ML19346A344: Difference between revisions
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==5.0 REFERENCES== | ==5.0 REFERENCES== | ||
[ | [ )) | ||
\ | \ | ||
N ~' 1/ T V Eichler, E S Napadensky, " Accidental Vapor Phase Explosions on Trans-poration Routes Near Nuclear Power Plants", NUREG/CR-0075, April 1977. . | N ~' 1/ T V Eichler, E S Napadensky, " Accidental Vapor Phase Explosions on Trans-poration Routes Near Nuclear Power Plants", NUREG/CR-0075, April 1977. . | ||
| Line 3,220: | Line 3,220: | ||
g 3.., | g 3.., | ||
PX n | PX n | ||
EV | EV ((h EV | ||
; L ; L ; L C I LATER [ | ; L ; L ; L C I LATER [ | ||
PX PX +- PX n n $,* n EV , , | PX PX +- PX n n $,* n EV , , | ||
| Line 5,626: | Line 5,626: | ||
ACNGS-PSAR- | ACNGS-PSAR- | ||
(( a) , Establishing the number of QA/QC personnel based upon the project | |||
- \'"'}/ schedule uto ensure' that personnel are available, qualified, snd certified te perfona quality-related inspections and evaluations. 1 b) Establishing the need for specially qualified QA/QC personnel based upon the. schedule for activities requiring special or unuoual expertise as far in advance of the activity as possible, t i | - \'"'}/ schedule uto ensure' that personnel are available, qualified, snd certified te perfona quality-related inspections and evaluations. 1 b) Establishing the need for specially qualified QA/QC personnel based upon the. schedule for activities requiring special or unuoual expertise as far in advance of the activity as possible, t i | ||
c) Establishing the number of QA personnel based upon the r umber and criticality of problems identified during routinc activities in order i to perform addit:lonal or supplemental inspections, reviews, or evaluations as r. squired to ensure implementation of project 57 | c) Establishing the number of QA personnel based upon the r umber and criticality of problems identified during routinc activities in order i to perform addit:lonal or supplemental inspections, reviews, or evaluations as r. squired to ensure implementation of project 57 | ||
Latest revision as of 19:21, 15 March 2020
| ML19346A344 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 06/18/1981 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML19346A338 | List: |
| References | |
| NUDOCS 8106190173 | |
| Download: ML19346A344 (475) | |
Text
{{#Wiki_filter:ACNGS-PSAR
- HOUSTON LIGHTING & POWER COMPANY ALLENS CREEK NUCLEAR GENERATING STATION - UNIT NO.1 PRELIMINARY SAFETY ANALYSIS REPORT AMENDMENT NO. 59 INSTRUCTION SHEET This amendment contains information pertaining to the PSAR Update. Each re-vised page bears the notation Am. No. 59, (6/81) at the bottom of the page.
Vertical bars with the number 59 representing Amendment No. 59 have been used in the margins of the revised pages to indicate the Incation of the revision on the page. The following page removals and insertions should be made to incorporate Amendment No. 59 into the PSAR. REMOVE INSERT (EXISTING PAGES) ( AMENDMENT NO. 59 PAGES) Chapter 1 Chapter 1 1* 1* 3* 3* 4* 4* F1.2-3 F1.2-3
,. , F1.2-33 F1.2-33
/ X.~ F1.2-35 F1.2-35 k F1.2-38a F1.2-38a F1.2-38b F1.2-38b Chapter 2 Chanter 2 3* 3* 3a* 3a* 26* 26* 2.2-5 2.2-5 2.2A-1 thru 10 2.2A-1 thru 10
- 2.2A-10a thru 10m 2.2A-11 2.2A-11 2.2A-12 2.2A-12 2.2A-13 2.2A-13 2.2A-14 2.2A-14 2.2A-15 2.2A-15 2.2A-16 2.2A-16 2.2A-17 2. 2A-17 2.2A-18 2.2A-18 thru 20 2.2A-19 -
2.2A-20 - 2.2A-23 2.2A-23
- 2.2A-24
- F1.2
/ ; \ ! IS-1 Am. No. 59, (6/81) 8106L90l'O. .
l 4 ACNGS-PSAR
# REMOVE
~
INSERT l f l ( AMENDMENT NO. 59 PAGES) , (EXISTING PAGES) i i ! Chapter 2 (Cont'd) Chapter 2 (Cont'd) { F1.2a - l !. F1.2b - F1.3 F1.3 - - F1.4 thru 1.15 ! F2.1 - F2.2 - F2,2-2 F2,2-2 j 1 I I l i l l f , i . I i 1 .' r r a i i t t l I A f i IS-2 Am. No. 59, (6/81) i i
I
'l F ACNGS-PSAR
-l>
.!- , A . REMOVE INSERT'
-/ - (EXISTING PAGES) (AMENDMENT NO. 59 PAGES) l
, 1 .! . Chapter. 3 Chapter 3 [
- - 1* - 1* f L 4* '4* !
5* 5* [
- 6* 6*
7* - 7* 8* 8* 8a*
- 9* 9*
13* 13*
, .14* 14*
! 15* 15* ! 16* 16* j 16a* - i 17* 17*
- 3.2 3.2-26 l j 3.5-11a 3.5-11a j i 3.5-12 3.5-12 l l 3.8-7a 3 8-7a [
i 3.8-9a 3.3-9a -l
- .3.8-9b 3.1-9b +
i 3.8-10 3 - 10 ! t I , Chapter 5' Chapter 5_ !
-t
.1* 1*
5* 5* f 6* 6* i Chapter 6 Chapter 6 i 1* 1* ( l ,
- - 3* 3* ;
6* 6* ['j 10a* l 10a* j i ! 6.2-32 6.2-32 l } 6.2-107g 6.2-107g ! ) - F6.2-26 (Sheet 27a of 31) i l F6.2-26 (Sheet 31) F6.2-26 (Sheet 31) j Chapter 7 Chapter 7 i- 1* 1* i la*- - l '2* 2_* i 3*_ 3* ! 4* 4* i 6* 6* 6 7* 7* l 8* 8*' i 9* 9* f- IS 3 Am; No. 59, (6/81)
t. f. it li
- 1 1
- . - ACNGS-PSAR l
.j
' ~
J j .- REMOVE INSERT i -' (EFISTING PAGES)- (' AMENDMENT No. 59 PACES) ( 1 , Chapter 7 (Cont' d) Chapter 7 (Cont'd) f- 1 10* 10* 11* N 11*- i t 11a* 11a*
; 12* 12*
13* 13* 7 l 14* l
- 14*'
t 7.3-30a 7.3-30a h 7.5-44g i 7.5-44g Chapter 8 Chapter 8
.. 1* 1* i ! 3* 3* l 1 t i
4 Chapter 9 Chapter 9 ;
,i . I 1* l
! 1* - } 2* . 2* 't 3*- 3* i ~4* 4*
.' 5* 5* )
+ 6* .6* ! j 7* 7* ;
- 8* 8* l l 8a* 9* i j- 8b* 10*
i' 9* 11* 10* ' 12 * { l 1' 11* 13* ! I - 14* h
- 15*
Chapter 10 Chapter 10 1* 1* Chapter 11 Chapter 11 1* 1* 1a* - 2* 2* 1 3* 3* 4* 4* I 5* 5* I 6* 6+ I- { 7* 7* IS-4 Am. No. 59, (6/81) f' 1 1
~
s: .
,~ ACNGS-PSAR-
^ REMOVE INSERT (EXISTING PAGES) . (AMENDMENT No. 59 PAGES) l i- Chapter 12 ! Chapter 10 1.- 1* 1* 3
- - 3* 3*
- 3a* +
,- 14
- 4*
l- '5* 5* Chapter 13 Chapter 13 1
- ll* 1*
l' 3* 3* ! 4* - -4* ' !' 5* i 5* , F13.1-2 F13.1-2 i -13.3-11 13.3-11 ,
- i. 13.3-12 13.3-12 13.3-17 13.3-17 l- 13.3-18 13.3-18 l '
i 13.3-19 13.3-19
'13.3-20 13.3-20 1 13.3-20a 13.3-20a 13.3-20b
- 13.3-20c
- '13.3-20d
- 13.3-20e
- 13.3-20f ;
l - 13.3-36 ' F13.3-1 F13.3-1
- - F13.3-2
- F13.3-3 [
- F13.3-4 ;
Chapter 14 Chapter 14 ;
'1* 1*
Chapter 15 Chapter 15 i ! '1* 1* !. ' 2* . 2* !. 3* 3* 1
'4*- 4*
5*. 5* 6*- 6* 9* 9* 15B-1 15B-1 15B-2 15B-2
@ 15B-3' 15B-4 IS-5 15B-3 15B-4 ,
Am. No. 59, (6/81) I
.a
. . . ~ - . . . ,..._. .., . -. - - . . . . ~ . - . . . . . . . . = . - . - _ . . _ . . _ ~ _ - . _ ~ . . . . . . . - . _ - . . I I i: l l ACNGS-PSAR I i *
~ REMOVE- . INSERT l (AMENDMENT NO. 59 PAGES) . j 1
,(QI". TING PAGES)
L. : ~ Chapter 15 (Cont'd) Chapter 15 (Cont'd) f 15F 15B-5 4 - 15B-6
- i. -. 15B-7
!: - - 15B-8 i t Chapter 17 Chapter 17-i
- Remove Chapter 17 Insert Chapter 17 in it's entirety in it's entirety j
}. Appendix C Appendix C f j 1* ~ 1* 2* 2* i 3*. 3* ! t- - 4* ! i 1 i i - C1.97-1 C1.97-1 > C1.97-2 C1.97-2 Appendix 0 Appendix 0 l9 1 1 Remove Appendix 0 in it's entirety Insert Appendix 0 in it's entirety-t i l t t l I i t I l' l l ! J l l IS-6 Am. No. 59, (6/81) l
3
. -E ;
.( ~.
l .. t. l l- ~ ACNGS-PSAR j i LIST 0F EFFECTIVE'PAGES 'I
~
'~e CHAPTER 1 INTRODUCTION AND' GENERAL DESCRIPTION-OF PLANT i Page ' Amendment 1* 59 2* 46 2a* 56 3* - 59 4* 59 , l i 37 l ii 33 I iii 7 ! iv 37 i y' 37 vi' 37
'vii 37 1.1-1 37
- 1.1 33 1.1-3 33 1.1-4 33 1.1-5 33 1.1-6 33
! 1.1-7 33 i 1.1-8 33' { 1.1-9 33 l l 1.1- 10 33 I 1.2-1 39 1.2-2 39 1.2-3 37 1.2-3a 37
-1.2-4 56 1.2-5 ' -
l 1.2-6 56 i 1.2 56 1.2-8 37
- 1. 2-9 26 1.2-10 37 1.2- 11 37 1.2-12 56 1.2-12a 23 1.2-13 37 1.2- 14' 37 j 1.2-15 37 '
l
-1.2-15a 37 l 1.2-16 37
- 1. 1.2- 17 37 I
1.2-18 37
- Effective Pages/ Figures Listings 1 Am. No. 59, (6/81)
r [ f ACNGS-PSAR v I EFFECTIVE FIGURES LISTING *
) CHAPTER 1 [
l
, < Q/ INTRODUCTION AND GENERAL DESCRIPTION OF PLANT l
+ *All figures, whether labelled " Unit 1" or !
" Unit 1 and 2" are to be considered applicable [
to Unit No. 1. ; Figure Amendment ( i
.1.2-1 -
37 i l 1.2-2 37 ! 1.2-3 59 I 1.2-4 37 l 1.2-5' 37 l 1.2-6 37 t 1.2-7 37 i ' 1.2-8 37 ! 1.2-9 37 l 1,2-10 37 j 1.2-11 37 l 1.2-12 37 ! 1 1.2-13 37 ! 1.2-14 37 I 1.2-15 37 ! 1.2-16 37 i 1.2-17a 37 l j- 1.2-17b 37 } 4 1.2-17c 37 ! 1.2-17d 37 [ 1.2-18a 37 ! 1.2-18b 37 l 1.2-19 37 i 1.2-20a 37 ! ! 1.2-20b 37 I i 1.2-21 37 l 1.2-22 37 i 1,2-22a '37 l 1.2-23a 37 ! ! 1.2-23b 37 ! .l 1.2 "4 37 l 1 1.2-25a 37 ! 1.2-25b 37 l
- 1.2-25c 37 j 1.2-25d 37 l l 1.2- 26 37 l 1.2-27(deleted) 37 '
i
- 1. 2- 28 37 I 1.2-29a 37 ;
1.2-29b 37 i 1.2-29c 37 l 1.2-30 37 i 1.2-31 37 i l 1.2-32 37 !
\ 1.2-33 59 l I
i 3 Am. No. 59, (6/81) l I
- l. .
. _ . _ . . _m.__.- . _ . . . _ _ . _. . . _ . _ . _ _ . . _ _ _ . - _ . _ _ _ _ . . . _ . _ . _ _ _ _ - - _
}.' . ACNGS-PSAR - ? F8FECTIVE FIGURES LISTING (Cont'd) CHAPTER 1-INTRODUCTION AND GENERAL Di3CRIPTION OF PLANT [ Figure Amendment
, i I
1.2-34 37 1.2-35 59 I 1.2-35a(deleted) 37 : j- 1,2-35b(deleted); 37 { l 1.2 39 1 1.2-37. 37
- . 1.2-38a 59 1,2-38b 59 !
h 1.2-39 37 h 1.7-la 37 - ! 1.7-1b 37 , l 1.7-2 - j- 1.7-3 - i b t t i' l0 1 l ! i 3 $ 5 i i } i 5 l 4-1 1 }- j 4 Am. No. 59, (6/81) 1 I. _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ , . . , _ _ _ _ , _ _ _ . , _ . .. _ . _ ~ . . . _ . . _ _ _ _ _ _ _ . _ _ . _ ._
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g..t" /4 ' Al i nf - a-3 1 NAN #I " ISI M' AM. N O,69 '6/1/81 (D) - DESIGN HOUSTON LIGHTING & POWER COMPANY LEGEND: Allens Creek Nuclear Generating Station A RE - RADIATION ELEMENT > (GM TUBES) GENERAL ARRANGEMENT PERSONNEL ACCESS BUILDING FLOOR PLANS 4 l FIGURE 1.2-38a 1 1
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i 1 ._ i NOTE: THESE SECTIONS DO NOT CUT THE PROPOSED LOCATION OF THE OSC. i AM. N J.59 6/1/81 (D) - DESIGN HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 ! GENERAL ARRANGEMENT l PERSONNEL ACCESS BUILDING ' SECTIONS FIGURE 1.2-38b L
. . . _ . . . _ _ _ _ _ . _ . . _ _ _ . . _ . _ _ . . . , .-z_~_-__._,__ . ~ . . . _ _ _ _ _ . _ . . _ . , . . _ .
I i 5 t, 1: ! ACNCS-PSAR' LIST OF EFFECTIVE PAGES (Cont'd) , i CHAPTER.2 i l: i i i L Page No. Amendment No. 2.1-39 - { !2.1-40' -
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p I I ! 2.2-1 36 + ! 2.2-2 42 .! ! 2.2-2A 51 .i' L 2.2-3 36 .i' I 2.2-4 36 ! .2.2-5. 59 } 2.2-5a- 43 2.2-6 . 43 { l 2.2-7 42 1 2.2-8 1
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1 2.2.A-10k 59 ! l ! 2.2.A-101 59 j 2.2.A-10m 59 ! 2.2.A-11 59 1 2.2.A-12 59 ! 2.2.A-13 59 j 2.2.A-14 59 l 2.2.A-15 59 j 2.2.A-16 59 l9 f 3' Am. No. 59, (6/81)
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I h ACNCS-PSAR L l-i LIST OF EFFECTIVE PAGES (Cont 'd) i- . CHAPTER 2 i 1: l 1-I' . i Page No. Amendment No. > I 2.2.A-17 59 I 2.2.A-18 (Deleted) 59 j 2.2.A-19 (Deleted) 59 ; 2.2.A-20 (Deleted) . 59 l 2.2.A-21 41 l , 2 . 2 . A-2 2 41 ] } 2 .2 .A-2 3 59 ! 1- 2.2.A-24 59 l t: {- t .2.3-1 2.3-2 1 2.3-3 } 2.3-4 1 2.3-4a 46 2.3-5 -- ) 2.3-6 ! 2.3-7 1 2.3-8 36 l , 2.3-8a 1 .l 2.3-9 1 9 2.3-9a_ 2.3-10 2.3-10a 36 36 1 1 ! 2.3 4 2.3-12 1 i 2.3-13 39 I I 2.3-14 -- i i 2.3-15 -- 1 l ! I I t i I I t I 1 l 1 l 1 l
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3a Am. No. 59, (6/81) l 3
l ACNGS-PSAR EFFECTIVE FIGURES LISTING CilAPTER 2 f SITE CllARACTERISTICS Figure No. Amendment No. j ! 2.1-1 36 i l 2.1-2 36 2.1-3 l 2.1-4 36 2.1-5 36 i 2.1-6 36 < 2.1-7 36 41 l 2.1-8A 2.1-8B 41 2.1-9 ;6 2.1-10 __
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! 1.1 (Appendix 2.2A) 41 1.2a (deleted) 59 1.2b (deleted) 59 1
'9 1.2 1.3 1.4 1.5 59 59 59 59 ! 1.6 59 1.7 59 1.8 59 1 1.9 59 :
4 1.10 59 ] 1.11 59 1.12 59 l 1.13 59 1.14 59 1 1.15 59 2.1 59 I I I 2.2 59 3.1 41 3.2 41 3.3a 41 f 3.3b 41 2.3-1 __
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ACNGS-PSAR
-s 2.2.3.2 Gas Pipelines When relocated, the 24-inch Texas Utilities Company pressurized natural gas line will pass about 9300 feet northeast of the nearest Category I 38(U) structure. The six-inch Shell Pipeline Company liquefied petroleum gas (LPG) line as well as the 8-inch crude line pass about 8000 feet northwest af the nearest Category I structure (see Figure 2.2-2).
lities Company pipeline operates at a maximum flow The rate 24-inch of 2.5 Texas x 10 ftUtj/ day at operating pressures ranging from 750 psi 8 to 900 psi. The design opetating pressure of the line is 975 psi. 41 Q312.5 The closest proximity of the line to any Category I plant structure is 9300 Q312.6 feet to the ultimate heat sir.k structure. Neither the liquified petroleum gas line (LPG) nor the crude oil line be-longing to Shell Oil pass under the lake. The closest pro.cimity of these lines to the lake is approximately 2500 f t from the lines to the dam. The closest proximity betweet; taese lines and Category I plant structures is approximately 8000 feet. The consequences of a rupture in the LPG line at its closest poin t to plant Category I structures have been evaluated assuming worst atmospheric dis-persion as well as formation of a low lying non-dispersive petroleum cloud along the terrain depression outlined by the 140-foot isocline which fol-luws the Allens Creek into the lake. p ( ) The 6" LPG has been evaluated under the assumption of the line carrying _/ pure propane. l 59 The hazards to the Allens Creek Plant Category I structurer and cooling lake dike from detonations of gaseous clouds resulting from breaks in proximate natural and liquefied petroleum gas lines have been analyzed 41 using conservative but realistic models. Q312.5 Q312.6 Results of the analyses which are presented in Appendix 2.2-A indicate that the plant's Category I structures will suf fer no adverse ef fects from the consequences of any pipeline rupture and subsequent detonation and/or deflagration of the cloud. 59 In the event that the applicant's nnalyses do not demonstrate the assurance recommended by Standard Review Plans 2.2.1-2,2.2 that the postulated rupture of Shell 6" LPG line need not be considered as a design basis event, physical changes to the site and/or environs to provide such 51 assurance will be made. No later than the submittal of the application for an operating license, the applicant will provide for staff review and approval physical measures to cope with the potential hazard. If necessary, the applicant will relocate the pipeline if acceptable resolution cannot be demonstrated by analysis or other alternate physical measures. 52
,.m I \ \s 2.2-5 (U)-Update Am. No. 59, (6/81)
i AChGS-PSAR jm (v, I APPENDIX 2.2A PIFEL1ht BREAK EVALDATIOh 1.0 E"aLDATION OF THE BREAK OF 6" LPC PIPELIhE The consequences of a comp 1.=te severance of the 6 inch LPG line has been evaluated for the case of 15.o line filled with volatile fluid and both with and without consideration of the use of a leak detection system. The choice of the volatile hazardous fluid for analysis is propane. There are two reasons for this choice. a) The line currently carries batches of gasoline, isopentane, normal butane ano isobutane, plus distillate, but could carry propane also, ann possibly other hydrocarbons. Found for pound, propane has ef fectiveley the same ThI mass equivalen'cy of other flammable hydro-carbons such as butane, irobutane, ethylene, butadiene, propylene, etc. (See Refe.rence 1) 59 b) There are several instances of unconfirmed propane-air cloud deflag-rations and at least one detonation, against which it is possible to compare results of the subsequent analyses. The analysis has been performed on the basis of the following assumptions: ['~'}j a) Double endeo rupture of the line occurs instantsnecosly and at the ( closest point to plant Category 1 structures (8000 feet). b) The released petroleum liquid gas mixturt escapes from the break at the critical velocity for two phase flow, and at the design pres <ure of the line, 1,000 psig (a conservative assumption since the operat-ing pressure is only 750 psi). 4 c) The temperature of the atmosphere is assumed to be 72 F. Higher temperatures would lead to higher vaporization of escaping propane, but the flow rate would be less due to the higher quality at the exit plane. d) Five percentile meteorology is assumed, Which is equivalent to a Pasquill F inversion with wind speed of 0.8 mps in the direction of the plant structures, e) A leak-detection syster is available which is capable of detecting a leak of approximately 20 bbis/ hr (or about 5% of operating flow) in 3 minutes after the leak occurs. After detection, line shutdown and isolation will be taken in 5 minutes. Normally the pumps at the Sheridan Station, located approximately 34.0 miles upstream from the Allens Creek crossing would be stopped and the line block valve at John Sue Junction would be closed. John Sue Junction is located approximately 15 miles downstream from the Allens Creek crossing.
/'~'N Doth actions would be accomplished by operator intervention rather than (v) automatically.
2.2Aal Am. No. 59, (6/81)
~ .
- AChGS-ptAR fx [ j horeover, there are block s alves straddling the Allens Creek crossing v vicinity as shown in Figure 2.2--2. The upstream valve "A" is located approximately at Station 4120+99 best of hwy 36 just past the western fork of Allens Creek. The other valve "B" is located at Station 3326+00, west
- of the Brazos River and approximately 20,000 feet downstream ot the first .
valve. These valves are currently manual operated, but can be upgraded to close remotely when a signal is given by the operator.
-1.1 CALCULATIch 0F FLOW RATE OUT OF ThE BREAK Two kinds of break require examination:
a) learcs larger than 5% of operatint, flow requiring 8 minutes to iso-late the section containing the break, and b) leaks smaller ' than 5% of operating flow. The large break scenario is examined first. The propane in the line will, upon the i cant of the break, decompress isenthalpically to a saturation pressure of 125 psia immediately because 59 of the very large speed of sound in the liquid. A decompressioit wavo 411 travel very rapidly away f rom the break leaving the fluid behind at tSe saturation pressure. . Since propane would issue from the break to atmosphere at 72 F, approximately 1/3 of it would quickly vaporize, cooling the re-mainder to its boiling point of about - 44 F. Hence the process of de-compression is described by the throttling process shown in Figure 1.1. From
.b) that figurar the exit plane quality, x, of the fluid can be estimated from: > v=v g +xV gg 3
where v=2.4 ft /lb, v =.0275 ft /lb, v =6.6 ft 3 /lb, v 3 f g fg=vg-vf Ilence x = 0.36 i lo estimate the flow rate out of the break Fauske's equation Reference 2 for critical two-phase mass velocity is used:
! G g = ( ga/(k dv /dp+k dx/dp+k g 2 3dv g /dp)) !
where kg= (1-x+ax)x k2 " "g (1+2 ax-2x) + v g(2ax-2a -2a x+a ) k = (1-x(a-2)-x 2(,_g)) 3 and a = (v /v )l/2 g Initially the break discharges approximately 568 lbm/sec. This discharge rate decays as the pressure in the line drops, until such time that inertial flow is established . (/ During the first 8 minutes following the break, it is very conservatively 2.2A-2 Am. No. 59, (6/81)
ACh65-PSAk m *
> \ .
At that time the pump assumed that the 566 lbm/sec remain constant. _( j pressurized side of the break is assumed to be about 200 psi dile the ; pressure on the other portion of the line is assumed to have fallen to
- 100 pai.
Inertial flow conservatively computed with single phase rather than two phase ; friction factors, is predicted to peak at 70 lbm/sec from the pump pressur-ized side of the break ano 30 lbm/sec from the other side. The full break scenario is depicted in histogram (See Figure 1.2) wherein the , solid line represents the tiow rates chosen for the analyses, and the dashed , line represents what is physically expected. 1.2 LALCOLAT10h 0F DETONALLE CLOUD SIZE To estimate the ' atmospheric'* ly dispersed propane and air cloud under Pasquill F conditions with .. f ps wind speed, a coc it ant average source of propane of approximately 160 lbm/sec has been assumed to continue to release 59 trun the tjreak indefinitely. At a specific volume of propane of app 5 *i""'*- i ly 9.0 ft /lbm, this corresponds to a source 3strength ot 900-1000 ft /sec ot propane. Conservatively assuming 1000 ft /sec, and further assuming that vertical dispersion is depressed to account for the heavier than air character of the ensuing cloud, the resulting cloud is depicted in Figure 1.3. l The centerline (directly downwind) com.entration of propane is determined by: [] 0
- kl " ## Y *E" where Q = 1000 it3(S1P) of propane per second 3
u = 2.6 ft/sec and ey, *z are the plume dispersion standard deviations d tained fron reference 3 4 The of1-centerline concentrations are determined by , X*X c3 exp 1/2 hy/,y)2 + (*/z)] It is very important to determine which are the proper disparsion standard deviations to be used in solving for the equilibrium concentrations. Dispersion standard deviations for instantaneous "puf f" releases and cantinu-ous constant releases are considerably dif ferent; the former being much smaller than the latter.
*1he reason for this is that the continuous release deviations account for a i part of averaging process caused by wind vagaries resulting in meandering of the plume about its axis. The ay and *z then account for this ef feet. For a Og puti release this acaadering averaging process is absent and the puff dis-The case of a pipeline break is neither a puff release -(j perses more slowly.
nor a continuous constant release, but rather a continuous release at an 2.2A-3 Am. No. 59, (6/81)
ACNGS-PSAR initially rapid decreasing rate, followed by an almost constant release. The meandering procesa would then be in ef feet and the proper dispersion standard deviations are the continuous release standard deviations. Table 1-1 lists pertinent centerline concentrations at several distances from the break, and also distances from the cloud axis at which the propans concentration f alls to flammable and detonable limits. Since lower volumetric flow rate result in smaller cloud formation due to in-creased atmospheric dispersion, the small break scenario is bounded by that previously described for the large break incident. 1.3 CALCULATION OF EFFECTS OF DETONATION 1he capability of t'he Allens Creek plant to withstand potential explosions is evaluated by determining the total quantity of detonable or deflagrable mix-ture contained in the detonable or deflagrable cloud. 7 3; The detonable volume is calculated from Figu9e 3 3 to be 1.155 x 10 f t 59 whereas, the deflagrable volume is 1.% x 10 ft . The total mass of pro-pane contpined in the larger deflagrable volume is estimated to be 9.14 x 10* lbm. One can further estimate that the quantity of prgpane contained in the over rich cloud region is as a minimum 2.14 x 10 lbm. Moreover , ciore than 90% of the mass of propane is dispersed past the lower flammable limit. 6 ( ) Thus to achieve the clouds depicted in Figure 1.3 in excess of 2 x 10 lbm of propane must be released. This is a larger quantity of propane that can escape diile the line is not isolated (8 minutes) and following isolation of the line. For instance if the line is isolated by stopping the pumps at the Shgridan Station and blocking the valve at John Sue Junction, at most 1.8 x 10 lbn of propane can be ejected. This fact demonstrates that the cloud depicted in Figure 1.3 is a very con-servative upper bound of real possible clouds, and that therefore if the plant can withrtand the consequences of its detonction or deflagration, it will be safe against any real explosion. The detonation is conservatively assumed to be at the extreme edge of the cloud closest to the plant. Thus detonation parameters are evaluated at a distance of 3700 ft. At this distance, detonation of the entire detonable cloud, calculated f rom Figure 1.3 assuming no iso 1.ation of the line would result in the following parameters at the niant anforv related arrncturen: Peak - Overpressure 0.6 psi Peak Dynamic Pressure 0.009 psi Peak Reflected Overpressure 1.22 psi Positive Phase Duration 0.281 msec.
/ T
( ) v 2.2A-4 Am. No. 59, (6/81)
1 ACNGS-PSAR
/ Peak Acceleration 0.048 g
}
(v/ (Horizontal or Vertical) Peak Velocity 0.064 Peak Displacement 0.0057 inches The plant can withstand these blast induced loadings. The parameters have i been conservatively evaluated using the methods described below. From an enthalpy of detonation release of 260 Kcal/lb of propane air mixtures of 4.9 percent (enthalpy of detonation is insegsitive to mixture ratios be-tween 4-5 percent), and a volume of 1.155 x 10 cubic feet , it is possible to compute the total energy released in a hypothetical detonation of the entire detonable cloud, assuming that the whole cloud is at a mixture averag-ing 4.9 percent. Fromthetotalweightofthgmixture(944,382 lbs) one com-putes the total enthalpy released (2.455 x 10 Kcal), divides it by 500 59 Kcal/lb TNT to derive the equivalent weight of . TNT. A 507. yield is conserva. tively uced to compute the TNT explosive equivalent yield, whie is then that of 127 tons of TNT. Since work by Iotti et al, shows that inG A the yield of a gaseous detonation is lower than that of TNT. Reference 4 / compares ' overpressures calculated by assuming gaseous point sources 5, /~to overpres-sures obtained by Kingery 6, / for the same yield, and those measured by Kogarko et al, 7,/ for the given gaseous detonation. This comparison shows that Kingery's result would have been comparable to those of References 5 and 7, if a TNT yield of 50 percent had been employed. Reference 8 / cites a -
/ yield of 7.5 percent. Thus a conservative estimate of the TNT equivalent of the detonation of the entire cloud can be obtained by using 50 percent yield
() and use of Kingery charts. It must be pointed out that the method chosen to compute the equivalent TNT yield of the detonation dif fers from that reconsnended in Reference 1. If one had used the methodology of Reference 1, one woulg have first computed the weight of propane in the detonable cloud (6.056 x 10 lbs), multiplied it by 2.4 to compute the equivalent weight of TNT (72.67 tons). As can be seen, the method chosen is more conservative. As previously de-scribed the small break event is bounded by the large break analyzed above, this the plant is safe against detonation of atmospherically dispersed clouds resulting from any size in the pipeline.
).4 GRAVITY SLUMPING EFFECT OF LPG CLOUDS Since there is a potential for propane cloud to flow along Allens Creek, spillover the 140-foot isocline near the site and reach the plant safety re-lated structures in suf ficient concentrations to pose a hazard, analyses have been performed. Therefore the following sections have discussed the different ef fects of the formation of low-lying propane clouds which may be governed by gravitationally induced flow with limited vertical dispersion.
1.5 CLOUD FORMATION AND DISPERSION MODEL fh
\J > This section examines the gravity slumping model associated with the propane 2.2A-5 Am. No. 59, (6/81)
ACNGS-PSAR (\ cloud resulting from 6 LPG inch pipeline break. For boiloff rate
/ )
b' B = 33,800 BTU 2/ f t hr it is possible to compute initial radii and height ' of the pure propane clouds formed at the break as a function of the initial release rate, W assumed to be a constant y
= Pf (h - h )_ p (ft) r"#* . ,
wB (1) i and # h = max (ft) init p u(h -h) g 7 (2) , i wherein B is the boilof f rate, h and hg are the vapor and liquid enthal- 59 pies of propane at thechosente$perature, p and pf are the vapor and ; liquid density of propane at that temperature, and u is the wind velocity.
- 8 The gravity spread of the cloud is then computed by .
i
\
dR(t) P a (3) dc
= 2g .( c(t) p,
- P ) h(t) + S aR where p(t) is the time varying cloud density, p is the density of ambient air, g is the acceleration of gravity, S ,ts the slope of the ground
( p) in ft/ft, and h(t) is the time varying cloud height. V R(t).is the radius of the cloud at time t, or alternatively the distance travelled by the cloud tront at time t after the cloud makes contact with banks confining it at an arbitrary distance W, where W can be taken as the , half width of the channel. Air is entrained at the surface of the cloud as it spreads. If the entrain-ment velocity is defined as V e
= r I
$ while the cloud slumps radially (4) dt and V
= dR -u while it slumps along the wind (5)
E direction, where u is the wind velocity the volum.tric entrainment of air during radial spreading is given by I dQ(t).= av 2wrdr (6) while during the subsequent phase of slumping along a channel of width 2W l 2.2A-6 Am. No. 59, (6/81)
~
I
- ACNGS-PSAR t
v/ assumed to be in the same direction as the wind direction, it is given by dQ(t) ,av 2Wdr (7) Integration of equations (6) and (7) using equations'(4) and (5) yield . Q(t) = 3 R (t) d f r radial spreading (8) and 2 -u (9) Q(t) = { W - TW [R(t).W] }a 59 for initial radial spreading tc a radius W followed by axial spreading down wind in a constant width channel (channel width = 2W). In equation (9) the parameter 7 is used to distinguish between triangularly shaped and rectangularly shaped channels ( 7 has value of unity for tri-angular channels and 2 for rectangular channels). No entrainment is considered for the upstream of the radial spread. ( With an entrainment coefficient a = 0.1 (see Reference 9) the conservation of mass and energy equations are solved as a function of tim numerically. The initial motion of the cloud is radial, and the entrainment velocity goes from a maximum at the cloud edge, to zero at the center. When the cloud edge reaches the banks, further motion is only down the channel, and the velocity from that point is maintained uniform at the computed value for any time for all channel sections downstream of a distance equal to half the , channel width. The velocity is decreased to zero tram that distance to the break location. The cloud properties at any point in time are computed as if the cloud were of homogeneous composition. The results are shown in Figures 1.4 and 1.5. Figure 1.4 details two cases. They are (1) an initial source of 100 lbm/sec. lasting 995 seconds tollowed by a flow of 30 lbm/sec. and (2) a constant flow of 100 lbm/sec. contained indefinitely. Figure 1.5 assumes an initial source of 568 lbm/sec. for the first 480 seconds, followed by a flow of 100 lbm/sec for another 515 seconds and a flow of 30 lbm/sec. for another 2750 seconds. From the results of Figure 1.5 it is obvious that initial cloud heights, re-gardless of wind velocity are so high that atmospheric dispersion would be sure to occur. Moreover, the 566 lbm/sec discharge has been conservatively assumed for 460 sec, whereas
- real discharge will decay from the initial 568 lbm/see to 100 lb/sec. h . e, the more realistic slumping results will be somewhere within the resulta presented in Figures 1.4 and 1.5. Figure 1.4 illustrates
[}\ / the effect of variability of source on cloud extent. U For constant flow rates there exists a wind velocity under which very long 2.2A-7 Am. No. 59, (6/81)
= _ ._ . ._. ._, - -. . _ _ _ .
ACNGS-PSAR
--{-
/N clouds could be predicted. The variability in flow from'the break combined (U ) with the presumed constancy in wind velocity during the intervals of time considered remove the possibility of these very long clouds. In fact for constant channel widch (300 f t), the longest distance down the channel fore-casted for the postulated full break is between 8500 and 9500 ft.
In reality the Allens Creek depression has channel width at the 140 feet isocline ranging from approximately 300 feet to more than 550 feet approxi-mately 4000 feet down channel from the pipeline crossing the Creek as shown in Figure 1.6. Figure 1.7 shows the distance which the cloud can reach for wind speeds of 6 f ps as a function of channel width for flow rates, but of the break corresponding to those used in Figure 1.5. Note that the final cloud height is always sufficient to exceed the 145 foot isocline, hence the effective channel can be considerably larger than 600 feet after the cloud travels more than 2000 feet down the channel. To compensate for the widening of the channel as the cloud height exceeds the 59 145 foot isocline, a rectangular channel of cross-sectional area roughly cor-responding to the average cross-sectional area of the Allens Creek channel from its lowest point to the 145 foot isocline (depth of channel approximate-ly equal to 25 feet so that cloud travel is overestimated) can be used to estimate the farthest reach of the cloud. Between Sections 14E and 11E of Figure 1.6 representing approximately 5000 feet of down channel distance the cross-sectional area up to the 145 foot isocline varies between,9000 square feet to 11200 square feet at Sections 14E and 13E to even larger values down channel. For instance, at llE it is more than 17500 square feet.2 Thus, an p average yalue can be taken to be linear between 9000 and 18000 ft or i 13500 it . For the depth of 25 feet, the equivalent width of the rectan-gular e.hannel is 540 ft. From Figures 1.5 and 1.7, therefore it is expected that the cloud resulting from the full break in the pipeline will not reach further than 6000 feet down the channel. To assess the hazards to the plant from the homogeneous cloud resulting from the gravity slumpimg calculation, the total quantity of propane released to the point wherein the cloud reaches a homogeneous average $d etonable concen-tration of 2.8% is calculated. This represents 3.23 x 10 lbm of propane, the detonation of which by the methodology of Reference (1) would correspond to the detonation of 0.39 IsT of TNT, which is roughly three times what had been conservatively estimated for the detonation of the atmospherically dis-persed cloud. This detonation would pose no safety hazard to the plant. The preceding gravity slumping analyses have assumed a considerable amount of an entrainment in the propane cloud. Experiments (See References 10 and
- 11) conducted with fluids of density differences not comparable to the prob-
'lem at hand; i.e, P 2 /gP <y 1; whereas, dCH/3g a 2.0 initially ano p / pg g > 1 always, inotcate that actual entrainment of the lighter Ek0YN may be considerably bss. Therefore, the gravity flow of pr'opane down the Allens Creek has been also addressed under a condition in which air entrainment is assumeo to be negligible.
For this condition, the characteristic of the propane flow down the Allens Creek are determined from the following equations which have been adapted from those 01 Reference (11) for 1-D flow in a constant channel {}/ V 2.2A-8 Am, No. 59, (6/81) l l _ _ _ _ _ _ - - . _ . - , _.I
L i ACh6S-PSAE
. f i s.
for O $ t f t (10) l; r = 1.26 (g' S)1/3 w t m i e E 2 e for L $ t g t (11) f h(ra ) = 0.794 (2 0 -)1/3 > g,w i t 1/3 i t
= 0.4 (g'wc 0 3} ;
I f 2 4/5 e for t 2t t (13) r m
= 1.05 2(S 81-)1/5 t 59 !
w e ' I 4
; 3 1/5 1/5 for t 2: t e
(14)
+
h(r m) = .954 (E E--) 3 t j , g, w l Therein w is the channel width, Q is the volumetric flow rate and t is the time at which flow changes from inviscid (i.e., no frictional ef fects), i , to viscous. The initial inviscid flow is due to the fact that the initial l Froude number is larger than unity; i.e., the flow is supercritical, but approaches critical very quickly. I c is the frictional loss which can be expressed as ; n= 0.075 for rough (15) c = g (1.486 h 1/6) 2 river beds 4
' n i I Also !
( i 9- 9 g' = g ( a) (16) p a > In equation (15) it is assumed that the width of the channel is suf ficiently , larger than the depth of the flow, h, so that the hydraulic radius is approx- i imately equal to the depth, h. i l For Allens Creek, the channel shape between the locations of the pipeline cross-over and tre plant vicinity can be taken as square bottomed with an average width of 50 feet for a depth of 6 to 7 feet. 2.2A-9 Am. No. 59, (6/81)
ACNGS-PSAR
.(m *j .In the preceding equations, r, is the distance downchannel of the front of 4
V]' the cloud at time, t. Solution of the above ' sets of equations yields the following cloud front, distance ard average depth as a tunction of time as shown in Figure 1.8. It should be noted that to achieve the distances and depths given by Figure 1.8, it is necessary for the source of propane to continue for times longer than the time of flow when the line is isolated. Two casas are then possible, a) If the line is isolated the maximum depth will occur at a distance of $3330 teet downstream of the break location and will equal 6.3 feet. Therefore, the depth will decrease as the cloud continues slumping, 59 b) The line is not isolated and flow continues indefinitely. Equations , (10) through (14) are applicable only until* steady uniform flow is achieved. The depth for such flow can be computed from Bresse's equation (See Reterence 12) dh "
~ l dr 2 VW (17) m 1 Ag O
l where i is the slope of the channel, A its cross-sectional area, R its hy-draulic radius, v the flow velocity, add . b = RC E (16) I All other terms are as previously defined. Equation (18) is entirely analogous to the equations of Reference (10) for flows with entrainment, which are reproduced below (2-fSR) g E-SR tan a + C 2 ! dr 1-SR (ggy 1 s ' 1 dR (1 + y S R )E - S R2g tan a + C 3R dr 1-SR (20) i m 1i
,O i L./
2.2A-10 Am. No. 59, (6/81)
Act.GS-l'S Ak l where the Richardson nuuber, h;, is defined as ( )} (p -p ) cos a i
"8 2 (21)
PV S g and S are shape cocificients, E is the entrainment coefficient, and 2 a is the angle of the sloping channel. 1or steady uniform flow, and no entrainment, it is necessary that 9 R tan u =C (22) 59 where 5 is taken as equal to 1.0 (i.e., no vertical acceleration of the 2 fluid) Equation (22) when solved yielos a depth oi 9.53 feet for a channel width of 50 tect. Consideration of entrainment, by solving equations (19) ano (20) would in- $j} / crease depth in the initial region of the cloud, where the Richardson num-ber is less then unity. That number becomes rapidly larger than unity how-ever, so that the effects of entrainment can be considered negligible (i.+ ., E -+ b a s h1 . 2 1.0) (See Reference 10). The steady uniform flow depth ot 9.53 feet, according to epations (10) through (14) would be reached at a distance of approxivately 50,000 feet downstream of the break (well into the lake) ano a tir..e of 35000 seconds.
'lhus it can be concluded that the depth ri the cloud in the proximity of the Allens Creek plant will be in all likelihood of the order of 7 feet or less (less if the line is isolated) and the width of the cloud will be about 50 teet.
'lhe cloud will be pure propane vapor (non-detonable). To determine the po-tential effects on the plant, the plant safety related structures of which are located 1600 feet away from the channel at its closest proximity, the amount of propane mixing with air and thus leaving the channel bed and being transporteu toward the plant by atmospheric dispersion, is determined from kL 1/3 cD
=
0.36 ( # ) (## ) 0.8 (23) AB AB where k, is the mass transter coetficient, L the go.pth from ttje break [ ,} location , c the molar density of air (2.589 x 10 # mole /ft ), v the (j cloud velocity which is about 2.2 fps, e the air density = 0.075 lb/ft 3 , 2,2A-10a Am. No. 59, (6/81)
- ~. - - _ _ __ __ _ _
Y ACNqS-PSAR a [3 a t 7 b 1:, e the air viscosity 1.05 x 10 -5 lb/ft sec, and i f , g i 1.823 1/3 5/12 Dg = _2.745 x lo ( ) (P g Pg) (T g l'g) . I (24) ;
. - (I-- + b )1/2 cm sec. i A B !
where H = 44.09 lb A t
. l M = 26.97 lb b
I gg = 37 6 O P A
= 42'atm 59 I
.l P b
= 36.4 atm !
T b
= 132 h i The heat transfer from the ground has been computed to be insuf ficient to heat th'e cloud too much above its -44 F temperature, as long as the source
% of propane is continuous. Eence the value of the mass transfer coef ficient, k and diffusivity DAB, have been computed at -44 F, for air at ,
-753 F. The length L has been chosen as 12,000 feet, this being the distance '
from the break to the plant along the channel. Longer distances would result in slightly lower values of k,. ; The computed values of D AB and k, are
-4 2 -1 D " * * '"
AB f
-4 2 .
k m
= 2. 8 x 10 lb/ft sec ;
i i b i l, {
-4 long line source of strength, Q = 2.8 x 10 l Assumgng an intinitely /sec (using a width of 50 ft and a propane den- '
, lb/ftseeor0.179fg sity ot' O.107 lb/ft at 70 F), the dimension of the detonable and the , deflagrable cloud downwind of the source have been computed for a wind speed l of 2.6 fps. The results are shown in Figure 1.9. l > - Assuming that the actual length of the cloud parallel to the plant and chan- j j nel is no longer than the 26,000 f t separating the pipeline rossing' location :
- and the lakeexit-ofghegliensCreek, the maximum detonable volume is com- I puted to be 1.23 x 10 ft . This volume contains 64,410 lbs. of propane !
[ at an average congentration of 4.9 percent. Its detonation is equivalent to { Q that ot 1.55 x 10 lbs. of Thl. , i k 2.2A-10b Am. No. 59, (6/81)
,w - --_ - . - , , , . - - - , . -r-- ,, - ----,+-yre vw----,w y---w, --
yy,, yvve,,---
r ACNGS-PSAR
, ")
Assuming that the center of the detonation is 400 feet from the channel, or 1400 feet from the plant nearest safety related structure, the peak overpres-sure to which the plant would be subjected is a tolerable pressure of 1.o psi. It is therefore coa.i.ided that the propane flow down the Allens Creek would not present an unaccec table hazard to the plant safety related structures. l.6 EVALUATION OF THE HAZARD OF THE DEFLAGRATION FROM LPG PIPELINE BREAK A vapor rloud of flammable concentration may burn (deflagrate) or detonate (if within the detonable limits) or both types of combustion may occur in case of transition from deflagration to detonation. Volumetriu explosions may also occur particularly if partial confinement of the cloV exists. More likely, pockets of gas in a cloud may explode volumetricallt if beated to the autoignition limit by radiated heat or shock waves. 59 In general, the deflagration, detonation, and volumetric explosions are the co:nmon modes of combustion to be considered. In our case , due to the uncon-fined nature of the cloud, the latter mode cannot occur, and it is only necessary to address detonations and deflagrations. For detonations, all of the thermodynamic properties, detonation velocities and flow properties behind the detonation front are calculated from standard thermodynamic equilibrium calculations. [} (' > If th and U2 are defined as velocity of the unburned gas and burned gas with respect to the stationary detonation wave, then with respect to a stationary o'us erve s U is the detonation velocity and W = U -U is the velocity of g 2 the gas dehind the detonation wave front. Further, the thermodynamic states behind the detonation front are described by the Hugoniot equation (25) AE=E 2
-E =
(P2+P)(v -v)2 Wherein p and v are the pressure and specific volume and E the energy and I and 2 denote the unburned and burned states, respectively. The actual detonation involves a passage through a family of Hugoniot curve which proceed from the curve corresponding to no chemical reaction wherein AE=E 2 -E g = Cy (T2 - T g) (C yis the average specific heat at constant volume between tem and T before and after the passage of the wave front) peratures Tto thei curve 2corresponding to complete chemical reaction in which case E -E =C (T2 - T ) - AE (26) p ( ) wherein AE 'is the energy released in the combustion process. 2.2A-10c Am. No. 59, (6/81)
ACNGS-PSAR Figure 1.10 shows the llugoniot curves. The llugoniot curve for the complete reaction is distinguished by two branches. The branch from A to V is the
" detonation" branch and the one from B to C is the " deflagration" branch.
It has been shown that the only possible stable detonation is the detonation proceeding at the minimum detonation velocity (see Reference 16). This is the Chapman Juguet detonation and the resulting detonation overpressure for stoichiometric propane air mixture has been determined to be p 2/P1 = 17 . 8 f rom an initial stata of 14.7 psi and 460 F. That value is closely correspondent to that presented by J H Lee (see Refer-ence 17). Point A corresponds to the overpressure resulting from an adia-
/p = 8.5 batic dif fersconstant slightlyvolume explosion.
from that quoted inhevarious value obtained of pkncluding Reference references, (1), which reports 8.34. The dif ference is due to the initial state assumed for these calculations which was O C, whereas others generally used 25 C. The intersection of the tangent line with the llugoniot curve for no reaction is the Von Neumann spike which precedes the C - J detonation pressure. Its computed value is approximately p2 /p1 = 28. 59 The detonation velocity which is given by
~
U =D=v 2
~
1 (27) s 1 1 !. V -# J
) 1 2 is computed to be D = 5330 fps or 1610 mp, which is in reasonable agreement with literature data.
In the detonation branch the flow velocity of burned gas, W, is in the same direction as the detonation wave. In f act , it can be shown that (see Refer-ence 18) D=W+C (28) where C is the velocity of sound in the burned medium. Since the detonation wave travels supersonically with respect to the unburned medium, no disturbance precedes it, hence the cloud size remains at its initial size as the detonation propagates; i.e, the gas expansion occurs afi :rv1rds and has the ef fect of a rarefaction wave which tries to overtake the detonation wave et sonic velocity in the burned medium, his observation has been made by others (see Reference 17, page 12). Thus, the ef fects of the detonation blast of the wave of the propane gas cloude has already been properly addressed in the preceding subsections. The portion of the curve between A and B corresponds to no real physical
..~ state. The portion of the Hugoniot curve between B and C represents a de- '
l ; c rea se in pressure and increase in volume, corresponding to a raref action. _J 2.2A-10d Am. No. 59, (6/81)
ACNGS-PSAR i < l ) The burned gas flow velocity, W, therefore is always negative; i.e., the
\- / burned gas no longer moves in the same direction as the wave, but away from it.
The consequence of this is that in this deflagrative process a pre-compres-sion wave is sent out into the explosive mixture to push that unburned gas with a velocity just sufficient to ensure that the gas may come to rest when it is swept over and barned by the deflagration front. Physically no deflagration occurs past the point C (Reference 16) which is known as a Chapman Jouguet deflagration point. In fact, weak deflagrations (essentially constant pressure explosions) are those to be considered for vapor clouds explosions, and are the ones that have been observed in experi-ments. Whereas the C - J detonation velocity represents the minimum of all detona-tion velocities, the C - J deflagration velocity is in fact the largest possible deflagration velocity, which is calculable from equation (27). 59 For the propane-air mixture studied (stoichimetric) the C - J deflagration velocity was computed to be 168 fps. For weak deflagration, however, the deflagration velocity is the same as the laminar burning velocity, 1.5 - 2.0 fps, hence thousand times less than detonation velocities. (, ~) However, the spatial deflagration velocity, which would be th t seen by a
\~ - stationary observer is higher than the deflagration velocity computed by equation (27). This is due to the displacement of the unburned gases ahead of the propagating flame caused by the specific volume increase across the flame front. Since this increase is about eight-fold, the spatial deflagra-tion velocity is roughly eight times that computed by equation (27). Hence, for a C - J deflagration spatial velocities of the order of 1300-1400 fps could occur. In actual tests however, (see References 1 and 19) the spatial flame velocity has been measured to be of the order of 30-40 fps, for mix-tures difficult to detonate (like propane air) and ten times larger for mix-tures rich in oxygen.
The ef fect of oxygen richness is not surprising since lack of inerts such as nitrogen has the ef fect of raising the Hugoniot curve for complete reaction to higher values. For instance, the C - J detonation point for a stoichio-metric propane-oxygen mixture would correspond to overpressure almost exactly double that occurring for the detonation of propane-air mixtures, with deto-nation velocities 30 percent higher. Similarly, the C - J deflagration velo-cities for oxygen rich mixtures is expected to be higher than that for pro-pane-air mixtures. Hence, C - J deflagrations of any kind exhibit velocities in excess of 1300 fps. The fact that no such spatial burning velocity has been observed con-firms that C - J deflagrations do not occur, but that only weak de flagra-tions; i.e., basically constant pressure burn.ing occur.
; j For these kind of deflagrations, the precompression sent into the unburned '% / medium and surrounding air is in the nature of basically an acoustic wave, 2.2 A- 10e Am. No. 59, (6/81)
ACNGS-PSAR
/ ..which does not steepen into an air shock unless spatial flame velocities of
( ]) 300 fps or more are achieved, as shown on Figure 1.11 which is taken f rom ' Reference (19). Likewise, there is no overpressure of great significance within the deflag-rating cloud. For typical spatial burning velocities of 30 fps or less, overpressure of about 1 psi will occur just ahead of the flame front. Hence, it is concluded that blast damages are insignificant for deflagrative burning of propane air clouds either near or far from the cloud. Hence, for a de-flagration the possible damage is limited to temperature. F In a detonation event, after passage of the initial detonation blast, the compressed products expand. Under the assumption that this expansion is isentropic, the final volume will be approximately 9.7 times. initial volume (assuming k = 1.25) (for an oxygen rich detonation these rates would be more than double). This expansion in turn can generate a second shock in the air ahead of the expanding products of the detonation, which follows the initial shock caused by hydrooynamic coupling of the detonation wave and the air. This second shock is one order of magnitude smaller than the first air shock
' (see Reference 20) and exhibits the same decay with distance f rom the center i 59 of detonation as the first and much stronger shock.
Since the Allens Creek plant has been shown to withstand the first shock . overpressures, it is also safe against the weaker second shock caused by the expanoing detonation products. O) During the preceding it has been tacitly assunmed that the flammable cloud is entirely tormed ot a stoichiometric mixture of propane and air (or oxygen).
'1his was of course the case for all experiments conducted (see References 17, 16, 19 and 20). In fact, the propane air cloud computed to occur as a result of atmosphere dispersion exhibits a range of mixtures which are only stoichi-metric in a region which at ground level is centered about the 1300 ft and 35W f t from the origin of the cloud. At closer distances to the origin of the cloud, ground level concentrations are over rich and farther away they are leaner than stoichiometric.
khereas the detonable limits for propane air mixtures are 7.0 and 2.6 percent by volune of propane, the limits of deflagration, are 9.0 and 2.2 percent. The volume of fgammgble mixture contained wichin the vapor cloud is c9mputed to be 1.96 x 10 ft , whereas the detonable volume is only 1.155 x 10 3 ft hence, 40 percent of the total flammable volume has concentrations below 2.8 percent, or below 0.7 times the stoichiometric concentration of 4.12 percent. The expansion ot a mixture of less than 67 percent stoichiometric concentra-tion (as well as that of concentrations about 120-130 percent of stoichio-metric mixtures is at least 30 percent less than the expaiwion of stoichio-metric mixtures. -lhis results basically from the lower flame temperature. The expansion given on Iigure 1.10 as point B is thus applicable only to stoichiometric concentrations. Conservatively.therefore, the final volume 5 of the deflagrated vapor cloud can be estimated by expanding the volume of 2.2A-10f Am. No. 59, (6/81)
ACNGS-PSAR
, x 2.8 percent or larger propane concentration to a final volume eight times
! j la ger, and the volume of lower concentration 8 t 3 v lume 5.5 times larger.
The resultant final volume would be 1.35 x 10 ft ,
Assuming that the original cloud can be represented as a hemi-ellipsoid of 4564 ft, 143 ft, and 28.6 ft dimensions in the downwind, crosswind and ver-tical direction, centered at 2734 f t, from the break, and that the deflagra-tion is centered at that point, the final dimension of the product cloud downward will be either 6716 f t from the break. The closest distance from the 6 inch pipeline to a plant safety related structure in Allens Creek is 8000 feet. Hence, it is concluded that the deflagration of the " worst" cloud ensuing from a catastrophic break in such a line poses no hazard to the plant. Finally we have to examine the deflagration of the homogeneous cloud calcu-lated by gravity slumping. Again assuming an expansion of 5.5 times for the lower temperatures resulting from the low concentrations applicable to the cloud, the length of the cloud will have increased 1.75 times. Hence, th: burned cloud can reach a distance which is approximately 2250 feet farthe down channel than its farthest reaches when unburned. 59 Since the cloud slumps by gravity, it follows the channel. 6000 feet down channel puts the furthest reaches of the cloud 3000 feet from the plant safety related structures and 2400 feet from the switchyard. Thus the burned cloud cannot reach plant safety related structures although it comes close fs to the switchyard, and the plant is safe. ! l-
\\ '/ l.7 GRAVITY SLUMPING EFFECT OF LPG CLOUD FORMED FROM SMALL BREAK OF THE PIPELINE The following scenarios to be examined are the gravity slumping of cloud formed from small leaks or breaks in the line. Analyses of the Ruff Creek incident (see Reference 21) wherein the cloud resulting from a 20.5 lb/see constant propane source was subjected to gravity slumping, indicated that the distances from the break which homogeneous clouds of 2.4% concentration are computed to reach are in general lesser than those calculated for larger flows out of the break, although wind conditions can make a cloud resulting from a smaller flow go farther than that resulting from a higher flow, there will be another wind condition at which the latter will go even farther.
This is shown by Figure 1.12 which compares cloud distances and final heights for breaks resulting in constant 100 lbm/see and 20.5 lbm/see in like chanects. Since even for small breaks the source of propane is expected to vary in time as the pressure in the line drops, a similar behavior of distance vs. wind velocity as exhibited in Figure 1.5 is expected for the smaller breaks as for the large breaks. Thus, a small break presents less hazard from the gravity slumping standpoint than the large break. 1.8 GRAVITY SLUMPING EFFECT OF THE EQUILIBRIUM ATM0 SPHERICALLY CLOUD
,m Ine gravity slumping of the equilibrium atmospherically dispersed cloud
[v) 2.2A-10g Am. No. 59, (6/81)
ACNGS-PSAR depicted in Figure 1.3 has been evaluated by considering that the continu-ously fed cloud begins its gr avity slumping from an elongated shape cloud which has an initial length equal to approximately 2500 feet, a width of 150 feet, and a height of 15 feet; in:tead of the pancake shape having the r and h. assumed previously. 'Ihis represents the over rich region oT"Ihe cloEiE which, when slumping under its own weight or because of ground slope, can entrain surrounding portions of the cloud which are already flam-mable or below flammable, and thus dilute its concentration by mixing to levels which are flammable. In turn, the region which is flammable initially also slumps and entrains non-flammable mixture thus diluting its concentration. The velocity of the sliimping, initially over rich volume of the cloud in the absence of significant slope in the ground is computed by assuming an average concentration of approximately 40% propane. The density of this mixture is about 12% higher than the surrounding mixture which is taken to be at an average concentration of 5%. The velocity of the front of this cloud is com-puted to be approximately 10.3 fps. The amount of mixture at 5% congentra-tien3 entrained per second at this velocity is approximately 1.8 x 10 59 ft /sec. The velocity of the 5% volume of the cloud which is slumping is only 5 fps, and the entrainment of non-flammable mixture .at ghis velocity with a width of 300 feet is computed to be about 1.2 x 10 ft /sec. O Of course the front velocity of the over rich region of the cloud slows down (U' ) from the initial 10 fps to 5 fps as the mixing drops its concentration to a concentration comparable to that of the initially flammable cloud, while the latter front velocity drops to essentially zero when its concentration drops to very low values. Assuming a linear variation in front velocity from the time at which the slumping begins to ghen it ends, the over rich cloud which had an original volume of 1.93 x 10 ft will have entrained essentially all of the 5% concentration gas that hasn't been diluted by its own slumping in approxi-mately 2 minutes, to form a. volume of approximately 1.35 x 10 ft 3 having a concentration which averages about 10% by volume, meanwhile, at the same time, the average concentration in the previously flammable volume will have dropped to less than 0.7%. The cloud will have advanced approxi-mately 1000 feet down the channel. The fraction now computed to be an aver-age 10% concentration will further slump and advance with a velocity vary-ing from 6 fps to zero while entraining the surrounding lesg righ mixture. Assuming an average surface area of entrainment of 5.4 x 10 ft (interf ace between 10% region and surrounding less rich region), the concen-tration will be dropped to about 5% in approximately 80 to 90 seconds, during which the flammable portion of the cloud will have advanced another 500 feet toward the plant. Thus the slumping of the equilibrium cloud formed by atmospheric dispersion has the ef fect of moving the flammable and detonable region closer to the plant by about 1500 feet, p
! ) \ s /
2.2A-10h Am. No. 59, (6/81)
i 1 ACNM-PSAR
,- m The preceding approximate calculation inherently makes the same assumptions
[ ) that are made in Section 1.2. 'w/ As such the cloud resulting from the slumping of the one depicted in Figure 1.3 is itself a very conservative +mr bound of physically possible clouds. The detonation of such cloud, now located 1500 feet closer to the plant, would still pose no hazard. More realist.ically, vapor clouds of smaller dimension can be expected. For instance, isolation of the line at the Sheridan Station and John Sue June. tion limits the maximum vapor cloud size to sixty percent of that shown in Figure 1.3. If such a realistic cloud were to move 1500 feet closer to the plant, then deflagrate, the farthest reaches of the burn marks would extend to a distance of 7150 feet from the break location, and hence pose no undue hazard to the plant safety related structures. 59 1.9 ATMOSPHERIC DISPERSION OF CLOUD INITIALLY MOVING BY GRAVITY The scenario to be examined is that of gravity slumping followed by atmos-pheric dispersion. Except for the analyses performed for the low wind velocities, it was shown that the cloud velocity during gravity slumping f alls below the wind speed rather soon. From that point on, .the large clouds can only be obtained by assuming that atmospheric dispersion does not occur. Thic would, of course only happen if the air flow is laminar (i.e., no disturbance), a condition which is extremely unlikely. Hence, at the point in which the cloud speed falls to the level of the wind speed, the more realistic assumption can be made that atmospheric dispersion begins. The concentration downwind can then be obtained from the equation for a continuous line source 2 x-1 4 - 2 0 exp (- z
)
+erf(L/2+yj _
21
- L.ua z 2ag . erf[(L/2-v)
'y 'y . (29) 3 in which Q is the source strength in ft /sec, u the wind speed, L the width of the source which is taken to be that of the cloud spread across the wind at the point in time at which the cloud speed equals the wind speed.
For points along the wind direction, y = 0, and equation (29) for ground level reduces to X= T3 2' Q er f- L/2 (30) 32 Lu , z Q2a'y We have calculated from the gravity slumping model the time at which the cloud advancing velocity falls below the wind velocity. We have also cal-culated the location and the concentration (which is of course homogeneous). To determine the atmospheric dispersion of the cloud from that point on we [sv} 2.2A-101 Am. No. 59, (6/81)
ACNGS-PSAR resort to a line source having a length, L, located perpendicular to the wind
.-- direction, and located at a fictitious (virtual) distance, upwind from the po int at which the cloud and wind velocity have been compared to be equal, which is equal to that which would result in a centerline concentration equal to the homogeneous cloud concentration computed by the gravity slumping model. This is shown in Figure 1.13.
The location of the virtual line source is chosen upwind of the point at which the cloud and wind speed become equal, by solving the following equation
/ )
ii =.dR/dt = f O UL (tI = d[6I dt) er (31) where X ud is the computed concentration, in volume fr ac t ion , at the time kn=wbdb)the c wind velocity equals the cloud velocity. In equations (29), (30), and (31) the downwind distance does not cppear explicitly, but it is found by trial and error solution for a and a , the dispersion parameters which are functions of that distande. foe a Pasquill F meteor-logical condition, assumed throughout this study, values of the dispersion parameters a and a with downwind distance are taken for two cases. In 59 case A,a = 2 a, ; aEd in case B, a =5a.z y y I The very same equation (29) has also been used to back calculate the width of _ a source and its distance upwind that would produce concentrations of 2.4% of propanc at a point on the ground located downwind on a line bisecting the line source. The results for the Allens Creek plant are shown in Figure 1.14 for dif ferent wind speeds. Figure 1.14 can b used for other source strengths if it is remembered, that in equation (29) is a single parameter. Hence, the ef fect of a smaller source is like that of increasing velocity; i.e., a source of 100 lb/see with a wind of 2.6 fps will give the same results as a source of 20 lb/sec in a
- 0. 5 f ps wind .
Figure 1.14, when used for Allens Creek, and a break giving 100 lb/sec. of propane, is limited to cases in which the wind velocities are in excess of 4.3 fps. This is because the equilibrium cloud velocity for such a constant break flow, never falls below 4.3 fps. This is illustrated in Figure 1.4, where for wind speeds exactly matching the cloud advancing speed (i.e., no entrainment) the cloud is seen to advance to very large distances. For wind speed s below 4.3 f ps , it is assumed that no atmospheric dispersion occurs. For wind speeds equal to or in excess of 4.3 fps, it is assumed that gravity slumping ceases and atmospheric dispersion takes over. For breaks producing lower flow rates (such as for instance, a break result-ing in 20.5 lb/ sec . ), the speed of wind at which atmospheric dispersion can be assumed to begin is lower. As shown in Figure 1.12 for such a break, the
^
equilibrium cloud speed is about 2.6 fps. Hence, Figure 1.14 for such a break should be used for wind speeds in excess of 2.6 fps. Since Figure 1.14
- has been developed for a flow rate of 100 lb/sec. and the flow rate in ques-2.2A-10j Am. No. 59, (6/81) ,
ACN,S-PSAR [ } tion is only 20.5 lb/sec, direct use of Figure 10 for such a flow, requires the use of a corrected wind speed of = 100 2.6 = 12.7 fps. (f 20.5 Returning to the case of Allens Creek with a break resulting in 100 lb/sec. flow of propane, it is noted. that during gravity slumping, the speed of the cloud falls to or below a wind speed of 4.3 fps very soon (i.e., at distances from the break of less than 400-300 ft.). At such point the homogeneous cloud concentration is below 20-30%. A line source producing such concentra- ' tion would have to be located about 1500 f t upwind from that point, i.e, 1000 ft. further upwind of the break location. From Figure 1.14, therefore, one would compute that concentration of 2.4% of propane would be achieved about 3000 f t. fron, the virtual line source loca-tion (wind speed > 4.3 fps) for the case of a* = .2 aY, and hence, 4000 ft. from the break location. Thus, the resulting cloud would be less severe than that examined in Figure 1.3. 1.10 POTENTIAL MISSILES FROM DETONATION It is difficult to assess the potential hazard to plant critical structures 59 resulting from missiles generated by the detonation either at the initial crater or propelled by the blast wave. Since the LPG line crosses the plant vicinity in an open area, there is little likelihood that substantial missiles may be generated other than from the place where the detonation is i
) postulated to occur.
kj ' Work by Ahlers y / on observed maximum debrin distance and equivalent yield, which is reproduced in Figure 1.15 shows that the range of missiles from the l 390 tons detonation would be on the average of 6000 feet, with source having a range of up to 12,000 feet. Hence the detonation of the vapor clouds could result in missiles reaching the plant. The missiles which travel the larger distancea, however, are expected to be ; the smaller since air drag will af fect the larger missiles proportionally more. Studies on several detonations 24, 25 / have shown that the size dis-tribution of ejecta (missiles) follows an exponential law. ;
-3.5 8= kr where 4 is the araal density and r the distance from the detonation center.
To ascertain what the probability is of a missile from even the impossible maximtsn detonation of the propane (390 tons) hitting critical plant struc-tures, we assume that the total weight of missiles is proportional to the volume of the crater which would be created by the detonation, had it been a TNT detonation near the surface. The crater in turn is proportional to the detonation yield. Roughly the volume of the crater can be estimated by the scaling law. Diameter (depth) of crater = Diameter (depth) of crater xW 1/3
}
for a 1 KT explosion j d 2.2A-10k Am. No. 59, (6/81)
ACNGS-PSAR t
) and the knowledge that a 1 KT surf ace detonation in dry soil results in
' diameters of 180 feet and depths of 35 feet.
3 Hence the total mass of ejecta at 95 lb/ ft will be: 95 lb/ft3 [2r /3 (180 x .73)2/4] (35 x .73) ft3 = 2.19 x 107 lb Assuming none of the mass f alls back within the crater , then the total mass must be given by
-2.5 M= 2r r a dr = 2rn r dr = .00786n J
80 80
= 2.79 x 10 9
It-ft 3/2 Hence the areal density 7,000 feet away from the center of the hypothetical (impossible) 390 ton detonation is 3 = 2.79 x 10 lb-ft 3/2 9
= 6.09 x 10
-5 lb/ f t 2 4.58 x 10 I3 ft 7/2 Since the area of critical structures is of the order of 10 ft , the total weight of missiles hitting these areas for the hypothetical maximum
^
detonation would only be 6 lb. w# Assuming that all of this macs is concentrated in one missile, and that missile travels at the maximum air particle velocity given by u= = 53 fps 7P (1 + 6p/7P ) where p is the peak overpressurs at the critical structure (1 psi), P the ambient pressure (14.7 psi) and C the ambient sonic speed, taken as 1,130 f t/sec , then the impact energy of this missile is 26 f t-lb, which is considerably below the energy required for penetration of the structures. It is concluded therefore that missiles from the propane cloud detonation would present no hazards to the Allens Creek Nuclear Island. 1.11 COMPARISON OF MODELS WITH ACTUAL OBSERVATIONS a) Ruf f Creek (See Reference 26) The actual wind condition during the Ruff Creek event is not reported. It is deduced to have been less than 23 fps; since this agrees with the description given that the wind was calm. ,y From Figure 1.12, further it can be deduced that the cloud spread ! I should have been 300 by something less than 2500 feet if it occurred by gravity alone, which would agree well with the burn marks observed 2.2A-101 Am. No. 59, (6/81) I
ACN",S-PSAR to be roughly 100 yards uide by more than one mile long, since the deflagration would have resulted in an expansion to roughly six to eight times the voluue. lience burn marks would appear at twice the distance computed for the cloud. The steady cloud speed exceeds the wind speed so long as the wind speeds are below 2.3 fps, except right at the beginning of the event. 11 atmospheric dispersion is assumeo to start at that instance of time and gravity spreaoing is arrested, similar results are obtained. A calculation of the extent of the cloud with Figure 1.14 yielded a length of the cloud equal to approximately 2000 feet. b) bevers, Texas (See Reference 27) Atmospheric dispersion started at the point at which gravity spreading i velocities f all below the wind speed of 8 fps, yield downwind dis- 59 tances at 2800 feet, which when coupled with further lateral spreading of the cloud to the 2.4% concentration, give a cloud a dimension of 2600 feet by 1200 feet wide. The model overestima!.es the longer di-mension which was observed to be roughly 1100 feet. Pure gravity flow, coupled with wind translation (no dispersion), with suppression of spreading upwind and enhancement downwind, would result in a cloud of 1200 feet width and 2600 feet length. The upwind front part of the cloud, however, must be more diluted in reality than cal-culated as a result of increased entrainment. l[ }
%~.) c) Austin, Texas (See Reference 28) l l
Because the wind speed is always above the cloud gravity induced l velocity, atmospheric dispersion takes place right away. The atmos-j pheric dispersion model yields a cloud dimension of approxirhately 1 2000 feet by lE0 to 200 feet which compares well with the observed 2400 x 200 toot cloud. Gravity flow in a channel 200 feet wide with no atmospheric dispersion would yield a cloud length of 1300 feet. 1 l 2.0 (1hlS SEQ 10N IS 1hlth110hALLi DELETED) e s
- A) 2.2A-10m Am. No. 59, (6/81)
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). 3.0 BREAK OF 24" hAlbRAL CAS LlhE The consequences of a complete severance of the 24 inch natural gas line have been evaluated on the basis of the assumptions listed in the following sections.
3.1 CALLULAT10h 0F FLOW RATE 001 0F TiiE BREAK a) Double enued rupture of the 24 inch pipe. b) Gas escapes from both ends of the ruptured line, c) Stagnation pressure upstream of break in both ends is the maxinum operating pressure (rnt design) of the , line, ie, 750 psi. Pressure can vary up to 900 psi, but the analysis is not very sensitive to the pressure assumed. d) The discharge from the break is divided into: initial i transient when the flow from both broken ends may be choked, ano steady flow (due to continued pump operation). 5 cas Itne boperates J at a maximum flow rate'of 2.5 x 10 hCP/ day, ie 2.gx10 it (@ standard temperature and pressure) / day = 9.25 x 41 lb lb/ day. Q f G 312.5
= .037 lb/ft3 (e standard temp. and pressure) since #ch4 (See Figure 3.1)
Thus the flow in the unruptured line to deliver such quantity will be 9.25 x 10 6 lb/ cay
=
. x t ay = M f thsec 3
2.1 lb/ft where 2.1 lb/tt3 is the density in the line at 750 psi & 105 F (see Figure 3.1). 5 Thus to deliver 2.5 x 10 McF/ day, the gas is pumped through the pipeline at 750 psi (105 F) with a velocity of 16.25 ft/sec. To determine whether choked flow exits, the natural Eas is assumed to behave as an ideal gas, even though this assumption can lead to errors of nearly 50 percent in the estimation of densities in the conservative direction. l For an ideal gas, the critical pressure, at khich the mass flow rate in the break area cannot be increased regardless how low the back pressure is maae is given by: _2/ Og V 2.2A-11 Am. No. 59, (6/81)
~ , . _ _ _ . _ _ _ _ , . _ . _ _ _ _ _ . . - _ _ , - - , _ _ . ~
1
.g)
J ACNGS-PSAR
?
V . , k/(k-1) ! P* = ( 2 ) P where Po is the upstream g kF1 stagnation pressure lhe critical density is given by i I i 1 P* = ( ) El pg ; k+1 i 41 i 1hus for the stagnation pressure of 750 psi, choked flow will exist as q soon as the exit pressure is dropped below 410 psi. Thus choked flow 312.5 will exist at the break. The shock wave which is created at the exit plane at the instant of the break will travel back through the line
! (both ends) at sonic speed until the whole line to the pump stations has been decompressed. At this time the flow can be assumed t.o unchoke.
To calculate the maximum flow rate out of the break, the equation of con- l ] tinuity at the critical (choking) plane can be used, thus ]' h (Ib/sec) = pVA E, where p ,V , and A are the critical density, velocity and area re- , spectively.
; For an ideal gas l = 2kgc k1 g / (k+1) . anu A = A for Mach No = 1.0 (choked flow).
l ~ ' A 105 F, usin), k ce1.3 for natural gas, and Rcr60 ft-lbf, lbm - R V = 1,110 ft/sec Using the ideal gas law, the etagnation density' corresponding to 750 psi i and 105 F, is
%= Po = 750 1144) -
3.18 ,
-RT o 60 (565)
P* = 0.629 (3.16) = 2.0 hote that the actual density for methane Po= 2.1 thus P *is really in the neighborhooo of 1.35; hence the estimated maximum flow rate using ideal gas law will be conservative by more than 30 percent
- v 2.2A-12 Am. No. 59, (6/81)
.-. __ . _ _ _ _ _ _ _ _ _ . - . _ . _ _ _ . ~ - . _ . _ _ . . . _ , . _ _ . . _ . _ _ _ _ . . . _
ACNGS-PSAR W = (2.0) (3.14) (1.11 x 10 )3 = 6,910 lb/see v The maximum mass flow rate can also be evaluated from l_/ E* - E k I 2- P A c
-R lk+1
( k-1 o o Substitution of the proper values would also yield 6,910 lb/sec. If true densities had been used, the maximum flow rate would have been 41 roughly 4,800 lb/sec. This last figure is obtained by using a critical density of 1.37 instead of 2.0 in the continuity equation. (Initial flow Q 312.5 would be higher for 900 pais pressure by roughly 20 percent.) Assuming that the break occurs midway between p' imping seations, and, that pumping stations are separaced by 65 miles, the time required for complete decompression of both ends of the ruptured line is calculated as: 32.5 mi (5280 f t/mi) a ~36 see m 2.5 minutes 1,100 ft/sec Thus for the firar 2.5 minutes following the' breek it may be assumed that both ends discharge natural gas at a rate of 4800 lb/see for a total of , j
.) 9,600 lb/sec. Thereafter only the pumped end will discharge (barring pump trips) at a rate of roughly 108 lb/sec.
In reality the upstream pressure will decrease with time with a correspond-ing decreasc in the flow rate out of the break. Further because of friction ef fects, the maximum flow rate out of the break will be lower, hence the model assumed is known to be conservative. The following figure compares the model used with what is expected in reality. 3.2 CALCULATION OF DETONABLE CLOUD SIZE The dimension of the detonable plume downwind of the break hdve been , evaluated for a Category F stability, and a constant, invariant wind speed of 2.6 ft/sec. For purposes of conservation buoyancy ef fects have been ignored in this section. Buoyancy and the jet momentum are concidered in Subsection 3.5. The centerline (directly downwind) concentration of the methane (excluding buoyancy) is determined by:
'E ,, a u 1 \
2.2A-13 Am. No. 59, (6/81)
ACN':S-PSAR Off-centerline concentrations are determined by 41
-s Q l \
312.5 y X = X cl Exp - [(y/ g + (z/a )2) g Since the flow out of the break varies with time it is necessary to discuss whi'h flow out of the break is chosen for subsequent analysis. During the initial 2.5 minutes of the event, flow out of the break is essentially sonic. Although a large porticn of the total mass of gas is released during this time period, (i.e. 70 percent of the total 56 quantity of gas emitted from the break in the 2.5 hrs. estimated to be required to termiccte flow is released during the initial decompression perioo), the velocity at the break, coupled with the buoyancy of the gas will propel it away from the surtace and past low atmospheric inversions so that less than 500 lbs. of this initial mass is calculated to fall within the flammable limits in the vicinity of the surface, but most of it will be dispersed in the upper levels. To establish therefore the configuration of a credible, low lying detonable , cloud in the vicinity of the plant (ie, cloud travelling toward the plant under worst meteorological conditions), it is assumed that the flow out of the break is a constant 108 lb/sec, corresponding to the condition existing after the initial 2.5 minutes transient. Suoyancy was neglected in this calculation. 41 Q g the potential cloud contiguration (neglecting buoyancy) is plutted in 312.5 ( } Figure 3.3a. The dashed portion represent the fraction of the cloud which \d f alls within the flammable limits (4.8 and 14.0 volume percent). The volume of this cloud is calculated ta be: Volume of flammable cloud = (Volume of ellipsoid at 4.8% = 2r/3 (2200)(106)(53)] - [ Volume of ellipsoid at 14.0% = 2r/3 (1150)(62)(30)] = 7 3
= 2.11 x 10 ft Density of 10% mixture = .069 lbm/ft height of cloud (assumed to be all at 101) = 1.46 x 106 lbm 3.3 CALCOLAT10h 0F EFFECIS FROM DEIchATION TO CATEGORY I STRUCT0FES From an enthalpy of detonation release of 1225 BIU = 310 kcal of 10 lbm Ibm 7 3 methane and mixture, and a total volume of 2.11 x 10 ft , it is possi-ble to compute the total energy released in a hypothetical detonation of the entire detonable cloud, assuming that the whole cloud is at a mixture averaging 10 pet ent.
O
? I U
2.2A-14 Am. No. 59, (6/81) i P
-- .n--.
ACNGS-PSAR 7 3 +;[m}
~
2.11 x 10 ft U 3 6 ft = 1.48 x 10 lbm [(.9) (12.4) + (.1)(27.0)J lb Total energy released = (1.48 x 10 0 lbm) (310 Kcal) = 4.60 x 100 Kcal ibm Equivalent TNT = 4.60 x 108 Kcal .90 x 106lb TNT = 450 tons TNT
=
41 500 kcal = .45 KT of TNT Q lb TNT 312.5 The preceding assumes that the detonation yield will be 100 percent. In actuality this will not be the case. Reference 2 cites yields of 7.5 per-cent. Work by Iotti, et al, shows that indeed the yield of a gaseous de' anation is lower than that of TNT. Reference 6 compares overpressures es.iculated by assuming gaseous point sources (Reference 9) to overpressures obtctned by Kingery (Reference 8) for the same yield, and those measured by Kogarko (Reference 7), et al, for the given gaseous detonation. This comparison shows that the Ktegery's results would have been comparable to those of References 7 and 9 if a TNT yield of less than 50 percent had been employed. Thus a conservative estimate of the TNT equivalent of the detonation of the entire cloud can be obtained by using 50 percent yield and K1ngery charts.
)
V A yield of 50 percent is conservatively used in computing all the blast wave, and seismic parameters at the plant side. These are tabulsted in the following table for distances of 3,600 feet, 5,400 feet, and 10,600 teet trom the plant critical structures to the closest point of the deton-able cloud (which is conservatively assumed to be the center of the detonation). The breaks are Sasumed to occur at the point in the line closest to Category I plant structures (9,300 feet for the relocated line). Assuming conservatively that plant critical structures can withstand 1.0 psi overpressure the closest distance between the 24 inch line and any of those structures should be approximately 7,200 feet. 3.4 POIENTIAL MISSILES FROM DETONATION The potential hazaros of plant critical structures resulting from missiles generated by the natural gas detonation have been evaluated. From Ahlers work (Figure 1.15) one can argue that missiles could reach the plant struc-tures, since the f arthest range for the 225 ton of TNT detonation is shown to be of the order of 11,000 feet. Moreover, again assuming that no mass falls within the expected crater (97tt.giameterand18.2feetdepth), the total mass of the ejecta, 8.5 x 10 lbm, is distributed about the crater in an areal density given j by; l 8 -3.5 2 6 = 6.9 x 10 r 1b/ft g 2.2A-15 Am. No. 59, (6/81)
- -__ - . - - \
ACNGS-PSAR j* } Hence at the Ib,600 feet distance, the areal density is expe=ted to be 6 4= 6.9 x 10 0 = 6.9 x 10 =
-6 (1.06 x'16"#*D) 1.22 x 10 14 51.62 x 10 lb/ft 2
4 Q For a distange of Sy400 feet, the corresponding areal density is expected to be 6 x 10 ~ lb/ft 312.5 4 hence for a critical plat area of 105 gg2, the total mass hitting the plant is less than 6 lbs, which even if concentrated all in one missile travellink at the air particle speed, calculated to be 24.6 fps, cannot damage the structures. 3.5 EFFECTS OF BLOYAhC) AhD M0hENTLM Since methane is lighter than air, the methane-air mixture formed from the escape of methane f rom the break will rise with respect to the surrounding air or denser mixture in accordance to Archimedes law. In the previous sections tuoyancy has been neglected for the sake of conservatism. Also neglected has beers the large momentum of the methane gas out of the break, 1 which would result in a high plume. j In spite of neglecting these effects it has been shown that the plant will withstand the hypothetical detonation of the ensuing cloud without deleter-fs ious ettects to Category I structures. i (N 4.0 BREAK Ih ThE 8-lhCH CKUDE LINE 1he consequences of a break in the crude oil lines may be conservatively 41 assumed by comparison with the consequences of a break in the 6 inch LPG 9 line which is identically routed. 312.4 The major dif ference between the crude oil lines and the LPG line stem from the lower design and operating pressure of the crude line (720 psi vs 1,000 psi cesign and 560 psi vs 750 psi operating). The size of the line (6" vs the 66" LPG), and operating flow (12,000 barrels / day vs 8,500 barrels / day). An identical time is required to shut of f flow to either pipe line af ter a break is identified. Even thouhh proportionately more fluid can escape from the break of the crude oil line than the LPG line, a smaller f raction of the crude vaporizes to give rise to potentially detonable clouds. Therefore the consequences of a break in the crude oil line will not have an adverse effect on the pisnt's Category I structures. , . 1 4 i
N c) 2.2A-16 Am. No. 59, (6/81) a i
- - - - - - - , -, -n - - . - - , - , , - - - - , , - - r.m -
y----- ,--. - - l
O O ACNGS-PSAR e TA3LE 1-1 3 DIhdNSION OF PROPANE PLINE D0'.INVIND OF 1000 FT /SEC S01'RCE, (6 INCH LPC LINE) STABILITY CATECORY F. WIND SPEED 2.6 FT/SEC Harizontal* Vertical
- Distance f rom Distance from Doenwind Plume Dispersion Centerline Plume Axis, ft Distance (ft) (Std Deviation) f t Concentration Plume Axis. ft.
To rich 'imit To lean limit To"~r ich limit To lean limit yrd yrf yld ylf zr1 zrf zId alf y : % _ 59 70.4 77.86 12.0 11 4 14.0 14.6 27 5.4 83.9 60.2 37 17.8 150 47.2 70.4 65 85.6 89.14 14.0 13.0 17. 't 1000 36 7.2 10.5 24.4 26.4 73.6 53 172.2 131 2 14.7 2000 72 14.4 11.8 0 25.2 27.7 9.0 58.5 0 126 133.5 11 7 2170 82.5 16.5 7.06 -- 25.4 28.3 18.0 7.56 35.3 -- 126.8 141.8 2500 90 0 -- 25.3 ?3.4 18.7 7.0 0 -- 126.5 142.2 2575 93.5 -- 25.0 28.6 20 6.12 -- -- 125 143.6 -- 3000 100 -- 21.4 27.2 24.4 4.11 -- - 106.9 136.4 -- 3500 122 18.4 26.6 27.7 3.5 -- - 92.2 133.0 -- -- 4090 138 -- 0 20.5 29.6 2.8 -" - G 102.6 -- 4290 147.8 -- -- -- 7.0 2 . 25 -- -- -- 35 5000 165 13 - - 0 2.2 - - -- 0 - F 5016 If ' .8 33.4 - - - - - - 5500 181.5 36.3 1.86 - W U 59 rf - rich deflagration limit
- rd - rich detonable limit if - lean deflagration limit Id - lean detonable limit Am. No. 59 (f. A1)
.. . - . - - . - - . - - . _ - - . - . . . . . . . . - . _ - - - . . . - - . . - . - - ~ . . . . _ _ _ _ . _ _ - . . . . . . , _ . .
t i e
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.i I f.. i f i i i i i i \ i I t i l PAGES 2.2A- 18 THRU 20 HAVE BEEN INTENTIONALLY 59 k l DELETED i e ll l l '! I j , 1 . i h I I . l i f T l i i I i ( 2.2A-18 Thru 2.2A-20 Ara. No. 59, (6/81) i i
ACNGS-PSAR
5.0 REFERENCES
[ ))
\
N ~' 1/ T V Eichler, E S Napadensky, " Accidental Vapor Phase Explosions on Trans-poration Routes Near Nuclear Power Plants", NUREG/CR-0075, April 1977. . 2,/ _ ASHRAE Handbook of Fundamentals, ASHRAE, New York, N Y 1967 3/ Turner " Workbook of Atmospheric Dispersion Estimates" Public Health Service Pu b. No. 999- AP-26, 1969 4/ Iotti R, Krotiuk and D R deBoisblanc " Hazards to Nuclear Plants from On (or Near) Site Gaseous Explosions", CONF 730?04, USAEC,1973 5/ H L Brode " Numerical Solutions of Spherical Blast Waves" J. App. P5 s. 26,, 766 (1954)' 6/ Kingery D " Air Blast Parameters Vs Distance for Hemispgerical TNT Bursts" BRL Report No.1344. Ballistic Research Laboratories Aberdeen Proving Gro und , Ma ry land ,19 66 7/ Kogarko, Adushkin and Lyamin, " Investigation of Spherical Detonations of Gas Mixtures" Combustion, Explosion and Shock Waves, l., No . 2, 19 6 5 59 " p p 2 2-34
~8/ D S Burgess & M C Za betakis, "De tonation of a Flammable Cloud Following a Propane Pipeline Break" Bureau of Mines Report of Investigation 7752, 1973 ,' N
(%- /) 9/ Lofquist, K, " Flow and Stress Near an Interf ace between Stratified . Liquid s", the Physics of ' Fluid s, 3,, P. 158, 1960. I
~--10/
Ellisen, T H, and J Turner, " Turbulent Entrainment in Stratified Flow", Journal of Fluid Mechanics, 195 9, p. 45 3. f 11/ Britter, R E, "The Spread of a Negatively Buoyant Plume in a Calm Environment", Atmospheric Environment, H,1979, p.1241-1247. J2/ M Bresse, "Cours De Mecanique Appliquee", Mallet-Bachelier, Paris (1860) . 13/ Abramovitch, G N "The Theory of Turbulent Jets", MIT Press Cambridge, ! Mass. 1963 , 14/ Smith, M (ed.) Prediction of the Dispersion of Airborne Ef fluents ASME Committee on Air Pollution Controls, c.1968 15/ Sutton, O G Micrometeorology, McGraw, c. 1953 16/ Courant and Friedricks " Supersonic Flow and Shock Waves" Interscience Publishers, New York, New York. i j i 4
%m,/
2.2A-23 Am. No. 59, (6/81)
a ACNGS-PSAR l p\ REFERENCES (Cont'd)
'u, M/ J H Ime et al, " Blast Ef fects from Vapor Cloud Explosions" McGill University, Montreal, Canada.
M/ Lewis and Von Elbe, " Combustion Flames and Explosions of Cases", Academic Press, New York, New York , M/ Lind, C D and R A Strehlow, Fourth International Symposium on Transport J of Hazardous Cargoes by Sea and Island Waterways, Jacksonville, Florida, 1975. , M/ Kogarko, Adushkin and Lyamin, " Investigation of Spherical Detonations in , Cas Mixtures" Combustion, Explosions and Shock Waves,1, No. 2,1965, l pp. 22-34. j 3/ SB Report PAR-78-1 " Pipeline Accident Report - Ruf f Creek, Pennsylvania . July 1977. i M/ Ahlers, E B " Debris Hazards, A Fundamental Study" DASA 1362, Illinois l Institute of Technology Research Institute, 011cago, Illinois t 23/ Ahlers, E B " Debris Hazards, A Fundamental Study" DASA 1362, Illinois i Institute of Technology Research Institute, clicago, Illinois 59 M/ Kaplan, K, Sears, P and Melichar, J "The Hazards to the Brunswick Steam [g} Electric Power Plant caused by an Explosion near the Sunny Point Munitions Terminal, Section D, Live and Inert Missiles" URS Report 700-1, () dated May 1968 M/ H L Brode "'Ihe Hazards to the Brunswick Steam Electric Power Plant caused by an Explosion Near the Sunny Point Munitions Terminal, Section D , Missiles from Accidental Explosion - Distant liazards." i
~
26/ NTSB Report PAR-78-1 " Pipeline Accident Report - Ruf f Creek, Pennsylvania" July 1977. , M / NTSB Report PAR-76-5, " Pipeline Accident Report - Near Devers", Texas, May 1975
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28/ NTSB Report SS-P-17 " Pipeline Accident Report - Austin, Texas", February 22, 1973 i t.m] 2.2A-24 Am. No. 59, (6/81)
ACNGS - PSAR O
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30 __ _~~~----.--.. . O I I 480 995 3745 TIME (SEC) AM NO.59, (6/81) HOUSTON LIGHTING & POWER COMPANY Allens Creek Huclear Generating Station Unit 1 FLOW RATE ISSUEING FROM THE BREAK VS TIME , FIGURE 1.2
. . _ _ - - - . . -. . __ . _~ .. .- .. :
ACNGS - PS AR
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) Unit 1 DETONABLE & DEFLAGRABLE CLOUD FORMATION FOR 6 INCH g
LPG LINE BREAK l FIGURE 1.3 ,
ACNGS PSAR 30,000 - SOURCE 100 LBM/SEC CHANNEL WIDTH 300 FEET SLOPE. 0006
, PEAK 78,500' AT APPROXIMATE 5 HOURS p ( ) TIME AT WHICH 2.4%
S CONCENTRATION OR 2.2% o CONCENTRATION IS REACHED ! 20.000 - (SECONDS) o ! m ; CONSTANT SOURCE g E _ ._._ VARIABLE SOURCE I y TO 2.2% o _ __ VARIABLE SOURCE l 1 TO 2.4% I . s
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- 2. INITIAL 100 LBM/SEC SOURCE DECAYING TO 30 LBM/SEC AT 995 SECONDS j AMJO. 59. (6/81) [
HOUSTON LIGHTING & POWER COMPANY I Allens Creek Nuclear Generating Station l [ T Unit 1 l FARTHEST DOWN CHANNEL REACH & HGT . j OF GRAVITY SLUMPING CLOUD AS A ! FUNCitON OF WIND VELOCITY I FIGURE 1.4 l
ACNGS - PSAR CHANNEL WIDTH = 300' ,A 9500 - (1657) SLOPE = .0006 (1280) ( ) o 8500 - TIME AT WHICH (1170)
$ 2.2% CONCENTRATION g IS REACHED (SEC) (1117) o E
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ACNGS - PSAR
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)
v ALLENS CREEK CHANNEL WITH A 140 FOOT ISOCLINti f.'GURE 1.6
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l SPATIAL FLAME VELOCITY (m/s) l AM. NO. 59, (6/81) l l HOUSTON LIGHTING & POWER COMPANY ! I Allens Creek Nuclear Generating Station ! Unit I f SPATIAL FLAME VELOCITY VS. OVER PR ESSUR E ! [ FIGURE 1.11 i~ !
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- HOUSTON LIGHTlHG & POWER COMPANY '
h Allens Creek Huclear Generating Station I'
) J Unit 1
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FIGURE 2.2 2 3
? I: ACNGS-PSAR i l LIST OF EFFECTIVE PAGES l CHAPTER 3 I I DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS > P.tge Amendment 1* 59 2* 58 3* 56 i' 4* 59 5* 59 6* 59
- 7* 59 h j 8* 59 8a 59 I
9 59 10* 56 11* 57 12* 56 12a 56 13* 59 l
- 14* 59 i 15* 59 16* 59 17* 59 18* 57 i
35 11 35
- iii 35 iv 35 .
v 35 ; t vt 35 vii 35 viii 35 , ix 35 x 35 xt 35 xii 35 i 1 xiii 37 , xiv 35
- xv 35
! xvi 44 l xvii 44 xviii 44 . xix 48 l l xx 35 ; , xxi 44 l 3 ' xxii 35 l xxiii 35 xxiv 35 xxv 35
- Effective Pages/ Figures Listing 1
1 Am. No. 59, (6/81)
~ . - _ _ . - . .. . - _ . - - _ - - - . . . - . . ----. -...- - -. - - . - . . - . - . . . . - . . ,
I i ACNGS-PSAR r O LIST OF EFFECTIVE PAGES (Cont'd) CHAPTER 3 Page Amendment ' 3.2-16 35 3.2-17 35 3.2-18 7 ! 3.2-19 35 3.2-20 - 3.2-21 35 l 3.2-22 35 3.2-23 35 3.2-24 57 3.2-25 57 3.2-25a 30 3.2-26 59 3.2-27 22 3.2-28 .53 3.2-28a 51 3.2-29 - 3.2-30 35 3.2-31 56 3.2-32 22 - 3.2-33 22 j 3.2-34 46 3.2-34a 46 3.2-34b 32 3.2-35 - 3.2-36 - 3.2-37 - 3.2-38 - 3.2-39 - I 3.2-40 - 3.2-41 21 3.2-42 21 3.2-43 7 3.2-44 56 3.2-45 56 3.2-46 10 3.2-47 11 3.2-48 10 3.2-49 35 3.2-50 11 1 3.2-51 37 , 3.2-52 47 i
- G 4 Am. No. 59, (6/81)
, . - , . - - , - - _ - - - - - - - , , , , - . , , , - - _ - , - . . , - - . , - - - - - - , - , _ - - - , - - - - , _ . - - . . . ~ , . . . _ - - - , - . . _ _ - , - , , , - , - . - . . , . -- - , . - . . - , . .
a ACNCS-PSAR r i i a EFFECTIVE PAGES LISTING (Cont'd) , [ b I l\ CHAPTER 3 Page Amendment No. I I 3.3-1 35 !. 3.3-2 35 l l 3.3-3 48 I I 3.3-4 35 [ 3.3-5 35 - 4 3.3-6 35 ! 3.3-7 35 -, 3.3-8 35 } l l 3.4-1 45 l 4 3.4-la 35 3.4-2 35 , 3.4-3 35 l 3.4-4 37 ! 3.4-5 35 i 1.5-1
- 35 }
3.5-2 48 [ 3.5-2a 35 i 3.5-3 58
; 3.5-4 35 '
I 3.5-5 58 l ,! 3.5-Sa 58 i . 3.5-6 37 i ! 3.5-7 35 I I 3.5-8 35 3.5-9 58 l 3.5-10 58 3.5-11 58 ' 3.5-11a 59 3.5-12 59 3.5-13 37 : j 3.5-14 35 l 3.5-15 35 l 3.5-15a 35 >
- 3.5-15b 44 l
- 3.5-16 49 l l 3.5-16a 58
! 3.5-17 39 3.5-18 58 1 3.5-19 58 : 3.5-20 58 f 3.5-21 35 l i i i I e , i 4 i i 5 Am. No. 59, (6/81)
i I i- i ,. l l ! ACNCS-PSAR i 2 l EFFECTIVE PAGES LISTING -(Cont'd) ! I' CHAPTER 3 j Page Amendment No. 1 3.6-1 35 3.6-2 35 ! 3.6-3 35 i .3.6-3a 35 l l 3.6-3b 35 i ! 3.6-3c 35 : 3.6-3d 35 3.6-3e 35 3.6-3f 37 , l 3.6-4 37 , 3.6-5 41 ! , 3.6-5(i) 37 3.6-Sa 47 3.6-5b 41 l , 3.6-5b(i) 48 i 4 3.6-Se 48 l 3.6-5d 41 j 3 3.6-6 37 : ! 3.6-7 37 l 1 3.6-8 35 ! ! 3.6-9 35 i . 3.6-10 35 i ! 3.6-11 35 i ! 3.6-12 35 l 3.6-13 35 I . 3.6-14 35 i 4 3.6-15 35 ! j 3.6-16 35 3.6-17 35 i 3.6-18 35 l
-3.6-19 35 l
, 3.6-20 35 l l 3.6-21 35 l l 3.6-22 35 5 3.6-23 35 ! 3.6-24 35 ! 3.6-25 35 ! 3.6 35 ! 3.6-27 35 3.6-28 35 3.6-29 35 l 3.6-30 35 3.6-31 35 3.6-32 35 l i { 6 Am. No. 59, (6/81) l t _ ..___ _-- , __ _ . _ _ _ ._. ~ ._,. , _ . _ ..__ _ _. - -.._.- _ _... -..,_,_._ , ~ ..,_ .
. .. . -. .-. - ~ . . . . . . - - - . -- -. - - . . . . - . . . . . . - . . _ . . _ .
p ACNGS-PSAR-i EFFECTIVE PAGES LISTING (Cont'd)
'\ .
CHAPTER 3 i 3 Pege Amendment No. 4. 3.6-33 35 L 3.6-34 35
- 3.6-35_ 35 3.6-36 35
- 3.6-37 35 3.6-38 35 ;
i 3.6-39 35
!- 3.6-40 35
- 3.6-41 35
- - -3.6-42 35 i 3.(,-4 3 35 ;
3.6-44 35 l 3.6-45 -35 ; 3.6-46 35 l
?
1
. 3.7-1 49 ;
I 3.7-2 35 ; 3.7-3 44 4 3.7-4 44
; 3.7-4a 44 3.7-5 35 ! 3.7-6 35 ; ; 3.7-7 (deleted) 35
, 3.7-7a ~(deleted) 35 i j 3.7-8 (deleted) - 35 i 3.7-9 (deleted) 35 e i 3.7-10 (deleted) 35 l' I 3.7-11 (deleted) 35 3.7-12 (deleted) 35 } 3.7-13 (deleted) 35 t 3.7-13a (deleted) 35 i 3.7-14 (deleted) 35 l !, 3.7-15 35 ! 3.7-15a 35 ? i 3.7-15b 35 ! I j 3.7-15e 35 3.7-15d 35 3.7-15e 35 i l 3.7-15f ~ 35 l
- 3.7-15g 35 3.7-15h 35 4
3.7-151 44 , 4
.3.7-151(1) 44 !
3.7-15j 35 : l 3.7.15k 35 ! l: 3.7-151 35 I 3.7-16 35 -
-l r
i !
}
I 7 Am. .No. 59, (6/81) ! l !
'ACNGS-PSAR EFFECTIVE PAGES LISTING (Cont'd)
CHAPTER 3 ' . i g odtent No. Page l' 3.7-17 35 l 3.7-18 35 : 3.7-19 35 t 44 i 3.7-20 3.7-20a 44 3.7-21 35 ! 3.7-22 44 44 i 3.7-22a 3.7-22b 35 I 3.7-23 44 l 3.7-24 44 l 44 .
- 3.7-24a 3.7-25 35 3.7-26 35 3.7-27 35
] i 3.7-28 35 i
- 3. 7- 2 8a 42 1
j 3.7-28b 42 3.7-28c 42
- 3. 7-2 8d 42 3.7-29 35 3.7-30 44 i t
- j 3.7-31 35 i- 3.7-32 35 I 3.7-3 2a 35 :
, 3. 7-3 2b 35 ! 3.7-32c 44 _j ! 3. 7-3 2d 44 l 3.7-33 (deleted) 37 i l i 3.7-34 (deleted) 37 l 3.7-34a-(deleted) 37 1 l !
? 3.7-35 (deleted) 37 3.7.A-1 50 l
! .3.7.A-2 49 ! ! 3.7.A-3 49 l 3.7.A-4 48 i l 3.7.A-5 54 l' 1 3.7.A-6 48 I 3.7.A-7 48 i I ! 3.7.A-8 48 j 3.7.A-9 48 { 3 3.7.A-10 48 i 3.7.A-11 48 l i I
- i. !
I i- 8 Am. No. 59, (6/81) ; i L
i i, i ACNCS-PSAR ( l EFFECTIVE PAGES LISTING (Cont'd) i f I CHAPTER 3 1- ; i - Page Anendraent No. ; f 3.7.A-12 - 48 l ! 3.7.A-13 48 l l 3.7.A-14 48 : ! 3.7.A-15 48 l l 3.7.A-16 48 i l 3.7.A-17. 48 l 3.7.A-18 48 I ! 3.8-1 54 l 3.8-2 54 ! i 3.8-3 41 !
- 3 . 8-4 57 !
l 3.8-4oa 57 l j 3.8-4a 35 i j 3.8-4b 35 ! j 3.8-4c 35 3.8-4d 35 3.8-4e ~15 i 3.8-4f 35 ! 3.8-4g 39 l 3.8-4h 35 ! 3 3.8-41 35 ! 3.8-4j 54 l j 3. 8-4 k 54 i 3.8-5 35 E r i 5 i k 1 . i i i ' t 4 i i l 1 a I l l I i i i ! 8a Am..No. 59, (6/81) ? l
., mar.-i----m_. -,wwv--~- -----.e--.,v----->--n.- -- - - + -n-'-,s--
^
j ACNCS-PSAR { l
,, LIST OF EFFECTIVE PAGES (Cont'd)
CHAPTER 3 l l i i ] Page Amendment l ! t 3.8-6 54 ? 3.8-7 54 i, 3.8-7 a 59 3.8-8 35 . 3 . 8 -9 44 ! I 3.8-9a 59 l l 3.8-9b 59 . l i 3.8-9c 57 i 3.8-10 59 : 3.8-11 54 l 2 3.8-12 54 i 3.8-13 54 l 3.8-13a 35 i i '3.8-14 35 l 3.8-14a 44 j 3.8-14b 44 , .3.8-15 44 3.8-16 35 3.8-17 35 l 3.8-18 54 l j 3.8-19 35 f
- 3.8-20 46 !
3.8-20s 46 I
- 3.8-20b 46 l 3.8-21 56 3.8-22 35 j 3.8-2 2a 35
- 3. 8-2 2 b 35
- 3.8-2 2c - 35 l 3.8-23 54 i
! 3.8-24 44-
, 3.8-24a 54
} 3.8-25 54 i 3.8-26 54 l 3.8-2 6a 54 i '3.8-27 35
- 3.8-28 35 l 3.8-28a 35
! 3.8-2 8b - 54 i 3.8-29 54 i 3.8-2 9a 54
! 3.8-29a(1) 54 1 i~ 3.8-2co 35 l -3.8-30 35 i 3.8-31 35 r l .3.8-31a 35 i 3.8-31 b 35
[ $ 3.8-32 35 3.8-33 35 i 9 Am. No. 59 -(6/81) l i ! l f l
d ACNGS-PSAR EFFECTIVE PAGES LISTING (Cont'd) CHAPTER 3 Page knendinent No. 3.9.A-1 35 . 3.9.A-2 35 3.9.A-2a 35 I. 3.9.A-3 21 3.9. A-4 21
- 3. 9 . A-5 35 f
3.9.A-6 - 21 3 . 9 .A-7 41 3.9. A-8 41 3.9.A-9 41 i 1 3.9.A-10 35 ! 3.9.A-10a 35
- 3. 9. A-11 35
. 3. 9. A-11a 35 3.9.A-12 35 3.9.A-12a 35 3.9. A-13 35 3.9.A-14 21 3.9.A-15 21 3.9.B-1 42 3.9.b-la 42 3.9.B-2 3 j 3. 9. B-3 3 1 3.9.B-4 42 3 3.9.B-5 35 3.9.B-Sa 35
- 3. 9. Br6 3 4 3.9.B-7 17
- 3. 9. B-8 41
! 3.9.B-8a 41
- 3. 9. B-9 ,
42 , 3.9.B-10 42
- 3. 9. B-10a 42 3.9.B-11 17
- 3. 9. C -1 41 3.10-1 41 3.10-1a 47 3.10-2 47 ;
I 3.10-2a 35 i 3.10-2b 35 3.10-3 35 s + I 13 Am. No. 59, (6/81) 1
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t i t r ACNGS-PSAR i I i l EFFECTIVE PAGES LISTING (Cont'd) i i t.itAPTER 3 i Pg Amendment No. l ! 3.10-4 - 4 3.10-5 35 4 3.10-6 39 i j 3.11-1 35 l ! 3.11-l a 35 l 3.11-2 35 1 3.11-3 44 3.11-4 35 8 3.11-5 35 ! 3.11-6 39 l
- 1. 3.11-7 35 i 3.11-7a 39 j
! 3.11-7 b 35 i l 3.11-8 35 1 3.11-9 35 ! 3.11-10 35 l 3.11-11 35 ! ! 3.11-11a 35 I 3.11-12 - 3.11-13 -
)'
i 3.11-14 45 I i 3.11-15 45 j 3.11-16 45 ! < 3.11-17 45 ! ! 3.11-18 48 l } 3.11-19 35 } l 3.11-20 35 l ! 3.11-21 - ! 3.11-22 35 , I 3.11-23 -
- 3.11-24 . -
l .s 11-25 - i 3.11-26 - l 3.11-27 17
- i l
i l 1 . : 4 l I i 1 14 Am. No. 59, (6/81) J + 1 i l
ACNGS-PSAR ; EFFECTI\'E FIGURES LISTINC* [ CHAPTER 3 I Figure Amendment No. t 3.2-1 3.2-2 26 i ! 3.4-1 13 3.4-2 37 I i 3.4-3 37 ! I 3.4-4 37 ! 3.4-5 (De leted) 37 j 3.4-6 37 i 3.5-1 37 I 3.5-2 58 i 3.5-3 1 3.5-4 1 '
- I 3.6-1 35 !
! 3.6-2 35 I 3.6-3a (Same on both page) 35 [ . 3. 6-3 b 35 !
- 3. 6-4 35 !
- 3.6-5 35 !
3.6-6 35 l ! 3.6-7 35 l I j 3.6-8 35
- 3. 6-8 a -
35 4 3.6-9 35 ) 3.6-10 35 i 3 6-11 35 , 3.6-12 42 ! j' 3.6-13 42 l
- 5 j 3.7-1 35 i
3.7-2 35 . 3.7-3 35 f i 3.7-4 35 I l 3.7-5 35 3.7-6 35 3 3. 7-7 35 i 3.7-8 35 3.7-9 35 .i 3.7-10 35 i
- 3.7-11 35 !
l ~ 3.7-12 35 ! 3.7-13 35 i ! 3.7-14 35 i ! l !
- All Figures whether labelled " Unit 1" or " Units 1 & 2" are to be considered ,
l applicable to Unit No. 1. , i f i 15 Am. No. 59, (6/81)
----.------e --. s- .--, -. , . , , ,,.-m. ,-,-,,,w-.--. ., , - e. . -, e.,e-ea- -w-,-v*e e ,e- v w--+wwe,-w vww-- er p we-w w ~m-~ ~ r v iv--- w--+e- &w.g , w y -y e--.
i ACNGS-PSAR i
) EFFECTIVE FIGURES LISTING (Cont'd)
CH APTER 3 s Figure Amendaent No. 3.7-15 35 3.7-16 35 3.7-17 44 3.7-17a 44 l 3.~-18 35 ! 3.7-14( Amendment No. not shown on page) 35 ! 3.7-20 35 3.7-21 35 l 3.7-22 35 ! 3.7-23 35 l 3.7-24 35 1 3.7-25 35 i 3.7-26 35 I 3.7-27 35 i 3.7-28 35 1 3.7-29 35 3.7-30 35 i 3.7-31 35 l 3.7-32 35 i i 3.7A-1 48 i 3.7A-2 48
- 3. 7A-3 48 3.7A-4 48 l 3. 7A-5 48 3.7A-6 48
- 3. 7A-7 48 l 3.7A-8 48
- r. 3. 7A-9 48 3.7A-10 48
! 3.7A-11 48 i 3.7A-12 48 1 3.7A-13 48 3.7A-14 48 3.7A-15 48 3.7A-16 48 3.7A-17 48 3.7A-18 48 3.7A-19 48 3.7A-20 48 3.7A-21 50 3.7A-22 50 3.7A-23 50 3.7A-24 50 3.7A-25 50
\
- All Figures whether labelled " Unit 1" or " Units 1 & 2" are to be considered applicable to Unit No. 1.
16 Am No. 59, (6/81) l
- i i i 3 ACNCS-PSAR l j .
I ; i j EFFECTIVE FIGURES LISTING (Cont'd) [ I CHAPTER 3 i I l Figure Amendment No. t , 3.7A-26 50 l 3.7A-2 7 48 4 3.7A-28 48' ! i 3. 7A-29 48 , ! 3.7A-30 48 i ! 3.7A-31 48 1 3. 7 A-3 2- 48 l 3.7A-33 48 , i 3.7A-34 48 - t !
- 3.8-1 54 l 3.8-2 54 3.8-3 54
- 3.84 26 ,
2 3.8-4a 30
- " 3. 8-4 b 30 I 3.8-4c 30
! 3. 8-4 d 30 I I I !6 1 I i i i ! 4 i
- f I
i 4 I i I i j i i I i
- All Figures whether labelled " Unit 1" or " Units 1 & 2" are to be considered
- - applicable to Unit No.1.
I ! . 17 Am. No. 59, (6/81) i
- - _ _ . . - . _ . - . _ _ . , _ - _ , . ~ _ . . _ _ _ . , _,_. _ _ . - , - . . . - _ _ _ _ _ . _ . . . , _ . . , . . - , _ . , - - , -
O O O ACNCS-PSAR TABLE 3.2-1 (Cont'd) ENVIRONMENTAL CAPABILIIY Scope AEC( quality g,) of Safety
- Quality Component
- Seismic (O Extreme (E Tornado / Flood (" Assurance Principal Component Supply Clas s Group Location Category Wind Missile Protection Program Cornents XXXVI Control Room Air Conditioning System P 3 NA B I b b c B XXXVII ECCS Area Exhaust System
- 1. Filters P 3 NA A I b b c B
- 2. Ductworks valves P 3 NA A I b b c B
- 3. Cables P 3 NA A I b b c B 35(C)
- g. ghaust Fans P 3 NA A 1 b b c B
- AreI*edeTbp$ ting CS P 3 NA A I b b c B XXXVIII Imak Detection System (Containment) - !'$7
- 1. Temperature element CE 2 NA C I b b c B (z)
- 2. Differential temperature switch CE 2 NA C I b b c B (z)
- 3. Differential flow indicator CE 2 NA C I b b c B (z)
- 4. Pressure switch GE 2 NA C I b b c B (z)
- 5. Dif ferential pressure indicator switch CE 2 NA C I b b c B (z)
- 6. Differential flow sumer CE 2 NA C I b b c B (a)
Other Leak Detection
." 1. Control Building Basemat P 2 NA B I b b c B 7 Sumps Level Indication y 2. Reactor I.uxiliary Building P 2 NA A I b b c B low Purity Sumps Level 59 Indication
- 3. DG Building Cubicle and P 2 NA S I b b c B Day Temp Sumps Level Indication XXXIX Civil Structures
- 1. Reactor Building including:
Base slab P 2 NA M I b b a B Shield Building Cylindrical Wall P 2 NA O I a a b B 22 s Dome P 2 NA 0 I a a b B
." O Steel Containment P 2 NA e I b b c B -: l 59 =
l C
AChGS-PSAR it would be necessary to penetrate the operating floor at an impact 35(U) s angle at five to fourteen degrees frei horizontal. The operating j tloor is composed of reinforced concrete three feet thick. (See (s_,,j
/ Figure 1.2-22). At these angles of impact, the mt.sile would not penetrate the operating floor.
To impact the walls of the hadwaste Building adjacent to the Turbine h5(U) building, it would be necessary for the missile to penetrate the reinforced concrete turbine pedestal and several reinforced concrete internal walls at the mezzanine level. Therefore, the missile would not have sutticient energy to reach these walls. 135(U) 3.5.2.2.4 Summary and conclusion l59 1he high trajectory turbine missiles are characterized by their nearly i vertical trajectories. The total damage probability of a high trajectogy turbine missile striking the safety related structures is less than 10 per unit year as listed in Table 3.5-4. In addition, the vulnerable safety related equipment area which is exposed to the potential turbine missile is redundant an.d physically well separated. Consequently the risk from high trajectory turbine missiles is insignificant. 58 The ACh6S turbine generator has been arranged in a peninsula orientation. bith the exception of the Radwaste Building, this configuration excludes all major systems important to safety from the low trajectory turbine missile s trike zones. The haowaste Building does not contain any essential systems required for sate st.utdown and is located below the turbine operating deck f , level. The location of the Radwaste System components relative to the tur-bine is such that they are adequately protected by the presence of the re-(\--) inforced concrete pedestal, internal building walls and the turbine opera-ting deck floor. Thus, the plant configuration complies with the guide-lines of Regulatory Guide 1.115, " Protection Against Low Trajectory Turbine hissiles." In addition to the above, due to the redunaancy and testing features of the turbine overspeed protection, quality control manufacturing processes, materials, and inspection program, the hypothetical turbine missiles are considered very remote. Consequently the risk of potential turbine missile damage to saiety related plant structures, systems, and components fer the facility is acceptably low. l t \ (G ()
\
3.5-11a (U)-Update Am. No. 59, (6/81)
d , ACNCS-PSAR I35(U) 3.5.2.3 Site Related Missiles i 3.5.2.3.1 Aircraf t Missiles (j s No commercial airport facilities are located in the vicinity of the site. The nearest major commercial facilities are located at Hobby and Houston Intercontinental, both of which are located approximately 50 miles from the site. The Eagle Lake facility is located approximately eleven miles west of the site. No military f acilities exist within 20 miles of the site. Consequently neither aircraft nor military projectiles are con 3idered cred-ible missiles for the site. Section 2.2.1.6 described r.irport locationa l35(U) ! in the site region. 3.5.2.3.2 Transportation Accidents Section 2.2.3.4 demonstrates that the consequences of missiles arising from l35(U) explosion of hazardous material being transported along Highway 36 and Gulf Colorado and Santa Fe line of the Santa Fe Railway is not significant. ' 3.5.2.3.3 Industrial Accidents Pipelines in the vicinity of the plant include a 24 inch natural gas pipe-line of the Texas Utilities Company and 8 and 6 inch oil pipelines of the 35(U) Shell Pipeline Corporation. The 24 inch pipeline will be rerouted on the opposite side of the cooling pond from the plant, approaching to within ap-proximately 2 miles of the plant. Section 2.2.3 demonstrates that the consequances of missiles arising from these sources is not significant. g No other major industrial facilities exist within 5 miles of the site. 3.5.2.4 Tornado Missiles Structures or components whose failur'e could prevent safe shutdown of the , reactor, or result in significant uncontrolled release of radioactivity from the unit , shall be protected from such failure due to design tornado wind loading or missiles by the following methods: " a) Structure or component is designed to withstand tornado loading (see Section.3.3.2) and/or tornado missile. b) Component is housed within a structure which is designed to withstand the tornado and/or missile loading.
+
v 3.5-12 (U)-Update Am. No. 59, (6/81)
ACNGS-PSAR n P,, = A 4 5 ps ig internal static pressure to envelope the conbined pres-
) sure induced by an accident that releases hydrogen geaerated from 100% 57 'O act ive fuel clad metal water react ion and tre pressure from post-acci-dent inert ing assuming carbon dioxide. See S e tion 6.2.1.3.4 for the derivation of the accident pressure. l 59 Fin = Pressure produced by an inadvertent actuation of the post-accident inerting system causing full inerting (with carbon dioxide). This internal static pressure is 25 psig, as derived in Section 6.2.1.3.5.
P st
= C ntainment Vessel structural acceptance test 57 (pressure = 110 percent of P in as required by hUREG 0718 Item 11.B.8.4.
b) Temperatures
)
T
=
Design (accident) t emperature inside Containment. khen coinci- 35(D) dent with P this temperature is 185 F (Table 6.2-1A). It is adjusted ac$ordingly when negative (accident) pressure occurs. For T under IbA and SbA, see Sections 6.2.1.3.1.3 and 54 6.2.1*3.1.4 respectively. I g
=
Operating lemperature (the tange is 60 to 80 F inside Containment and 51 to 95 F in the annulus). During SRV blowdown the increasa temperature in the suppression pool [
} is included in 1 . During construct.on T is specified
( / as the construction temperature. 1
* = lemperature inside the containment associated with the pressure F . khen coincident with P this temperature is 195 F, as dNeribed in Section 6.2.1.3N .
1.
= lemperature inside the containment associated with the inadvertent actuation of the post-accident inerting system causing full inert-ing (with carbon dioxide). This temperature is 95 F, as described in Section 6.2.1.3.5.
T
* = Ambient temperature in the containment during the Containment 57 Vessel structural acceptance test. This ranges from 30 F to 96 F.
c) Lead Loads D = Dead loads; they shall include the following:
- 1) keight of vessel shell, penetrations, hatches and locks.
The dead weight of the polar crane and its runway.
- 3) beight of platforms, walkways, equipment, piping, ventilation
('"] duct, cable and trays, conduit, etc. These loads are generated (' either by direct attachment to the vessel, or through support-ing structures. 35 (C) 3.8-7a (C)-Consistency (D)-Design Au. No. 59, (6/81)
ACNGS-PSAR
,Cs 3.8.2.3.2 Load Combinations (V )
3.8.2.3.2.1 Containment vessel Shell l9 The design of the Containment will include consideration of the load combi-nations listed below. Stress limits for these loading conditions are dis-cussed in Section 3.8.2.6. l54 a) Construction and Test Conditions
- 1) D+LII} + TII) + WII)
Vessel Test 35(C)
- 2) D + L( 2 ) + T,( 2 ) + Pg ( 3 ) + F, .
Wau h
- 3) D + L(2)(4) , 7 (2) , po (2) , pn o
Vessel Test
- 4) D + P,g +Tst (PAIS actuation) 59 b) Normal Operating or Shutdown Conditions l35(D)
- 5) D+L+T o + P, + R, + Fn
- 6) D + L + Ig(5) + Pg +P +R g +F n
- 7) D+P,+lin g PAIS Inadvertent 57 Actuation c) Severe Environment Loads SRV blowdown l57
- 8) D + L + T,(5) + P, + Pbd + R, + F, + F,q, 9 D + L(6) + T (6) , p (6) , g (6) , y ,y MWUg @
- 9) o o o n ego 1
d) Extreme Environmental Loads 35 SRV blowdown (D) [57
- 10) D + L + To (5) + Po +Pbd + Ro + F, + Feqs 44 e) Abnormal - Severe Environmental Loads
'"*II (8)
- 11) D + L + T,(7) + P,(7) +P *** " *""
bd ps' sc sation or (9)(13) +F or P ch n + F,qo +R a "EE
- 12) D + L + 1 (10) , p (10) + R +
* *
- F"(10)+ F*S 54 l57 Long Term LOCA
- 13) D + L + T,(11)(14) , p (14) , p (12) Intermediate l57 break, A1:S
+P or Pch(9)(13) + R, + Fn + F,q, sc f 14) D + L + T (15) , p (15) ,bdp (8) , p (9) l57 a ch
( a p (iO) n y ego Small break { (C)-Consistency i (D)-Design 3.8-9a Am. No. 59, (6/81)
ACNGS-PSAR
- 15) D+L+T +P +R a +F n +F ego Negative Pressure l 57 7, a e i
!, ) D+L+T +P +R +F n Accident at pene- l 57 N/ 16) o o o +Yj+Feq tration sleeve 17 ) D + L + T (b) + P (6) + R (6) + F Post-accident l 57
+y Pa flooding 54 f) abnormal - Extreme Environmental Loads L
- 18) D + L + T (7) + P (7) + PPd(0) + Pps, Pont swell l57 steam condensa-P sc rP i9)(13)a+ R, + F, + F ch egs tion or chugging 19 ) D + L + T,0 0 ) + p ( 10 ) + R, + Fn ( L ng term LOCA l 57 eqs '
- 20) D + L + Ta (1I)(I4) + P (14) + p (12) Intermediate l 57
+P sc rP ch (9)(13) + aR, + P +bp break, ADS eqs 54 n
- 21) D+L+ ( 5) + p (15) + p ( 8) + p (9) Small break 59
+R +F xF a n eqs
- 22) D+L+T +P +R a +F n +F eqs Negative pressure l57 a e Accident at
- 23) D+L+To+Po+Ro+F n +Yj+Feqs penetration sleeve l 57
( ; Degraded core 57
- 24) D + P,9 +Tmv v) :
t 3.8-9b Am. No. 59, (6/81) i i
ACNCS-PSAR values of R The design accident loads discussed a$ov(thermal e, result lead) from on the penetration's. a postulated pipe break inside the drywell. They 35(C)
/' h
( ) would not occur simultaneously with the loads "Y." which are for a pipe break-3 ing at a penetration. The intent of load combinations (e) 16 and (f) 23 is to cover the design of 57 local areas around penetrations for any pipe break postulated to occur at a pene t ra t ion . In such a case, design accident loads (pressure, tem pe ra t ure , and fluid) would not be acting simultaneously on the overall vessel. The loads "Y" which would be acting at the penetration already include the
- thermal ef fects on the penetration of the postulated pipe rupture (se definition item (g) in Section 3.8.2.3.1).
35(C) Local areas will be designed by investigating the applicable loads combined as in the above listed loading cases. Local areas to be investigated in-clude penetration nozzles and the surrounding shell, anchorage details, crane ruaway and floor framing brackets, and the dome knuckle. Investigations of these are discussed further in Section 3.8.2.4. 3.8.2.3.2.2 Bottom Liner l 59 The containment bottom liner plate will be designed in accordance with the applicable rules listed in the ACI-ASME ( ACI-359), Division 2 Code issued in January 1975. It will also be designed in accordance with selected sec- 54 tions of ASME Code, Section III, Division 1, Subsection NE applicable to strength, buckling and low cycle fatigue for cases where SRV negative pressure occurs. The load combinations shown in Table 3.8-3 and the
/o b) additional combinations shown in Section 3.8.5.3.2(a) related to NUREG 0718 Item II.B.8.4 are applicable to the liner plate design except that load factors for all load cases shall be taken to equal to 1.0.
59 l 54 3.8.2.4 Design and Analysis Proceedures The design and analysis of the Containment will be the responsibility of the selected containment vendor. The scope of the vendor's responsibility includes the design and analysis of the vessel shell and bottom liner, the vessel anchorage, the crane runway, dome platforms, intermediate floor eupport seats, personnel locks, equipment hatch, and penetration nozzles. The penetration internals discussed in Section 3.8.2.1.2 are not included. The vessel vendor will be required to report fully on the actual completed design and analysis, and a summary of this will be available for the FSAR. Containment Vessel design and analysis procedures will vary somewhat according to the selected vendor. Ilowever , the following discussion rep-resents, in general a typical example of the approaches utilized, and, in several areas, it represents specific requirements. 1 Q3.20 (A)
%/
3.8-10 (C)-Consistency i Am. No. 59, (6/81)
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t i t ACNGS-PSAR l I LIST OF EFFECTIVE PAGES i f
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CHAPTER S REACTOR COOLANT SYSTEM i 4 M Amendment j 1* 59 2* 56 }
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*Ef fective Pages/ Figures Listing l i
4 - 1 1 Am. No. 59, (6/81)
1 ACNGS-PSAR ! EFFECTIVE FIGURES LISTING
- v CHAPTER 5 REACTOR COM. ANT SYSTEM
- All figures, whether labeled " Unit 1 and 2" are to be considered applicable to Unit No.1.
Fi gure Amendment No. 5.1-1 18 5.1-2 - 5.1-3a 42 5.1-3b 31 5.1-3c 31 5.2-1 42 5.2-2 - 5.2-3 - 5.2-4 -
- 5. 2-5 26 i 5.2-Sa 26 5.2-5b 26
! 5.2-6 26 5.2-7 32 5.2-8 34 5.3-1 56 5.3-2 56 5.4-1 - 5.4-2 - 5.5-1 5.5-la 34 5.5-2a - 5.5-2b - 5.5-3 - 5.5-4 - 5.5-5 - 5.5-6 - 5.5-7a 42 5.5-7b 42 5.5-Ba 5 5.5-8b 5.5-8c 5 i 5.5-9 - 5.5-10a 42 5.5-10a(i) 42 5.5-10b 42 l 5.5-10e 42 l 'CI 5.5-11e 5.5-11b 5.5-13c 42 42 42 5 Am. No. 59, (6/81) 1
i ACNGS-PSAR 1 l, EFFECTIVE FIGURES LISTING i CHAPTER 5 (Cont'd) i Figum- . Amendment No. l t 5.5-12 , 5.5-13a 26 l 5.5-13b 26 l' l l 5.5-14 26 l ! 5.5-14a 26 l 5.5-14b 26 l 5.5-15 - i i 5.5-16 34 ! 5.5-17 f
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ACMGS-PSAR LIST OF EFFECTIVE PAGES CBAFTER 6 ; i ZNGINEEE D SAFETY FEATURES Amendment No. L*I* i 59 g, 58 2* 58 3* 58 4* 58 5* 59 , b* 58 . 6a 58 I* 58 d* 58
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. gan 58 1 l 13, 58 14* 58 g 37 I 46 i 3 111 37 IV 56 1 56
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58 x111 58 xilia 37 xiv 56
*" 56 xvt 56 xvil 56 xviii 5 xviita 46 6.1-1 37 l 6.1-2 37 6.2-1 37 6.2-2 56 6.2-3
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Am. No. 59,(6
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r ACNCS-PSAR i LIST OF EFFECTIVE PAGES (Cont'd) ! CHAPTER 6 l e g Amendment No. ) l 6.2-29f(i) 37 6.2-29g 37 ; 2
- 6. 2-29h 37 6.2-29h(i) 37 j
- 6.2-29i 26
! 6.2-29j 26 l l 6.2-29k 6 6.2-291 37 6.2-291(i) 37 ! 6.2-29m 26 [ 1 6.2-30 26 ! 23 I 6.2-30a j 6.2-30b 21 ;
- 6.2-30c 23 !
6.2-30d 23
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6.2-31 7 ! 6.2-31a 7 { i 6.2-32 59 : ! 6.2-32a 57 ! ) 6.2-33 - l 1 6.2-34 30 t i i 6.2-35 37 ! 6.2-35a 30 ! l 6.2-35b 31 l 6.2-36 - ; 4 6.2-37 7 (' i 6.2-38 - ? 6.2-39 - l 6.2-40 46 1 6.2-40a 21 l 6.2-41 26 6.2-41a 26 j 6.2-42 37 i 1 6.2-42a 16 l 6.2-42b 37 l 6.2-42c 37 ! 6.2-43 37 , l 6.2-43a 37 l ! 6.2-44 46 i i ! 6.2-44a 37 i 6.2-45 37 i r 6.2-46 37 I 6.2-46a 37 i. 3 Am. No. 59, (6/81) , , 4 1 i
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LIST OF EFFECTIVE PAGES (Cont'd) l CHAPTER 6 ; i i
& Amendment No.
t i 6.2-95 5 ; 3 6.2-96 37 ; i 6.2-97 37 l 6.2-98 37 : 6.2-99 37 , + 6.2-10 0 37 s 6.2-101 37 ! 6.2-10 2 37 I 6.2-103 57 : 6.2 -10 4 57 ! 6.2-105 57 6.2-106 57 i 6.2-107 57 i 6.2-107a 57 , 6.2-107b 57 i 6.2-107c 57 l
- 6. 2-107d 57 i
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6.2-107h 57 l 6.2-10 8 57 ! 6.2-109 57 ; 1 6.2-109a 57 l 6.2-109b 57 , 6.2-10 9c 57 : I 6.2-109d 57 ; 1 6.2-110 (Intentionally deleted) 57 ; I 6.2-1 11 58 !
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j 6.2-113 (deleted) 37 l 6.2-114 23 6.2 -1 15 17
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6.2-117 (deleted) 37 .- 6.2-118 5 6.2 -1 19 5 i 6.2-119a 5 i 6.2-120 37 6.2-121 5 6.2-122 23
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~ ACNGS-PSAR EFFECTIVE FIGURES LISTING (Cont'd) CHAFTER 6 I Figure No. Amendment No. 6.2-26 (9:eet 12) 57
- 6.2-26 (Sheet 13) 57
. 6.2-26 (Sheet 14) 57 6.2-26 ('heet 15) 57 4 6.2-26 (Sheet 16) 57
- 6.2-26 (Sheet 17) 57 l 6.2-26 (Sheet 18) 57 l 6.2-2 6 ( Shee t 19) 57 i
6.2-26 (91eet 20) 57 . 6.2-26 (Sheet 21) 57 ' , 6.2-26 (Sheet 22) 57 6.2-26 (Sheet 23) 57 6.2-26 (Sieet 24) 57 i 6.2-2 6 (Sheet 2 5) 57 , 6.2-26 (Sieet 26) 57
- 6.2-26 (Sheet 27) 57 i 6.2-27 (9 eet 27a) 59
! 6.2-26 ( Shee t 28) 57 6.2-26 (Sieet 29) 57 < 6.2-26 (Shee t 30' 57 , l 6.2-26 (Sicet 31) 59 h-l i 4 10a Am. No. 59, (6/81) 'f g J t r w= mw , w - er - +**'-1---'re-erTm*v---'--+ m- 5 mw---- - - - - * - - * *-*-er-**-" -- - - - - -
ACNGS-PSAR [ j The Standby Gas Treatment System is represented by input tables describing y/ f an size, system resistance, and pressure dependent flow chara::teristics. The analyses were performed for the steam line break described earlier in Section 6.2.1.3.1 with the same minimum Engineered Safety Features as oe-scribed there. In addition, only one of the two 100 percent SGTS subsys-tems is assumed to be operating. Table 6.2-5 lists the import ant assumptions used in the Shield Building annulus transient analyses. A heat transfercoefficientbetweenthecontainmentagmosphereandthecon-tainment steel is taken to be equal to 250 BTU /hr.-f t - F. This number represents a typical condensation heat transfer coefficient on the containment steel. This heat trans fer coef ficient is higher than a convec-tion heat transfer coefficient and will result in a faster heat t rans fer rate to the annulus. This is the reason that a conservative condensing heat trans fer coef ficient is used instead of a more realistic convection heat trans fer coef ficient. Inleakage to the annulus is based on a design basis leak rate of 45 cfm at -
-2 inches wg which will be verified by test. During phase 3 operation the l SGTS exhaust rate is equal or less than 2000 cfm at all times. This is the 17 value used throughout the phase 3 period for LOCA dose calculations given in Chapter 15. Q2-58 (D ) Figure 6.2-22 shows the annulus pressure ve rsus time for the above de-(d scribed conditions. Figure 6.2-23 is a linear time scale replot of the annulus pressure transient given in Figure 6.2-22. Figure 6.2-24 shows the corresponding annulus temperature versus time. All of the above curves are based on an assumption of zero percent initial humidity in the annulus, as this assumption results in the highest pressure and temperature transients in the annulus. These plots show that the temperature in the annulus will reach 130 F after 2 hours. The annulus reaches a peak of -1/4 in. H2 O at 50 minutes and would remain negative during the transient. The results show that the design criteria set out for the SGTS in Section 6.2.3 are met.
6.2.1.3.4 Post-Accident Containment Pressure Calculation Based on a transient event which results in 100 percent metal water reac- 57 tion of the active fuel clad, the following conditions are calculated for this event: containment pressure = 42.5 psig, containment temperature
=
133 F and suppression pool temperature = 183 F. See the response to 59 NUREG 0718 Item II B 8.4(a) in Appendix 0 for details of the analyses. 57 6.2.1.3.5 Containment Pressure Calculation Inadvertent Actuation For the purposes of containment structural evaluation, the following con-ditions are assumed to result from inadvertent actuation of the Post-Accident Inerting System during normal operation: j] Containment pressure = 26.5 psig l59 V Containment / Suppression pool temperature = 95 F 57 6.2-32 Am. No. 59, (6/81)
Y % Penetration Penetration Applicable Justi Number Type CDC System Service (E)/(I)/(N) d I-Later XI 56 RWCU - Sample Line I Not rec and imr lowing but ust accider M-137A & B IVd 56 Fire Protec tion N Availat M-141 VI essenti is not helpfu: follow: /N) M-138A & B M-142 IVd VI 56 Fire Protection N Availal essenti ( j is not helpfu: follow, M-139A & B IVd 56 Seismic Fire Protection N Availal essenti is not helpfu followi M-140A & B IVd 56 Seismic Fire Protection N Availal essentz is not helpful followi M-143 & IV & 56 Post Accident Inerting I Not rG M-149 VII diste use; m
) tiatei j
4 Isolation Maxit Valve Clost .ficction for Tag Line Size Valve Openi
- )/(I)/(N) Number Valve Type (In.) Operator Time <
iuircd during EV-71201 Cate 1/2 Solenoid iedictely fol- EV-71203 Gate 1/2 Solenoid cn cccident EV-71205 Cate 1/2 Solenoid ful for post- Later Gate 1/2 Solenoid <t sampling EV-71200 Gate 1/2 Solenoid EV-71202 Cate 1/2 Solenoid EV-71204 Cate 1/2 Solenoid Later Gate 1/2 Solenoid >ility is non- 2FP-V188-S1 Gate 4 Motor Std (N( el as system 2FP-V190-S2 Check 4 - - required, 2FP-V546-S2 Cate 4 Motor Std (Nc or desirable 2FP-V548-S1 Check 4 - - ng accident. >ility is non- 2FP-V229-S1 Gate 4 Motor S td (N< al as system 2FP-V231-S2 Check 4 - - requi re d , 2FP-V545-S2 Gate 4 Motor Std (Ni or desirabic 2FP-V547-S1 Check 4 - - ng accident. aility is non- 2SF-V107-S1 Gate 6 Motor Std (N4 al as system 2SF-V110-S2 Check 6 - - required, L or desirable ng accident. illity is non- 2SF-V108-S1 Cate 6 Motor Std (N ill as system 2SF-V111-S2 Check 6 - - required, or desirabic ng accident, quired for imme- (Later) Gate 8 Motor Std (N< post-tecident (Later) Gate 8 Motor Std (2 anun11y ini- (Later) Gate 8 Motor Std (R cystem (Later) Gate 8 Motor Std (2 (Later) Check 4 - - (Later) Check 4 - -
?
AQIG8-PSAR TABLE 6.2-12 (Cont'd) [ l um Isol 30/ Valve Position si ng Flow Power (Di ge cc)_, Valve Location Direction Normal Accident Failure Parad (See Not Inside out Open Opep or Closed Closed B,C,RM Inside Open Inside Open Inside Open " " Outside Open Outside Open Outside Open Outside Open its 2) Outside Shield Bldg In Closed Closed As is B,C,RM Inside Containment - - - Reverse' sta 2) Inside Containment Closed Closed As is B,C,RM Inside Drywell - - - Reverse >to 2) Outside Shield Bldg In Closed Closed As is B,C,RM Inside Containment - - - Reverse- >ta 2) Inside containment Closed closed As is B,C,RM f Inside Drywell - - - Reverse >ta 2) Outside Shield Bldg In Closed Closed As is B,C,RM Inside Containment - - - Reverse ato 2) Outside Shield Bldg In Closed Closed As is B,C,RM Inside Containment - - - Reverse >to 2) Outside Shield Bldg In Closed Open or Closed As is B,C,RM , >to 2) Outside Shield Bldg Closed Open or Closed As is B,C,RM' >to 2) Inside Containment closed Open or Closed As is B,C,RM-ste 2) Inside Containment Closed Open or Closed As is B,C,RM Inside Containment - - - Reversd Inside Drywell - - - Reversd j I I I 5
I l
- tion Reopen Appendix
,nal By J Applicable Bypass rerca Manual Type C Figure Penetration Leakage set ^r) Only Test (See 6.2-26) _. Location _ Ba rrier s
;s 5,6,7) (Sheet / Item) (See Note 11)
Ye8 Yes 27A / 31 Aux Bldg A,B n i. hl Il n 99 II ,, II ft n '. Il 99 Yes Yes 27 / 22a Aux Bldg A,E flow - Yes Yes No flow - No Yes Yes 27 / 22a Aux Bldg A , B, ' flow - Yes Yes No flow - No Yes Yes 27 / 22b Aux Bldg A , B, flow - Yes Yes Yes 27 / 22b Aux Bldg A , B, flow - Yes Yes Yes 27A / 30 Aux Bldg A,B, Yes Yes Yes Yes Yes Yes flow - $'; flow - No i s
s t Y I.. e Sheet 12 of 15 1 0 k ja Potcntial I Bypnos Path -Remarks l Yes , o 59 i I
- i 1
i Yes Penetrations 137A & B and 141 are j located on the same pipe run. . t 57 : Yes -Pencerations 138A & B and 142 are located on the same pipe run. i. Yes Yes
-Yss 59 t
6.2-107g Am. No. 59, (6/81) I *
\.
. . . , . - _ . - - , , , , , ..--n,.-.,.,,,,.. .-,,c., , ,.,,-,-,,,-,,,..__,,w.,,,,
l ACNGS - PSAR ANNULUS DRYWELL CONTAINMENT SHIELD WALL ,f'~) O A L h 14 j M M 149 X 2 - M-143 2 -Q BV ' 1T 1 I I a 6 s 6 a 6 c.
.b PX PX PX
- 30. POST ACCIDENT INERTING SYSTEM CONTAINMENT ANr!ULUS SHIELD WALL PX PX PX n n n EV , ,
pg- , , EV
) ; J JL s s b I - L/ ER 2 PX PX pg PX n
r, r1 y,g EV s.'.* EV L L j L b I LATER * [ PX n P,X r g 3.., PX n EV ((h EV
; L ; L ; L C I LATER [
PX PX +- PX n n $,* n EV , , jf , , EV
); ,
JL J L , b i - LATER 2 1 %~
- 31. RWCU SAMPLING SYSTEM AM. NO. 59, (6/81) n HOUSTON LIGHTING & POWER OMPANY
/ ' Allens Creek Nuclear Generatin, Station \ Unit 1 CONTAINMENT ISOLATION VALVE ARRANGEMENTS SHEET 27A OF 31 FIGURE 6.2-26
ACNGS - PSAR [ PRESSURE DRYWELL CONTAINMENT ANNULUS SHIELD WALL 'y/ VESSEL l d b d b d b
- ;4 l I-241A l l I2418 l 74 ;
I s
- I 224A l l I-224B PX lfl4 -f PX PX HYDROCEN ANALYZER PANELS PX PX PX
' r . i r , r d b d b d b r 7,A I2o2A '
l l242B ! >C: k
- M 4 I 251 A l l I-2518 l l4 :
i r , , i r d b d b d b u a u PX PX PX CONTAINMENT HYDROGEN ANALYZER SAMPLING SYSTEM AM. NO. 59, (6/81) HOUSTON LIGHTING & POWER COMPANY l Allens Creek Huclear Generating Station f]) \ Unit 1 CONTAINMENT ISOLATION l VALVE ARRANGEMENTS l SHEET 310F 31 ! FIGURE 6.2-26 I
E t
' ACNGS-PSAR' [
i EFFECTIVE PACE LISTING I CHAPTER 7 l 1- t i 331RUMENTATION AND CONTROLS . i Pg Amendment No. { I 1* 59 ! ! 2* 59 [ { 3* 59 j ' 4* 59 ! 5* 40 I ) 6* 59 [ j 6a* 43 j 7* 59 }* ! 8* 59 i 9* 59 !! , 10* 59 ! 11* 59 !
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- Effective Pages/ Figures Listings l 1 A:n. No. 59, (6/81) l l__._. _ _ . . _ _ _ _ _ . - _ . _ _ - , . _ . . _ . . . - - - - _ _ . - - - . - _ - - - - , -. ......- - -- _ -- - - -'
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il .. I. ACNGS-PSAR f' j, EFFECTIVE PAGE LISTING (Cont'd) i CHAPTER. 7 Ame ndme nt No. [ Pg I. xxit 37 i 37 f ! .xxiii I xxiv 37 f xxv 37 ! l xxvi 37 l l 37 xxvii xxvili 37 xxix 37 xxx 37 , xxxi 37 xxxit 37 l' 37 xxxiii xxxiv 37 xxxv 37 xxxvi 37
.xxxvii 37 l xxxviii 37
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xlvi 37 ; 7.1-1 37 ; i ! 7.1-2 37 l 7.1-3 37 l 7.1-4 37 { 7.1-5 44
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7.1-10 37 ; ! 7.1-11 37 > ' 45
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I I ~ ACNGS-PSAR l t ! EFFECTIVE PAGE LISTING (Cont'd) ! CHAPTER 7 l l Page Amendment No. 7.1-17 37 ! 37 ! 7.1-18 7.1-19 37 }
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ACNGS-PSAR EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 7
- ' l 4
Page Amendment No. l 7.2-1 35 l 7.2-2 35 7.2-3 35
! 7.2-4 35 l 7.2-5 39 7.2-6 35 7.2-7 56 7.2-8 35 7.2-9 35 7.2-10 40 7.2-11 35 7.2-2.2 35 7.2-13 35 ,
7.2-14 35 7.2-15 35 7.2.16 35 i 7.2-17 35 7.2 83 35 7.2-19 35 7.2-20 35 7.2-21 35 7.2-22 35 7.2-23 35 7.2-24 35 7.2-25 35 7.2-26 35 7.2-27 35 7.2-28 35 i 7.2-29 35 7.2-30 35 7.2-31 35 l 7.2-32 35 7.2-33 35 7.2-34 35 7.2-35 35 7.2-36 35 7.2-37 35 7.2-38 35 7.2-39 35 7.2-40 35 7.2-41 35 , 7.2-42 35 7.2-43 35 7.2-44 35 7.2-45 35 I 4 Am. No. 59, (6/81) l
i ACNGS-PSAR 1 EFFECTIVE PAGE LISTING (Cont'd) i CHAPTER 7
- Page Anendaent No.
7.3-1 37 7.3-2 - 37 7.3-3 37 , 7.3-4 37 7.3 37 7.3-6 45 7.3-7 37 7.3-8 37 7.3-9 37
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l ACNGS-PSAR t EFFECTIVE PAGE LISTING (Cont 'd) j CHAPTER 7 Page No. Amendment No. I 1 , 7.3-51 37 7.3-52 37 7.3-53 37 7.3-54 44 7.3-55 37 7.3-56 37 7.3-57 37
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I 7.3-74 37 7.3-75 37 4 7.3-76 37 ; 7.3-77 37 7.3-78 37 i 7.3-79 37 7.3-80 37 ' 7.3-81 37 7.3-82 37 7.3-83 37 1 7.3-84 37 7.3-85 37 , 7.3-86 37 7.3-87 37 7.3-88 37 7.3-89 37 7.3-90 37
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1 I i-l ACNGS-PS AR , i ' EFFECTIVE PAGE LISTING (Cont'd) l CHAPTER 7 l f- i l Page No. Amendment No. l !. 7.3-94 37 l- 7.3-95 37 I 7.3-96 37 i 7.3-97 37 l 7.3-98 37
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ACNGS-PSAR EFFECTIVE PAGE LISTING (Cont'd) OR APTER 7 ; Page No. . Amendment No. 7.3-13 6 37
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7.3-13 8 37 7 . 3 - 13 9 37 e 7.3-140 37 7.3-141 37
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{ 7 . 3 - 14 7 37 l 7.3-14 8 37
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7 . 3 - 16 3 37 l 7.3-16 4 37 7.3 - 16 5 37 7.3-16 6 37 7.3- 16 7 37
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d i ACNGS-PSAR EFFECTIVE PAGE LISTING (Cont'd) CH APTER 7 Page No. Amendment No. 7.3-179 37 7.3- 18 0 37 7.3-181 37 7.3- 18 2 37 7.3-183 37
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ACNGS-PSAR EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 7 Page No. Amendment No. i 7.5-4 39
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;' EFFECTIVE PAGE LISTING (Cont'd) ;
CHAPTER 7 l- Page 'No. Amendment No. I 7.5-26 37
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- i l' EFFECTIVE PAGE LISTING (Cont'd)
CHAPTER 7 1 l l'
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EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 7 i i-l ,t ; i' Page No. Amendment No. l e 7.6-27 37 I 7.6-28 37 j 7.6-29 37 > 7.6-30 37 ! l 7.6-31 37 ! 7.6-32 37 7.6-33 37 7.6-34 37 l 2 7.6-35 37 7.6-36 37 7.6-37 37 [ 7.6-38 37 I 7.6-39 37
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- . ACNGS-PSAR t
EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 7-4 e i t Page No. Amendment No. , 7.6-73 37 j ^ 7.6-74 37 7.6-75 37 j. 1 7.7-1 37 l 7.7-2 37 i 7.7-3 37 7.7-4 37 7.7-5 37
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-, . . .. . . . - , - . . ...- .- .. . . . - . - _ - -.. ... . ~ - _ ~ . . a i;. ;
I ACNGS-PSAR i . ing the valves and monitoring _ the valve closure via the annunciators 37(G)
; in the Control' Room (10% and 90% closed indicators are provided).
j) Manual- Reset Capability F -1 solation Valves classified as non-essential, will require manual, 57 i operator initiated re' set upon clearance of all automatic isolation ! signals. Isolation valves classified as intermediate will require i manual operator initiated -reset without clearance of the accident s closure signal, but will not be possible in the presence of a system ! failure signal. Isolation valves classified as essential will re- i ceive .no automatic containment ' isolation signals but will be ; ! manually controlled b,v the operator and/or required system. per- ! formance interlocks. 59 i l 1
~
l 2 j L i i j > 4 e i h t t I i ; t t i i : i i i I i .t
.t i !
t i [-. \ , I i l I 7.3-30a (G)-GESSAR Am. No. 59, (6/81) . I 1 ' . - - . . - , - - . . - - . - - - ~ -.-~_n~--,--n-----.--~ .. -n-----.+---_-_.--n,---,-~.--en.--- - - - - , . - - - - - - - - - - - , . --,-i
, ____._...._____._.._m_ _ _ . ..- ... _ __ _,-. .. . . _ . _ . . ~ . m,. _ . . _
h
.,s 1 g
i 1 n. ACN(S-PSAR ' f I TABLE 7.5-0 (Cont'd) l ' VARIABLE PANEL
- VARI ABLE TYPE RANGE CATECORY NUMBER COMMENTS 1
.i t
i .26 Effluent Rau activity C See Table 12.2-5a . ! ' l 57 . :[ i 27 Main Feedwater Flow D 0-110% 3 P680-0 3D design flow
" l 28 Condensate D Bottom to Top . 3 (la ter)
Storage Tank Imvel ! 29 . Containment ' D 0-110% 2 P601 59 - < Spray Flow design flow P601 .
~
30 Drywell Pressure D Previously listed as Variable 8 31 Suppression Pool D Previously listed as Variable 18 ! .} t Water Level 1 ' 32.. . Suppression Pool D Previously listed as Variable la i r Water Temperature } '; Drywell Atmosphere D 4 0-44 0*F 2 P871 , u 33
, Temperature P872 57 v' . -
0-15" w.g. 2 P601-19B
- E 35 Main Steamline D j isolation valves 0-5 psia leakage control .{
system pressure 2 9 g
$ ]
a-
^
, w t v" . a 4
ACNGS-PSAR 6 EFFECTIVE PAGES LISTING
=
CHAPTER 8 , t ELECTRIC POWER Amendment No. 1* 59 45 2* 9 3*
. 35
. 35 LL kII 35 35
-* . 43 s vt 35 8.1-1 43 8.1-2 45 8.1-3 43 8.1 -4 43 8.1-5 43 8.1-Sa 8.1-6 45 >
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t L f ACNGS-PSAR ! EFFECTIVE PAGES LISTING (Cont'd) f i CHAPTER 9 ! _Page, Amendment No. 9.1-1 37 ! 9.1-la 37 i 9.1-2 37 f 9.1-2a 37 9.1-3 53 , 9.1-3a 53 9.1-4 53 9.1-5 53 , 9.1-Sa 53 l 9.1-6( Delete d) 53 ! 9.1-6a 53 9.1-6b 53 , 9.1-7 48 i 9.1-7a 23 I 9.1-8 39 f ! 9.1-9 37 i
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ACNGS-PSAR EFFECTIVE PAGES LISTING (Cont'd) CHAPTER 9 3. , Pg Amendment No. ! 9.2-7 46 - 9.2-8 48 ! 9.2-8a 37 ! 9.2-9(amendment number not shown on page) 37 l 9.2-10 37 [ il 9.2-11 37 ;
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ACNGS-PSAR g- EFFECTIVE PAGE LISTING (Cont'd) CHAPTER 9 1 Pg Amendment No. 9.5C-29 40 9.5C-30 40 j 9.5C-31 40 I 9. 5C-3 2 40 9.5C-33 40 9.5C-34 40 9.5C-35 40 9.5C-36 40 9.5C-37 40 9.5C-38 40
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ACNGS-PSAR EFFECTIVE FIGURES LISTING CHAPTER 9 All figures, whether labelled " Unit 1" or " Units 1 and 2," are to be considered applicable to Unit No.1. Figure Amendment No. 9.1-1 - 9.1-2 (Deleted) 53 9,1-2a 53 5 .-3 (not assigned) 37 9.1-3a 37 9.1-3b - 42 9.1-4 - 9.1-5 9.1-6 - 9.1-7 - 9.1-8 - 9.1-9 - 9.1-10 39 9.1-11 39 9.1-12a (Deleted) 53 9.1-12b (Deleted) 53 9.1-12c (Deleted) 53 O 9.1-13 9.2-1 (not assigned) 37 37 ; 9.2-la 37 9.2-1 b 37 ' 9.2-2 (not assigned) 37 9.2-2a 37 9.2-2b 37 , 9.2-2c 37 9.2-3 (not assigned) 37 ; 9.2-4 37 9.2-5 37 i 9.2-6 37 i 9.2-7 37 l 9.2-8 (not assigned) 37 9.2-8a 37 9.2-8b 37 9.2-8c 37 9.2-9 (deleted) 37 - 9.2-10 37 t
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ACNGS-PSAR EFFECTIVE FIGURES LISTIleG (Cont'd) CHAFTER 9 Figure Amendment No. 2 9.2-15 46 9.2-16 1 9.3-1 37 9.3-2 37-9.3-3 37
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l' I ACNGS-PSAR EFFECTIVE FIGURES LISTING (Cont'd) l CHAPTER 9 l Figure Amendment No. i f 9.5-IC-49 40 9.5-1C-50 40 i i ! 9.5-1C-51 40 I 9. 5-1C-5 2 40 I t 0 i 1 l 4 I i 4 i e 1 l l ! l
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ACNGS-PSAR EFFECTIVE PAGES LISTING CHAPTER 10 STEAM AND POWER CONVERSION SYSTEM Page Amendment No. 1* 59 35 2* i 35 35 il iii 35 iv 1 y 35 10.1-1 35 10.1-2 35 10.1-3 35 10.2-1 35 10.2-2 35 10.2-2a 35 , 10.2-3 35 10.3-1 35 10.3-2 35 10.4-1 35 10.4-2 35 10.4-3 37 10.4-4 35 10.4-5 35 10.4-6 35 10.4-7 35 10.4-8 35 10.4-9 35 10.4-10 42 10.4-10a 45 r 10.4-10b 37 10.4-11 35 10.4-12 35 10.4-13 35 10.4-14(Amendment number not shown on page) 35 10.4-15 35 10.4-16 35 ; 10.4-17 35 I L
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l ACNCS-PSAR EFFECTIVE PAGE LISTING (Cont'd) j CHAPTEE 11 l g Amendment No. ! 11.2-33 37 11.2-34 37 11.2-35 37
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CHAPTER 11 Amendment No. Pg 37 11.3-4 37 11.3-5 37 f 11.3-6 37 11.3-7 37 ! 11.3-8 37 11.3-9 39 11.3-10 37 i 11.3-11 37 l 11.3-12 42 < 11.3-13 42 l 11.3-13a 42 11.3-14 42
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t ACNCS-PSAR 1 EFFECTIVE PAGE LISTING (Cont'd) CHAPTR 11 Amendment No. l P_ age l 5 11.4-8 5
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*All figures whether labelled Unit 1" or, Units 1 and 2 should be considered ,
applicable to Ur.it No.1 Figure No. Amendment No. 11.1-1 - 11.1-2 - 11.1-3 37 l
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d ACNGS-PSAR l El'FECTIVE FIGURES LISTING (Cont'd) i l CHAPTER 11 Figure No. Amendment No. f Section 11.6 , Appendix A Plate 1 i Plate 2 - Plate 3 . i Plate 4 - Plate 5 Plate 6 - Plate 7 - Plate 8 - Plate 9 Plate 10 - t Plate 11 - f f 4 i l i i l l ) 1 i i 7 Am. No. 59, (6/81)
ACNCS-PSAR ; LIST OF EFFECTIVE PAGES CHAPTER 12
'N RADIATION PROTECTION g Amerdaent No. l 1* 59 2* 58 1 2a* 58 ,
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l ACNGS-PSAR I l EFFECTIVE PAGES LISTING (Cont'd) ' CHAPTER 12 Pg Amendment No. 12.2-21 35 12.2-22 35 12.2-23 35 12.2-24 35 12.2-25 35 12.2-26 39 1 12.2-27 39 12.3-1 35 12.3-2 35 12.3-3 35 12.3-4 35 12.3-5 - 35 12.3-5a 35 12.3-6 35 12.3-7 35 1 Unnumbered ( App 12A - Cover Sheet) 18 j Unnumbered (App 125 - Cover Sheet) 18 Unnumbered ( App 12C - Cover Sheet) 18 12C-1 18 12C.1-1 18 12 C.1-2 18 12C.1-3 18 12C.2-1 18 12 C. 2-2 18 12 C . 2-3 18 12C.2-4 18 12 C . 2-5 18 12 C. 2-6 18 12C.2-7 18 i 12 C. 3-1 18 12 C . 3-2 18 12C.3-3 18 12C.3-3a
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considered applicable to Unit No.1. Fi gure Amendment No. 12.1-1 37 ! 12.1-2 37 [ i 1241-3 37 , 12.1-4 37
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s j ACNG8-PSAR j EFFECTIVE FIGURES LISTING * (Cont'd) CHAPTER 12 i Figure Amendment No. j , 12.1-20s 37 i 12.1-20b 37 , 12.1-21 37 l 12.1-22a 37 ; 12.1-22b 37 l'2.1-22c 37 12.1-22d 37 12.1-22e 37 12.1-22f 37 l 12.1-23 (deleted) 37 l 12.1-24 _ 12.1-25 - l , 12.1-26 35 : I 12.2-1 _ i 12.2-2a 32 i 12.2-2b L 12.2-3 35 i 12.2-4 35 !
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I 4 l' I ACNGS-PSAR LIST OF EFFECTIVE PAGES 'O l i CHAPTER 13 CONDUCT OF OPERATIONS Page Amendment l 1* 59 2* 57 l 4 3* 59 4 4* 59 j 5* 59 i i 58 ii 55 iii 55 iiia 55
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ACNGS-PSAR LIST OF EFFECTIVE PAGES (Cont'd) CHAPTER 13 _Page Amendment 13.3-26 55 13.3-27 55 13.3-28 55 1 3 . 3- 29 55 13.3-30 55 13.3-31 55 13.3-32 55 13.3-33 55 13.3-34 55 13.3-35 55 13.3-36 59 13.3A-1 thru 13.3A-16 55 i-v (Appendix 13.3B) 55 1-49 (Apppendix 13.3B) 55 A-i 55 A-1 55 B-i 55 B-1 thru B-7 55 C-i 55 C-1 and C-2 55 13.4-1 33 13.5-1 42 13.6-1 16 13.7-1 33 13.7-2 33 13.7-3 33 13.7-4 33 4 Am. No. 59, (6/81)
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_ __ _ . ~ . , ACNGS-PSAR f 13.3.2.5' Planning Responsibility Houston Lighting & Power Company, the State of Texas and local counties : presently .have qualified personnel assigned responsibility for the development 57 of fixed nuclear facility emergency plans. The final ACNGS Emergency Plan, , and the State and loca'l plans approved by the NRC and FEMA, will describe these persons responsible for maintaining, reviewing, distributing and i j updating applicable emergency plans. HL&P will ensure that these ! responsibilities are executed. i . 13.3.2.6 Location of Key ACNGS Personnel During Alert or Greater ; Emergencies For Alert Emergencies, the Technical Support Center (TSC) will be activated to support - the Control Room. Other centers such as the Baergency Operations < Facility (EOF) and Operations Support ehnter (OSC) may be brought to standby status. All of these facilities are a2tivated for the Site Area and General 59 r Emergencies. Thus, many emergency functions will be performed in the TSC during Alert Emergencies and transferred to the EOF if the event escalates to a Site Area or General class event. I Table 13.3-2 provides, in matrix format, the location of key ACNGS personnel during Alert, Site Area and General Emergencies. l 13.3.3 EMERGENCY CLASSIFICATION SYSTEM Emergency conditions at ACNGS will be classified into four categories which will cover the entire spectrum of probable and postulated accidents. The four
- classes will be
Unusual Event Alert , Site Area Emergency General Emergency The Unusual Event and Alert categories are intended to provide early and prompt notification to the onsite and offsite emergency response organizations (see Section 13.3.2) that minor events have occurred or are in progress which could lead to more serious consequences if the plant status is in some way further complicated or which might be indicative of more serious conditions 55 which are not yet fully realized. The Site and General Emergency categories ' are intended for more severe situations, which indicate that significant offsite effects are likely and require immediate action from both onsite and offsite emergency response organizations. The identification of a Site or General emergency should be. assessed as quickly as possible and steps taken to I mitigate the event and its effects and returning the plant to a safe status. The decision to declare a particular emergency class is the responsibility of the Operating Supervisor at ACNCS, as Emergency Director. 'His decision will be based, to the extent feasible, on readiy available information about plant and offsite conditions which would indicate potential or actual hazards. The final ACNGS Emergency Plan will specify the criteria for declaring each emergency classification, as well as the provisions for upgrading the classification level and the corresponding response in the event of change of severity of the emergency condition. 13.3-11 Am. No. 59, (6/81) i
ACES-PSAR t [N
~13.3.3.1' Unusual Event Events within the Unusual Event class will represent off-normal conditions in .
and around the plant. Rese events will not, by themselves, constitute signi- [ ficant emergency conditions and will have no offsite radiological consequences. -
- Some of these events could, however, indicate a potential degradation in the level of phnt safety and/or could escalate to a more severe condition if i appropriate.onsite action is not taken. ,
he primary purpose of notifying of fsite agencies of an Unusual Event is to put the offsite response organization on standby, provide them with current j information and provide unscheduled testing of the offsite communication i link. An Unusual Event classification would initiate the augmentation of the onsite response resources to assist in the assessment and mitigation of the [ event. Recommendations will specify that no offsite actions are necessary. - t 13.3.3.2 Alert I . l f Events within the Alert class will indicate an actual or potential degradation , in the level of plant safety. The purpose of declaring this emergency class ;
.will be to assure that offsite emergency personnel and monitoring teams are ' ' 55 ready to respond if needed. his class may also serve as an unscheduled test i.
of the activation of the onsite and offsite emergency response facilities and l the related communication systems. He response of offsite agencies will be to bring key elements of the emergency response organization into standby I status, including offsite monitoring teams. This class of emergency will also initiate the activation of the Technical f Support Center and the Operations Support Center. he near-site Baergency Operations Facility will be brought to standby status. ! i 1 In addition to the manning of the onsite response facilities, this emergency I l class might require that radiological and meteorological assessments be made I j and reported to the offsite agencies. No public action would be recommended ' in this emergency class. As a precautionary measure, visitors to the Allens Creek Iake and State Park will be evacuated. i 13.3.3.3 Site Area Emergency ; i i Events within the Site Area Bnergency class involve actual or probable major failures of plant functions needed for protection of the public. he purpose I of declaring the Site Area Emergency class is to assure the manning of all
- emergency response facilities, the dispatching of monitoring teams and the
- assembling of personnel required for evacuation if such action becomes l- necessary. Declaration of this emergency class will also be used as a means
[ of informing the offsite agencies and the public that significant events are
- taking place. %e response of of faite agencies following-such a declaration will be to consider implementing protective actions and to assess information i 13.3-12 Am. No. 59, (6/81) i
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ACNCS-PSAR N the public. IIL&P will erk closely with appropriate State and local governuents to ensure that, before the ACNGS begins operation, the capability will exist to notify the public in the 10-mile EPZ within 15 minutes of noti-fication to of f site organizattuns. An evaluation will be made to determine the specific system which will provide this capability. This system may in-clude st rens, in-residence tone alert devices, alarms connected to electric meters and/or multiple telephone call-up techniques. Arrangements will be made f or broadcastind emergency instructions to the pub-lic via radio and/or television following the initial notification. The pub-lic will have received prior information as to how they will be notified in the event of an accident and what protective actions might be taken. 55 13.3.5 EMERGENCY RESPONSE FACILITIES i Emergency f acilities, as well as special systems, will be established at and near the ACNCS for assessing an event, directing response and recove ry ef forts, mitigating accident consequences and informing the public. The facilities and systems will be described in detail in the final ACNGS Eme rgency Plan. 13.3.5.1 Cont rol Room During an emergency, the Control Room will be the location in which actions
. ~ . are taken primarily to bring plant systems under control. The Control Room is
[ ) the location in which an accident event is initially recognized, classified and assessed, and notification procedures initiated. The Control Room will \s_,/ have communications with all other emergency f acilities and will be equipped with terminals of data systems. The Cont rol Roam, being inside the plant, is designed as a totally safety grade f acility. 13.3.5.2 Technical Support Center An Onsite Technical Support Center (TSC) will be provided. Details of the TSC I59 a re described below. 13.3.5.2.1 TSC Function 57
*e functions of the TSC are to:
a) Provide a location for plant management and technical support pe rsonne l to work to support operations personnel during emergency conditions, b) . eform EOF functions durir,g emergencies requiring EOF activation until t hs EOF is fully manned and functional. t D) 13.3-17 Am. No. 59 (6/81) ,
m ACNr;S-PSAR It is expected that the TSC complex will be used for routine plant functions. l I For example, the NRC of fice may be the Resident IE Inspectors office and the 57 work area with its displays may be used for training. Such uses will not interfere with rapid activation of the TSC for its emergency functions. 13.3.5.2.2 TSC Iocation The TSC is located on the northeast side of the Control Building at el.187', as shown on Figure 1.2-3. Walking time between the TSC and Control Room is i 59 i well under two minutes. The two areas and the hall between them are in the
- same ventilation envelope as described in Section 13.3.5.2.6, so there would be no need to don protective gear to pass from one area to the other. There are no major secuity barriers between the two areas.
i 57 13.3.5.2.3 TSC Staffing and Training his will be provided in the ACNGS Final Baergency Plan. f 13.3.5 2.4 TSC Size f 4 The TSC is a complex consisting of the following directly adjacent areas. A ) working space of approximately 75 sq. ft. per person was used as a basis for j the layout of the TSC. The design meets the minimum requirements of 25 i persons as required by NUREG-0696, February 1981. The layout meets the
! minimum requirements for square footage per person, and the location of the TSC (adjacent to the Control Complex) adds additional benefit.
j a) Open working area for 22 people of approximately 1580 sq. ft. divisible i into up to 5 separate rooms by the use of moveable room divideris. This area also contains the SPDS displays and other plant data displays. These displays, as a minimum, will be Type A, B, C, D, and E variables specified in Regulatory Guide 1.97 Rev. 2 and meteorological variabbs as specified in Regulatory Guide 1.23. i 59
. Copying equipment and plant data displays are located in this work ; area. A portion of this work area is available to be separated out as a conference roo.a, if required. The remaining work areas will each l' have a data display to allow addressing the parameters in the data acquisition system. This area will be used by technical personnel to 4 enable them to carry out their function of supporting the operators in the Main Control Room.
b) Office for NRC representatives of approximately 150 sq. f t. This area j_ provides working space for 2 NRC people as well as being used for l private NRC consultations. 1 Data is available to NRC personnel, as well as other TSC personnel, in i any of the sections. of the work area described in (a) above. Hard I copies of any display can be made by the video copiers or the ERIS data
.J 13.3-18 Am. No. 59, (6/81) i
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, ,/'~') acquisition system can ce requested to print information on the printer located in time work area. c) Commanications equipment room of approximately 158 sq. f t. Tne area is normally manned by two people.. These persons will perform tasks su6h i as teletype operations, telephone operator functions, and secretarial ; functions.
.c d) Document Storage Room of 180 square f eet.
e) Living Area, including kitchen, sleeping area, supply storage, and , repair shop f or storing tools and spare parts for the TSC instruments 59 f displays. This area is 900 sq. f t. and is not considered part of the TSC working area. , f) Secondary Alarm Station. This station will be continuously operational. 13.3.5.2.5 TSC Structure l The Control Building in which the TSC is located is designed to withstand the full range of natural phenomena specified for safety-related structures for ACNGS. This exceeds the structural requirements f or the TSC. l t 13.3.5.2.6 TSC Haoitability 57 Ine TSC is served by the Control Room Ventilation System, so it is haoitable
. N to the 'ssue extent as the Control Room. See Section b.4.
13.3.3.2.7 TSC Commanications Tne TSC will be provided with reliable voice communications to the Control Room, OSC, EOF and NRC. Details of TSC communications will be provided in the ACNGS Final Emergency Plan. I J. J . 5. 2. 8 TSC Instrumentation, Data system Equipment and Power Supplies Plant data will be available for display in the TSC. The set of parameters to ' be displayed in the TSC has not been finalized, as HL&P intends to abide by the results of the BWROG efforts in this regard when approved by the NF.C. As ' a minimum, the SPDS (see Section 7.5.1.6) and post-accident monitoring instrumentation (see Appendix C) will ne available in the TSC. This three located in the information Main Work Area will be anddisplayed one in theon a minimum TSC conferenceof room, f our CRT's,igure (see F 1.2-33 and 59
, 1.2-35. In addition, a hard-copy printer will be available' in the Main I
, Working Area, as well as two video copiers. The TSC displays are not Class IE or seismically qualified, but are provided with reliable backup power from the BOP diesel generator. 1 I. L (~
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\_ / 13.3.5.2.9 Records Availability and Management The TSC will have a records storage area in which up-to-date plant records deemed useful for accident or of f-normal condition diagnostics will be stored. A list of the types of these records will be provided in the ACNGS Final Emergency Plan. 13.3.5.3 Emergency Operations Facility An Emergency Operations Facility will be provided. Details of the EOF are described below. 13.3.5.3.1 EOF Function The ACNGS Emergency Operations Facility (EOF) will be controlled and operated by HL&P and will serve as the location for performing the functions of: a) management of overall emergency response, b) coordination of radiological and environmental assessment , c) determination of recommended public protective action, and d) coordination of emergency response activities with Federal, State and s local agencies. \s./ The EOF will be activated for Site Area and General Emergencies, and will be brought to standby status _ for the Alert Bnergency. 59 It is anticipated that the EOF will be used as the post-accident recovery center. The EOF space may be used for other purposes during normal operations, however, provisions will be set forth to assure that the emergency functions of the EOF are in no way degraded by those activities and that all necessary systems meet required availability. These provisions will include adequate security protection of the facility during normal and emergency conditions. 13.3.5.3.2 EOF Location The location of the EOF will be described in more detail in the ACNCS Final Bnergency Plan. In selecting the actual location, consideration will be given to the following factors: a) Whether the location provides optimum functional and availability characteristics for carrying out the licensee functions specified for the EOF (i.e., overall strategic direction of licensee onsite and support operations, determination of public protective actions to be ps \ N. j 13.3-20 Am. No. 59, (6/81)
ACNGS-PSAR n)- ( s N- '/ recommended by the licensee to of tsite of ficials, and coordination of the licensee with Federal, State and local organizations) b) Whether the EOF f unctions would be interrupted during radiation releases for which it was necessary to recommend protective actions for t
. the puDlic to of fsite of ficials.
The siting of the EOF will be coordinsted with State and local authorities. It is anticipated that the EOF will be located on the site of HL&P's Service Center at Katy, Texas (Figure 13.3-1). The location is on property owned and controlled by HL4P and is approximately 19 miles from the ACNGS. Inis location is easily accessible from the Interstate 10 Highway. Presently under construction at the Katy Service Center is an additional 15,000 square feet of office and conference room space. A plot plan of the i service' center is shown on Figure 13.3-2. The present service center of fice will be modified to include a coffee bar/ meeting room of about 875 square feet. The additions and modifications are scheduled f or comp'letion by early 1982. A conceptual floor layout for the EOF using the modified Katy Service Center 59 is shown on Figu res 13.3-3 and 13.3-4. Should the location of the EOF be changed at a later date, the new location
- and its description will be submitted to the NRC for approval prior to making D7 the change.
13.3.5.3.3 EOF Staffing and Training When tne EOF ts activated, it will be staffed by HL&P, Federal, State, Local and otner emergency personnel designated by the ACNGS Emergency Plan. A designated senior HL&P official will manage licensee activities in the EOF to support the HL&P otficial managing activities in the Technical Support Center and the senior reactor operator serving as shif t supervisor in the Control Ro om. The EOF will be staf fed to provide the' overall management of licensee resources and the continuous evaluation and coordination of licensee activities during and af ter an accident. Upon EOF activatica, designated personnel will report direct ly to the EOF to achieve full functional operation within I hour. The EOF staf f will include personnel to manage the licensee onsite and of fsite radiological monitoring, to perform radiological
- evaluations, and to interface with of fsite officials. The specific number and type of personnel assigned to the EOF may vary according to the emergency
-class. The staffing for each emergency class shall be fully detailed in the
- ACNGS Final Emergency Plan. Operating procedures and staf f training in the use of data systems and instrumentation will contain guidance on the limitations of instrumentation including whether the information can be relied i. A l l i 13.3-20a Am. No. 59, (6/81)
ACNGS-PSAR jy s- upon following serious accidents. The EOF staff will participate in EOF activitia c drills, conducted periodically in accordance with the Emer;;ency Plan. These drills will include operation of all facilities that will be used .
- to perform the EOF functions. -
13.3.5.3.4 EOF Size
- l The EOF building or building complex will %e large enough to provide the following:
a). Working space for the personnel assigned to the EOF as specified in the ACNGs Final Emergency Plan, including State and local agency personnel. A working space of approximately 75 sq. f t. per person will be used as a basis for size and layout of the EOF. The conceptual EOF layout provided on Figure 13.3-3 assumes approximately 25 persons from the licensee,10 persons from State and Local agencies, 9 persons from the NRC and 1 person from FDIA. b) Space for EOF data system equipment needed to transmit data to other locations; c) Sufficient space to perform repair, maintenance, and service of 59 equipment, displays, and instrumentation; d) Space for ready access to communications equipment by all EOF personnel who need communications capabilities to perform their functions; e) Space for ready access to fut.ctional displays of EOF data; f) Space for storage of plant 'ra. c rc'a and historical data or space for means to readily acquPS ,5t <1 splay those records; g) A separate room to at; same; . > at least 5 NRC personnel will be provided; h) Conference rooms; i) A space to brief sclect groups of about 25 per-ons, such as a press pool or government officials; j) A secured entrance k) Sufficient space outsidt for parking private vehicles, mobile laboratorfes and trailers. Electrical power and sanitary hook-up will be available for mobile laboratories and trailers.
'(
8 s x/m 13.3-20b Am. No. 59, (6/81)
ACNE,S-PSAR A 13.3.5.3.5 EOF Structure (v) The EOF structure will be well engineered for the design life of the ACNGS, in accordance with the criteria in NUREG-0696, Table 2. , 13.3.5.3.6 EOF Comn2nications The EOF will have reliable voice comm2nications f acilities to the TSC, the control room, NRC, and Sta'te and local emergency operations centers. The normal commanication path between the EOF and the control room will be through i th e TSC. _ The primary functions of the E05 voice communications facilities will be: 1 a) Comn2nications with the designated senior licensee manager in charge of the TSC, I b) manage licensee emergency response resources, c) coordinate radiological monitoring, 59 d) coordinate offsite emergency response activities, and e) disseminate'information and recommended protective actions to responsible government agencies. The EOF voice communications facilities will include reliable primary and frs i b.tekup means of commanication. Voice connanications may include private s_,/ telephones, dedicated telephones, commercial telephones and radio. A means for ECF telephone access to commercial telephone services that bypasses any local telephone switching facilities that may be susceptible to loss of power during emergencies will be provided. Spare commercial telephone lines to the ple.ac will be available. The EOF comnanication system will include designated telephones for use by NRC personnel. Tne licensee will also furnish the access facilities and cables to the NRC f or the Emergency Notification System (ENS) and the Health Physics Network (HPN) telephones. 4 Facsimile trenssission capability between the EOF, tne TSC, and the NRC Operations Center shall be provided. 13.3.5.3.7 EOF Instrumentation, Data System Equipment, and Power Supplies The Emergency Response Information System (ERIS) is a data acquisition system that provides data acquisition and information display capabilities in various locations such as SPDS TSC, EOF, OSC, and other of fsite facilities. The EOF will contain equipment for the acquisition, display, and evaluation of all radiological, meteorological and plant system data necessary to determine protective measures recommended to of fsite authorities. This equipment will also be used to evaluate the magnitude and effect ot actual or potential radioactive releases and to project of fsite doses. fs 4 13.3-20c Am. No. 59, (6/El)
ACNCS-PSAR
/
Data will be transmitted to the EOF from either the existing plant process (V) computer system or a separate computer-based system, using CRTs located in the An evaluation is currently underway to determine which data acquisition EO F. system will be used. The evaluation is scheduled to be completed prior to the sof tware development for the existing plant process computer system. Video graphics will be utilized to provide graphic displays of necessary plant status information. Video copier (s) will be provided in the EOF to ensure
-that hard copiec of video displays can be obtained when needed.
The data acquisition system will gather, store, and display the data needed in the FP to analyze and exchange information on plant conditions with the designated senior licensee manager in charge of the TSC. The system will perform these functions independently from actions in the control room without degrading or interfering with control room and plant functions. Trend-information display and time-history display capability will be provided. The SPDS will be displayed in the EOF. The EOF data set shall include radiological, meteorological, and other environmental data as needed to: a) Assess environmental conditions, b) Coordinate radiological monitoring activities, and 59 m c) Recommend implementation of offsite emergency plans.
~
As a minimum EOF data set, sensor data of the Type A, B, C, D, and E variables specified in Regulatory Guide 1.97, Revision 2, and of those meteorological variables specified in proposed Revision 1 to Regulatory Guide 123,
" Meteorological Measurements Programs in Support of Nuclear Power Plants," and in NUREG-0654, Revision 1, Appendix 2, will be available for display in the EOF. All data that are available for display in the TSC, including data transmitted from the plant to NRC, will be part of the EOF data set. The accuracy of data displayc in the EOF will be equivalent to that for the data displayed in the TSC.
Data storage capability will be provided for the EOF data set. The sample frequency will be chosen to be consistent with the use of the data. Capacity to record at least two weeks of additional post-event data with reduced time resolution will be provided. A sufficient number of data display devices shall be provided in the EOF to allow all EOF personnel to perform their assigned tasks with unhindered access of: Plant systems variables, In plant radiological variables,
- Meteorological information, and
- Offsite radiological information.
v 13.3-20d Am. No. 59, (6/81)
ACNGS-PSAR f ? The total EOF data system shall be designed to achieve an operational unavailability goal of 0.01 during all plant operating conditions above cold shutdown. The EOF electrical equipment load will not degrade the capability or reliability of any safety-related power source. Circuit transients or power supply failures and fluctuations will not cause a loss of any stored data vital to the EOF functions. 13.3.5.3.8 Records Availability and Management i The EOF will have ready access to up-to-date plant records, procedures, and ' emergency plans. The EOF records will include, but shall not be limited to:
- Plant technical specifications.
- Plant operating procedures,
- Energency operating procedures,
(
' - Final Safety Analysis Report,
- Up-to-date records related to licensee, State, and local emergency response plans, 59
- Of fsite population distribution data.
- Evacuation plans,
- Environs radiological monitoring records, and
- Licensee employee radiation exposure histories
- Up-to-date drawings, schematics and diagrams showing conditions of plant structures and systems down to the component and in-plant locations of these systems.
These records will either be stored and maintained in the EOF (such as hardcopy on microfiche) or shall be readily available via transmittal to the EOF from another records storage location. 13.3.5.4 operations Support center 55 Appropriate space will be designated onsite for the assembly of operations personnel whose support is required in or near the plant, but not in the Control Room or TSC. The preliminary location of the OSC is in the personal 59 access building as shown on Figures 1.2-38a and 1.2-38b. Supplies such as protective clothing, respiratory protection, portable lighting and l57 ! communications equipment will be provided. The OSC will be described in 55 detail in the final ACNGS Baergency Plan. 13.3-20e Am. No. 59, (6/81)
m. ACNGS-PSAR f] 13.3.5.5 News Media Center HIAP will provide a location at or near the EOF to serve as a News Media Center (letC) in which to conduct press conferences and briefings during an ene rgen%y. The !#fC will be activated for the Site Area and General Energency levels and will be brought to standby status for the Alert Bnergency. He 14fC will be large enough to accommodate 300-400 news media representatives. A backup location outside the plume exposure EPZ will be available, such as an auditorium or civic center. A small briefing room will be made available in the EOF in which to conduct briefings with small select groups. In the'19fC, information packets or " press kits" will be available providing informatiori 55
-about the licensee, the plant and plant surroundings. Visual aids will be provided.
HIAP will designate an official company spokesperson to interf ace with the l news media and the spokespersons of offaite response organizations. HIAP j spokespersons will be trained in conducting press conferences and briefings i and will be knowledgeable of plant operations and the Bnergency Plan. l 13.3.5.6 - Safety Parameter Display System i The SPDS is described in Section 7.5.1.6. 57 13.3.5.7 Data Transmission ~55 The ACNQi will be equipped with the capability to transmit plant data to the ( g TSC and EOF. %e design of this system will be described in the final AQiGS Bnergency Plan. 57 13.3.5.8 First Aid Facility A first aid room equipped with the first aid equipment and supplies which are appropriate for a major industrial _ facility will be provided at the AQiGS. At least one individual onsite will be trained and qualified in advanced first aid methods. 13.3.5.9 Decontamination Facility ! 55 Personnel decontamination facilities, consisting of showers and sinks which drain to the radwaste system, will be provided. ACNGS personnel will be trained in decontamination methods. First aid to injured individuals will, in most cases, be performed in conjunction with any necessary decontamination. However, if immediate treatment of the injury is 4eemed necessary, that treatment will take precedence over decontamination. This philosophy will also extend to transportation and offsite treatment of contaminated, injured individuals. 13.3-20f Am. No. 59, (6/81)
f ACNGS-PSAR TABLE 13.3-2 a LOCATION OF KEY ACNGS PERSONNEL DURING ALERT OR GREATER EMERGENCIES l Baergency Class Pers onnel Alert Site Area General A. Bnergency Director TSC EOF EOF Bl. Radiation Protection Mgr. TSC TSC TSC B2. Radiological Emergency Mgr. (EOF) EOF EOF C. Public Af fairs Spokesperson TSC EOF EOF D. Recovery Mgr. TSC EOF EOF E. Vendors, A-E's, Construction TSC(EOF) TSC or EOF TSC or EOF 59 F. Site Support Mgr. TSC(EOF) EOF EOF C. Consultants, Mobile labs, etc. TSC(EOF) EOF EOF
\j H. Fire Brigade, Damage Control, OSC OSC OSC s Repair, etc.
I. Federal, State & Local Agencies TSC(EOF) EOF EOF J. NRC Site Team TSC(EOF) TSC & EOF TSC & EOF Note: Centers in parentheses indicate that this center is normally not l activated for the emergency class, but may be activated for use by the personnel if necessary.
- s 13.3-36 Am. No. 59, (6/81)
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l I r 1 Am. No. 59 (6/81) ' L
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ACNGS-PSAR , 4 l l 1 LIST OF EFFECTIVE PAGES l
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ACNCS - PSAR EFFECTIVE PAGE LISTING (Cont'd ) , CHAPTER 15 Page Amendment No. 15.1-36 35 15.1-37 42 15.1- 18 35 15.1- 35 15.1-40 35 15.1-41 35 15.1-42 35 15.1-43 35 15.1-44 35 15.1-45 35 15.1-45a 35 15.1-46 35 15.1-47 35 15.1-48 35 15.1-48a 35 15.1-49 35 15.1-50 35 15.1-50a 35 15.1-50b 35 15.1-50c 35 C 15.1-51 15.1-52 35 35 15.1-53 35 l 15.1-53a 35 15.1-54 35 15.1-55 42 15.1-56 42 15.1-57 1 15.1-58 39 15.1-59 15.1-60 15.1-61 l 15.1-62 - 15.1-63 15.1-64 39 15.1-65 35 15.1-65a 35 15.1-66 35 15.1-67 35 15.1-68 35 15.1-69 28 15.1-70 35 15.1-70a 35 15.1-70b 35 15.1-70c 12 4 15.1-70d 35 15.1-71 35
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ACNGS - PSAR EFFECTIVE PAGES LISTING (Cont'd) CHAPTER 15 Page Amendment No. 15.1-llla 28 15.1-111b 28 15.1-111c 28 15.1-llld 28 15.1-112 28 15.1-113 35 15.1-113a 35 15.1-ll3b 12 15.1-113c 12 15.1-114 35 15.1-115 35 15.1-116 35 15.1-117 35 15.1-118 - 15.1-119 - 15.1-120 28 15.1-121 28 15.1-122 35 15.1-123 35 9 15.1-124 15.1-125 15.1-126 35 35 35 15.1-127 35 15.1-128 35 15.1-129 35 15.1-130 41 15.1-131 41 j 15.1-132 35 15.1-133 41 15.1-134 (deleted) 41 15.1-135 35 15.1-136 35 15.1-137 35 15.1-138 35 15.1-139 35 ' 15.1-140 35 l 15.1-141 35 15.1-142 35 15.1-143 35 15.1-144 35 ! 15.1-145 35 ; 15.1-146 41 ; O . 1 l l 4 Am. No. 59, (6/81)
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- ACCIDENT ANALYSI3 i 1 t
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i ACNGS- PSAR !O i EFFECTIVE PAGES LISTING APPENDIX 15B { 1 j ALLENS CREEK NUCLEAR GENERATING STATION i 8' 3ELIABILITY ANALYSIS PROGRAM 4 4 pAES Amendment No. t P l ' 15B- 1 59 15B-2 59 ! i 15B-3 t 59
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ACNGS-PSAR APPENDIX 153 L! ALLENS CREEK NUCLEAR GENERATING STATION V RELI ABILITY ANALYSIS PROGRAM 57
1.0 INTRODUCTION
-' i ltem II.B.8(1) of the " Proposed Licensing Requirements for Pending Applications for Construction Permits and Manufacturing Licenses" (NUREG 0718) requires that a site / plant specific probabilistic risk assessment be performed, and that the results be evaluated and considered in the design of the facility.
The' following' discussion describes the proposed pgram to be applied to the 59 Allens Creek Nuclear Generating Station ( ACNGS) in response to NUREG 0718 includirg an outline of the program scope, methodology, schedule, quality assurance procedures and the means by which the reliability analyses will be 57 integrated into the ongoing design process. , 2.0 PROGRAM OBJECTIVE i 59 he aim of the Reliability Analysis Program is to seek improvements in the i reliability of core and containment heat removal systems that are significant and practical and do not impact excessively on the plant. The reliability Analysis Program will be specifically structured to meet this objective, recognizing the advanced state of design (70% engineering complete) and fabrication (major plant components fabricated and in storage). ' 3.0 SCOPE
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The ACNGS Reliability Analysis Program is similar in scope to the Interim Reliability Evaluation Program (IREP) being performed by NRC for several oper- 57 atirg plants,except that it will include the calculation of fission product release quantities for the accident release categories. It will focus on core 1 and containment cooling systems together with all pertinent support systems. !59 Individual accident sequences and their probabilities will be analyzed to 1 identify the initiating events and plant system and component failures which I57 are the dominant contributors to core damage risk. Event tree and fault tree analyses will be perf ormed for core and containment cooling systems and all l59 g support systems to identify common-mode failure mechanisms. ' he list of initiating. events to be considered will be determined durirg the initial phase of the study. However, as a minimu=, the initiating events will 57 encompass loss of coolant accidents (small, intermediate and large) and tran-sient events including loss of feedwater, loss of of fsite power and turbine trip. _ We interdependence of support systems will be considered with any initiating event that may lead to of fsite radiological releases in excess of 59 10CFR100. We following key safety-related systems will be included in the
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program: 't h ' ( /. Reactor Core Isolation Cooling System
.. Residual Heat Removal Cooling System
. High Pre ssure Core Spray Cooling System
. Low Pressure Core Spray System
. Automatic Depre ssurization System-
. Containment Spray System 57
. Containment Isolation System
. Reactor Protection System
. Electric Power (AC and DC) System
. Essential Service Water System
. Essential 01111ed Wa ter System
. Essential HVAC Systems
. Standby Liquid Control System Other systems are expected to be added to the above representative list as the core and containment cooling systems are evaluated with respect to interactions and dependencies.
HL&P has identified an alternative heat removal system based on a study recently performed to determine design features which would reduce the pro bability of occurrence of a degraded core. Results of the PRA study will be inspected to determine whether an obviously superior alternative is feasible. ( - A preliminary outline of the report is given in Table 15B-1. 4.0 PROGRAM ORGANIZATION AND RESPONSIBILITIES 59 HL&P is responsible for the PRA study and will ensure that the study will be
. performed by engineers who are highly qualified and experienced in risk -
assessment methodology. HIAP engineers will be actively involved in this study and will provide direction in the development of the program. Prior to
. decisions relating to identified design improvements, HL&P will appoint a third party to conduct a peer review on the study.
HL&P retains ultimate audlority and re sponsibility of implementing design improvements as a result - f this study. 5.0 METHO DOLOGY The approach to be used in the program will employ event tree / fault tree methodology similar to that used in WASH 1400 and other comprehensive plant 57 risk studies. The major tasks involved are discussed below. nv 15B-2 Am. No. 59, (6/81)
ACNGS-PSAR 5.1 INITIATING EVENT SELECTION
) A list of initiating events will be established that, together with system failures, have the potential f or causing core damage and of fsite radioactivity releases. This will be accomplished th rough a screening of the accidents and transients identified in PSAR Chapter 15 and in WASH 1400 to identify the basic set of initiating events requiring operation of the key safety systems for core 57 protection and release mitigation. The frequency of these initiating events will be estimated based on available data including WASH-1400, EPRI NP-801, and, where it exists, pertinent plant specific or site-specific information (e.g., frequency of loss of of fsite power on the HL&P system) .
Initiating events will be classified according to the safety system response required for accident mitigation. Initiating events hi.ving the same or similar safety system response requireuents will be grouped together and a unique event tree will be developed for each such group of events. Initiating events including LOCA 3 transients and recirculation line breaks will be evaluated. 59 Failures during cold shutdown, severe natural phencaena and fire will also be considered. 5.2 EVENT TREE DEVELOPMENT For each unique group of initiating events, an event tree will be con- l structed, identifying the safety systems required to mitigate the event and h57 the expected ef fect on ability to maintain core and containment integrity given success or failure of each safety system involved. The full event tree ,g, will be reduced to reflect safety system interdependencies and required se-( ; quences of operation. L ,' 5.3 SYST&! FAILURE MODES AND EFFECTS ANALYSIS (FMEA) '59 For each safety system involved, a FMEA will be conducted to identify and tab-ulate component and common cause failures and their ef fect on system operabil-ity for each initiating event. The FMEA will provide documentation of the basis for f aelusion or exclusion of specific failure modes in the system fault 57 tree analysis. Failure modes will include mechanical and electrical faults, operator error, maintenance or testing outages, etc. Particular attention will be paid to potential common cause failures which could disable multiple compo-nents. Common cause failure mechanisms to be investigated include environmen-tal factors, operator or maintenance errors, passive failures and system inter-actions. 5.4 SYSTEM FAULT TREE ANALYSIS (FTA) :59 Us ing the FME A as input , fault trees will be constructed for each safety system identifying the f ailures (basic events) and their logical combinations which will re sult in system unavailability (t op event) . The fault tree will be 57 analyzed to determine the minimal cut sets and failure combinations which are the dominant contributors to system unavailability. Using the appropricte com-ponent f ailure data, a quantitative assessment of overall systcm unavailability and of' the dominant cut sets will be perf ormed. n I \
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/ - The fault tree analysis will be performed using a computer program to perform
~\ ,j cut set determination and quantitative analysis and to provide computer-
- generated graphical representation of the fault tree using standard logic 57 symbols.
~5.5 D ATA BASE DEVEIDPMENT j 59 l i
l A component failure data base for' use in system fault tree analysis will be !
-. developed from recognized reference sources including WASH-1400 and IEEE-500.
In addition, prototype-specific failure data will be requested from vendors of , selected components (e.g. , diesel generators) being supplied to ACNGS. The ! data base will identify the types of components and estimated median failure { rates on demand and, where appropriate, per hour of continuous operotton. 57. j Error ranges will be assigned to each median value to re flect the uncert,ainty in the data base. The data base will include methodologies to adjust failure ; data to account for varying testing and' surveillance strategies. Test and i j' maintenance unavailability contributions will be included based on proposed Technical Specifications operating and maintenance procedures. I I Human error rates will be estimated for required or corrective actions by , l control room operator and for maintenance- or testing operations which are . I included as failure modes in the system fault trees. Available human error I e and performance data, including those provided NUREG/CR 1278 will be used. ! 6 L 5.6 ACCIDENT SEQUENCE Pa0BABILITIES 59 I The unavailability of each system will be calculated by inputting the i appropriate failure rate data into the system FTA. The various accident sequences,' as represented by the branches on the event trees will then be i quantified by inputting the system failure probabilities determined from the 57 - quantitative FTA. Each individual accident sequence will be classi-fled according to release category and the total probability of a given ' release category will be obtained by the summation of all accident sequence i probabilities assigned to that category. ; i 5.7 ACCIDENT SEQUENCE FISSION PRODUCT RELEASES l 59 j The release categories of WASH-1400 will be re-examined and modified if neces-l sary to account f or any changes due to the Mark III containment configura- l tion. ' A radiological release source term to the environment will be calcula-ted for each release category. 57 I J J t f 15B-4 Am. No. 59, (6/81)
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\ ~- All quantitative results will be reported in terms of point values of a prob-ability distribution function, including expected (mean) or median (50th per- 57 centile) value and upper (95th percentile) and lower (5th percentile) uncer-tainty bounds. These point values will be detecmined based on a propagation 1 59 uf component failure data, including er ror ranges, th ro ugh the f ault trees and event t ree s. The uncertainty propagation will be performed using standard 57 statistical d istribution f unctions (e.g. lognormal) or numerical (e.g. Monte Carlo) technique s.
5.9 SENSITIVITY ANALYSIS l59 The re sults of the study will be reviewed to ider.tify the accident sequences l which are the dominant contributors to overall risk and, within those se- i q ue nc es, the significant system and component failure modes. Comparisons with existing risk studies, including WASit-1400, will be made to identif y and explain any significant differences. The senqitivity of the re sults to assumptions regard ing component or common 157 cause '.ailures will be evaluated by varying the assumed failure rates of key basic events which appear with a high frequency in the dominant event sequen-ces and determining the resultant effect on system failure rates and overall re su lt s. 6.0 SCHEDULE
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( ) The Reliability Analysis Program will commence in mid-1981. The initial phase l
\ ~- / of the program is expected to take approximately 15 months, as shown on Figure !
15.B-1, and will consist of a base line reliability analysis of the present '57 ACNGS design. The overall study, ir.cluding radionuclide release quantifica-tion, will be completed withJ - e5r years of CP issuance. The overall design of ACNGS is approximately 70% complete as of May 1981. Fabrication of the NSSS is rapidly nearing completion. Similarly, supporting i balance of plant major features are in fabrication and would be expected to be ! essentially completed prior to completion of the PRA study. l
- 59 If the ACNGS construction permit is received in the March-June 1982 period , !
the PRA study would be completed about March-June 1984. Design and fabrication will by then be essentially complete f or core and containment heat removal systems and support systems. Ho wever , the re sults will be used on a case-by-case basis to determine whether major redesign, repurchasing, and refabrication are warranted. A
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ACNGS-PSAR 7.0 59 APPLICATION OF RESULTS TO FINAL DESIGN TU ) here are ~ currently no established regulatory requirements or acceptance 57 criteria for judging the acceptability of quantitative system reliability analyses. Sus the need for implementing changes in design or operating, testing or maintenance procedures to achieve improvements in system reliability will be based on judgmental acceptance criteria which are not l 57 i directly related to licensing requirements. nese acceptance criteria will be established during the initial phase of the program and will include both quantitative and qualitative considerations of potential design changes on plant cost, schedule and availability. Followtg completion of the' base line reliability analysis, the results will 57 be retiewed and various options available f or improvement in reliability will ' be evaluated with respect to the established acceptance criteria. Re commend-ations will be made regarding changes in design or operating procedures and the reliability analysis will be revised to reflect those selected for imple-mentation. i Routine design changes will also be evaluated on an ongoing basis. A determin-ation will be made regarding the effect of any proposed design change on the reliability analysis results. If the change is expected to affect reliability, the results will be reviewed for acceptability and need for 59 further modifications determined. In this manner, the Reliability Analysis Program will be kept current with respect to design modifications and a mechanism will be in place to evaluate reliability related changes for [ acceptability as the design is finalized. 57
'(j Results of the study will be utilized to improve reliability of component .
selection, specification and testing and to improve systems interaction. Results will also be used to identify improvements to be considered in the future for the following areas: maintenance procedures, operator training and operating feedback. j , 59 . 8.0 QUALITY ASSURANCE l Results of the study will be used to identify those areas where additional
- quality assurance activity would improve reliability. He results of the program, including all calculations will be subject to review and verifica-tion in accordance with normal ACNGS quality assurance program practices.
57 j Documentation will be maintained current so that all results can be reproduced and all assumptions checked from original references. 1 [ 1
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ACNGS-PSAR TABLE 15B-1 [ r OUTLINE OF RELI ABILITY ANALYSIS REPORT
.'v I. INTRODUCTION ,
II.
SUMMARY
III. )ETH0DG.0GY OVERVIIN A. Event Trees B.- Fault Tree s C. Quantification of Accident Sequences D. ' Containment Failure Analyses E. Fission Product Release Analyses F. Treatment of Uncertainties IV. SYSTDI DESCRIPTIONS 4 59-A. Performance Requirements B. Actuatio n C. Environment Considerations . D. Dependency Diagrams for Support Systems V. CORE MELT PROBABILITIES A. Dominant Sequences B. Dominant Cut-Sets I VI. PLANT MODIFICATIONS THAT ADDRESS DOMINANT SEQUENCES A. Improvement in Reliability Expected B. How Factored into Design, Equipment Purchase, Fabrication, i Procedures, Operation, etc. C. Basis for Not Implementing More Reliable Alternatives 3 a i 15B-7 Am. No. 59, (6/81) i j
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ACNCS-PSAR- -. i . !: TABLE 15B-1 (Cont 'd): -! i t f: VII. FISSION PRODUCT RELEASE ANALYSIS-i 6'
=A. Release Groups l B. Containment Failure Probabilities a
i- C. Fission Product Release Fractions i . Sf- > l D. Total Radioactive Release from Containment to Environment i i for the Various Release Gioups. i i l j , VIII. APPENDICES (DETAILS OF- STUDY)
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( ) TABLE OF CONTENTS V CHAPTER 17 OUALITY ASSURANCE Section Title Pag e 17.0 QU ALITY ASSURANCE 17.0-1 17.0.A HOUSTON LIGHTING & POWER COMPANY 17.0-1 17.0.B EBASCO 17.0-2 17.0.C GENERAL ELECTRIC 17.0-3 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 17.1-1 17 1.1A ORGANIZATION 17.1-1 17.1.1A.1 Manager, Quality Assurance 17.1-5 17.1.lA.2 Project Quality Assurance Manager, Allens Creek 17.1-5
/'~'N 17.1.lA.3 Houston Quality Assurance Manager 17.1-6 '\
17.1.1 A.4 Project Quality Assurance General Supervisor 17.1-7 17.1.lA.5 Supervisor, Quality Systems. 17.1-7 17.1.1A.6 Discipline Project Quality Assurance Supervisors 17.1-7 1 17.1.1A.7 Procurement Project, Quality As surance Supervisor 17.1-7 17.1.lA.8 fanager, Allens Creek Project 17.1-7 17.1.1A.9 Project Engineering Manager 17.1-8 17.1.lA.10 Supervising Project Engineer (s[ 17.1-8 17.1.1A.11 Project Construction Manager 17.1-8 17.1.1A.12 Project Purchasing Manager 17.1-9 17.1.1A.13 Project Controls Manager 17.1-9 17.1. l A.14 Project Administration Supervisor 17.1-9 17.1.1A.15 Project Controller 17.1-9 ()N_ /
- 17.1.1 A.16 Project Environmental Engineer 17.1-9 17.1.1A.17 Project Nuclear Fuel 17.1-9 i Am. No. 59, (6/81)
ACNGS-PSAR TABLE OF CONTENTS (Cont'd) Section Title Pag e 17.1 MA QUALITY ASSURANCE PROGRAM 17.1-10 17.1.3A DESIGN CONTROL 17.1-14 17.1.4A PROCUREMENT DOCUMENT CONTROL 17.1-16 I 4 17.1.5A It!STRUCTIONS, PROCEDURES, AND DRAWINGS 17.1-17 17.1.6A DOCUMENT CONTROL 17.1-19 17.1.7A CONTROL OF PURCHASED MATERIAL, EQUIPMENT, AND 17.1-20 SERVICES j 17.1.8A IDENTIFICATION AND CONTROL OF MATERIALS, PARTS 17.1-23 > AND COMPONENTS , 17.1.9A CONTROL OF SPECIAL PROCESSES 17.1-23 17.1.10A INSPECTION 17.I-25 17.1.11A TEST CONTP.0L 17.1-26 t 17.1.12A CONTROL OF MEASURING AND TEST EQUIPMFNT 17.1-28 17.1.13A HANDLING, STORAGE AND SHIPPING 17.1-29 17.1.14 A INSPECTIONS, TEST AND OPERATING STATUS 17.1-29 17.1.15A NONCONFORMING MATERIALS, PARTS, OR COMPONENTS 17.1-30 17.1.10A CORRECTIVE ACTION 17.1-30 17.1.17A QUALITY ASSURANCE-RECORDS 17.1-31 17.1.18A AUDITS 17.1-32 i V , 11 Am. No. 59. (6/81)
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1 Ta ble Title Page l 17.0.5-1 ACNGS-Specific. Modifications To Ebasco Topical 17.1-33 - 1- Report ETR-1001 Revisloa 9. i 17.0.C-1 CE Position On Regulatory Guides 1.58, Rev. I 17.1-36 , And 1.4 6, Re v. 0 17.1.2A-1 Program Compliance Matrix 17.1-38 f' 1 j 17.1.5A-1 Project Procedure Matrix 17.1-40 l o i
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! Figure Title , 17.0.B-1 EBASCO Project QA Orgsnization i
- 17.1.1A-1 External QA Relationships ,
17.1.1A-2 Allens Creek Project Organization 17.1.1A-3 IIIAP Project QA Organization f l 4 1 l i t i I 1 1 : 1 . l i f i I i 1 l l l iv Am. No. 59, (6/81) I
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ACNCS-PSAR 17.0 QUALITY ASSURANCE Each section in. Chapter 17 has been assigned suf fixes to the section 33(U) numbe rs to identify that the section is applicable to the respective organizations as follows: A - Applicant, Houston Lighting and Power Company (HL&P), B - Architect Engineer / Constructor, Ebasco Services Incorporated (Ebasco), C - Nuclear Steam Supply System and Nuclear Fuel Supplier, General Electric Company (GE). Ebasco and HL&P corporate management responsible for quality related activities (ie: engineering, construction, quality assurance) will periodically visit the Allens Creek site to evaluate and assess the quality attitude and motivation of managers, supervisors, technicians, foremen, craf t personnel and other cognizant personnel. Seminars, group meetings, discussions, etc. will be held as necessary to ensure that personnel are 59 aware of executive management support for the quality assurance program. Ebasco and llL&P corporate management support is further delineated in the corporate policy statements in the front of the respective Nuclear QA Program Manuals. A brief introduction has been prepared to introduce HL&P, Ebasco and CE, respectively. 17.0.A HOUSTON LIGHTING & POWER COP.PANY llouston Lighting & Power Company (HL&P) as the Applicaat, has the Quality Assurance responsibility for design, engineering, procurement, fabrication, [m construction, preoperational testing and operation of the Allens Creek
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Nuclear Generating Station (ACNGS). Although HL&P will delegate certain of its QA activities and authority to its contractors, it retains responsibi-lity for the QA program controlling all aspects of the ACNGS. The HL&P Quality Assurance Program requires that HL&P, its prime contractors, subcontractors and vendors comply with the criteria established by 10CFR50 57 Appendix B. It is the intent of HL&P to comply with ANSI N45.2 and the applicable daughter standards and implementing Regulatory Guides. Furthor-more, HL&P shall assure through programmatic direction that Ebasce and all of its subcontractors and suppliers performing nuclear safety rel< ted work comply with 10CFR50 Appendix B, ANSI N45.2, and the Regulatory Guides as referenced herein consistent with their scope of work. In addition, HL&P will comply with 10CFR50 Appendix A and Regulatory Guide 1.29, Revision I for the identification of items "important to safety," and these items 59 will be under the control of the QA Program. Progrcmmstic direction is defined as the role of HL&P in establishing the program requirements and ensuring the adequacy of the prime contractors QA Program. The programmatic direction consists of review and approval of the system features initially and continued monitoring of those systems during implementation and further refinement or revision of the systems if the systems need strengthening. 57 , The HL&P QA Program is described in the corporate Nuclear Quality Assurance Program Manual (NQAPM). The NQAPM requires the establishment of a Project Quality Assurance Plan for each project to describe the QA program to be j implemented during the design and construction phase of each project and an Operational Quality Assurance Plan to describe the QA Program to be imple-17.0-1 (U)-Update Am. No. 59, (6/81)
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( ) ACNCS. The Project QA Plan (PQAP) is described in Chapter 17.1 of the v' PSAR. The Operational QA Plan will be described in Chapter 17.2 of the FSAR. The PQAP specifies requirements applicable to prime contractors and ilL& P. The llL&P quality assurance staff shall assure through implementation review that the HL&P staf f and contractors are complying with the QA pro-gram _ and the PQAP. Implementation reviews are performed by qualified personnel based on. experience, educational level, training, and proficiency 57 examinations. Certifications are issued for specific discipline oriented activities. The implementation reviews use prepared checklists and tech-niques such as interviews with personnel performing the activities, obser-vations of actual work in progress, and reviews of final fo rm. The combination of the QA programs as described in the NQAPM, PQAP, and OQAP as aagmented by definitive procedures provide llL&P with the assurance that its quality commitments are m t. 17.0.B EBASCO The Quality Assurance program for safety related activities and services performed by Ebasc'o in the design, engineering, procurement, and construc- 33(U) tion of the Allens Creek Nuclear Generating Station is now described in tha Ebasco Nuclear Quality Assurance Program Manual for the Allens Creek Pro-ject. This manual is a modified version of Ebasco's Topical Report No. ETR-1001, Revision 9, which was accepted by the NRC on January 5,1981. 59 The ACNGS specific modifications to ETR-1001 are described in Table 17.0.B-1.
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( ) Later NRC approved revisions to ETR-1001 may be incorporated when deemed necessary. If necessary to define any additieaal clarifications, or modi- 45(U) fications to the project Nrelear Quality Assurance Program Manual because 33 of IIL&P contract requirements or to suit the unique Project conditions, they will be submitted for NRC acceptance in accordance with established 46 provisions which require execution of an authorization forin involving (U) approval of specified authorities to assure, among other things, that safety and/or quality are not sacrificed or compromised. l 59 The Ebasco Quality Program defined herein assures that structures, systems, and components important to safety as defined in Section 3.2 of this PSAR, are reliable and possess a high degree of quality. 'Ihis objective is achieved by the implementation of the Ebasco Nuclear Quality Assurance Manual which defines the policy, procedures, and requirements by which Ebasco will design, procure and erect the Allens Creek Nuclear Generating l 59 Station. Implementation of the Ebasco Nuclear Quality Assurance Manual
- provides a quality program which is in compliance with the requirements of the Code of Federal Regulations, 10CFR50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants," and ANSI N45-2-1971, " Quality Assurance Program Requirements for Nuclear Power Plants".
The Ebasco QA organization and division of responsibility for ACNGS is summarized on Figure 17.0.B-1. 59 O V) l 17.0-2 (U)-Update Am. No. 59, (6/81)
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- :AL ELECTRIC The Quality Assurance Program for safety related acti'rities and services for the Allens Creek huclear Generating Station is described in the 33(g)
General Electric huclear Energy Divisions BhR Quality Assurance Program Deseription, hELO-11209-04A. 14 9 4 In addition, General Electric has been asked to review Regulatory Guides ) 1.56 Rev. I and 1.146 Rev. O, specifically for the AChCS Project. These 59 positions are shown in Table 17.0.C-1. 4 4 E t i i 1 J t 1 1 i f i i (. i 17.0-3 (U)-Update Am. No. 59, (6/81)
_=. . - . . . ACNGS-PSAR 17.1 QUALITY AFSURANCE DURING DESIGN AND CONSTRUCTION 17.1.1A ORGANIZATION 33(U) The major organizations involved in the ACNGS are: c) Houston Lighting & Power Company (HL&P) As the Applicant, HL&P has and retains the overall responsibility for the engineering, design, procurement, fabrication, construction, pre-operational testing, operation and QA activities for the ACNGS. HL&P will audit the activities of Ebasco, GE, consultants and other contractors to assure that their QA Programs are implemented and have suf ficient authority and organizational freedom to be effectivelv implemented. HL&P will perform surveillance of the activities of Ebasco, GE, consul-tants and other contractors during the manufacturing, fabrication and const ruction of the ACNGS. b) Ebasco Services Incorporated (Ebasco) As the Architect-Engineer, Ebasco is delegated the responsibility to h) y/ provide HL&P with engineering, design, procurements and QA services. As the constructor, Ebasco, is delegated the responsibility to provide r HL&P with construction and QA services at the site, Ebasco has the responsibility to provide an acceptable QA program to HL&P for the activities that have been delegated to Ebasco. These delegated activities include the following: 57
- 1) design and engineering
- 2) procurement activities
- 3) home office QA activities
- 4) vendor surveillance activities
- 5) construction activities
- 6) site QA/QC activities c) General Electric Company (CE)
As the Nuclear Steam Supply System (NSSS) and Nuclear Fuel Supplier, GE is delegated the responsibility to provide HL&P with the engineering', design, procurement, fabrication and QA services for the NSSS and Nuclear Fuei. Gl: has the responsibility to provide an acceptable QA program to HL&P for the activities that have been delegated to GE. I a (U)-Update 17.1-1 Am. No. 59, (6/81)
i L
' ACNGS-PSAR These delegated activities include the following:
- 1) . design and engineering' activities I
... 2) procurement activities t
~3) fabrication activities
, 4 )' vendor surveillance activities
- 5) QA activities i
'd)' Consultants HL&P may utilize the services of qualified consultants or other con-tractors 1to assist in the performance of anpropriate quality tasks, such as audits, inspections, interpretations of test results, reviews, 57 etc.
Figure 17.1.lA-1.111ustrates how the above companies interrelate for . the'ACNCS. Figure 17.1.lA-2 is an organization chart showing the organizations ~ within HL&P with responsibility for the engineering, design, procure-ment, construction, operation, and quality assurance activities for j ACNGS. Figure 17.1.lA-3 is an organization chart of the HL&P Quality Assurance
. group for the ACNCS.
The HL&P Project Quality Assurance Manager shall be located at the construction site and is responsible for directing and managing the site QA program. He is f ree f rom non-QA duties. The HL&P Project Quality Assurance Manager is responsible f or providing the programmatic direction and administering the policies, goals, objectives, and methods for the Allens Creek Project which are described in the Project 59 Quality Assurance Plan. Programmatic direction is defined as the role ! of HL&P in establishing the program requirements and ensuring the ! adequacy of the quality assurance program for HL&P and the prime j contractors. The Project Quality Assurance Manager reports to the ' Manager, Quality. Assurance, who reports directly to the Executive Vice President and has the independent authority to identify quality-related , pro blems, to initiate or recommend solutions, to control existing i nonconf ormances, to verify implementation of approved dispositions, and when necessary to stop work. 17.1-2 Am. No. 59, (6/81)
t; ACNGS-PSAR
, The HL&P Executive Vice President reviews and approves the Project quality Assurance Plan and has ultimate responsibility for Quality Assurance activities. The Project Quality Assurance Plan interfaces with the corporate Quality Assurance program objectives by describing specific Quality Assurance controls to be established by HL&P and the prime contractors on the Allens Creek Project.
Two levels of control have been implemented by HL&P to monitor the effectiveness of the Quality Assurance Programs for the Allens Creek Project: (1) Corporate level control relates to the overall activities 57 and perf ormance of HL&P, Ebasco, subcontractors and suppliers. This is administered through the direct involvement of the HL&P Executive Vice President and through audits of project activities. (2) Project level control relates to monitoring the specific activities and perfornance of HLAP, Ebasco and its subcontractors. This is accomplished through review of documents, and implementation reviews that establish QA system f eatures (e.g. procedures, apecifications) . HL&P Executive Manageme.t is also involved through participation in the Quality Assurance Program Evaluation Committee (QAPEC). The QAPEC is comprised of HL&P management (and consultants as required) whose purpose is to provide a review and evaluation of the HL&P nuclear QA p ro gram. The primary responsibilities of the QAPEC is to assess the effectiveness of tae HL&P nuclear QA Program from the management
'~N vie wpoint inc luding:
59
\~ / 1) review ni NRC reports
- 2) review of trend analysis reports
- 3) review of selected audit reports
- 4) review of management QA audits
- 5) review major changes to methods and systems being implemented as part of the nuclear QA Program Designated QA individuals are involved in day-to-day plant activities important to saf ety, (i.e. the QA organization routinely attends and participates in daily plant work schedule and status meeting to assure they are kept abreast of day-to-day work assignments throughout the plant and that 57 there is adequate QA coverage relative to procedural and inspection controls, acceptance criteria, and QA staffing and qualification of personnel to carry out QA assignments) .
Figures 17.1.1A-3 and 17.0.B-1 show the project QA organization and indicate l 59 which personnel are "onsite" and "of fsite". The PSAR Section 13.0 shows project personnel frcm other organizacions. The criteria for determining staf fing for the QA organization includes: 57 m \ s' 17.1-3 Am. No. 59, (6/81)
ACNGS-PSAR- (( a) , Establishing the number of QA/QC personnel based upon the project - \'"'}/ schedule uto ensure' that personnel are available, qualified, snd certified te perfona quality-related inspections and evaluations. 1 b) Establishing the need for specially qualified QA/QC personnel based upon the. schedule for activities requiring special or unuoual expertise as far in advance of the activity as possible, t i c) Establishing the number of QA personnel based upon the r umber and criticality of problems identified during routinc activities in order i to perform addit:lonal or supplemental inspections, reviews, or evaluations as r. squired to ensure implementation of project 57
- requ i rement s.
Staf fing projections are periodically reviewed based upon the project schedule and are re-reviewed and revised, as nec. ssary, as the project schedule , chang es. QA management personnel part'eipate in short and long range scheduling activities. Staffing levels for QA/QC are a prime consideration in determining the level of ef fort for quality-related activities. Prior to allowing quality-related activities to be conducted, adequate numbers of qualified QA/QC personnel must be available. Adequate QA/QC staf fing must be available to prevent QA/QC personnel f rom being resnired to perform inspections without adequate preparation time or uncer pressure to complete inspections within a scheduled time period. Adequate QA/QC staff must be available to allow for prompt closcout of open nonconformances and proper f-~g follow up to ensure corrective action has been taken. ( .
.)
i 1 1 s - . s s_ - 17.1-4 Am. No. 59. (6/81)
ACNGS-PSAR {
'/ ~ 17.1.1A.1 Manager, Quality Assurance The Manager, Quality Assurance, has the authority and responsibility to iden-tify, initiate, recommend, or provide solutions to quality related problems and verify the implementation and ef fectiveness of the solutions. This posi-tion has the authority to "stop work" for cause in engineering, design, pro-curement, fabricatica, construction, and operation phises of the nuclear plant. The minimum requirements established for thi osition are:
a) a college degree in a field of enjineering or science or equivalent experience, b) familiarity with nuclear power generation facilities and the related operations. c) knowledge of the industry's quality assurance standards and regulatory requirements. d) management experience and familiarity with HL&P corporate organizations. The Manager, Quality Assurance, provides technical guidance, project direc- 57 tion, and administrative direction to: a) Project QA Manager, Allens Creek I \ b) Houston QA Manager c) Operations QA Manager The Manager, Quality Assurance, reports to the Executive Vice President. 17.1.lA.2 Project Quality Assurance Manager, Allens Creek The Project Quality Assurance Manager, Allens Creek (Project QA Manager) must as a minimum have: a) a college degree in a field of engineering or science, or equivalent experience, bi familiarity with nuclear power generation facilities and related ope rat iot, s. c) knowledge of the Quality Assurance standards and regulatory require-ments. d) management experience and familiarity with HL&P corporate organizations. l O \ , (j 17.1-5 Am. No. So, (6/813
ACNGS-PSAR (J / The major responsibilities of the Project QA Manager are: a) administer QA policies established by management and easure the proper planning, development , implementation, coordination and administration of the Project Quality Assurance Plan. b) provide programmatic direction on QA related matters to HL&P and con- 57 tractor management and interf ace with NRC.
~
c) coordinate activities relating to auditing and vendor surveillance in conjunction with the IIL&P llouston Quality Assurance Manager. The Project QA Manager has the authority to solve quality-related problems and 159 to verify the implementation and effectiveness of the solutions. He has the authority to "Stop Work" f or cause on any quality-related activity of the Allens Creek Project. 17.1.lA.3 flouston Quality Assurance Manager The !!ouston Quality Assurance Manager reports on all technical and administra-tive matters directly to the Manager, Quality Assurance. This organizational arrangement providec independence from coat and scheduling influences. The llouston Quality Assurance Manager is responsible f or directiry: allIILeP
-s llouston of fice auditing, vendor surveillance and technical support activi-
[ } ties, lie has the authority to "Stop Work" for cause on any quality-related
\s_ / activity of the Allens Creek Project.
The llouston Quality Assurance Manager as a minimum has: a) a college degree in a field of engineering or science, or equivalent exp erie nce. b) Familiarity with nuclear power generation f acilities and the related operations. c) knowledge of the industry's Quality Assurance standards and regulatory requ i rement s. ' d) management experience and f amiliarity with IIL&P corporate organizations. i The major responsibilities of the liouston Quality Assurance Manager are: i a) provide administrative guidance and direction for the HL&P Quality Assurance audit pro gram, b) direct the llL&P vendor surveillance program. c) provide technical support in the review of specifications, procedures, ! manuals, procurement documents, etc. ' V i 17.1-6 Am. No. 59, (6/81)
ACNGS-PSAR
~
l 1 (,) 17.1.1A.4 Project Quality Assurance General Supervisor The Project Quality Assurance General Supervisor reports directly to the Project QA Manager. He is responsible for technical direction and administra-tive guidance to the discipline Quality Assurance personnel, providing pro-grammatic direction to Ebasco and interfacing with the NRC. lie has the autho-rity to "Stop Work" for cause on any activity related to f abrication and construction. 17.1.lA.5 Supervisor, Quality Systems The Supervisor, Quality Systens reports directly to the Project QA Manager. lie is responsible for providing technical direction and administrative guid-ance to the site Quality Systems personnel; developing and administering the HL&P Project QA Plan; evaluating the Ebasco QA/QC program; administering the 57 IIL&P site QA personnel training and certification program; administrative control of HL&P quality assurance procedures and providing mechanisms to correct the QA programs as necessary. He has the autho-ity to "Stop Work" for cause on any activity related to fabrication or construction. 17.1.lA.6 Discipline Project Quality Assurance Supervisors The Discipline Project Quality Assurance Supervisors report to the Project Quality Assurance General Supervisor. They are responsible for technical direction and administrative guidance to the Discipline Quality Assurance personnel in the.r respective discipline group; coordinating implementation (nV) reviews; interf ace with NRC during audits; identifying deficiencies; reviewing and approving procedures applicable to their respective discipline; and pro-viding programmat ic direction to Ebasco. They have cuthority to "Stop Work" for cause on any activity related to fabrication or construction. 17.1.lA.7 Procurement Project Quality Assurance Supervisor The Procurement Project Quality Assurance Supervisor reports directly to the Project QA Manager. lie is responsible for providing technical direction and administrative guidance to procurement Quality Assurance personnel, coordina-ting the resolutions of vendor problems identified by HL&P, coordinating with site discipline Quality Assurance functions for input to vendor sur-veillance/ audit activities and providing programmatic direction to Ebasco regarding vendor surveillance and auditing functions. He has the authority to "Stop Work" for cause on any activity related to engineering, design, or procurement. 17.1.lA.8 Manager, Allens Creek Project The Manager, Allens Creek Project reports to the llL&P Vice President, Nuclear Engineering and Construction. He has overall responsibility for the engineer-ing, construction, procurement, cost, schedule, and start-up of the Allens Creek Project. v 17.1-7 Am. No. 59, (6/81) J
J. ACNGS-PSAR
' i
; i
\__/ He directs the personnel assigned to the Allens Creek Project in the perfor-mance of their activities to ensure that design and engineering, procurement construction, and start up meets the requirements of the project specifica-tions, procedures and policies. Ensures that interfaces and communication with and support by HL&P and Ebasco parent organizations are adequate to assure competent performance of project-related activities. He has authority to "Stop Work" for cause in all activities of the project.
17.1.lA.9 Project Engineering Manager The Project Engineering Manager reports to the Manager, Allens Crcek Project. He directs project engineering personnel in the performance of the KL&P review of the design and engineering work performed by the prime contractors. The 57 Project Engineering Manager ensures that adequate. engineering planning and coordination of solutions to problems and work priorities are established by the prime contractors. He can recommend "Stop Wo-k" for cause in the engi-neering and design of all items. 17.1.lA.10 Supervising Project Engineer (s) The Supervising Project Engineer (s) report to the Project Engineering Mana-ger. They direct in their area of responsibility the daily activities and interf ace with prime contractors. These activities include adequate engineer-ing planning, coordination of problems, work priorities and activities of the
/ ' 3l HL&P Project Engineering group assigned to each Supervising Project Engineer.
( The Supervising Project Engineer (s) monitor the Prime Contractors resolution of pertinent QA noncompliances, participate in the HL&P Incident Review Com-mittee and identify and resolve critical problems in their area of responsibi-lity. Direct the coordination and interface between design engineering and other project disciplines, ensure the HL&P review of ACP design documents and recommend "Stop Work" for cause in the engineering and design of all items within their area of responsibility. 17.1.1A.11 Project Construction Manager The Project Construction Manager reports to the Manager, Allens Creek Project. He is responsible for monitoring the total construction effort and maintaining liaison between HL&P and the Prime Contractors Management. The Project Construction Manager provides technical direction and cdministrative guidelines for HL&P and Prime Contractors in the areas of construction, secu-rity, start up, accounting, construction control, and ensures that the prime contractor's management properly implements the dispositions to various non-conformances as determined by the engineering resolution. Reviews and ap-proves as applicable procurement documents, drawings, specifications and construction interface schedules with subcontractors and ensures that con-struction conforms to the plans, specifications and procedures that govern work activities. Has the authority to "Stop Work" for causing relating to construction. A [ i N_.) 17.1-8 Am. No. 59, (6/81)
ACNGS-PSAR 7 17.1.lA.12 Project Purchasing Manager The Project Purchasing Manager reports to the Manager, Allens Creek Project. He is responsible for the overall coordination and administration of pur-
' chasing and subcontracting activities for the Allens Creek Project including the development and implementation of procedures, vendor selection, contract negotiations and preparing purchase orders.
17.1.1A.13 Project Controls Manager The Project Controls Manager reports to the Manager, Allens Creek Project. He is responsible for providing a detailed project budget and schedule integrat-ing engineering, construction and start-up. The Project Controls Manager has no directly-related quality assurance responsibilities on the project. 17.1.1A.14 , Project Administration Supervisor The Project Administration Supervisor reports to the Manager, Allens Creek 1icject. He is responsible for coordination of support to the Allens Creek Project Team from Ebasco and HL&P, processing and distribution of project 57 mail, development of project procedures and adrinistrative support. 17.1.1A.15 Project Centroller i' The Project Controller reports to the Manager, Allens Creek Project. He is responsible for the coordination and execution of the accounting and financial administration. The Project Controller has no direct quality assurance re-sponsibilities on the project. 17.1.1A.16 Project Environmental Engineer The Project Environmental Engineer reports to the Manager, Allens Creek Pro-ject. He is responsible for the environmental protection of the environs of the plant and for the acquisition of all local, state, and federal permits and approvals exclusive of NRC licensing. 17.1.1A.17 Project Nuclear Fuel The Project Nuclear Fuel group reports to the Director, Nuclear Fuels. They are responsible for fuel procurement, fuel management and providing technical support on nuclear fuel related matters. O
\v) i 17.1-9 Am. No. 59, (6/81)
~ ACNGS-PSAR f i
( /
/ 59 17.1.2A QUALITY ASSURANCE PROGRAM The llL&P Project Quality Assurance program for the Project has been developed in accordance with the criteria of 10CFR50 Appendix B, ANSI N45.2 and Regula-tory Guides as referenced herein, to provide programmatic direction on quality requirements for the prime contractors and subcontractors during design and construction. 57 The nuclear safety-related structures, systems and components covered by this program are listed in Section 3.2, Table 3.2-1, column designated " Quality As surance Program" . In Table 3.2-1, GE has the responsibility for Quality Assurance (QA) for items designated "GE" in the " Scope of Supply" column in Table 3.2-1 until delivery of the component to the site. Ebasco QA retains the responsibility for QA of all itcms designated "P" and the GE' items upon receipt at the project site.
Items listed in Table 3.2-1, the Q-list, will be maintained in compliance with 10CFR50 Appendix A and R.G. 1.29, Revision 1. IIL&P shall approve additions or deletions to Table 3.2-1. The criteria for and management of the the items on 59 the Q-List are described in Chapter 3. Measures shall be established for the initiation, control and maintenance of the Q-list by Ebasco. These measures shall include provisions for signature approval by Engineering and QA, con-trolled distribution of the lists to identify responsible personnel, and [^N assurances to preclude use of obsolete lists. l l The IIL&P Quality Assurance program f ar the Allens Creek Project is described l 57 by the IIL&P Project Quality Assurance Plan (PQAP). A letter signed by the Executive Vice President in the front of the PQAP makes the requirements of the PQAP mandatory. Procedures are reviewed by project QA personnel during 59 preparation for inspections, surveillance, implementation reviews, and audits to ensure consistency with project requirements. Additionally, selected procedures are reviewed and concurred with by the project QA organization prior to issuance. The plan requires that written procedures, training and c er t if ica t ion, issuance of specifications and drawings, and work and inspec- l tf on planning be accomplished in advance of performing nuclear safety-related activities. IIL&P Project Quality Assurance ensures through procedure reviews that this advance preparation is accomplished. The Project Quality Assurance Plan for the Allens Creek Project is structured in accordance with the NRC regulatory position of the Regulatory Guides as 57 described in Appendix C of the PSAR and with ANSI N45.2.12. (Draft 3, Rev. 4 - February, 1974) . The IIL&P QA Program and Procedures which are used to implement the quality related activities for each major organization and the reference to the applicable criteria of 10CFR50 Appendix B are listed in Table 17.1.2A-1. Verification that plans and procedures are properly implemented is accomplish-ed by llL&P Quality Assurance through audits, inplementation reviews and p regular management assessment of the Quality Assurance Program. U 17.1-10 Am. Nn. 59, (6/81)
g'" 3 ACNGS-PSAR L / The project QA organization and the necessary technical organization partici-pate clearly in the QA program definition stage to determine and identify the extent QA controls are to be applied to specific structures, systems, and components. Implementation reviews shall be performed by HL&P Discipline Quality Assurance 59 personnel using prepared checklists to evaluate the ef fectiveness of compli-ance to the Quality Assurance Program at the Allens Creek Nuclear Generating Station site during construction. The implementation reviews use techniques such as interviews with personnel perf orming the activities, observations of actual work in progress, and reviews of final form. Implementation reviews are perf ormed at the construction site by personnel qualified based upon experience, education level, training, and proficiency 57 examinations. Certifications are issued for specific discipline oriented activities. This qualification and certification program is documented in written procedures. Personnel performing quality control functions at the site anri at vendor f acilities are qualified in accordance with ANSI-N45.2.6 (Regulatory Guide 1.58, Revision 1). Audit personnel will be qualified in 59 accordance with Regulatory Guide 1.146. It is the policy of HL&P as applicant, to assure that the design, engineering, l57
<-~3 procurement, fabrication, construction, preoperational testing, and opera- l 59
/ ) tion of ACNGS are in conf ormance with project specifications, procedures,
\_/s codes, and NRC regulations. It is the responsibility of each organization assigned to the Allens Creek Project to ensure that project proced' ral review methods include provisions to ensure that the requirements stated in this manual are incorporated into project procedures. The Project Quality Assur-ance Plan establishes activities and procedures which identify, initiate and verify the resolution of nuclear safety-related quality problems. The imple- 57 menting procedures call for the resolution of quality problems at the lowest possible authorized level. However, if a dispute is encountered in the resolution of a quality problem which cannot be resolved at lower levels, the HL&P Project QA Manager presents the problem ultimately to the HL&P Executive Vice President for resolution.
Allens Creek Project Quality Assurance is responsible for conducting a quality oriented Indoctrination program for new personnel that have quality-related functions. The HL&P Project Quality Assurance Plan requires that prior to performing activities affecting quality the personnel are trained in the applicable procedures. The training, qualification and certification programs are established such that: 59 a) Personnel responsible f or performing quality af fecting activities are instructed as to the purpose, scope, and implementation of the quality related manuals, instructions, and procedures. p
)
LJ 17.1-11 Am. No. 59, (6/81')
ACNGS-PSAR [V I b) Personnel verifying activities affecting quality are trained and qualified in the principles, techniques, and requirements of the activity being perf ormed. c) For formal training and qualification programs, documentation includes the objective, content of the pregram, attendees, and date of a ttendance. d) Proficiency tests are given to those personnel performing and verifying activities affecting quality, and acceptance criteria are developed to determine if individuals are properly trained and qualified. 59 e) Certificate of qualifications clearly delineates (1) the specific func-tions personnel are qualified to perf orm, and (2) the criteria used to qualify personnel in each function. f) Proficiency of personnel perf orming and verifying activities af fecting quality is maintained by retraining, reexamining, and/or recertifying ao determined by management or program commitment. g) The description of the training program provisions listed above satisfles the regulatory position in Regulatory Guide 1.58, Revision 1. IIL&P Quality Assurance audit s are perf ormed to ensure compliance with these c ri t eri a. G The Project QA Manager and the Houston Quality Assurance Manager are directly responsible f or assuring effective implementation of the Quality Assurance program. The qualifications for these positions are defined in Sections 17.1. lA.2 and 17.1. lA.3. The HL&P Project Quality Assurance Plan requires the prime contractor (Ebasco) to submit all procedures which control nuclear safety-related construction 57 activities to lilAP Project Quality Assurance for review. Procedures are reviewed by Project QA personnel during preparation f or inspection, surveil-lance, implementation reviews and audits to ensure consistency with project requ irement s. Additional selected procedures are reviewed and concurred with prior to issuance. It is the responsibility of IIL&P Project Quality Assurance to determine that the prime contractor's procedures require proper equipment, environment and other prerequisites to perf orm the associated activity. These requirements are verified through implementation reviews by HL&P Discipline QA and audits by HIAP Houston QA. The results of the HlAP implementation reviews and audits are presented in a monthly report to the HL&P Executive Vice President. Regular executive management review of the monthly activities and the direct involvement of the HL4P Executive Vice President assures that an objective progran assessment of the Allens Creek Project Quality Assurance program is being perf ormed. l 59 V 17.1-12 Am. No. $9, (6/81)
t ACNGS-PSAR S_
\ [ \s./ _ HL&P Project Quality Assurance reviews and documents concurrence with the Ebasco Quality Assurance manual and audits are performed by HL&P Houston Quality Assurance to ensure compliance.
When required to implement / originate quality activities (10CFR50, Appendix B), HL&P will apply the same controls, as is specified in this Chapter 17, for Principal Contractors. Additional QA/QC requirements will be described for 59 the operational phase, which includes preoperational and startup in the FSAR. HL&P is committed to maintaining the Project Quality Assurance Plan as an ef fective and meaningful document to provide directions to HL&P and the prime contractors on the Allens Creek Project. When proposed substantive changes to this Project Quality Assurance Plan affect the docketed Quality Assurance Program description, HL&P will notify the NRC of the change (s) for their 57 review and acceptance prior to implementation. Organizational changes of a substantive nature will be reported to the NRC within 30 days of announcement. Table 17.1.2A-1 is a matrix showing 10CFR50, Appendix B criteria compared to 3 appropriate sections of the QA Program and Plan. This matrix illustrates how } the HL&P QA Program snd Allens Creek QA Plan are in compliance with the j Regulatory cri ter.'s. HL&P positions on Regulat'ory Guides (RCs) are enumerated in Appendix C. lf~' 59 i : (, i e i l i i ( 17.1-13 Am. No. 59, (6/81) i
ACNGS-PSAR [ ,} 17.1.3A DESIGN CONTROL Q llL&P has the overall responsibility for design and engineering of the Allens Creek Project and imposes the requirements of 10CFR50, Appendix B, Criterion III, Regulatory Guide 1.64 (Rev. ?) and ANSI N45.2.11-74 on the prime con-tractors and applicable subcontractors. IIL&P has contracted with Ebasco and General Electric to perform the design, engineering, and design verification. IIL&P Engineering performs reviews of selected elements of the completed design, design documents, and specifica-tions to ensure that contractual requirements are met. The llL&P Project Engineering Manager is responsible %r ensuring that project engineering activities are conducted in accordance with approved engineering procedures. The project engineering organization provides programmatic direc-tion and overview of the Ebasco engineering activities. The llL&P project engineering activities are conducts. in accordance with approved project procedures. When llL&P has direct responsibilities or assumes direct responsibility for conducting design activities, these activities will be conducted in accordance with the requireinents of this section and/or the FSAR Section 17.2.3. IIL&P contractort are required to provide the following design control measures in their quality assurance programs: I a) A design control system is established to document the methods of accomplishing and controlling essential design activities. 57 b) Design documents such as calculations, diagrams, specifications, and drawings are prepared and records developed such that the finM design is traceable to its sources. c) Design activities, documents, and interfaces are controlled to assure that applicable input such as design bases, regulatory requirements, codes, and standards are incorporated into the final design. d) Design input requirements, including design criteria, are documented and their selection reviewed and approved. e) Design documents include an indication as to their importance to safety and shall specify the quality characteristics, including materials, parts, equipment and processes, that are essential to functions of s t rue r.u re s , systems, and components. Design documents also include, as appropriate, acceptance criteria for inspections and tests. f) Design control measures are applied to items such as seismic, stress, thermal, hydraulic, radiation, and accident analyses, as they apply to the development of design input or as they are used to analyze the i design. ' i U ! 17.1-14 Am. Fa. So, (6'81) l
c ACNGS-PSAR
. g) Safety-related and/or seismic Category I designs are verified for adequacy and accuracy through independent objective review of design documcats by individuals competent in the subject activity. This verification may include the use of alternate or simplified solution methods or qualification testing, as appropriate.
h) Design changes, including engineering, vendor, and construction origi-nated changes, are controlled in a manner commensurate with the control imposed on the original desiga.
- 1) Document distribution is controlled such that all individuals using a design document or its results and/or conclusions for f urther design work can be notified if the document is revised or cancelled.
j) Design documentation includes evidence that design control requirements have been satisfied. k) Errors and deficiencies in approved design documents, including design methods (such as computer codes), that could adversely affect struc t-ures, systems, and components important to safety are documented; and l action taken to assure that all errors and deficiencies are corrected.
- 1) Deviations f rom specified quality standards are identified and proce-dures a re established to ensure their control.
57 7-~s m) A documented check to ensure dimer.3 tonal accuracy (including tolerance j j f or accept / reject criteria and inspectability) and the completeness of
\s_ ,/ the drawings and specifications.
n) A system to en'aure design requirements from engineering specifications and drawings for that system, component, or structure are included in inspection documents and that the cognizant engineering group perf orm an engineering evaluation and signoff on deviations identified on the
- inspection document s.
o) A system is established to require that design specifications and drawirms are reviewed by individuals knowledgeable and qualified in QA/QC techniques to assure that the documents are prepared, reviewed, and approved in accordance with written procedures and that the documents contain the necessary QA requirements such as inspection and test requirement s, acceptance naquirement s, and documenting of inspection and test results. HIAP Houston Quality Assurance performs audits of HL&P, Ebasco, and General Elec tric to ensure that design controls, requirements, specifications, and documents are in accordance with the design control criteria. In addition HL&P Proiact Quality Assurance reviews quality / construction proce-dures to ensure that the quality requirements of the design specifications are inc orpora ted. HL&P Project Quality Assurance also perf orms implementation reviews to ensure that the work is accomplished in accordance with the design f'~'s ruquirements and to ensure that field changes to the design are processed in ( ) accordance with the design control criteria. v 17.1-15 Am. No. 59, (6/81) L
i ACNGS-PSAR 17.1.4A PROCUREMENT DOCUMENT CONTROL
. ,m
[ To assure that nuclear safety-related items are purchased in a planned and
\s_ ,/) controlled manner, the HL&P Project Quality Assurance Plan establishes '
basic requirements which are to be used by HL&P in preparing procurement procedures for the Allens Creek Project. Ebasco performs procurement activities for nuclear safety-related equipment, materials, and services, 57 exclusive of the NSSS contract, which is performed by General Electric. Ebasco and General Electric ensure through contract, vendor surveillance, and audit that their suppliers comply with the established requirements. The basic requirements are: a) Written procedures are established clearly delineating the sequence of actions to be accomplished in the preparation, review, approval, and control of procurement documents. 1 Q17.9 b) A review of the adequacy of quality requirements stated in procure-ment documents is performed by qualified personnel knowledgeable in the QA requirements. This review is to determine that all quality requirements are correctly stated; they can be inspected and con- 33(U) trolled; there are adequate acceptance and rejection criteria; and the procurement document has been prepared in accordance with QA Program requirements. ~ c) Documented evidence of the review and approval of procurement docu-ments is provided and available for verification. s
\' ) d) Procurement documents identify those QA requirements which must be complied with and described in the supplier's QA Program to meet 10CFR Part 50, Appendix B. This QA Program or portions thereof shall be reviewed for adequacy by qualified personnel knowledgeable in QA.
, e) Procurement documents contain or reference applicable design bases [33(U) technical requirements including regulatory requirements, component and material identification, drawings, specifications, codes and
' industrial standards, including their revision status, tests and inspection requirements and special process instructions for such activities as fabrication, cleaning, erecting, packaging, handling, shipping, storing, and inspecting.
f) Procurement documents contain as applicable , requirements which I identify the documentation to be prepared, maintained, submitted, and made available to the procuring agent for review and/or approval, such as drawings, specifications, procedures, inspection and test records, personnel and procedure qualifications, and material and 33(U) test reports. g) Procurement documents contain the requirements for the retention, control, and maintenance of records. 't O)\- ' (U)-Update 17.1-16 Am. No. 59, (6/81)
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ACNGS-PSAR h) Procurement documents contain the procuring agency's right of ( access to vendor's facilities and records for source inspection Q) and audit. i) Changes and/or revisions to procurement documents are subject to at least the same review and approval requirements as the original document. j) I Purchase documents for spare or replacement parts of safety-related gg7,9 structures, systems, and components are revtewed for adequacy of quality requirements by qualified personnel knowledgeable in QA. The review is to determine the adequacy relative to the quality as-surance requirements and acceptance criteria of the original design. k) The evaluation and selection of suppliers are determined by qualified personnel in accordance with written procedures.
- 1) A written procedure acceptable to HL&P shall be used for source evaluation, m) Procurement documents, records, and changes thereto are collected, stored, and maintained in a systematic and controlled manner.
HL&P Engineering is responsible for review and approval of Ebasco procure-ment specifications. Engineering also coordinates with HL&P Procurement QA for performance of a quality assurance review. HL&P Procurement QA co-ordinates with Ebasco and IIL&P Engineering in the review of the procurement (%j ( package. 57 v In addition, llL&P Discipline QA is responsible for reviewing field procure-ment nackages to ensure that all quality assurance requirements have been included, ilL&P llouston Quality Assurance is responsible for performing audits and vendor surveillance to verify that the requirements have been implemented and that they are ef fective. 17.1.5A INSTRUCTIONS, PROCEDURES, AND DRAWINGS The llL&P Project Quality Assurance Plan requires HL&P, the prime contrac-tors, and their suppliers to establish and implement a Quality Assurance Program which is in compliance with 10CFR50, Appendix B. The program is e f fect ive in verifyin; that the defined activities are accomplished and documented in accordance with written procedures, instructions, and drawings and that they provide quantitative and qualitative acceptance criteria. 57 Procedures for the review, approval, and issuance of documents (including procedures, instructions, specifications, and construction drawings), and ch anges thereto are established and described to assure technical adequacy and inclusion of appropriate quality requirements prior to implementation, s Selected documents are reviewed and concurred with by the project QA organization for Quality Assurance related aspects. V} 17.1-17 Am. Nn. 59, (6/81)
as ACNGS-PSAR HL&P Project Quality Assurance reviews the Ebasco Allens Creek Project . ,r x Quality Assurance Prograq. To measure the effectiveness of the Quality (' Assurance Program, HL&P has . implemented a monitoring program consisting of sudits which are performed by 4 L&P Houston Quality Assurance and imple-mentation review and trend analysis performed by the HL&P Project Quality Assurance Department. HL&P Houston Quality Assurance also audits HL&P 57 organizations and General Electric for compliance with their respective Quality Assurance programs. Table 17.1.5A-1 is a matrix showing HL&P procedures that are used on the Allens Creek Project compared to the appropriate 10CFR50, Appendix B criteria. This matrix illustrates how the requirements of the applicable 10CFR50, Appendix B criteria are addressed in the written procedures used on the Allens Creek Project. 59 b
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l l O 17.1-18 Am. No. 59, (6/81)
ACNCS-PSAR 17.1.6A DOCUMENT CONTROL ,' i The HL&P Project Quality Assurance Plan and implementing procedures require Q) that HL&P, the prime contractors, and subcontractors implement a document control system for nuclear safety-related items for the Allens Creek Project. The established system ensures that design, engineering, procure-57 ment, fabrication, construction, and QA/QC procedures, plans, and changes . thereto are reviewed and appioved by procedurally authorized groups and ' that the documents are issued, maintained current, and controlled by the use of controlled liste of document holders to ensure that superseded documents are replaced in a timely manner. The document control system applies to the control of quality-related documents, including as a minimum: a) Design documents; such as calculations, drawings, specifications and analysis b) Procurement documents c) Instructione and procedures for such activities as fabrication, construction, installation, t e'.t inspection, modification, operation, maintenance and refueling 59 d) As-built drawings e) Safety Analysis Reports N I f) QA/QC manuals J ' g) Nonconformance reports h) Audit reports Project procedures will be developed to ensure that drawings are provided to indicate the as-built configuration. The as-built drawings will stand alone and delineate actual location - elevation, azimuth, etc.; actual com-ponent identification or numbering; and dimensions and other relevant in- ; fo rma t ion. When changes occur anhsequent to issuance of as-built drawiags, procedures vill require a ae-review and reissue of the drawings. Measures are established and documented to control the issuance of docu-ments, such as instructions, procedures, and drawings, including changes thereto, sich prescribe activities af fecting quality. These measures shall assure that documents, including changes, are reviewed for technical adequacy and the inclusion of appropriate quality requirements and approved for release' by authorized personnel and are distributed to and used at the 57 location dere the prescribed activity is performed. Changes to documents are reviewed and approved by the same organizations that performed the original review and approval unless other organizations are specifically designated. The reviewing organization has access to pertinent background information upon sich to base its approval and shall have adequate under-standing of the requirements and intent of the original document. 17.1-19 Am. No. 59, (6/81)
AChGS-PSAR Those participating an an activity are made aware of and use proper and current instructions, procedures, drawings, and engineering requirements for performing the activity. Partsctpating organizatior.s have proceduces Jf)N
\
fur control of the documents and changes thereto to prec lude the possible use of outdated or inappropriate documents.
Document control measures provide for: a) Identification of individuals or organizations responsible
.ier preparing, reviewing, approving, and issuing documants and rivisions thereto; b) identifying the proper documents to be used in performing the activity; 57 c) coordination ano control of interface documents; d)- ascertaining that proper documents are being used;
,' e) establishing current and updated distribution lists The document control system includes a listing identifying the current reviston of instructions, procedures, specitications, drawings, and l procurament documents. This list is updated and distributed to cognizant ' responsible personnel. i HL&P Discipline Quality Assurance performs implementation reviews at the
/
construction site to ensure that document control systems are in place and et tect ively implemented. HL&P Quality Assurance audits are pe rformed >k % to ensure compitance with these criteria. i 59 17.1.7A C0hlh0L OF PURCHASEL MATEklAL, EQUIPMENT, AhD SERVICES The ht&P Quality Assuranca Plan and implementing procedures require that HL&P, prime contractors, and subcontractors define and document the system and requirements for the control of nuclear safety-related purebased matertal, equipment, and services. ( Control and verification of supplier's activities during fabrication, inspection, teati.:g, and shipment of materials, equipment, and components are planned and performed as early as possible, as required, to assure conformance to the purchase order or contractual requirements. These procedures provide for:
- a) Requiring the supp1Lir to identify processes to be utilized in tultilling procurement requirements. 57 b) Reviewing documents required to be submitted by the procure-Sint requirements. '
c) Specifying the characteristics or processes to be witnessed, inspected, or verified and accepted based upon the fabrica-tion schedules; the method of surveillance, and the extent i /' T' ; fss 17.1-20 Am. No. 59, (6/81) g g w. ,m7m,, .gg ,-,-y-- - - - - - - - w - yyrg -+
ACNCS-PSAR of documentation required; and those responsible for imple-e menting these procedures, d) Audits, surveillance, and/or inspections which assure that the supplier complies with quality reouirements and his QA program. For items determinea to be impo rt ant to safety where specific QA controls cannot be imposed in a practical manner, an evaluation will be made to determine special quality verification requirements to be applied during 59 installation or testing to provide the necessary assurance that the item (s) meet project requirements. Control and verification of organizations performing service is accom-pltshed by technical vert tication of data provided, surveillance and/or audit of the activity, anc review of objective evidance such as certifi-cat ions, report s, etc. The selection of suppliers is based on evaluation of their capability to provide items or services in accordance with the requirements of the procurement documents prior to award of contract. Procurement source evaluation and selection measures are implemented by HL6P and Ebasco and provide for identification of the organizational responsibilities for determining supplier capability. Measures for evaluation and selection of procurement sources, and the results thereot, are documented and include one or more of a) through (g d) below: 57 i ) k/ a) Evaluation of the supplier's history of providing an identical or similar product or service which performs sattstactorily in actual use. 'Ih e suppler's history shall reflect current capability. b) Supplier's current quality records supported by documented qualitative and quantitative information which can be objec-t ive ly evaluated, c) Supplier's technical ano quality capability as determined by a direct evaluation of his facilities and personnel and the implementation of his approved quality assurance program. d) Evaluation of bid documents including review for technical adequacy, quality assurance, and commercial considerations. Procurament of spare or replacement parts for struetures, systems, and components important to safety is subject to QA program controls, ' to codes and standards, and to technical requirements at least equal to the original technical requirements or any properly reviewed and approved revisions there to. A receipt inspection is planned and implemented to assure: (O v/ l i 17.1-21 Am. No. 59, (6/81')
AChGS-PSAR a) Timely inspectton of stems upon eceipt. (,,j
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b) The material, component , or equipment is properly' identified and corresponds to the identification on the purchase document and receiving documentation. c) haterial, components, equipment , and acceptance records sattsfy the inspection instructions prtor to installation or use. d) Specified inspection, test, and other records are accepted and available at the Allens Creek Project prior to installa-tion or use where required. e) Items accepted and released are identified as to their inspection status prior to forwarding them to a controlled storage area or releasing them for installation or further work. f) Coordination of receipt inspection with vendor surveillance activttres to ensure the required vendor inspection has been 57 pertormad and deficiencies have beten resolved prior to shipment. Supplier's certiticatec of conformance are evaluated by audits, vendor inspection, or tests to ensure that they are valid. Supplier's racords will include a description of those nonconformances from the procurement -(q 4 requirements dispostt toned " accept as is" or " repair".
\~ / Ebasco receiving inspection ensures that, for nuclear safety-related items received at the Allens Creek Proja ' ,
- r e is 4: companying documantation that indicates review Snd concurrence by the prime con-tractor or designee, that the item complies with catablished requirements or has an authorized waiver prior to shipment. HL&P Quality Assurance audits are performed to ensure compliance with these criteria.
HL&P houston Quality Assarance ensures by an overview of the Ebasco vendor surveillance function that source surveillance and inspection are performed in accordance with the quality assurance program. In addition, HL&P Discipline QA performs implementation reviews of activities commancing with receiving inspection at the site to ensuring proper controls of purchased material and equipment are exercised. HL&P Houston Quality Assurance performs audits of these activities to ensure overall compliance. 59 n v 17.1-22 Am. No. 59, (6/81)
ACNGS-PSAR U] [ 17.1.8A IDENTIFICATION AND CONTROL OF MATERIALS, PARTS AND COMPONENTS The HL&P Prnject Quality Assurance Plan requires that prime contractors and suppliers establish written procedures which identify, control, and_ ensure traceability of materials, parts, rnd components including partially assembled components. Prime contractors and suppliers procedures shall include the documented verification of correct identification of materials, components, and subassemblies, and that the identification does not affect the function or quality of the ites prior to release of the itema for acaembly or installation. Specific procedures have been developed to: a) Establish controls to identify and control materials (including consumables), pa rts, and components (including partially fabricated subas semblies) . b) Provide a method for identification of quality relatea materials and partu and to provide traceability to zhe appropriate drawings' 57 specifications, purchase orders, manuf acturing, and inspection documents, deviation reports, and physical and chemical mill test i repo rt s. c) Travide a method for identification and control of-incorrect or defective items. This system will include verification and documentation prior to release for fabrication, assembling, shipping, and installation.
) HL&P Project Quality Assurance ensures that the above criteria are s- / incorporated into the Ebasco quality / construction procedures during the procedure review and then follows up with implementation reviews to ensure
- complia nc e.
In addit'.on HL&P Houston Quality Assurance perf orms audits for evaluation of a the conf ormance to identification and control criteria. 17.1.9A CONTROL OF SPECIAL PROCESSES The HL&P Project Quality Assurance Plan requires that written procedures be established by prime contractors and subcontractors for the activicies associated with all special processes. For special processes the qualification of personnel, procedures, and equipment relating to specific code s, sta nda rd s, specifications, and contractual requirements shall be 57 documented and maintained current. Special procer 4s are defined as the processes where direct inspection if impossible or cisadvantageous and which must be carefully controlled and 17.1-23 Am. No. 59, (6/81) 1 1
ACNGS-PSAR
/~'N monitored to ensure the required results. Special processes for the Allens (w; ) Creek Project include:
a) welding b) heat treating c) cadwelding d) concre te placement e) nondestructive testing f) chemical cleaning Organisational responsibilities are defined in the Allens Creek Project procedures for qualification of special processes, equipment, and personnel. These responsibilities include the provision to assure that special processes are performed by qualified personnel using procedures qualified and approved in accordance with applicable codes, standards, or other requirements. Procedures are established for recording evidence of acceptable accomplishment 59 of special processes using qualified procedures, equipment, and personnel. The QA organization verifies the recorded evidence and documents the result.
,- ,s Specini processes are performed under controlled conditions by qualified
( ) personnel using procedures qualified and approved in accordance with \s_ ,/ applicable codes, standards, or other requirements. For special processes not covered by existing codes or standards the specific equipment, personnel qualification, and procedure quolification requirements are defined prior to application of the special process. Records are maintained for the qualification of procedures, equipment, and personnel associated with special processes. Records are in suf ficient detail 57 to clearly define the procedures, equipment, or personnel being qualified; criteria or requirements used for qualification; and the individual approving the qualification. lil4P Discipline Quality Assurance ensures that the special process control criteria are met by the review of all Ebasco special process procedures and performance of implementation reviews to ensure compliance. IlL&P llouston Quality Assurance performs audits of special process activities to ensure compliance with all aspects of the Quality Assurance program. 59 (Av) 17.1-24 Am. No 59, (6/81) m w -- w P
ACNCS-PSAR 17.1.10A INSPECTION rN i i The HL&P Project Ouality Assurance Plan requires the prime contractors to ( "/ establish and implement an inspection operation whose activities are inde-pendent from the group performing the activities being inspected. The training, qualifications, and certifications of inspectors includes criteria from appropriate codes, standards, and the prime contractors pro-cedures and shall be documented and kept current. Inspection activities relating to construction, fabrication, installation, and testing are doc-umented, kept current and identify all mandatory inspectior ! ld and test points and the criteria to be witnessed by authorized inspectors. Opera-tions and inspections (including rework, replaced items) are performed in predetermined , documented sequences, and deviations or deleticas must be accomplished in accordance w[th approved and documented systems.. InspeC-tion procedures include all required inspection operations defined by che specifications, drawings, codes, and standards. These procedures provide for the following: a) Identification of characteristics and activities to be inspected b) A description of the method of inspection 57 c) Identification of the individuals or groups responsible for per-forming the inspection operation d) Acceptance and rejection criteria [ e) Identification of required procedures, drawings, and specifications V and revisions f) Recording inspector or data recorder and the results of the inspec-tion operation g) Specifying necessary measuring and test equipment including accuracy requirements and verification of calibration h) Evaluation of inspection results The project QA organization participates in the definition of the scope of the inspection program. Procedures provide criteria for determining the accuracy requirements of inspection equipment and criteria for determining when inspections are required or define how and when inspections are 59 pe r f o rmed . Procedures are established to identify in pertinent documents, mandatory inspection hold points beyond which work may not proceed until inspected by a designsted inspector. Where direct inspections are impossible or disadvantageous, in process monitoring is specified in the inspection procedures and both direct and in process monitoring are used when control is inadequate without both. All required procedures, specifications, and drawings are made available 57 to the inspectors prior to performing inspection. If mandatory inspection , [ hold points are required beyond which work cannot proceed without specific (y l 17.1-25 Am. No. 59, (6/81)
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ACNCS-PSAR consent. of the designate - representative, the specific hold points will be indicated in appropriate documents. Inspection results are documented, 57 (Ql U evaluated, and their acceptability determined by a responsible individual or group. Personnel performing quality control functions at the site and at vendor facilities are and will be qualified in accordance with ANSI-N45.7.6 59 (Regulatory Guide 1,58, Rev.1). HL&P Discipline Quality Assurance ensures that inspection control _ criteria are complied with by review and approval of the inspection procedures and 57 by implementation reviews of ILnspection in each discipline activity. HL&P liouston Quality Assurance performs audits of HL&P and Ebasco inspec-- tion activities to ensure compliance with these criteria. 17.1.llA TEST CONTROL The HL&P Project Quality Assurance Plan requires that' a test control pro-gram be developed and documented by the prime contractors and subcontrac-tors which demonstrates that the f acility performs in accordance with the Allens Creek Project requirements and specifications. The training, certification of personnel, calibration and certification of test equip- 57 ment, system or component status, environmental conditions, inspection hold points, and configuration of .the items tc, be tested are included in the procedures. Test results are documented, evaluated, and the accep-tance status determined by the authorized departments. (/ ) 4 A test control program will be established to include proof tests prior 59 to installation and preoperational tests. Procedures provide criteria for determining acurracy requirements of test equipment and criteria or determining when a test is required and how and when testing activities a r e pe r fo rmed . Test procedures or instructions provide for the following as required: a) The inclusion of requirements and acceptance limits contained in applicable design and procurement documents. b) Instructions for performing the test c) Tcst prerequisites such as calibrated instrumentation, adequate test equipment, and instrumentation including their accuracy require- 57 ments, completeness of item to be tested, suitable, and controlled environmental' conditions, and provisions for data collection and storage d) Mandatory inspection hold points for witness by Owner, contractor, or inspector (as required) i e) Acceptance and rejection criteria 7'Nt f) Methods for documenting or recording test data and results ( iv/ . i 17.1-26 Am. No. 59, (6/81)
ACNGS-PSAR g) Provisions for assuring that test prerequisites have been met (- ( j h) Evaluation of test results
%/
HL&P Discipline Quality Assurance ensures inclusion of adequate test con-trol criteria by review of the Ebssco quality / construction testing procedures. They also perform follow-up implementation reviews to verify that the controls are implcmented and et fective, llL&P llouston Quality Assurance audits both HL&P and Ebasco activities to verify QA program compliance. The test control activities are an example of a case in which HL&P Discipline Quality Assurance monitoring activities and the Operational Quality Assurance monitoring activities will interface and in some instances overlap. IIL&P Project Quality Assurance procedures will speci- l 59 fically define the responsibilities for this transition period. l 57 59 m m v 17.1-27 Am. No. 59, (6/81)
- . , . . , . . _. ~. . . - .. .~ . .
ACNGS-PSAR
- 17.1.12A CONTROL OF/ MEASURING ~AND TEST EQUIPMENT
- )-
ls,,,/ . The HL&P[ Project Quality Assurance Plan requires the establishment, docu-- .
- mentation, and implementation of a. Measuring and Test Equipment Control System.- The' system is to include calibration techniques, specifications and ,
accuracy, frequency, and maintenance of all measuring instruments and test
. equipment used in the . measuring, inspection,-and monitoring ~of nuclear safety-related. items. Calibration and maintenance data shall be filed and
~ kept current. ~ Calibration standards are .to be' traceable to nationally
- i- recognized standards. If standards do not exist, the basis for calibration ,
of-the equipment is to be documented. If measuring or test equipment is-found to be out of calibration, an investigation is required to be performed to determine the validity of the use of the instrument and whether measure-ments or tests are required to be reperformed. 2 Equipment is identified and traceable to the calibration test data and suitably marked to indicate calibration status. Markings include the.last day calibrated and next calibration due date. 57 Measuring and test equipment is calibrated at specified intervals based on the required accuracy, purpose, degree of usage, -stability characteristics, and other 'conditior.s af fecting the measurement. Calibration of this equip-ment is against standards that 'have an accuracy of at least four times the required accuracy of the equipment being calibrated, or when this is not ] possible, have an accuracy that assures the equipment being calibrated will be within required tolerance and that the basis of acceptance is documented and authorized by respor.sible management, i(
\~- Calibratin, standards will, when possible , have greater accuracy than 4 standards being calibrated. Calibrating standards with the same accuracy may be used if it can be shown to be adequate for the requirements and the
, basis of acceptance is documented -and authorized by responsible management. '
!!L&P Disciplin ' io.' 'ty Assurance reviews and documents concurrence with Ebasco calibratton pr ocedures to ensure these criteria are incorporated.
~
In addition implementa' ion res as are performed to ensure compliance. IIL&P Houston Quality Assurance audits the measuring and test equipment con-trols to ensure compliance to the QA program in this area. } j 59 d a l' i V .J } 17.1-28 Am. No. 59, (6/81) y -m.- , 4.- -, , , . - , , , , - -,,,--,----+,~,,-w , , - ,3.m- , , - - - , ,,,,.e=,,,,,, ,,<,--,w.-- - - , - ,,
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I ACNGS-PSAR r3 ( ) 17.1.13A HANDLING, SIORAGE, AND SHIPPING ( <' The HIAP. Project Quality Assurance Plan requires that for nuclear safety-related items, written procedures be developed in accordance with design requirements, specifications, and standards to control the cleaning, handling, storage, packaging, shipping, and preservation to preclude damage and deteri-oration by environmental conditions. The activities are to be accomplished by appropriate trained and experienced personnel. HIAP Discipline Quality Assurance reviews and documents concurrence with 57 construction procedures for receiving, handling, storage, and cleaning to ensure that the appropriate criteria of Regulatory Guide 1.38 and ANSI N45.2.2. are included. Periodic implementation reviews are conducted to ensure compli-ance to the procedures. HIAP ~ Houston Quality Assurance performs audits to ensure overall program ccapliance. 17.1.14A ^ INSPECTIONS, TEST, AND OPERATING ST ATUS i 59 The HL&P Project Quality Assurance Plan requires that the prime contractor and subcontractors indicate the current inspection, test, and operating status of nucle.ir safety-related items through the use of stamps, markings, tags, or o ther suitable means. During the startup and testing activities, HL&P is respor sible for complying with this section f or inspection status, test
/ ) statun, and operating status. Procedures include the requirements for:
I a) Controlling the application and removal of inspection status indicators such as tags, markings, labels, and stamps. 57 b) Documenting the status of nonconf orming, inoperative, or malfunctioning structures, systems, and components to prevent inadvertent use. c) Defining and documenting the use, application, removal, and status of inspection tags, labels, and markings which identify the status of inspections or tests perfonned or attest to the acceptability of the structure, system, or component. d) Controllin3 the altering of the sequence of required tests, inspec-tions, and other operations important to safety. e) Providing a system for the indication of the inspection, test and operating status of structures, systems and components throughout fabrication, installation and test. 59 HIAP Discipline Quality Assurance personnel review and document concurrence with these procedures and mnduct periodic verification to assure compliance. g Houston Quality Assurance audits both HL&P Project Quality Assurance and . Ebasco to verify compliance.
. /m
( V} 59 17.1-29 Am. No. 59, (6/81)
l l ACNCS-PSAR q. j' j 17.1.15A NONCONFORMING MATERIALS, PARTS .OR COMPONENTS v , he HIAP Project Quality Assurance Plan requires that HIAP and the prime contractors' Quality Assurance Program include a system which is documented by writte i procedures for the identification, segregation, and disposition of 57 nonconforming materials, pa rt s, and components. The procedures shall specify the preparation and handling of nonconformance documents, segregation require-ments, and which groups a re responsible for review and disposition of the items. %e procedures also provide identification of authorized individuals for independent review of non-conformances, including' disposition and closeout. 59 I Documentation identifies the nonconf orming item; describes the nonconf ormance, the disposition of the nonconformance, and the inspection requirements; and includes signature approval of the disposition. Nonconformances are corrected and resolved prior to initiation of the preoperational test program on the item. Rework, repairs, and subsequent reinspection and tests are conducted in '~
' accordance with the original inspection and test requirements or accepted
- alternatives and shall be performed in accordance with controlled procedures and contain mechanisms for providing inf orination to the identifying group as to the disposition of the nonconf ormance. :
For NSSS items, IlI4P coordinates nonconformance resolution through CE. HIAP Project Quality Assurance reviews for concurrence the proposed disposi-tion of selected Ebasco nonconformance reports and performs an evaluation of j Ebasco nonconformance trend analyses. l Qi - Procedures are established by HIAP to report significant deficiencies during the design, construction, and operations phase to HIAP executive management and to the Nuclear Regulatory Commission in accordance with 10CFR50.55(e), 10CFR21, and 10CFR71, where applicable. Compliance of these activities with Project Quality Assurance Plan requira- , ments is ensured through the performance of audits and implementation review =. I ' l 59 17.1.16A CORRECTIVE ACTION i Re IllAP Project Quality Assurance Plan requires that a system be established i and documented by HL&P and the prime contractors which defines the responsibi-lities, authorities, and methods used by specific groups involved in the evaluation of nonconformances and trending to determine the need for correc-l tive action. The system includes measures to identify the cause of signifi-cant conditions adverse to quality, measures to ensure that the root causes 57 are corrected, and measures to ensure that timely action is taken. Follow-up is perfomed to ensure the effectiveaess of corrective action and that appro-priate levels of management are informed of the results. HIAF Project Quality j Assurance performs a review for concurrence of selected Ebasco nonconformance 4 reports and corrective action reports. HIAP Project Quality Assurance also , performs trend analyses to determine the need for corrective action. F) L 17.1-30 Am. No. 59, (6/81) i 4
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ACNCS-PSAR N
'lhe results of the trend analyses are included in the QA monthly activity u,/ report which is sent to the Executive Vice President. 59 Compliance of these actions with Project Quality Assurance Plan requirements is verified by IIL&P Quality Assurance through the performance of audits and 57 implementation reviews.
l59 17.1.17A QUALITY ASSURAfCE RECORDS The HL&P Project Quality Assurance Program requires that a Quality Assurance record system be developed by HL&P and the prime contractors for the Allens Creek Project. The record system provides evidence that activities relating to quality are defined, implemented, and that inspection and test documents contain a description of the type of observation, reference to nonconformance reports, evidence relating to status of observation, date, and inspector identification. The Project Quality Assurance Plan requires that HL&P and prime contractors establish requirements to ensure tha t records generated during the design, procurement, construction, properatic a -I and start-up testing are identifiable, re trievable and mee t the requirements of 10CFR50, and ANSI N4 5.2.9 as endorsed by Regulatory Guide 1.88, Revision 2. 59 As an alternative, records may be maintained for the Allens Creek Project in a two hour rated fire resistanc file room meeting NFPA No. 232 including the A, following provisions: i i 's / L' 57 a) An automatic fire suppression system and an early warning fire detec-tior. system is utilized. b) Records are stcred in fully enclosed metal cabinets. c) Smoking, eating, and drinking should be prohibited within the records storage facility. d) Work not directly associated with record storage or retrieval is prohi-bited within the records storage facility. c) Ventilation, temperature, and humidity contrr1 equipment is controlled where they pene trate fire barriers bounding the storage facility. Compliance with Project Quality Assurance Plan requirements is verified by HL&P Quality Assurance through the performance of audits and implementation reviews. l59 O k j 17.1-31 Am. No. 59, (6/81)
ACNGS-PSAR g l ) 17.1.18A AUDITS \ /
'~'
The HL&P Project Quality Assurance Plan establishes the requirement that HL&P, prime contractors, and subcontractors develop, document, and implement audit activities which are structured in accordance with the requirements of ANSI 57 N45.2.12 for the Allens Creek Project. As required by the A,NSI standard, results of audits are presented for review to management of the audited or-ganization and the HL&P Executive Vice President. Where indicated, HL&P perf orms f ollow-up action, including re-audit of the deficient areas. Audits are conducted by personnel qualified in accordance with ANSI N45.2.23 and the results analyzed by QA. Audit reports indicate any quality problems and the 59 ef fectiveness of the audited QA Program. Reaudits of deficient arecs are conducted as necessary to assume implementation of corrective action and recurrence control. Audit results are reported to management for review and assessment. HLAP has the ultimate responsibility for the auditing of the quality related activities on the project. This responsibility is fulfilled by Ibuston Quality Assurance, which audits the activities of HL4P, its prime contractors, and their suppliers and subcontractors. The prime contractors and subcontractors perform quality related audits of internal activities and suppliers of material, components and systems. 57 HLAP and Ebasco perform supplemental audits when required, based on such
~~s factors as significant changes in the Quality Assurance Program, results of !
(\ )
,/ '
trending programs, or investigations into the root causes of problems.
~
The HL&P Project Quality Assurance Plan requires that each year an independent outside firm shall conduct an overall audit of the Allens Creek Project Quality Assurance activities. The audit results are presented to the H14P Executive Vice President and the Project QA Manager. The audit results will be used by HL&P management to evaluate the effectiveness of the Quality Assur-ance program and to determine the need for changes in the Quality Assurance programs of HL&P and its contractors. l59 o I
\v/
17.1-32 Am Na. 59, (6/81)
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ACNGS-PSAk y TABLE 17.0.B-1
'N_ /
- ACNGS-SPECIFIC MODIFICATIONS TO ERASCO TOPICAL REPORT ETR-1001, h2 VISION 9
- 1. General Where the work " client" appears within the appropriate sections of EBASCO's Nuclear Quality Assurance Program Manual it shall be understood to mean " Houston Lighting & Power Company".
- 2. Section QA-I-2 Organization and Responsibilities In Paragraph 2.1, a direct line of communication for quality related matters has been established between Ebasco's Chief Quality Assurance Enginer.r or his alternate and HL&P's Quality Assurance Manager.
- 3. Section QA-I-5 Quality Assurance Evaluation of Suppliers /Contracto,rs
- a. Paragraphs 3.1.2 and 5.1 are modified to allow for alternate methods of evaluation and qualification of supplier's capabilities by methods other than audits by Ebasco. Such methods are detailed as follows:
i) Audits of suppliers by HL&P or others qualified to do so. 59
/; 11) Historical data is available substantiating the capability of the
( ' ' '
) supplier to provide products which have performed satisf actorily in actual use and were fabricated in accordance with an acceptable quality assurance program. Such historical data shall only qualify suppliers who have provided identical or similar products in the past.
- b. Paragraph s 2.3, 4.1, and 5.1 are modified such that in the event Construction Contractors are awarded a contract before review and approval of their quality assurance manual or their facility, but prior to start of any safety-related work, the following shall be complied with :
- 1) The " Terms and Conditions" section of the Purchase Ordet will stipulate that the award of the contract is predicated on: 1)
Submittal of construction contractors quality assurance manual for review and comment by Purchaser, 2) a satisfactory quality assurance audit by Purchaser of the construction contractors Quality Assurance Program. If the manual review and/or audit are unsatisfact , and if, in the opinion of the Purchaser there is , no hope of successful corrr_ctive actions, the terms of th l contract will permit Purchaser to absolve himself of the contract.
- 11) A visit will be made to the home offices of the contractor to discuss the techniques they intend to use in implementing their
,- program at the construction site.
)
. -33 Am. No. 59, (6/81)
ACNGS-PSAR f~~N J ABLE 17.0. B-1 (Cont'd) ( ) A / 111) Records of past lor. similar jobs shall be examined at the contractors office to verify implementation of constructors quality assurance program or at least make an evaluation of the contractors qualifications and capability. iv) ,The results of the preceeding reviews will be forwarded to HL&P for their concurrence.
- 4. Section QA-I-6 Quality Assurance Records All Quality Assurance documents maintained by Ebasco will be inspected for legibility, proper identification, and microfilmability. The requirement for vendor quality assurance records to be legible and microfilmable will be imposed on vendors through Ebasco's Purchase Order Specification.
- 5. Section QA-II-4 Purchasing 59 Section QA-II-4 is modified such that Ebasco proposes only the recommendations for purchasing and has no responsibility for issuance of the purchase orders. HL&P has assumed this responsibility and the system of contrel for issuance of these purchase orders is outlined in HL&P's Qun11ty Assurance Program Manual. Otherwise the remainder of this
[ Section is applicable in its entirety.
\
- 6. Section QA-II-5 Supplier Surveillance i
Paragraph 3.4 is modified such that af ter issuance of a purchase order and prior to start of fabrication, the Project Quality Assurance Engineer " prepares the Vendor Quality Assurance Plan for approval by HL&P QA. Af ter HL&P approval, the PQAE forwards the Vendor Quality Assurance Plan to the Vendor Quality Assurance Supervisor for use by Vendor Quality Assu. ance Representativec.
- 7. Section QA-III-4 Construction Site Procurements
- a. In Paragraph 2.7.3 the following is a clarification of the term
" Direct Evaluation":
Under certain circumstances and to assist the vendor evaluation group and to expedite the vendor evaluation process, the QA Site Supervisor j or qualified members of his staff who he may appoint, may perform 4 facility audits, primarily in their local geographic areas.
- b. In Paragraph 3.1.2 delete subparagraph (b) .
- 8. Section QA-III-ll In spe'e tion t
pg j The Ebasco QC organization reporting to the Quality Program Site Manager is responsible for the performance of inspection activities during construction. 17.1-34 Am. No. 59, (6/81) !
-~. _.-. . _ - _. _
- ACNCS-PSAR 6p . TABLE 17.0. B-1 (Cont 'd)
I
! 9. Section QA-III-14 Control of Receiving, Handling and Storage 1 . ..
t
- The f ollowing is a clarification of Paragraph 2.0; i
- l. When vendor surveillance-is not required f or certain items purchased by -
the site' organization, recef .ing inspection will include the review of Certified Material Test. Report s, NDE Record s, etc. In these cases the i review of such documents will be the responsibility of the Quality 39
- i. Control Site Organization. In turn, this operation would be audited by j the Quality Assurance Site Group.
I 1 T 4 .) i 1 1 i-
!l ,
+ l t a i t i-l l f' i i ' l l l' + } 1 : l 1 i i I 17,1-35 Am. No. 59, (6/81) i
ACNGS-PSAR TABLE 17.0.C-1 t [3 1 ( ~) GE POSITION ON
~
REGULATORY GUIDES 1.58, REV. 1 AND 1.46, REV. 0
, Regulatory Guide 1.58 " Qualification of Nuclear Power Plant Inspection Examination and Testing Personnel".
Comply with the provisions of Regulatory Guide 1.58, Se ptembe r , 1980, in-cluding the requirements and recommendations in ANSI N45.2.6-1978, except it is recognized that in lieu of the required education and experience requirements expressed in Position 6 of Reg. Guide 1.58, that the alternate mechanism can be i ercised as described in Position 10 of the Reg. Guide. These additional controls vill be forward fit at the CP issue date and will not be back fit. Appropriate certification / qualification records are kept of the inspection, examination, and testing personnel. Regulatory Guide 1.146 " Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants". 59 Comply with the provisions of Regulatory Guide 1.146, August, 1980, includ-ing the requirements and recommendations in ANSI N45.2.23-1978, except that it is recognized that in lieu of the education and experience controls described in Section 2.3.1 of ANSI N45.2.23, it is agreed that GE will 7-~3 ( ) have documented objective evidence (i.e., procedures and records of written
\s / tests) demonstrating that lead auditors possess the required educational and experience skills to fulfill Reg. Guide 1.146 positions. These con-trols will be forward fit at the CP issue dat and will not be back fit.
0% ; (
'N_/
17.1-36 Am. No. 59, (6/81) J
I 1 ACNGS-PSAR-(j TABLE 17.1.2A-1 PROGRAM COMPLIANCE MATRIX 57 ; SECTIONS ADDRESSING 10CFR$0 APPENDIX B REQUIREMENTS 10CFR50 APPENDIX B HL&P NUCLEAR QUALITY l CRITERIA ASSURANCE PROGRAM ACNGS QA PLAN *
?
- 1) Organization 1.0 1.0, 3.2
- 2) QA Program 2.0 1.0 .:
- 3) Design Control 3.0 4.0
- 4) ' Procurement Docuteent Cont rol 4.0 5.0, 7.2 ;
o i
- 5) Instruction, Procedures, and Drawings 5.0 3.0, 6.3 J l'
- 6) Dacument Control 6.0 3.4, 6.3 7.0, 8.0 l
- 7) Control of Purchased 59 Material, Equipment, l and Services 7.0 5.0, 6.2, 6.3
- 8) Identification and Control of Materials, Parts, and ,
Components 8.0 6.2, 6.3 l
- 9) Control of Special Processes 9.0 6.2, 6.2 5 i
- 10) Inspection 10.0 6.2, 6.3
- 11) Test Control 11.0 6.2, 6.3
- 12) Control of Measuring and Test Equipment 12.0 6.2, 6.3 r
- 13) Handling, Storage, and l Shipping 13.0 6.2, 6.3 l l [
l 14) Inspection, Test, and ;
- Operat ing St atus 14.0 6.2, 6.3 {
i l l 4
'd j 17.1-37_ Am. No. 59, (6/81) l.
I
I ACNGS-PSAR TABLE 17.1.2A-1 (Cont'd) SECTIONS ADDRESSING 10CFR50 APPENDIX B REQUIREMENTS 10CFR50 APPENDIX B HL&P NUCLEAR QUALITY l 57 CRITERIA AS5URANCE PROGRAM ACNGS QA PLAN
- l
- 15) Nonconforming Items 15.0 3.5, 6.2, 6.3 59 I
- 16) Corrective Action 16.0 3,5, 6.2, 6.3 4 17) QA Records 17.0 6.2, 6.3, 7.0 1 18) Audits 18.0 6.2, 8.0 1
- NOTE: Approximately July 1,1981 the Allens Creek Nuclear Generating Station Quality Assurance Plan will be rewritten and issued with eighteen sections consistent with the 18 Criteria of 10CFR50 Appendix B.
i, i i t 4 i i 1 ~ l 4 17.1-38 m. No. 59, (6/81)
, , - . - , . , - , , , - , - - - . . , - , - - - , , , , - . , - ,.,n--- ., .. -----,--- . - . -
e .. - . w-, -- . - - -. , - .. - - -- - , - , ,
i ACNCS-PSAR , O, s TABLE 17.1.5A-1 PRaJECT PROCEDURE MATRIX.
- Procedure. 10CFR50 App. B
~I.. Project Procedures Number Criterion
. 1) INTRODUCTION [ Project Description and Policy ACPP-1Q II Project Organization ACPP-2Q I ~ Purpose and Scope of the ibnual ACPP-3Q II, V t Issuance and ' Control of the Manual ACPP-4Q II, V, VI
-Issuance and Control of Procedures ACPP-SQ ,II, V, VI
- 2) PROJECT ADMINISTRATION
-Correspondence Processing and ACPP-52Q VI
- Control
-Telephone 111nutes ACPP-53Q VI Meetings and Meeting Minutes ACPP-54Q VI
- q. Trip and Trip Reports ACPP-56Q VI
- Issuing and Controlling Project ACPP-58Q II, V, VI j Di rec t ives i
Administrative Training ACPP-59Q II i ] 3) COST /SCllEDULE ] Processing'the Cost Estimate ACPP-101 N/A j Change Request i Annual Budget Development and ACPP-102 N/A Control Project Estimate Review ACPP-103 N/A l 4) ENGINEERING REVIEW i Introauction ACPP-EIQ III Design Review ACPP-152Q III i' , i
~-.
4 17.1-39 Am. No. 59, (6/81)
l
-m ACNGS-PSAR
\
\j TABLE 17.1 5A-1 (Cont'd)
Procedure 10CFR50 App. B
- 1. Projact Procedures Numbe r Criterion
- 4) ENGINEERING REVIEk (Cont'd)
Design Review Authority ACPP-153Q III, I AE Design Change Notice (DCN) ACPP-154Q III, VI Processing Design Change Control within IIL&P ACPP-155Q III Transmittal of Engineering ACPP-156Q III Correspondence Engineering Review of Procurement ACPP-157Q III, IV 59 Documenta Engineering Training ACPP-158Q II Designation and llandling of Con- ACPP-159Q VI fident tal Security Docu n.snts Processing Supplier Deviation Requests ACPP-160Q III, IV
- 5) PROCUREMENT Ceneral Procurement ACPP-201Q IV, VII Establishing Bidders List ACPP-202Q IV, VII Inquiry Issuance ACPP-204Q IV , VII Proposal Evaluation and Supplier ACPP-205Q IV , VII, XVII Selectton l
Purchase Order, Preparation, Changes ACPP-206Q IV , VII ( Approval, and Issuance l Training ACPP-208Q I l Procurement File System and File ACPP-210Q IV , VI Control l 1 bcument Review ACPP-211Q IV, VII 1 0 O 17.I-40 Am No. 59, (6/81)
ACNCS-PSAR f .k]j , TABLE 17 1 5A-1 (Cont'd) Procedure 10CFR50 App. B
- 1. ' Project Procedures Number Criterion
- 5) ~ PROCUREMENT (Cont'd)
Document Control ACPP-212Q IV, VI Material Control Organization ACPP-213Q VII, VIII, XIII, XIV Preventative Maintenance ACPP-214Q VII, VIII, XIII, XIV Spare Parts ACPP-215Q IV, VII, XIII, XIV-
- 6) ACCOUNTiUC Invoice Review and Approval ACPP-251 N/A
- Vendor Back Cha rge ACPP-252 N/A 59
{ 7): LICENSING Reporting Design and Construction ACPP-301Q XV , XVI Deficiencies to NRC . llandling of NRC Inspection Reports ACPP-302Q VI and Immediate Action Letters Review of NRC Inspections and ACPP-303Q V1 Enforcement Bulletins and Circulars
- 8) CONSTRUCTION (Later)
- 9) STARTUP/ WARRANTY (Later)
- 10) PROJECT DIRECTIVES Use of Nuclear Division Procedure AC-PDIR-1 III, VI, VII NDP-130 for Evaluation of Reportable Defects and Deficiencies Authorization to Approve for Release AC-P DIR-2 I, VI Certain Project Correspondence Storage Procedure for Early Deliveries AC-PDIR-3 Vill, XIII, XIV or Material _for the Allens Creek Q Nuclear Generating Station 17.1-41 Am. No. 59, (6/81)
j'~~s ACNGS-PSAR k/s / TABLE 17.1.5A-1 (Cont'd) Procedure 10CFR50 App. B II. ' Project Site Quality Procedures (PSQP) Number Criterion
' Organization & Responsibility of Project QA PSQP-Al I, II Personnel Project Site Quality Procedures PSQP-A2 V, VI flandling of NRC Inspection Reports PSQP-A3 XV, XVI Control of Site Documentation PSQP-A4 VI Non-Nuclear Site Quality Assurance PSQP-A5 N/A Document - Reviews PSQP-A6 V, VI 59 Stop Work PSQP-A7 XV, XVI Trend Analysis Administration PSQP-A8 XVI Implementation Review PSQP-A9 II, IV thru
/\ ryII kw/ I Audit Overview PSQP-A10 XVIII Vendor Surveillance Overview PSQP-All IV, VII Construction QA - Operations QA Interface PSQP-A12 II, XI, XVII III. IIL&P !!ouston Quality Assurance Procedures Indoctrination & Training of IlL&P llouston QAP-2.1 II QA flome Office Personnel Procedure for Qualification and Certification QAP-2.2 II of Surveillance Personnel Training and Qualification of Audit Personnel Q AP-2.3 II Procedure for Document Review QAP-3 1 III Procedure for Procurement Document Review QAP-4.2 IV Standard Definitions and Abbreviations Q AP-5.1 V Standard Format for Writing and Controlling Q AP-5.2 V f
i s 17.1-42 Am. No. 59, (6/81) l l l s" *
~ l ACNCS-PSAR
- f' s ,,/ TABLE .17.1. 5A-1 (Cont 'd) .
Procedure 10CFR50' App. B III..HL&P Houston Quality Assurance Procedures Number Criterion (Cont'd)
' Issuance and Control of Documents QAP-6.1 VI 4
Procedure for Vendor Quality Surveillance QAP-7.3 VII HL&P Vendor Surveillance QAP-7.4 VII l Second Party Vendor Survelliance Category I QAP-7.5 VII
~
Second Party Vendor Surveillance Category II Q AP-7. 6 VII Review of Nuclear Steam Supply System Quality QAP-7.7 VII Assurance Records Packages 59 Control of Nonconformances QAP-15.1 XV
' Stop Work Procedures QAP-15.2 XV Corrective Action QAP-16.1 XVI
}
1 - Audit Filing QAP-17.1 XVII
.lli4P Audit Program QAP-18.1 XVIII Auditing QA Programs QAP-18.2 XVIII Joint Auditing of QA Programs QAP-18.3 XVIII IV Records Management Systems Procedures Records Management Respcasibilities & 1-2 I Interfaces Preparation and Periodic Review of RMS 1-3 V Procedures Records Management Personnel Training 1-4 II
' Records Center Micrographic Section 2-1 XVII Flow of Nuclear Correspondence with RMS T1-1 XVII Center-
\s 17.1-43 Am. No. 59, (6/81)
i i ACNGS-PSAR TABLE 17.1.5A-1 (Cont'd) i Procedure 10CFR50 App. B III. IIL&P llouston Quality Assurance Procedures Numbe r Criterion (Cant'd)
, Document Logging T2-1 VI Log Maintenance T2-2 VI Document Distribution T2-3 VI Storage & Maintenance of Nuclear Records T2-4 XVII l
Document Checkout T2-5 XVII 59 l Correspondence Serial Number Assignment T2-6 VI l , I ! Correspondence Serial Number Corrections T2-7 VI l I Subject File Number Assignment T2-8 XVII l NSSS Data Package llandling T2-10 XVII l t I
, t i
i 4 l l L i ! i l I I L i i l l ) i l l O 17.1-44 Am. No. 59, (6/81)
CORPOF QUALITY PROGRAM SITE MANAGER
/
QUALITY ASSURANCE RECORDS SITE CONTROL SUPERVISOR SUPERVISOR
- REVIEW OF FIELD CHANGE
- RECEIPT, INDEX, ETC. OA
- COC REQUESTS RECORDS SUF
- REVIEW OF DESIGN CHANGE
- ESTABLISH AND MAINTAIN WIT NOTICES PROJECT StTE FILE
- REVIEW AND APPROVE FINAL
- REVIEW RECORDS RECEIVED NDE FROM FIELD OC/QA e lSSUE GA PLANS FOR SITE
- PROCESS DEFICIENCY PURCHASE ORDERS REPORTS
- DEVELOP GUIDELINES FOR OUALITY PROCEDURES AND INSTRUCTIONS
- DEVELOP StTE OA/OC PROCEDURES & CHECKLISTS
- RSVIEW AND APPROVE NON-CONFORMANCE REPORTS
= SITE PURCHASE ORDER REVIEW HOME OFFICE SITE )* \-
i
ACNGS - PSAR 5 EBASCO [ LATE QA CRGANIZATION I PROJECT OA
' ENGINEER A
- SPECIFICATION REVIEW
- PURCHASE ORDER REVIEW s
- DCN REVIEW
- VENDOR SURVEILLANCE COORDINATION s
e VENDOR SITE QUALITY CONTROL SURVEILLANCE AUDIT SITE COORDINATOR SUPERVISOR SUPERVISOR )RDINATE VENDOR
- PLAN AND PERFORM SITE
- PERFORM INSPECTIONS VEILLANCE ACTlVITIES AUDITS
- ESTABLISH INSPECTION HOLD H RECEIPT INSPECTION . REPORT AUDIT RESULTS POINTS
- TREND ANALYSIS
- WITNESS TESTS
- PERFORM NDE
- PERFORM SOILS AND CONCRETE TESTS
- REVIEW MANUFACTURE AND CONTRACTOR RECORD PAlKAGES
- WITNESS HANDLING OPERATIONS
- PERFORM CORRECTIVE ACTION VERIFICATIONS
- VERIFY DOCUMENT CONTROL PROGRAM
- MONITOR WELDING OUALIFICATIONS
- CONTROL OF MTGU AM. NO. 59, (6/811 HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 ,
EBASCO PROJECT 1 y QA ORGANIZATION i f FIGURE 17.0.B-1 ,f
o I l EXECtJTIVE VICE PRESIDENT J L l DIR ECTOR VICE PRESIDENT NUCLEAR NUCLEAR ENGINEERING FUELS AND CONSTRUCTION PROJECT MANAGER NUCLEAR ALLENS CREEK FUEL PROJECT Jk JL I f 4 GE-NEBG MANAGEMENT GE-N F&SD MANAGEMENT 1 1f OUALITY GUALITY f ASSURANCE PROJECT ASEURANCE MANAGER MANAGER MANAGER J L J L J L 1 e 3 I i I,
ACNGS - PSAR I MANAGER QUALITY ASSURANCE PROJECT OUALITY ASSURANCE MANAGER i J L EBASCO MANAGEMENT PROJECT CHIEF QUALITY MANAGER ASSURANCE ENGINEER Jk J L J L GE-NPSD MANAGEMENT 4 OA LINES OF COMMUNICATION l 4 PROJECT LINES OF COMMUNICATION i V l l PROJECT MANAGER J L l l AM. NO. 59, (6/81) HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 EXTERNAL QA RELATIONSHIPS FIGURE 17.1.1 A-1
I i j i DIRECTOR NUCLEAR FUELS PROJECT PROJECT F NUCLEAR ENGINEERING PUI FUEL MANAGER M P ADMI SUI SUPERVISING I ' PROJECT COr ENGINEER (S) i I. f I < + i t i i j
ACNGS - PSAR t E.vECUTIVE VICE PRESIDENT VICE PRESIDENT MANAGER NUCLEAR ENGINEERING QUALITY AND CONSTRUCTPON ASSURANCE I I MANAGER PROJECT ALLENS CREEK OUALITY ASSURANCE PROJECT MANAGER ll ROJECT PROJECT ICHASING CONSTRUCTION kNAGER MANAGER I OJECT PROJECT ISTRATION CONTROLS RVISOR MANAGER I ( l PROJECT ECT d ENVIRONMENTAL AROLLER ENGINEER L AM. NO. 59, (6/81) HOUSTON LIGHTING Si POWER COMPANY Allens Creek Nuclear Generating Station i Unit 1 l ALLENS CREEK PROJECT ORGANIZATION L l FIGURE 17.1.1 A-2 i l
g, . . _ _ -- . I t I s l l l OPERATIONS OA MANAGER A _ NDE/ISI PROJECT OPERATIONS QA QUALITY ASSURA( SUPERVISOR GENERAL SUPERV[ i l PLANT OA
^
SUPERVISOR MECHANICAL i I PROJECT OA j SUPERVISOR CIVIL 4 -
- M HOME OFFICE SITE PROJECT QA SUPERVISOR ELECTRICAL 8
I s i
ACNGS - PSAR , l EXECUTIVE VICE PRESIDENT A MANAGER QUALITY ASSURANCE A PROJECT HOUSTON QUALITY ASSURANCE OA MANAGER MANAGER
/ A PROCUREMENT SUPE RVISOR ICE PROJECT QA Al>DITS & TECHNICAL 50R SUPERVISOR SERVICES
/ / A SUPERVISOR SUPERVISOR QUALITY VENDOR SYSTEMS SURVEILLANCE f
1 [ I AM. NO. 59, (6/81) l HOUSTON LIGHTlHG a t'0WER COMPANY Allens Creek Huclear Generating Station Unit 1 HL&P PROJECT F QA ORGANIZ ATION , FIGURE 17.1.l A-3 1
. - . . - ~ . - - - . . - ~ . . - _=... .. - ~ ... . ... _ - .- - .. - ._ .._._-. - _. _ - . - _- -
ACNCS-PSAR i j. LIST OF EFFECTIVE PACES l APPENDIX C t i'
.en 1
Page No. Amendment No. j !' 59 ; ! 1* 59 I i 2* 59 ! 3* 59
- j. 4* ,
I
~
42 ! i 42 [ j 11 42 l
- 111
' 58 [
iv 42 ! v , 42 vi 42 vii 42 i viii l 1 i 17 ; l C1.1-1 17 l j C1.2-1 35 i l C1.3-1 l 17
- C1.4-1 35 l
C1.5-1
- C1.6-1 35 C1.7-1 35 l 42 .
C1 8-1 35 l C1.9-1 C1.10-1 35 i l 17 ! C1.11-1 C1.12-1 35 l l 35 1 l C 1.13 -1 C1.14-1 17 l j 17 : l C 1.15-1 C1.16-1 35. r j 35 l C1 17-1 ' C1.18-1 35 ! 17 C1.19-1 ' C1. 2 0-1 42 35 C1.21-1 l C1. 2 2-1 31 35 C1.23-1 C1.24-1 17 j 35 ; i C1.25-1 C1. 2 6-1 42 , j 42 : j C1.27-1 j C1.2 8-1 46 I C1.29-1 42 l 42
- C1. 2 9-2 t
f- *Ef fective Pages/ Figures List 1 t i 1 Am. No. 59, (6/81) ! i ^
=
l t ACNGS-PSAR l l LIST OF EFFECTIVE PACES (Cont'd) - .; 4 APPENDIX C i e Page No. . Amendment No. l C1.30-1 45 l C1.31-1 42 ,
- C1.31-2 (deleted) 42 C1.31-3 (deleted) 42 C1.32-1 35 i C1.33-1 42 C1.34-1 17 I C1.35-1 35 C1.36-1 22 i '
1 C1.37-1 45 ! C1.38-1 46 C1.38-2 42 C1.39-1 46
- i. C1.40-1 17
! C1.41-1 17 i C1.42-1 35
! C1.43-1 46 C1.44-1 56 C1.45-1 17 l 22 ! C1.46-1 i C1.47-1 35
- C1.48-1 35
- C1.49-1 35 C1.50-1 35 i C1.50-2 35 C1.51-1 35 [
, C1.52-1 58 C1.53-1 35
; C1.54-1 46 4 C1.55-1 17 ,
I C1.56-1 18 C1.57-1 22 i C1.58-1 46 C1.59-1 42 C1.60-1 42 ' 17 C1.60-2 C1.61-1 22 C1.62-1 22
- C1.63-1 22 C1.63-2 22 l 46 C1.64-1 C1.65-1 35 C1.66-1 42 C1.67-1 22 C1.66 45
) 2 Am. No. 59, (6/81) i
AChCS-PSAR 4 i i LIST OF EFFECTIVE PAGES (Cont'd) APPENDIX C i Page No. Amendment No.
. C1.68-1-1 42 l C1.68-2. (deleted) 45 i 4 C1.68-3 (deleted)
I C1.68.1-1 42 , C1.68.2-1 42 l C1. 69-1 44
- C1.6 9-2 44 C1. 69-3 4 i 4 .
i C1.6 9-4 I C1. 7 0 -I 35 C1.71-1 35 C1.72-1 17 C1.7 3-1 22 C1.74-1 22 , C1.7 5-1 35 l 3 C1.76-1 35 C1.78-1 35 C1.80-1 35 C1.84-1 , 42 C1. 85-1 42 1 S C1.86-1 35 C1. 8 8-1 47 } C1.8 9-1 35 C1.91-1 42 i C1.92-1 42 2 C1. 9 2-2 42 C1.93-1 35 C1.94-1 47 ' C1.95-1 42 C1. 96-1 35 C1.97-1 59 C1. 9 7-2 59 C1.98-1 42
-C1.98-2 42 C1.99-1 42 C1.100-1 42 i C1.101-1 42 C1.10 2-1 35
! C1.104-1 42 C1.10 5-1 42 C1.106-1 42 C1.10 8-1 42 C1.109-1 42 C1.110-1 42 C1.111-1 42 i i.
- 3 Am. No. 59, (6/81) 1 i
._.._>_._._m..___..___.._ _ . _ . _ . - _ . _ _ _ _ . . _ _ _._. _ ._ _.. _ _ -.-- _.._ _. - _ __ -_____ _ _ _ _ _
4 h i I 4 1 ACNGS-PSAR f ,. i'O 1 LIST OF EFFECTIVE PAGES (Cont'd) APPENDIX C l t ! Amendae Page No. , nt No. l l 1 42 l l C1.112-1 42 l l C1.113-1 35 j C1.114-1 42 ; i C1.115-1 46 5 C1.116-1 42 l C1.117-1 42 ' C1.118-1 42 ! C1.120-1 42 ! C1.122-1 42 i C1.123-1 48 l ; j C1.124-1 42 i l C1.126-1 42 ! C1.127-1 48 i j C1.130-1 i ! ( l 9 , t i f k I i I ,. t I I i i 1 i l 4 4 Am. No. 59, (6/81) f
ACNGS-PSAR REGULATORY CUIDE 1.97 42(U) 57 ( w/
) Rev. 2, 12/80 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENT.
Applicant's Position: 42(U) The ACNGS design will meet Regulatory Guide 1.97, Rev. 2, except for the following items listed in Table 1 of that guide:
-BWR Core Thermocouples: The primary indicator of core cooling for the BWR is reactor vessel water level, of which multiple indication is available in the main control room. Thermocouples are not re-liable devices, and their potential to confuse the operator with erroneous or ambiguous information is deemed to outweigh any per-ceived benefits of having them available. The detailed arguments 57 against thermocouples are presented in attachment 1 of a letter from D. B. Waters (BWROG) to D. E. Eisenhut (NRC) titled "BWR Emergency Procedures Guidelines, Rev. I and responses to related geestions",
dated 1/31/81.
-Radiation Exposure Rate:
HL&P will develop a plan for the selection and location of radiation monitors in containment penetration areas and in areas where access f') to service safety equipment is required. This plan will be (j developed in conformance with the provisions of Reg. Guide 1.97, identify any exceptions and provide justification for any excep ons noted. This plan will be submitted to the NRC prior to the pro-curement of any of these monitors. 59
-Cooling Water Temperature to ESF components: A range.of 30-120 F is provided versus the 30-200 F of the guide. ACNCS uses lake water to cool ESF components. It is inconceivable that the lake could heat up to 200 F, and arbitrarily extending the instrument range diminishes its accuracy over the expected range, decreasing its usefulness to the operator. Thus, the 30-120'F range was selected.
The following are clarifications to the items listed in Table 1 of the guide: 57
-Drywell Sump Level and Drywell Drain Sump Level: These are inter-p;eted to mean the Low Purity Drywell Sump and High Purity Drain Tank Drywell.
-Containment and Drywell Oxygen Concentration: These are inter-preted to be required only for pre-inerted containments.
-Suppression Chamber Spray Flow: This is interpreted to mean Con- 59 fm tainment Spray flow, l I
'V C1.97-1 (U)-Update Am. No. 59, (6/81)
ACNGS-PSAR
-Drywell Spray Flow: ACNGS has no drywell sprays, i
-Isolation Condenser System Shell Side Water Level: ACNCS has no l isolation condenser.
57
-HPCI Flow: This is interpreted to mean HPCS flow.
r I
-Core Spray System Flow: This is interpreted to mean LPCS flow.
-High Radioactivity Liquid Tank Level: This is interpreted to mean l Liquid'Redwaste System Collection Tanks Level.
-Type A paraneters for ACNGS are given in Section 7.5.1.4.2.
i r i I' I v 1 Cl.97-2 Am. No. 59, (6/81)
i 1 ACNGS-PSAR EFFECTIVE PAGES LISTING APPENDIX 0 Page Amendtent No. 1* 59 .; 2* 59 3* 59 4* 59 5* 59 > 0-1 59 0-11 59 0-111 59 0-tv 59 0-v 59 0-1 59 0-2 59 0-3 59 0-4 59 0-5 59 0-6 59 0-7 59 0-8 59 0-9 59 0-10 59 0-11 59 0-12 59 0-13 59 0-14 59 0-15 59 0-16 59 0-17 59 0-18 59 0-19 59 0-20 59 0-21 59 0-22 59 0-23 59 0-24 59 , 0-25 59 0-26 59 l 0-27 59 l 0-28 59 0-29 59 1 0-30 59 0-31 59 0-32 59 0-33 59 0-34 59 0-35 59
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ACNGS-PFAR EFFECTIVE PAGES LISTING (Cont'd) APPENDIX 0
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ACNGS-PSAR TABLE OF CONTENTS APPENDIX 0 Ites No. Title Page Appendix 0 Responses to NUREG 0718. " Licensing 0-1 Requirements for Pending Applications for Construction Permits and Manufac-turing Licenses" (March, 1981) NUREG 0718 Category 1 0-2 NUREG 0718 Category 2 0-3 NUREG 0718 Category 3 0-4 II.B.8.1 Rulemaking Proceeding On Degraded 0-5 Core Accidents (Core / Containment Heat Removal Reliability) II.K.2.16 Impact of RCP Seal Damage Following 0-6 Small-Break LOCA with Loss of Offsite Power II.K.3.13 Separation of HPCI and RCIC System 0-9 Initiation Levels-Analysis and i Implementation II.K.3.16 Reduction of Challenges and f ailures of 0-12
. Relief Valves - Feasibility Study and System Modification II.K.13.18 Modification of ADS Iogic - Feasibility 0-15 Study and Modification for Increased Diversity for Some Event Sequences II.K.3.21 Restart of Core Spray and LPCI Systems 0-17 on Low Level - Design and Modification II.K.3.24 Confirm Adequacy of Space Cooling for 0-19 HPCI and RCIC Syecess II.K.3.28 Verify Qualification of Accumulators on 0-20 ADS Valves II.K.3.44 Evaluation of Anticipated Transients 0-21 with Single Failure to Verify No Significant Fuel Failure
- II.K.3.45 Evaluate Depressurization with Other Than 0-23 Full ADS
[ *) l l O-1 Am. No. 59, (6/81)
ACNGS-PSAR TABLE OF COhith1S (Cont'd)
, [-sT
/ Ites No., Title h
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hbR E 0718 Category 4 0-25 1.A.4.2 Long-term Training Simulator Upgrade 0-26 I.C.9 Long-term Program Plan for Upgrading 0-35
- of Procedures 1.D.1 Control Room Design Reviews 0-38 1.D.2 Plant Safety Parameter Display 0-71 Console 1.D.3 Safety System Status Monitoring 0-72 II.B.1 Reactor Coolant System Vents 0-73 11.b.2 Plant Shielding to Provide f.ccess to 0-75 Vital Areas and Protect Safety Equipment for Post-Accident Operation 11.B.3 Post-Accident Sampling 0-79 (N I1.b.b.3 Rulemaking Proceeding on Degraded 0-82 Core Accidents (Hydrogen Control)
, (v) II.D.) Testing Requirements 9-90 II.D.3 Relief and Safety Valve Position Indication 0-93 II.E.4.2 Isolation Dependability 0-94 II.E.4.4 Purging 0-100 II.F.1 Additional Accident Monitoring 0-107 Instrumentation 11.F.2 Identification of and Recovery from 0-110 Conditions Leading to Inadequate Core Cooling II.F.3 Instrumentation fo) Monitoring 0-111 Accident Conditions (Reg. Guide 1.97) 11.K.I.22 Describe Automatic and !!anual Actions 0-112 4 for Proper Functioning of Auxiliary Heat Removal Systems m en FW System is not Operable O II.K.3.23 central Water Level Recording 0-114 0-11 An.. No. 59, (6/81)
l ACNGS-PSAR l l
- TABLE OF C0hTEhT5 (Cont'd)
,, Item ho. Title Pagg III.A.I.2 Upgrace Licensee Emergen? 0-115 Support Facilities 111.D.I.!.
Primary Coolant Sources side the 0-116 Containment Structure 111.D.3.3 In-Plant Radiation Monitoring 0-123 111.D.3.4 Control Room habitability 0-124 h0 REG 0718 Category 5 0-126 1.C.5 Procedures for Feedback of Operating, 0-127 Design and Construction Experience l 1.F.! Expand QA List 0-135 1.F.2 Develop hore Detailed QA Criteria 0-136 11.B.8.2 Rulemaking Proceeding on Degraded 0-153 Core Accidents (3' Penetration) II b.8.4 kulemaking Proceeding on Degraded 0-154 l ('~' Core Accidents (Containment Capabil.y) 11.E.4.1 Dedicated Penetration 0-169 11.J.3.1 Organization and Staffing to oversee 0-170 Design and Construction i 1 i L 3 0-111 Am. No. S'9, (6/81) 1 _ . . - _ , --- . _ . , . , , - - . - - - - - - . . - - - , . -,..,....,n_ -
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- i. t LIST OF TABII.S i Table No. Title P_ age j I.A.4.2-1 Manpower Estimate During Construction 0-29 I.A.4.2-2 Comparison of Allens Creek and Black Fox O-30
Control Room Design j
- 1.D.1-1 System vs. Criteria 0-63 !
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- 1. D .1-2 . Panel Assignment Conclusions 0-65 ,
1 II.B.2-1 Core Inventory of Isotopes for 0-77 : ACNGS Unit No. I at Shutdown
+ t l . II.E.4.4-1 Guidelines for Demonstration of O-102 l a Operability of Purge and Vent Valves j III.D.1.1-1 Sununary of Lesak Reduction Criteria 0-119 ! 7
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; 11.B.8.4-1 - Systems Ncce sary to Ensure Containment &-168 :
f { Integrity During a Postulated Accident
; Vent .
[ 1 0-183 ! II .J . 3.1-1 Allens Creek Estimited Staf fing Levels 1, i d i i
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f ACNGS-PSAR l LIST OF FIGURES !
)
j Figure No. Title Page ! t + 1 j I.A.4.2-1 Allens Creek Control Room Layout 0-33 l 1.A.4.2-2 Black Fox Control Room Layout 0-34 i i II.B.3-1 location of Post-Accident RCS Sample Point 0-81 1 II.B.8.3-1 General Arrangement of Carbon Dioxide 0-88 i Inerting System (CDIS) (
! t II.B.8.3-2 Containment Pressure, Temp. History For 0-89 j The Inadvertant Actuation of The CO !
Post-Accident Inerting Sys. l s
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3 0 E 7 ACNGS-PSAR i APPENDIX 0
.l r
f- s RESPONSES TO NUREG 0718, " LICENSING REQUIREMENTS.FOR PENDING APPLICATIONS FOR 1 i { CONSTRUCTION PERMITS AND MANUFACTURING LICENSES" (MARCH, 1981) l
- }
NUREG 0718 lists the TMI Action Plan (NUREG 0660) items that the NRC Staff- ; } 4 will require to be addressed by near-term construction permit (NTCP) 57 l applications. The Action Plan items are divided into five categories of level !
- j. of detail to be provided in the NTCP submittal. This appendix gives the ACNGS !
j responses to the Action Plan items by NUREG 0718 category. I i I I'
- r i
f i , l l
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? l I i ! k i I l l I ! 1 [ 1 1 l I > i f f ) l 1 i I i l 4 i i , I ! i i' i j 0-1 Am. No. 59, (6/81) l ! i i i
ACNGS-PSAR NUREG 0718 CATEGORY 1 "A requirement of a type not applicable to the pending CP or ML applications for any of the following reasons:
- a. It can only be addressed in operating liccar, applications or by licen-sees;
- b. It is not directed to CP or ML applicants; 57 i c. It does not apply to plants of the type new pending;
- d. It has been (or will be) superseded by a more restrictive requirement in 3
the Action Plan or in the regulations;
- e. It has already been completed.
RESPONSE ; No response is required for NTCP applications, j t I 1 1 0-2 Am No. 59, (6/81) < l
ACNGS-PSAR NUREC 0718 CATECORY 2 (, ~, ) "A requirement of the typs customarily left for the operating license stage. The applicant should indicate its recognition of the need for development of operating license or final design requirements and should provide a commitment to implement such requirements in connection with its application for approval of the final design."
RESPONSE
57 The Action Plan items in NUREG 0718 Category 2 concern mostly operations. IIL&P has reviewed these items and has concluded that there is nothing to preclude their implementation at the OL stage of licensing. These will be addressed in detail in the FSAR. rN v
) \~_-) ;
O-3 Am. No. 59, (6/81)
_ .- -. - -. .-_ ~- -
-ACNGS-PSAR NUREG 0718 CATEGORY 3 .m I " Studies (and other research and development activities to provide design lfN- / development information) of the type customarily lef t for review at the final stage. However, to satisfy 50.35(a)(3) the staff believes that items in this category should be completed as early as is practicable so tcat the results ;
- can be most ef fectively .taken into account in developing final design de-tails. The applicant should provide sufficient information to describe the ;
nature of the studies, how they are to be conducted, the completion dates, and ! a program to assure that the results of such studies are factored into the , final design." AESPONSE
$7 ,
With the exception of Items II.B.8.1 (PRA study) and II.K.3.24 (Adequacy of i HPCI and RCIC space cooling), the NUREG 0718 Category 3 items are being , resolved between the NEC and the BWR Owners Group (BWROG). These items are being addressed generically and are being conducted so as to envelope all l P l ants participating in the studies. HL&P is participating in the BWROG i ef forts in this regard and concurs with the positions and recommendations of the Owners Group on the Category 3 items. HL&P commits to incorporate into the ACNGS design the resolution of these items agreed to between the NRC and BWROG. L A summary of the BWROG work done to date on each item is given herein. G J i k e L 8 6 s + t v 0-4 Am. No. 59, (6/81) i
ACNCS-PSAR (N 11. B .8 RULEMAKING PROCEEDING ON DECRADED (DRE ACCIDENTS NUREG 0718 REQUIREMENT
** Applicant shall:
(1) cosmit to performing a site / plant-specific probabilistic risk assessment and incorporating the results of the assessment into the design of the facility. The commitment must include a program plan, acceptable to the staff, that demonstrates how the risk assessment program will be scheduled so as to influence system designs as tney are being developed. The assessment shall be completed and submitted t o NRC within two years 57 of issuance of the construction permit. The outcome of this study and the NRC review of it will be a determination of specific preventive and mitigative actions to be implemented to reduce these risks. -A prevention feature that must be considered is an additional decay heat removal system whose functional requirements and criteria would be derived from the PRA study. It is the aim of the Commission through these assessments to seek such improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant, Applicants are encouraged to take steps that are in harmony with this aim."
RESPONSE
HIAF commits to performing an ACNGS specific Reliability Analysis Program l (RAP) study. The program status of this study, including schedule, is given in Appendix 158. 59 Os (v) 0-5 Am. No. 59. (6/81)
ACNCS-PSAR /
- II.K.2.16
)
N.__j ITEM II.K.2.16 IMPACT OF PfP SEAL DAMICE FOLLOWING SMALL-BREAK LOCA WITil LOSS OF OFFSITE POWER NUREG 0718 REQUIREPENT Applicants shall address the requirements set forth in the Commission Orders issued to operating B&W plants in May 1979 and set forth in Item B.4 of NUREG 0626 regardirs the impact of reactor coolant pump seal damage following ; a small break loss-of-coolant accident with loss of of fsite power. Applicants with B&W-designed plants shall provide suf ficient inf ormation to describe the j nature of the studies, how they are to be conducted, the completion dates, and l the program to a ssure that the result", of such studies are f actored into the t final designs. I SECY-81-20B ITEM II.K.3.25 EFFECT OF LOSS OF AC POWER ON PUMP SEALS Applicants with BWR plants shall addreas the requirements set forth in Item B .4 o f NU REG-062 6. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the 57 completion dates, and the program to assure that the results of such studies are f actored into the final designs. NUREC 0626 ITEM B.4 ('~'} The licensees should determine by analysis or experiment, on a plant specific g'-'j basis, the consequences of a loss of cooling water to the reactor recirculation pump se 21 coolers. The pump seals should be designed to s stand a complete loss of alternating current power for at least two hours. Adequacy of the seal design should be demonstrated. (a) Na t u re o f St ud y This concern relates to the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. Adequacy of the seal design should be demonstrated. The recirculation pump design incorporates a dual mechanical shaf t seal assembly to control leakage around the rotating shaf t of the recirculation pump. Each assembly consists of two seals built into a cartridge that can be replaced without removing the motor from the pump. Each individual seal in the cartridge is designed f or full pump design pressure and can adequately limit leakage in the event that the other ecal should fail. Even though General Electric uses two different recirc pump configurations, the seal designs are essentially the same. Both designs use hyd ostatically balanced mechanical shaf t seals. Subsequent discus sion in this memorandum is applicable to both pump designs. (m\ s /
\j 0-6 Am. W. 59, (6/81)
J ACNGS-PSAR ; II.K.2.16
./'~' The recirc pump seals require f orced cooling due to the temperature of ;
the primary reactor water and due to ti.e friction heat generated in the i sealing surf aces. For all BWR/6 Reactors two systems accomplish this f orced cooling: .(1) the equipment protection closed cooling water system and (2) the seal purge system. Cooling water, provided by tia equipment protection closed cooling water (EPCCW) system, flows through a heat exchariser around the seal assembly. This EPCCW flow cools primary reactor water which flows to the lower seal cavity thereby maintaining ' the seals at the correct operating temperature. The seal purge system
- injects clean, cool water froe the control rod drive system into the j lower seal cavity. This seal purge flow also provides an efficient cooling function f or the seals. I j 'The seal cooling system described above is examined to determine the
! consequenses of a total loss o' cooling on the effectiveness of l { recirculation pump shaf t sealing. {
' ( b) Conduct of Study '
Under normal conditions, with the primary reactor system at or near rated ' i temperature and pressure and the recirc pumps either operating or
; secured, both EPCCW and seal purge are operating. These two systems j maintain the seal temperatures at approximately 1200F.
' 57 ' Recirculation pump vendor test data have shown that the pump seals may begin to deteriorate when seal temperatures exceed 2500F. If an event ] occurs whe:e both closed cooling water to the pump seal heat exchanger i and control rod drive seal purge flow are totally lost, the recire pump j seals will heat up. Vendor test data, taken while operating at , approximately 5300F/1040 PSI A, indicate that the seals will heat up, reaching 250*F approximately 7 minutes'af ter the total loss of cooling. l Similar test data indicate that if either one of the seal cooling systems { is operating, the seal temperatures remain well below 2500F and no seal deterioration should occur. l If both closed cooling water and seal purge are totally lost, and if the
- seals heat up to exceed 2500F, seal deterioration may occur, resulting in prima ry coolant leakage to the drywell. In onder to evaluate the fluid loss through a degraded seal, an analysis was perf ormed using ti.e j- RELAP-4 computer program (see Reference 1). ,
; This analysis modelled the fluid leakage path as a series of fluid i volumes with interconnecting junctions, each having appropriate initial conditions. Also, the model assumed gross degradation of the mechanical
' seals. Gnoss failure of these seals encompasses warpage, tractures and grooving of the seal faces due to excessive thermal gradients and dirt. i 0-7 Am. No. 59, (6/81)
ACNGS-PSAR 11.K.2.16 O ' The results of this leakage analysis show that, even with gross degradation of the seals, the leakage would be less than 70 gallons per minute. This amount of leakage is within normal reactor fluctuations and the normal vessel water level control systems will easily compensate for it. Also, 70 GPM is much less than the bounding vclues of loss-of-coolant accident analyses, hence there are no adverse ef fects on LOCA analyses. (c) Completion Date
; The study is complete, and was transmitted to the NRC in Reference 1.
(d) Program for Implementation of Results . The study concluded that the leakage through a grossly failed RCP seal is of no consequence to any of the LOCA analyses. Therefore, no changes are required to implement the results. REFERENCES
- 1. NEDO-24083, " Recirculation Pump Shaf t Seal Leakage Analysis", November 1978. (Licensing Topical Report) 0 0
O 0-8 Am. Na. 59, (6/81)
_t ACNGS-PSAR II.K.3.13 ITEM II.K.3.13 SEPARATION OF HPCI AND RCIC SYSTEM INITIATION LEVELS - l- Q : ANALYSIS AND IMPLEMENTATION , NUREG 0718 REQUIREMENT 4. Applicants with BWR plants shall address the requirements set forth in Item A.1' of NUREG-0626 as they apply to HPCS and RCIC systems. Applicants.shall 5~ provide sufficient information to describe the nature of the studies, how they , are to be conducted, tne completion dates,_and the program to assure that.the , results of such studies are factored into the final designs. i. NUREG 0626' ITEM A.1' Currently, the reactor core isolation cooling (RCIC) system and the high . pressure coolant injection (HPCI) system both initiate on the'same low water level signal and both isolate on the same high water level signal. The HPCI system will restart on low water level but the RCIC will rot. The RCIC system L is a low-flow system when compared to the HPCI system. The initiation levels
~
of the HPCI and RCIC system should De separated so that the RCIC system initiates at a higher water level than the HPCI system. Further, the RCIC ' system initiation logic should be modified.so that the RCIC system will restart on low water level. These changes have the potential to reduce the 57 number of challenges to the HPCI system and could result'in less stress on the
; vessel from cold water injection. Analyses should be performed by GE to i evaluate these changes. The analyses should be submitted to staff and changes should be implemented if justified by -the analyses.
. RESPONSE
- a. . Nature of Study t
f This~ concern covers two aspects of the HPCS and RCiC r.ystems. The first 4. concern is with the initiation levels of these two systems, and requests analysis to determine if benefit could be obtained from allowing the RCIC
- 4. system to initiate from a higher water level than the HPCS. The second concern is with automatic restart of the RCIC system, and requests analysis to determine if benefit could be gained by introducing this
. feature.
As previously confirmed in discussions with the NRC the fundamental issue j of the separation of initiation setpoints (water level) is the potential benefit of reducing the number of thermal cycles on the reactor vessel and internals resulting from HPCI. operation. It is noted that the Allens l
- Creek plant employs HPCS which does not inject via the feedwater nozzle, consequently the fatigue usage on this component is reduced. Thus the j i study of this issue, which was based mainly on the BWR/4 HPCI arrangement '
1~ ~is conservative for Allens Creek. i = 0-9 Am. No. 59, (6/81) i
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-ACNGS-PSAR II.K.3.13 Analysis was also made to evaluate the proposed logic change for the RCIC system which permits this system to restart automatically following 7-~s}
\s /- isolation from high water level. This evaluation considered the logic
. changes involved, ef fect on system availability, impact on design reliability and the operater/ equipment interface. ,
- b. Conduct of Study a) Setpoint Separation The analyses conducted are for typical BWR/3 and 4 designs where the HPCI and RCIC systems inject via the feedwater spargers. Later plant designs (BWR/S and 6) have a separate injection location for HPCS and are less limiting in comparison to the typical BWR/3 and 4 configuration. Dif ferences in the thermal fatigue analyses are identified were appropriate.
The discussion of the study addresses the potential for reducing the , thermal cycles due to HPCI and RCIC initiation. The transients considered are those cited in PSAR Chapcer 15. Two classes of transients can cause RCIC and HPCI initiation:
- 1. Initiation of HPCI and RCIC on low water level af ter feedwater is tripped on high reactor water level. For these transients, the inventory is slowly lost due to decay heat steam generation.
- 2. Initiation of HPCI and RCIC following a sudden loss of g feedwater. For these transients, inventory loss is rapid with N' HPCI and RCIC initiation occurring approximately 20 seconds 57 after event initiation.
The details of this study are provided in Reference 1. b) Automatic Restart of RCIC System NUREC-0626. Item A.1, requires evaluation of changes to the Reactor Core Isolation Cooling System to allow automatic restart following a trip of the system at high reactor vessel water level. The evaluation of this change showed that it would contribute to improve system reliability and that it could be accomplished without adverse effect on system function and plant safety. The recommended change would be to relocate the existing high level trip from the RCIC l turbine trip valve to the steam supply valve. Once the level l reaches a predetermined high level the steam supply valve would be ! closed. One additional relay in the logic circuitry would be required to accomplish the new function. Closure of the steam supply puts the system in a partial standby configuration because of the existing interlocks associated with closure of this valve. Very little modification to the logic circuitry is required to automate v 0-10 Ac. No. 59, (6/81)
ACNGS-PSAR II.K.3.13 s' realignment of the system in preparation for low water level initiation. This approach was one of several options considered. v The details of this study are provided in Reference 2.
- c. Completion Date Completed.
- d. Program for Implementation of Results a) Separation of HPCS AND RCIC Setpoints The results of the analyses for this issue indicate that no significant reduction in thermal cycles can be achieved by separation of these setpoints. It is therefore proposed that the current design values be retained.
b) Automatic Restart of RCIC System The re sults of the analyses for this issue indicate that the proposed logic change would contribute to improved system reliability, be of l 57 assistance to the plant operator and generally enaance safety. This change can be incorporated into the design, and will be upon NRC l approval of the BWROG study, and will be described in the FSAR. l 59
/'N Ref e rence s:
( k'
)
- l. Letter from R.H. Buchholz (GE) to D.G. Eisenhut (NRC) dated October 1, 1980 and titled "NUREG-0660 Requirement II.K.3.13".
57
- 2. Lettet from D.B. Waters (BWR Owners' Group) to D.G. Eisenhut (NRC) dated Decembe r 29, 1980 and titled "BWR Owners' Group Evaluation of NUREG-0737 Re quire men t s" .
v 0-11 Am. No. 59, (6/81)
ACNGS-PSAR II.K.3.16 ITEM II.K.3.16 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF VALVES - rs FEASIBILITY STUDY AND SYSTEM MODIFICATION
\
4
~~- NUREG 0718 REQUIREMENT Applicants with BWR plants shall address the requirements set forth ia Item A.4 of NUREG-0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into the final designs.
NUREG 0626 ITEM A.4 The record of relief valve failures to close for all BWRs in the past three yearo of plant operation is approximately 30 in 73 reactor years (0.41 failures / reactor year). This has demonstrated that the failure of a relief valve to close would be the most likely cause of a small-break LOCA. The high failure rate is the result of a high relief valve challenge rate and a 57 relatively high failure rate per challenge (0.16 failures / challenge). Typically, five valves are challenged in each event. This results in an equivalent failure rate per challenge of 0.03. The challenge and failure rates can be reduced in the following ways: (1) Additional anticipatory scram on loss of feedwater, (2) Revised relief valve actuation setpoints, jp} (3) Increased emergency core cooling (ECC) flow, \ (4) Lower operating pressures, (5) Earlier initiation of ECC systems, (6) Heat removal through emergency condensers, (7) Of fset valve setpoints to open fewer valves par challenge, (8) Installation of additional relief valves with a block or isolation valve feature to eliminate opening of the safety / relief valves (SRVs), consistent with the ASME code, (9) Increasing the high steam line flow setpoint for main steam line isolation valve (MSIV) closure, (10) Lowering the pressure setpoint for MSIV closure, (11) Reducing the testing frequency of the MSIVs, A N~_- l 0-12 Am. No. 59, (6/81)
1 4
.ACNGS-PSAR II.K.3.16 s t, (12) More stringent valve leakage criteria, and r i >
/
(13) Early removal of~ leaking valves. CE should investigate the feasibility and contraindications of reducing challenges to the relief valves by use of the' aforementioned methods. Other methods should also be included in the feasibility study. Those changes which ; are shown to reduce relief valve challenges without comprising the performance of the relief valves or other systems should be implemented. Challenges to i the relief valves should ha reduced substantially (by an order of magnitude). ' i
RESPONSE
I i s.. Nature of Study ' i 1 This report documents a study performed in response to NUREG-0626 item A.4 which requires an evaluation of the feasibility and contraindications { of reducing challenges to the relief valves by various methods in BWRs. The report reviews potential methods of reducing the likelihood of stuck ,, open relief valve (SORV) events in BWRs and estimates the reduction in such events that can be achieved by implementing these methods. Reducing the likelihood of S/RV challenges will directly reduce the likelihood of a SORV. In addition, attention is also given to 57 modifications which could reduce spurious SRV blowdowns and to modifications which could reduce the probability of SRVs to stick open ;
- when challenged. l I u_,l ~
l ] b. Conduct of Study . i Although the study was precipitated b/ the consideration of reducing
, challenges to the Safety / Relief Valves (SRVs), it was recognized that the j true objective was to reduce the incidence of Stuck Open Relief Valve (SORV) events. In line with this approach the study also considered
( reducing the causes of spurious blowdowns and reducing Ohe probability of '
; SRVs to stick open when challenged. The goal of the study was to -
identify feasible modifications to BWR design and operation, which reduce l the frequency of uncontrolled blowdowns by a factor of ten relative to ' the BWR/4 case, which was used as a base for evaluation. l The details of this study are provided in Reference 1. For the BWR/6 j plants such as ACNGS it was concluded that no changes are required to ; achieve a factor of ten reduction (relative to operating experience) l because: j e j 1. desida features which reduce SRV challenges are already incorporated. ) II .
- 2. the two-stage Crosby valves to be used are less likely to stick open I
due to design differences from the three-stage Target Rock valves on f which the operating experience is based. ' l L (s_ /
+
l. 0 13 Am. No. 59, (6/81) v w r e, w *w - * - - - --
<w, -,~v + - -
--m,---n-ew-*- 'um e __-ev
- ACNGS-PSAR l
II.K.3.16
- c. Completion Date 9 Complete
- d. Program for Implementation of Results The study indicates that the required factor of ten improvement relative to operating experience is met by the present design. Thus, no changes are required to implement the results.
57 Re f erences
- 1. Lettcr from D.B. Waters (BWROG) to D.G. Eisenhut (NRC) dated March 31, 1981 and titled "BWR Owners Group Evaluation of NUREG 0737 Requirements."
~
0-14 Am. No. 59, (6/81)
ACNCS-PSAR II.K.3.18 ITEM 11.K.3.18 MODIFICATION OF ADS LOGIC - FEASIBILITY STUDY AND MODIFICATION , '~'} FOR INCREASED DIVERSITY FOR SOME EVENT SEQUENCES s NUREG 0718 REQUIREMENT Applicants with BWR plant shall address the requirements set f orth in Item A.7 of NUREG-0626. Applicants shall provide sufficient inf o rmation to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are f actored into tre final designs. NUREG 0626 ITEM A.7 The ADS actuation logic should be modified to eliminate the need for manual actuation to assure adequate core cooli ng. A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme which should be conside red is ADS actuation on low reactor vessel water level provided no HPCI or HPCS system flow exists and a low pressure ECC system is ru nni ng. This logic would complement, not re plac e , the existing ADS actuation logic. RESPONSd
- a. Nature of Study A feasibility and risk assessment study was made to examine possible
,-~s modifications to the ADS initiation lo 'c, which would eliminate the need f or manual initiation to assure adequate core cooling. For some non-line 57
[ } ( ,/ break events which are further degraded by assuming non-availability of all high pressure injection systems, manual depressurization of the reactor is required in order to employ the low pressure injection systems. This study examines the advantages and disadvantages of a number of possible ADS initiation logic modifications.
- b. Conduct of Study Five ADS logic alternatives were considered: the current design, and four logic modifications. These four modifications are 1) elimination of the high drywell pressure trip, 2) addition of a timer that bypasses the high d rywell prussure trip requirement af ter a certain length of time, 3) addition of a suppression pool temperature trip in parallel with the high drywell pressure trip, and 4) the addition of high pressure system flow measurement and logic in parallel with the high drywell pressure trip.
Each of the 9ptions' is evaluated on the basis of whether it assures aiequate i ure cooling without operator action for isolations and SORV's. Each option is also evaluated for its capability to assure adequate core cooling without operator action. For these analyses it i s assumed that all high piessure systems have failed and the ADS must depressurize the vessel and allow the low pressure systems to inject. The modeling used in these analyses is the same as that used in NEDO-24708. /Sl \ The details of this study are provided in Reterence 1. v 0-15 Am. No. 59, (6/81)
ACNCS-PSAR 11.K.3.18
- c. Completion Date The study is comple';e and was transmitted to the NRC by Reference 1.
- d. Program for Impleraentation of Results s
> The BWROG conclu jed that an ADS modification which adds a bypass timer on i ECCS initiation level or removal of the high drywell pressure trip would [ be beneficial. These changes would not have any major impacts on the plant design. They can be readily incorporated, and will upon NRC/BWROG 4 resolution of the item, 57 REFERENCES i
< l. Letter from D.B. Waters (BWROG) to D.G. Eisenhut (NRC) dated March- 31, 1981 and titled "BVR Owners Group Evaluation of NUREG 0737 Requirements."
i k e
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1 4 l f 0-16 Am. No. 59, (6/81) l-
ACNGS-PSAR II.K.3.21
~x ITEM II.K.3.21 RESTART OF CORE SPRAY AND LPCI SYSTEMS ON LOW LEVEL - DESIGN
( ) AND MODIFICATION s~ NUREC 0718 REQUIREMENT Applicants with BWR plants shall address the requirements set forth in Item A.10 of NUREG 0626. Applicants shall provide suf ficient inf ormation to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are f actored into the final designs. NUREG 0626 ITEM A.10 The core spray and LPCI system flow may be stopped by the operator. These systems will not restart automatically on loss of water level if an initiation signal is still present. The core spray and LPCI system logic should be modified so that these systems will restart if required to assure adequate core cooling. Because this design modification af fects several core cooling modes under accident conditions, a preliminary design should be submitted for staf f review and approval prior to making the actual modification. RESPONSE 57
- a. Nature oof Study In this item, the NRC Suggested Certain Modifications to the Core Spray
/T (CS) and the Low Pressure Coolant Injection (LPCI) Emergency Core Cooling ( ) Systems (ECCS) that are provided as part of the BWR ECCS network. The
' ~ ' '
NRC suggestions center on incorporating additional control system logic to provide automatic system restart f rom a low reactor water level signal following actions by the operators to terminate system operation. The NRC concern is that the reactor operators may terminste ECCS operation when a high reactor water level condition exists but may neglect to reinitiate the systems if a low condition recurs. The study, which covers the Allens Creek plant design, includeds the LPCI and both the low and high pressure core spray systems. Intuitively, it might appear that additional ECCS automation would be purely beneficial since this would supposedly provide added protection against operator errors and omissions. H9 wever, these perceived benefits of extended system automation must be measured against the very real penalties of increased system complexity, reduced system reliability and restricted operator flexibility for dealing with unanticipated events. These considerations are not amenable to precise quantification and control system design decisions must of necessity involve judgements as to relative importance of these competing influences. (p) v 0-17 Am. No. 59, (6/81)
'ACNCS-PSAR II.K.3.21
- b. . Conduct of Study
's / :In order to determine if any overall benefit is to be derived f rom the postulated design changes it is necessary to consider the integral nature of the ECCS network, and how the ECCS interacts with other plant systems. The study provides an overview discussion of the generic GE
- ECCS design philosophy and design practices as they govern ECCS initiation and operator control of these systems. The need for operator override is identified, ar.d how this feature provides for improved overall system reliability. Considerable significance is attached. to the complexity of logic and hardware, which would be required to deal with relatively long-term transients involving core and containment cooling, on a purely automatic basis. Several long-term transient scenarios are presented to support this contention.
The details of this study are provided in Reference 1. 57
- c. Completion Date
'The study is complete and was transmitted to the NRC Reference 1.
- 3. Program for Implementation of Results The study concluded that while changes to the LPCI/LPCS logic would not have a net positive sa ety ef fect, modifications to the HPCS logic to assure a restart ois Icw reactor water level would. This can be readily incorporated into the ACNGS design and will upon NRC/BWROG resolution of
[sI this item. REFERENCES
- 1. Letter f rom D.B. Waters (BWR Owners' Group) to D.G. Eisenhut (NRC) dated Decembe r 29, 1980 and titled "BWR Owners' Group Evaluation of NUREG-0737 i Requirements".
i ' b U-0-18 Am. No. 59, (6/81) 4
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r; ACNGS-PSAR II.K.3.24 j] ITEM II.K.3.2 4 CONFIRM AD2QUACY OF SPACE COOLING FOR HPCI AND RCIC SYSTEMS NUREG 0718 REQUIREENT Applicants with BWR plants shall address the HPCI and RCIC systems requirements set f orth in Item B.3 of NUREG 0626. Applicants shall provide suf ficient in-f onnation to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are f actored into the final d.esigns. NUREC 0626 ITEM B.3
' Long-term operation of the RCIC and HPCI systems may require space cooling t 57 maintain the pump room temperatures within allowable limits. The licensees should verify f or each plant the acceptability of the consequences of a complete loss of alternating current power. The kCIC and HPCI sys ems should be designed to withstand a complete loss of alternating current power to their ;
support systens, including coolers, f or at least two hours. .
RESPONSE
The HPCS and RCIC systems are designated as safety-related for ACNGS, and as such, are designed to operate independent of of fsite power *. As with all safety systems, HPCS and RCIC are serviced by a safety-related cooling system, which is itself independent of of fsite power. The cooling system is sized and ps s designed to maintain a suitable environment for HPCS and RCIC components
;( j following a loss of of fsite power.
*The BWR Owners Group received clarification that the intent of this item was l 59 to assume loss of of fsite power only. This was confirmed by letter, D B 157 Waters (BWROG) to D.C. Eisenhut (NRC) " Clarification of NUREG 0737 Items II.K.3.24 and II.K.3.25, dated 1/23/81. 59 O
i 1 l V . 0-19 Am. Na. 59, (6/81)
ACNCS-PSAR II.K.3.28 ITEM II.K.3.28 VERIFY QUALIFICATION OF ACCUMULATORS ON ADS VALVEF NUREG 0718 REQUIREMENT
" Applicant s with BWR plants shall provide in format ion t o assure that the ADS valves, accumulat ors, and associat ed equipmant and inst rument at ion will be capable of performing their int end;4 funct ions during and following a.1 accident sit uat ion while t aking no credit for nonsafet y related equipment or inst rument at ion. Air (or nit rogen) leakage through valves must be account ed for t o assure that enough inventory of compressed air (or nit rogen) will be available to cycle the ADS valves. Appl icant s shall commit that these requirement s will be met in the final design at the OL stage.
In addressing this it em prior to CP issuance, applicant s should not e that safet y analysis report s claim that air (or nit rogen) accumulat ors for the ADS valves provide suf ficient capacit y (inventory) to cycle these valves open five t imes at design pressures. Also, General Elect ric has st at ed that the emergency core cooling systems are designed to withst and a host ile environment and st ill perform their funct ions for 100 days following an accident ." 57
RESPONSE
The present ADS air accumulators are sized to cycle the ADS valves twice aga inst 70% of cont ainment design pressure (or five t imes against cont ainment atmospheric pressure) plus component leakage for seven days. Post accident
,y access to replenish the air supply (assuming t hat the supply compressors are
( ) inoperative) is being confirmed as part of the post accident shielding study __' in response to Item II.B.2. The radiat ion environment al qualificat ion for the ADS air accumulators and associated components for at least 100 days will be confirmed by this study as well, ilL&P is participating in the BWR Owners Group ef fort s to agree wit h t he NRC on a uniform design basis for ADS air accumulat or siziag. The result s of this e f fort will be adopt ed for ACNGS and design changes made if necessary. 1
,~s i t 1
' l l l
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0-20 Am. No. 59, (6/81)
-)
ACNGS-PSAR II.K.3.44 ITEM II.K.3.44 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE FAILURE TO f d~'s VERIFY NO SIGNIFICANT FUEL FAILURE i\~-) NUREG 0718 REQUIREMENT Applicants with BWR plants shall address the requirements set forth in Item A.14 of NUREG 0626. Applicants shall provide sufficient information to describe the nature of the studies, how they are to be conducted, the completion dates, and tne program to assure that the results of such studies are factored into the final t.esigns. NUREG 0626 ITEM A.14 I For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery. Transients which result in a stuck-open relief valve should be included in this category. , RESPONSE
- a. Nature of Study Analyses of the worst anticipated transient (loss of feedwater event) with the worst single failure (loss of a high pressure inventory makeup 57 or heat removal system) were performed to demonstrate adequate core cooling capability. It is shown that, for the BWR/2 through BWR/6 1
f plants, adequate core cooling is maintained for these worst-case q
\s_,}/ conditions. Analyses of further degraded conditions involving a stuck open relief valve in' addition to the worst transient and single failure were also performed. The results show that, with proper operator i action, the core remains covered and therefore adequate core cooling is achieved.
- b. Conduct of Study j Of the two alternate criteria allowed in Item A.14 of NUREG 0626, the study demonstrated that for the combination of anticipated transients with the worst single failure, tha reactor core remains covered until stable conditions are achieved. The following assumptions are also made:
,- a. A representative plant of each BWR product line, BWR/2 through BWR/6, is used to represent all of the plants of that product line. l The BWR/6 analyses are applicable to ACNGS.
; b. The anticipated transients as identified in NRC Regulatory Guide 1.70, Revision 3 were considered.
1 4 4 v 0-21 Am. No. 59, (6/81) i l
ACNGS-PSAR II.K.3.44
- c. The single failure is interpreted as an active failure.
/ i t s
~~'
/ d. All plant systems and components are assumed to function normally, unless identified as being failed.
For a BWR/6 plant such as Allens Creek the study indicates that the worst combination is a loss of feedwater with failure of the HPCS. The study further shows that even for the further degraded condition of a stuck open relief valve in addition to the worst single failure / worst transient combination, the core cr.n be kept covered. The details of this study are provided in Reference 1.
- c. Completion Date The study is complete, and was transmitted to the NRC in Reference 1. 57
- d. Program for Implementation of Results The study concluded that there are no anticipated transient / single failure combinations which result in significant fuel damage. There fore ,
no changes are required to implement the results. REFERENCES
- 1. Letter from D.B. Waters (BWR Owners' Group) to D.G. Eisenhut (NRC); dated December 29, 1980 and titled: "BWR Owners' Group Evaluation of A)
( \~ / NUREG-0737 Requirements". -
/ \ \J 0-22 Am. No. 59, (6/81)
L ACNGS-PSAR II.K.3.45 fx
;j ITEM 11.K.3.45 EVALUATE DEPRESSURIZATION WITH 'OTHER THAN FULL ADS
)
N_^s
- NUREC 0718 REQUIREMENT Applicants .with BWR plants shall address the requirements set forth in' Item A.15 of NUREG-0626. Applicants shall provide sufficient . information to describe the nature of the studies, how they are to be conducted, the completion dates, and the program to assure that the results of such studies are factored into i the final designs. ,
NUREG 0626 ITEM A.15
' Analyses to support depressurization modes other than full actuation of the A06 (e.g. , early blowdown with.one or two SRVs) should be provided, Slower depressurization would reduce the possibility of exceeding vessel integrity 1.mits by rapid cooldout.
RESPONSE
- a. Nature of Study This feasibility study addresses NUREG-0626, Item A.15, and provides 57 l an evaluation of alternate modes of reactor depressurization than full l actuation of the Automatic Depressurization System (ADS). The study I includes the BWR/6 product line and therefore the Allens Creek plant.
(""' b. Conduct of Study
\ ,,f Depressurization by full ADS actuation constitutes a depressurization ,
from about 1050 psig to 180 psig in approximately 3.3 minutes. Such l an event, which is not expected to occur more than once in the life- ! time of a plant, is well within the design basis of the reactor pressure vessel. This conclusion is based on the analysis of several ! transients requiring depressurization via the ADS valves. Results of these analyses indicate that the total vessel fatigue usage is less i than 1.0. Therefore, no change in the depressurization rate is necessary. However, to comply with NUREG-0626, Item A.15, reduced depressuriza tion rates were analyzed and compared with the full ADS actuation. The 1-alternate modes considered cause vessel pressure to traverse the same pressure range in 1) depressurization case 1 { ranges from 6-10 minutes
' depending on plant size and ADS capacity and 2) depressurization case 2 (ranges from 15-20 minutes). The case 2 depressurization bounds the possible increase in depressurization time by producing an undesirably a long core uncovered time. The case 1 depressurization gives the results of an intermediate depressurization. These modes are achieved by opening a reduced number of relief valves.
The details of this study are provided in Reference 1. O' V 0-23 Am. No. 59, (6/81)
,i-.
1 ACNCS-PSAR l I II.K.3.45 c.- Completion Date-
\;
The study is complete and was transmitted to the NRC in ~ Reference 1.
- d. Prostram' for Implepentation of Results _ ,
I The study concluded that there is no benefit to be derived from the use 57 1 of reduced blowdown rates. Therefore, no changes are required to im-l plement the results. t REFERENCES l 1 ^
- 1. Letter from D.B. Waters (BWR Owners' Group) to D.G. Eisenhut (NRC);
- j. dated December 29, 1980 and titled: "BWR Owners' Group Evaluation ,
. of NUREG-0737 Requirements".
j i 1 $ r 1 1 .' i 5 i j i l j I I i ! 4 i i s != 0-24 Am. No. 59, (6/81) f.
ACNGS-PSAR / s NUREC 0718 CATEGORY 4 N I
'"A requirement to demonstrate that any additional design, development and inplementation necessary to satisfy the requirement (or to satisfy the goals of the task whose requirements are to be developed in the future) will be satisfactorily completed 'by the operating license stage. This is the type of information customarily required at the construction permit stage to satisfy 50.35(a)(2), or to satisfy ALAB-444 with respect to generic issues."
RESPONSE
Responses to the applicable Category 4 items are given herein, including PSAR level of information and detail where design is involved. It should be noted that for most of these items there are no questions as to the ability to im-plement the requirement prior to issuance of an operating license. The following Category 4 itema are not applicable to ACNGS: II.E.1.2 Auxiliary Feedwater System Autonatic Initiation and Flow indica- 57 tion applicable to PWRs only. II.E.3.1 Reliability of Power Supplies for Natural Circulation - applicable to PWRs only. II.E.5.1 B&W Reactors Design Evaluation epplicable to B&W NSSS plants only. ,/ m ( ) II.E.5.2 B&W Reactor Transient Response Task Force applicable to B&W NSSS
'v' plants only.
II.G.1 Power Supplies for Pressurizer Relief Valves, Block Valves and Level Indicators applicable to PWRs only. II.K.2.9 Procedures and training to initiate and control AFW independent of integrated control system applicable to B&W NSSS plants only. II.K.2.10 Hard wired safety grade anticipatory reactor trips applicable to B&W NSSS plants only. II.K.3.ll Control use of PORV supplied by Control Components Inc applicable to PWRs only. b'i t ; G 0-25 Am. No. 59, (6/81)
ACNGS-PSAR l I.A.4.2 ,
-m ITEM I .A.4.2 LONG-TERM TRAINING SIMULATOR UPGRADE
, NUREG 0718 REQUIREMENT Applicants shall describe their program for providing simulator capability for their plant s. In addition, they shall describe how they will assure that their proposed simulator will correctly model their control room. Applicants shall provide suf ficient inf ormation to permit the NRC staf f to verif y that they will have the nt :essary simulator capability to carry out the actions described in this Action Plan item as well as Action Plan Item ll.K.3.54.
Applicants shall submit, prior to the issuance of construction permits, a 57 general discussion of how ':he requirements will be met. Suf ficient details shall be prescated to provide reasonable assurance that the requirements will be implementea properly prior to the issuance of operating licenses. RESPONSE i HL&P intends to use the Black Fox simulator for operations training in aupport o f ACNGS. The simulator will be used for - i (1) continuing training of licensed personnel (2) initial attainment of cold licenses (3) supplemental training for advancement f rom RO to SRO ; I ,7 ~ S (4) training of home office and plant staf f engineering personnel and f q l auxiliary operators. In addition, the licensed operator training program will meet the requirements of the following documents:
- 1. ANS 3.1, 4/10/81, Standard for Qualification a nd Training of Personnel for Nuclear 'ower Plant s.
- 2. 10 CFR Part SS, Operators Licenses.
- 3. REG. GUIDE 1.149, Nuclear Power Plant Simulators for Use in Operator 59 Tr a ini ng .
These requirements will be accomplished in a timely manner to support sta rt up and operation. Table I .A.4.2-1 provides an estimate of the manpower schedule to support operator training and assignment. The ACNGS license candidate training progr a will be typical of that defined in ANS 3.1, Appendix A. The control room at both Black Fox and ACNGS are Nuclenet 1000 advanced control room designs ( Ref erence Item 1.D.1 - Control Room Design Reviews) . 1 The control room floor plans for the primary control panels are the same (see Figures I .A 4.2-1 and L .A.4.2-2) . A comparison of functions is found in Table I.A.4.2-2. A general discussion of the similarities and differences in the panels is discussed below. ,m,
/
0-26 Am. No. 59, (6/81)
-_~ -. _- . ,,] - ACNGS-PSAR \v- ).
Panels P680, P601, P678 I.A.4.2 Panel P680 is the principal plant operations control console, and the
' day-to-day normal plant operations functions, short response functions, and reactivity control functions are located on this panel. ,
The functions / systems on this panel are the same for both plants. The layouts-are essentially the same, except for the condensate pumping system, feedwater 59 pumping systes, and generator controls. These systems are the responsibility of che utility, which makes them utility un que.~ The other systems are designed by General Electric. ' I Panel P601 is the reactor core cooling benchboard and contains the NSSS safety ; t systems and NSSS long response functions. The operator uses this panel during . abnormal conditions and testing. The functions / systems on this panel are the same for both plants. The layouts are essentially the same, except for the ; , minor differences in layouts of the RHR system. The panel is designed by l General' Elec tric. ; , Panel P678 is the Standby-Information Panel. It is used as a support function , f or the P680 panel. It contains no control functions. The functions / systems are the same; however, the layouts are different due to utility preferences. , The dif ferences are minor and do not interfere with operator performance. , t , (j$_ During simulator training operators' using these ponels will learn how to use l the CRT displays and their interaction with the associated controls. In '57 : addition, the operators will learn to use the benchboards during the transients and accidents. The control rooms are essentially duplicated for these very important functions so that the operaters will gain experience ' directly applicable to ACNGS through the simulator training. Panels P870, P800, P868, P869, P877 t Panel P870 is the BOP control benchboard and contains the BOP long response functions and non-frequent use functions. The panel is designed by the individual utility to be plant specific. The functions / systems on this panel are essentially-the same; however, those- functions / systema that are not on P870 are found on other panels in the control room. The layouts of the panels are dif ferent because of differences in utility philosophy.
; Panel P877 contains controls for the standby diesels and support systems. The functions / systems are the same; however, the layouts are different because this papel is designed by the utility. 59 Panel P800 is the BC
'%d (5) ROD CONTROL AND INFORMATION SYSTEM (6) NEUTRON MONITURING SYSTEM (7) STEAM BYPASS AND PRESSURE REGULATOR SYSTEM (8) MAIN TURBINE CONTROL SYSTEM (9) CENERATOR CONTROL SYSTEM Human f actors engineering principles and criteria were used to evaluate human-machine interf aces in analyzing performance requirements for plant control functions and f or the allocation of f unctions to categorize these nine systems. Allocation categorias consisted of: (1) Automatic operation by plant systems equipment (2) Manual operation by control room Operators and/or plant Technicians (3) Some combination of (1) and (2) \ Am. N. 59, (6/81) 0-43 ACNCS-PSAR !.D.1 / T The design evaluation allocation criteria considered the capabilities and g y/ limitations of the Operator (s) and Systems, along with cost-benefit con-siderations of automating in those instances where the Operator and system could perform a given task approximately equally well. Factors comparable to those Listed in Table B-1 of Appendix B to NUREG-0659 were used in making the allocation of functions. 2.1 Operator / Technician Processing Capabilities Plausible human rolee of Operators, Technicians, and Supervisors (e.g. , control manipulator, instrument monitor, supervisor, decision maker, communicator, equipment repairer, coordinator) have been defined. Qualitative inf ormation processing capability in terms of load, accuracy, rate, and time delay have been prepared for each Operator / Supervisor inf ormation processing function. 2.2 System Processing Capabilities Plausible system roles of Control Complex equipment (autowatic con-trol of reactor flux, reactor trip system, engineered safety feature) ' have been defined. Inf ormation processing capabilities and control . function response times of control systems equipment have been de-fined consLiering load, accuracy, rate, and time delay for processing and re spor.se. p 2.3 Responsibility f or Plant Saf ety \V] The overall responsibility f or the top-level assessment of plant 57 operating and saft ty status has been allocated to the human Oper-a tor (s) . The rationale for this allocation is based on the cognitive abilities of humans, which cannot be duplicated by a machi ne. The inf ormation requirements to exercise this responsibility determine methods for transf er of plant systems data and_ inf ormation to the - operator (s) in the Control Complex. 2.4 Results of Allocation i l A suaimary description of the allocation of functions follows: l l Reactor Watee Cleanup System - This system is operated manually. ' This is an instance where the operator or machine can perf orm approx-imately equally well, and the system objective is achieved by manual operation. 'Ihis operation does not overload the Operator, and it was not cost-beneficial to automate. O 0-44 Am. No. 59, (6/81) ACNGS-PSAR I.D.1 [ Condensate Pumping System - This system is primarily manual. The ~~} ( j operator must reach a decision on when and how much water to pump. V The decision is based on the Operator's ability to observe a wide variety of stimuli and to reach a judgment based on those observa-tions. 'Ihis in an activity in which the Operator is superior to the 57 machine. Once the Operator takes the manual action, other actions in the system are carried out automatically, such as maintaining the hot well water level. The automatic operation is best suited to the machine. Feedwater Pumping and Reactor Level Control System - During power 59 operation this system operates automatically since its function is to monitor and perf orm the routine task of maintaining proper reactor water level. Reactor Recircuhtion System - This system can be operated semi-auto-maticall> or manually, and is another example of approximately equal capability between the operator and the machine to perform a task. Manual operation, when used, does not overload the operator. 57 Rod Control and Information System - The Operator manually initiates the action for operation of this system, based on a judgment of when it should be operated. This judgment is reached af ter considering a wide variety of information, a task in which the operator excels. Once control action is initiated by the operator, the system 59 functions automatically to ensure rods do not exceed established (n) V limits while being withdrawn. This automatic function is ideal for the machim. The portion of the PCIS which controls Control Rod t sequences and patterns during Startup, Shutdown and power operation namely the Rod Pattern Control System is not initiated by the Operator. This system is a hard wired scheme for which the Operator has limited bypassing ability f or a limited number of control rods. ' , 57 l Neutron Monitoring System - The Operator must insert and withdraw l 59 IRM's and SRM's during startup and shutdown. Also during these 37 phases of operation the Operator must change IRM ranges. Once the limits within which this system must operate are established by the 59 Operator, the system perf orms its monitoring and trip functions automatically. This is a monitoring function in which machines excel. Main turbine Control System - This system combines manual and automa-tic operation. The Operator manually initiates system operation; the system then operates automatically up to predetermined hold points' 57 to permit the Operator to monitor the system's performance and reach a judgment on whether automatic operation should be continued to the next hold point. This system thus combines the most desirable as-pects of operator and machine control. Steam Bypass and Pressure Regulator System - The Operator sets the limits appropriate f or the phase of operation, and r.he system 59 l operates automatically within those limits. m i u . 0-45 ACNCS-PSAR I.D.1 ) Cenerator Control System - Synchronizing of the unit could be operat- l x' ed approximately equally well either manually or automatically. Manual operation was chosen as preserving the greatest flexibility for integrating balance of plant controls for this system into the Control Complex. 57 Load control is automatically within the limits of the Reactor Recir-culation Control. There are other systems essential to safe operation which are not included in the primary operator interface with plant control, such as ECCS , the long-term residual heat removal system and most other l 59 safety systems. Since NRC requirements dictated that these systems $7 operate automatically, an allocation of functions was not perf ormed for these systems, except for those used f or surveillance testing and manual intervention. 59 The Combustible Gas Control, Spent Fuel Pool Cooling and Spent Fuel Pool Service Water Systems are the only safety systems which are manually actuated. These systems are simple to operate and their immediate operation is not required. In addition, the control rod drive hydraulic system is manually operated because it is such a simple system that automation is not justified. The systems on the B0P control benchboard (P870), including Turbine l [ T and Generator Lube Oil, Steam Supply and Drains, Circulating Water, ( j/ Condenser Of f-Gas and Condensate and Feedwater Auxiliary Systems, are long-response systems. They are generally manually initiated during start up then operate automatically.
- 3. Verification of Functional Allocation The verification of functional allocation is a detailed assessment and analysis of each allocation to ensure that the correct functional alloca- 57 tion has been made. The verification of functional allocation defines the design requirements and specifications f or the systems required by the Control Complex as well as the specifications for quantity of operators, f or the interface between operators and a system, f or the operational procedures (including emergency procedures), ar.d for maintenance require-ments.
3.1 Verification of Functions Allocated to Machines For each system function allocated to a machine, the perf ormance requirements of the system, or equipment to execute the function, ' have been defined. The perf ormance requirement considers such char-acteristics as response time, accuracy, reliability, and operator interf ace or display requirements. Points regarding tne design of Control Complex systems are: A t i \ ) %/ 0-46 Am. No. 59, (6/81) l i i ACNGS-PSAR f 1.D.1 i x/ .) 1. Display Systems The design requirements f or display systems consider established design criteria. Futhermore, the design requirements for display systems contain criteria to display signals that directly and accurately reflect the inf ormation to be transferred to the operator. These signals are to the extent practicable a direct measurement of the desired variable. Displayed parameters are selected from which the operator can determine if the systems are 57 performing their design functions or are responding to operator comma nd s.
- 2. Control Systems The design of control systems considers the design criteria presented in Appendix A of 10CFR50: General Design Criteria 20-39. Utilizing this analysis data and the design criteria previously described, prima ry and seconda ry Operator Interface panels were defined as:
- a. Prima ry Operator Interf ace i Control Console (P680) l 59
,_ 11 Standby Information Panel (P678) I b \s,,/ 111 Reactor Core Cooling Benchboard (P601) iv Stand by Diesel Panel (P877) 57 v BCP Control Benchboard (P870) vi BOP Auxiliary Control Benchboard (P800) vii Supervisory Monitoring Console (P679) 59
- b. Secondary Operator Interface l 57
, All other panels on which are located controls or displays 59 which must be overtly employed by the operator, as opposed to the maintainer. The design criteria were then applied to the Operator Interf ace panels. Function Placement 57
- 1. NUCLENET Control Console (P680)
- a. Normal (af ter prestart) plant operations functions t
O i \j 0-47 Am. No. 59, (6/81) ACNGS-PSAR I.D.1 N
- b. Short response functions f c. Frequently used and/or reactivity controls ,
- d. P.eactor Protections System Operator Interf ace t
Note 1: Only non-divisional systems related to a, b and c above, l except Nuclear Steam Supply Shutoff Manual Initiation at l
- System level.
L Note 2: Exclude functions not related to above. I
- 2. Standby Inf ormation Panel (P678)
- a. Support Inf ormation of the Display Control System 57 i
i b. No Process Control I
- 3. Reactor Core Cooling Benchboard (P601) 4 l
- a. NSS Safety systems i
i l ' b. NSS Long response f unctions - c. Standa rd design with no licensing impact
- d. Maintained divisional integrity i
I '1 4 3 i .i j t i ) i l r .0-48 Am. No. 59, (6/81) , I 1 ACNGS-PSAR 1.D.1
- 4. BOP Control' Benchboard (P870) l59 fN i )
)$~s# - a.- BOP Long Response functions 57
- b. ' Non-frequent use functions
- c. Maintained divisional integrity 5
- 5. Standby Diesel - Generator Panel (P877) l. 9
- a. Sefety related diesel generators 57
- b. Support systems f or a.
- c. Maintain divisional integrity .
- 6. BUP Auxiliary Control Panel (P800) l59
- a. All other safety systems
- b. BOP long response functions 57
- c. System level monitoring and control of a. and b.
1'
- d. Maintain divisional integrity ,
I I
- 7. Supervisory Monitoring r,onsole -(P679) (see page 0-50). l59 Table 1.D.1-1 illustrates the form for recording the application of these major criteria to each system, leading to a conclusion as to the assign- ;
ment cf functions to the Operator interf ace panels. Table I.D.1-2 shows the panel assignment conclusions for the BWR process systems which perform t the various functional objectives. <-~ With the Systems assignment to panels determined, the next step was to determine the order of placement of those Systems on the panels. Working on each panel individually, applying the design criteria, a logical order i of placement of Systems, upon that panel was deduced. 3.2 Verification of Functions Allocated to Humans 57 The most critical portion of the analysis is the verification of functions allocated to humans. Detailed analysis of functions ass-igned - to humans has detennined the suitability of the human-machine interface for the performance of the assigned function. Evaluation of the Operator's workload has determined if Operator overload condi- . tions. exist. The product resulting f, rom the analysis of functions allocated to humans should determine requirements for: "
- 1. Operator training
- 2. Operator procedures
- 3. Optimal Control Complex human-machine interf ace and control room configuration
- 4. Control Complex staffing.
. 's. 0-49 Am. No. 59, (6/81) ACNGS-PSAR I.D.1 /~~N Initial Work Station Layout
- l. Control Console (P680): l59 "Ihe most critical controls and displays, should be placed in the center of the Operator's work station."
In a Nuclear Power Plant the most critical controls and displays are those which are used to control and monitor the intended performance of the reactor core. In the BWR, these are the Rod Control & Infor- 57 mation System, the Reactor Protection System and the Neutron Monitor-ing System. These were, therefore, placed nearest the center of the Console. There must be water to act as moderator for the fission process and cool the core. In the BWR, steam is generated within the core and, af ter being scrubbed and dried, carried off to directly drive the Tu rbin e-Ge ne ra to r. l59 With the reactor core at the center of the Console, as the point of' reference, if water comes in, and steam goes out, there exists the 57 lef t to right expectancy of: water into the core; water and steam in the core; and, steam out of the core. 159 Therefore, the water system groups were placed on the lef t side of s- s the Console, and the steam sysi em groups on the right. _, The Reactor Water Recirculation System controly reactivity, as a function of flow. It was placed on the lef t side of the Console, . nearest the center. The Condensate Pumping and Feedwater Pumping and Level Control Systems indirectly control reactivity. They were
- placed next to the Recirculation System. The remaining water system, the Reactor Water Clean-Up System, which bears a functional relation-ship with another system was placed on the far lef t side of the
< Console. (This functional relationship will be explained during the discussion of the other system) . 57 The reactor's pressure control is performed by the Steam Bypass and Pressure Regulator System. Pressure directly affects reactivity; therefore this system was placed on the right side of the Console, . nearest the center. The Turbine Electrohydraulic Control (EHC) i System controls steam utilization by the Turbine. It was placed next to the Steam Bypass and Pressure Regulator System. The Generator is directly coupled to the Turbine, and was therefore placed next to the Turbine. ( . V) ( i o-50 Am. No. 59, (6/81) ACNGS-PSAR I.D.1 There are two more systems which were placed on the Console. One of which had been included in the previous analysis. The Perf ormance i _,1 Monitoring System is an operational aid which provides the capabil-ities of:
- a. NSS perf ormance calculations, Sequence of Events, Status alarm, and, Post-incident data recall
- b. BOP perf ermance calculations and logs
- c. Displays of NSS perf ormance calculations results
- d. Means of displaying operations information to supervisocy per-sonnel
- e. Means of generating new display f ormats, in the field, for both computer systems.
The Perf ormance Monitoring System's Operator Interf ace was placed on the f ar right side of the Console. The other system was the new Display Control System (DCS), so named because it was to be used to provide, per General Design Criteria 1.b.3, (Section 1.3.1) information displays which bring operations 57 data to the Operator. r~~ x There were ten color CRTs placed on the Console, one to be associated ( ) with each of the System groups, and one to be used primarily by the \- / Perf ormance Monitoring System, with switching capability to the DCS. The Operator's controls f or the DCS were located on the Console, in a manner to be described later.
- 2. Reactor Core Cooling Benchboard (P601):
Order of system placement on P601 was based on the sequence and f requency of operation, as well as the relationship of a particu-lar system to other systems. Of those system assigned to P601, there is one system which bears a functional relationship with the RWCU. It is the CRD Hydraulic System. During fueling of the reactor, there are times when it is neither desirable nor practicable to operate the Control Rods. Since the Control Rods are hydraulically operated via controlled leakage carbon seals, when the CRD Hydraulic System is operated, water inventory in the reactor vessel is increased, if not compensated for. One of the functions of the RWCU is to compensate for water level increases, during reactor startup, by providing a controlled drain. When the operator starts up the CRD Hydraulic System af ter an outage, he must g-1 v / 0-51 Am. No. 59, (6/81) ACNGS-PSAR I.D.1 x control reactor water level through the RWCU. This f unctional rela-l tionship establishes the need for the CRDilS and the RWCU to be in L / close proximity to each other, even though on two separate panels. IIence, the CRD Hydraulic System must be located on P601 at the end closest to the Console, and that end of P601 must be located in close 57 proximity to the lef t side of the Console. Panel arrangement and Key plan are both anchored by this relationship. , I In a nuclear plant, the integrity of the Nuclear Steam Supply is of l vital importance. Leakage from both controlled and uncontrolled I sources must be monitored to verif y the degree of that integrity. Ccntrolled leakage is collected in Equipment Drain Sump (s) befcre being pumped to the Clean (low conductivity) Radwaste. Uncontrolled 59 leakage is collected in Floor Drain Sumps before being pumped to the Dirty (high conductivit y) Radwaste. The frequency of monitoring and recording the leakage collected and pumped out to Radwaste dictates l5,, that the inf omation should be as close as possible to t..e Operator. This f unction is therefore located on P601 next to the CRDilS. l57 The next most frequently used functions are those of the !!ain Steam Systen: Safety / Relief Valves; Main Steam Line Isolation Valves; and, the Steam Line Drains. These functions are located next to the CRDilS a nd "'in Sumps. The Standby Liquid Control System has very few 9perator Iaterface devices, and, in point of fact, has never been deliberately operated to inject negative reactivity into the core. 57 ( 1 The SLC System controls and displays were located next to the Main ( ,/ Steam System. Core standby cooling is f unctionally allocated to: the Residual Heat Removal System ( RHR); the Low Pressure Core Spray System (LPCS); the Reactor Core Isolation Cooling System (RCIC); and, the High Pressure Core S pray. These systems were assigned to locations on P601 in that o rd e r. l 59
- 3. BOP Control Benchboard (P870):
Order of system placement on the P870 was also based on the sequence and frequency of operation, as well as the relationship of a particular system to other systems. 1 l During power generation, che in-house electrical loads are usually either totally or partially supplied from the Generator output. {57 Switching of the power source to these Auxiliary Electric Systems normally occurs immediately af ter the Turbine-Generator is synchron-ized to the Grid and loaded. There is a relationship therefore between the Generator and the Auxiliary Electric System. (3 \ ] v 0-52 Am. No.59, (6/81) ACNGS-PSAR I.D.1 I!ence, the Auxiliary Electric System must be located on P870 at the ) end closest to the Console, and that end of P870 must be located in m j close proximity to the right side of the Console. As before, panel arranhement and Key plan are both anchored by this relationship. 57 This logical order of placement of systems was continued to locate the remainder of the systems on P870 in the following order: Turbine , l Te st; Turbine-Generator Auxiliaries; Steam Systems, Condensatc/ Feedwater; Air Removal System; Of f Gas System; Circulating Eter 59 System; and Process Radiation Monitoring. BOP Auxiliary Contral Panel (P800) Systems lacated on Panel P800 are automatic with the exception of the control room emergency filtration inlet selection, as explaired in ' the control room habitability Section 6.4. The panel contains indi-cation to monitor if a system is perf orming its design function (cognitive task) and controls to start or stop a system (Reg. Guide 1.62) and where required select a mode of operation. The panel has a general layout from lef t to right 57 Fuel Pool Service 1*ater Fuel Pool Cooling Suppression Pool Cooling Upper Pool Dump to Suppression Pool HVAC System Level Controls in an arrangement similar to back row ['N s i HVAC panels P853, P864 Diesel Generator Panel P877 The Diesel Generator Panel contcins controls for the Division 1 and 2 l59 standby diesels and support systems such as fuel oil transfer. The - operation of the standby diesels is automatic as loss of power or LOCA. This panel contains displays to verify the system is operating according to design such as voltage and frequency. The panel contains controls to synchronize the diesels on to the Auxiliary Electrical Distribution System. The panel is located next to the HPCS diesel which is on one end of P601. This allows the operator to address the entire plant standby power system at one station. 57 Containment Isolation Panel P868, P869 The primary purpose of these panels is two f ald: 1) To supply the operator with a display by which one can determine that f or a given event the port ion of the Containment Isolation system required to opecate has indeed operated; 2) To supply a location of control for the safety r< .ated isolation valves of non-safety systems, such as ! i ,/ Am. No. 59, (6/81) o-53 U ACNGS-PSAR I.D.1 jm service water,1n a location readily accessible to the operator ( ) without violating the requirements of separation between safety AM . releted components and non-safety related components (Rog. Guide 1.7 5) . It can be seen f rom the above that the back row panels have a support function ta the front row panels and contain displays and controls 57 . f or equipe.ent addressed by the operator on a f ess f requent basis than front row controls and displays. Back row Panels Combustible Cao Con
- col Systems located on Panels P871 and .P872 are l" for control of hydrogen that may be generated as a result ot an 1 a ccident. The operator will be required to address this panel any-where f rom 30 minutes into an accident to several days into an acci-dent. This is a very long response compared with the other systems.
The systems that make up the Comesdhle Gas Control System are for control, monitoring and recording of the Hydrogen concentrations in the contsinment. Other Back row Panels The safety system back row panels contain controls f or individual components with the various BOP safety system These systems, with the exception of the Combustible Gas Control System, are automa- [ tically initiated and operated. The operator addresses the back row panel f or testing components or systems or in the case that the operator may wish to operate the particular system in an arrangement different from its normal alignment. The Accident Monitoring Panel and ESF Status Panel contain specifics of the general information displayed e n the f ront row panels, for example the back row 57 Post-Accident Monitoring Panel displays five localized Suppression Pool temperature and Bulk Suppassion Pool temperature and the front row Panel displays the same Bulk Suppression Pool tempe;ature. , Front Row Rack Row Interf ace The interf ace betwen the front row and back row panels is different for various modes of operation. a) During system setup previous to reactor criticality the operator will align service water and service steam systems. b) Once the reactor has reached criticality the operator has all the controls needed for normal start-up, operation and shutdown on che front row panel group. Testing of the various plant systems can be done from the back row panel during this mode of operation. \ v Am, No. 59, (6/81) 0-54 w ACNGS-PSAR ,- s c) In the event of an cccident as assessed by " Accident Analyses" ( ) Chapter 15 of the ACNGS PSAR or the sequence of failure events 57 N_/ s f or transients and accidents analyzed to develop upgraded emer- - gency procedures no operator initiated control is required f or at least the first 20 minutes. An operator need only monitor that l 59 the systems are perf orming their function. If an accident exists l 57 where hydrogen may be generated an operator will go to panels P871 l 59 and P872 to monitor and initiate systems to control combustible 1 57 gas control systems. Control of all other BOP safety systems at I the system level (Reg Guide 1.63) 1s on panel P800. An operator 1 59 may choose to' go to the Post Accident Monitoring Pancis P880, t 57 P885, P892, P896 to verify a display located on one of the front l , row panels. An operator may choose to manipulate controls for ; 1 safety related hVAC systems which are located on panels P847, l 59 P848, P863, and P864. Manipulation of component controls is not j 4 necessary for the system to perform its safety function. This is true even in the event of a single random failure. ! 3.2.1 Subfunction and Task Definition L For each function allocated to humans, all subfunctions and tasks including cognitive tasks that must be performed to achieve the function have been defined and arranged in se- l quence of performance. Manual tasks are specific with regard l to actions and information transfers from system to human required to complete the task. The plant procedures used by : _/ the control room Operator / Technicians have been reviewed to \ determine that they provide adequate guidance to perf orm the plant control functions according to the allocation of func-57 tions. l 3.2.2 Operator Task Analysis All requirements for Operator tasks have been analyzed to ensure that they do not exceed human capabilities. All time-critical functions allocated to the Operator have been analyzed to define the time requirements needed to success-fully perf orm each task. These analyses serve as the basis for specifying the size of the operating crew required, the human perf ormance charac-teristics required f or normal and emergency operations, the operational procedures required for abnormal and emergency operation, and the training requirements for Operators. , Based upon the data just derived, the anthropometric data of the intended user population, and the criteria previously stated, a full-scale mockup of the Primary Operator Interf ace panels was constructed. Sheet styrofoam was used to form the v 0-55 Am. No. 59, (6/81) , l f ACNGS-PSAR s panels. The front surfaces representinh the control and fN . display areas were covered with a material whose texture is ( ) compatible with the use of " velcro" fasteners. Systems analysis had detersined, in meeting the systems design objectives, which f unctions were allocated to the Operator, and which were allocated to the control system. The manner of implementation of those allocations was yet to be tested. , Tbn assumed control and display functional devices, selected for consistency with the design criteria, were photograph-ically reproduced. Small pieces of "velero" fastener material l were adhered to the backs of the devices, to permit their placement (and rearrangement) on the mockup. The system's Operator Interf ace Devices were placed on the ! Console and Benchboards in accordance with the design cri- 4 l teria, and in the same order in which they were selected for i location on the penei. The devices were rearranged many times, to provide as nearly as possible, the optimum Operator . o rientation. ! 3.2.3 Cri_tical Task Analysis Operational analyses was then perf ormed, by simulating opera- , tion of each system, using system Operating Procedures. The l System Operating Procedures used were those in effect in a 57 [ plant having nearly identical system (s) design. \s_-)/ Operational analysis was then perf ormed for integrated plant operation, using the plant procedu re s. As a result of these analyses, device location and arrangement were more nearly optimized. A task analysis was conducted for those tasks and modes of operation that are likely tojhave an adverse effect on plant safety if not accomplished la accordance with system require-ments. These tasks are identified as critical tasks. An ; analysis of critical tasks was done to identify:
- 1. information required by Operator / Technician, including cues tor task initiation !
t
- 2. Inf ormation availabic to Operator / Technician
- 3. evaluation process
- 4. decision reached af ter evaluation t
+ \, .
- 1 1 0-56 Am. No. 59, (6/81) t
- . . _ - , m , _ , ACNGS-PSAR
- 5. action taken
[ ) \ '~' /
- 6. ir Jy movements required by action taken
- 7. werkspace envelope required by action taken 8 workspace available 9 location and condition of work environment 57
- 10. frequency and tolerances to action
- 11. time base and time margins (time margins must be adequate to cover variances in human responses) 12 . teedback inf orming operator / Technician of the adequacy of the actions taken
- 13. tools and equipment required
- 14. number of personnel required, their specialty, and ex- l 59 pertence 57
- 15. job aids or references required 16 communication required, including types of communication ] 59 fm i I
1 C ') 17. special hazards involved
- 18. Operator interaction where more than one operator is involved 57
- 19. operational limits of personnel (performance) l
- 20. operatio. sal S imits of machines and systems The cri tical task a talysis also included analyzed accident l 59 conditions During the operational analyses, careful notation was made of the Operator's isdorn.ation needs for cach phase of system 57 opera tion. This data would be used to select input variables to the Display Control System (DCS). and, to help assign the variables to the various system foraats. The immediate use of l the data, however, was as a basis for assignment of hardwired, ,
backup information devices to the Standby Information Panel. i A l I Am. No. 59, (5/S1) 0-57 ACNGS-PSAR 57
- 1. Standby Inf ormation Pane 7 "P678) 7_x 59 k, Until the calculated DCS reliability (1.995) can be verified operationally, it is necessary to provice sufficient hard-wired information displays (as well as a DCS Configuration /
Statuu Display) to allow continued steady state power ope rat ion s, reasonabic power maneuvers in the Run Mode, or a safe shutdown, without reliance on the DCS. The Standby Inf ormation Panel serves no other purpose. The re a re no process controls or annunciators on the panel. There are no displays which were not determined to be necessary, as a result of the operational analyses. The Star.dby Information Panel stands behind the Control Con-sole. Initially, it was intended to be in the direct view of a standing operator. It was later determined that the front silhouette of the Control Console could provide a visual path for the seated operator. The standby information displays for each system controlled from the Control Console were located, accordingly, on the panel. The Standby Inf ormation Panel is located four feet behind the Control Console to allow clearance for CRT removal from, and replacement in, the Control Console, but still maintain the inf ormation displays within the visual range of a licensed 57 Operator. Oi 2. Supervisory Monitoring Console (P679) ;b The Supervisory Monitering Console allows supervisory personnel access to the same data available to ths Operator, without creating a disturbance f or the Operator, by looking over his shoulder. The DCS and the PMS have communications links, theref ore, all data in the DCS is 1 cvailable to the PMS. Supervisory personnel wishing to access DCS data may do so on two color CRTs. communicating via a free-standing, multi-function keyboard which is identical, in all but physical appearance, to the keyboard supplied the Operator, for PMS commenication. The Supervisory Monitoring Console is centrally located between, but at the opposite end of, the Benchboards from the Control Console. This provides supervisory personnel with a dependent visual access to all of the Primary , Operator Interf ace. i C\ b 0-58 Am. No. 59, (6/81) r ACNCS-PSAR
- 3. Display Control System (DCS) p --
( ) The total design for the DCS required approximately 35 x- / man years of ef fort. Some sof tware enhancements con tinued for almost 7 years af ter the design initiation. Display foonat research and development extended over a period of more than 3 years. As a result of studies performed by General Electric, the DCS formats employ the follow 1.ig color coding:
- a. Green - Used or. , for lines and symbols in process diagrams to represent static system com-ponents, i.e. , pumps, motors, valves, and, piping which are not dynamically presented in the given format. Selected f or this association because the display elements make up the langer part of the display, and, a green hue has been demonstrated to be the least visually fatiguing of the available hues.
- b. Cyan - Uaed as a supporting hue and applied to alphanumeric identification, scales, and bo'rder s. 57
- c. Yellow - Applied to all dynamic process variable
[s} display elements, such as bar graphs and (,_ ,/ digital data. Selected f or this appli-cation because of the intensity of its hue. Yellow allows the Operator to scan the display and easily identify dynamic i nf o rmatio n.
- d. Red - Restricted to use as a visual cue for abnormal conditions. Should any variable exceed process limits, the data (bar graph and/or digital) norma 11.v displayed in '
yellow, changes to red. Selected because of the traditional, pre-established.psy-chological associations (populational stereotype) with such conditions, and because intensity allows minimal visual sea rch.
- c. White - Used as a reference mark on scales, ad-jacent to bar graphs, to indicate process limits, or, to present low confidence data -
/' s . \s_ ' 0-59 Am. No. 59, (6/81) ACNGS-PSAR
- f. Magenta - May be used in place of red.
7 g. V) l Dark Blue - Shall not be used, due to its visual loss against the normal background color.
- h. Black - Nomal background color.
Initial f ormat definition began f rom the data gathered during the operational analyses. A family of 63 formats was gene-rated for the process System Groups, depicting various levels - of each system's operation. Further analysis was perf ormed to determine the relationship of these formats to reactor opera-tiono phases. The DL design was a continuing process, as stated above. At 57 this point, however, the Operator's controls for retrieval of operational information via the DN could be defined and located. Each of the ten CRTs, on the Control Console, would have two multi position selector switches. One switch would serve for Syatem selection and one for Fomat selection, thus providing capability of displaying any System Format on any CKr. Two momentary push-buttons would provide a Menu Display and Format Change Enable. It is not necessary that the com-puter system, which drives the displays, attempt to follow Format Selection until the operator has placed the Format Select Switch J n the position of the Format desired. The Operator inf orms the ? <.mputer that the System and Format Q selected are those desired for viewing by depressing the Format Change Ena' o le switch. This group of f our switches is I59 mounted next to each of the CRTs which they control, including the CRT which is normally assigned to the PMS. Included, for the PMS CRT is a fif th switch (momentary push-button) for assignment of that CRT to the DCS, when necessary. One of the positions of the Format Select Switch is designated "Ma st er" . When any, or all, of the Format Select Switches are in this position, the Operator has simultaneous control of those CRTs from a " Master Display Select Matrix" located at his lef t hand, when seated at the center of the Control Con-sole. The informational needa data, derived from the opera-tional analyses, showed what Information the Operator needed 57 to either overtly employ, or have available to him, during which phase of plant operation. The Master Display Select Matrix is used, by the operator, to inform the computer which phase of reactor operation he is performing. The computer then displays those Syctem Formats determined to be most meaningful to that phase of operation. Thus the Operator is only required to perform a single action to have appropriate dr.ta re trieved , and displayed to him. O V 0-60 Am. No. 59, (6/81) ( ' ACNCS-PSAR ', s y 3.2.4 Work Station Design Analysis l57 s_s For. each work station in the control complex, the time 159 sequence of operator activities and the time required for 4 inf onsation exchange or transfer to the operator has been defined. The analysis verified that the Operator is capable , L of completing all tasks and that all tasks are capable of being perf ormed using the work station design. 3.2.5 Operational Sequence Analysis An analysis and evaluation of Control Complex sequences of operations, flow of decisions, physical .transmissiona of data and inf ormation, receipts of information, storage of informa-57 tion, monitoring of systems and interactions among operational crew members, work stations, and systems has been conducted. The purpose of the analysis was a validation of the Control Complex capability to successfully complete the intended functions of the design, in both the time and space domain. 3.2.6 Workload Analysis A workload analysis for all critical functions was conducted to appraise the extent of the Control Complex operator work- l 59
- loads. The analysis was cased on the sequential accumulation of task times. Application of this technique permits an 57 evaluation of 'the capability of the control complex operator (s) l 59 to perform all assigned tasks in the time required to maintain plant safety. -
- The detailed workload analysis divided the Operator's tasks
- into categories corresponding to perceptual motor channels such as vision, lef t hand, right hand, f eet, cognition, audi-
. tory, and voice channels. The purpose of thia level of detail ' 57 was to ensure that the Operator is not required to perform more than one task at a time if two or more tasks require the simultaneous use of a single perceptual-motor channel nearly 75 percent of the time. 3.2.7 Human-Error Analysis A human-error analysis was conducted for each perceptual-j motor channel workload of 75 percent or greater as defined iy j the results of the workload analysis. i l The purpose of the human-error analysis was to investigate the probability of error during high workload conditions and to j evaluate the consequences resulting from these errors. J 4 0-61 Am. No. 59, (6/81) } y , v y -w,m -g ---- -- --e ,-~--e,-n-- r,,:,.,r,r,ye -- ., .,_nw.r--, - , - , - *w;- vw - w- _ -- . . . . .--- . ._.- _ _ -._= - ACNGS-PSAR 3.2.8 Work Station Link Analysis b A work station link analysis has been conducted for each work ( station used by the Operator to perform critical tasks. The analysis defined the frequency and criticality associated with ! each of the interactions occurring between Operator and equip-ment and/or between one Operator and another. The defined p f requency and criticality of the interactions are then used to 57 evaluate the design adequacy of the work station layout in i terms of time and space utilization. This analysis achieves a near optimal design f or the work station, such as the spatial ! correlation of displays with controls to provide the Operator with feedback information as required by General Design criterion 13, Instrumentation and Control. I l i I i e 1 f 1 i I I 0-62 Am. No. 59, (6/81) f -- . - , . .,+-r -.g.,e-- ,w,- , , --m - , . ,e- ,,,w .[ ; t r { v [I t.L ,' i ; , , ,'n*, 7 9 7 , 5 $ 5 .l % / R -r E H T O x x x 0 0 8 P 0 x x x x x x x x x L 7 8 I - M P - r . A F N T 1 ) 0 x x x x x x x x x EGN NE 6 P ( MID* RC AY 0 x x x x mC S 8 x x x x x x x x 6 e wERAS P - m + l w a w n ) - o t r m . lt i ra x x x x x x x x x x x x x x x x x x x x x - aa et y tr t s re op f e ~ Ar . e NO (P - y v y d e y e g t e t a x x x x x x x x x x x y l l e se m M w ~ l o x x x x x x x x x x x x x x x x x x x x x x x a N y n w A o - i I - R s - i 4 E v s R T I i e x x z x x x x x x x x A R D Y S C - vP - S. S V t "N en ) C M ra s x x x x x x x x , A E u ( T R f:oloo , - S Y C S e e - l v q o i . r t y s x x x x x _ t ct - n ai o e - C R + - t n e x x x x x x x x x x x w u e x x x x x x x q s e U g r F - - m e. e g n x x x x x x x x x x x x x x x x x x x x g o . s l ~ n o p s + e t R r x x x x x x x x x x x o w h e S e t t - l . t s p s n s i o a . c & l t n y u f n r e y n - o i . t ! o i s o S S e i e t S o m s F s r rr s ! l n r n yC i , a l t a M e i e et y u oo t o S . r l m e g sr a r a y e e t D l p t l s S a ri n M . n e e a t n rD rv - r e W S t t n s n e - ri i n r t t o . s o w R e a i e eo o t a s o y co u a or oI . d na C n t s M o t sg p t & sm et a e d s s yi S ei F e y om o o e m a ne na W ic 0 a n S jC m B e . r H C r v r r r . P / S C s . S nn ei s U n cp P Ht u es eR i r b e e v C G e t g an m a
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A e e g g e g m P o A R V l i r P. h jC a e r c H e e o o or \.)< M t p t S S S eS tC s sC r r re s p n T C C p iA S e eA d d d n y u o C C C po wV H i iV y y yi ,. d S C S E E Ut SH U DDH H H Hb [ti o[ N. . . ke i ACNGS-PSAR ,s TABLE I.D.1-2 \ l V PANEL ASSIGNMENT (I)NCLUS10NS Panel System NUCLENET* Control Nuc. Boiler Process Instr. Console Recire Rod Control & Inf ormation FW Ievel Control bkutron Monitoring Rx Protection System Rx Pressure Control Performance Monitoring System 57 Rx Water Cleanup (RWCU) Cbndensate Pumping FW Pumping Turbine - Generator Rx Core Cooling CRD Hydraulic System . Benchboa rd Standby Liquid Control System (SLC) Residual Ikat Removal (RHR) Iow Pressure Core Spray (LPG) High Pressure Core Spray & Power Supply (HP CS) Rx Core Isolation Cooling (RCIC) /'~'} ( j Pressure Relief valves 4 Main Steam System BOP Control Benchboard FW Heaters, Vents & Drains Condenser, Air Removal l59 l Of f Gas System Tbrbine A1xiliaries & Test Cire Water Condensate System Feedwater System Radiation Monitoring Stean Supply ; Drain System Aurtliary Electric Distribution BCP Aux Control Spent Fuel Pool Cooling Denchboard Suppression Pool Cooling 57 Control Room HVAC Standby Gas Treatment System ECCS Area Filtered Exhaust ECCS Area Fan Coolers Upper Pool Dump to Suppression Pool All other Safety-Related HVAC Systems l Combu'stible Gas Drywell Hydrogen Mixing 7s , Control Panel Hydrogen Recombiners Hydrogen Monitoring System ( v) 0-65 Am. No 59, (6/81) ACNGS-PSAR . D .1 ; TII CONTROL ROOM DETAILED HUMAN FACTORS REVIEW gq BACKGROUND he Allens Creek Control . Room Evaluation Task Team performed a preliminary assess- , ment 'of the A11 ens' Creek Nuclear Generating Station. Unit 1 control room for the pur-pose of identifying additional human factors engineering design conditiona that could [ . provide a basis for further improvements.- < ' i Houston Lighting & Power Company's instrumentation and controls engineers assigned ' to Allens Creek built a full size mockup of the front row psneis of the Nuclenet , i I i 1000 Control Complex. A human factors engineering evaluation was performed on the l mockup, .We control room was evaluated as is, without regard to planned changes in layout f and changes required as a result of %ree Mile Island. [ %e evaluation _ consisted of the application of human factors engineering design [ i checklists. %e checklists were developed and prepared by the General Electric l l Boiling Water Reactor (BWR) Control. Room Owners Group. : As defined by the enecklists, the purpose of the control room survey is te review i i and assess the adequacy of the arrangement and identification of importanc controls [ ! and displays, the usefulness of audio and visual alarm systems, plant status infor- i mation provided procedures and training with respect to limitations of existing ' 4 instrumentation, information recording and recall capability, the control room lay-out and environment, and other areas of human factors engineering that potentially j impact operator effectiveness, he ultimate objective is to identify essential ' modifications at the operator-control room interface to minimize the potential for human error. ! \ %e control room survey methodology as currently recommended by the BWR Control j' j Room Owners Group consists of four phases. The four phases are: (1) control room I review utilizing checklists which compares the engineering aspects of the control room with established human factors engineering criteria, (2) operator interviews, (3)- LER analyses, and (4) emergency procedures walk-through. i Phase I of the control room survey is conducted by the survey team using checklists ' which are titled A) Panel Layout and Design, B) Instrumentation and Hardware, C) Annunciators, D) Computers, E) Procedures, F) Control Rc>n Environment, - , G) Maintenance aid Surveillauce, and H) Training and Manning. 59 , his essentially agrees with the ten major topics which will be included in Draft - NUREC- 0700. The esjority of the topics in Draft NUREG-0700 are found in BWR Owners [ ! Checklist A and B as there are distinct subsections within the BWR Owners Group check- [ lists for the Draft NUREG-0700 copics. , t BWR Owners Group 57 NUREG
- 1) Control Room Werkspace Panel Layout and Design l Control Room Environment
- 2) krkplace Envi_*onment '
- 3) Annunciators sud Auditory Signak Annunciators
- 4) Controls Instrumentation and Hardware
- 5) Visual Displays Instrumentation and Hardware 3
- 6) Panel Layout Panel Layout and Derign .
- 7) Control / Display Integration Panel Layout and Design , ,
' Instrumentation and Hardware i V)8)9)- Proc Labels'and Location Aids ss Computers Panel Layout and Design Computers _
- 10) Data Recording and Retrieval Instrumentation and Hardware
} 0-66 Am. No. 59, (ti/81)' _{ v ~ 7 , ,ACNGS-PSAR 1.D.1 Houston' Lighting & Power Company's survey team has. performed the control room , SM g review phase. The checklists that-were considered applicable to the Allens Creek ,l JcockupwereA): Panel. Layout and Design, B) . Instrumentation and Hardware, and ,Lbs_,eC)) Annunciators.. The other ' phases of the survey (2-4) cannot be performed on the Allens Creek cont.rol room at this time becausa they involve operating history end emergency procedure availability. ~ Each checklist' item is presented in the form of a question for _ consideration by each survey team member. As each specific quescion is evaluated, the team member actually doing the evaluation of that question indicates the relative. degree l_ of ' compliance (no compliance, somewhat compliant, mostly compliant, full compliance, ' not applicabic). Following each checklist item is space for the person performing the evaluation.to enter comments. For each specific checklist item, these comments will identify items.or components of non-compliance, the scope of review, or any qualifying statement judged to be appropriate to the evaluation. 1 Evaluation Approach y The evaluation of the aliens Creek control room mockup was conducted in three phases, describcd aa follows:
- Phase 1 - Documentation Collection This phase include 3 collection and review of control room panel drawings, control and display design conventions, human factors engineering reference material (MIL STD 14723, NUREG CR-1580) board profiles and dimensions, piping diagrams, control panel instrument lists, control wiring diagrams, '
and flow diagrams. e Phase 2 - Data Collection 57 l [\s_p During this phase, checklists containing human factors criteria were applied. These checklists addressed: . - control room layout ! - panel layout and design + - annunciators j - controls - instrumentation j - displays and mimics f l.
- Phase 3 - Analysis and Reporting The final phase of the evaluation consisted of (a) review of the data col-Iceted, (b) identification of the items needing human factors enginaering enhancement, (c) prioritization of the items needing human factors engineer-i 1
i L ing enhancement, and (d) reporting of the results. The items identified i as needing human factors engineering enhancement were consolidated into ! the following groupe:
- . annunciators
- labels - panel layout + - displays > - instrumentation %, / 0-67 Am. No. 59, (6/81) , .. - ,_ - , _ . . . _ _ __ - , _ - . . - _ _ _ _ , _ _ _ _ - . - _ , _ _ _ . ~ . . . _ _ _ _ . ACNGS-PSAR ,- The items needing human factors engineering enhancement were prioritized ; according to the following: x- - Category I - corrective action is recommended to minimize the potential for error - Category 17 - corrective action should be censidered to reduce the potential for error - Category III - no corrective action is necessary Results -- Identification of Human Factors Engineering Discrepancies The results of the control room survey will be objectively evaluated by Houston Lighting & Power Company to determine what changes should be made for minimizing facters engineering con-the potential for operator error as a result of human siderations. The schedule for this evaluation is such that the changes deemed , necessary wi11 be incorporated prior to fabrication of the control room. The following is a summary of the general human engineering discrepancies that are applicable in some degree to the panels reviewed by the eveluation task team. i 57
- Labels
- 1) Labels are not consistent in nomenclature and abbreviations.
- 2) Labels are not size coded in a hierarchical system for components, major systems, and associated subsystems.
- 3) Labels a:e not consistently positioned.
s
- Panel Layout l s_
1 7
- 1) There is a lack of consistency in color coding.
- 2) The use of lines of demarcation needs to be expanded.
- 3) Mimics need to be added to enhance the system flow path.
- 4) Lines of demarcation are needed to separate primary and secondary j systems. i
- 5) Limits on enthropometric design are exceeded.
- 6) Layouts are not consistent in operational sequence.
- Displays
- 1) Indicators are not scaled in process units that relate to system operation. i
- 2) There is no normal range or setpoint markings on indicators. There are no markings to show safe or unsafe ranges and expected or unex-nected range of operation.
l
- Instrumentation
- 1) There is a lack cf consistency in switch coding for pumps, fans, dampers, valves, and breakers.
- 2) All switch positions are not clearly marked.
- 3) Switches for emergency or abnormal use are nat clearly identifiable.
- 4) Keylock switches require use of keys for normal operation. The key is the switch handle.
- - 5) Backup indication is nit easily correlated with indicators. 0-68 Am. No. 59, (6/81) [ ,--g ACNGS-PSAR 'f 4 \/ U e Annunciators
- 1) Abbreviations used on annunciator windows are inconsistent with the abbreviations used on instrument nameplates.
- 2) Alarm prioritization does not exist for balance of plant systems.
Results -- Enhancements to Resolve Human Factors Engineering Discrepancies Based on the preliminary assessment of the Allens Creek control room, the following enhancements will be made to resolve the human factors engineering discrepancies:
- 1) A standard list of abbreviations is being developed to include instrumenta-tion on the panels, annunciators, and the computer system.
- 2) Hierarchical labeling will be implemented.
- 3) Specific guidelines are being developed for labels and will be implemented.
This will include label consistency, accuracy, and board placement rela-tive to labeled components.
- 4) Lines of demarcation will be added to functional control and display groups.
- 5) Mimics and lines of demarcation will be added to improve flow paths.
- 6) Panel laycut will be improved in terms of control / display relationships, operational sequence, functional relationships, and anthropometric standards.
- 7) All switch positiens will be identified, and switch coding will be consistent throughout the control room.
- 8) Instruments required for emergency and abnormal conditions will be identi-fied and clearly marked.
,$2)- 9) A color coding standard is being developed for indicating lights, mimics, and computer system displays. 57
- 10) The annunciator system will be improved in terms of prioritization using color coded windowa and proximity to associated controls and displays.
- 11) Indicators will be scaled in process units relating to system operation, and ranges of operation will be provided.
Control Roon Evaluation Task Team i The Allens Creek Control Room Evaluation Task Team consisted of two licensed oper-ators, five instrumentation and controls engineers, and a human factors engineering consultant. The team was broken taco four groups for the purpose of the survey. A minimum of two groups evaluated each of the six front row panels. Although each person filled out his own checklist, the two people on each group were not required to agree on the assessment of the panel. i l l _ (~'% k. 0-69 Am. No. 59, (6/81) . _ _ _ . _. . . _. .. . _ . . .. . . . ___. _ . _ _ . _ _._. ~ _ _ _ . _ _.___ _ _ _ _ . _ [ [ i ACNGS-PSAR Conclusion , l The Allens Creek control room uses advanced technology which incorporates 57 ' l human fcctors engineering principles in design and operation. A control - room design review has been conducted based on a full-scale mockup, ' i deficiencies identified, and enhancement activities are being undertaken.
- Completion of the enhancements will result in a control room which meets
. the intent of NUREG/CR-1580 and NUREG-0659. Results of these actions will . be forwarded to NRC upon completion. i a-i f i 4 N l 4 'f . l t I t' I i ? e 4 o-70 - Am. No. 59, (6/81) l t 4 ~ .-- , - , - - ,,,-m ..,.-.mw.,y,.r---,--.,,wn.- n ., ,,r- , mw,,,- ,,nw_--n,-wvnnr-,,mwnw w..,,w m+,e,w,w me m , m ,w,,,mrw-ww_ i i ACNCS-PSAR I.D.2 ,4./~'} PLANT SAFETY- PARAETER DISPLAY CDNSG.E \] : ITEM NO. I.D.2 NUREG 0718 ' REQUIREME NT ~ ~ Applicants shall describe how they intend to meet the staff criteria. con-tained in NUREG-0696 for the plant safety parameter display console. Ap- ' plicants. shall, to the extent. possible, provide prelimir.ary design inf ormation at a level consistent with that normally required at the construction permit s tage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. > ' Applicants shall also demonstrate that the design concept is technically f easible, and wid11n the state-o f-the-art, and that there exists reasonable - assurance that the requirements will be implemented properly prior to the 57 issuance of operating licenses.- . RES PO NSE j.
- A Safety Parameter Display System (SPDS) will be provided, and is described in i - Sec tion' 7.5.1.6. The system will have the capability of displaying the full range of important plant parameters and data trends on demand. The system 59 will also indicate when plant parameters are approaching or exceeding process limits. The SPDS will be designed in conformance with the guidance of j NUREG-0696, February 19 81, Section 5.
fk The' design concept of the SPDS is known to be technically feasible and within the state-of-the-art. HL&P has no questions or concerns as to the ability to l 57 implement die SPDS design prior to OL issuance. The SPDs will be a ' computer-based system of high quality'and reliability. It will be capable of func- 59 y tioning properly in the environments that are present during transient and' accident conditions. l 1 1 h t t t V 0 Am. No. 59, (6/81) _ ACNGS-PSAR s ~'S ( ) I.D.3 u/ ITEM I.D.3 SAFETY SYSTEM STATUS MONITORING NUREG 0718 REQUIREMENT " Applicants shall describe how their design conf onns to Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Sy s t ems. " Applicants shall, to the extent possible, provide preliminary design inf ormation at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, appli- 57 cants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state of-the-art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses."
RESPONSE
Regulatory Guide 1.47 was a pre-TFE design basis for ACNGS, and the design meets the guide without any exceptions. The design includes automatic indication of the bypassed and operable status of safety systems. To the 59 extent practical, inputs to the Safety System Status Monitoring system will be direct measurements of the desired variables. Details are given in Section 7g 7.2.2.2.2.2, and systems covered by the guide are shown on Table 7.1-2. l57 ( ) w/ f~N k l v 0-72 Am. Na. 59,(6/81)
~ I
ACNGS-PS AR
II.B.1 ITEM II.B.1 REACTOR COOLANT SYSTEM VENTS NUREG 0718 REQUIREMZNT
" Applicants shall modify their plant designs as necessary to provide high point reactor coolant system and reactor vessel head vents that can be remote-ly operated from the control room. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normal-ly required at the construction permit stage of review. Where new designs are [
involved, applicants shall providc a general discussion of their approach to j57 meetirs the requirements by specifying the design concept selecced and the i supporting design bases and criteria. Applicants shall also demonstrate that
~
the design concept is technically feasible and within the state-of-the-art , and that there exists reasonable' assurance that the requirements will be implemented properly prior to the issuance of operating licenses." j l
RESPONSE
i Ve nting capability of the ACNGS reactor vessel is addressed in two parts (refer to Figure 5.1-3a):
- 1. Up to the main steam line nozzles: The presence of non-condensible gases !
in the vessel below the main steam line nozzles could interfere with ! continued core coolirg, so the capability for ventirg this region is l essential. This can be accomplished by opening any one of the 19 safety t fS relief valves on the main steam lines (which may be open already depend-ing on the mode of core shutdown cooling in use). These valves and their ( )
operators are safety grade, seismically and environmentally qualified for accident conditions, have positive position indication in the control i room (see Item II.D.3), and are powered from the onsite electrical system 159 and operable from the control room. Eight of the valves have a safety L 7
related air supply, providing redundant venting capability. P In addition, this region of the vessel can be vented through the RCIC steam supply line which connects to main steam line A, without opening 59 the SRV's. This path is through the RCIC steam turbine exhaust , which discharges to the suppre ssion pool.
- 2. Above the main steam nozzles: The presence of non-condensible gases in the vessci above the main steam line nozzles will not interf ere with continued core cooling, and as such venting this region of the vessel is 57 not considered to be a safety concern. Even so, there are two means of venting this space:
- a. Normally open 2" reactor head vent line and valve B21-F005, which 59 discharges to main steam line A (which can be vented to the suppre ssion pool / containment via any one of three safety relief 57 valvo s) .
- b. Narmally closed 2" reactor head vent line and series valves B21-F001 159 and B23-F002, which discharges to the drywell high purity drain tank. j
( '- ) These valves are safety grade and their operators are Class lE, 57 seismically and environmentally qualified, but are not powered from the onsite electrical system. They are operable from the Main Control Room. 0-73 Am. N,. 59, (6/81)
4 ACNGS-PSAR II.B.1
,q.
t \ Consideration has also been to the potential for the accumuistion of 5 / ' non condensible gases interf ering with the operation of the ECCS. In the post-LOCA condition, it is possible for non-condensible gases to come out of solution while operating the WHR system. It is expected that these gases would be swept through the RHR system, but some gases could 59 i potentially accumulate in the upper portions of the RHR heat exchanger during the steam condensing mode of RCIC operation should substantial amount of non-condensibles be generated. The upper portion of the RHR heat exchangers are provided with separate 1" vent lines to the suppre ssion pool f or the removal of non-condensible gases. The isolation valves on these lines are Class lE, and are operable from the Main Control Room. Procedures f or the use of thase lines will be summarized in the FSAR. All of the above venting paths lead to the containment via the suppression pool, which is the basis for hydrogen mixing analyses. The control of large i amounts of hydrogen in containment is discussed in the response to Item II. B.8.3. 57 The above supplements the PSAR inf onnation on capability for RCS venting, and is consistent with preliminary design inf ormation normally required at the CP , stage of review. There is no new, novel design, and there are no concerns regardirs technical feasibility, state-of-the-art or ability to implement the intended RCS venting design. n's/ } t 4 s
\w/ )
0-74 Am. No. 59, (6/81)
ACNCS-PSAR II.B.2
N i
s _- ITEM II.B.2 PLANT SHIELDING TO PROVIDE ACCESS TO VITAL AREAS AND PRUIECT SAFETY D)UL H4ENT FOR P'O SI-ACCIDENT OPERATION NUREG 0718 REQUIREENT
" Applicants shall (1) perform radiation and shielding design reviews of spaces 57 around systems that may contain highly radioactive fluids and (2) implement plant design modifications necessary to permit adequate access to vital areas and protect safety equipment. Applicants shall, to the extent possible, provide preliminary design inf onnation at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a general discussion of their approach to meeting the requirements by specifying the design concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state-of-the-art , and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses."
RESPONSE
A post-accident radiation shielding review is being conducted for ACNGS, and is scheduled for completion in December,1981. Its purpose is to ensure that:
,- s 1. access and required occupancy is possible to areas where plant personnel
/ j must perform post-accident functions, such as the control room, onsite 59 _,' technical support center, sampling station, sample radiochemical analysis laboratory, etc. , such that the GDC-19 specified dose design basis will not be exceeded.
- 2. radiation levels f or which saf ety-related equipment is qualified to function post-accident are not exceeded. 57 The basic assumptions of the study are:
- 1. 100% of the core inventory of noble gases and 25% of the halogens are dispersed into the drywell and containment air volumes. 50% of the halogens and 1% of all other ff c.sion products (except noble gases) are dicyersed into the reactor and suppression pool water. The core 59 inventory of isotopes available for relesse is given in Table II.B.2-1.
Instantaneous release and mixing are assumed. I i
- 2. The following systems are instantaneously filled with contaminated fluid per 1 above: 57
- a. Residual Heat Removal (Suppression Pool Cooling, Containment Spray and Low Pressure Coolant Injection).
- b. High Pressure Cote Spray
, _ _ c. Low Pressure Core Spray
! d. Reactor Core Isolation Cooling x I 0-75 Am. No. 59, (6/81)
' ACNGS-PSAR f
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- e. MSIV Leakage Control '
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- f. _ Standby Gas Treatment
.g. Hydrogen Analyzer 57
- h. Containment Gas Sampling
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- 1. Post- Accident Liquid Sampling
- j. - ECCS Filtered Exhaust
- k. 'Drywell and Containment 4
Systems listed in NUREG-0737, Item II.B.2, Section (2) which are not considered as sources terms are listed as follows with justification: 59. l-
- a. Hydrogen Recombiner System: ACNGS utilizes thermal recombiners which 1 are completely internal to the containment .
- b. Gaseous and Liquid Radwaste Systems: The radwaste systems are isolated 'from the containment and other systems which may contain primary coolant af ter an accident.
-c. Chemical and Volume Control System: This is a PWR system.
4 [\ \ TIIe study will verify the adequacy of the existing design and indicate where changes will need to be made. If changes are required to meet acceptable operator and/or equipment dose levels in certain locations, the following options are available: 57
- 1. move the offending radiation source to a less sensitive location
- 2. . move the targer equipment or operator control / work station to a location with an acceptable radiation field.
i
- 3. place additional shielding around the offending radiation source.
- 4. place local r51elding around the targec equipment or operator control / work station.
- 5. purchase equipment designed to withstand the newly specified radiation
! environme nt . In selecting the option to be used emp. asis will be placed on minimizing
, building structural modifications, since the buildings potentially affected are mostly designed and are early in the construction sequence.
If problems are encountered as a result of the shielding analysis, they are
~
expected to be of a physical or design detail nature rather than questions of te chnical feasibility o state-of-the-art. Since the shielding study will be finished in 12/81 and e CP for ACNGS is not expected before the n, HL5P is [\,_,/\ assured that any necessary changes can be er'fectively implement d prf or t o construction, and that the specifications of NUREG 0737, Section II.B.2 will be implemented prior to OL issuance. 59 0-76 Am. No. 59, (6/81)
4 ACNGS-PSAR TABLE II.B.2-1 V CORE INVENTORY OF ISOTOPES FOR ACNGb UNIT #1 AT SHUTDOWN Inventory Isotope in Ci I 131 1.05 x 10 88 I 132 1.49 x 10 I-133 1.72 x 10 I 134 2.22 x 10 8 I 135 1,76 x 10 Br 82 1.39 x 10 5 Br 83 1.07 x 10 Br 84 1.68 x 10 7 7 Br 85 2.33 x 10 Xe 133* 6.99 x 10 68 Xe 133 2.02 x 10 Xe 135* 6.08 x 10 59 Xe 135 3.50 x 10 87 Xe 137 1.68 x 10 Xe 138 1.59 x 10 8 Xe 139 1.41 x 10 8 Xe 140 6.93 x 10
\ Xe 141 2.11 x 10 Kr 83* 1.08 x 10 77 Kr 85* 2.32 x 10 6 Kr 85 1.07 x 10 7 Kr 87 4.19 x 10 Kr 88 6.05 x 10 3 7
Kr 89 7.13 x 10 Kr 90 7.31 x 10 7 Kr 91 4.71 x 10 6 Cs 134 6.96 x 10 6 Cs 136 3.21 x 10
- Cs 137 1.28 x 10 87 Cs 138 1.85 x 10 7
Rb 88 6.11 x 10 Rb 89 7.82 x 10 7 Sr 89 8.05 x 10 6 Sr 90 8.82 x 10 8 Sr 91 1.03 x 10 0 Sr 92 1.11 x 10 n Y 9J 9.27 x 10 6 1 Y 91 1.10 x 10 8 , , ('v) i i 0-77 Am. No. 59, (6/81)
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ACNGS-PSAR
- I.B.2 TABLE II.b.2-1 (Cont'd) ;
Inventory l luotope in ci ; 8 Zr 95 1.67 x 10 l 8 - Zr 97 1.65 x 10 8 hb 95 1.69 x 10 8 ho 99 1.81 x 10 { j 6 j Ru 103 1.46 x 10 7 ku 106 6.83 x 10 7 4 Ag 110* 9.72 x 10 6 Te 129* 4.65 x 10 7 le 129 2.81 x 10 8 Te 132 1.49 x 10 le 134 1.69 x 10 8 0 ba 139 1.78 x 10 8 Ba 140 1.69 x 10 Ba 141 1.74 x 10 0 8 ' r ha 142 1.40 x 10 59 , 8 La 140 1.80 x 10 f 0 Ce 141 1.77 x 10 f 8 Ce 143 1.49 x 10 l 8 Ce 144 1.26 x 10 ; 6 , Pr 143 1.53 x 10 I i 7 hd 147 6.01 x 10 8 Tc 99* 1.59 x 10 , 0 le 101 1.70 x 10 , I i l l i l l i l i 1 l l 0-78 Am. No. 39, (6/81)
ACNGS-PSAR II.B.3 ITEM II.B.3_ POST-ACCIDENT SAWLING (m\
\j NUREG 0718 REQUIREMENT !
" Applicants shall (1) review the reactor coolant and containment atmosphere sampling system designs and the radiological spectrum and chemical analysis f acility designs, and (2) modify their plant designs as necessary to meet the requirement s. Applicants shall, to the extent possible, provide preliminary design inf ormation at a level consisteat with that normally required at the construction pe rmit stage of review. Where new designs are involved , ,
applicants shall provide a general discussion of th?.ir approach to meeting the requirements by specifying the design concept selected and the supporting 57 design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the state-of-the-art, and that there exists reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses."
RESPONSE
The capability for post accident sampling of reactor coolant and the containment atmosphere, along with onsite analysis capability, will be provided, consistent with the requirements of NUREG 0737, II.B.3. Details are 1 59 as follows: 57 A. Sample Collection
- 1. Liquid: The capability to collect liquid samples from the reactor coolant syatem (sample location for the RCS is shown on Figure 59 gv; II.B.3-1) and suppression pool will be provided. The length of the sample lines will be as short as possible to minimize plateout.
Sample collection will not require an isolated auxiliary system to be 57 placed in operation. The sampling operation under post-accident conditions utilizing Regulatory Guide 1.3 source terms will not result in a personnel dose of greater than the dose specified in 59 GDC19 (5 rem whole body, 75 rem extremeties) . See Item II.B.2. The sample can be collected within one hour of the request f or .a samp_c. l 57 The sample re turn line will be to the suppre ssion pool. Provisions 59 1 will be incorporated to minimize radioactive release and the spread of contamination from the sample station.
- 2. Gaseous: The capability to collect containment atmosphere samples will be provided through the containment /drywell H2 sampling system described in Section 7.5.1. The same dose to workers criteria and sampling time as for liquid sampling will be met for the containment 57 atmosphere sampling operation. The sample return line will be to the containment atmosphere.
B. Sample Analysis Analysis of the post-accident samples collected per A above will be in the Personnel Access Building which will be designed and shielded such that the required analyses can be perf ormed without interf erence from 59 e.ternal radiation sources. Radiological analyses for certain radionuclides that are indicators of core damage (e.g. noble gases,
-[ iodines and cesium and non-volatile isotopes) will be performed in the
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.} 'N . counting' room and chemical analyses in the laboratory. - The . chemical . l 59 sample analysis stations are equipped with fume hoods which are exhausted 3'- ')~ ~ to the outside through HEPA filters. Doses to workers involved in ' sample
= analysis will not exceed those specified f or the sample collection and transportation operation. - . Time f or the sample analyses ~ will not exceed 57 the following:
- radiological: two hours
- - boron: two hours, if boron injection was initiated l 59 i - chlorides: twenty-four hours i 57
- total dissolved gas or hydrogen: two. hours
- dissolved oxygen: verification that dissolved oxygen is < 0.1 ppm if 59 l' chloride concentration exceeds 0.15 ppm _ 3 i- .
. Accuracy, range and sensitivity will be adequate to provide pertinent I -- data to the operator in order to describe radiological and chemical j status of the reactor coolant system.
- 57 There .are no questions regarding technical f easibility or ,
state-of-the a rt rega rding the post accident sampling capability, nor are > there any concerns as to the ability to implement the design prior to OL issuanc e. 1 . (s / i s i i e i l I t ) 1. i i i. i j I [ L f ~~s 4 i i 0-80 Am. No. 59. (6/81)
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l i v/ LOCATION OF POST-ACCIDENT I l RCS SAMPLE POINT l FIGURE II.B.3-1 l 1 o-81
r ACNGS-PSAR [ f II.B.8.3 [x- - ITDI II. B.8 RULDIAKING PROCEEDING ON DEGRADED CORE ACCIDENTS l , \ -l' s NUEEC 0718 REOUIREMENT - I
" Applicant shall:
(3) provide a system for hydrogen control capable of handling hydrogen gener- i ated by the equivalent of a 100% fuel-clad metal water reaction." ,
RESPONSE
HIAP commits to provide a hydrogen control system. Currently a number of dif-
~ ferent methods are being considered throughout the industry and it is expected 57 that these efforts will, in the future, produce valuable data upon which to select an optimum means of hydrogen control. Further, it is expected that the ;
pending rulemaking on degraded cores will determine the necessity for such a ; system. For the purposes of meeting the stated requirement, a post-accident inert ing system using CO2 as an inerting agent is proposed. However, for the reasons siven above the basic design and need for this system will be under continuiag review. - The post-accxdent inerting system is to be capable of handling the hydrogen generated by the equivalent of oxidation of 100% of the active fuel cladding. i This system assures that containment integrity is not endangered due to the combustion of hydrogen genera *.ed by the reaction between the fuel cladding and j _s the reactor coolant. Control of hydrogen combustion will be accomplished by the injection of carbon dioxide into the containment after event initiatio n, but before significant hydrogen transport to the containment. The carbon di-oxide concentration will be sufficient to render all mixtures of hydrogen and air inert , and incapable of sustaining combustion. The following criteria will be used to design the Post-Accident Inerting Sys-t em: (a) The hydrogen from a transient resulting from the reaction ci up to 100 percent of the active fuel cladding with the reactor coolen'. is assumed to 1 s tart evolving from the suppression pool surface in no lest than 45 minutes from the reactor scram. Analyses will be performed and submitted j to the NRC for review two years from construction permit issuance in order ; 59 to demonstrate that the 45 minute period fo,r hydrogen evolution from the 1 suppression pool encompasses a majority of those sequences that are major . contributors to risk (as identified in the PRA performed as described in App. ISB) in ACNGS (including at least analysis of the most probable small break accident) . ! (b) The system will have the capability for providing adequate mixing of the carbon dioxide with the containment atmosphere to assure the prevention of 57 l the combustion of hydrogen in the containment atmosphere. (c) The system will have provisions f or both long term sampling of carbon di-4 oxide and oxygen concentrations, and the addition of more carbon dioxide as necessa ry. i
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0-82 -Am. No. 59, (6/81)
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ACNGS-PSAR I II.B.8.3 O. ~ (dL Alternating current power will not be required for the systen to perform its irart ing function. 1~ (e) De system will-be single active failure proof, for either intended or inadvertent operation. (f) Inadvertent full inerting, exc'luding seismic and design basis and other j . loadings, will not produce stresses in the steel containment in excess of *
.the limits set forth -in II.B.8(4)(d) . . (g) Inadvertent full inerting of the containment will not ef fect the saf e shut
[ down of the. plant. s' (h) We system will be protected from tornado and external missile hazards. i l ' (1). ne. quantity of carbon dioxide to be injected will be limited to the ' amount reoaired to provide 61 volume percent carbon dioxide concentration 57 within tho drywell and containment for the accident condition, plus the . ] additional allowance for carbon dioxide solubility in water. - 1 (j) We containment isolation valves associated with the system will be clas-y sified as intermediate with no system integrity isolation func ion. (k) Design and operation of the system will be such that buildup of ice on the nozzles, valves or in the lines will be prevented, and the isolation j valves operability following injection will be assured. l 59 Functional Description The Post-Accident Inerting System (PAIS) will be designed so that the PAIS
- components located inside the containment can withstand conditions produced by the PAIS design basis transient.
The inert ing system will be initiated manually. We system will annunciate audibly and visually in the main control room and by special tone alarm in the 57
. c ontaincent . A time delay will be prvvided for containment evacuation, and
-continued restoration 'of water level, thus preventing unnecessary op'eration.
A second audible and visual annunciation in the main control rooni plus special 1 tone containment alarm will '.:3 given to confirm system operation based on the detection of carbon dioxide flow in the main header. . We system will be
-designed so that an inert condition is reached after a 15 minute discharge
- time.
. In general, maintenance on equipment of significant importance to safety are l i not undertaken during plant operation. If such maintenance is performed, it 59 will be as allowed by'the Technical Specifications and under technical s}}