustion Engineering, Inc., May 1987.
- 5. CEN-203, Rev. 1 " Response To NRC Action Plan Item II.K.3.30 g
-- Justification of Small-Break LOCA Methods", Combustion Engineering, Inc., March 1982. Rev. 1 includes Supplements 1, 2, 3, and 4 dated February 1984, November 1984, December 1985 and November 1986, respectively.
O i l-O Amendment G A-164 April 30, 1990}}