ML20003G922
| ML20003G922 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 05/31/1981 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20003G921 | List: |
| References | |
| NUDOCS 8105040222 | |
| Download: ML20003G922 (500) | |
Text
{{#Wiki_filter:. ._ .- _= _ -. ACNGS-PSAR (~)' HOUSTON LIGHIING & POWER COMPANY ALLENS CREEK NUCLEAR GENERATING STATION - UNIT NO. '1 p PRELIMINARY SAFETY ANALYSIS REPORT 4 ( AMENDMENT NO. 57
\ INSTRUCTION SHEET This amendment contains information pertaining to New Fuel and PSAR Update. '
Each revised page bears the notation Am. No. 57, (5/81) at the bottom of the page. Vertical bars with the number 57 representing Amendment No. 57 have been used in the margins of the revised pages to indicate the location of the revision on the page.
;
The following page removals and insertions should be made to incorporate
. Amendment No. 57 into the PSAR.
REMOVE INSERT _(EXISTING PAGES) (AMENDMENT N0 4 _57 PAGES) Chapter 3 Chapter 3 1* 1*
! 4* 4*
8* 8* 9* 9*
; 11* 11* i
) 3.2-24 3.2-24 3.2-25 3.2-25 3.2-26 3.2-26 4 3.8-4 3.8-4 , 3.8-4oa
- 3. 8- 7a 3. i- 7a
) 3.8-9a 8 3.8-9a l 3.8-9b 3.8-9b 3.8-9c 3.8-9c i , 3.8-10 3.8-10 3.8-73 3.8-73 4 3.8-73oa F3.8-18 Chapter 6 Chapter 6
- 1* 1*
l 3* 3* l- 4* 4* 6* 6* 6a l 10* 10* l 10a* 6.2-32 6.2-32 6.2-32a 6.2-32a 6.2-60 6.2-60 6.2-61 6.2-61 6 2-61a 6.2-61a 6 2-61b 6.2-61b IS-1 Am. No. 57, (5/81) { / OIOVO2.11
ACNCS-PSAR REMOVE INSERT (EXISTING PAGES) (AMENDMENT No. 57 PAGES) Chapter 6 (Cont'd) Chapter 6 (Cont'd) 6.2-62 6.2-62 6.2-64 6.2-64 6.2-71 6.2-71 6.2-71a 6.2-71a 6.2-75 6.2-75
- 6.2-75a
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l Cha p t er_,7, Chapter 7 + 1* 1* 6* 6* 9* 9* 10* 10* 11* 11*
- 11a*
15* 15* 21* 21* F7.1- 2 F7.1-2 7.3-20 7.3-20 7.3-21 7.3-21 l 7.3-22 7.3-22 l 7.3-23 7.3-23 7.3-24 7.3-24 7.3-25 7.3-25 7.3-26 7.3-26 7.3-27 7.3-27 7.3-28 7.3-28 O, ' 7.3-29 7.3-29 7.3-30 7.3-30
-- 7.3-30a 7.3-169 7.3-169 IS-2 Am. No. 57,' (5/81) l .
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ACNGS-PSAR REMOVE INSERT (EXISTING PAGES) (AMBOMENT NO. 57 PAGES) O Chapter 7 (Cont'd) Chapter 7 (Cont'd) 7.5-11 7.5-11
!. 7.5-12 7.5-12 7.5-18 7.5-18 7.5-22 7.5-22 7.5-23 7.5-23 a 7.5-24 7.5-24 7.5-24a 7.5-32 7.5-32 7.5-33 7.5-33 7.5-44a
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- 7.5-44 f
- 7.5-44g 7.5-44h
- 7.5-44L
- 7.5-44j 7.5-44k j -
7.5-441
- 7.5-44m 7.5-44n F7.5-9 F7.5-9 Chapter 12 Chapter 12 1* 1*
2a* 2a* 12.1-1 12.1-1 12.1-la 12.1-la
- 12.1-lb 12.2-Sa 12.2-Sa
- 12.2-19a Chapter 13 Chapter 13 1
l 1* 1* 2* 2* 2a* 3* 3* 4* 4* , l 5* 5* l v v l 13.1-1 13.1-1 tj 13.1-2 13.1-2 13.1-3 13.1-3 13.1-4 13.1-4 13.1-5 13.1-5 IS-3 Am. No. 57, (5/81)
ACNGS-PSAR REMOVE INSERT (EXISTING PAGES) (AMENDMENT NO. 57 PAGES) chapter 13 (cont'd) chapter 13 (cont'd) 13.1-6 13.1-6 13.1-7 13.1-7 13.1-8 13.1-8 13.1-9 13.1-9
- 13.1-10 13.1-10 13.1-11 13.1-11 13.1-12 13.1-12 13.1-12a 13.1-13
, 13.1-12b 13.1-14 ! 13.1-13 13.1-15 L ? ., pe -
13.1-43 13.1-43
- 13.1-50 13.1- 51 F13.1-1 F13.1-1 F13.1-2 F13.1-2
, F13.1-3 F13.1-3 F13.1-4 F13.1-4 F13.1-5 F13.1- 5 F13.1- 6 F13.1- 6 F13.1-7 F13.1-7
, F13.1-12 F13.1-12 i
' ] F13.1-13 13.3-17 F13.1-13 13.3-17 13.3-18 13.3-18 13.3-19 13.3-19 13.3-20 13.3-20 t - 13.3-20a F13.3-1 chapter 15 chapter 15 1* 1* 6* 6* 9* ' - 10* F15.1.43-1 F15.1.43-1
- 15B-1 4 - 15B-2
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- F15B-1 l IS-4 Am. No. 57, (5/81)
ACNGS-PSAR 1 1 REMOVE INSERT (EXISTING PAGES) (AMENDMENT NO. 57 PAGES) Chapter 17 Chapter 17
- t. 1* 1*
j' 2* . 2* 3* 3* i 17.0-1 17.0-1 j 17.0-2 17.0-2 17.0-3 - 17.1-1 thru -72 17.1-1 thru -75 l F17.1.1A-1 F17.1.1A-1 .
- F17.1.1A-2 F17.1.1A-2 I F17.1.1A-3 F17.1.1A-3 l Appendix C Appendix C 1* 1*
i 3* 3* ; C1.97-1 C1.9 7- 1 j C1.9 7-2 Appendix 0 Appendix 0 t - 1*
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3* , l' - 4* j - 0-1 0-2 1 0-3 7 1 - 44 45 I - 0- 6 I 47 48 0-9 3
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ACNGS-PSAR REMOVE INSERT , (EXISTING PAGES) (AMENDMENT No. 57 PACES) n Anoendix 0 (Cont'd) Appendix 0 (Cont'd)
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ACNGS-PSAR lO REMOVE (EXISTING PAGES) INSERT (AMENDMENT NO. 57 PAGES) Anoendix 0 (Cont'd) Appendix 0 (Cont'd)
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IS-7 Am. No. 57, (5/81)
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l ACNGS-PSAR REMOVE INSERT (EXISTING PACES) (AMENDMENT No. 57 PAGES) Annandix 0 (Cont'd) Aopendix 0 (Cont'd)
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ACNGS-PSAR LIST OF EFFECTIVE PAGES CHAPTER 3 O DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS
)
1 M Amendment i 1* 57 l 2* 56 ' 3* 56 4* 57 5* 49 6* 48 7* 49 8* 57 9* 57 10* 56 11* 57 12* 56 12a 56 13* 42 14* 47 15* 48
-16* 44 16a* 39 17* 54 18* 57 i 35 1i 35 i iii 35 i iv 35 v 35 vi 35 vii 35 viii 35 . ix 35 I
x 35 xi 35 xii 35 xiii 37 xiv 35 xv 35 xvi 44 j xvii 44 xviii 44 I xix 48 xx 35 xxi 44 1 xxii 35 xxiii 35 xxiv 35 xxv 35
- Ef fective Pages/ Figures Listing i
1 Am. No. 57, (5/81) 1
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ACNGS-PSAR I,IST 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 3.2-22 35 3.2-23 35 3.2-24 57 3.2-25 57 3.2-25a 30 3.2-26 57 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 3.2-34 46 3.2-34a 46 I O 3.2-34b 3.2-35 3.2-36 32 3.2-37 - 3.2-38 - 3.2-39 - 3.2-40 -
- 3.2-41 21 j 3.2-42 21 i 3.2-43 7 3.2-44 56 3.2-45 56 3.?-46 10 3.2-47 11 3.2-48 10 3.2-49 35 i 3.2-50 11 l
3.2-51 37 3.2-52 47 s 4 Am. No. 57, (5/81)
~ -.
EFFECTIVE PAGES LISTING CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUI" MENT AND SYSTEMS Amendment Page G' 42 3.7-28c 3.7-28d 42 1 3.7 35
-3.7-30 44 3.7-31 35 3.7-32 35 3.7-32a 35
~3.7-32b 35 3.7-32c 44 3.7-32d 44 37 3.7-33 (deleted) 37 3.7-34 (deleted) 37 3.7-34a (deleted) 37 3,7-35 (deleted) 50 3.7.A-1 3.7.A-2 49 3.7.A-3 49 3.7.A-4 48 3.7.A-5 54 3.7.A-6 48 3.7.A-7 48 3.7.A-8 48 3.7.A-9 48 3.7.A-10 48 s
3.7.A-11 48 3.7.A-12 48 3.7.A-13 48 3.7.A-14 48 3.7.A-15 48 3.7.A-16 48 3.7.A-17 48 3.7.A-18 48 3.8-1 54 3.8-2 54 3.8-3 41 3f 3.8-4 57
- 3.8-4oa _ , .
- 3. 8-4a 35 35 I 3.8-4b 35 3.8-4c
(' 35 3.8-4d 35 3.8-4e 35 3.8-4f 39 3.8-4g 35 . 3.8-4h 35 3.8-41 3.8-4j 54 54 3.8-4k 3.8-5 35 8 Am. No. 57, (5/81) E__________ _ -
.. - . - -. .. - --. ... . .-~ . - - . - _ _ - - - . - .- .w ACNGS.- PSAR LIST OF EFFECTIVE PAGES (Cont 'd)
CHAPIER 3 I
- Page Amendment 3.8-6~ 54 3.8-7 54 ,
! 3.8-7r. 57 -
- .3.8-8 35 1 3.8-9 44 3.8-9a 57 4
3.8-9b 57 3.8-9c 57 , 3.8-10 57 3.8-11 54
- 3.8-12 54 3.8-13 54
3.8-13a 35 3.8-14 33 .
- 3.8-14a 44 1
3.8-14b 44
- 3.8-15 34 3.8-16 35 4
3.8-17 35 I 3.8-18 54 1 3.8-19 -35
- 3.8-20 46 3.8-20a - 46 i 3.8-20b 46 ;
j 3.8-21 56 1 3.8-22 35 I 3.8-22a 35 3.8-22b 35 3 8-22c 35
; 3.8-23 54
3.8-24 4 '+ i 3.8-24a 54 ! 3.8-23 54 l 3.8-25 54 ! 3. 8 -26.- 54 2 3.8-27 35 i 3.d-28 35 3.8-28a 35 3 8-28b 54
- 3.8-29 54 l
3.d-29a 54 3.8-29a(1) 54 i 3.d-2sb 35 !- 3.8-30 35 l 3.8-31 35 ll 3.3-314 35
' 3.8-31b 35 d
3.8-32 3.8-33 35 35 9 Am. No. 57, (5/81) r,__,,.- .
ACNGS - PSAR-LIST OF' EFFECTIVE PAGES (Cont'd) CHAPTER 3 Page Amendment- . 35 3.8 44 3.8-72a 57 3.8-73 '
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3.8-73 os 57 3 8-73a 54 54 3.8-73b 35
.3.8-74 35 3.8-75 32 3.8-75 35 3.8-77 35 3.8-78 26
'3.8-78a 26 3.8-78b 26 3.8-78c 3.8-79 54 54 i
3.8-79a 35 3.8-80 (deleted) 46 3.8-81 46 3.8-81a 46 3.8-81b 46 3.8-81c 54 3.8-82 54
3 8-82a ' 35 3.8-83 37 3.8-84 41
3.9 41 3.9 .a 41 < 3.9-lb 41 3.9-1e 41 3.9-2 35 3.9-3 '~ 35 3.9-3a 35 3.v-3b 35 3.9-3c 41 3.9-3d 41 3.9-3d (i) 35 3.9-Je 35 3.9-3f 35 ' 3.9-4 54 ! 3.9-5 54 3.9-5a 41 ! 3.9-6 41 I 3.9-6a 49 3.9-60 42 !' 3.9-5c i . l 11 Am. No. 57, (5/81) \ i
ACNGS-PSAR EFFECTIVE FIGURE LIST
- CHAPTER 3 Figure No. Anendment No. l 3.8-5 34 3.8-6 -
3.8-7 - 3.8-8 - 3.8-9 - 3.8-11 34 3.8-12 26 3.8-13 26 3.8-14 26 3.8-15 . 26 3.8-16 26 3.8-17 35 3.8-18 57 3.9-1 - 3.9-2 - 3.9-3 56 3.9-4 36 3.9- 5 56
- 3.9-6 56 3.9-7 '56 3.9-8 56
;
3.9-9 56 3.9-10 56
-3.9-11 56 3.9-12 56 3.9-13 56 3.11-1 17 i
I
)
i
- All Figures whether labelled " Unit 1" or " Units 1 & 2" are to be con-sidered applicable to Unit No. 1.
18 Am. No. 57, (5/81)
- w. e, - , - -e-,-m . , ,mm-- w -,,---- ---ww ---_-w,we, +w- 4-, w,~m--- -p-m-ew-ww-e *- -ww~
l
# ACNCS-PSA1 s._,
TABLE 3.2-1 (cont'd) ENVIRONMENTAL CAPABILITY II E Plood II' ur e Safety I* Quality Component
- Seismic Entreme Tornado /
of Nissile Protection Program Comments Supp y Class Ctsup location Category wind Principal Component XXV Component Cooling Water System
- 1. Piping and valves forming part of primary containment b b 5 boundary P 2 B A,C 1
- 2. Piping and valves, that serve potentially radio- c NA P Other D A,C NA b b active components XXVI instrument and Service Air Systems
- 1. Vessels, accumulators, sup-porting safety-related c 3 P 3 NA AC I b b systems
- 2. Piping and valves in lines betwest. above accumulators c B P 3 EA A,C 1 b b and safety-related systems
- 3. Piping and valves forming 3 part of containment boundary P 2 NA AC I b b c 4 ADS air storage bottles l 57 P 3 NA A I b b c 3 5 35tu)
- 5. Piping and valves between ADS nitrogen storage bottles and ADS safety / relief valves P 3 NA AC 1 b b c g 12VII Diesel Generator Systems
- 1. Day tanks . P 3 NA S 1 b b c 3
- 2. Piping and valves, Puel 011 System P 3 NA S.O I b b b,c 8
- 3. Piping and valves, Diesel Cooling System P 3 C S I b b c 3 4 Pumps, Diesel Cooling System P 3 C $ I b b c B
-, 5. Pep mators, Cooling Water
*7 System P 3 NA S,0 I b b c 3 5 6 Diesel-generators P 2 NA S I b b c 3
- 7. Electric modules with safety functions P 3 NA S I b b c 3
- 8. Cable, with safety function P 3 MA S I b b c 3 g 22
- 9. Diesel fuel storage tanks P 3 NA ) I b b a B l
- 10. Air Receivers, Diesel Air
" Starting System P 3 C S I b b c 3
- 11. Pipe & Valves Diesel Air 2 Starting System P 3 C S I b h c 5 0
*2. Compressor & tbtor, Diesel 005.6 Air Starting System P Other NA S NA b b c NA
*" 13. Diesel Generator HPCS CE 2 NA $ 1 b b c 5 N
m 14 Air Starting System (tanks and piping) GE 3 C S I b b c 3 IIVIII Combustible Cat. Control Systems ms ._ 8"* M 1. Hydrogen Recombiner System P 2 NA M*
- 2. Hydroger. Purge System P Other NA C
A.C.E I NA b b b c 3 b c NA
- 3. Drywell Mixing System P 2 NA 1> ,C b 4.
I b c 3 Hydrogen Analyzer System CE/P 2 NA D.C.A b E I -b c 3
?Y
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. n i
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C 0
- -. . - . ~ . . , . - . -. --. ,-
gg . 1 p.g .
. /g g\
/ \ / 1 ACNCS-PSAR.
TABLE 3.2-1 (Cont'd) ENVIRONMENTAL CAPABILITY Scope AEC Quality U ,' of Safety (*) Quality ComponentI *) Seismic III Extreme (8} Tornado /(h) Plood Assurance Principal Componentg ,) Supply Class Group Location Category Wind Missile Protection Program C,===n te XIII ' Standby Cas Treatment System
- 1. Pilters P 2 NA R I b b c B
- 2. Ductwork, valves- P 2 NA C.E.R I .b b c B
- 3. Cable with safety functions P 2 NA C.E.R I b b c B
, 4. Exhaust fans P 2 NA R I b b c B 10 ' 5. Ductwork between Puel. Q1-8.21 Handling Building and SCTS P 3 NA R 1 b b c B'
- 6. Lines penetrating FHB envelope P 3 NA R I b b c B l 40 (C) l57 III ECCS Equipment Area Cooling System
- 1. All components with safety functions P 3 - A I b b 'c B IIII Power Conversion System
- 1. Main Steam Line (MSL) from i second isolation valve to . .
and including shutoff valve P 2 B A I b b c B
- 2. Branch lines of MSL between the second isolation valve 4
, and the NSL shutoff valve,
, from branch point at MSL to 4
w and including the first A valve in the branch line P 2 B I b b c 8 26
- 3. Main feedwater line (MFL)
D from second isolation valve 17,22 i g to and including shutoff valve P 2 B A I b b c B
- 4. Branch lines of KFL between 12 .2 the second isolation valve 4 and the shutoff valve, from the branch point at MFL to and including the first
. b" , , valve in the branch line P 2 B A I b b c B
{{ p 5. Main steam line piping be-' e,,,,s , tween the HSL shutoff valve ,3 sma y y9 and the turbine main stop i 0 -k valve P Other D T NA NA NA NA NA (r) '
% 3 =,
-g 6.
7. Turbine bypass piping Branch lines of the MSL be-P Other D T K4 NA NA NA NA 4 y tween the MSL shutoff valve g and the turbine main stop O valve P Other D T NA NA NA NA N1
?
^
/~% } .
h
\- \. \~
ACNCS-PSAR - TABLE 3.2-1 (Cont'd) ENVIRONMENTAL CAPABILITT Scope I) AEC( Quality of IC Quality Component
- Seismic ('}
Safety Extreme Tornado / Flood Assurance l'rincipal Component4 ,) Supply Class Group Location Category Wind Missile Protection Program Coussents 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 Pans P 3 NA A I b b c B 8
' AreI'eDeINS P 3 NA A I b b c B 57 XX1YIII Imak Detection System (Containment) .
- 1. Temperature element CE 2 NA C I b b c 8 (z)
- 2. Differential tenperature switch GE 2 NA C I b b c B (z)
- 3. Differential flow irdicator GE 2 NA C I b b c 5 (z)
- 4. Pressure switch GE 2 NA C I b b c *B (z)
- 5. Differential pressure u indicator switch GE 2 NA b b o C I c B (z)
Y 6. Differential flow sumer CE 2 NA C I b b c B (z) 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 Dome P 2 NA 22 O I a a b B Steel Containment P 2 NA e I b b c B Q gEI Other Leak Detection uE 1. Control Building Basemat P 2 NA B I b b c B 7 Sumps 14 vel Indication E 2. Reactor Auxiliary Building P 2 NA A I b b c M[
!aa Purity Sumps Level Indication B
57
~
- 3. DC Building Cubicle and P 2 NA S I b b c B Day Tamp Sumps Level Indication 3:=
r-s
ACNGS-PSAR. i \ Aj Lines which extend into the drywell will have a penetration which will . be designed to maintain the drywell integrity after a LOCA. Drywell penetrations will l'e used in combination with a cold type containment
~
penetration assembly. Refer to Figure 3.8-4. Anchorage points ~for penetrations are not limited to the buildings being penetrated and will be selected according to ability'to withstand loadings imposed by the movements in the process line. d) Fuel Transfer Penetration (GE)
-e) Equipment and Personnel Access Penetrations Two personnel access locks will be provided. Each lock will be a welded steel assembly having two double gasketed doors in series. Provision will be made to pressurize the space between the gaskets. The doors will be mechanically interlocked to ensure that one door _cannot be opened until the second door is sealed. Provisions will be made for deliberately violating the interlock by the use of special tools and procedures under strict administrative control. Each door will be equipped with quick acting valves for equalizing the pressure across the doors. The doors will not be operable unless the pressure is equalized. Pressure equalization is possible from every point at which the associated door can be operated. Each door will be designed so that ,\
h/ with the other door open, it will withstand and seal against design and
~
testing pressures of the Containment. There will be visual indication outside each door showing whether the opposite door is open or closed and whether its valve is open or closed. In addition, limit switches will be provided to indicate remotely whether doors are open or closed. Provision will be made outside~each door for remotely closing and latching the opposite door so that in the event that one door is accidentally lef t open it can be closed by remote control. The air-locks will have nozzles installed which will permit pressure testing of the lock at any time. An interior lighting system and a communications system capable of operating from an emergency power supply will be included. An equipment access hatch with an inside diameter of about 20 feet will be provided. This will be a welded steel assembly with a double gasketed, flanged and bolted cover. Provision will be made to pressurize the space between gaskets, f) Dedicated Three Foot Penetration A dedicated three (3) foot diameter containment penetration assembly will be incorporated into the plant design. The space provision to reflect the penetration assemoly is available in the present containment 57 i design. The dedicated 3 foot diameter containment penetration assembly will consist of a capped penetration in the Steel Containment and the Shield Building located at a center line elevation of (+) 217.00 feet 3.8-4 Am. No. 57, (5/81)
l i' ACNGS-PSAR
- ..,m f\ \
'~ ') and an azimuth angle of 122 , as shown in Figure (3.8-18). The Elevator Access Room on RAB floor elevatien (+) 202.33 feet is dedicated for the Containment penetration assembly, and allows space for a future l
outboard isolation valve, if required. The space inside the Containment
- on floor elevation (+) 207.33 feet is dedicated for the Containment l penetration assembly, and a future inboard isolation valve, if required.
The penetration assembly and the welded caps will be designed in accordance with the requirements of the ASME Section III, Subsection NE, and seiseic Category I. The penetration assembly and the welded caps will be protected f rom natural phenomena in the same manner as the Containment Steel shell. Periodic tests and inspection will be performed in accordance with the normal plant operation procedure. Test connections as per 10CFR50, Appendix J, _ shall be provided. 57 There are no instruments required for the penetration assembly. The following materials will be considered. Caps: Carbon Steel ASME SA-333 grade 6, Seamless or Carbon Steel ASME SA-155 [ KCF-70, Class I b Penet ration Carbon Steel ASME SA-333 Assembly: grade 6, Seamless or Carbon Steel ASME SA-155 KCF-70, Class I 3.8.2.1.3 Containment Vacuum Relief System 3.8.2.1.3.1 Desc ription 26 The containment vessel is designed for a negative internal pressure of (-)0.8 psig. To prevent exceeding the containment. design negative pressure , redundant (single f ailure proof) vacuum relief system is provided to connect the containment volume to the annulus volume bounded 23 by the f ree standing steel containment and the shield building.
-'g N
3.8-4oa Am. No. 57, (5/81)
i l ACNGS-PSAR l 7 P, = A-45 psig internal static pressure to envelope the combined pres-sure induced by-an accident that releases hydrogen generated from 100% ( j v active fuel clad metal-water reaction and the pressure from post-acci-dent-inerting assuming carbon monoxide. See.Section 6.2.1.3.4 for the i derivation of thi -accident pressure. l Pg ,= Pressure produced by an inadvertent actuation of _ the post-accident inerting system causing full inerting (with carbon dioxide). This- l internal static pressure is 25 psig, as derived in Section 6.2.1.3.5. 57- l l P = Containment Vessel structural acceptance test ! at
'(pressure = 110 percent of P in as required by NUREG 0718 Item II.B.8.4) b) Temperatures 35(D)
T
= Design (accident) tempereture inside Containment. khen coinci-
-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 8 6.2.1 3.1.4 respectively.
'T 9
= Operating Temperature (the range is 60 to 80 F inside Containment and 51 to 95 F in the annulus). During SRV
- blowdown the increased temperature in the suppression pool is included in T . During construction T is specified n
\g as the construct, ion temperature.
1 T
= Temperature inside the containment associated with the pressure P . When coincident with P this temperature is 195 F, as d Ucribed in Section 6.2.1.3N.
.T.
= Temperature inside the containment associated with the inadvertent i
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. 57 . i T at
= Ambient temperature in the containment during the Containment 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.
- 2) The dead weight of the polar crane and its runway.
- 3) keight of platforms, walkways, equipment, piping, ventilation g duct, cable and trays, conduit, etc. These loads are generated
[ either by direct attachment to the vessel, or through support- 35 (C) - ( ing structures. (C)-Consistency 3.8-7a (D)-Design Am. No. 57, (5/81)
ACNCS-PSAR
,m 3.8.2.3.2 -Load Combinations ! )
() 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) Constructica and Test Conditions
- 1) D + L( 1 ) + 'III} + WII)
- 2) D + L(2} + T ,I2) + gP ( }+Fn Crane Test
- 3) D + L(2)(4) .i (2) , p (2) , y
- 4) . 57 D+Pst + T sp (PAIS actuation) b) Normal Operating or Shutdown Conditions l3S(3)
- 5) D+L+T o + P, + Rg +F n
- 6) D + L + I (5) , po , p d+Ro +F n n
, 7) D+ P*-"+I*" PAIS Inadvertent Actuation 57 j c) Severe Environment Loads D + L + T (5) , p
- 8) 9 ,p d+R g+ F, + F,q, 9 Re M ng b p
- 9) D+LI6) + T (6) , p (6) + R (6) , p ,p o o o n ego I
d) Extreme Environmental Loads 35
- 10) D + L + To (5) + Po +Pbd + Ro +F n
+F eqs 4
! e) Abnormal - Severe Environmental Loads Pool swell (8)
- 11) D + L + Ta (7) + Pa (7) +P bd
+P ps, P se Steam Conden- l57 sation or (9)(13) or P ch + F, + F ego Ch"EEI "E
- 12) D + L + 1 (10) , p (10) + R + F (10)+ F j57
# N 54 Long Term LOCA (12) Intermediate
- 13) D + L + T,( 11)( 14 ) , p ( 14 ) , p l57 break, ADS
+ P,c or Pch(9)(13) + R, + F, + F,q,
- 14) D + L + 1 ( W + P* ( 15 ) , p (8) , p (9) l57 Small break q, +R,+F(iO)+F,q, n
(C)-Consistency l (D)-Design ! 3.8-9a Am. No. 57, (5/81)
ACNGS-PSAR
- 15) D+L+T + P, + R, + F, + F,q, Negative Pressure l a
n( V) 16) D + L + T, + P, + R, + F n +Yj+# ego Accident at pene-tration sleeve l 17 ) D + L + T (6) ,p (6) + R (6) + F Post-accident ! o Pa
+y flooding ego 54 f) Abnormal - Extreme Environmental Loads
- 18) D + L + T (7) + P (7) + P (0) + P , Pool swell l
*** ""**~
pse , p chT9)(13)(Ra +F n +F eqs tion or chugging 19 ) D + L + T,(10) ,p (10) + R, + Fn ng erm LOCA l5 eqs
- 20) D + L + T (11)(14) + P (14) ,p (12) Intermediate l5-
+P sc # + + + ** '
4 ch a n eqs , 21) D + L + T (15) ,p (15) , p (18) ,p (9) Small break l 5 ~,
+ R, + F )xF
- 22) D+L+T +P +R +F +F Negative pressure l57 a e a n eqs
- 23) D + L + T, + P, + R,,+ F, + Y) + Feqs en r ti sleeve
- 24) D+P,+T, Degrcded core l 57 i
O 3.8-9b Am. No. 57, (5/81)
ACNGS-PSAR g' Notes: (1) Temperature and live load (including snow), during construc-
/ tion. The wind load on the shell will be considered only if
( the shield building does not provide protection during contain-ment erection. Snow and wind shall not be combined in the same loading equation. (2) . Ambient pressure, temperature and live load during test. (3) Pressure test _specified for the structural acceptance of the vessel. (4) Use load factor 1.25 for the crane lifting load. (5) T' is adjusted accordingly during SRV blowdown. o (6) Use live load, the temperature, the pressure and the pipe 35(C) reactions during shutdown, start-up, refueling or post-accident as the load combination might call fo r . (7) These are pressure and tersperature distribution during the l 54 first 100 seconds of the short term LOCA (Figures 6.2-5 and 6.2-6). (8) Single SR valve blowdown only (First Actuation). l 54 (9) It includes all local pressures in the suppression pool region p plus reactions from pipes, structures and protuberances.
\ h V (10) The pressure, the temperature and the pool water level during the long term postulated design accident (later than 100 sec- 54 onds after LOCA) . The maximum values of P and F (water dump from upper pool during accident) may
- ot be Eoincidental and therefore the worst feasible combination will be considered.
(11) The pool temperature increase due to the activation of Auto-matic Depressurization System (ADS) will be considered. (12) Use pressure resulting from the blowdown of eight ADS valves. (13) P P and P may occur sequentially while not sEmu,lt$fieouslygh 54 (14) These are pressure and temperature distribution during the intermediate break accident (IBA). (15) These are pressure and temperature distribution during the small break accident (SBA) l. l Load Combinations (e) 11,12 and (f) 18,19 cover the design of the overall 54 b vessel for the accident fluid pressure cases, include the accident press-ure accident temperature and accident fluid pressure on the vessel shell; 35(C) l the seismic loads on the shell and the penetrations; and the associated ' k / ' u 3.8-9c (C)-Consistency Am. No. 57, (5/81)
. _ _ ~.
ACNGS-PSAR 6 c,q .- values of k th The design accident loads t[
. discussed ~a$ov(e,ermal result fromload) on-the penetrations.
a postulated pipe break inside the drywell. They 35( M would not occur simultaneously with the loads "Y." which are for a pipe break-ing at a penetration.
.J
The intent of load. combinations (e)' 16 and (f) 23 is to cover the design of l54 local areas atound penetrations for any pipe break postulated to occur at a ' penetration. In such. a case, design accident loads (pressure, temperature, and fluid) would not be acting simultaneously on the overall vessel. The loads "Y" which would be acting at the penetration already include the 35(< !
' thermal effects on the penetration of the postulated pipe rupture (see l
. definition item (g) in Section 3.8.2.3.1).
Local areas will be designed by investigating the applicable loads combined. as in the above listed loading cases. Local areas to be investigated in-1 clude penetration nozzles and the surrounding shell, anchorage details, crane , runway and floor framing brackets, and the dome knuckle. Investigations of these are discussed further in Section 3.8.2.4. 4 3.8.2.3.2.1 Bottom Liner The containment bottom liner plate sill be designed in accordance with the applicable rules listed in the ACI-ASME (ACI-359), Division 2 Code Issued j in January 1975. It will also be designed in accordance with selected sec-tions of ASHE 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 Tab.le 3.8-3 are appli-j[(. cable to the liner plate design except that load factors for all load cases 54 shall be taken to equal to 1.0. 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 li ner , the vessel anchorage, the crane runway, dome platforms, intermediate floor s uppe,r t 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. However, the following discussion rep-resents, in general a typical example of the approaches utilized, and, in 1 several areas, it represents specific requirements. Q3.20 b d 3.8-10 (C)-Consistency Am. No. 57, (5/81) 1
ACNGS-PSAR Y, = Loads generated by the missile impact during the postulated break. 5 , F = Hydrostatic pressure due to post-accident flooding of the
- Pa steel containment-or its related moments and forces.
B = Buoyancy force due to Probable Maximum Flood 30.2 P ba
= Pressure loads due 1, 2, 8 or 19 Main Steam Safety Relief 54 Valve (SRV) blowdown under normal' plant operating or accident conditions.
35(U) PP , = Pool swell loads, including pipe and other structure reactions resulting from the pressure. P = Steam condensation oscillation loads, including the direct and the' feedback effects. 54 Pd = Chugging loads, including the direct and the feedback effects. , For a more exact definition of the above loads, refer to Section 3.8.3.3.1 a). P t
= Containment Vessel Structural acceptance test pressure, as described in Section 3.8.2.3.1(a) and 3.8.2.8(a)
The following loads shall also be considered for the design of the steel
-/ i containment vessel anchorage concrete: \
G) P,,= A 45 psig internal static pressure to envelope the combined pressure induced by an accident that releases hydrogen generated from 100% active fuel clad metal water reaction and the pressure from post-accident inerting assuming carbon dioxide. See Section 6.2.1.3.4 for the derivation of the accident pressure. 57 Pg = 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 g = Containment vessel structural acceptance test pressure (= 110 percent of P g , as requhed by MM 0718 Item II.B.8.0 T = Temperature inside conteinment associated with the pressure P . When coincident with P , this temperature is 195 F, a%" described in Section 6.2.U3.4. T.
= Temperature inside the containment associated with the inad-vertent actuation of the post-accident inerting system causing full inerting with carbon dioxide. This temperature is 95 F as described in Section 6.2.1.3.5.
(v ) . i 3.8-73 (U)-Update Am. No. 57, (5/81) 1
ACNGS-PSAR
= Ambient temperature in the containment during the contain-T* 57 ,- ment vessel structural acceptance test. This ranges from
( 30 F to 96 F. c) Other Category I Buildings The definitions and symbols applicable -to the Category I buildings other than the Reactor Building are as specified in Section 3.8.4.3.1. 3.8.5.3.2 Loading Combinations a) . Reactor Building Mat The load combinations to be used in the design of the Reactor Building mat are as shown in Table 3.8-3. Furthermore, the following 1. ad com- -54 bination will be considered under Extreme Environmental catego::y to include the ef fect due to Probable Maximum Flood
- 1) 1.0D + 1.0L + 1.0Po + 1.0To + 1.ORo + 1.0B Additional load combinations for the steel containment vessel anchorage concrete to account for possible results c* the degraded core rulemaking and commitments associated with that include the following:
Test Condition i 1)D + Pst+ Tat Vessel Test (PAIS Actuation) g.) 57 hormal Service Condition,
- 2) D + Pio + Tin PAIS inadvertent acte.ation Factored Abnormal Condition 3 ) D + 1. 5 P , + T,y degraded core event b) Other Category I Building The load combinations to be used in the design of Category I building mats are as indicated in Sections 3.8.4.3.2.1.1.2 and 3.8.4.3.2.1.2.1.
c) In addition to the load combinations referenced above, the fol-lowing load combinations are utilized to check all Category I building foundations against sliding and overturning due to earthquakes, winds and tornadoes, and against floatation due to 35(D) fl oods . i 3.8-73ca (D)-Design Am. No. 57, (5/81)
i t m (s. N B.9 4' clo "
- ELEV ACCESS ROOM
< ..N
%"4 4 El. 217.00' - /
PENETRATION *
/
ASSEM Pat.Y j TYPE Y (M I32407)
'R W c w EitTEA .
RM FL EL
~
209.33' = = g
= =
HOLDIN6 PtlMf RM stAggl FL El 201.M' to.3 i E ll Rest %ETERIN6 ll PUMPPRE(QAT PtW TANK fL EL107,33'- FUEL TRAW REACTOR - STORA6E i T. CON 1 AINME4T l l: l-I 1 ( k i 3
ACNGS - PSAR { i W '4P (TYP) 9
/4,TELDED/ /
DEDICATED SPACE FOR FUTURE ISOLATION VALVES & PLATFORMS ( W/1,5* HEAD ROOM on Et 202.33') IO
! /4
/ RAP; ROOF EL 20'2.n* .
Io 4 gi / -
/Y / \* i i GERVICE p' ELEVR Mo t C
[
\ (4 ' -57AI A RAI N
o,#
\
f
\ -
-DEDICATED SPACE FOR FUTURE VALVES &
PLAT FORMS iFER POOL I - 6HLEt.D P;LillDING AREA 5 - STE6L CONTAINMENT 180' PART PLAN RCB El 201,3T E RAP; EL 102.M'(HOT TO SCALE) SK-R\/-M AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuc ea Generating Station THREE FOOT CONTAINMENT PENETRATION DEDICATED FOR DEGRADED CORE RULE MAKING FIGURE 3.C-18
ACNGS-PSAR fN, 6.2.1.4 Testing and Inspection NN' ] There will be leak tests conducted to verify that the Containment Sys tem will be capable of performing the functions for which it was designed. These tests will be designed to insure that the initial performance objec-tives have been met and that this performance capability will be maintained during the plant lifetime by the respective means of preoperational leak rate testing and post-operational leak rate testing. 6,2.1.4.1 Provisions for Testing i In order to facilitate leak testing and to accommodate an integrated leak rate test at any tire during plant li fe , the following provisions will be made in plant design to allow testing in accordance with 10 CFR 50 Appendix J type A, B and C tests: a) Penetrations (Type B Tests) Electrical penetrations will be provided with double seals and will be capable of being tested separately at the peak calculated pres-sure. All Containment closures which are fitted with resilient seals or gaskets will also be capable of being tested separately at the peak calculated pressure. The covers on flanged closures , such as the equipment access and personnel air-lock doors will be pro-vided with double seals and with a test tap which will allow pres-surizing the space between the seals without pressurizing the en- ,f ) tire Containment System. In addition, it will be possible to pres-Nm/ surize between the air-lock doors without pressurizing the Contain-ment. t i [n\
\
%./
)
6.2-32a Am. No. 57, (5/81)
kCNG8-PSAR
, g) The isolation valves'shall be designed to meet leak tightness stand-l i ards consistent with the overall leak tightness of the contaiament.
b h)
~
The system shall- be designed with redundancy and physical separa-tion so that no single active' failure can: result -in loss of Con-tainment integrity .~ In addition, the dependability of isolation
' valve signals will be enhanced by the use of a diversity of actua- 57 tion parameters, i) The system shall be designed to withstand the environmental con-ditions which' accompany an accident without. loss of function.
j) Isolation valve closure speeds and leak tightness will limit radio-logical-effects from exceeding guidelines values established by - 10 CFR 100. i k) -Provisions shall be made for periodically testing the operability !. and leak tightness of the isolation valves to the extent necessary-- to ensure that the system will meet its performance requirements in the event of an accident requiring Containment isolation. 4
- 1) . Containment, isolation valves will be selected on the following -
general bases: 3-
- 1) For penetrations other than ventilating and gas control systems - g a
11 a) Cate valves will be used for sized 2 1/2" and larger b) Gate valves will be used for 2" aad smaller (600 lb minimum rating). c) Globe valves will be used for 1,500 lb and higher ratings, I 2" and smaller valves. 1
- i. 2) Butterfly valves will be used on ventilating and gas control i
i systems and on certain fluid systems including the Closed Cool-ing Water System, Non-Essential Chilled Water and the Fuel Pool Cooling and Cleanup System. 57 1 { 3)- Valve materials and pressure ratings will generally be based
- 'upon its system service requirements and transient effects
- where applicable.
I' 4 m) . The type of operator used on isolation valves will generally be i-based on the following: (. ! 1) If valve must close. under high dif ferential pressure-motor operator will be used to ' avoid need for oversized air operator , 2) If valve must " fail safe" either open or closed either an air
- or electro-hydraulic operator will be used. Motor operated 57
- valves will fail "as is" i
i, !~ ! 6.2-61 Am. No. 57, (5/81) a _ _.
ACNGS-PSAR
- 3) Where fast closure is required air or electro-hydraulic opera- l 57
( ) tors will be used. (' n) Main steam isolation valves will be selected based on the following criteria:
- 1) Meet all Code and seismic requirements of Safety Class 1 3 Q6.8
- 2) Inboard and outboard valves to have dif ferent means of power closure.
- 3) Closure time of 10 seconds under downstream line break l 57 conditions.
- 4) Expected in service seat Icakage less than that dictated by radiological considerations with reasonable service periods between valve overhauls
- 5) Remain functional under emergency ambient conditions
- 6) Valve operation can be tested during normal operating condi-tions
- 7) The valves and valve operators will be ulected, taking opera- l ting experience into account, to meet the above criteria with the best reliability obtainable based on existing knowledge and O state of technology.
/
4 ( ) 6.2.4.2 System Design O The Containment 1 solation System will consist of the valves and controls required for the isolation of lines penetrating the Containment. With re- 91-9 3 s pec t to the Containment barrier, Figure 6.2-26 (1 thru 31) schematically shows the location of irolation valves on all fluid lines penetrating the Containment. 4 The number and type of leakage barriers availabic to lines which could be 8 open to the containment or drywell atmosphere and interface with the out- Ql-9,3 side atmosphere following a loss of coolant accident are covered in Sub-ecction 6.2.4.3.4 - Bypass Leakage. Table 6.2-12 summarizes the Containment penetrations and contains informa-tion as to: a) Categorized as to valves required to operate post accident (i.e., essential, intermediate, or non-essential). b) Open or closed status under normal operating conditions and acci-dent situations. 57
c) The parameters sensed to initiate isolation valve closure. b) d) The closure time and sequence of timing for principal isolation (/ valves to secure Containment isolation. 6.2-61a Am. No. 57, (5/81)
ACNGS-PSAR j e) The capability of manual (Main' Control Room) isolation valve re-setting by operator following DBA.
's ]d n
r f) Applicable General Design Criteria, g) Penetration type as represented in Figure 3.8-4. h) Region in which penetration terminates. Isolation signals are the _following: A Reactor Vessel low water level (level 2) B Reactor vessel low-low water level (level 1) C High drywell pressure D Main steam line high radiation 57 E Main steam line high flow F Main steam line high space ambient temperature or high dif ferential temperature (in RAB) G Main steam line low pressure at turbine inlet H Containment ventilation exhaust high radiation d I RCIC System steam line to turbine
- 1) Steam line high space ambient temperatore or high dif ferential temperature j 2) Low steam line pressure
- 3) High steam flow
- 4) High turbine exhaust pressure I
J Reactor Wat. J1eanup System high dif ferential flow K Reactor Water Cleanup System high space ambient temperature or i high dif ferential temperature L RRR System area high space ambient temperature or high dif ferential i temperature M Condenser low vacuum N Main Steam Line high space ambient tempature or high dif ferential ( temperature (in the steam tunnel in the TGB)
-f RM Remote Manual from Control Room (dm i
I 6.2-61b Am. No. 57, (5/81)
ACNGS-PSAR Refer to Section 7.3 for discussion of the instrumentation and controla
-('~'s associated with the Containment Isolation System.
t
' '- '#)
The Containment isolation signals will, in general, close those fluid pene-trations which support systems not required for emergency operation. Those fluid penetrations supporting Engineered Safety Feature Systems will be provided with remote operated isolation valves which may be closed from the Control Room, if required. In addition, associated system interlocks may operate the isolation valves. 57 In general, on fluid lines connected with the Reactor Coolant Pressure Boundary or Containment atmosphere, two isolation valves will be provided; one inside and one outside the Containment. On influent lines, the valve inside Containment will usually be a creek valve except for IN ducting. On lines which are normally or always closed during reactor operation, manual leaked closed valves will normally be used. Actuation air or motor operated valves will be employed on lines, but are open during reactor operations and which must be closed during accident conditions. Lines connected to closed seismic Category I systems inside the Contain-ment and lines penetrating the Containment below the pressure Suppression Pool water level will utilize one or more isolation valves outside the Containment. 3 (O 1 1 l
-s d
j 6.2-62 Am. No. 57, (5/81) j
ACNGS-PSAR
' }A} Tne isolation valving located within the Containment or drywell will be \s, / designed to witnatand the post-accident environment without suf fering any loss of integrity.
Saction 3.11 presents a discussion of the environmental conditions, both normal and accidental, for which the Containment Isolation Valve System will be designed. The section also discusses the qualification tests that will be required to assure the performance of the isolation valves under those environmental conditions. 6.2.4.3 Design Evaluation 6.2.4.3.1 Redundancy Each fluid line required to be isolated following an accident will be provided with redundancy of valving, electrical power and control circuitry such that no single active failure can result in loss of isolation of that line. Lines whien are part or the ECCS and which take suction from and discharge back to the suppression pool are required to be open following an accident. Only one isolation valve is provided for each of these lines. Each line which will be potentially open to Containment atmosphere following an accident will contain a minimum of two isolation valves in series. Where two electrically automatic valves are employed on a particular line, each valve will be powered from a separate onsite electrical po ~r source and each valve will receive an s automatic isolation signal from a separate Containment isolation channel. The
/N actuation instrumentation will meet the requirements of IEEE-279 (See Section ;V I 7.3).
- 6. 2.4. 3. 2 Compliance witn General Design Criteria As stated in Section 6.2.4.1, the design of isolation valving for lines penetrating the Containment shall follow the requirements of General Design criteria 54 tnrough 57 and Regulatory Guide 1.11 to the greatest extent practicable consistent with safety and reliability. Although a literal interpretation of Criteria 54 through 57 in some cases is not practical, it is possible to demonstrate adequate isolation provisions on some other defined basis.
Table 6.2-12 lists tne applicable GDC for each fluid line penetrating the Containment and Figures 6.2-26 show their arrangement. 57
- 6.2.4.3.2.1 Evaluation Against Criterion 55
{ The Reactor Coolant Pressure Boundary (as defined in 10 CFR 50, Section 50.2(v)) will consist of the Reactor Pressure Vessel, pressure retaining appurtenances attached to the vessel, valves and pipes which extend from the Reactor Pressure vessel up to and including the outermost isolation valve. The lines of the deactor Coolant Pressure Boundary which penetrate the Containment will be capable of isolating the Containment thereby precluding any significant release of radioactivity.
~>
6.2-64 Am. No. 57, (5/81) _ . _ . __ _ . _ . . . - , ., __ , _ -y . ,
ACNGS-PSAR bypasses the annulus region must be maintained within a small fraction [] of the Containment design leak rate (0.5 nercent per day) . This Icak-y/ age could be released to the environment without filtration or treatment by the SGTS.
- ' A potential source of such bypass leakage is through paths which could
.coasnunicate the Containment atmosphere directly with the unfiltered areas of Fuel Handling or Auxiliary Building atmosphere outsice the controlled Icakage areas following a LOCA. Penetrations in this category include the l following lines:
a) Main steam lines b) Containment ventilation supply and purge lines c) Lines which are connected to non-Category I piping systems on either side of the isolation valving and which may not contain a water seal during isolation conditions. Table 6.2-12 identifies those lines which are considered potential paths 57 for bypass leakage. An evaluation was made of all lines which penetrate containment to determine the number and types of barriers to bypass leak-age provided for each line. The types of bypass leakage barriers con-sidered were as follows: 8 Bypass Leakage Barrier Types 91~9 [ 4 37.1 a) Isolation valve outside containment b) Isolation valve inside containment 01-9 c) Closed Category I piping system inside containment 37.2 d) Closed Category I piping system outside containment c) Water seal in line f) Line beyond isolation valve outside containment vented to annulus 17 Line terminates outside containment in, or is enclosed by, area g) served by safety-related filtration 57 h) Main steam isolation valve leakage control system Leakage barriers of types c, d, e, f, g 'or h were consideced to ef fectively eliminate any bypass leakage. Leakage barriers of types a or b were con-sidered to limit but not eliminate bypass leakage. Therefore, lines con-taining any of the bypass leakage barriers c through g were not considered as potential bypass leakage paths. l57 l Due to their large size, special provisions will be incorporated in the design of Containment Ventilation System air purge and inlet lines to en-dure that leakage past the seats of the isolation valves will be vented ' Cs]/. to the annulus. As shown on Figure 9.4-3, the isolation valving on these i 6.2-71 Am. No. 57, (5/81)
AChGS-FSAR lines will include a valve inside the Containment and two valves outside 1/~'h the Shield Building. The piping between the two outboard valves will be ( ) provided with a small vent line open directly to the annulus. Sincs the
'- annulus will be held at a negative pressure, any small amount of seat leak-age past the Containment isolation valves will be vented to the annulus.
! ' Guard pipes are not considered potential unfiltered containment bypass leakage paths because they will be designed to withstand the consequences of a process pipe rupture without loss of integrity. All welds which serve as a boundary between the containment or drywell atmosphere and the outside atmosphere will be nonleaking to. 1 penetration welds which must meet in-service inspection requirements and be leak tested in accordance with Appendix J of 10 CFR 50. See Ammended Figure 3.8-3 for penetration descria*Jon. Pera . nir Locks As described in Section 3.8.2.1.2 two personnel air locks will be provided. 8 Each lock will be a welded steel assembly having two double-gasketed doors Q1-9 in series. Provision will be made to pressurize the space between the 37.2 gaskets. The doors will be mechanically interlocked to.be,sure that one door cannot be opened until the second door is sesled. Each door will be designed so that with the other door open, it will withstand and seal against design and testing pressures of the containment vessel. The per-
- sonnel air locks are part of the steel containment system, and are not con-siderd a potential source of unfiltered by pass leakage. Refer to Section f- 16.4.7 for periodic leak rate testing requirements for the containment sys-tem. The air-locks will have nozzles installed which will permit pressure (S\m_-) testing of the lock at any time to assure leak tightness of the outermost door.
I i I 't
(
\,./
6.2-71a Am. No. 57, (5/81) i
ACNGS-PSAR instrumentation is described in Section 7.5. f 6.2.5.2.3 Drywell Hydrogen Mixing Subsystem (DHMS) The design basis for the drywell containment mixing subsystem shall be to control the hydrogen content in the Mark III containment in accordance with the position stated in Regulatory Guide 1.7 Revision 1, 1976. The system will be Safety Class 2 seismic Category I. It will be completely redundant with duplicate piping, equipment and instrumentation located in the contain-ment. The compressors are physically separated from each other by missile 45(C) shield walls. The DHMS is shown in Figure 6.2-65. It consists of two redundant sub-systems. In the direction of airflow, each subsyste.n consists of an air compressor, cooling coil, check valve , motor-operated turterfly valve and 37(D) connecting piping that penetrates the drywell ceiling. Design data for principal system components are shown in Table 6.2-25. The required flow rate fe diluting the drywell containment atmosphere following LOCA is 500 cfm. The flow enters the top of the dr,well and pressurizes tne drywell sufficiently to uncover the suppression pool vents. Thus, the resulting air flow, including entrained steam, will exit near 37(C) the bottom of the drgell in a uniform peripheral pattern. The entrained steam will be condensed in the suppression pool. The cooling coil in the DHMS is an air-to-water heat exchanger and is used to cool the compressed supply air. Water to the cooling coil is supplied 37(D) j from the Containment Heat Removal System (See Sect ion 6.2.2) . s-Tne DHMS valves F014, F024, F016 and F026 functica as follows : Each compressor will have both a check valve and an Au motor operated but-terfly valve located in series in the 6" discharge line. See Figure 6.2-65 for tne valving diagram. The butterfly valve will be normally closed and will be interlocked to remain closed until start of the respective compress-o r wh ich is in series. The check valve provides redundant drywell pressure 37(D) integrity in case of loss of power to the butterfly valve. Both series valves will have position indication in the Control Room. j 6.2.5.2.4 Hydrogen Recombiner Subsystem (HRS) A thermal recombiner system will be used to control hydrogen when the con-centration in the Containment approaches four percent following a LOCA. 37(D) The system will consist of two redundant recombiners permanently installed inside the containment. This recombiner design does not require any mechanical containment pe ne t rat ions . 57 In the direction of airflow, each recombiner unit will consist of an inlet preneater section, a heater-rr. combination section and an exhaust chamber. Design data for principal system components are shown in Table 6.2-25. 37Q) The heating elements in the recombiner unit are energized manually. Na-tural convection is induced as the trsaperature of the process gas rises. The resulting flow enters the unit through the inlet louvers and passes 37(D) (C)-Consistency
- esign 6*2-75 (U)-Update Am. No. 57, (5/81)
I i j 1 A ACNGS-PSAR
;-
a The warmed stream rises and passes
!$ first through the preheater section.
through the electric heater section. The gases reach a suf ficiently high 37(D) 1 e i ! 1 1 4 l
)
I
l 1 l 2 i d i 1 ' 1 i l r 4 i
I f I i I i
- l i l t n l
l ! I
I l l l , l G 1 6.2-75a (D)-Design Am. No. 57, (S/81)
t.
~
Itoistion Nasimum Valve Cleeure/ Fenetration Fenetration Applicable Justification for Tag List Stae Velve Opeatag Number Tvoe CDC System Service (E)/(I)/fM) (E)/(II/(N) Number Valve T m (In.) Operator Time (Sec) M-1 1 55 Main Steen Line "C" N Not required ist '1821-F022C 32 Clobe (Note 8)' 26 Air 5 safe shutdown 1321-F02SC-51 Clobe (Note 8) 26 Air 5 It21-F067C-SI Clobe (Note 8) 1-1/2 notor Std (Note 2) 1E32-F N 1J-S1 Cate (Note 4) 1-1/2 _ unter - Std (hete 2) n.2 1 55 Main Steam Line "A" N Not required for 1821-F022A-S2 Clobe (Note 8) 26 Air 5 safe shutdown IB21-F028A Clebe (Note 8) 26 Air 5 1821-F067A-Si Clobe (Note 8) 1-1/2 Motor Std (Note 2) IE32-F001A-S1 Cate (Note 8) 1-1/2 Notor Std (Note 2) I M-3 1 55 Main Steam Line "f"' N Not required for IB21-F022D-S2 Clobe (Note 8) 26 Air 5 safe shutdown IB21-F0280-SI Clobe (Note 8) 26 Air 5 IB21-F06 ?D-51 Clobe (Note 8) 1-1/2 - Motor- Std (Note 2) 1E32-F MIN-31 Cate (Note 8) 1-1/2 Motor Sed (Note 2) i M-4 1 55 Main Steam Line "B" N Not required for 1821-F0225-S* Clobe (Note 8) 26 Air 5 safe shutdown 12 1-F0283-31 Clobe (Note 8) 26 Air 5 4 1821-F0675-51 Clobe (Note 8) 1-1/2 - Motor Std (Note 2) - IE32-F001E-81 Cate ~ (Note 4) 1-1/2 - notor Std (Note 2) M-5 11 55 Main Steam Drains N Not required for la21-F016-52 Cate 3 Motor Std (Note 2) to Condenser safe shutdown 1321-F019-S1 - Cate 3 Notor Std (Note 2) Later Later 56 Main Steam Line N Not required for 1821-V317-$2 Clobe 1 Mand- - Fressurtaing Con.section safe shutdown IB21-V318-SI Clobe 1 Mand = for Appendix J Test 4 _ _ _ _ _ m__. __ _
I M -PSAR TABLE 6.2-12 CONTAllOENT PEN 5TRM}0N AND IS01ATION VALVE INFORMATION Isolation Reopen Appendix Valve Fosition Signal by J. Applicable-Flow Power (Diverse Manual Type C Figure
.. Valve Location Direction Narms! Accident Teilure Farsaeter) Only Test (See 6.2-2f (See Notes 5,6,7) .(Sheet / Iter Inside Drywell Out Open Closed . Closed B,D E.F,G,M,N,RM Yes Yes 1 / 1 Outside Shield Bldg Open . Closed Closed B,D,E,F.C.M N,RM Yes Yes Outside Shield Bldg Open Closed As Is B , D , E , F.G ,M ,N , RM Yes Yes Outside Shield Bldg Closed closed As Is RM* Yes Yes or Open Ins (de Drywell Out Open Closed Closed 'B.D.E.F,G,M,N,RM Yes Yes 1 /1 Outside Shield Bldg Open Closed Closed B,D,E,F,G,M,N,RM Yes Yes . Outside Shield Bldg Open Closed As Is B,D,E,F,G,M,N,RM Yes Yes Outside Shield Bldg Closed Closed As Is RM* Yes Yes or Open 1
Inside Drywell Out Open- Closed Closed B , D ,E . F , G ,M , N , RM Yes Yes 1 / 1 Outside Shield Bldg Open Closed Closed B,D E F G.M,N,RM Yes Yes
! Outside Shield Bldg Open Closed As Is 3, D, E , F,G ,M,N, RM Yes Yes Outside Shield Bldg Closed Closed As Is RM* Yes Yes or Open Inside Drywell Out Open Closed Closed B,D,E,F,G,M,N,RM Yes Yes 1 / 1 Outside Shield Bldg Open Closed Closed B, D,E , F,G ,M,N .RM Yes Yes outside Shield Bldg- Open Closed As Is B.D.E.F.C M,N,RM Yes Yes Outside Shield Bldg Closed Closed As Is RM* Yes Yes or Open Inside Drywell Out Open Closed As Is B , D, E , F ,G ,M,N , RM Yes Yes 1 / 1 Outside Shield Bldg Open Closed As.Is B,D,E,F,G,M,N,RM Yes Yes Inside Drywe',1 In Closed Closed Closed - -
Yes 1 / 1 Outside Shield Bldg Closed Closed Closed - - Yes
-n , ,- - -- , ,- -
m i Sheet 1 of 15 Bypass Potential L* 8' 37Pass Penetration f i' T8 Path Remarks location (See Note 11)
*
- No Steam tunnel
*' NO Steam tunnel 57 l
*
- No Steam tunnet t
** 80 Steam tunnet 4
'* No Steam tunnel A,B.H No Line used to perform Appendix J i Steam tunnel Type C toits on the main stese lines and main steam line drain valves.
6.2 103 Am No. 57, (5/81) b
1 solation Nominum Valve Closure / Fenetrition renetration
- App!! cable Justification for Tag Line Sise ' Valve Opealog .
Number Type CDC System Service IE)/(I)/(N) (E)/f11/ fM) Number Valve Tyne (In.) Operator Timur (Bac) Valve Location M-8 11 55 Feedwater "A" N Not required for 1821-F032s $1 Check 20 . Air - Outside Shield tids safe shutdown except 1821-F0104-31 Check 20 - - 1aside Dryuell the Class I portion M-9 11 55 Feedwates "B" N Mot required for 1821-F0328-52 Check 20 - Air - Outside Shield Blde safe shutdown except IB21-F0105-$2 Check 20 - -
- taside Dryuell the Class I portion M-11 Vc 56 RHR Fump "A" Suction E Performs a safety 2k12-F004A-Si Cate 24 Motor Sed (Note 2) Outside Shield Blds function; essential for vessel injection and pool cooling M-12 Ve 56 RHR Fump "B" Suction E Performs a safety 2E12-F0048-52 Cate 24 Motor. Std Qtote 2) Outside Shield Bids function; essential for vessel injection a and pool cooling M-13 VC 56 RHR Fump "C" Suction E Performs a safety 2E12-F105-S2 Cate 24 Motor Std (Note 2) Outside Shield 814g function; essential for vessel injection and pool cooling '
M-14 IV 56 RHR Fump "A" Discharge E Pump minimum flow 2E12-F024A-S1 Cate 18 Moto 90 Outside Shield Blds to Suppression Pool line - essential to 2E12-F011A-31 Clobe 4 Motor ' Std (Note 2) Outside Shictd Bida protect pump for 2E12-F064A-31 Cate 8 Motor 8 Outside Shield Bldg safety function 2E12-F066A-S1 Check 8 - - Outside Shield Bids 2E12-F017A-31 Belief 1a2 - - Outside Shield 31dg M-15 IV 56 RER Fump "B" Discharge E Pump minimum ficw 2E12-F0248-S2 Cate 18 Motor 90 Outside Shield Blds to Suppression Pool line - essential to 2E12-F064B-52 Cate 8 Motor 8 Outside Shield Blds - protect pump for 2E12-FO!!B-52 clobe 4 Motor Std (Note 2) Outs 1Je Shield 31ds ' safety function 2E12-F0465-$2 Check 8 - - Outside Shield Bldg 2E12-F0178-S2 Rstief 1x2 - - Outside Shield Bids I
h
/
j . MMDS-PSAR
. TABLE 6.2-12 (Cont'd) i _
' Sheet 2 of 15 -
Isolation Reopen Appendis gy J Applicable Bypass Potential Velve Position signal Leakage Bypase Flow Power (Diverse Manual Type C Figure Penetration Only Test (See C .2-26) Location Barriers Path Remarks tirection Normal Accident e 7,,a, .1,,lule, _ Parameter) (See Note !!) (Se e Notes 5.6,7) (sheet /tten) Steau tunnel A e 3,E No la - - - Reverse flow - No I / 2
- ~ - - Reverse flow - No No I / 2 Steam tunnel A ,5,E No la - - - Reverse (Iow -
- - -- Reverse flow - No Aux Bldg A . D.E .C No Out Open Open As is RM Nots 12 No 3 / 3c St.
3 / 3c Aus 814g AeD,E,G No Cut Open Open As is RM Note 12 No 3 / 3c Aux Blds A. D.E .C No Out Open Open As is RM Note 12 No No 4 / 3d Aux Bids A,D,E C No In Closed Closed As is B C.RM Yes 8,C.RM* Yes No Closed Closed As is Open Open As is RM* Note 12 No
- - - Reverse flow - No Closed Open or - - - No Closed Aux Bldg A.D.E.C No in Closed Closed As is B,C,Rh Yes No 4 / 3e Open Open As is RM* Note 12 No Closed Closed As is B.C.RM* Yes No
- - - Reverse flow - No Closed Open or - - - No closed 6.2-l% Am. No. 57, (5/st)
I Isolation Masimum Velve Closure / Fenetration Penetration Applicable Justification for Tag . Line Siza Valve Opeatag Number Type CDC System Service (E)/ft)/fN) (E)/ft1/(N) Number Velve Tree (In.) Operator Time (Secl Valve Locattom s M-16 IV 56 RHR Fump "C" Discharge E Pump minimum flow 2Ea2-F064C-52 Cate 8 Motor. 8 Outside Shield 314g . to Suppression Pool line - essential to 2E12-F021-52 clobe 18 Motor Std (Note 2) Outside Shield 2143; protect pump for 2E12-F046C-51 Check 8 - - Outside Shield 3143 safety function 2E12-F036-S1 Relief 4x6 - - Outside Shiald 3143 2E12-F101-S2 Relief 1x2 - - Outside Shield steg RHR Fump "C" Discharge IE12-F042C-S2 Cate . 12 Motor 22 Outside Shield side M.17 & 1S IV & Vil 55 E Safety function - Inside Drywell to LPCI core injection IE12-F041C-S2 Check 12 Air - M-19 & 20 IV & VII 55 RHR Heat Exchanger E rerforms a safety 2E12-F027A-51 Cate 12 Motor Std (Note 2) Outside Shield Bldg outlet to 1.PCI function - contain- IE12-F042A-$1 Cate 12 Motor 22 Inside costalement ment spray and core injection - essen- 2E!2-F028A-S1 Cate 12 Motar 90 laside Containment tiah for pressure control 2E12-F037A-;? Clobe 12 Motor Std (Note 2) laside Containment IE12-F041A-Si Check 12 Air - Inside Dryuell 2E12-V001A-51 Y-Fat Globe 2 Motor Std (Note 2) . Inside Containment 2E12-F025A-S! Reltaf 1 x 1-1/2 -
.- laside Cantainment M- 21 IV $6 RHR HEIR "A" Relief E RHR HEIR essential 2E12-F103A-S1 Check 1/2 - -
Outside Shield Blds Valve Discharge for safety function 2E12-F104A-31 Check 1-1/2 -
- Outside Shield Sids (
2512-F055A-31 Relief 8 x 10 - - OutsideShield31ds{ 1
^k 1
M- 33 & 23 IV & V11 55 RHR Heat Exchanger E Performs a safety 2E12-F0273-S2 Cate 1* Motor Std (Note 2) Outside Shield tids } Outlet to LPCI function - contain- IE12-F042a-S2 Cate 12 etor 22 laside containment : ment sprey and cota $ injection essen- 2E12-F0288-S2 Cate 12 Motor 90 Inside Containment ~ i tial for pressure control 2E12-F0373-S2 Clobe 12 Motor Std (Note 1) laside containment IE12-F0418-$2 Check 12 Air - loside Drywell 2E12-v0013-S2 Y-Pat Clobe 2 Motor Std (Pote 2) laside Containneet 2E12-F0253-S2 Relie f 1 x l-1/2 - - tooide Containment I'
+-
V 5
.m ACNCE-PS&E YABLE 6.2-12 (Cont'd)
Sheet 4 of 15 Isolation Reopen Appendis valve Position Signal By- J Applicable Bypass Potential Flow Peer (Dive rse Manuel Type C L** W 8 Direction Normal Figure Pesetration IFPass Accident Failure Parameter) Only ' Test (See 6.2-26) ~focation Ba rrie rs Path temerks (See Notes 5,6,7) (Sheet / Item) (See Nots !!)
~
la Closed open or - - - No 5 '/ 3e A.D.E,,C No Aux Blds Closed Reverse flow - No Reversar tiow - No Closed Open or - - - No
. closed Out Closed Closed As is B C.RM Yes No 3 / 3b Steam tunnel A,B,Q No Closed Closed As is B,C,RM Yes No In Closed Closed As is RM Yes No 4 / 3d Aux Bldg A.D.C No Closed Closed As is RM Yes No $1 In Closed Closed As is RM Yes No 5 / 3e Aux Bldg A.D.C No Closed Closed As is RM Yes No 6.2-106 As. No. 57, (5/S1)
J anses-Fees TAstA 6.2-12 (Cont'd) Sheet 3 el 1$ Isolation Reopen Appendix Signal Applicable Bypass Potential Valve Position By J Leakage Bypass Flw Fwer (Diverse Manual Type C Figure Fenetration marriers Fath Remarks Direction Normal Accident Failure Parameter) only Test (See 6.2-26) Location-
,(See Notes 5,6,7) (sheet / Item) (see Note 11)
Aux Bldg A,.D.E.C No 13 Open Opea As is RM* Note 12 No 4 / 3e Closed Closed As is B,C.RM Yes No
- - - Reverse flow - No Closed Open or - - -
No Closed - No Closed ogny - - No 2 / 3a Aux Bldg / cont A..B eD.C No In Closed Open As is RM* No
- - - Reverse flow - No Aux stdg/ Cont A 8 D.C No
> 13 Open Open As is RM Note 12 No 2 / 3. Closed Open or As is Rh* No No Closed Closed Open or As is RM* No No Clos ed $ Closed Closed As is RM* Yes No
- - - Reverse flow - No Closed Open As is RM* No No Closed Open or - - - No Closed A . P.E .C No la - - - Reverse flow - No 4 / 3d Aux Bldg
- - - Reverse
- low - No Closed open or - - - No Closed A.B.D G No in Open Open As is RM Note 12 No 2 / 3a Aus sids/ Cont Closed Open or As is RM* No No Closed Closed open or As is RM* No No Closed Closed Closed As is RM* Yes No
- - - Reverse flow - No Closed Open As 13 RM* No No Closed Open or - - - No Closed l
I 6.2 105 Am. No. 57, (5/81) J
r_ _ - - _ . _ _ - . _ _ . . _ - . _
.,i
'1 I
1 t f Isolation Maximum Valve Ciceural Finetration Fene tration Applicable Justification for Tag ;Line Sise Valve - Opening l Number Type CDC System Service JE)/(1)/(N) (E)/(1)/(N) Number Valve Type fle.) Oeerator Time (Sec) Valve tacation M-24 IV 56 RHR MIR "B" Relief E RNR HEXR essential 2512-F0553-52 Relief 8 x 10 - - Outside shield side! Valve Discharge for safety function 2E12-F1033-52 Check ' 1-1/2 - f l
=. -
Outside Shield 3143: 2E12-F104B-52 Check 1-1/2 - - Outside Shield 8143 2312-F025C-82 Relie f 1 x 1-1/2 - - :
- Outende Shield 34dg!.
I M 25 11 55 RHR Shutdown Suction 1 Safety grade, but IE12-F009-52 Cate 20 Motor 13 Inside Drywell ! Lt e not redundant - 1512-F008-51 'Cate 20 Motor 33 Outside Shield 3148-essential for pres- > sure control M 16 & 27 IV & Vill 56 RHR Heat Exchanger - E Main heat sink 2E12-F074A-$1 Clobe 2 Motor Std (Note 2) Outside Shield tids: Shell Thermal Vent during isolation - 2E12-F073&-St. Globe 2. Motor Std (Note 2) leside Dryuell . essential to pro-tect heat exchanger from overpressuriza- , tion for use in safety function . 1 M-28 & 29 IV & Vill 56 RHR Heat Exchanger = E Main heat sink 2E12-F0745-52 Clobe 2 Motor Std (Note 2) Outside Shield Blds - Shell Thermal Vent during isolation - ' 2E12-F073s-82 clobe 2 Motor Std (Note 2) Inside Dryuell essential to pro-tect heat exchanger from overpressuriza-tion for use in 5-safety function.
. . t - _ -- -
n isolation Maxiom Valve Closure / Finetration renetration Applicable Justification for Tag . Line size Valve Opening Number Type CDC System Service (E)/(1)/(N) (E)/(11/(N) Number Valve Troe (In.) Operator line (See) Valve Location M-35 Vc 56 HPCS Suction from E Safety Svtem - 2E22-F015-S3 Cate 24 ' Motor 1 in./see o tegde Shield sleg Suppression Pool essential for vessel - injection i M-36&,8 & 37 IVe & VII 55 HPCS Core injection E Safety System - IE22-F004-S3 Cate 12 Motor 27 ottside Shield 314g essential for vessel IE22-F005-53 Testable 12 Air -
- Inside h fuell -
injection Check M-39 IV 56 HPCS Minimum Flow / Test E Safety System - 2E22-F023-S3 Globe 12 Motor Std (Note 2) - Outside Shield side essential for vessel 2E22-F012-83 Cate 4 Motor 5 Outside Shield tids injec tion -2E22-F014-33 Relie f 1x2 - - Outside Shield 314g 4 M-41 Vc 56 LPCS Suction from E Safety System - 2E21-F001-31 Cate' 24 Motor Std (Note 2) . Outside Shield Blds Suppreselon Pool essential for vessel injection M-42A.5 & 43 ivc & VII 55 LPCS Core Injection E Safety System - IE21-F005-S1 'Cate 12 Motor 37 Outside Shield Sids essential for vessel 1821-F006-$1 Testable 12 ' Air - Inside Dryvell injection Check M-44 IV 56 LPCS Minimum Flow / E Safety System - 2E21-F012-51 Clobe 12 Motor Std (Note 2) ~ Outside Shield Bids Test Line essential to protect 2E21-F011-S1 Cate 4 Motor Std (Note 2) Outside Shield 5143 pump for safety 2E21-F018-S1 Relief I-1/2 x 2 - - Outside Shield Bids d function 2E21-F031-SI Rettof 'l n 2 - - Outside Shield Blds M-50 Ve 56 RClc Fump Suction E Necessary for core 2E51-F031-S1 Cate 8 Motor Std @ote 2) - outside Shield Blds from Suppression cooldown following Pool isolation from the - turbine / condenser and feedwater makeup. Not operable long term L dw
. _ . _ . _ . = - . _ _ _ _ m _
V i I , M -PSdB ; I l TABIA 6.2-12 (Cont'd) , i I' Sheet 5 of 15 j Isolation Roopen Appendix Valve Position Signal By J Applicable Bypass Potential l Flow Power (Diverse Manual Type C , Figure Leauge Bypass Direction Normal Accident Failure Parameter) only Test (See 6.2-26) . Fenetration Location Barriers Fath g ,g, (See Notes 5,6,7) (Sheet /Ites) (3ee Note 11) a Out. Closed Open or As is - RM* No No 6 / 4 Aux Bldg A.D.E.C _ No Closed I t
- In ' Closed Open As is RM* No No 6 / 4 - Aux Bldg / Cont AeB.D.C No i - - -
Reverse flow - No la closed Closed As ir A.C.RM* Yes No 6 / 4 Aux Bldg A.D.E.C No Closed Open As is RM* No No - Closed Open or - - - - 1 Closed
$(
Out Open open As is RM Note 12 No 7 / 5c Aux Bids AeD.E.C No In Closed open As is Rfe No No 7 / Sa Are Bldg / Cont A.B.D,C No
- - - Reverse fice - No J 3, . Closed Closed As is B.C.RM* Yes No y / 5b Aws Bldg. A , D,E ,C No Open Open As is RM* Note 12 No Closed Open or - - - No Closed Closed Open or ' - - - No Closed Out Closed Closed As is A.I.RM* No do 9 / 6c Aux Bldg A.D.E,C No or open 6.2-107 Am. No. 57, (5/81) 1 l
s _v -. -, e mr. . .-r .
l Isolatton Maxime Valve Closure / F nett: tion Penetration Applicable Justification for Tag Line Stae Valve Opening Number Type _ CDC System Service (E) / (1) /(N) (E)/(I)/(N) Numbe r Valve Type (In.) Ocerator Time (See) Valve Location M-51 IV 56 RCIC - Vacuum Breaker E Necessary for core 2E51-F078-$1 Cate 1-1/2 Motor Std (Note 2) Outside Shield Sids cooldown following isolatican from the turbine / condenser and feedwater makeup. Not operable long term M-52 IV 56 RCIC - Test Line E Necessary for core 2E51-F019-S1 Clobe 2 Motor 5 OutsLJe snteld Bids cooldown following ; isolation from the i turbine /cendenser and feedwater makeup. Not operable long tern M-53 11 55 RCIC - Head Spray E Necessary fur core IE51-F013-S1 Cate 6 Motor 15 Outside Shield Bldg cooldown following isolation from the 1E51-F0bf - 52 Check 6 Air - Inside Drywell turbine / condenser and IE12-F019-S1 Check 6 - - Outside Shield Blds feedwater makeup. 1E12-F023-Si Cate 6 Motor - Outside Shield Bldg Not operable long ters IE51-F065-$1 Oteck 6 - - Outside Shield 81ds M-5 5 11 $5 RCIC Steam Line to E Necessary for core IE51-F063-S2 Cate 10 Motor 12 Inside Drywell Turbine cooldoen following isolation from et.e IE51-F076-S2 Clobe ! Motor Std (Note 2) Inside Drywell turbine / condenser and feedwater makeup. 1E51-F064-S1 Cate 10 Motor 12 Outside Shield sids Not operable long term g, t
;
l AOICS-PSAR q q J TABLE 6.2-12 (Ccat'd) l Sheet 6 of 15 4 M isolation Reopen Ap,endix Valve Positi<m Signal By J Applicable Bypass Potential 1 Pwe r (Dive rse Manual Type C Figure Penetration Leakage Bypass 71ow Dir'ction Normal Accident Failure Parameter) Only Test Dee 6.2-26) Location Barriers Path Remarks (See Notes $,6,7) (Sheet /Ites) _(See Note 11) Out Open Open As is C,1,RM Note 12 No 10 / 6d Aux Bldg A . D,G No i No 9 / 6c Aux Bldg A.D E.C iso in Closed closed As is RM* No
$7 h
No No 8 / 6a Steam tunnel A,B D No la Closed Open or As is A,RM* closed
- - - r.everse flow - 13 o i
- - - Reverse flow - No i Open or As is llM No No in Closed ;
closed
- - - Reverse flow - No Out Open Open or As is 1,RM* No Yes 8 / 6b Steam tunnel A B.D No l
I closed ' Closed Open or As is 1,RM* No Yes 4 closed Open Open or As is 1,RM* No Yes j c losed 6.2-107a Am. No, $7, ($/81) i l t t 4
h
' r;
- )
Isolation Man tam ' Valve Closure / Fanetration Fenetration Applicable Jussification for Tag Line Sise Valve ope ning Number Type CDC System Service M I)/(r) (t) /(1) /(N) N uvree r Valve Type (In.) Ove rat or Time (3 c) __ Yahe Loennian ! 4 y M-56 111b 56 RCIC Turbine Exhaust E Necessary for core 2E51-F068-S1 Cate 12 Motor Std (Note 2) Outside shield alg Line cooldown following isolation from the 2E51-F077-51 Cate 1-1/2 Motor Std (Note 2) Outside shield sig turbine / condenser and ,e feedwater makeup. d Not operable long term .. M-58 11 55 RWCU Fump* Suction from N Not required during IC33-F001-S2 Cate 6 Motor 15 Inside Drywell Recirc Loop and leanedtately f ol- 1G33-F004-Si Cate 6 Motor 15 Outside Shteld tids y; lowing an accident. , Necessary in long i term recovery, iz I il s-M-59 IIIe 56 RWCU Fump sypass to N Not required follow- 2G33-F028-S2 Cate 4 Motor 15 !aaide Centalement / Condense r in an accident 2C33-F034-S1 Cate 4 Motor 15 Outsies shield tids / dl Q M-60 IIIa 56 RWCU Regen Heat N Not required during 2C33-F040-S2 Cate 6 Motor 15 Inside Coatstamena Exchanger to Feedwater and immediately fol- 2C33-F039-$1 Cate 6 Motor 15 Outside shield tids A} loving an accident. May be desirable in j p]', long term accident c M-61 V 56 RWCU Filter Demineraliser N Not required for C36-V011-$2 Cate 2-1/2 Air Std (Note 2) laside Contaiampet 'j safe shutdan C36 V012-S1 Cate 2-1/2 Air Std (Note 2) Outside shield 9148 M-62 Illa 56 RWCU - Fump Discharge N Not required during 2c33-F054-$1 Cate 4 Motor 15 Outside shtold DI4 [ and immediately fol. 2c33-F053-S2 Cate 4 Motor 15 Inside Contatamme .J lowing an accident. , j-May be desirable in <- long term reci.-wry y M-63 III 56 CRD - Supply Line E No credit is taken 2Cll-F083-S1 Cate 2 Motor Std (Note 2) Outside shield BI4 y for reflood but it is desirable 2C11-F122-S2 Chec k 2 . . Inside Costa d n
- 1 u !.:
R.I f*e * ? s< . ....<..m.
E d3CS-PSAR h g TA51J2 6.2-12 (Cont'd) Sheet 7 of 15 Isolation Reopen Appendin valve Position Signal By J Applicable Bypass rotentist Flow Power (Diverse Manual Type C Figure Penettstion Leskage Bypass Direction Norms 1 Accident Failure Only (See fr.2-26) Barriers Fath gemerks Persmeter) _ Test Locstion (See Notes 5,6,7) (Sheet /1 tem) (See Note 11) In Open Open or As is RM Yes do 10 / 6d Aux Bldg A D.E.C No closed Open Open or As is I.C.RM No 80 closed out Open Closed As is A,RM,Ke Yes No 11 / 7a Steam tunr.el A.B.E No Open Clesed As is A.C.RM,J,Ke Yes No 5 Out Closed Closed As is A,RM.J.K* Yes No 11 / 7b Steam tunnel A,B,E Mo Closed Closed As is A.C RM,J,K* Yes No Out Open Closed As is A.RM.J,K* Yes No 12 / 7c Steam tunnel A.B.E No Open Closed As is A.C.RM.J,K* Yes No l i In Open Closed As is B.C.H,RM Yes Yes 12 / 7d Aux Bldg A,P Yes l Open Closed As is B C.H,RM Yes Yes In Open Closed As is A ,C,RM J ,R* Yes No 13 / 7e Steam tunnel A.B.E No Open Closed A. is A.RM.J.K* Yes No In Open Closed As is RM Yes Yes 13 / a Aux Bldg / Cont A,5 Yes or open
- - - Reve-se flow -
Yes 6.2-107b Am. '.to . 57, (5/B1)
,, ,L .: -
='!"--
LJ
.~. .m ll Isolatica lessimum Valve Closure /
Psas t rat tion . . Penetration Applicable Justification for Tag Line Sise Valve Opening Number Tvoo -CDC ' system Service (E)/(11/fM) (I)/ ff)/ fM) Number Valve Tvne ilm.) Operator Ttas (sec) Valve tacatian M 49' IV 56 Equipment ruotection 1 See penetratione 2CC-V010-51 Butte rfly 12 , sootor 15 - Outside' shield ends Closed Coo 1 Lag Water $1 84 M-71 IV 56 FPC - Upper Containment N Not essential for 2PC-Y101A-Si tutterfly to - Air . 15 ' Outside Shield 314g Pool - Containment safe shutdoun 2FC-V102-52 Testable Checa - 10 - ' - Imatdo Contatement Booster Pump Discharge M- . 2 IV 56 Fuel Fool Skism r N Not essential for 2FC-V1078-S2 Butterfly - 14 Air- 15 Imaide Contatsument Drain safe shutdoun 2FC-V107A-S1 Butterfly 14 Air 15 Outside shield Fids M-74 Iva 56 FPC - upper Pool Drain N There u111 be no 2FC-Villt-52 Butterfly to Motor 15 . Inside Contatement' to Condenser transfer from 2FC-V11LA-SI Butterfly 10 botor 15 Outside Centainment'. upper RCB to main condenser post
- accident 1
- t XX 56 inclined Fuel Transfer N Not essential for 2F42-F002 Cate 24 Meaual -
Inside Conrainment Tube safe shutdoun 2F42-F004 Cate 24 Manuel - Outside Shield Blds 2F42-F003 Plus 4 IIntor 30 Outside Shield steg i N- 79A in B IVd 56 condensate Storase and N Not assumed avail- 2CT-V027 SI- Cate 6 peotor Std (Note 2) Outside Shield 814g Transfer able in ECCS analy- 2CT-v028-52 Check 6 - '- ' 1maide Contatement sis. Condensate is not required inside containment from condensate transfer system after an acc ident. l i 4 0 i 4
., . . _ _ . _ . . - - __..... ~ _ ._ ..m.. . ,
l AM-PSAR TABLE 6.2-12 (Cont'd) Sheet e of 15 leolation Reopen Appandia Valve., Position Signal By J Applicable BFPass Potential
- Flow Fouer Manual Type C Penetrattan Leakage - Bypass (Diverse . Figure Fath Remarks Direction Normal Accident Failure A rameterl' Only Test (see 6.2-26) tocation Barrters (S** Not* II)
(See Notes 5,6,7) (Sheet / Item) la Open Open Aa is B,C.BM* Yes Yes 19 / 15a Aux .' Ids / Cont A,B . Yes Fenetrations 69 6 87 are located (Nots 14) on the sees pipe run. la Open Closed Closed B C.RM Yes No 16 / 11a Fuel Handling Blds A,B,E No . 4 - - - Reverse flow - No I out Open Closed Closed C,B,RM Yes No 16 / Ilb Fuel Handling Bldg A.B.E No ' Open Closed Closed C.B RM Yes No Out Closed Closed As is C.B.RM Yes No 16 / 11c Steam tunnel A.B.E No Closed Closed As is C.B RM Yes No 4 Closed - - No 17 / 12 FHB/ Cont A.B.E No In or Out Closed Closed Closed Closed Closed - - No I Closed closed Closed RM - No Yes 18 / 13 Aux Bldg A,8 Yes In Open Closed As is C.B H RM Yes
- - - Reverse flou - Yes i
6.2-107c ~ Am. No. 57, (5/81)
i isolation Maatsum ' Valve Closure / Justification for Tag Line Stae Valve Opening P2netration Penetration Appitcable Jalve Type Operator Time (Sec) Velve Location Number Type CDC System Service (E)/(1)/(N) (E)/(f } /(Ni Number (In.) 56 Demineralised Water N Not assumed available 20W-V045-S1 Cate 4 Motor Std (Note 2) Outside Contatusent M-80A & B IVd Check 4 - - Itaide Containment in ECCS analysis. 20W-VC40-S2 Demin water is not required inside con-tainment after en acc ident. 2CC-V033-31 Butterfly 14 Motor 15 Outside Shield Stdg M-81 IV 56 Equipment Protection i Used for normai Clased Cooling Water ope ration only. Not required f or DBA, 2CC-V030-S2 Buttently 14 Motor 15 Inside Drywell M-82 VII 56 Equipment Protection I Closed Cooling Water 1ut is necessary for 1 recirrulation, clean- 2CC-V2 74-S1 Butterfly 6 Motor 15 Outside Shield tids M-83 IV 56 Equipment Protection Closed Cooling Hater up system operation. Inside Drywell 2CC-V275-S2 Check 6 - - M-84 VII 56 Equipment Protection 1 Closed Cooling Water tyd 56 Auxiliary Closed Cooling N Auxiliary closed 2CC-VO6 7-51 Butterfly 6 Motor 15 Outside Shield 31d3 ] M-86A & B 2CC-V068-S2 Check 6 - - Inside Containment l Water cooltag water system IVd 56 Auxiliary Closed Cooling N is used to cool the 2CC-V075-Si Butterfly 6 Motor 15 Outside Shield Elds ! M-85A & B 2CC-V074-S2 Motor 15 Inside contalement Water RWCU nonregenerative Butterfly & heat exchanger. This system is not needed after an ac-c ident . Check 12 - - laside Drywell j M-8 7 Vil 56 Equipment Protection 1 See Penetrations 81- 2CC-V013-52 Closed Cooling Water 84 M-89 IVd Spare (Potable Water) IVd 56 Station Air N Not use ' ming 2SA-V045-S1 Cate 3 Motor Std Olote 2) Outside Shield sids , M-91 A & B 3 Inside Containment l plant operation. 2SA-V046-S2 Testable - - Serves no safety or Check shutdown func tion. 1
v
+
1 solation Nasiam valve Closure / Pinetrition Penetration Applicable Justification for Tag Line Sise . Valve Opening Number Type CDC System Service (E)/f1}/(N) (E)/(1)/(N) Number Valve Tyne' fin.) Operator Time (Sec5 Valve Location M-80% & 3 IVd 56 Demineralized Water ni Not assumed available 2DW-v045-Si Cate 4 tsotor Std (Note 2) . Outs 1Je Containment in ECCS analysis. 2DW-V046-52 Check 4 - - Inside Contatammet Demin water is not required inside con-tainment af ter an accident. IV 56 Equipment Protection 1 Used for normal 2CC-V033-51 Dutterfly 14 Notur 15 outside Shield slag M-81 closed Cooling Water operation only. Not - M-8.I ' VII 56 Equipment Protection I requir3d for DnA, 2CC-V030-S2 autterfly 14 Motor 15 Inside Drywell closed Cooling Water but is necessary for M-83 IV 56 Equipment Protection I recirculation clean. 2CC-V274-Si sutterfly 6- noter 15 Outside Shield aids Closed Cooling Water . up system operation. 56 Equipment Protection 1 2CC-V275-S2 Check 6- - - ^ 1aside Drywell M-84 VII , Closed Cooling Water M-86A & B IVd 56 Auxiliary Closed Cooling N Auxiliary closed 2CC-V067-SI sutterfly 6 Notor .15 Outside Shield Blds Water cooling water system 2CC-V068-S2 Check - 6 - - Inside Castainment M-85A & B IVd 56 Auxiliary Closed Cooling N is used t, cool the 2CC-V075-St Butterfly. 6 Motor 15 outside Shield alds Water RWCU nonregenerative 2CC-V074-52 Butterfly 6- Motor 15 Inside Contatement hear exchanger. This system is not needed af ter an ac-cident. 56 Equipment Protection See Penetrations 81- 2CC-V013-52 Check 12 - - Inside Drywell M-87 Vil i closed Cooling Water 84 M- 89 IVd Spare (Potable Water) M-91A & B IVd 56 station Air N Not used during 2SA-V045-SI cate 3 Motor Std (Note 2) Outside Shield Sids . plant operation. 2SA-V046-52 Testable 3 - - Inside containment Serves no safety or Check shutdoun function. 4 1 1
ACuGS-PSAR T*BLE 6.2-12 (Cont'd) Sheet 9 of 15 Isolation Reopen Appendia Valve Position signal By J Applicable Bypass Potential Flow Power (Diverse Manual Type C Figure Penetration Leakage Bypass Direction Normal Accident Failure Parameter) Only Test (See 6.2-26) Loca t ion Marriers _.Fath Remarks (fee Notes 5,6,7) (Sheet /tten) (See Note 11) la closed Closed As is B,C,H,RM Yes Yes .9 / 14 Aux Bldg A,B y,,
- - - Reverse tiow - Yes Out Open Open As is B,C,RM* Yes Yes 19 / 15e Aux Blds A,B ye, (Note 14)
Out Open Open. As is B,C RN* Yes Yes 19 / 15a (Note 14) In Open Open As is 3,C.RM* Yes Yes 19 / 15a (Note 14) 5! in - - - Reverse flow - Yes 19 / 15a closed As is Yes Yes 20 / 15b Aux Bldg ,A , B ye. It Open A.C.RM
- - - Reverse flow - Yes Cat Open Closed As is A,C,RM Yes Yes 20 / 15b Open Cicsed As is A.C.RM Yes Yes in - - - Reverse flow -
Yes 19 / 15a Aux Bldg A B. Yes Fenetrations 69 & 87 are lM ated on the same pipe run B,C,H,RM Yes Yes 21 / 16 Aux Bldg AB Yes in Closed Closed As is
. - - Reve rse flow -
Yes 6.2-107d Am. No. 57, (5/el)
.,a . c .-==-mm
I; Isolation Masimum Valve . CIseere/ PInetration Penetration Applicable Justification for Tag Line Sise valve Opeatag Number Type CDC System Service ' (E)/(I)/(N) (I)/(I)/(N) Number valve Tyne (In . ) Operarot Time (Sec) Valve Emestian M-72 XV 56 Station Air N Not used during plant 2SA-V047-51 Clobe 2 Motor Std (Note 2) Inside Contatammet operation. Serves no 2SA-VO48-52 ' Testable Check 2- - - laside Dryuelt
'd i~
safety or shutdown function. 'f4-1 M-13A & S IVd 56 Instrument Air N Safety-related sys- 21A-v018-51 Clobe 2 Diaphragm 30 Outside Shield slagj tems requiring tastr 21A-V019-52 Check 2 - - Inside Costatammet j air are equipped with ; accumulators so that ! containment isolation } valves may be closed. M-94 XV 56 Instrument Air N Safety-related sys- 21A-v238 '41 Clobe 2 Diaphragm 30 Instae Containment . tems requiring teatr 21A-V239-S2 Check 2 - -~ 1aside Dryuell I 4 air are equipped with accumulators so tinat s containment isolation valves may be closed. PA 7A & B IVd 56 Drywell and Contain- N Not required for safe HS-v040-S2 tall 2-1/2 Air, 13 Inside Drywell M-100 VI 56 mer.t He tump shutdown HS-VO46-S2 Ball 2-1/2 Air 13 Imeide Containmost HS-V047-51 Bell 3 Air 15 Outside Shield steg M-98A 6 8 IVd 56 Containment LP Sump N Not required for safe LS-V096-S2 Ball 3 Air 15 Inside Containment shutdown LS-V097-S1 Ball 3 Air 15 Outside Shield sids; M-99A & a IVd 56 Contairement Decon Sump N Not required for safe . DS-V031-52 Ball 3 Air 15 Inside containment shut town DS-V032-S1 Ball 3 Air. 15 Outside Shield Blds El-102A & S IVd 56 Drywell LP Sump N Not required for safe LS-V073-S2 Ball 2-1/2 Air 13 Inside Drywell M-101 V1 56 shutdown LS-V076-51 Ball 2-1/2 Air 13 Outside Shield Sidg M-103A & 5 IVd 56 Drywell 011 Sump N Not required for safe LS-ViO3-52 Ball 2 Air 13 Inside Drywell M-104 KV 56 shutdown LS-V106-S1 Sall 2 Air ,3 Outside Shield 3143 4 i :
AIMiG-PSAB YAILE 6.2-12 (Cont'd) Sheet 10 of 15 Isolation Reopen Appendia Yalve Position Signal By J Applicable Bypass Potential Flow Feuer (Dive rse Manual Type C Figure Fenetration taakage Bypass Direc tion Normja Ace n dent Failure Parsneter) Only Yest (See 6.2-26) Location Barriers Path meesrks (See Notes 5.6,7) (Sheet /tten) (See Note i t)
.1a Closed Closed As is B.C.H,RM Yes No 21 / 16 Containment -
No Drywell penetrations Reverse flow - No In Open Closed Closed B,C,H,RM Yes Yes 21 / 17 Aus Bldg A,3 Yes
- - - Reverse flow - Yes in Open Closed Closed B.C.H,RM Yes No 21 / 17 Containment -
No Drywell penettettoes $f Reverse flow - No Out Open Closed Closed ' B.C.H ,RM Yes Yes 22 / 18a Aux Bldg A,8 Yes Open Closed Closed B C.H RM Yes Yes Open Closed Closed B.C.H ,RM Yes Yes Out Open Closed Closed B.C.H .Ist Yes Yes 22 / IBc Aux Bldg A,B yes Open Closed Closed B,C.H,RM Yes Yes i Out Open closed closed B,C,H RM Yes Yes 22 / 18b Aus Bldg A,8 Yes Open Clo=ed Closed B ,C .H ,RM Yes Yes out Ope n Closed . Closed B.C.H,RM Yes Yes 22 / 18c Aux Blds A,5 Yes Open Closed Closed B.C.H,RM Yes Yes Out Open Closed Closed B,C,H,RM Yes Yes 23 / 18d Aux Bldg A,8 Yes Open Closed CloseJ B,C,H.RM Yes Yes 6.2-107e. An. No. 57, (5/81) i
?
4 9 Isolation Maximum Valve Closure / Penetration Penetration Applicable Justification for Tag L2ne Size Valve Ope ning J umber Tree CDC Sysren Service (E)/(I)/(N) (E)/fil/(N) Number Valve Type (I n . ) Op rator Time (Sec) Valve location M-118 VI 56 Annulus LP Sump N Not required for safe 2LS-V088-S1 Ball 2-1/2 Air 13 Outside Shield Sids shutdown 2LS-V08 7-52 Ball 2-1/2 Air 13 Annulus M-Il9 V1 56 Annulus LP Sump N Not required f or safe 2LS-V086-S1 satt 2-1/2 Air 13 Outside Shield Sids shutdown 2LS-Se5-S2 Ball 2-1/2 Air 13 Asmulus
' 12C3 & B IVd 56 ADS Air (Inst Air) E ADS valves serve am 2IA-V406-S1 Clobe 1 Motor Std (Note 2) Outside Shield tids M- 131 IV 56 E important sa * : y 21A-V428-S1 Testable Check 1 - -
Inside Containment M-12'A & B IVd 56 E function in v .nging 21A-V430-SI Clobe 1 Motor Std (Note 2) Inside Containment M-12 3 IV 56 E down the reactor 21A-V432-51 Testable Check 1 - - Inside Drywell af ter an accident 2 iA-V407-S2 Clobe 1 Motor Std (Note 2) Outside Shield Olds 21A-V429-52 Testable check 1 - - Inside Containment 21A-V431-S2 clobe 1 Motor Std (Note 2) inside Containment 21A-V433-S2 Testable Check 1 - - Inside Drywell M-125 Va 56 Suppression Pool Pump N Not required for safe 2SP-P003-SI Cate 10 Motor Std (Note 2) Outside Shield Bldg Suction f rom Suppres- shutdown; used only 2SP-P002-S2 Cate 10 Motor Std (Note 2) Outside Shield 31ds ston Pool during cleanup of f pool water. I M-126 Vb 56 Supp re s s io.a Pool Pump N Not required for safe 2SP-P001-51 Cate 10 Motor Std (Note 2) Outside Shield slos Suction from Suppres- shutdown; used only 2SP-v005-$2 Testable Check 10 - - Inside Containment ston Pool d ring cleanup of i pool water. IVd 56 Chilled Water from N Not es.ential for 2CN-V301-Si Butterfly 6 Air .5 Outside Shield Bldg M-12 7 Nonessential HVAC safe shutdown 2CN-V201-S2 Butterfly 6 Air 15 Inside containnent IVd 56 Chilled Water to 2CN-V300-SI Butterfly 6 Air 15 outside Shield Bldg M-128 Nonessential HVAC 2CN-V200-S2 Check 6 - - Inside Containment l
+
t
)
7 -__ - - . -. . _ _ _ . . _ - _ _ _ _ _ _ _ - _ . _ . _ _ . . _ _ _ _ - _ . - - _ - - - - - - - - _ _--___ _ - .. _-_. - _ _ _ _ _ _ - - - _ _ _ _ _ . - - . _ - - - --- -. _ eJ Agggg.pngs TABLE 6.2 12 (C+it'd) Bheet 11 of 15 Isolation Reopen Appendix valve Position signal By J Applicable Bypass Potential Flaw Power (Dive rse Meaual Yype C Figure Penetration Leakage - Bypass pirection . Normal Accident ratture Parameter) only Yest (see 6.2-26) tocation Ba rr t
- r
- Path mensrk.
(See Notes 5.6,7) (Sheet /tten) (8** #*** I U Out Onen . Closed Closed Yes No 23 / IBe Aux Bldg
- No Fenetrates shield Building only.
B.C.H.RM Lpen Closed Closed B.C.N RM Yes No Out open Closed Closed B.C.H,RM Yes No 23 / 18e Aux Bids
- No Feastrates $hield Building only.
Open Closed Clossi B.C.H.RM Yes No Open R}r Note 13 Yes AB Yes la Open As is 24 / 19 Aux Blds
- - - Reverse flow -
Yes open Open As is RM* Note 13 No AB Yes
- - - Reverse flow - No Open Open As is RM* Note 13 Yes
- - - Reverse flow - Yes Open Open As is RM* Note 13 No (
Reverse flow - No out open Closed As is B.C.RM* Yes No 25 / 20s Fust Handling Bldg AsI #8 Open Closed As is B.C.RM* Yes No in Open Closed As is B,C.RM* Yes No 25 / 20b Fuel Handlind Bldg A*I*E E'
- - - Reverse flow - No Out Open Closed Closed B C.H,RM Yes Yes 26 / 2ta Aux Bldg A*B Yes Open Closed As is B.C.H,::M -
Yes Open Closed Closed Yes Yes 26 / 21b Aux Bldg A,B Yes in B.C.H.RP
- - - Reverse flow Yes Yes 6.2-107f Am. No. 57, (5/al)
Isolation. Nastain Valve Closure / ' Fenetritico Fenetration Applicable Justification for - Tag Line Sise Valve . Openiss Numbe r Tree CDC System Service G)/(11/ G) (t)/(I)/(M) Number Valve Tvoe fla.) Omerator Time (Esc) Valve 1.ecation M-136 later Later RWCU - Sample Line I Not required during Late r later later. later later Later , and immediately fol-lowing an accident but useful for post-acctdent sampling M-13?A & % IVd 56 Vire Protection N Availability is non- 2FF-V188-SI Cate 4 Motor Std Osote 2) outside Shteld Blds M-ILI VI essential as system 2FF-Vl90-S2 Check 4 - - laside Containment
- 1. not required. 2FF-V546-S2 Cate 4 Motor Std (Note 2) Inside Costalammet helpful or desirable 2FF-V548-SI check 4 - .
Inside Drfuell tottoutes accident. M-138A h B IVd 56 Fire Protection N Availability is non- 2FF-V229-SI Cate & Motor Std (Note 2) Outside Shield Blds M-142 VI essential as system 2FF-V231-52 Check & = = Inside Containmoet is not required. 2FF V545-52 Cate 4 Motor Std (Note 2) Inside Containment .. helpful or desirable 2FF-V547-$1 Check 4 4 - 1meida Drywell follouing accident. M-139A & B IVd 56 Seismic Fire Protection N Availability is non- 2SF-V107-31 Cate 6 Motor Std (Note 2) . Outside Shield Bida essential as system 2SF-Vil0-52 Check 6 - - Inside Coutstament is not required, helpful or desirable following accident. M-It04 & B IVd 56 Seismic Fire Protection N Availability is non- 2SF-V108-St Cate 6 Noter Std (Note 2)' Outside Shield Blds essentatt as system 2SF-V111-52 Check 6 - - Inside Cuntainment is not required , helpful er desirable following accident. m
- a
i M -PSAE Tant r 6.2 12 (Cost'd) Sheet 12 et 15 Isolation Reopen Appendix valve Position $1gnal By .I Applicable Bypass Potential Flow Power (Diverse Manual Type C Figure Fenetration Leakage typass Direction Normel Accident Fellure Parameter) Only Yest (see 6.2-26) Lucation marriers Fath Ramarks (See Notes 5,6,7) (Sheet /lten) (dee Note 11) la tJ r Closed Closed later Later Yes Later later later Later E*J1: This is a separate poet hter Yut-2 issue. la Closed Closed As is B.C.RM Yes Yes 27 / 22a Aux Blds A,3 Yes Femetratione 137A & B and 141 are Reverse flow - Yes located on the same pipe run. Closed Closed As is S.C.RM Yes No
- - - Reverse flow - No SE la closed Closed As is B C.RM Yes Yes 27 / 22a Aux Bids A,8 Yes penetrationt 1384 & B and 142 are Reverse flow -
Yes located on the same pipe rum. Closed Closed As is B C.RM Yes No
- - - Reverse flow - No I
la Closed closed As is B C.RM Yes Yes 27 / 22b Aux Bldg A,5 Yes
- - - Reverse flow - Yes a
la closed Closed As is B,C,RM Yes Yes 27 / 22b Aux slag A,a Yes
- - - Reverse flow - Yes r
6.2-107g Am. No. 57, (5/41) R 4 a
y V W U isalat1on Masimum Valve . Closure /
- Perstration Penetration Arplicable Justification for Tag Line Sise Valve . Opentag-Numbe r Type CDC System Service (E)/(I)/(N) (E)/ft)/(N) . Number Valve Tvoe (In.) Operator' Time (Sec) Valve 1.ocatica 3-200/207. V/XV 56 Containment Air Supply N Not required for 2CV-V004-S2 Betterfly - , 36 Air- 10 laeide costalement accident mitigation 2CV-V005-52 Spectacle Fig ~ 36 - -
Inside costalement 2CV-V006-S2 Butterfly 18 Air 5 Inside Contalement 2CV-V001-51 Spectacle Fig 36 - - Outelde Shield B14g - 2CV-VOC2-SI Butterfly 18 Air 5 Outside Shield Blds 2CV-V003-81 Butternly. ~ 36 Air 10 Outside shield Blds'; H-201/208 V/XV 56 Containment Exhaust Air N Not required for 2CV-V0ll-S2 Butterfly 36 Air 10 Inside containment - accid 6at mitigation 2CV-V012-S2 Spectacle Fig 36 . - - laside containment 2CV-V013-S2 Butterfly 18 . Air 5 Inside containment -'j 2CV-VA10-S1 Butterfly . 18 Air 5 Outside Shield Bids 2CV-V007-S1 Spect.cle Fig 36 - - Outside Shield 31d3 h 2CV-V008-S1 Butterfly 18 Air 5 Outside Shield 31ds 2CV-V009-S1' Butterfly' 36 Air 10 Outside Shield tids ,' 5-202 VI - SGTS Air Return E Safety-Related 2SG-V010-S1 Butterfly 24 Motor 120 Outsade Shield tids System a-203 VI - SCTS Ai-. Exhaust E Sa fe t y-Rela ted 2SG-V007-SI Butterfly 24 Motor 120 Outside Shield Blds System a 204 VI - SCT5 Air Return E Safety-Related 2SC-V020-S2 Butterfly 24 Motor 120 Outside Shield 31d3 System E405 VI - SCTS Air Exhaust E Sa fe ty-Rela ted 2SG-V017-S2 Butterfly 24 Motor 120 Outende Shield Bids System
~
It-206 VI - AVMS Air Exkaust .N Not required for 2SC-V022-S1 Butterfly 10 Air 10 Outside Shield Bldg accident mitigation 2SG-V023-S1 Butterfly 10 Air 10 outside Shield Blds , EJ15 XVI 56 Containment vacuum E Safety-Related 2CV-V014-S! Butterfly 24 Air 10 Annulus Relief Air System RCB-ACV-15-$1 Check - 24 - - Inside Containment T __J
,-mm _. . _ _ _ ,. __ _ _ . . - _ . . . . . _ - m , .
AGICS-PSAR YAll.g 4.2 12 (Cont'd) E Sheet 13 of 15 ' Isolation Reopen Appendix
- u. .
Valve Pos[ tion Signal Bv J Applicable BYPtse . Fotential
- Flou Feuer (Diverse Manual Yype C Figure Pene tration Leakage bypass Directin Normal . Accident Failure Parameter) Only Test (See 6.2-26)~ 1.ocation Barriers fath Ramerks (See Notes 5.6,7) (Sheet / Item) (See Note 11)
In Closed- Closed B C.H RM' Yes No 23 / 23e FHB/RCB A,B,F ' ~ No . V001, V005 are open manually prior to - Open Closed.. - Closed atosed - - No Pre-entry purge, otherwise they are closed.' Open Closed Closed B.C.H RM Yes No closed Closed Closed - - No Penetration 207 is the outboard vent Open closed Closed B,C,H,RM Yee No : to annulus. Open Closed Closed B.C.H,RM Yes No I Out- . Open Closed Closed B.C.H,RM Yes No 28 / 23a FHB/RCB A sEsF No V007 & V012 are open meaually prior to Closed Closed Closed - -
.No pre-entry purge, otherwise they are closed.
Open Closed Closed B,C,H,RM Yes .No V010 is used for dry well pre-entry
. Closed Closed Closed B,C,H,RM Yes No purge during refueling, otheruise it y Closed Closed Closed . -
No is closed. Open Closed Closed . B,C.H,RM Yes No Fenetracian 200 is the outboerd vent open Closed CloseJ B.C.H,RM Yes No to annulus. In Closed Open As is 'M* No No 29 / 24b Fuel Handling Bldg A No Shield Building per.etration only Out Clused Open As is RM* No No 29 / 24b Fuel Handling Blds .A No Shield Building penetration only In Closed Open As is RM
- No No 29 / 24b Fuel Handling Bldg A No _ Shield Building penetration only.
-~0ut Closed Open As is RM* No No 29 / 24b Fuel Handling Blds A No Shield building penetration only Out Open Cicted Closed B C.H,R*8 Yes No 29 / 24a Fuel Handling Bldg A No Shield building penetration only Open Closed Closed B.C.H.RM Yes No in Closed open or Closwd RM* No Yes 28 / 23b Containment A,B,G . No Closed Revers. flou -
Yes
.6.2-107h~ Am. No. 57 (5/81)
r' ~ l~ ' - r - , s. k j
.[
s. t E.. Isolation 'Masimum Valve . Closure /- Penetration Penetration Applicable Justification for Tag . Line size Valve Ope.ing Number Type CDC Sveten Service (f./(t)/(N) (E)/(1)/(N) Numbe r Valve Twee (In 1 Onerator Time (see) Vaive Location H41 A & B v 56 Containment H Purge 1 Nonsafety. May be 2CP-V003-81 Cate 3 Motor . 15 ' outside Shield alds ;
'2 required for post- 2CP-V004-S2 Cate 3 Motor 15 laside Containment i accident H control (BTP CSB 6 4)
H- 317A & B V 56 Contat.nment H Purge 2 I Nonsafety. May be 2CP-V001-31 Cate '3' not'or 15 .Outside Sh'ield Blds i Makeup Air required for post- RCB-ACV-18-S2 Check 3 .- - Inside Containment accident R2 ' "I**l t>220 XVI 56 Containment vacuum E Safety-Related System 2CV-v016-S2, Butterfly 24 Air 10 Annulue Relief Air RCB-ACV-16-S2 Check 24 - -
' Inside Containment 1-341A & B XII 56 Hydrogen Monitoring E The hydrogen analyser EV-62028-S2 Solenoid 1/2 - ' Note 16. . Inside Cor.tatement Sample from Containment system is a safety- EV-62032-b2 Solenoid 1/2 . Note 16 Outside Shield 314g
.related system re-quired to mitigate the consequences of an accident.
1- 3+!A & B XII 56 Hydrogen Monitoring E The hydrogen analyser . EV-62027-S1 Solenoid 1/2 - Mote 16 Inside Containment Sample from Containment system is a safety- EV-62031-S1 Solenoid 1/2 - Note 16 Outside Shield Blds related system re-quired to mitigate the consequences of an accident.
am ca pgAs
/
YABLE 6.2 12 (Cont'd)
. Sheet 14 of 15.
Isolation Reopen ApM adiz . valva Position Sigest By J Applicable ~ Bypass . Potential Flow Power (Diverse - Manual . Yype C Figure Penetration Leakage Bypese , . Accident Failure Parameter) Only Yest (See 6.2-26) Location Barriers ' Path jemarks -
- ' Eyc t iol - Normal (See Notes 5,6,7) .(Sheet / Item) (See Note 11)-
Out, Closed Closed As is B.C.RM Yes Yes 30 / 25b Fuel Nandling R!ds A,a Yes: . Closed . Closed ~ As is B.C.RM - Yes j a In Closed Closed As is B.C.RM Yes Yes 30 / 25a Fuel Hand!!ng Bldg A,B .Yes
- - - Reverse flow - -
- In Closed Open or Closed RM* No Yes 28 / 23b' Containment A,B,C No .
Closed
- - - Reverse flow - -
1 Out Closed Open Closed RM Yes No - RAB 4,S.D No Closed Open Closed RM Yes No = l ? i Closed Open Closed .Yes No RAB 'A,B,D No j Out RM - Closed Open Closed RM Yes No = 4 s i 4 5 I 6.2-108 h. No. '57, (5/81) 1 7' 3 a t i + ii 1
--- ~>---
.)
. ,1
- a I
i Isolation Masimum . Valve closure /" Pynetration . Penetration Applicable Justification for Tag Line Stae Valve Opening Numbe r Type CDC System Service (E)/(II/(N) (I)/(I)/(N) Number Valve Tyne (In . ) herator Time (Sec) Valve taca' tion' I-224A & B XI 56 Hydrogen Monitoring E The hydrogen analy- Later-62030-32 Check g/2 . . . Inalde Contatumsat - u turn to Containment zer system is a . EV-62034-32 solenoid 1/2 . $
. Outside shield tids safety-related sys-tem required to mitigate the conse-quences of ar. acci-dent. Isolation is -
provided. 56 Hydrogen Monitoring E The hydrogen analy- 1,ater-62029-St .. . *: heck 1/2 -- - Inside containment : 1-251A & B II
.teturn to containment ser system is a .EV-62033-51 botenoid 1/2 - 5 outside shield Bldg' safety-related sys-tem required to mitigate the c Mae-quences of an eact-dent. Isolation is
.provided.
v
. . , . ~ . . . - . _ . _ _ _ _ . . . . _ _ . . . . . . _ _ . _ _ . . . . _ . _ . _ , . _ _ . . . . _ _ . . _ _ . . ~ . . . . . _ _ . . _ . - . _ . . _. _ V .
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. TAgt2 6.2 12 (Cont'd) e Sheet 15 of'15 t
- l. Isolation Reopen Appendix i Valve Position -Signal By J- Applicable . Bypass Potent ial
, Flow ..
, Paimr (Diverse M.snual Type C ' -Figure Penetration . 1,eakage Bypass
! Direction ' Ror.3d o -. 4ceteent Fallere _ Parameter) Only Test (See 6.2 26) Location Barriers Path Reestka (See potes 5,6,7) -(Sheet / Item) (See Note <11) . I
. No
. . In - - - Reverse Flow - No - RAB A,B,D
- j. Closed Open Closed RM Yes No --
57 s. 1 In - - - Reverse flow - No - RAS A,8.D No y Closed Open . Closed RM Yes No =' s a
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i-r 1 6.2'.io, -.Am. No. 57. <5/st) i t f J
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ACNGS-PSAR (O
'v'
) NOTES TO TABLE 6.2-12 and Figure 6.2-26 (1) thru (31)
- 1) The secondary method of closure of the main steam-lines is spring action.
- 2) Standard closure time for motor-operated gate valves is approximately 12 in./ min based on nominal pipe diameter. Standard closure time for motor
> operated globe valves is approximately 4 in./ min. based on nominal pipe diameter.
- 3) Initiatic,n of valve closure time is to be simultaneous with the receipt of the isolation signal.
- 4) All motor-operated isolation valves remain in last position upon failure 57 of valve power. In general, iir-operated valves, diaphragm valves and sulenoid valves fail closed.
- 5) The notation used to designate isolation signals is:
Notation Signal A Reactor vessel low water level (Level 2) B Reactor vessel low-low water level (Level 1) 4
/ C High Drywell pressure l57 D Main steam line high radiation E Main steam line high flow F Main steam line high space ambient
temperature or high differential temperature 57 in RAB G Main steam line low pressure at turbine inlet H Containment ventilation exhaust high radiation I RCIC System steam line to turbine !
- 1) Steam line high space ambient temperature or high differential temperature
- 2) Low steam line pressure
- 3) High steam flow
' /~N
.U 6.2-109a Am. No. 57, (5/81)
ACNGS-PSAR s Notation Signal [ b
\s_,/ I RCIC System steam line to turbine (Cont'd)
- 4) High turbine exhaust pressure J Reactor Water Cleanup System high differential flow 57 K Reactor Water Cleanup System high space ambient temperature or differential temperature L RHR System area high space ambient temperature or high dif ferential temperature M Condenser low vacuum N Main steam line high space ambient temperature or high dif ferential temperature 57 (in the steam tunnel in TGB)
RM Remote manual from control room
*' System interlock Remote-manual valves can be opened or closed by remote-manual switch
/'~'} 6) during any mode of reactor operation except when automatic signal is
( _ ,/ present.
- 7) Remote-manual valves have two sets of lights - one at hand switch and one at status board.
- 8) Main steam line isolation valves require that both solenoid pilots be de-energized to close<! valves. Accumulator air pressure plus spring act to close valves when both pilots are de-energized. Voltage failure at only one pilot does not cause valve closure. 57
- 9) Testable check valves are designed for remote opening with zero differential pressure across the valve seat. The valves will close on reverse flow even though the test switch may be positioned for open. The valves open when pump pressure exceeds reactor pressure even though the test switch may be positioned for closed.
1
i i
~
v 6.2-109b Am, No. 57, (5/81)
I I
.I ACNGS-PSAR 1 l
~~s - 10) Key for Figure 6.2-26 (1) thru (31)-
! )~
'\_ / A - Air-operated Valve M - Motor-operated Valve BV - Block Valve AX - Auxiliary Tap LC - Telltale Leak Connection PX - Pressure Tap VT - Vent
- 11) Possible Bypass Leakage Barrier Designation:
A) Isolation valve outside containment B) Isolation valve inside containment C) Closed Category-I piping system inside containment D) Closed Category I piping system outside containment E) Water seal in line F) Line beyond isolation valve outside containment vented to annulus G) Line terminates outside containment in, or is enclosed by, an area 57 served by safety-related filtration. H) Main steam line isolation valve leakage control system
- 12) The essential systems are normally open and remain open during and post accident so that the essential system may perform its safety function.
These1 isolation valves will close only if system failure is sensed or by remote manual operation. Reopening af ter such a closure is by remote manual operation or by' automatic operation on system demand. If the isolation valves were closed before the accident, they would open when the safety system is started as required by system demand.
- 13) ADS air supply isolation valves close only on loss of air pressure.
Reopening is by remote manual operation only.
- 14) System should remain in operation after an accident (other than LOCA) as long as its integrity is maintained. For this flow elements will be installed both on the supply and return leg. If the tro flows are the same (except for "no-flow" indication), integrity is maintained, i
1
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i 6.2-109c Am. No. 57, (5/81)
I- , ', f' : e d' < f l i. ACNGS-PSAR i '.
- 4.
-f15) 'All'. essential' penetrations are sliown, a11' outstanding penetrations will !
~
[ either be used.:as sparas or will' not have any assigned function or'are Ldryvell. penetrations.-
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ACNGS-PSAR l l l
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i THIS PAGE HAS BEEN INTENTIONALLY DELETED 57 I Y l l l l l 6.2-110 Am. No. 57, (5/81) l l
ACNGS PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL j PX
/ /
TEST CHAMBER ! FORMED BY M - 1 (TYP) H ! PX MAIN STEAM ,, rp
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AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY p( Allens Creek Nuclear Ger. orating Station Unit 1 CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 10F 31 FIGURE 6.2 26
ACNGS - PSAR /3 3A. RHR INFLUENT LINES I'a) CONTAINMENT a I A A A A 'Ia A tr A A A n s a m
, , , , CHIELD WALL x{ , x 7 M 7 M
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) Unit I /
CONTAINMENT IFOLATION VALVE ARRANGEMENTS - SHEET 2 OF 31 FIGURE 6.2 26
1 ACNGS - PSAR l PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL
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- FROM SUPPRESSION POOL -
l HOUSTON LIGHTlHG & POWER COMPANY O Allenr. Creek Nuclear Generating Station Unit 1 CONTAINMENT ISOL ATION VALVE ARRANGEMENTS - SHEET 3 OF 31 FIGURE 6.2 26
PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIE LD WALL
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AR NO.57, W81) 3E. RHR DISCHARGES TO SUPPRESSION POOL HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Huclear Generating Station Unit 1 CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 5 OF 31 FIGURE 6.2 26
.1
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS h, VESSEL SHIELD WALL l
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HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuc en Generating Station CONTAINMENT ISOLATION VALVE ARRANGEMENTS - ShdET 6 OF 31 FIGURE 6.2.26
ACNGS - PSAR (" i PRESSURE VESSEL DRYWELL CONTAINMENT ANNULUS SHIELD WALL l D I
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, / Unit 1 CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 7 OF 31 FIGURE 6.2-26 1
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALf. O
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AM. NO. 57, (5/81)
- 6. BCIC 300' TON LIGHTING & POWER COMPANY l
q v AHens Creek Nuc eo Generating Station CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 8 OF 31 FIGURE 6.2 26 l
ACNGS - PSAR O PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIE LD WALL
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Unit 1
;
- CONT AINMENT ISOL ATION
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l VALVE ARRANGEMENTS - SHEET 9 OF FIGURE 6.2 26
ACNGS - PSAR PRESSURE ORYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL r
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6D. / AM, NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY O Allens Creek Huclear Generating Station Unit 1 CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 10 0F 31 FIGURE 6.2 26
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIE LD WALL
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AM. NO. 57, (5/81)
- 7. RWCU HOUSTON LIGHTlHG & POWER COMPANY Aller: Creek Huclear Generating Station
( Unit 1 CONTAINMENT ISOLATION VALVE ARR ANGEMENTS - SHEET 11 OF 31 FIGURE 6.2-26 l
l ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL 1
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AM. NO,57, (5/81) HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuc en Generating Station CONTAINMENT ISOL ATION VALVE ARRANGEMENTS - SHEET 12 OF 31 FIGURE 6.2-26
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL l G
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AM. NO. 57, (5/81) l HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station i Unit I CONTAINMENT ISOL ATION VALVE ARRANGEMENTS - SHEET 13 0F 31 l FIGURE 6.2-26
ACNGS - PS AR 7
\ PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL
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- 9. STANDBY LIQUID '
CONTROL INJECTION f l y 6 / DRAIN / H M! T
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I i AM. NO.57, (5/81) HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Genero+1ng Station
} Unit 1 CONTAINMENT ISOLATION l VALVE ARRANGEMENTS - SHEET 14 0F 31 FIGURE 6.2 26 l 1
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PENETRATIONS AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY Allens Creek Huc ear Generating Station CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 15 OF 31 FIGURE 6.2 26
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430 GALLONS (MIN) WATER INVENTORY y PX VENT 118. FUEL POOL / SKlMMER DRAIN
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1 AM. NO. 57, (5/81)
- 11. FUEL POOL COOLING & CLEAN-UP HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuc ea Generating Station v
CONTAlHMENT ISOL ATION VALVE ARR ANGEMENTS - SHEET 16 OF 31 FIGURE 6.2 26
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL DRAINS FROM /
/ UPPER POOLS / j
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HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Huc ear Generating Station i CONTAINMENT ISOLATION l VALVE ARRANGEMENTS - SHEET 17 OF 31 FIGURE 6.2 26
1 ACNGS - PSAR l l Q, G PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL l
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- 14. DEMINER ALIZED WATER AM. NO. 57. (5/81)
HOUSTON LIGHTING & POWER COMPANY O Allens Creek Huclear Generating Station Unit 1 CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 18 OF 31 FIGURE 6.2 26
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL (b /
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15A. CLOSED COOLING WATER EQUIPMENT PROTECTION / /
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AM. NO. 57, (5/81) HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Huc ea Generating Station CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 19 0F 31 FIGURE 6.2-26
ACNGS - PSAR O PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL
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158. CLOSED COOLING WATER / AUXILIARY j i AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY l Allens Creek Huc ea Generating Station CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 20 OF 31 FIGURE 6.2-26
ACNGS- PSAR DRYWELL CONTAINMENT ANNULUS SHIELD WALL v
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s AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY (lOg l Allens Creek Huclear Generating Station Unit 1 CONTAINMENT ISOL ATION VALVE ARRANGEMENTS - SHEET 21 OF 31 FIGURE 6.2 26
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL
/%
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ACNGS - PSAR l rm l PRESSURE DRYWELL CONTAINMENT ANNULUS VESSEL SHIELD WALL
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- 20. SUPPRESSION POOL HOUSTON LIGHTlHG & POWER COMPANY
, Allens Creek Huclear Generating Station Unit 1 CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 25 OF 31 FIGURE 6.2 26
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- O 21. NON ESSENTIAL SERVICE CHILLED WATER HOyT A n CN eel' " j"I Ge erat ng at n Unit 1 CONTAINMENT ISOL ATION VALVE ARRANGEMENTS - SHEET 26 OF 31 FIGURE 6.2 26
ACNGS - PS AR DRYWELL CONTAINMENT ANNULUS SHIE LD WALL V /
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- 22. FIRE PROTECTION SYSTEM HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuc ear Generating Station C.ONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 27 OF 31 l FIGURE 6.2 26 1
ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS [ VESSEL SHIELD WALL \%s )
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- 23. CONTAINMENT VENTILATION SYSTEM Allens Creek Nuc ear Generating Station CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 28 0F 31 FIGURE 6.2-26
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ACNGS - PSAR PRESSURE DRYWELL CONTAINMENT ANNULUS
- VESSEL SHIELD WALL l v s
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- 25. CONTAINMENT PURGE HOUSTON LIGHTING & POWER COMPANY Allens Creek Huclear Generating Station Unit 1
( CONTAINMENT ISOLATION VALVE ARRANGEMENTS - SHEET 30 OF 31 FIGURE 6.2-26
1 i ACNGS - PSAR l
) /
PRESSURE VESSEL DRYWELL CONTAINMENT ANNULUS SHIELD WALL 1 l l '
- >g LI-241 A l l I2418 l l4 ;
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l l-224B l s lC : l / HYDROGEN ANALYZER PANELS 8' T r T ' l 242A 7, ' l l2428 ! >G:
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\
ACNGS-PSAR / N The Standby Gas Treatment System is represented by input tables describing
) f an ' size, system resistance, and pressure dependent flow characteristics.
v "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 de-scribed there. In addition, only one of the two 100 percent SGIS subsys-tems is assumed to be operating. Table 6.2-5 lists.the important assumptions used in the Shield Building annulus transient analyses. A heat transfer coefficient between the containment agmosphereandthecon-tainment steel is taken to be equal to 250 BTU /hr.-f t - F. This number represents a typical condensation heat transter coefficient o the containment steel. This heat transfer coefficient is higher than a convec-tion heat transfer coefficient and will result in a faster heat transfer rate to the annulus. This is the reason that a conservative condensing heat transfer coefficient is used instead of a more realistic convection heat transfer coefficient. Inleakage to the annulus is based on a design basis leak rate of 45 cfm at
-2 inches wg which wil? Sa verified by test. During phase 3 operation the 17 SGIS exhaust rate is equa) or less than 2000 cfm at all times. This is the Q2-9.58 value used throughout the phase 3 period for LOCA dose calculations given in Chapter 15.
(p \~ s') ' Figure 6.2-22 shows the annulus pressure versus time for the above de-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 af ter 2 hours. The annulus reaches a peak of -1/4 in. H,0 at 50 minutes and would remain negative during the transient. The result s 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-tion of the active fuel clad, the follouing conditions are calculated for this event: containment pressure = 42 psig, containment temperature
= 145 F and suppression pool temperature = 195 F.
6.2.1.3.5 Containment Pressure Calculation Inadvertent Actuation 57 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: Containment pressure = 25 psig
\
Containment / Suppression pool temperature = 95 oF y ,/ 6.2-32 Am. No. 57, (5/81)
ACNGS-PSAR , 6.2.4 CONTMNMENT ISOLATION SYSTEM
/ n \,.
bl 6.2.4.1 Design Bases The Containment Isolation System will be designed to prevent the release of radioactive air or steam to the environment as a result of a loss of cool-ant accident inside the Containment. Isolation of the Containment will be accomplished by automatic isolation of all fluid systems entering the Con-tainment not serving accident consequence limiting functions. Actuation of the system will be automatic upon receipt of the appropriate signals but it can also be manually initiated from the control room. Systems penetrating containment will. be categorized as follows: a) Essential : required to mitigate the consequences of an accident (i.e. , credit taken for system action in the accident analyses) 57 b) Intermediate: could be useful in mitigating the consequences of an accident or during short term recovery, but credit was not taken for system action in the accident analysis . c) Non-essential: use is not required; helpful or desirable following an accident 4 Those lines which serve the engineered safety features will not close auto-ma tically. However, these lines may be closed manually from the Control Room to isolate any safety feature system if necessary. Additionally, (]J these lines may be closed as required by normal Engineered Safety Features System Oper;stion. 57 The Containment Isolation System meets the following specific design bases: a) Containment isolation valves shall provide the necessary isolation , of the Containment in the event of accidents or other coaditions when the free release of Containment contents cannot be permitted. b) The design of isolation valving for lines penetrating the Contain-ment shall follow the requirements of General Design Criteria 54 through 57 to the greatest extent practicable consistent with safety and reliability c) Isolation valving for instrument lines which penetrate the Contain-ment shall conform to the requirements of Regulatory Guide 1.11, i and General Design Criteria 55 and 56. d) Isolation valves, actuators, and controls shall be protected by shielding or location against damage by missiles. e) Design of 6he Containment isolation valves and associated piping and penetrations shall be seismic Category I .
' '\
.f) Containment isolation valves and associated piping and penetrations
- shall meet the requirements of the ASME Boiler and Pressure Vessel s,,) Code, Section III, Classes 1 or 2, as applicable.
r 6.2-60 Am. No. 57, (5/81)
ACNGS-PSAR 3 EFFECTIVE PAGE LIST CHAPTER 7 INSTRUMENTATION AND CONTROLS PAGE NO. AMENDMENT NO. 1* 57 la* 43 2* 46 3* 44 4* 56 5* 40 6* ' 57 6a 43 7* 44 8* 37 9* 57
, 10* 57 11* 57 11a*
57 12* 40 13
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16* 57 37 17* 39-18* 37 19* 37 20* 21* 39 4 22* 23* h 37 24* 37 i 44 11 37 iii 37 iv 37 v 37 vi 37 vii 37 viii 40 ix 37 x 37 x1 37 xii 37 xiii 37 xiv 37 xv 37 I xvi 37 ! xvii 37 l xviii 37 xix 37 xx 37 xxi 37
- Effective Pages/ Figures listings 1
Am. No. 57, (5/81)
r ACNGS-PSAR
.c EFFECTIVE PAGE LI5I CHAPTER 7 INSTRUMENTATION AND CONTROLS PAGE NO. AMENDMENT NO.
7.3-1 37 7.3-2' 37 7.3-3 37
'7.3-4 37 7.3-5 37 7.3-6 45
- 7.3-7 37 7.3-8 37 7.3-9 37 7.3-10 45 7.3-11 M 7.3-12 37 i 7.3-13 45 7.3-14 56 7.3-15 37 7.3-16 37 7.3-17 37 7.3-18 45 7.3-19 37 l 7.3-20 57 7.3-21 57 i 7.3-22 57 7.3-23 57 7.3-24 57 7.3-25 57 7.3-26 57
. 7.3-27 57 7.3-28 57 7.3-29 57 7.3-30 57 7.3-30a 57 7.3-31 45 7.3-32 37 7.3-33 37 7.3-34 37
- 7.3-35 37 7.3-36 37 l 7.3-37 37 7.3-38 49 7.3-38ca 48 7.3-38a 43 i 7.3-38b 43 7.3-39 37 7.3-40 37 0 7.3-41 7.3-41a 49 49
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6 Am. No. 56, 3/81
I ACNGS-PSAR J.r. F.TIVE PAGE LISTING CHAPTER 7 INSTRUMENTATION AND CONTROLS
- J AMENDMENT NO.
PAGE NO. 37 7.3-151 37 7.3-152 37 - 7.3-153-37 7.3-154 37 7.3-155 37 3- 7.3-156 37 7.3-157 37 7.3-158 37 7.3-159 37 7.3-160 37 7.3-161 37 7.3-162 37 7.3-163 37 7.3-164 37 7.3-165 37 7.3-166 37 7.3-167 37 7.3-168 57 7.3-169 37 7.3-170 37 7.3-171 7.3-172 37 37 7.3-173 37 7.3-174 37 7.3-175 37 7.3-176 37 7.3-177 37
; 7.3-178 37 7.3-179 7.3-180 37 37 7.3-181 37 7.3-182 37 7.3-183 37 7.3-184 37 7.3-185 37 7.3-186 37 7.3-187 37 7.3-188 37 7.3-189 7.3-190 43 7.3A-1 thru 4 (deleted) 39 7.3B-1 and 7.3B-2 (deleted) 39 O
9 Am. No. 57, (5/81) i
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ACNGS-PSAR EFFECTIVE PAGE LISTING CHAPTER 7 INSTRUMFNTATION AND CONTROL PAGE NO. AMENDMENT NO. l 7.4-1 35 ! 7.4-2 35 l 7.4-3 35 7.4-4 35 7.4-5 45 7.4-Sa 45 7.4-6 35 7.4-7 46 7.4-8 46 7.4-9 45 7.4-10 45 7.4-11 35 7.4-12 44 7.4-13 35 7.4-14 35 I 7.4-15 35 7.4-16 35 7.4- 17 35 7.4-18 35 7.4-19 35 7.4-20 35 7.4-21 56 7.4-22 35
% 7.4-23 35 j 7.4-24 35 7.5-1 27 7.5-2 37 7.5-3 37 7.5-4 39 7.5-5 37 7.5-6 37 7.5-7 37 7.5-8 37 7.5-9 -
37 7.5-10 37 7.5-11 57 7.5-12 57 7.5-13 37 l 7.5-14 37 7.5-15 ) 37 7.5-16 37 7.5-17 37 7.5-18 57 7.5-19 37 7.5-20 37 7.5-21 37 7.5-22 O( 7.5-23 7.5-24 7.5-24a 57 1 57 I 7.5-25 37 10 Am. No. 57, (5/81)
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ACNCS-PSAR
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! EFFECIIVE PAGE LISTING CHAPTER 7 AMENDMENT NO.
PAGE NO. 37 7.5-26~ 37 7.5-27 37 7.5-28 37 7.5-29 37-7.5-30 37 7.5-31 57 7.5-32 57 7.5-33 37 7.5-34 37 7.5-35 37 7.5-36 37 7.5-37 37 7.5-38 37 7.5-39 37 7.5-40 i 37 7.5-41 7.5-42 37 j O 7.5-43 7.5-44 7.5-44a 37 37 57 7.5-44b 57 7.5-44c 57 7.5-44d 57
7.5-44e 57 7.5-44f 57 l 7.5-44g 57 7.5-44h 57 7.5-441 57 7.5-44j 57 3 .- a 7. 5- 44k 57 7.5-441 57
- 7. 5- 44m 57 7.5- 44n 57 7.5- 45 37 7.5-46 37
7.5- 47 37 7.5- 48 37 7.5- 49 37 7.5- 50 37 7.5- 51 37 7.5- 52 37 7.5- 53 37 7.5- 54 37 g 37 j 7.5- 55 i 37 i 7.5- 56 l 11 Am. No. 57, (5/81)
.ACNGS-PSAR b
' EFFECTIVE PAGE LISTING
. CHAPTER 7 AMENIEENT NO.
PAGE NO. i 37 f 7.5-57 37 l 7.5-58 37 7.5-59 37
.7.5-60 37 i 7.5-61 37 7.5-62 37 7.5-63 37 i 7.5-64 37 .-
7.5-65 37 i 7.5-66 37 7.5-67 37 7.5-68 37 7.5-69 37 7.5-70 t g- ?, i i i 1 2 i l i 1 i i 11a Am. No. 57, (5/81) I 1
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ACNGS-PSAR - EFFECTIVE FIGURES LISTING
- CHAPTER 7 m
INSTRUMENTATION AND CONTROLS (% l FIGURE NO. _ AMENDMENT NO. 37 7.1-1[Sh'eet' 7.1-2 l ok 2) _ 57 7.1-2 (Sheet 2 of 2) 44 7.1-2 (Notes) 37 7.1-3
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37
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7.1-5 37 7.1- 6 37 7.1- 7 37 7.1- 7 37 7.1-8 37 7.2-1 35 7.2-2 35 7,2-3a 35 7.2-3b 35 7.2-3c 35 7.2-3d 35
, 7.2-3e 35 7.2-3f 35 7.2-3g 35 7.2-4 35 7.2-5 35
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35 7.2-10b(Amendment,-NuadEr not(shovh on page) 35 7.2-10c 35 7.2-10d 35 7.2-10e 35 7.2-10f 35-7.2-10g 35 7.2-10h 35 7.2-101 35 i 7.2-10j 35 7.2-10k 35 l 7.2-101 35 7.2-10m i 35 7.2-10m(1) 35 7.2-10n 35 , 7.2-10o 35 7.2-10P 35 7.2-109 35 l 7.2-10r 35 7.2-10s 35 7.2-10t 35 7.2-10u 35
- All figures whether labelled " Unit 1" or " Units 1 and 2",
applicable to Unit No. 1. are to be considered 15 I Am. No. 57. (5/81) i
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' ACNGS-PSAR EFFECTIVE FIGURES LISTING
' CHAPTER 7 D
). INSTRUMENTATION AND CONTROLS FIGURE NO. AMENDMENT NO.
7.5-3 . 37 7.5-4 37 7.5-5 37
- 7. 5- 6 37 7.5-7 37 7.5-8 37 7.5-9 57
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t 21 Am. No. 57, (5/81)
, r l i i !- Reference General Design Criteria for Nuclear Power Plant s Section 10CFR50 Appendix A (Tanuary 4.1972) l 13 17 18 19 20 21 22 23 24 .25 26 27 28 29 30 33 34 35 37-38 41 42 .43. 44 45 ,46 [ i i Reactor Protection System 7.1.2.1.1 X- X X X X X X X X X X l ESF Systems ECCS 7.1.2.1.3 X X X X X X X X X X ,X X X .X CRVICS 7.1.2.1.2 X X X X X X X X X X X X SCTS (5) 7.1.2.1.18 .X X X X' X X X X X' l I combustible cas control (5) 7.1.2.1.26 ) X X X X X X< X' ECCS Area Exhaust System (5) 7.1.2.1.19 X X X X X Containment Spray 7.3.2.9.2 1 X X X X X X X X X X: X-i ESF Support Systems (5) l Standby Power 7.1.2.1.20 X X X (3) (3) (3) (3) (3) ! Essential Services cooling 7.1.2.1.21 X (3) (3) (3) (3) (3) X .X X, Control Room Air Conditioning 7.1.2.1.22 X X (3) (3) (3) (3) (3) ECCS Area Fan coulers 7.1.2.1.23 X (3) (3) (3) (3) (3) Containment Vacuum Relief System 7.3.2.7 X X X X X X Suppression Pool Makeup Systen 7.3.2.8.2 X X X X X X X. X Safety-Relief Valve System 7.3.2.10 X X X X X X Systems Required for Safe Shutdown RClc 7.1.2.1.13 X X X X X X XN X' X X X 'X X Standby Liquid Control 7.1.2.1.14 X X X X X X X Reactor Shutdown Cooling (RHR) 7.1.2.1.17 K X X X X X X X X X X X ~ X X X Other Systems Required for Safety Refueling Interlocks 7.1.2.1.5 X Reactor Vessel Power Generation , ( Instrumentation 7.1.2.1.7 X 7.1.2.1.25 X X X X X Proeess Radiation Monitoring (6) % X X
- RWCU , 7.1.2.1.15 i Leak Detection 7.1.2.1.16 % X X X X X -X X X X X X Neutron Monitoring System 7.1.2.1.4 APRF , 7.1.2.1.4.4 X X X X X X X X' X IRM 7.1.2.1.4.2 X X X X X X X X X' control Systees Rod Control and Information System 7.1.2.1.6 RP RP . RP RP RP RP X X X X X RP Recirculation Flow Control 7.1.2.1.8 X X Feedwater Control 7.1.2.1.9 K-Radwaste Control (5) 7.1.2.1.27 X Pressure Regulator and TG 7.1.2.1.10 (2) (2) (2) (2) (2)
ACNCS-PSfJL FIGRI 7.s-2 , CODES AND STANDARDS k. Regulatory Guides Institute of Electrical , Divis tom of React or St endards. NRC Electronic Enzineers - Standards 54 61 63 64 1.6 1.7 1.9 1.21 1.22 1.29 1.32 1.41 1.45 1.47 1.52 1.53 1.56 1.62 1.66 1.75 279 308 323 336 338 344 352 379 384 38 7 lf 6 K X X X X X X X X X X x 57 k *1. 4
;
X X X X X X X X X X X X X X X X X 'A X X X X X X X X X X X X X X X l :.{ X X X X X X X X X X X X X (4) I 4;. X X X X X X X X X X X X X (f'"
- i. E X X X X X X X X X X X X (4) x , ? h':
X (7[ X X X X X X X X X X X X X q X (3) X X X X X (3) I I X X X (4) X l 57 . [ (3) X X X X X (3) I I I I (4) X 'j (3) I I X X X (3) I X X X X (4 ) I t ,' (3) X X X X (3) I X X X (4) X X X X X X X X X I I I I X (4) I I 'a '= X X X I X X X X X X X X X X X X X X X I X I X I X I X (4) (4) I X I X I X 57
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37(G) . '4'? xx x X X x Xu X X X X r x x Xu X 3 X X X X X X Y s X X X X X X X X X X X X X X X X X X I (1) X t X X K r X X X X X X X X X X X X X X X X X X X X X X X' X X X X X X I I I I X 39 ' X X X X X X X X X X (U)
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*See Table 7.1-2 for applicability dates of IEEE Standards and Appendix C for Regulatory Guides q j*- .%_
(G) - CESSAR 03 M '
*Y (U) - Update , + , '
Am No. 57 (5/81) 1
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ACNGS-PSAR 7.3.1.1.2 Containment and Reactor Vessel Isolation Control System f- s .
' i Instrumentation and Controls !
7.3.'1.1.2.1 System' Identification
'The Containment - And Reactor Vessel Isolation Control System I5C will in-clude the sensors, channels, ssttches, and remotely activated valve closing
.sechanisms associated with the valves, which, when closed will affect
' isolation oc the Containment or. Reactor Vessel, or both.
The _ purpose aof the system is to prevent the release of significant amounts of radioactive materials from the fuel and RCPB by automatically isolating the appropriate pipelines that penetrate the Containment while avoiding spurious closure of particular isolation valves as a result of single i f ailure. 7.'3.~1.1.2.2 Power Sources 37 i' (G) ' Power for the instrument channels and logics of the isolation control sys-tein.are supplied from the four divisionally separated 125 volt de busses. The solenoid operated valves. and solenoid operated air valves are powered by two au busses. These are the RPS "A" and "B" busses discussed in Sec-tion'7.2 Motor-operated isolation valves receive power from standby ac , busses. Power for the operation of two valves in a line is supplied from separate sources. Table 6.2-9 lists the power supply for each isolation valve.
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- 7.3.1.1.2.3 Equipment Design Table 6.2-12 lists those pipelines that will penetrate the Containment and indicates the types and locations of isolation valves inscalled in each i
4 pi peline. Figure 6.2-26 identifies these pipelines. Pipelines that pe ne- 57 , trate the Containment and drywell and directly communicate with the reactor
; vessel will generally have two isolation valves, one insite the drywell and ,
one outside the Containment. Lines that will penetrate th* Contain.nent but i not connect di rectly to the reactor vessel will have one, two or three iso-j lation valves depending on the system. Lines ending in the Suppression o - Pool, such as RCIC pump suction, will have one valve outside the Contain-i ment. . Fluid lines will normally have two valves, ' e outside and one in- r side the Containment. Ventilation lines will norma ' y have two valves, one inside the Containment with either one inside tue annulus space, or one 57 outside the Containment. See Section 6.2.4 for a more complete discussion 1 of the system. These automatic isolation valves are considered essential for protection against the gross release of radioactive material in the ., event of a breach in the Reactor Coolant Pressure Boundary. Power cables will run in raceways from the electrical source to each motor- , operated isolation valve. Solenoid valve power ' rill go from its source to the control devices for the valve. The main steam line isolation valve i controls will include pneumatic piping and an accumulator for those valves 37 that use air as the emergency motive power source. The design and perfor ~ (U)
'~'N mance of these valves are described in the topical report APED-5750 (Refer- rD) ence 1). Pressure, temperature, radiation and water level sensors will be 57 '
N- mounted on instrument cabinets in either the Containment, the Auxiliary 4 (D)- Design (G)-CESSAR E (U)-Update 7.3-20 Am. No. 57, (5/81)
A6NGS-PSAR Building or the Turbine Building. Valve position switches will be mounted 57 7'"x son air, electro-hydraulic and motor operated valves. The switches will be ( l cacased to protect them from environmental conditions. Cables from each
\~ ' sensor will be routed in conduits and cable trays to the Control Room. I All signals transmitted to the Control Room will be electrical; no pipe from the nuclear system will penetrate the Control Room. The sensor cables cnd power supply cables will be routed to cabinets in the Control or Elec-trical Equipment Room. where the logic arrangements of the system are formed. The elementary diagram for the Nuclear Steam Supply Shutof f System ' 37(U) is shown in Figure 7.3-21 c) Initiating Circuits During normal plant operation, the instrument channels, division logic and actuators that are essential to safety are energized.
When ah*ormal conditions are sensed, the instrument channel trips, > which cuases the division logic to respond and the actuators to change state to initiate isolation. Loss of two or more logic power supplies also initiates isolation. All sensors located in Non- 57 Category I buildings, will utilize electronic isolation devices to prevent system degradation. For the main steam line isolation valve control, four instrument channels are provided for each measured variable. The instrument channels trips are combined into a two-out-of-four logic using iso-lation modules to assure that no single failure in one channel can f-~s prevent the safety action by disabling another channel nor can a single failure one division logic prevent isolation from the re- , (s/) m mainder of the system. The division 1 actuator controls the "A" 37(c) solenoid on the outboard valve and division 4 the "B" solenoid. The division 2 actuator controls the "B" solenoid on the inboard valves and division 3 the "A" solenoid. For each valve to close automatically, both of its solenoids must be deenergized. The main steam line drain valves and reactor water sample valves also operate in pairs. The outboard valves close if the division 1 logic trips. The inboard valves close if the division 2 logic trips. The reactor core isolation cooling system, reactor water cleanup system and residual heat removal system isolation valves are each j controlled by two logic circuits, one for the inboard valve and a 4 second for the outboard valve. The control system for the automatic isolation valves is designed to - provide closure of valves in time to minimize the loss of coolant from the reactor used on the release of radioactive material from the Containment. A secondary design function is the prevention of l uncovering the fuel as a result of a break in those pipelines that the valve isolatas and thereby restrict the release of radioactive material to levels below the guidelines of published regulations. b)
\ %- (G)-GESSAR (U)-Upda te 7.3-21 Am. No. 57, (5/81) e
l i ACMGS-PSAR I b) Isolation Functions and Settings e t ( / The isolation trip settings of the Reactor Vessel Isolation Control i System are listed in Table 7.3-9. Table 6.2-12 which lists all the l37(U) pipelines that penetrate the Containment, shows the signals that initiate autonctic isolation. The safety functions of these isola-tion signals are discussed in the following pr.ragraphs and the func-tional control diagram (Figure 7.3-12) illustrates how these signals will initiate closure of isolation valves. l57
- 1) Reactor Vessel low and Low-Low Water Imvel A low water (level 2) level in the Reactor Vessel could in- !57 dicate that reactor coolant is being lost through a breach in the Reactor Coolant Pressure Boundary and that the core is in danger of becoming overheated as the reactor coolant inventory diminishes. Reactor Vessel low (level 2) water level initiates closure of various valves. The closure of these valves is in-tended to isolate a breach in any of the pipelines in which the valves are contained, conserve reactor coolant by closing off process lines or prevent the escape of radioactive materials l37(U) from the Containment through process lines that are in communi-cation with the Containment interior.
The Reactor Vessel low-low water level isolation settings
- (level 1) was selected low enough to allow the removal of heat 57 p from the reactor for a pre-determined tiae following the scram i and high enough to complete isolation in' time for the operation of Emergency Core Cooling Systems in the event of a large break in the Reactor Coolant Pressure Boundary. Isolation of the fol-loving pipelines is initiated when the Reactor Vessel water level falls to this second setting:
a) All four main steam lines b) Main steam line drain 37(G) c) Reactor water sample line d) RHR reactor shutdown cooling supply c) RHR reactor head spray f) Reactor water cleanup g) RHR shutdown cooling discharge to radwaste h) Drywell drain discharge to radwaste i) Drywell biced-of f pressure vent j) RHR process sampling 7.3-22 (G)-GESSAR (U)-Update Am. No. 57, (5/81)
ACNGS-ISAR
-k) Containment Normal Ventiistion Supply
- 1) Containment Normal Ventilation Exhaust (Q)
- 2) Main Steam Line High Radiation High radiation in the vicinity of the main steam lines could indicate a gross release of fission products from the fuel.
High radiation near the main steam lines initiates isolation 37(C) of the following pipelines: a) All main steam lines, b) Main steam line drain 1 c) Reactor water sample line The high radiation trip setting is selected high enough above background radiation levels to avoid spurious isolation, yet low enough to promptly detect a gross release of fission pro-ducts from the fuel. (For more information on the high radia-tion set point, see Section 7.6.1.3, " Process Radiation Moni-toring System".)
- 3) Main Steam Line High Space Ambient Temperature or Differential Temperature. ,
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High ambient temperature in the space in which the main steam lines are located in the RAB and TGB would indicate a breach in a main steam line. Also, such a breach may be indicated by high differential temperature between the outlet and inlet for this steam line space. The automatic closure of various valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear systems process barrier. When either type of high temperatures f57 occur in the main steam line space, the following pipelines are isolated: a) All four main steam lines b) Main steam line drain The RAB and TGB main steam line space high temperature trips 57 are set high enough above the temperatures expected during operations at rated power to avoid spurious isolation, yet low enough to provide early indication of a steam line break.
- 4) Main Steam Line High Flow Main steam line high flow could indicate a break in a main steam line. Automatic closure of vacious valves prevents excessive loss of reactor coolant and release of significant amounts of radioactive material from the teactor Coolant Pres-(p) sure Boundary. On detection of main steam line high flow, the x_ ,/ 37(U)
(G)-GESSAR (U)-Update 7.3-23 Am. No. 57, (5/81)
ACNGS-PSAR l57 f3 (V l following pipelines are isolated: a) All four main steam lines b) Main steam line drain The main steam line high flow trip setting was selected high enough to permit isolation of one main steam line for test at rated power without causing an automatic isolation of the. other steam lines, yet low enough to permit early detection of a steam line break.
- 5) Main Steam Line Low Pressure at Turbine Inlet 57 Low steam pressure at the turbine inlet while the reactor is operating could indicate a malfunction of the nuclear system pressure regulator in which the turbine control valves or tur-bine bypass valves open fully. This action causes rapid de-pressurization of the nuclear system. For part-load operating conditions, the rate of decrease of nuclear system saturation temperature could exceed the allowable rate of change of vessel temperature. A rapid depressurization of the reactor vessel while the reactor is near full power could result in undesir-able differential pressures across the channels around some fuel bundles of suffielent magnitude to cause mechanical defor-
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mation of channel walls. Such depressurizations, without ade-quite preventive action, could require thorough vessel analysis or core inspective prior to returning the reactor to operation. To avoid these time consuming requirements following a rapid depressurization, the steam pressure at the turbine inlet is monitored. Pressure falling below a preselected value with the reactor in the RUN mode initiates isolation of the following pipelines: a) All four main steam lines 44 b) Main steam drain line QO32.1 The low steam pressure isolation setting was selected far enough below normal turbine inlet pressures to avoid spurious isolation, yet high enough to provide timely detection of a pressure regulator malfunction. Although this isolation func
- tion is not required to satisfy any of the safety design bases for this system, the discussion is included to complete the listing of isolation functions.
- 6) High Drywell Pressure 57
.: High pressure in the drywell could indicate a breach of the Reactor Coolant Pressure Boundary inside the drywell. The
~s s comatic closure of various valves prevents the release of g
significant amounts of radioactive material from the drywell. On detection of high drywell pressure, the following pipelines are isolated: (U)-Update 7.3-24 Am. No. 57, (5/81)
ACNGS-PSAR a) Drywell drains discharge to radwaste b) RHR head spray c) RHR Shutdown Cooling System 37(( d) RHR shutdown cooling discharge to radwaste e) 4 RHR process sampling f) Drywell Purge System g) '37(G) Containment Ventilation System The drywell high pressure isolation setting was selected to be as low as possible without inducing spurious isolation trips. l57
- 7) Reactor Water Cleanup System High Differential Flow t
High differential flow in the Reactor Water Cleanup System ' could indicate a breach of the nuclear process barrier in the cleanup system. The cleanup system flow at the inlet to the l37(G) heat exchanger is compared with the flow at the outlet of the filter /deoiineralizer. Higher flow from the vessel initiates isolation of the cleanup system-l A
- 8) Reactor Water Cleanup System Area High Space Ambient Tempera-f ture or High Differential Temperature 57 High temperature in the area of the Reactor Water Cleanup System could indicate a breach in the nuclear process barrier in the cleanup system. High area ambient temperature or high -
differential temperature in the area ventilation system l$7 initiates isolation of the Reactor Water Cleanur System.
- 9) RHR and RCIC System Area High Space Ambient Temperature or
, High Differential Temperature 57 High temperatures in the area of the RHR System and RCIC System could indicate a breach in the nuclear process barrier in the RHR Shutdown Cooling System or RCIC System. High area ambient 57 cemperature or high differential temperature in the arec venti-lation system initiates isolation of the faulted RHR Shutdown Cooling System or RCIC System i
- 10) Condenser Low Vaccum 37(U)
Low condenser vacuum may indicate that primary coolant is being (D) ' lost through the condenser or a loss of the condenser as the primary heat sink. This condition will initiate closure of all four main steam lines and the main steam line drains.
(G)-CESSAR (U)-Update (D)- Design
- 7.3-25 Am. No. 57, (5/81)
ACNGS-PSAR [N 11) Containment Ventilation Exhaust High Radiation
\ ~ 'v) Containment Ventilation Exhaust High Radiation may indicate an accident or a leak in the Reactor Coolant Pressure Boundary. l 57 The automatic closure of the Containment ventilation isolation ,
valves will prevent the. release of significant amounce of radioactive material from the Containment environment. The magnitude and general location of a leak in the Reactor Coolant Pressure Boundary can be determined by the operator with the information provided by the leakage detection system.
- 12) Reactor Core Isolation Cooling Steam Line Low Pressure and High-Flow and High Turbine Exhaust Pressure.
Low pressure or high flow in the steam supply line to the RCIC turbine inlet could indicate the breach of the nuclear process barrier. It should be noted that the pressure sensors will be at the inlet to the inboard flow element (in the drywell) and the flow will be monitored both in the drywell and 'in the 57 auxiliary building. High turbine exhaust pressure, as measured between the turbine exhaust rupture diaphragms, could also in-dicate a loss of system integrity. Any of the above signals will initiate the isolation of this cooling system. c) Instrumentation p Sensors providing inputs to the Containment and Reactor Vessel Iso-k)) lation Control System will not be used for the automatic control of ^ the Process System. This setup separates the Protection and Process Systems. Diverse isolation initiation signals are used to increase 57 the dependability of system operation. Channels are physically and electrically separated to reduce the probability that a singic physical event will prevent isolation. Redundant channels for one monitored variable will provide inputs to the redundant division logics only through isolators. The functions of the sensors in the 37(G) Isolation Control System are shown in Figure 7.3-12. Tabic 7.3-9 lists instrument characteristics. The sensors are described in the 37(U) following paragraphs.
- 1) Reactor vessel low water level signals are initiated from four level transducers. They sense the difference between the pres-sure caused by a constant reference leg of water and the pres-sure caused by the actual water level in the vessel.
Four pairs of sensing lines, attached to taps above and below i the water level on the reactor vessel, are required for the measurement arid terminate outside the drywell and inside the 37(G) containment. They are physically separated from each other and tap off the reactor vessel at widely separated points. This arrangement assures that no single physical event can prevent isolation. {A
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j 7.3-26 (G)-CESSAR (U)-Update (D)-Design i Am. No. 57, (5/81) I l
- . .__ . _ - _ . .___.-,_,..m.
ACNCS-PSAR m 2)' Main isteam . line radiation will be monito' red by four radiation 37((
'fV h; monitors, which are described .in Section - 7.6.1.3, "Proce.as Radiation Monitoring System".
Containment ventilation radiation monitors initiate isolation signals on high radiation level. These devices are-located in $7. the exhaust l ductwork. This arrangement assures that no single physical event can prevent isolation.
- 3) : High flow in _ each main steam line is sense'd by four dif feren-
. tial pressure transducers that sense the pressure difference across the flow element in that line.
High' dif ferential flow is monitored by six flow transducers connected to three flow nozzles in the reactor water cleanup 57 system. These devices indicate the failure of the system piping integrity. i
- 4) High ambient and differential temperature in the vicinity of the main steam lines for both the RAB and TGB areas are detect-ed by dual element thermocouples located along the main steam 57
) lines and on space air handling units. The temperature
- .
, elements are located or shielded so that they are sensitive t air temperature and not the radiated heat from hot equipment 37(G) ' i The main steam line space temperature detection system is de- ! signed to detect leaks and perform isolation and alarm 57 [%\ functions. ,
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- 5) Main steam line low pressure is sensed by four pressure trans-ducers that sense pressure downstream of the outboard main d
steam line isolation valves. The sensing point is located as close as possible to the turbine stop valves. i
- 6) Drywell pressure is monitored by pressure transducers that are mounted in instrument cabinets outside the containment. In-s't rument sensing lines that terminate .a the auxiliary building connect the sensors with the drywell interior.
- 7) High ambient temperature and differential temperatures in '
- he j spaces occupied by the reactor shutdown cooling cystem piping, reactor isolation cooling syetem piping and the reactor water cleanup system piping outside the drywell is sensed by thermo-couples that indicate possible pipe breaks. Temperature sen-57 sors in the equipment area and the inlet and outlet ventilation l ducts of the RHR shutdown cooling system, RCIC system and the 4-reactor water cleanup system actuate temperature switches, which result in isolation.
- 8) There are four independent main condenser vacuum transducers -
4 for the purpose of providing an isolation signal to the main , . steam isolation valves. Each vacuum switch has its own isola-l tion valves. Each vacuum switch has its own isolation (root valve) and pressurizing source connection for testing. The (c)-cESSAR 7.3-27 Am. No. 57, (5/81)
ACNGS-PSAR
-s wiring and separation requirements for these switches are in l' '
accordance with IEEE-279. The vacuum switch setting is select-( ed so that it is compatible with safe turbine and main condenser operating at design conditions Logic The main steam line isolation (MSLI) valve closure logic is two-out-of-four. The logic for this control is shown in Figure 7.3-14. The variables that initiate automatic closure of the MSLI valves are: a) Low reactor water level b) High main steam line radiation c) High main steam line flow 37(G) d) High main steam line tunnel temperature e) High main steam line tunnel differential temperature f) Lov turbine throttle pressure in RUN Mode g) Low main condenser vacuum This same logic controls the main steam line drain valves. Other isolation valves are controlled by drywell high pressure and reactor low water level signals. In this arrangement, the pressure sensors and the water level sensors are combined in a two-out-of-two logic for the inboard valves and a two-out-of-two logic for the outboard valves. This logic is shown in Figure 7.3-12. These same pressure and water level logics are used with process radiation monitor upscale and inoperative signals to produce other isolation actions, including initia-tion of the Standby Gas Treatment System. The reactor water cleanup isolation valves are controlled by two logics, using high flow. high area temperature, high area differential temperature, and low water level signals. A Containment isolation signal will also initiate the following actions to isolate the Containment (See Table 6.2-12).
- 1) Annulus Vacuum Maintenance System a) Close inlet and outlet isolation valves b) Stop the fans
- 2) Containment Air Supply System a) Close inlet and outlet isolation valves (G)-GESSAR 7.3-28 Am. No. 57, (5/81)
ACNGS-PSAR 37(G) .
,_sq b) Stop supply and exhaust fans
]
s,,s c) Stop containment cooling fans 3)- Drywell Purge System a) Close drywell normal purge valves I 57 b) Stop drywell purge fans c) Stop drywell cooling fan p The Containment will be isolated with the closing of the pre-i . ceding valves as demonstrated in Section 6.2.4. 1 The logic that affects isolation of the Containment or reactor vessel will also af fects the following Containment Systems and
- j. their supporting systems:
4
- 1) Standby Gas Treatment "
- 2) ECCS Area Filtered Exhaust 1
- 3) Control Room Air Conditioning System The output of the logics will actuate remotely operated equip-ment such as valves, fans and dampers associated with the
- [ effective isolation of the Containment.or raactor vessel or both.
i e) Bypasses and Interlocks l An automatic bypass of the main steam line low pressure signal will 37(G) be affected in the startup mode of operation. (See isolation func- . tions and settings). i. t
- Interlocks will be provided from position switches on tLe drywell drain sumps to the Radwaste System to turn off the drywell drain sump pumps if the isolation valves close.
f) Redundancy / Diversity
- Isolation valve actuation will be accomplished by a diversity of 57 i measured variables.
g) Actuated Devices 37(G) j Table 6.2-12 itemizes the type of closing device provided for each isolation valve. To prevent the Reactor Vessel water level from falling below the top of the active fuel as a result of a pipeline break, the valve closing mechanisms will be designed 'o meat the minimum closing rates specified in Table 6.2-12. ! 1 l l (G)-GESS/.R 7.3-29 Am, No. 57, (5/81) t ww~-s- - - w - r .e- ..---.,,w.e-- ..-,..y ..w .%- .m.e-w<----, .nm.e,e-nw,,,-,oww-,,---m-- y.,-ev----,-en--
//?NGS-PSAR
(~3l The main steam line isolation valves will be spring-closing, pneuma-( tic, piston-operated valves. They will close on lo?s of pneumatic b' pressure to the valve operator. This is fail-safe oesign. The con-trol arrangement is shown in Figure 7.3-13. Closure time for the , valves will be adjustable between 3 and 10 seconds. Each valve will be piloted by two three-way, packless, direct-acting, solenoid-operated pilot valves both powered by ac. An accumulator located close to ecch isolation valve will provide pneumatic pressure for valve closing <n the event of failure of the normal air supply system. The logic elements for the instrumentation used in the systems de-scribed are highly reliable. The logic elements are selected so that the continuous load will not exceed 50% of the continuous duty rat-ing. Table 7.3-10 lists the minimum numbers of instrument channels needed to ensure that the isolation control system retains its func-tional capabilities. h) Separation Sensor devices are separated physically such that no single failure (open, closure, or short) can prevent the safety action. By the use of conduit and separated cable trays the same criterion is met from the sensors to the logic cabinets in the Control Room. The logic 37(( cabinets are so arranged that redundant equipment and wiring are not
,.~ s present in the same cabinet. Redundant. equipment and wiring may be present in control room bench boards, for separation is achieved by (d) x surrounding redundant wire and equipment in metal encasements. From the logic cabinets to the icolation valves, separated cable t' ays or conduit are employed to complete adherence to the single fai.*ure criterion.
4 i) Testability The isolation valve instrumentation is capable of complete te sting during power operation. Flow, temperature, pressure, radiat!.on, low 57 water level and condenser vacuum initiation instrument channels are tested by cross comparison between the related channels. 'Any dis-agreement between the meter readings for the channels would indicate a failure. The instrument channel trip setpoint is verified by manually superimposing a test signal and observing the channel meter and the indicator light on the output of the trip device. The logic is tested by automatic pulse testing. The radiation measuring amplifier is provided with a test switch and internal test source by which trip availability may be verified. Control Room indications include annunciation, panel lights and computer methods. In addition, the functional availability of each isolation valve may be confirmed by completely or partially closing each valve individually at reduced power using test switches located in the Control Room. 't is possible to test the MSIV closure during power operation. Test.ng can be accomplished by individually clos-d (G)-GESSAR 7.3-30 Am. No. 57, (5/81) !
#- m u. _ _ - - - + --
ACNGS-PSAR p ing the valves and monitoring the valve closure via the annunciators in the Control Room (10% and 90% closed indicators are provided). 37(c j) Manual Reset Capability Isolation Valves classified as n:,n-essent'lal, will require manual, operator initiated reset upon clearance of all automatic isolation signals. Isolation valves classified as intermediate will require 57 manual operator initiated reset without clearance of the accident closure signal, but will not be possible in the presence of a system failure signal. i f I 4 (G)-GESSAR 7.3-30a fa. No. 57, (5/81)
ACNGS-PSAR TABLE 7.3-9 ( x._,/
)
INSTRUMENTATION SPECIFICATIONS FOR CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM Trip 37(G) Isolation Function Senson Range Accuracy Setting Reactor vessel level 3 Leve 0-210 +6 in. See Figure trip transmitter in. 7.3-ile Reactor vessel level 2 Level. 0-210 +6 in. See Figure trip transmitter in. 7.3-11c Main steam line Radiation FSAR high radiation monitor Main steam line Thermocouple: Temperature 50-350 F +2 F FSAR l57 space high temperature Differential temp 50-350 F +2 F FSAR' _ l57 Main steam line high Differential 0-150 ~
+2% 120%
flow pressure psi rated transmitter flow Main steam line low Pressure 0-1500 +1% 880 psig pressure transmitter psig Drywell high Pressure 0-25 +0.05 2 psig pressure transmitter psig psig Containment ventilation Radiation FSAR l57 exhaust high radiation monitor
Condenser low vacuum Pressure FSAR l57 transmitter Alarm Only Reactor shutdown Thermoccuple: cooling system Temperature 50-350 F +2F FSAR l57 space high Differential 50 F +2 F FSAR temperature temp [57 Reactor cleanup Thermocouple: Temperature 50-350 F +2 F FSAR 57 system space high temperature Differential 50-150 F ~1 F FSAR
*The temparature alarm and/or isolation setpoint shall be established by n applicant after calculating a heat balance for the normal operational room environment, and then introducing the heat release caused by alarm and/or 57
(\.s) isolation limit leakage. This leakage limit (s) is (are) based on radio-logical release limitations. 7.3-169 (G)-GESSAR Am. No. 57, (5/81)
ACNGS-PSAR
/ individual case will be analyzed to insure that the required reliability of the protective function will not be jeopardized by adverse affects on
( ')i
'~ the accuracy, repeatability and sensitivity of the instrumentation.
7.5.1.3 Abnormal Transient Occurrences The ranges of indicators and recorders provided are capable of covering the 37( extremes of process variables and of providing adequate information for all abnormal transient events. 7.5.1.4 Accident Conditions The DBA-LOCA is the most extreme postulated operational action event. Information readouts are designed to accommodate this event from the stand-point of operator action, information, and event tracking requirements and, therefore, will cover all other design-basis events or incident require , ments. In addition, some parameters are monitored ' over ranges which far exceed those predicted for the DBA. A complete list of post-accident moni- 57 toring instrumentation is given on Table 7.5-0. Conformance to Regulatory Guide 1.97 is discussed in Appendix C. 7.5.1.4.1 Initial Ace'ident Event The design basis of all engineered safety features required to mitigate the accident event condition takes into consideration that no operator action p or assistance may be assumed for the first 10 minutes of the event. This ,g requirement, therefore, makes it mandatory that all protective action ' necessary in the first 10 minates be " automatic." 'Therefore, although continuous tracking of process variables is immediately available, no operator action based on them is required. 37G 7.5.1.4.2 Post-Accident Tracking Af ter 10 minutes, operator action is optional, based on the information available. The following process instrurentation provides information to the operator af*er a loss-of-coolant accident for his use in monitoring reactor condi-tisns and safety system actions, and is designated Type A.
- 1) Containment and drywell hydrogen concentration
- 2) Suppression Pool Temperature 57
- 3) Control Room Air Intake Radiation Level
- 4) Lake Level Tais safety-related display instrumentation will be equipped with redundant or diverse channels, will be energized by onsite emergency power supplies I and will be qualified for the accident environment. At least one channel
-O will be recorded. All meters and recorders used to display and record the 32 (j parameters listed above will be qualified to function satisfactorily after, but not necessarily, during, an SSE event without maintenance or repair. ,
f
;
(G)-GESSAR 7.5-11 Am. No. 57, (5/81)
ACNGS-PSAR I
) .,r x, Of the instrumentation lit ad above, certain instrumentation is required
( J- by the operator to determine when past LOCA conditions exist which would
\ _/c necessitate a required manual action in conformance to system criteria.
This instrumentation will be designed to conform with IEEE-279-71. The 32 instrumentation in-this category and the required manual action are as follows:
~
- 1) Containment 2H ; initiation of combustible gas control system.
- 2) Suppression pool temperature; initiation of the suppression pool cooling mode of the RHR system.
- 3) Control Room Air Intake Radiation Level; operating emergency intake and exhaust valves of the control room air conditioning system.
The instruments will be seismically qualified to function satisfactorily af ter, but not necessarily during, a seismic event without maintenance or re-calibration. The remaining post accident monitoring instrumentation will meet the follow-
, ing criteria:
32
- 1) General Functional Requirements The nuclear power generating station Post Accident Monitoring system shall function with precision and reliability to continuously i
[),/ (, display the appropriate nonitored variables. This requirement shall apply for the full range of conditions and ' performance associated with design basis events. Regulatory Guide 1.97 is addressed as 57 described in PSAR Appendix C.
- 2) Single Failure Criterion The accident monitoring system should be designed with redundant channels such that the loss of a single monitoring channel should not prevent the operator from determining the nature of an accident, the functioning of the engineered safety features, or the need for operator action and the response of the plant to the safety measures in operation. In lieu of redundancy, diverse channels (e.g.,
saturated pressure, saturated temperature) or several channels measuring the spatial distribution of a parameter (e.g. , containment temperature distribution) may be utilized where the loss of one channel will not mean the loss of all parameter information. Where 32 redundancy is used, one channel of each redundant set of channels shall be recorded and powered fro ~ the station d.c. system. khere other arrangements are used in lieu of redundancy, each channel shall be recorded and energizea from the station d.c. system. t / Ns/ 7.5-12 Am. No. 57, (5/81)
l l ACNGS-PSAR
- a. The isolation valve position lamps indicate valve closure.
. j' \ The power source is the same as for the associated motor
{) . operator. These lamps are ~ on the RCCS Benchboard.' )
- b. ' The main steam line flow indication will be downscale.1 This information .is provided on ' the PPC. The - power . source ' is instrument AC from the Class IE Power . System buses.
- c. Annunciators for the containment and reactor vesni isolation
- system variables and trip logic will be in the tripped state.
These annunciators are located on the RCCS Benchboard. The power source .is DC from the station battery. E
- d. Process Computer logging of trips. '
- 3) ' Operation of the emergency core cooling and the RCIC systems
;
'following the accident may be verified by observing the following indications, all of which are on the RCCS Benchboard:
- a. Annunciators for high pressure core spray, low pressure ~ core
- spray, residual . heat removal, automatic depressurization system,
- and reactor core isolation _ cooling system sensor initiation 37(g) logic trips will be actuated. The power source is DC from a station battery.
- b. Flow and pressure indications for each emergency core cooling A system are provided and are operable before and af ter an SSE.
j-( The power sources are independent and'come frca the same Class IE Power System buses as the driven equipment.
- c. RCIC isolatian valve position lamps will be indicating that these valves are open. The power source is from the same bus that powers the valve operator.
l~ d. Injection valve position lights will be indicating that the
- t. valves are either open or closed. The power source is the same l - as for the valve operator.
, e. Direct indication of relief valve position is provided by light , matrices, which are actuated by pressure measurement in the re-lief discharge pipe. The power source is Class IE. 57
- f. Relief valve . position may be inferred from reactor pressure indications (also shown on the PPC). The power source is from the Class IE Power System buses.
- g. . Relief valve discharge pipe temperature monitors will be responding to temperature changes. The powr.r source is from an instrument AC bus.
1
- h. Process computer logging of trips in the reactor core cooling network. Power source is the computer power supply from the i plant ' auxiliary AC power.
I 7.5-18 (G)-GESSAR Am. No. 57, (5/81)
ACNGS-PSAR
'f) Temperature alarms and indication for all air handling unit heating coils discharge
. g)' Outside air intake radiation indicator and alarms 37(U h) Outside air intake and return air smoke detector alarm i) High chlorine concentration alarm
- Transfer from the operating emergency filter train will be automatic upon loss of flow. The operator will be able to manually transfer emergency filter trains upon indication of malfunction within the operating train.
The emergency filter trains, will not normally be in operation, but can be tested during normal plant operation. The display instruments and controla provided are listed in Table 7.5-6. Also shown are the types of components and their locations. As the detailed system design is not yet. complete, only preliminary data is provided. The a completed table and control panel layouts will be presented in the FSAR. 7.5.1.4.2.10 ECCS Area Fan Coolers The ECCS Area Fan Coolers are provided in areas where ECCS pumps and equip-mant are located. (See Sections 7.3 and 9.4 for system description). The operator has available the following display instrumentation and con-p trols for the ECCS Area Fan Coolers: a) In the Control Room there will be area temperature indications and l area low flow alarms , b) The operator will have control switches for each of the fan cooler motors in the Control Room. An area fan will start automatically when associated ECCS equipment starts, or it can be started manually from the Control Room. The display instruments and controls are listed in Table 7.5-7. Also shown are the types of components and their locations. As the detailed system is not complete, only preliminary data is provided. The completed table and control panel inputs will be provided in the FSAR. 7.5.1.4.2.11 Drywell and Containment Conditions Drywell and containment conditions are indicated and/or recorded by the 37(G) instrumentation described below, a) Drywell and Containment Pressure Monitoring There are two containment and two drywell pressure monitoring chan-nels. One of each is used to cover a pressure range during normal reactor operation, and the other two are used to cover the pressure range following a loss-of-coolant accident. These devices will g/ . l 57 (G)-GESSAR (U)-Update 7.5-22 'Am No. 57, (5/81)
ACNGS-PSAR provide a continuous readout of containment pressure in the control 57 4
/ i room. The instrumentation consists of tour separate transmitters
( /
! and four recorders with adjustable alarr points. These recorders are mounted on the Standby Information Panel. In addition to the above four sensors, there are four channels of extended range trans- 5; mitters are required by NUREG 0718 (II.F.I).
b) Drywell and Containment Temperature Monitoring The drywell temperature is monitored by two tenperature channels. One is an indicating channel and one is a recording channel. A full range adjustable alarm is provided in the main control room. The suppression pool temperature is monitored at five locations by four sensors (for a total of 20 units). (These sensors are broken down into four safety channels (5 sensors per channel). Sensors are monitored individually and as an average temperature per safety channel). These temperatures are recorded and indicated in the Control Room. The indicators for these temperatures are mounted 57 on the Post Accident Monitoring Panels (P880, 885, 892 and 894) and Auxiliary Control Panel (P008). c) Suppression Pool Water Level , The suppression pool water level is monitored by four sensors. All four of these provide indication and recording functions and logic inputs for upper RCB pool dump initiation (on low low level), and {S -) g two of four provide alarm functions (high and low level). The indi- 57 V cators are located on both the BOP Auxiliary Control Panel (P800) and Post Accident Monitoring Panels (P880 and P885). Annunciation is on the BOP Auxiliary Control Panel (P800). In addition to the above four channels of extended range, sensors from the top of the ECCS suction to 5 feet above normal suppression pool water level will be provided. d) liydrogen Monitoring
- 1) Initiating Circuits Eight hydrogen sempling locations have been selected for post LOCA detection of hydrogen in the containment and the dryvell. 57 The locations were determined by considering two .nodels. The first model assumes hydrogen diffusion to be identical to neutron diffusion. Buoyancy effects were neglected in applying isotropic diffusion. The use of this model yielded locations above the suppression pool and at the bottom of the drywell.
The second model considers the effects of free convection. The buoyancy forces lif t the hydrogen from the lower regions of the containaent and drywell to higher regions. The influence of trapping was also considered. This model provided five loca-tions: 1) The top of the containment, 2) Near the top of the
! \
7.5-23 Am. No. 57, (5/81)
ACNGS-PSAR
,-- s pressure vessel, 3) Top outside of drywell, 4) Top outside of l ) drywell (opposite), 5) Near the Reactor Water Cicanup filter / \s_ / -demineralizer area. Both models which provided the sampling 57 locations assumed that no mechanical mixing occurred.
Each redundant hydrogen monitoring system consists of sample and return lines, isolation valves, hydrogen analyzers, grab sample cylinder and sample pumps. The equipment excluding the isolation valves and piping is located _in the reactor auxiliary building. Each sample location is provided with a separate sample line to each analyzer system. The hydrogen concentra-tion is determined in the analyzer and the volume percent is recorded in the Control Room. The analyzer has a range of 57 0-10 percent hydrogen with an accuracy of + 3.0 percent of full scale. The analyzer will be capable of continuous operation in the environment it will be located in. The concentration is recorded during sampling and an. alarm is actuated if the concentration at any sample point exceeds 3.0 volume. A block diagram of the Hydrogen Monitoring System is shown in Figure 7.5-9. Sample points are shown in Figure 7.5-9a. Applicable design codes and standards are indicated on Figure 57 7.1-2.
- 2) Redundancy The hydrogen monitoring subsystem consists of two identical
{Q'} sampling and analyzer trains. Each train can monitor any of the eight sampling points. This system is designed to withstand a
- single active failure and perfonn its safety funct ion.
- 3) Separation Each analyzer is physically separated and is powered from a 1 separate emergency Class 1E power source. 57
- 4) Hydrogen Monitoring Test and Calibratian Low concentration H2 and N2 mixtures cat! be introduced into each analyzer for zero adjustaent and scale calibration to provide greater accuracy. The calibration can be completed from the control room.
- 5) Operational Considerations
, The hydrogen subsystem shall he manually activated on the 37(U) occurrence of a loss-of-coolant accident and shall remain enabled after initiation unless turned off with a key - locked switch. Hydrogen concentration shall be recorded from 0 to 10 percent with an accuracy of + 3 percent of the full scale. An alarm is activated when any sample point
-~s indicates above 3 volume percent. No action outside the 57 Control Room is necessary for system operation.
[V) i l 7.5-24 Am. No. 57, (5/81) l
- _ . - . . . . _ - - ._. _. __ - . _ ._, . ,. .l
-% T f
l ACNGS-PSAR
- 6) Environmental Considerations
, V '1he hydrogen monitoring equipment .is located in the Drywell, Containment, Auxiliary Building and Control Room. Equipment i located in the Drywell and Containment is designed to remain - L functional in the environment resulting from a loss-of-coolant , l accident. See Tables 3.11-1 through 3.11-5 for a description t of the Containment, Drywell and Control Room environments. ! i I i t i i i i l l l. i 5
) i i l 1 -i i i I i i i , i i
- 7. 5-24a Am. No. 57, (5/81) e----<,,-e-----..we,-,,-----r--w=.<- , . , .-*mwr,..we.w r e - ww -,w e- . m y w- e
ACNGS-PSAR
'~'N The following Nuclear Boiler instrumentation shall be provided on the remote shutdown control panel:
a) Level indicatar R010 reading from level indicating transmitter switch N026D (LITS). b) Reactor pressure tranmaitter N006 connected to feedwater level reference condensing chamber D004B. 7 ) c) Reactor pressure indicator R011 reading from pressure transmitter N006. d) Power supply for pressure transmitter and level indicating trans-mitter switch. 7.5.1.5.1.5.4 Recirculation Flow Control System The following valve shall have remote shutdown control: F023A - Motor operated valve (recirculation pump suction) 7.5.1.5.1.5.5 Balance of Plant Systems BOP systems instrumentation and control devices necessary to support the shutdown of the reactor will be installed on the Remote Shutdown Panel.
, -- Manual Initiation for starting and loading the Diesel Generators and closed
( cooling Water Systems will be part of the Remote Shutdown Panel. 37(U),
(D) The Remote Shutdown Panel layout will be presented in the FSAR. 7.5.1.6 Safety Parameter Display System A Safety Parameter Display System (SPDS) will be provided. It will contain the minimum parameter set from which plant safety status can be determined so that the operator can determine plant safety status quickly and readily at a single display location. It is not intended to provide full problem diagnostic capability, but rather serves as an indicator that the plant is either in a safe condition or that an off-normal condition exists and fur-ther action should be taken to identify and correct it. The set of para-
$7 meters to be displayed on the SPDS will be those determined by the ongoing ef forts of the' BWR Owners Group when approved by the NRC. The plant functions to be presented on the SPDS will include, but not necessarily be limited to:
- Reactivity Control - Radioactivity in Containment
- Reactor Core Cooling - Containment Integrity
- Reactor Coolant System Intagrity
! Trending display of process parameters will be available on the SPDS. The SPDS will be displayed in the Main Control Room on a dedicated CRT on pg one of the f ront row panels, which makes it readily accessible and max-( ,) imizes its usefulness to the operator. The SPDS will also be displayed on a dedicated CRT in a Technical Support Center. Capability to display the (G)-GESSAR (U)-Update (D)-Design 7.5-32 Am. No 57, (5/81)
'ACNGS-PSAR
, 'SPDS in the ' EOF will be provided.
$ Class IE instruments serve as backup to 'the non-IE SPDS are availabe in the Main Control Room. The backup instruments are not arranged in an identical grouped' fashion as the SPDS, but are located such that they are useful' for operational and post-accident tracking and control, . as will be verified by j the Control Room design review. j An availability design goal of 99% when the reactor is above the cold shut-down. status is established for the SPDS, including the TSC display. An availability goal of 80% is ' set for when the reactor is in the cold shut- : down or refueling conditon.- A technical Specification of SPDS avail- i ability 'is not deemed necessary since the backup displays are available in well designed locations in the Main Control Room.
7.5.2 ANALYSIS t
7.5.2.1 General I The safety-related and power generation display instrumentation provides adequate information to allow an operator to make correct decisions as
~
bases for manual control actions permitted under normal, abnormal trans- ! ient, and accident conditions. The use of the Nucienet design will in-l prove the availability of the plant by providing the operator with more readily accessible information and control of the various plant operational
- parameters. This is accomplished by the logical organization of functional
] plant system indicators, displays, controls, and a computer display system into a human-engineered operator interface. The implementation involves
- the use of five modular console / panel /benchboards. The Nucienet design includes the PGCC control room design (Fe!. NEDO-10466).
37(G) A complete description and analysis of, and design criteria applicable to the specific hard-wired indicators, displays, and controls, for the various safety-related systems, are described elsewhere in this chapter with the systems theyiserve. Redundancy and independence or diversity are provided ] in all of those information systems which are used as a basis for operator- ' controlled safeguards action. The complete failure of the Display Control System, which serves as an ac-tive part of. the operator / plant interface, does not degrade the quantity or ! quality. of necessary information, presented by hard-wired devices, needed I to determine the status or action of plant safety systems. Some safety-l related process information is displayed and/or analyzed by this non- ' safety class Display Control Syste, as well as by the conventional hard-wired instruments. In all cases where safety-related information is shared this way, the DCS is isolated from the safety-related circuitry so that no DCS failure can inhibit or affect that circuit. i 7.5.2.2 Normal Operation Subsection 7.5.1.2 described the basis for selecting ranges for instru-mentation and inasmuch as abnormal transient or accident conditions mon-itoring . requirements exceed those for normal operation, the normal ranges s are covered adequately. 7.5-33 CESSAR I Am. No. 57, (5/81) _ _ . . _ . _ ~ - - - _ _ , _ . _ , . _ _ _ . . _ _ _ - . .
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AONGS-PSAR-i I TABLE 7.5-0 (Cont'd) ' i VARIABLg PANEL VARIABLE TYPE RANGE CATEGORY - NUMBER COMMENTS i Ib Level of A 134'-137'4" 1 P800-Div. 1 j; Suppression Pool P800-Div. 2 P880-Div. 1 [ ] P885-Div. 2 l 1 lc Radiation Monitoring A 5x10 1 P800-Div. 1 At Inlet To Control 5x10-3 Ci/cc P800-Div. 1 Bldg. P800-Div. 2 P800-Div. 2 1 - P844-Div. 1
- P844-Div. 1 P867-Div. 2-
."~ P867-Div. 2 57
~ Y' P831-Div. 3 P831-Div. 3 P832-Div. 4 P832-Div. 4 ) i ld Level of Lake A 1 P880-Div. 1 l P885-Div. 2 l P892-Div. 3 i~ P800-Div. 1 'i 4 P300-Div. 2 ! F t , u, j e
- 5 i .
(
._ . _ _ _ _ _ _ _ _ _ _ _ . . . _ _ . ~ . . . _ . . _ _ _ _. __ . _ _ _ _ . . . _ . _ _ . _ _ _ _ _ _ _ _ _ , . . _ . . . . .
ACNGS-PSAR TABLE 7.5-0 (Cont'd) VARIABLE PANEL
. VARIABLE TYPE RANGE CATEGORY NUMBER COMMENTS 2 -Neut ron Flux B 10-Sto 100% 1 P680-6C Full Power P680 SRM & APRM P669 P670 P671 P672 t
3 Cont rol Rod Posit ion B Full in or 3 P680-050 Mimic Not full in 4 RCS Soluble B 0-1000 ppm 3 - Grab Sample & Anslysis - See It em 7 Boron. Conc. II.B.3 m , h5 Coolant Leve l in Reactor B Bottom of Core support 1 P601-20B Div. 1 57 i plat e to lesser of top P601-17B Div. 2 of vessel or cent er of P892 Div. 3 main steamline , P896 Div. 4 7* RCS '.'ressure B 0 psig-1500psig 1 P601-20B Div. 1 P601-17B Div. 2 P892- Div. 3 P896 Div. 4 e b O
j t
\,_/).
4 ACNCS-PSAR l- TABLE 7.5-0 (Cont'd) T 4
'v mIABLE PANEL I
-VARIABLE TYPE RANGE CATEGORY- NUMBER COMMENTS
- 4
- 8 Drywell Pressure B 12 psia - 45 psia. 1 P601-17B Div.11-(110% design) P601-20B Div. 2 PS92 Div. 3 P896 Div. 4 9 Drywell B Bottom to Top 1 P601-22B LP-S75-12 LP Sump; High Purity Drywell
- Sump Level P300-Div. 1 Drain Tank "
P800-Div. 2 l 10 Primary Cont ainment B 10-80 psia 1 P601-208 Div. 1 Pressure P601-17B Div. 2 - P892 Div. 3 i u
'
- PS96 Div. 4 vi 57' i E P868
! g 11 Primary Cont ainment B Closed or not closed 1 i Isolat ion Valve i Position (excluding check valves) P869 See not e on 1.9 7-4 Para. 1.3.1B 12 Radioact ivit y C 1/2 Tech Spec 1 P870-60C _ Handled by CRT l Concentration or Limit to 100 times Rad Monitoring Radiation Level Tech Spec Limit Panels I in Circulat ing in R/hr a Primary Coolant 2 a , o
' .U
V 6
. - .=
_ . ~ . . . . - _ . . ._ _ _ __ _ _..m. . . _ .. . _ . - _. _ . . _ . _ _ _ ___ . . _ . . _ _ . _ . . _ _ _ . . _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ ._. _ ...
/"'N .
s.- ACNGS-PSAR TABLE 7.5-0 (Cont'd) I VARIABLE PANEL TYPE RANGE CATEGORf NUMBER COMMENTS ! VARIABLE 13 Analysis of C 10 Ci/gm to 3 - Grab Sample & Analysis; See Item Primary Coolant 10 Ci/gm(or II.B.3. (Gamma Spect rum) TID-14844 Source term in
- coolant volume) 15 RCS Pressure C Previously listed as Variable 7 i
15 Primary Cont ain- C 1 R/hr t o 10 5 3 P870-60C ' Previously list ed as Variable 9 j ment Area Radiat ion R/hr 17 Drywell Drain Sumps C See also Variable Ib l y Level (Ident ified & Unidentified Leakage) 57L l s~ Suppression Pool Bot t om Of ECCS Suct ion P-800-Div. 1 Refer to previous list ing; Variable S , j' 18 C F-800-Div. 2 Water Level line t o 5 f t above .. normal wtr Ivl P-800-Div. 3
- P-800-Div. 4
- 19 Drywell Pressure C Refer to previous list ing; Variable 7 20 RCS Pressure C Previously . list ed as Variable. 70 -
1 j
.a
, .N or ' i :: i i 1 i
-s s
- j. ACNGS-PSAR J
TABLE 7.5-0 (Cont'd) .
I VARIABLE PANEL VARIABLE TYPE RANGE CATEGORY NUMBER COMMENTS 1 21 Primary Cont ain- C 10 psia to 3x 1 ! ment Pressure design pressure for concret e 4X design pressure 1 for steet 22 Cont ainment C 0-30% 1 P871 Recorder
- and Drywell (Capabilit ies of P872 Recorder
, H2 Concentration operat ing from
- 12 psia io design l press)
)
." 24 Cont ainment C See Table 12.2-Sa
) d Y E f fluent Radioacti- i ! E vit y 57 Noble Gases (from , ident ified release i points including 3 standby gas ! Treatment Syst em I Vent) l 8 1 if i j 5 . i n
- vi
- ,
e i 1
_____;____._--___--_________.._. . _ . .. . _ . _ . _ . . .. ._ _ . _ . . - , ,__.; _- . D l , s
ACNGS-PSAR f f ' TABLE 7.5-0 (Cont ' d) i
-VARIABLE PANEL TYPE RANGE CATEGORY NUMBER COMMENTS
, VARIABLE
- 26 E f fluent Radioact ivit y - C See Table 12.2-5a 27 Main Feedwater Flow D 0-110% 3 P680-03D design flow 28 Condensate D Bot t om t o Top 3 (Lat e r) j .
Storage Tank Level . 30 Drywell Pressure D Previously' listed as Variable 8 f i i 31 Suppression Pool D Previously listed as Variable 18 Water Level T '32 Suppression Pool D Previously list ed as' Variable la 57. i } Wat er Temperat ure j
[ 40-4400F 2 P871 33 Drywell Atmosphere D , P872 Temperat ure 0-15" w.g. 2 P501-19B 35 Main Steamline D isolat ion valves 0-5 psia 1eakage cont rol
! system pressure I
i E l .U a . C 4 i e 1, .
/""N s -
U .
- - ACNGS-PSAR
- . TABLE 7.5-0 (Cont 'd)
~
i , VARIAB LE PANEL VARIABLE TYPE RANGE CATEGORY NUMBER COMMENTS i 36 Primary System D Closed - 2 P601-19C j Sa fet y Relief not closed P601-19C , l Valve Posit ions, P601-19C including ADS or P601-19C flow through or P601-19C pressure in valve P601-19C 4 lines P601-19C P601-19C I P601-190 i P601-19C i P601-19C P601-19C ] y
- P601-19C i
l Y P601-19C 57 P601-19C j h P601-19C l ! P601-19C P601-19C ! l P601-19C i 4 I i 39 RCIC Flow D 0-110% design 2 P601-21B l i j 40 HPCS Flow D 0-110% design flow . 2 P601-16B i
.E e
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~
1
_ . . = _ _ . _ _ _ _ _ . _ _ _ _ ._.___ ._.. _ _ _ _. _ _ _ N. j ACNCS-PSAR TABLE 7.5-0 (Cont 'd) VARIABLE PANEL VARIABLE TYPE RANGE CATEGORY NUMBER COMMENTS 42 LPCI System D 0-110% design 2 P601-178 Flow flow P601-20B 43 SLCS Flow D 0-110% design 2 P601-198 flow 44 SLCS Storage D Bot t om t o 2 P601-19B , tank level Top 45 RHR System D 0-110% design 2 P601-20B Flow flow P601-20B
-a P601-20B j [n P601-17B 5 P601-17B 57
$ P601-17B l
46 RHR Heat D 32-300F 2 P601-17B Heat Exchanger Exchanger Out let B001C Temperature B001A B001B ' B001D
.s i ,
(N i Es
- 0; l
) i i t i t _ _ _ _ _ _ _ _ _ i--
, - - - , . r _- _ _ _ _ _ _ _
% 3 ACNCS-PSAR TABLE 7.5-0 (Cont'd)
VARIABLE PANEL RANGE CATEGORY NUMBER COMMENTS VARIt.BLE TYPE 32-2000F 2 P880 47 Service Water D P835 Temperature (Cooling ' P892 Water Temp to ESF Syst em Component s) 0-110% 2 P601-22B 48 Service Water D P800 Div. 1 Flow (Cooling
; Water Flow to ESF P800 Div. 2 P800 Div. 3 System Component s)
Top to Bottom 3 Lat er' Tank # 49 Liquid Radwaste D IC-T1-Il , w Co11ect ion Tank IC-T1-12 v, (High Radioactivity IC-T3-ll 57 E Liquid Tank) Level IC-T2-11 d HP-T2-11 l HP-T2-12
- HP-T1-11 i LP-TI-l1 i LP-T1-12 LP-T3-11 I. LP-T2-11 DD-TI-12 DD-T1-11 f DD-T2-12 DD-T2-11 j
1 ) ^ S a 1 f 4
_- - - . - - . . . . - . - - , . _ . . . . . ~ . t N f J ACNGS-PSAR TABLE 7.5-0 (Cont'd) i i VARIABLE PANEL VARIABLE TYPE RANGE CATEGORY NUMBER COMMENTS. 4 50 Emergency Ventilat ion D Open-Closed 2 P363 Applicable it ems are all' indicat ed j Da. aper Posit ion Status P864 on these panels P847 I P848 4 P847 i
. P847 I
P847 )- P847 P847 I P847 ; i u P847 i i vi P847 i 5.' P847 57 E P847 l P847 P847 P847 P847 P847 P847- ! P847
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ACNGS-PSAR l TABLE 7.5-0 (Cont'd) VARIABLE PANEL TYPE RANGE CATEGORY NUMBER COMMENTS. VARIABLE 2 P870 Voltage j 51 Stat us of D Voltage, Standby Power Current, Pressure P870 ~ Volt age j
& other energy P870 Current sources import ant P870 Current
! to safety P870 Current P870 Voltage i hydraulic, pneumat ic Volt age P870 i i P870 Current P870 Current P870 Current I P877 Volt age ! P877 Volt age 2 4 P877 Current-y' P877 Current j; P877 Current 37 8 P877 Voltage P877 Current P877 Voltage . 52 Drywell E 1 R/hr to 1 P870 See Appendix 0, It em II.F.3 (Primary Containment) 107 R/hr P844-Dts. 1 Area Radiation P867-Div. i i
! P831-Div. 3 i
P832-Div. 4 6 l g i 1 g: e I g i 5 s b l
= . . . - - - - . . - . . . . . - -
\ N, s
P j ACNGS-PSAR TABLE 7.5-0 (Cont 'd)
4 VARIABLE PANEL
; VARIABLE TYPE RANGE CATEGORY NUMBER COMMENTS i "l b
i 53 Cont ainment E- 1 R/hr to 1 P870 , See Appendix 0, It em II.F.3 (Secondary Containment) 107 R/hr P844-Div. 1 Area Radiat ion P867-Div. 2 i P831-Div. 3 i P832-Div. 4 i i 55 Airborne dadioactive E See Table , Mat erials Released 12.2-5a From Plant > l 56 Environs Radiat ion E 3 - Sampling and analysis capability y and Radioactivity Meteorology E 3 P895 Refer to Regulatory Guide 1.97-15 for l 457 j .p
specific details on requirements and 57 ranges l 58 Accident Sampling E 3 - See It em II.B.3 Capabilit y i i I
- Discont int. t y in variable number indicat es t hat an it'em from Regulatory Guide 1.9'i, Table 1 is not list ed.
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ACNGS-PSAR l LIST OF EFFECTIVE PAGES I CHAPTER 12 RADIATION PROTECTION Page, Amendment do. 1* 57 2* 56 2a* 57 3*' 45 4* 37 5* 37 i 37 ii
- 46 iii 37 iiia 37 iv 37 v 37 vi 37 12.1-1 57 12.1-la 57 12.1-lb 57 12.1-2 35 -
12.1-3 41 pg 12.1-3a 41 12.1-3b
-Q 12.1-3c 41 41 12.1-4' 41 12.1-5 35 12.1-6 39 12.1-7 37 12.1-8 35 12.1-8a 35 ; 12.1-8b 35 12.1-9 35 12.1-10 -
12.1-11 39 12.1-12 35 12.1-12a 35 12.1-13
- 35
! 12.1-14 22 12.1-14a 35 1 12.1-15 35 12.1-16 35 12.1-17 35 i 12.1-18 35 12.1-19 41 ! 12.1-19a 41 12.1-20 45 12.1-20a 45 12.1-21 45
- Effective Pages/ Figures Listings i
, 1 Am. No. 56, (2/81) 2 _ . . - _ _ . - _ _ _ _, . _ - . . . _ _ . . _ _ _ . . . . . . _ _ , . , _ . . . . . . _ - . , _ _ _.-____..___..m...
- ACNGS-PSAR LIST OF EFFECTIVE PAGES (Cont 'd)
- m CHAPTER 12
, h Amendment No. l 12.1-50 45 12.2-1 43 12.2-la 43 12.2-lb 46 , 12.2-2 41 12.2-3 37 12.2-4 39 1 12.2-5 39 . 12.2-Soa 39 l 12.2-Sa 57 i 12.2-6 17 12.2-7 17- ). -12.2-8 42 4 12.2-8ca 42 12.2-8a. 39 i 12.2-8a(i) 35 12.2-8b 35 12.2-8e 17
- 12.2-9 1 1 12.2-10 -
! 12.2-11 - 12.2-12 - 12.2-13 - 12.2-14 35 12.2-15 35 i 12.2-16 46 1 12.2-17 35 12.2-18 21 12.2-19 35 i 12.2-19a 57 j 12.2-20 33 i I 4 l' I I 1 2a Am. No. 57, (5/81)
ACNGS-PSAR 12.0 RADIATION PROTECTION
) 12.1 SHIELDING v
12.1.1 DESIGN OBJECTIVES The primary design objective of the plant radiation shielding is to protect plant operating personnel and the general public against radiation exposure from the reactor, power conversion, auxiliary, and waste processing systems during normal operation, including anticipated operetional occurrences, postulated accident conditions and maintenance. This objective will be accomplished by designing the shielding to perform the following functions: a) . Limit inplant exposure to radiation of plant personnel, contractors and authorized site visitors to as far below the limits set forth in 1 10 CFR 20 as practicable for normal operation, including anticipated lQ12.;
. operational occurrences and maintenance in conformance with Regulatory Guide 8.8. I 3 lQ12.1 b) Limit radiation exposure of Control Room personcel, in the unlikely event of an accident to allow habitability of the Control Room, as specified in 10 CFR 50, Appendix A, Criterion 19 c) Limit exposures to the general public offsite from direct and air-p)
( V scattered radiation to a small fraction of the limits set forth in 10 CTR 20 during normal operation and anticipated operational occur-rences, and to within the limits specified in F1 CFP.100 for postu-lated accident conditions d) Provide barriers for restricting personnel access to high radiation areas and for controlling the spread of contamin ..v.- The plant ra-diation shielding will also be designed to protect certain plant components from excessive radiation damage or activation. e) Provide adequate access to areas designated as requiring access for post-accident operation and ensure that radiation limits for safety- 57 related equipment are not exceeded in the event of accident. To accomplish this objective, the plant shielding will function to: a) Limit radiation heating of bulk structural concrete, such as the drywell shield wall, by the use of the reactor shield wall b) Reduce neutron activation of equipment, piping, and other materials such as by the use of the reactor shield wall c) Lim. radiation damage to equipment and materials to below the sper ic Integrated Life Dose limits. To comply with the above objectives, the plant shielding is designed to f a*1enuate radiation levels throughout the plant, from direct and scattered (v f radiation to the dose limits specified in Table 12.1-1. l 12.1-1 Am. No. 57, (5/81) 1
.~ . . . -. -- -. - . - -
-ACNGS-PSAR In part the criteria which will be used in determining shielding require-
[]. l ments for pumps and valve galleries will be the following: ,' a) The dose rate at one foot from the equipment in question b) The annual exposure time anticipated for personnel with respect to the specific equipment. Exposure time will be classified according to the modes.of activity of the exposed individual rela-tive to the equipment during the exposure period. These are:
- 1) Function or operation of equipment f
- 2) _ Control _or surveillance of equipment
- 3) Maintenance of equipment
- 4) Incidental or background radiation dose due to equipment 17 Average exposure distances will be determined for each of the exposure Q2--
time modes and the resultant dose rates esiculated for each distance 14* using their respective dose rates at one foot away. In conjunction with these dose rates, the dose contributed by radioactive fluid in connecting pipes will be added. It should be emphasized that normally the pipe dose contributions easily exceed those of the pump or valves. At the present time the following guidelines are applicable respecting shielding requirements of pumps and valve galleries, Fut they are by no means exhaustive: i C l a) All pumps and valve galleries involved . che transmission of fluids of reactor coolant nuclide concentraticas, reactor steam nuclide concentrations, and condensate nuclide concentrations prior to demineralization will be shielded. b) All pumps and valve galleries involved in the transmission of fluids of condensate nuclide concentrations after demineralization will generally not be shielded. c) Notwithstanding paragraph b) all pumps and valve galleries involved in the transmission of fluids which generate significant crud build-up, plate-out, or settling deposition either as the fallout of functional use or as an adventitious fallout will be shielded; in particular, functional process clean-up conditioning and removal equipment such as spent resin lines, backwash tank lines, concen-
- tration bottoms lines, filter sludge transfer lines, and the like will be shielded, whereas lines which could generate adventitious crud formation will be allocated space for shielding in the event that cr'ud build up or other fallout produces unacceptable dose rate levels. .,
'i A radiation and shielding design review will be performed for the spaces around systems that as the result of an accident, contain highly I radioactive material in compliance with the requ rements of NUREC-0660 57 [ { 12.1-la Am. No. 57, (5/81) f -k u . , - - ,,,,e,,, ,,,<-,w , , , , ,, ,.- - . - - ,-,,-.-,-----vm- - , - - .,~,+,w---v4--.- av-
ACNGS-PSAR i item II.L.2. This review effort will assume post-accident releases of radioactivity equivalent to Regulatory Guide 1.3 and identify the g resultant levels of radioactivity in all systems external to containment that have the potential for containing radioactive fluid as a result of the accident. The activity in containment and in these systems will be utilized to identify the doses associated with the continuous occupancy of the control room and the technical support center and with the infrequent 57 occupancy of areas requiring identified operator actions (i.e. post accident sampling). The activity will also be.used as a basis for the qualification of safety-related equipment. 12.1.2 DE'AGN DESCRIPTION The plant will be divided into radiation access zones, based on the max-imum zone dose rate levels listed in Table 12.1-1. The anticipated access
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12.1-1b Am. No. 57, (5 /81)
ACNGS-PSAR1
,-Q
).
%/ -a) To alert personnel in areas in which airborne radioactive contamina-tion is suspected of being equal to or greater than setpoint levels.
Set point levels.are-established with tne intention of limiting concentrations to those set forth in 10CFR20, Appendix B, and the guidelines established in (NBSH) National Bureau of Standards lland-book # 69, in order to reduce occupational exposure via the inhala-tion pathway to as low as practicable.
-b) The system will provide a continuous record of airborne radio-nuclides in the form of particulates, iodine and gases. Records
.will be maintained in accordance with Section 20.401 of 10CFR20, which requires each licensee to maintairs records of radiation exposures to individuals for whom personnel monitoring is required under Section 20.202 of 10CFR20.
c) Concentrations exceeding limits established in Part a and with due consideration to dilution from other air discharge inputs will be a necessary condition for exceeding MPC within an equipment space discharging into an exhaust header monitored by an airborne unit ~, but not a sufficient condition. The unit, upon alarm will require - the initiation of a more comprehensive survey of the area serviced by said unit. d) In the case of the units servicing the outside air intakes of the control room complex, the system will initiate closure of the untreated air straam and divert intake air through filter units. e) To provide extended range noble gas monitoring and particulate and iodine sampling capability at potential accident release points. 12.2.4.2 System Description The exhausts listed on Table 12.2-Sa are considered to be potential release points for accident monitoring purposes. The 3 stage continuous Sir monitors (see Section 12.2.4.2.1) at these points are provided with ex-tended range noble gas monitors, as shown on the table, and with post-accident iodine and particulate sampling capability. These monitors are all of fline, and utilize isokinetic nozzles for sample extraction from the exhaust point. These monitors are of high quality, and are powered from 57 a vital instrument bus. Real time and trend display of the release from each point is available on demand in the form of equivalent Xe-133 con-centration at the radiation monitoring console, and a trend alarm is provided for each release point.
'Other airborne radiation monitoring locations are listed on Table 12.2-5b.
12.2.4.2.1 Three Stage Continuous Air Monitor Figure 12.2-1 shows a typical 3 stage continuous air monitoring unit. Each three stage continuous air monitoring unit will consist of particulate C/ iodine, 12.2-Sa Am. No. 57, (5/81)
C *% i V V ACNGS-PSAR TABLE 12.2-Sa
~
ACCIDENT RELEASE MONITORING POINTS Noble Gas Monitor
- Release Point Areas Served Range (uCi/cc)
- 1. Personnel Access Building o PAB potentially hot areas, such as 10 102 (PAB) Filtered Exhaust Decontamination Area, Sample Analysis / Room & Hot Machine Shop
- 2. Radwaste Building Exhaust o Entire Radwaste Building 10 102
- 3. Fuel Handling Building o Entire-FHB (except under fuel 10 102 (FHB) Exhaust handling accident conditions)
- 4. Plant Vent Stack o Containment Ventilation Exhaust 10 103 57 4
ro o Turbine Building Exhaust i e o Reactor Auxiliary Building (except ECCS areas under accident conditions and switchgear areas)
- 5. ESF Exhaust o Standby Gas Treatment System Exhaust 10 105 o ECCS Area Filtered Exhaust z o Annulus Vacuum Maintenance System
.' Exhaust u
." o CRDM Facility Exhaust Y
v 1 Notes: (1) Ranges are selected in accordance with NUREG 0737 Table II.F.1-1. j l (2) All accident release points have particulate and iodine samples collection capability, i'
.... . . . . . _ . - . -~ . - . - - - -_
ACNGS-PSAR
- LIST OF EFFECTIVE PACES CliAPTER 13
, C) CONDUCT OF OPERATIONS
- 8. age. Amendment
-1* 57 2* 57 3*
4* 57 i 5* 57 i 33 i ii 55 iii .55 iiia 55 iiib 55 iv 55 v 57 13.1-1 57 13.1-2 57 13.1-3 57 13.1-4 57 1 13.1-5 57 13.1-6 57 f 13.1-7 57 13.1-8 57 i 13.1-9 57 13.1-10 57 13.1-11 57 3 13.1-12 57 13.1-13 57 13.1-14 57 13.1-15 57 } , j 13.1-16 33 (deleted) i 13.1-16a 33 (deleted) I 13.1-17 33 (deleted) l~ 13.1-17a 33 (deleted) 13.1-18 33 (deleted) 13.1-19 33 (deleted) 13.1-19a 33 (deleted) 13.1-20 33 (deleted) 13.1-21 33 (deleted) 13.1-22 33 (deleted) 13.1-23 33 (dcleted) 13.1-24 33 (deleted) i 13.1-24a 33 (deleted) l 13.1-25 33 (deleted) l 13.1-26 33 (deleted)
*Ef tective Pages/ Figures Listing i
!s l 1 Am. No. 57, (5/81) i
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ACNGS-PSAR-f'N LIST OF EFFECTIVE PAGES (Cont'd)
- l CHAPTER 13 Page Amendment 13.1-27 33 (deleted) 13.1-28 33 (deleted) 13.1 33 (deleted) "
13.1-30 33 (deleted) 13.1-31 33 (deleted) 13.1-32 33 (deleted) 13.1-33 33 (deleted) 13.1-34 33 (deleted) 13.1-35 33 (deleted) 13.1-36 33 (deleted) 13.1-37 33 (deleted) 13.1-38 33 (deleted) 13.1-39 33 (deleted) 13.1-40 33 (deleted) 13.1-41 33 (deleted) 13.1-41a 33 (deleted) 13.1-41b 33 (deleted) 13.1-41c 33 (deleted) 13.1-41d 33 (deleted) 13.1-41e 33 (deleted) O 13.1-41f 13.1-41g 13.1-41h 33 (deleted) 33 (deleted) 33 (deleted) 13.1-41i 33 (deleted) 13.1-41j 33 (deleted) 13.1-41k 33 (deleted) 13.1-411 57 (deleted) 13.1-41m 57 (deleted) 13.1-42 57 (deleted) 13.1-43 57 13.1-44 26 13.1-45 41 13.1-45a 41 13.1-46 11 13.1-47 11 13.1-48 33 13.1-49 - 13.1-50 57 13.1-51 57 13.lA-1 33 13.lA-2 48 13.lA-3 46 13.IA-4 46 13.lA-5 46 13.lA-6 46
- 13. l A-7 46 l 13.lA-8 46 x
I' i + 2 Am. No. 57, (5/81)
.. - - .- -. - . . ~.. ...
ACNGS-PSAR
- . LIST OF EFFECTIVE PAGES (Cont'd)
CHAPTER 13 s. 7 Pg Amendment I 13. l'A-9 48 l 13.1A-10 46 i 13.1A-11 48 l 13.lA-12 48 j 13.lA 46 ! 12.1 A-14 (deleted) 46 ! 13.lA-15 (deleted) 46 13.2-1 42 i 13.2-la 42 13.2-2 33 ! 13.2-3 33 i 13.2-4' 45 13.2-4a 42 l- ^ 13.2-5 33 13.2-6 42 13.2-7. 33 13.2-8 33 i 13.3-1 55 . 13.3-2 55 j 13.3-3 > 55 4 13.3-4 55 13.3-5 55 l 'E 13.3-6 13.3-7 55 55 , 1 13.3-8 55 g i 13.3-9 55 { 13.3-10 55 { 13.3-11 55 j 13.3-12 '55 l 13.3-13 55 1 13.3-14 55 I 13.3-15 55 l 13.3-16 55 l- 13.3-17 57 j 13.3-18 57 13.3-19 57 .; 13.3-20 57 i 13.3-20a- 57 I 13.3-21 55 i 13.3-22 55 i 13.3-23 55
- 13.3-24 55
{ 13.3-25 55 13.3-26 55 . I 13.3-27 55 l 1 ! 13.3-28 55 ], 13.3-29 55
;( l3.3-30 55 1
I i . 3 Am. No. 57, (5/81)
1 i { ACNGS-PSAR LIST OF EFFECTIVE PAGES (Cont'd) 4 CHAPTER 13 i- ! g . Amendment 13.3-31 55 13.3-32 55 .- i
... 13.3-33 55- '
13.3-34 55 > 13.3-35 55-13.3A-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 I B-1 thru B-7 55
'C-i 55 C-1 and C-2 55
! 13.4-1 33
- i. 13.5-1 42 13.6-1 16 1
i l 13.7-1 33
- 13.7-2 33
! 13.7-3 33 13.7-4 ; 33 1
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4 9 t 4 Am. No. 57, (5/81) I
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ACNGS-1SAR l EFFECTIVE FIGURES LISTING CHAPTER 13 ' CONDUCT OF OPERATIONS ( Figure Amendment 13.1-1 57 13.1-2 57 13.1-3 57 13.1-4 57 (deleted) 13.1-5 57 (deleted) 13.1-6 57 (deleted) 13.1-7 57 (deleted) ! 13.1-8 48 (deleted)
- - 13.1-9 33 (deleted) i 13.1-10' 33 (deleted) 13.1-11 33 (deleted)
- ~ 13.1-12 57 (deleted) 13.1-13 57 (deleted) 13.2-1 33 (deleted) 13.2-2 33 13.3-1 57 i
i 1 4 l 4 i i i i. 1 i i i I l r i l 5 Am. No. 57, (5/81) '
- .
ACNGS-PSAR s g LIST OF FIGURES Eigure Title 13.1-1 Corporate Organization 13.1-2 Allens Creek Project Management Organization I 13.1-3' Allens Creek Project Construction Organization j l
-13.1-4 Dele ted l l
13.1-5 Deleted 57 ; 13.1-6 Deleted 13.1-7 Deleted 1 13.1-8 Deleted 13.1-9 Deleted Deleted () 13.1-10 13.1-11 Dele ted 13.1-12 Deleted 4 13.1-13 Deleted 57 13.2-1 Deleted 13.2-2 General Electric Program i ] 13.3-1 Technical Support Center Arrangement 57 j l 4 1 i hu y (U)-Update Am. No. 57, (5/81) 1 _ _ . . - . . . _ . . . . , ~ . - - _ _ - - . , - _ _ - - - . . . . - . _ . . . . . . _ _ _ . _ . . _ - _ _ . , _ . _ . . - . - . . . . . . - - . _ _ _ . _ _ _
ACNGS-PSAR , 13.0 ORGANIZATIONAL STRUCTURE OF APPLICANT 57 (D ) The Allens Creek Nuclear Generating Station will be owned by Houston Lighting
& Power Company (HL&P). HL&P has traditionally retained overall responsibility for the design and construction of its generating units. HL&P has retainti ,
primary responsibility for the purchase, start up, and substation design for
! its generating units. Ebasco Services Incorporated has been assigned the responsibility for detailed design and construction of many HL&P generating 33 )
units. Construction of generating unit additions has been accomplished by HL&P employees and subcontractors, many of which were managed by Ebasco. HL&P has also performed the start-up and testing of the generating plant additions. This work has been done by employees of the Energy Production
. Department with assistance from the Ebasco Betterment Engineers.
L For the Allens Creek Nuclear Generating Station, Ebasco Services Incorporated j has been selected to perform the architect / engineer and construction management services. As in the past, HL&P will retain full responsibility for the func-tional performance of the design of all systems in the plant. HL&P will also directly perform all procurement of equipment and construction contracts, and plans to perform start up and t.esting basically with HL&P Co. employees with
' technical assistance and direction from the General Electric Company and
.Ebasco.
General Electric Company will design, f abricate, and deliver the nuclear steam
j
\ supply system, turbine generator unit and its auxiliaries and nuclear fuel sufficient for che initial core loading.
13.1 ORGANIZATIONAL STRUCTURE OF APPLICANT i 13.1.1 MANAGEMENT AND TECHNICAL SUPPORT ORGANIZATION 4 I 13.1.1.1 Design and Operating Responsibilities , i The following paragraphs summarize the degree eo whien design, construction, l and preoperational activities have been accomplished, and describe specific 57
- responsibilities and activities relative to technical support for operations, i
13.1.1.1.1 Design and Construction Activities (Project Phase) i 13.1.1.1.1.1 Principal Site-Related Engineering Work j_ a) Meteorology 4- A preoperational meteorological monitoring program was established at the site to provide those meteorological factors that bear upon plant 4 design, operation, and safety. The program was conducted by Dames & i Moore until 1975 with the responsibility for the program being assumed j by the HL&P Environmental Air Quality Group at that time. m/ e a 13.1-1 Am. No. 57, (5/81)
1
^
l ACNGS-FSAR-g b) Geology and Seismology L The geological and seismological investigations and evaluations were 1 conducted by Dames & M9 ore. The geotechnical engineering investiga-i tions and analysis for the plant location and ultimate Heat Sink were i conducted by Ebasco. A major portion of the field boring and sampling program and specialized' laboratory and field studies associated with-this work were done by subcontractors under the supervision of Dames &
Moore. The geotechnical engineering and analysis for the reservoir and reservoir-related facilities were conducted by Ebasco. 4 [ .c) Hydrology Dames & Moore has developed the probable maximum flood from offsite areas ACNGS based upon hydrologic investigations of the Brasos River 4 Basin previously made by the Forth Worth, Texas, District Office of the U.S. ; Army Corps of Engineers. Physical data, previous reports, and unpublished engineer,ing studies, together with technical guidance, were
- made available by both the Fort Worth and Galveston District Corps of
Engineers offices. De' tailed information concerning hydrology is given in Section 2.4. , d) Demography 57 Dames & Moore performed demographic studies relative to population within 50 miles of the plant, as discussed in Section 2.1.3. 1 e) Environmental Effects l
, A preoperational monitoring program has been developed to enable the
, collection of data necessary to determine possible impacts on the en-vironment due to construction activities and to establish a preopera-i tional base line from which to evaluate future environmental monitor-
~ing. This program is described in the applicant's Environmental Report.
An Environmental Protection Control Plan provides for periodic review of all construction activities to ensure that those activities conform to the environmental conditions set forth in the Environmental Report. If harmful effects or evidence of irreversible damage are detected by j the monitoring program, the applicant will provide the NRC with an i i_ analysis of the problem and a plan of action to be taken to mitigate ; 4 detrimental effects or damage.
- Preoperational radiological monitoring activities have been initiated.
l l 13.1.1.1.1.2 Design of Plant and Ancillary Documents All correspondence among HL&P, Ebasco and GE which affect the design interface ! are received and processed by each of the parties according to internal proce-j dures. Periodic audits between Ebasco and GE are held to provide assurance that design interface are met. HL&P participates in these audits.
l a { 13.1-2 Am. No. 57, (5/81) i
1 ACNGS-FSAR [h Qj 13.1.1.1.1.3 Review and Approval of Plant Design Features Houston Lighting and Power, as the owner, is responsible for the overall design and engineering of the Allens Creek Nuclear Generating Station. Although Ebasco Services, the architect engineer and GE, the Nuclear Steam System Supplier, perform the design, engineer'.ng and design verification of the facility HL&P accomplishes its responsibilities by reviewing key elements of the design. The basic design criteria is developed by Ebasco under the scrutiny of HL&P. In developing this criteria, Ebasco uses criteria from the following sources: a) Basic design and interface criteria supplied by GE as part of the NSSS. b) Previous design and construction experience of Ebasco. c) Previous operations and maintenance practices of HL&P. d) All applicable codes, standards, regulatory guides. e) Industry experience and practice HL&P review of this design criteria is achieved through project procedures. Ebasco uses this information to develop a System Design Description (SDD) for
/ each major plant system. This description includes the following parts: 57 I
5 N-- 1) functions
- 2) design requirements
- 3) design description
- 4) system limitations, setpoints, and precautions
- 5) operation
- 6) casualty events and recovery procedures
- 7) maintenance HL&P reviews and approves each SDD. The SDD's form the basis for design drawings and procurement specifications.
Ebasco prepares detailed design drawings based on the above described SDD's. HL&P performs mandatory review on drawings which affect key elements of the plant such as general arrangements and flow diagrams. Other detailed design drawings are reviewed on a spot basis to ensure compliance with basic design criteria. HL&P has specific procedures to achieve this performance. The Quality Assurance Program for Allens Creek to assure compliance is found in Chapter 17. 13.1.1.1.1.4 Development of Safety Analysis Reports Overall responsibility for preparation of the Preliminary Safety Analysis a ['s ( ) ,,
\ Report rests with HL&P Nuclear Licensing. Preparation of individual sections was assigned to the cognizant technical groups within Ebasco and within General Electric for the NSSS systems.
13.1-3 Am. No. 57, (5/81)
.ACNGS-FSAR 13.1.1.1.1.5 Review and Approval of Material and Component Specifications
\
(/ All safety-related project specifications are reviewed in accordance with the Quality Assurance Program for Allens Creek as discussed in Chapter 17. During the design and construction phases of the Allens Creek Nuclear Generat-ing Station, several organizations participate in the procurement of material and components,. However, HL&P does not relinquish the overall responsibility for assuring adequacy for the procurement program of each party involved. The Project Manager, Allens Creek, is responsible for all aspects of the procure-
- ment schedule and the control of procurement documents by HL&P and for coor-dinating the review and approval of these documents by Ebasco and GE. GE is responsible for the procurement of equipment, material, and services related to the NSSS scope of supply. Ebasco's scope of responsibility relates to the equipment, material, and services for the balance of plant (BOP) systems not included in the GE scope of supply. HL&P retains.the right to review and approve the specifications used by Ebasco to purchase BOP material and com-ponents.
The system requirements related to each compcnent specification are developed in System Design Descriptions as noted in Section 13.1.1.1.1.3. Other requirements related to the particular piece of equipment are contained in , industry codes and standards, and regulatory guides. Other requirements are based on Ebasco's design practices, HL&P experience, and industry experience. 57 HL&P reviews all specifications in accordance with internal procedures. The O personnel required to review such specifications are also described in pro-
'\ l cedures. HL&P conducts audits of specification review activities within GE, Ebasco, and HL&P. The Quality Assurance Program in relation to this activity is given in Chapter 17.
13.1.1.1.1.6 Procurement of Materials and Equipment Engineered equipment inquiries are assembled by Ebasco, including the HL&P standard commercial terms and conditions, schedular information, and the approved specification, and distributed to HL&P approved bidders. Bidders for safety related equipment are expected to have: (1) a quality assurance plan in effect, and (2) applicable stamps and licenses. Upon receipt of bids, technical adequacy and specification compliance is reviewed by engineering, quality assurance plans and testing metnods are reviewed by quality assurance personnel, and pricing and terms are evaluated by purchasing. Ebasco provides and evaluation and recommendation as to a vendor after resolving with each bidder to the extent possible unacceptable or unfavorable bid exceptions. HL&P reviews the recommendation, making the ultimate selection. In no case, however, will a selection be made for a vendor whose equipment does not meet the necessary requirements of the specification as viewed by the orQinating engineer or the quality assurance engineer. Ebasco then prepares the purchase order which is signed and issued by HL&P. Safety related materials procured in the field by the Ebasco site organization must also be reviewed and approved by quality assurance personnel. 13.1-4 Am. No. 57, (5/61)
ACNGS-FSAR l s 13.1.1.1.1.7 Management and Review of Construction Activities ( j) HL&P construction management will perform the following management and control activities at the construction site, a) Construction reviews and approves construction-related documents prior to implementation.
- 1) Field requests for design changes are reviewed for potential cost and schedule impacts and for assurance of proper engineering reiew.
- 2) Construction-originated drawings are reviewed for cost impact, economical design, and conformance with the overall construction plan.
- 3) Recommendations for field purchases are reviewed to ensure confor-mance with engineering specifications and drawings and general project procurement guidelines.
- 4) Construction procedures are reviewed to ensure conformance with engineering specifications and HL&P standard policies. Addi-tionally, the procedure is analyzed to ensure that the most efficient and economical method of performing the work has been 57 selected, b) HL&P construction engineers monitor field activities. This includes
+Q.\ J inspection of ongoing conscruction activities for conformance to specifications, drawings, and procedures. Construction plays an active V part in resolution of problems identified during field monitoring. Review of the security, safety, and environmental programs for com-pliance with established project guidelines is also included in the' general monitoring program. c) The contractor's cost and schedule performance is monitored to keep construction management and project management informed of project status. The construction engineers evaluate ongoing construction activities for cost, schedule, and quality problems and actively participate in identifying and selecting resolutions tc eliminate or reduce these problems. I i 13.1.1.2 Organizational Arrangements l l 13.1.1.2.1 General HL&P is engaged in the production, transmission, distribution, and sale of I electric energy for lighting, heating, cooling, and power purposes to resi-
- dential, commercial and industrial customers. HL&P's service area is in and around Houston, Texas, and includes approximately 5,600 square miles. HL&P operates electric generating plants with a total capacity of 12,169 MW. The Company has experience in the design, construction, startup, testing, operat- ;
ing, and staffing of modern generating facilities. d 13.1-5 Am. No. 57, (5/81)
ACNGS-FSAR I~ - The corporate organization, which provides- line responsibility for operation
- of the company, is shown on Figure 13.1-1. Ultimate responsibility for
[%-V) - design, procurement, construction, testing, quality assurance, and operation rests with an Executive Vice Presidene reporting to the President and Chief Executive Officer. The' Executive Vice President assigns responsibilities to , the various HL&P organizations described below. 13.1.1.2.2 Operations Nuclear. Operations reports to the Executive Vice President and is responsible for activities related to operation and maintenance of the nuclear generation I stations. 13.1.1.2.3 Nuclear Engineering and Construction The Vice President, Nuclear Engineering and Construction, reports to the i Executive Vice President and is responsible for power plant engineering and construction activities as they relate to HL&P's nuclear power facilities. 13.1.1.2.3.1 Allens Creek Project Department The Manager, Allens Creek, is' responsible to the Vice President, Nuclear Engineering and Coastruction, for the management, coordination, scheduling, cost control, engineering, construction and startup of the Project. To 57 accomplish these goals, the Manager, ACNGS is supported by other discipline departments within HL&P. ,k The members of the Project Managenent Organization are assigned to the team by various departments within the HL&P. These assigned team members are respon-
- sible through the Project Management Organization, an'. ultimately through the
- Manger, ACNGS, for the technical adequacy of work executed in support of ACNGS.
j They also serve as inerfaces between their individual line organizations and j Project Management to ensure the accomplishment of the project-related work i pertinent to their departments. The organization, responsibilities, and qualifications of the Project Management Organization are described in Subsection 13.1.1.5. j 13.1.1.2.3.2 Nuclear Services Department
- The Manager of the Nuclear Services Department is responsible to the Vice President, Nuclear Engineering and Construction for the design and engineering
- of the nuclear system, as well as for corporate health physics.
I 13.1.1.2.3.3 Nuclear Licensing l The Manager, Nuclear Licensing is responsible to the Vice President - Nuclear i Engineering and Construction, for the identificatica and interpretation of , l requirements set forth by the NRC. 1 1 C 13.1-6 Am. No. 57, (5/81)
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. q 13.1.1.2.4 Quality Assurance Department The Manager, Quality Assurance is responsible for providing the programmatic
, directica, and adminictering policies, goals, objectives and methods which are
. described in the Project Quality Assurance Plan. The Project Quality Assur-ance 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. The Manager, Quality Assurance, reports on all technical and administrative matters directly to the Executive Vice-President of HL&P. Thia organizational arrangement provides independence from cost and scheduling influences.
i The major responsibilities of the Manager, QA are: a) Administer QA policies established by management and ensure the proper i planning, development, implementation, coordination and administration of the Project Quality Assurance Plan. b) Provide programmatic direction on QA related matters to HL&P and i contractor management and interface with NRC. I c) Coordinate activities relating to auditing and vendor surveillance in
conjunction with the HL&P Housten Quality Assurance Manager. The Manager, Quality Assurance has the authority to solve quality related 57 problems and to verify the implementation and effectiveness of the solutions. He has the authority to "Stop Work" for cause on any quality-related activity. 13.1.1.2.5 Nuclear Fuel The Director, Nuclear Fuel reports to the Executive Vice President and is responsible for assigning an enginc.er to the AC Management Team. This engineer is responsible for delivery and utilization of nuclear fuel. 13.1.1.2.6 Other Corporate Support Organizatons Although many other organizations within the Company not directly reporting to ; , the Executive Vice President furnish staf f and/or technical and administratim support to the Project, their actions as related to the Project are controlled I by the Executive Vice President through the Manager, Allens Creek. The major other corporate support organizations are discussed below: 13.1.1.2.6.1 Power Plant Engineering i The General Manager of Power Plant Engineering is responsible to the Vice President, Power Plant Engineering and Construction-Fossil Projects, for the coordination of the engineering design and review of the BOP mechanical sys-tems and their associated instrumentation and control systems. all electrical
. equipment and all civil engineering activities related to th site, bulidings and structures. Poser Plant Enginaering assigns engineers to the ACNGS Project Engineering Team to support these activities frous the following disciplines: mechanical, electrical, instrumentation and control, and civil.
I 13.1-7 Am. No. 57, (5/81) t.
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ACNGS-FSAR 13.1.1.2.6.2 Environmental Protection The Manager, Environmental Protection, reports to the Vice President, Power System Development and is responsible for the areas of site selection criteria, radioactive dispersion, thermal effects, air and water quality considerations, and environmental surveillance, including meteorological monitoring, geophysical testing, hydrological evaluations, and all offsite operational effects of the nuclear power plant. He is responsible for preparation of the Environmental Report and environmental considerations required in support of all safety analysis reports. His responsibility also inr.ludes acquisition of all local, state, and federal permits and approvals, exclusive of NRC licensing. , 13.1.1.2.6.3 System Engineering The Vice President, System Engineering and Operation, is responsible for contractual activities as related to the design, engineering, and constructxon of the switchyard at the nuclear power plant projects. The Manager, Engineer-ing Design and Development, is responsible for the engineering and design of the electrical auxiliary systems and the overall electrical protection of the switchyard and for the studies and specification of generator characteristics as they pertain to proper integration of the unit into the system. 13.1.1.2.6.4 Consulting Engineering Department The Consulting Engineering Department is responsible to the Vice President, 57 Power Systems Development for providing support as necessary to the other h' company departments. Liaison with other utilities is maintained through personal contacts and work with industry organizations. The Consulting Engineering Department does not routinely participate in design review activi-ties. One of the consulting engineers provides guidance and support far the ACNGS startup organization upon request. 13.1.1.3 Qualifications Engineering personnel are provided to the Allens Creek Project Management Organization from the staffs of various other departments as described in Section 13.1.1.2.3.1. Within HL&P, the person whose job position most closely corresponds to that identified as " engineer in-charge" as defined by ANSI N18.1-1971 is the Manager, Allens Creek. The resume of the Manager, Allens Creek is provided in Section 13.1.1.4. The education and experience of Allens Creek Project Management Organization personnel are listed in Table 13'1-2,. 13.1.1.4 Qualifications of Corporate Personnel Resumes of key personnel involved in ACNGS are provided in this section. HL&P organizational charts are provided as figures at the end of Section 13.1. Corporate technical resources are shown on Tables 13.1-1 and 13.1-2. 13.1-8 Am. No. 57, (5/81)
l l ACNCS-FSAR w a) Executive Vice President ! George W. Oprea, Jr. , Executive Vice President, is a 1952 graduate of Rice University with Bachelor of Arts and Bachelor of Science degress in Electrical Engineering. He joined HL&P that year in the Distribution Planning Section of the Engineering Department. He later worked in Computer Applications Engi-neering for System Planning, and in March 1965, was named Superintendent of I the Engineering Planning Division. He became the Energy Control Center Project Manager in March 1967; Manager, Energy Control Dispatching Department
- in June 1970; and Managar, Energy Control and Nuclear Program in April,1971.
He was elected Vice President of Operations in November 1971, Group Vice President of Opernions in January 1973, and Executive Vice President and member of the Brurd of Directors in December 1974. Mr. Oprea is a registered professional engineer in Texas, and a senior member of the Institute of Electrical and Electronic Engineers. He is a past director
- and past president of the Engineers Council of Houston, a member of the Association of Computing Machinery and Society of Information Display. He a also holds membership in Edison Electric Institute Executive Committee on Nuclear Power, EEI Engineering and Operating Division Executive Committee the Houston Chamber of Commerce and the American Nuclear Society. Mr. Oprea is a 57 l director of the Slurry Transport Association, the Atomic Industrial Forum, and the American Nuclear Energy Council. He is a counselor for the Texas A&M Research Foundation and a retired captain the the Naval Reserve.
~ b) Vice President, Nuclear Engineering and Construction
- Jerome H. Goldberg, Vice President, Nuclear Engineering and Construction graduated from the U.S. Merchant Marine Academy where he studied marine engineering. He obtained a master's degree in nuclear engineering from Massachusetts Institute of Technology. After a tour of duty with the U.S.
Navy, Goldberg joined Bethelen Steel Corp., at its Quincy, Massachusetts shipyard where the company was involved with the designing and building of nuclear powered cruisers and destroyers. Later, he worked for General Dynamics Corporation, where he was involved in all facets of engineering and construction of nuclear power plants for sub-marines. In 1971, Goldberg joined Stone & Webster Engineering Corporation in Boston, i Massachusetts, as a nuclear engineer, assigned to work on a major pressurized water reactor project. He has also served as project manager and construction manager on various nuclear power plant projects. In 1976, he became chief engineer of the Engineering Mechanics Division and in 1978 became a Vice
- l. President, Construction.
O - i V 13.1-9 Am. No. 57, (5/81)
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ACNGS-FSAR p c) Manager, Allens Creek Paul A Horn, Acting Manager, aliens Creek received the Bachelor of Science degree in Mechanical Engineering from Texas A&M University in 1969, and after continuing his education at Texas A&M, he received the Master degree of , Mechanical Engineering in 1970. He joined the Energy Production Department of HL&P upon graduation and after training in various department areas was assigned to the W.A. Parish Generating Station. There he directed routine performance tests of condensers, pumps, and heaters, supervised certain plant maintenance and modifications, and coordinated routine plant start-ups and shutdowns of fossil fuelad gas turbines, supercritical steam units, and drum boiler units. In March 1972, Mr. Horn was transfereed to the HL&P Nuclear Program where he participated in the writing of the NSSS specification and performed the Allens Creek NSSS evaluation for mechanical equipment. In August 1972, he was appointed Project Engineer for the Allens Creek Nuclear Generating Station. He was then promoted to Project Manager on June 17, 1974. During the postponement of Allens Creek in 1975 and 1976, Mr. Horn also assumed the Project Management responsibility for HL&P's 6-60 MW Gas Turbine facility at Greens Bayou, the 100 mile fuel oil pipeline system connecting our 57 power stations, and the five fuel oil pumping and storage facilities. He had responsibilities for these projects during the entire construction and startup phases and then devoted full time to Allens Creek af ter 1976. In March 1980 the Project Manager title was changed to Manager, Allens Creek, and Mr. Horn b was made Acting Menager.
\}) d) Manager, Nuclear Services Dr. J. R. Sumpter, Manager, Nuclear Services, received a Bachelor of Science in Engineering Science from Per.nsylvania State University in 1965, a Master of Science in Nuclear Engineering from the University of Michigan in 1967, and a PhD in Nuclear Engineering from Texas A&M University in 1970. His disserta-tion concerned the study of xenon oscillations during power reactor transient operations. He was employed in the summers in 1964 and 1965 at the Bettis Atomic Power Laboratory in the mechanical and nuclear design of naval reactors. l In the summer of 1967, he worked at the Los Alamos Scientific Laboratory as a research physicist involved with the theoretical and experimental study of !
critical assembly designs. Dr. Sumpter was employed as a Nuclear Analyst with Sargent & Lundy Engineers from 1970 to 1972. He had responsibilities involving radwaste systems design, health physics, shielding, radiation monitoring system design, equipment procurement, overall plant engineering design, and associated licensing for several nuclear power stations. He joined HL&P in August 1972. In March 1973 he was promoted to Supervising Engineer, Nuclear Safeguards and Licensing, and was responsible for ensuring ! that the design and operation of all HL&P nuclear power plants conformed to ) all applicable NRC regulations and criteria. In February 1975 he was promoted i
to his present position as Manager, Nuclear Services. l t I x 13.1-10 Am. No. 57, (5/81) i l l
ACNCS-FSAR a m e) Manager, Electrical Engineering Wade H. Morgan, Manager of Electrical Engineering, was graduated from Texas A&M University in 1948 with a Ba:helor of Science degree in Electrical Engineering, and joined HL&P in June of that year. With maintenance and operating experience'at progressive levels of responsibility, he was appointed , , Plant Superintendent of the W. A. Parish Plant in 1964, Superintendent of Electrical Maintenance in 1967, Senior Electrical Engineer in 1971, Principal 4 Engineer in 1972, Manager of Projects in 1977, and Manager of sngineering in 1978. Mr. Morgan is a member of a number of engineering organizations. < ! f) Manager, Mechanical / Civil Engineering James E. Bouvier, Manager of Civil / Mechanical Engineering, graduated from the University of Houston in 1967 with a Bachelor of Science degree in Mechanical
- Engineering, received a Master's degree in Mechanical Engineering in 1970, and a Master's degree in Business Administration in 1979.
Mr. Bouvier was employed by the National Aeronautics and Space Administration in the area of fuels, gas turbo machinery, and hydraulics research and dave-lopment from 1967 to 1973. He joined Bechtel Power Corporation in 1973 and l was involved in the design of both nuclear and fossil power generation pro- 57 jects. Mr. Bouvier joined HL&P in January,1977, as a Senior Engineer in the Engi-neering Department supervising the Allens Creek Nuclear Project. He was promoted to Supervising Engineer in March,1978, with responsibilities for the Allens Creek Project, Freestone Project, and Mechanical Engineering Staff. In February,1980, Mr. Bouvier was promoted to Principal Engineer in the Mechanical Division. He was responsible for the management and administration of the Mechanical Division of the Power Plant Engineering Department, which included all aspects of mechanical engineering. He also participated in the Site "X" A/E evaluation. Mr. Bouvier was promoted to Manager of the Civil / Mechanical Engineering Division in the Power Plant Engineering Department in October, 1980. His responsibilities include management and administration of the Mechanical and Civil Divisions. g) Director, Nuclear Fuel l Richard P. Murphy, Director, Nuclear Fuel, has a Bachelor of Science a gree in Mathematics from St. Mary's University in San Antonio, and a Master of Science degree in Nuclear Engineering from Texas A&M University. In addition, Mr. Murphy has completed approximately 40 hours of additional course work required for a PhD in Nuclear Engineering at Texas A&M.
;
O 13.1-11 Am. No. 57, (5/81) l
ACNGS-FSAR I Mr. Murphy joined HL&P in 1972 as an Engineer in the Engineering Design l Section, where he was involved in estimating dose rates and shielding require-
"f ments, technical evaluation of NSSS proposals and various assignments in the l Fuel Management Section. In February 1974, he transferred into the Fuel l Management Section, where his responsibilities included economic and technical evaluation of fuel proposals, preparation of fuel contracts and specifications and fuel design review. In August 1974, he was promoted to Supervising Engineer of the Fuel Management Section in the Nuclear Program, and, in June 1977, was promoted to his present position.
h) Manager, Nuclear Licensing Mr. Cloin C. Robertson received his Bachelor of Science degree f rom West Point in 1958 and Master of Science in Nuclear Engineering from MIT in 1965 and a Nuclear Engineer degree from MIT in 1967. After graduate school, Mr. Robertson was employed as an engineer at the KAPL Division of General Electric Company where he performed safety analyses and core thermal / hydraulic analyses on the reactors for the Nimitz class attack carriers. From January 1973 to May 1974, Mr. Robertson was on the staf f of the New York Atomic Energy Council as Nuclear Facilities Specialist where he performed 57 technical safety reviews of nuclear power plants located in New York State. From May 1974 to March 1977, was Siting Program Manager for New York Energy ( (3 V) Research and Development Authority where he managed the state wide environmental and technical studies to identify and evaluate sites for major new electrical seration plants. He also coordinated and directed the authority's i.cerests in the West Valley reprocessing plant and waste burial facility. He joined Stone & Webster in March 1977 as a Senior Licensing Engineer, and was appointed Supervisor of Engineered Safety Systems and Analysis Group in January 1978. Previously, he served as Licensing Engineer for Stone & Webster's Standard Plant (SWESSAR) and in Operations Services Division related to various Licensing activities for operating nuclear plants. He joined HL&P's staff as Manager, Nuclear Licensing, in March of 1981, 13.1.1.5 Project Management Organization 13.1.1.5.1 The Project Management Organization Under the direction of the Manager, Allens Creek, is the organizational and management structure singularly responsible for the direction of the design and construction of the Allens Creek plant. The Project Organization is composed of individuals assigned full time to the Allens Creek project from several departmental organizations. The Team members receive directions as to activity and schedule from the Manager, Allens Creek, however, rely upon their department or discipline management for technique and training. The Allens Creek Project Management Organization is physically located together and [,'T operates as a unit controlled by Project specific procedures and QA plan. As ( / activity would dictate, the group would relocate to the site to ensure com-municate and timely support of construction needs. The Project Management Organizational chart is shown on Figure 13.1-2 13.1-12 Am. No. 57, (5/81)
ACNGS-FSAR
.g , 13.1.1.5.2 Responsibility Assignments 4
( . Key individuals within the Team are shown below. A . I 13.1.1.5.2.1 Engineering Project Manager The HL&P' Project Engineering Manager is responsible for overseeing, coor-dinating and administering the ACP engineering effort. This effort is within the limits of the work contracted with Ebasco and other suppliers of en-gineering services to assure that project engineering requirements are met. 13.1.1.5.2.2 Project Purchasing Manager The Project Purchasing Manager is responsible for the overall coordination and - , administration of purchasing and subcontracting activities for the ACP. This includes, but is not limited to, engineered and field purchased equipment and material, expediting, material control and warehousing. 13.1.1.3.2.3 Construction Project Manager The Project Construction Manager is responsible for monitoring the total construction eff'rt and maintaining liaison between HL&D and the Prime Con-57 tractora Management. 13.1.1.5.2.4 QA Project Manager The Project QA Manager is responsible for developing the programmatic direc-tion policies, goals, objectiven and methods which ensures the proper planning, development, implementation, coordination and administration of the Quality Assurance program for the ACP during engineering, design, procurement, fabrication, construction, testing, start up and operation a+.tivities. Programmatic direction is defined as the role of HL&P in setting the program requirements and in ensuring the adequacy of the contractor's quality assurance program. The programmatic direction consists of a review and
; approval of the system features in the initial phase and continued monitoring of these . systems during implementation and further refinement or revision of
, the system if the systems need strengthening. 13.1.1.5.2.5 Project Controls Marager The Project Controls Manager is scsponsible for providing a detailed project
- . schedule integrating engineeting, construction and start up. Ensures that a j budget is developed from the project cost estimates and integrated with t'ae project schedule. Develops and implements a cost and schedule monitoring system to assure that cost targets and schedules milestones are met.
13.1.1.5.2.6 Project Controller j The Project Controller is responsible for the coordination and execution of the accounting and financial administration of the ACP. I-1
,-~
LU i r ! 13.1-13 Am. No. 57, (5/81) i' ^ i
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ACNGS-FSAR 13.1.1.5.2.7 Supervisor, Project Administration the Project Administration Supervisor is responsible for coordination c f support to the Allens Creek Project Team from Ebasco and HL&P, processing and distribution of project mail, development of project procedures and adminis- ! trative support of the ACP. 57 13.1.1.5.3 Qualificati<.ans The qualifications of key individuals within the Team are shown on Table ' J 13.1-2.
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l 1 l
! I 1
i i
l 5 l l l 4 i l 13.1-14 Am. No. 57, (5/81) 1 i
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. ACNGS-PSAR 1 4 1 e i i t 1 l 4 i 1 I f i l PAGES 13.1-15 THRU 13.1-42 HAVE BEEN 57 INTENTIONALLY DELETED j i I i, i l
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I i I l 13.1-15 Am. No. 57, (5/81)
ACNGS-PSAR
.g 57 Q
l
'13.1.2: OPERATING ORGANIZATION
~[ ThisLsection will' describe' the structure, functions and responsibilities of the operating organization. I~ '
- 13. l.~ 2.1 Plant Organization The general plan for the staffing and operation of Allens Creek Nuclear Generating Station is pre,sented hereia.' Houston Lighting & Power Company, using ANSI N18.1-1971.and 10 CFR Part 55 as a guide, has reviewed the actual l33(U)
.and proposed nuclear plant operating staffs of other utility companies and i
has compared the duties of nuclear plant operating crews with those now
! being performed in HL&F Company's large fossil plants. The proposed organ-l izational chart is shown'in Figure 13.1-13.
p 13.1.2.2 Personnel Functions, Responsibilities and Authorities I The functions and responsibilities of all personnel positions at Allens Creek Nuclear Generating Station, including a specific succession to responsibility for overall operation of the plant in the event of absences, ,
- incapacitation of personnel or other emergencies, is described in this l
section. The line of authority for all personnel positions is shown in j Figure 13.1- 13. 13.1.2.2.1 Plant Superintendent The principle responsibility for all phases of plant management, including
- operation, maintenance and technical supervision, rests with the Plant Superintendent. He is also responsible for adherence to all requirements of the operating license and technical specifications.
J ) 1 j 13.1-43 (U)-Update 4-
.Am, No. 57, (5/81)
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ACNGS-PSAR TABLE 13.1-2
' PROJECT TECHNICAL QUALIFICATIONS AND RESOURCES Experience (man years)
Discipline Assignment Name Degree Total Related- Nuclear Power Manager Allens Creek P. A. Horn (act) MSME .12. 9. Project Administration Proj. Administrator L. F. Apice11a HS 13 10 Accounting Proj. Comptroller D. J. Boone CPA 15 1 Planning & Control Proj. Cost Analyst J. D. Mann HS 12 5 Purchasing Proj. Purch. Manager D. R. Var *.er BSEE 6 3 Material Control L. C. 'lanning HS 22 3 Material Control H. J. Johnson MBA 38 1 Quality Assurance Proj. QA Manager P. W. Stoerkel(act) BSNE- 24 9 Operations Operations Engr. F. J. Comeaux BSNE 3 1 i- Operations Engr. S. P. Niehaus BSNE 4 0 Fuel Resources Fuel Engineer B. A. Handly MSNE 6 6' Environmental Environmental Engr. H. P. Hardy MSLUP 4 4 57 ) Engineering C/I/E Supv. Proj. Engr. M. R. Mouser MSME,MBA 10 4 Equipment Qualif. R. J.-Cooper (act) BSME 4 3
- Special Projects M. A. Towner BSCE 5 3 Civil Civil Proj. Engr. B. P. Wilkerson BSCE 6 4 4 Lead Civil Engr. N. S. Thakkar MSCE.MBA 15 1 Civil Engr. C. A. Rater BSCE 6 4 U I&C -
IEC Proj. Engr. P. A. Ranzau BSIE 8 4 I&C Engineer R. D. Simpson BSNE 4 1
- 0. Electrical Elect. Proj. Engr. V. G. Perret ESEE 9 4 Electrical Engr. D. J. Hansen BSEE 2 2 Engineering M/N Supv. Proj. Engr. J. N. Bailey BEE 8 5 Mechanical Mech. Proj. Engr. S. G. Cartwright BSME 8 8 i Mechanical Engineer P. D. Marinshaw B SN E*,MS ME 7 2
+ Mechanical Engineer D. L. Harrell BSME 3 2 Nuclear Design Nuclear Proj. Engr. M. D. Gaden BSNE 10 8 Nuclear Engineer D. M. Chamberlain BSNE 3 2 3 Nuclear Engineer C. R. Granton BSNE 3 2 I Nuclear Engineer D. E. Root MSME 7 7
> Piping Engineer S. M. Head BSNE 4 4
,a Piping Engineer R. J. Cooper BSME 4 3 2 Health Physics Lead Health Physics M. M. DiGenova MSNE 8 5 3
,o Health Physics S. W. Woolfolk BSPHY 4 1
, , Health Physics L. A. Smith PHD 6 2
- j. y Licensing Hearing Coordinator L. J. Klement BSNE 7- 7 j Licensing Engineer T. A. Peterson BSES 5 5 3 U i
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)
d ACNGS-PSAR TABLE 13.1-3 CORPORATE R E'S O U R C E S i Function Discipline Highest Degree (No, of Professionals) Experience (Years) HS B_S, MS PHD Power Plant Nuclear i Nuclear Management 2 5 1 119 100 Engineering i Civil 28 5 0 189 67 Nuclear 16 6 0 142 119 Mechanical 30 7 1 289 98 l
,U Electrical 44 1 0 349 95 f
, e Other
~
15 1 0 133 49 l h 7 6 4 134 134 57 Health Physics Quality Assurance 52 20 0 0 449 470 l t Nuclear Construction 14 13 2 0 362 130 4 I Nuclear Operations 25 15 0 0 258 144 g 2 Nuclear Fuels 3 3 30 30 si
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\v VICE PRESIDENT NUCLEAR MANAGER ENGINEERING & QA h( CONSTRUCTION o :d /
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-~y the public. HL&P will vork closely with appropriate State and local I ) governments to ensure that, before the ACNGS begins operation, the capability . \,d will exist to notify the public in the 10-mile EPZ within 15 minutes of noti-fication to of f site organizations. An evaluation will be made to determine the specific system which will provide this capability. This system may in-clude sirens, in-residence tone alert devices, alarms connected to electric meters and/or multiple telephone call-up techniques.
1 Arrangements will be made f or broadcasting 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 Emergency f acilities, as well as special systems, will be established at and near the ACNGS for assessing an event, directing response and recovery efforts, mitigating accident consequences and informing the public. The
- f acilities and systems will be described in detail in the final ACNGS Emergency Plan.
13.3.5.1 Control Rons
- 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 f\)
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have communications with all other emergency f acilities and will be equipped with terminals of data systems. The Control Room, being inside the plant, is designed as a totally safety grade facility. 13.3.5.2 Technical Support Center On Onsite Technical Support Center (TSC) will be provided. Details of the TSC are described below. 13.3.5.2.1 TSC Function 57 The functions of the TSC are to: a) Provide a location for plant management and technical support personnel to work to support operations personnel during emergency conditions. , l b) Perform EOF functions during emergencies requiring EOF activation until the EOF is fully manned and functional. G' 13.3-17 Am, No. 51, (5/81)
ACNGS-PSAR It is expected that.the TSC complex will be used for routine plant functions. [< ~sI For example, the NRC office may be the Resident IE Inspectors office and the
\s # 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 Location The TSC is located on the northeast side of the Control Building at el.187', as shown on Figure 13.3-1. Walking time between the TSC and Control Room is 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. Thare are no major security barriers between the two areas. 13.3.5.2.3 TSC Staffing and Training This will be provided in the ACNCS Final Emergency Plan. 13.3.5.2.4 TSC Size The TSC is a complex consisting of-the following directly adjacent areas: a) Open working area for 25 people of approximately 1200 ft2 This area i also contains the SPDS and other plant data (Reg. Guides 1.23 and.1.17) displays. 57 [) b) Conference r3om of approximately 400 ft2 Telecopying equipment and
\,,,/ some plant data displays are also located in the conference room.
c) " Repair shop" for storing tools and spare parts for the TSC displays. d) Communications equipment room. e) Document storage room. 4 f) Office for NRC representatives, which can be used for private NRC consultations. g) Living area, including a kitchen, sleeping room and supply storage. h) The Security Secondary Alarm Station. 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 spc ified for safety-related structures for ACNGS. This exceeds the structural requirements for the TSC. l l 1 As _/ l I 13.3-18 Am. No. S7, (5/81)
ACNGS-PSAR r 13.3.5.2.6 TSC Habitability The TSC is served by the Control Room Ventilation System, so it is habitable 9
') to the same extent as the Control Room. See Section 6.4 13.3.5.2.7 TSC Commurications The 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.
'13.3.5.2.8 TSC Instrumentation, Data System Equipment and Power Supplies 57 I Plant data vill 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 the regard when approved by the NRC. As a minimum, the SPDS (see Section 7.5.1.6) and post-accident monitoring e
instrumentation (see Appendix C) will be available on CRTs in the TSC, with the SPDS displayed on a dedicated CRI.
- - The TSC displays are not Class 1E or seismically qualified, but are provided with reliable backup power from the BOP diesel generator.
13.3.5.2.9 Records Availability and Management The TSC will have a records storage area in which up-to-date plant records
'sFSred.
deemed A listuseful for accident of the types of these records or off-normal condition will be provided in the ACNGS diagnostics will be
;
O' c inal Emergency Plan. 13.3.5.3 Emergency Operations Facility i An Emergency Operations Facility (EOF) will be established near the ACNCS for the management of overall emergency response, the coordination of radiological assessments, and for management of recovery operations. The EOF will be designed to provide assistance in the decision making process to protect the public health and safety and to control radiological monitoring teams. The overall management of licensee emergency response will be based in the EOF. Working spaces for Federal, State and local response organizations will be 55 , provided in the EOF, thus making it the center of coordination for all organizations involved. The EOF will be activated for Site Area and General Emergencies and will be i~ brought to stsndby status for the Alert Emergency. It will be located outside the plant security boundary, but probably not f arther than five miles from the ACNGS. The EOF will be habitable under accident conditions, including those that would require evacuation of the plume exposure EPZ. The EOF will be equipped with communication links, plant safety status data, and radiological and meteorological data. 13.3-19 Am. No. 57, (5/81) 4_. . . . _ _ . . _-_,,,c..,,,m , , - _ _ _ . . . __w,.y.,. -.
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l i ACNGS-PSAR I l The design and location-of the EOF will be described in the final ACNGS EmergencyfPlan. HL&P presently plans to' design and develop the EOF in , accordance with the functional and design criteria specified in NRC final l(es)
- guidance documents.
t 13.3.5.4 operations Support Center 55 Appropriate space will be designated onsite for the assembly of operations l
. personnel whose support' is required in or near the plant, but not in the Control Room or TSC. .The exact location of the OSC has not yet been
' determined, but it is expected that an area in the Personnel Access BJilding* 57 such as the Conference Room or Lunch Room, will be designated for this purpose. Supplies such as protective clothing, respiratory protection, portable lighting and communications equipment will be provided. The OSC will be described in detail in the final ACNGS Emergency Plan.
1 1 13.3.5.5 News Media Center 7 i HL&P will provide a location at or near the EOF to serve as a News Media Center (NMC) in which to conduct press conferences and briefings during an emergency. The NMC will be activated for the Site Area and General Emergency levels and will be brought to standby status for the Alert Emergency. The NMC 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 MMC, 55 information packets or " press kits" will be available providing information about - the licensee, the plant and plant surroundings. Visual aids will be
\ provided.
i HL&P will designate an official company spokesperaon to interface with the news media and the spokespersons of offsite response organizations. HL&P spokespersons will be trained in conducting press conferences and briefings and will be knowledgeable of plant operations and the Emergency Plan. 13.3.5.6 Safety Parameter Display System l The SPDS is described in Section 7.5.1.6. 57 l ! 13.3.5.7 Data Transmission The ACNGS will be equipped with the capability to transmit plant data to the j TSC and EOF. The design of this system will be described in the final ACNGS 57 Emergency Plan. 1 1 4
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,,, 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 ACNGS. At least one individual onsite will be trained and qualified in advanced first aid methods. 13.3.5.9 Decontamination Facility Personnel decontamination facilities, consisting of showers and sinks which 55 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 deemed necessary, that treatment will take precedence over decontamination. This philosophy will also extend to transportation and offsite treatment of contaminated, injured individuals. i 1 l
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ACNGS-PSAR APPENDIX 15B I i ' (j ALLENS CREEK NUCLEAR GENERATING STATION RELIABILITY ANALYSIS PROGRAM
1.0 INTRODUCTION
Item 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 incorporated into the design of the facility. This fol-lowing discussion describes the proposed program to be applied to the Allens Creek Nuclear Generating Station (ACNGS) in response to NUREG 0178 including an outline of the program scope, methodology, schedule, quality assurance procedures and the means by which the reliability analyses will be integrated r into the ongoing design process. 2.0 SCOPE AND OBJECTIVES The ACNGS Reliability Analysis Program is similar in scope to the Interim 57 Reliability Evaluation Program (IREP) being performed by NRC for several oper-ating plants,except that it will include the calculation of fission product release quantities for the accident release categories. The objective of the program will be to seek such improvements in the reliability of core and A containment heat removal systems as are significant and practical and do not ( impact excessively on the plant. Individual accident sequences and their i V probabilities will be analyzed to identify the initiating events and plant system and component failures which are the domi- nant contributors to core damage risk. The list of initiating events to be considered will be determined during the initial phase of the study. However, as a minimum, the initiating events will encompass loss of coolant accidents (small, intermediate and large) and tran-sient events including loss of feedwater, loss of offsite power and turbine trip. The following key safety-related systems will be included in the program:
. Reactor Core Isolation Cooling System Residual Heat Removal Cooling System
. High Pressure Core Spray Cooling System
, . Low Pressure Core Spray System
. Automatic Depressurization System
. Containment Spray System
. Containment Isolation System
. Reactor Protection System
. Electric Power (AC and DC) System Essential Service Water System f .
! . Essential Chilled Water System j . Essential HVAC Systems
. Standby Liquid Control System
( v) 15B- 1 Am. No. 57, (5/81) i
ACNGS-PSAR 3.0- METHODOLOGY k}j l The approach to be used .in the program will employ event tree /f ault tree methodology similar to that used in WASH 1400 and other comprehensive plant risk studies. The major tasks involved are discussed below. 3.1 INITIATING EVENT SELECTION A list will be established of initiating events which, together with system failures, have the potential for_ causing core damage and of fsite radioactivity releases. This will be accomplished through 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 protection and release mitigation. The frequency of these initiating events will be estimat ed based on available data including WASH-1400, EPRI NP-801, and, where it exists, pertinent plant specific or site-specific information i (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 having the same or similar safety system response requirements will be grouped together and a unique event tree will be developed for each such group of events. Some of the initiating events listed in Table 1 of the NRC acceptance criteria are not included for detailed evaluation in the ACNGS PRA study. Specifically:
a) Failures during cold shutdown operation (Item 5): It is not expected 57 that such scenarios are major or even visible risk contributors. When
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(A) the plant is in the cold shutdown condition, reactor decay heat is only a fraction of the immediate post-shutdown amount which figures into the ! LOCA and transient events. Also, the low temperature of cold shutdown makes the primary system itself a large heat sink for decay power, so the system could withstand an extended time without decay heat removal before fuel damage conditions would even begin to be approached. The
forgiveness" of the plar.t under such conditions was demonstrated in the 4/19/80 Davis Besse incident. Thus, it does not appear worthwhile t o pursue this category of initiating events.
b) Severe Natural Phenomena and Fire (Items 6-10): Although it is theoretically possible to incorporate these as initiat-1 ing events into the reliability study, HL&P does not believe that the state of-the art is to the point where it could properly contribute in a quantitative manner. Qualitatively, it is expected that the con-servative licensing design basis used for there events in design render them insignificant contributors to true plant risk, particularly for a site like ACNGS, which has no unusual site characteristics. t O 15B-2 Am. No. 57, (5/81) i
.ACNGS-PSAR 3.2i . EVENT TREE DEVELOPMENT -)
p , For each unique group of initiating ' events, an event , tree will'be con-F lstructed, identifying the-safety systems required to mitigate the event and the expected :effect on ability to maintain ' core and containment- integrity given success or f ailure of each safety system involved. The full event tree 4
- j. will be reduced to reflect safety system interdependencies and required se-quences of operation. l t
i: 3.3 SYSTEM FAILURE MODES AND EFFECTS ANALYSIS (FMEA)
. For each' safety system involved, a FMEA will be conducted to identify and ' tab-ulate component and common cause failures and their effect on system operabil- '
. .ity for each initiating event. The FMEA will provide documentation of the basis for inclusion or exclusion of specific failure modes in the system fault-
'_ tree analysis. Failure modes will include mechanical and electrical faults, i
!_ operator error, maintenance-or testing outages, etc. Particular attention will be paid to potential common cause failures which could disable multiple compo-i- nents. Common cause failure mechanisms to be investigated include environmen-
- tal factors, operator or maintenance errors, passive failures and system inter-4 actions.
3.4 SYSTEM FAULT TRFL ANALYSIS (FTA) ) Using the'FMEA as input, fault trees will be constructed for each safety system 57 j identifying the failures (basic events) and their logical combinations which will result in system unavailability (top event). The fault tree will be r analyzed to determine the minimal cut sets and failure combinations which:are the dominant contributors to system unavailability. Using the appropriate com-ponent failure data, a quantitative assessment af overall system unavailability and of the dominant cut sets will be performed, i The fault tree analysis will be performed using a computer program to perfora ! cut set determination and quantitative analysis and to provide computer-
- j. generated graphical representation of the fault tree using standard logic
- . symbols.
3.5' DATA BASE DEVELOPMENT A component f ailure data base for use in system '..11t tree analysis will be ! developed from recognized reference sources incl 2 ding WASH-1400 and IEEE-500, 2 In addition, prot _ type-specific f ailure dat a 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 operation. ! Error ranges will be assigned to each median value to reflect the uncertainty j in the data base. The-data base will include methodologies to adjust failure i data to account for varying testing and sutveillance strategies. Test and {~ maintenance unavailability contributions will be included based an proposed ! Technical Specifications o.>erating and maintenance procedures. ! t c i \ 15B-3 Am. No. 57, (5/81) p P _ . . _ . . . _ _ .. .., _ _ _ . _ _ . . _ , . ~ . _ _ _ . _ _ , . _ - . -
ACNGS-PSAR Human error rates will be estimated for required or corrective actions by [\ : control room operator and for maintenance or testing operations which are ( ) included as failure modes in the system fault trees. Available human error and performance data, including those provided NUREG/CR 1279 will be used. 3.6 ACCIDENT SEQUENCE PROBABILITIES . The unavailability of failure probability of each system will be calculated by inputting the appropriate failure rate data into the system FTA. The various accident sequences, as represented by the branches on the event trees will then be quantified by inputting the system failere probabilities determined from the quantitative FTA. Each individual accident sequence will be classi-fied according to release category and the total probability of a given release category will be obtained by the summation of all accident sequence probabilities assigned to that category. 3.7 ACCIDENT SEQUENCE FISSION PRODUCT RELEASES The release categories of WASH-1400 will be re-examined and modified if neces-sary to account for any changes due to the Mark III containment configura-tion. A radiological release source term to the environment will be calcula-ted for each release category. The present intent is to use the M and C computer codes for this purpose, but these may be modified to reflect radio-nuclide release behavior data available at the time the work is performed. 3.8 UNCERTAINTY ANALYSIS 57
All quantitative results will be reported in terms of point values of a prob-
) ability distribution function, including expected (mean) or median (50th per-centile) value and upper (95th percentile) and lower (5th percentile) uncer-tainty bounds. These points values will be determined based on a propagation of component f ailure data, including error ranges, through the f ault trees and event trees. The uncertainty propagation will be performed using standard statistical distribution functions (e.g. lognormal) or numerical (e.g. Monte Carlo) techniques.
- 3.9 SENSITIVITY ANALYSIS The results of the study will be reviewed to identify the accident sequences which are the dominant contributors to overall risk and, within those se-quences, the significant system and componert failure modes. Comparisons with existing risk studies, including WASH-1400, will be made to identify and explain any significant differences.
The sensit ivity of the results to assumptions regarding component or common cause failures 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 results. s V 15 B-4 Am. No. 57, (5/81)
ACNGS-PSAR 4.0 SCHEDULE '
/ \
() The Reliability Ar.alysis Program will commence in mid-1981. The initial phase 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 ACNGS design. The overall study, including radionuclide release quantifica-
- tion, will be completed within two years of CP issuance. Thereaf ter, the reliability analysis will be kept updated to reflect changes in design includ-ing any changes recommended as a result of evaluation of the findings of the base line study.
5.0 APPLICATION OF RESULTS TO FINAL DESIGN There are currently no established regulatory requirements or acceptance criteria for judging the acceptability of quantitative system reliability analyses. Thus the need for implementing changes in design or operating, testing or maintenance procedures to achieve improvements in system reliability will be based on judgemental acceptance criteria which are not directly related to licensing requirements. These 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. Following completion of the base line reliability analysis, the results will be reviewed and various options available for improvement in reliability will , be evaluated with respect to the establishcd acceptance criteria. Recommend- , ations will be made regarding changes in design or operating procedures and
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;
4 the reliability analysis will be revised to reflect those selected for imple-mentation. 57 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 reliability analysis will be revised and the results reviewed
< for acceptability and need for further modifications as described above. 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.
6.0 QUALITY ASSURANCE Since use of results of the Reliability Analysis Program as described hereir is not considered a licensing requirement, the quality assurance requirements of 10CFR50 Appendix B do not apply to the program. However, the results of the program, including all calculations will be subject to review and verifica-tion in accordance with normal ACNGS quality assurance program practices, i Documentation will be maintained current so that all results can be reproduced t and all assumptions checked from original references. i j G l 15B-5 Am. No. 57, (5/81) l
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' EFFECTIVE PAGES LISTI'NG CHAPTER 17 QUALITY ASSURANCE
;- g Amendment
.1* 57 2* 57 3* 57 i 33 11 45 iii 45 17.0-1 57 17.0-2 37
'17.1-1 57 17.1-2 57 4
17.1-2a 57 17.1 'i 57 17.1-4 57 17.1-5 57 17.1-6 57 17.1-7 57 17.1-8 57 17.1-9 57 17.1-10 57 17.1-11 3j 17.1-12 57 17.1-13 57 17.1-14 57 17.1-15 57 17.1-16 57 17.1-17 57 17.1-18 57 17.1-19 57 17.1-20 57
, 17.1-21 57 ; 17.1-22 57
- 17.1-23 57 17.1-24 57 l 17.1-25 57
- 17.1-26 57 17.117 57 17.1-27a 57 17.1-28 57 l 17.1-29 57 i 17.1-30 57 17.1-31 57 11.1-32 57 !
j 17.1-33 57 l 17.1-34 57 17.1-35 57 l 17.1-36 57 17.1-37 57 17.1-38 57 17.1-39 57 17.1-40 57 l 1 Am. No. 57, (5/81)
ACNGS-PSAR EFFECTIVE PAGES LISTING ) l CHAPTER 17 QUALITY ASSURANCE l g Amendment 17.1- 41 57 17.1- 42 57 17.1- 43 57 17.1- 44 57
'17.1- 45 57 17.1-46 57 17.1-47 57
, 17.1-48 57 17.1-49 57 17.1-50 57 17.1-51 57 17.1-52 57 17.1-53 57 17.1-54 57 17.1-55 ' 57 17.1256 57 17.1'57 - 57 17.1-58 57 17.1-59 57 17.1260 57 17.1061 57 ., i 17.1-62 57 k 17.1-63 57 17.1-64 57 r 17.1-65 57 1 17.1-66 57 17.1-67 57 17.1-68 57 17.1-69 57 17.1-70 57 17.1471 57 i 17.1272 57 l 17.1-73 57 l 17.1-74 57 17.1-75 57
2 Am. No. 57, (5/81)
ACNGS-PSAR EFFECTIVE FIGURES LISTING g CHAPTER 17 QUALITY ASSURANCE l Figure Amendment 17.1.1A-1 57 17.1.1A-2 57 17.1.1A-3 57 17.1.1B-1 33 i 17.1.13-2 33 17.1.2B-1 45
- 17.1.3B-1 33 17.1.7B-I 33 17.1.10B-1 33 17.1.1C-1 33(deleted)
, 17.1.10-2 33(deleted) 17.1.1C-3 33(deleted) i 3 4 i i 3 Am. No. 57, (5/81) 1
. ,. . , . .y -r_..-.-_ _.~- _ .. ._- - . __ . - . . _ _
ACNCS-PSAR L. 1- 17.0 QUALITY ASSURANCE g
;{ l Each -section in Chapter '17' has been assigned suf fixes to the Lsection jd numbers to. identify that the sectioniis applicable to the respective .
(organizations as follows: . A - Applicant, Bl-. Architect / Engineer'(A/E),- 4
- v. 33(U) lC - Nuclear Steam Supply. System -(NSSS) Supplier.
A'brief introduction has been prepared to introduce the - applicant A/E and
~
~
NSSS vendors respectively. 17.0.A. . HOUSTON LIGHTING & POWER COMPANY ^ Houston _ Lighting 6 Power Company (HL&P) as the ' Applicant, has the -Quality-. Assurance responsibility for design, engineering, procurement, fabrication,
- -
construction, preoperational testing and operation of the Allens Creek j
" Nuclear Generating Station-(ACNGS).. Although HL&P .will delegate certain of -
" its QA activities and authority to its contractors, .it retains reeponsibi-lity for the QA program. controlling all ~ aspects of the ACNGS. , s The HL&P Qual'ity Assurance Program requires that HL&P, .its prime con-
- tractors subcontractors and vendors comply with the criteria established
[ by 10 CFR 50 Appendix B. It is the intent 'of HL&P to comply with ANSI N45.2- l j l' and the applicable daughter standards and implementing Regulatory Guides. ~ Furthermore, HL&P shall assure . through programmatic direction .that Ebasco i s, and all of ~ its subcontractors and suppliers performing nuclear safety j related work comply with 10 CFR 50 Appendix B, ANSI N45.2, and the Regula- 3 tory Guides as referenced herein consistent with their scope of work. Programmatic direction is defined as the role of HL&P in establishing -the 3 . program requirements and ensuring the adequacy of the prime contractors QA Program. The programmatic direction consists of review and approval of the 57 L system features initially and continued monitoring of those systems during . l' implementation' and further refinement or revision of the systems if the j systems need strengthening. i The HL&P QA Program is described in the corporate Nuclear Quality Assurance Program Manual (NQAPM). The NQAPM requires the establishment of a Project l Quality Assurance Plan for each project to describe the QA ' program to be implemented 'during the design and construction phase of each project and an
- . Operational Quality Assurance- Plan to describe the QA Program to be imple-I mented during the preoperational testing and ' operational phases of 'the ~
ACNGS. The Project QA Plan (PQAP) is described in Chapter 17.1 of the PSAR. The Operational QA Plan will be described in Chapter 17.2 of the
-FSAR. The PQAP specifies requirements applicable to prime ' contractors and HL&P. The HL&P quality assurance staff shall assure through implementation l
review that the HL&P staff and contractors are complying with the QA pro-
- l. gram and the PQAP. Implementation reviews are performed by qualified l-l- personnel based on experience, edu:acional level, training, and proficiency
!- 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 form.
l The combination of the QA programs as described in the NQAPM, PQAP, and
. 0QAP as augmented by definitive procedures provide HL&P with the assurance (U)-Update-17.0-1 Am. No. 57, (5/81) we.e w w
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ACNGS-PSAR
, that .its ' quality commitments are met. 57 I 17.0.B" -EBASCO O '
.The Quality Assurance program for safety related activities and services performed by Ebasco in the - design,' engineering, procurement, and construc-tion of the Allens Creek Nuclear Generating Station is nca described in the Ebasco Nuclear Quality Assurance Program Manual for the Allens Creek Pro- 33(U) '
; ject. This manual is a revision to Ebasco's Topical Report No. ETR-1001,
- .which was accepted by the NRC on May 12, 1975. The revision will consist of modifying the Ebasco Site organization. The Site Construction Quality j' Control activities will be under -the administrative and technical control of the Ebasco Quality Assurance Engineering Department instead of the.
4
. Construction Quality Control Department. This organizational structure is 45(U) described in Table 17.1.2B-4. All Ebasco quality assurance related activi-1' ties performed prior to January 1, 1977 were done in accordance with the program described in this chapter. All Ebasco quality assurance activities subsequent to January 1,1977 will be performed in accordance with the _
U) latest HL&P and NRC accepted revision of Ebasco's Topical Report No. l
.ETR-1001, which'at present is Rev. 8, except for the site Construction lS7
. Quality Control- organizational changes described above and other approved modifications listed in Table.17.1.2B-3. Later NRC approved revisions to ETR-1001 may be incorporated when deemed necessary. ' M U) b If necessary to define any additional clarifications, or modifications to j the project Nuclear Quality Assurance Program Manual because of HL&P i contract requirements or to suit the unique Project conditions, they will
- D be submitted for NRC acceptance in accordance with established provisions 46(U)
- ' which require execution- of .an authorization form involving approval of specified authorities to asrure, among other things, that safety and/or f quality are not sacrificed or compromised. Approved changes will be incorporated in above referenced Tables, as required.
i 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. This objective is achieved by the implementation of the Ebasco Nuclear Quality Assurance Manual which defines the policy, procedures, and requirements by which i Ebasco will design, purchase and erect the Allens Creek Nuclear Generating Station. Implementation of the Ebasco Nuclear Quality Assurance Manual provides a quality program which is in compliance with the requirements of r 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".
.17.0.C GENERAL ELECTRIC l The Quality Assurance Program for safety related activities and services for the Allens Creek Nuclear Generating Station is described in the 33(U) i General Electric Nuclear Energy Divisions BWR Quality Assurance Program t Description, NED0-ll209-04A.
l49 i 4 O V i' ' (U)-Update 17.0-2 Am. No. 57, (5/81) L f
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ACNGS-PSAR p)- ( s_,/- 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 17.1. l A- ORGANIZATION 33(U) The major organizations involved in the ACNGS are: l a) 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 Ebaaco, GE, consultants and other contractors to assure that their QA Programs are implemented and have sufficient authority and organizational freedom to be ef fectively implemented. HL&P will perform surveillance of the activities of Ebasco, GE, consul-tants and other contractors during the manufacturing, fabrication and construction of the ACNGS. b) Ebasco Services Incorporated (Ebasco) As the Architect-Engineer, Ebasco is delegated the responsibility to
/"'s provide HL&P with engineering, design, procurements and QA services.
~
l (\' ) As the constructor, Ebasco, is delegated the responsibility to provide 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 57 delegated activities include the following:
- 1) design and engineering
- 2) procurement activities i
- 3) home office QA activities
- 4) vendor surveillance activities
- 5) construction activities
-6) site QA/QC activities c) General Electric Company (GE)
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 Fuel, GE has the responsibility to provide an acceptable QA program to HL&P for the activities that have been delegated to GE. w. (U)-Update 17.1-1 Am. No. 57, (5/81)
-- - - - . . - - -- . -- . - - , . . _ _ - - - - . . - - - . . - . - . . - - . . - .~
ACNGS-PSAR f% k) v These delegated activities include the following:
- 1) design and engineering activities
- 2) procurement activities
- 3) f abrication activities
- 4) vendor surveillance activities
- 5) QA activities 2
d) Consultants HL&P may utilize the services of qualified consultants or other con-tractors to sasist in the performance of appropriate quality tasks, such as audits, inspections, interpretations of test results, reviews, etc. 57 Figure 17.1.lA-1 illustrates how the above companies interrelate for i the ACNGS. Yigure 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 ACNGS. Figure 17.1.lA-3 is an organization chart of the HL&P Quality Assurance g'~' group for the ACNGS. wY The Project QA Manager, Allens Creek is responsible for providing the
programmatic direction and administering policies, goals, objectives and methods which are described in the Project Quality Assurance Plan. I 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 concrois to be established by HL&P and the prime contractors on the Allene Creek Project.
l Two levels of control have been implemented by HL&P to monitor the j effectiveness of the Quality Assurance Programs for the Allens Creek Project: (1) Corporate level control relatec to the overall activities } and performance 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 performance 4 of HL&P, Ebasco and its subcontractors. This is accomplished through
review of documents, and implementation reviews thst establish QA i system features (e.g. procedures, specifications). ] i
- o 17.1-2 Am. No. 57, (5/81) i
ACNGS-PSAR .I Designated QA individuals are involved in day-to-day plant activities im- < -( U portant to' safety, (i.e.'the QA organization routinely attends and participates in daily plant work schedule and status meetings to assure they are kept abreast of day-to-day work assignments throughout the plant and that 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 17.1.1B-la show the project QA organization and indicate which personnel are "onsite" and "offsite". The PSAR Section 13.0 shows pro-ject personnel from other organizations. The criteria for determining staffing for the QA organization includes: a) Establishing the number of QA/QC personnel based upon the project 57
- schedule to ensure that personnel are available, qualified, and cer-tified to perform quality related inspections and evaluations.
b) Establishing the need for specially qualified QA/QC personnel based upon the schedule for activities requiring special or unusual exper-tise as far in advance of the activity as possible. c) Establishing the number of QA personnel based upon the number and criticality of problems identified during rcutine activities in order to perform additional or supplemental inspections, reviews, or eval-uations as required to ensure implementation of project requirements.
, C} Staffing projections are periodically reviewed based upon the project schedule and are re-reviewed and revised, as necessary, as the project schedule changes.
QA management personnel participate in short and long range scheduling ac-tivities. Staffing levels for QA/QC are a prime consideration in determining the level of effort for quality related activities. Prior to allowing quality l related activities to be conducted, adequat.e numbers of qualified QA/QC per-i sonnel must be available. Adequate QA/QC staffing must be available to pre-
- vent QA/QC personnel from being required to perform inspections without ad-equate preparation time or under pressure to complete inspections within a
, scheduled time period. Adequate QA/QC staff must be available to allow for i prompt closeout of open nonconformances and proper followup to ensure correc-tive action has been taken. O l'i.1-2a Am. No. 57, (5/81)
ACNGS-PSAR l b)\ \. 17.1.lA.1 Manager. Quality Assurance 1 The Manager, Quality Assurance, has the authority and responsibility to iden- 1 tify, initiate, recommend, or provide solutions to quality related problems and verify the implementation and effectiveness of the solutions. This posi-tion has the authority to "stop work" for cause in engineering, design, pro-curement, fabrication, construction, and operation phases of the nuclear plant. The z.inimum requirements established for this position are: a) a college degree in a field of engineering or science or equivalent experience. b) -familiarity with nuclear power generation facilities and the related operations. c) kr.owledge of the industry's quality assurance standards and regulatory requirements. d) management experience and f amiliarity 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 b) Houston QA Manager N c) Operations QA Manager The Manager, Quality Assurance, reports to the Executive Vice President. 17.1.1A.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, b) familiarity with nuclear power generation facilities and related operations. c) knowledge of the Quality Assurance standards and regulatory require-ments. d) management experience and familiarity with HL&P corporate organizations. .f \ 17.1-3 Am. No. 57, (5/81)
ACNGS-PSAR O
- (v;' The major. responsibilities of the Project QA Manager 'are:
a) administer QA policies established by management and ensure 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-l tractor management and interface with NRC. ! c) . coordinate activitien relating to auditing and vendor surveillance in conjunction with the HL&P Houston Quality Assurance Manager. The Project Manager has the authority to solve quality-related problems and to verify the implementation and effectiveness of the solutions. He has the authority to "Stop Work" for cause on any quality-related activity of the Allens Creek Project. 57 17.1.lA.3 Houston Quality Assurance Manager The Houston Quality Assurance Manager reports on all technical and administra-tive matters directly to the Manager, Quality Assurance. Th6s organizational
- arrangement provides independence from cost and scheduling isfluences.
The Houston Quality Assurance Manager is responsible for directing all HL&P p Houston of fice auditing, vendor surveillance and technical . support activi-ties. He has the authority to "Stop Work" for cause on any quality-related
! t s' activity of the Allens Creek Project.
The Houston Quality Assurance Manager af a minimum has: a) a college degree in a field of engineering 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. i The major responsibilities of the Houston Quality Assurance Manager are: I a) provide administrative guidance and direction for the HL&P Quality Assurance audit program. b) direct the HL&P vendor surveillance program. I c) provide technical support in the review of specifications, procedures, i i manuals, procurement documents, etc. I 17.1-4 Am. No. 57, (5/81)
_, _ _._ _ _. ._~.- . _ _ _ . _ _ - - - - - - - - - - - - _ _ _ _ ACNGS-PSAR 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. He has the autho-rity to "Stop Work" for cause on any activity related to f abrication and construction. 1
, 17.1.1A.5 Supervisor. Quality Systems The Supervisor, Quality Systems reports directly to the Project QA Manager. ,
He is responsible for providing technical direction and administrative guid-ance to the site Quality Systems personnel; developing and administering the 3L&P Project QA Plan; evaluating the Ebasco QA/QC program; administering the 57 , HL&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 authority to "Stop Work" for cause on any activity related to fabrication or construction. 17.1.1A.6 Discipline Project Quality Assurance Supervisors . The Discipline Project Quality Assurance Supervisors report to the Project l Quality Assurance Generel Supervisor. They are responsible for technical direction and administrative guidance to the Discipline, Quality Assurance personnel in their respective discipline group; coordinating implementation aN reviews; interface with NRC during audits; identifying deficiencies; reviewing
- and approving procedures applicable to their respective discipline; and pro-
- viding programmatic direction to Ebasco. They have authority to "Stop Work"
] for cause on any activity related to fabrication or construction. f 17.1.1A.7 Procurement Project Quality Assurance Supervisor i j The Procurement Project Quality Assurance Supervisor reports directly to the l Project QA Manager. He 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 j regarding vendor surveillance and auditing functions. He has the authority to j "Stop Work" for cause on any activity related to engineering, design, or
- procurement.
l . { 17.1.1A.8 Manager, Allens Creek Project 1 l The Manager, Allens Creek Project reports to the HL&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. 17.1-5 Am. No. 57, (5/81)
ACNGS-PSAR
/
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 requirerants of the project specifica- ' tions, proceduras and policies. Ensures that interfaces and coassunication 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 4 The Project Engineering Manager reports to the Manager, Allens Creek Project. He directs project engineering personnel in the performance of the HL&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 Work" 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 interface with pr me contractors. These activities include adequate engineer-i ing planning, coordination of problems, work priorities and activities of the
% HL&P Project Engineering' group assigned to each Supervising Project Engineer.
The Supervising Project Engineer (s) monitor the Prime Contractors resolution , w of pertinent QA noncompliances, participate in the HL&P Incident Review Con-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 heign of all items within their area of responsibility. 17.1.lA.11 Project Construction Manager The Project Construction Manager reports to the Manager, Allens Creek i 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 administrative guidelines for HL&P and Prime Contractors in the areas of construction, secu-rity, start-up, accounting, construction control, and ensures that the prime 1 contractor's management properly implements the dispositions to various non- l 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-l struction conforms to the plans, specifications and procedures that govern
- work activities. Has the authority to "Stop Work" for causing relating to
, construction.
[] l \v/ 17.1-6 Am. No. 57, (5/81) i I l
ACNCS-PSAR f3 k 17.1.1A.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 Project. 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 administrative support.
17.1.1A.15 Project Controller The Project Controller reports to the Manager, Allens Creek Project. He is
) responsible for the coordination and execution of the accounting and financial s administration. The Project Controller has no direct quality assurance re-sponsibilities on the project.
17.1.lA.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. , 0 ! 17.1-7 Am. No. 57, (5/81) L
,J . - km o a w i
ACNGS-PSAR
\
V 17.1.lB ORGANIZATION The Ebasco organization which includes the quality assurance organization satablished for the ACNGS 'is shown on Figure 17.1.1B-2 as amended by ETR-1001, 49 Figure 1.2-4, Revision 3. Reporting to Ebasco's President, through the Senior Vice President of Engineering and Construction, are seven independent lines of cuthority; the Vice President of Construction; the Vice President of Nuclear Engineering (who is in charge of Materials Engineering and Quality Compliance l33(U) Department and the Licensing Department); the Vice President of Engineering (who is in charge of engineering and design section); the Vice President of Purchasing; the Vice President of Plant Operation and Betterment; the Vice President of Consulting Engineering; and the Vice President of Projects (who is in charge of the project management and coordination). The Ebasco Project Manager for HL&P reports to the Vice President of Projects through the Manager of Projects. The Ebasco Project Manager is responsible for all matters relat-ing to the overall execution of the project and is the primary contact with the Client. o 5 I The Ebasco Quality Program is prepared by the Quality Program Committee. This is a permanent committee representing each Vice President's department. This committee is responsible for and has authority to make changes to the policies End procedures in this program. Ebasco's quality program and policy is es-tablished and documente.1 in the 'Tbasco Quality Assurance Manual for Nuclear Power Stations" which is the Quality Assurance manual to be implemented for 16 the Allens Creek Nuclear Generating Station and referred .to hereafter as the OEbascoQualityAssuranceManual. The primary responsibility for implementing the quality assurance policy rests with Ebasco's Vice President - Nuclear. In addition, the Vice President - Nuclear is responsible for Ebasco's Nuclear Consulting Department and Licensing Department. The Ebasco Materials Engineering and Quality Compliance Department is techni- 5 cally and administratively independent of the Construction and Engineering Q1 11,11 Organizations. Reporting to the Vice President - Nuclear is the Director of Ql-17.23 Materials Engineering and Quality Compliance who is responsible for adminis- q1 11,11 tering the Quality Assurance Policy of Ebasco and the Materials Engineering Ql- 17. 2. 4 2 and Quality Compliance functions. He has been vested by Corporate Management with the authority to enforce the requirements of the quality assurance policy. This Director has the unqualified support of Corporate Management. His decisions may not be overridden by other individuals or groups without the written consent of Corporate Management. Reporting to the Director of Materials Engineering and Quality Compliance is a 5 Chief Materials Application Engineer, a Chief Quality Assurance Engineer and a Chief Vendor Quality Compliance Representative. The Chief Materials Applica-4 tions Engineer is responsible for metallurgical functions such as the develop-ment of materials and welding specifications, and processes such as welding 5 and joining. The Chief Vendor Quality Compliance Representative is responsi- The q1 11,11 ble for implementation of the Ebasco Vendor Quality Compliance Program. 91,17, Chief Quality Assurance Engineer directs the Quality related 2.20 t'
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x-(U)-Update 17.1-8 Am. No. 57, (5/81)
I 1 ACNGS-PSAR ' (/ activities of the Materials Engineering and Quality Compliance Department, including overall program planning. He is also responsible for the Ebasco 5 Vendor and site quality compliance programs, and the implementation of the Ebasco site-quality compliance program. He or his designee also represents Ebasco in AEC and in Client audits - In addition, he designates a Project Quality Assurance Engineer for each nuclear project. l The quality assurance policy functions at two levels, (a) Quality Compliance 5 and'(b) Quality Control. Quality Compliance, (a), is an auditing and advisory q1 11,1 function over engineering, construction, purchasing, vendor and sub-contractor Q1-17.2.1 quality control operations. It is administered by the Chief Quality Assurance Ql-17.2.2( Engineer and the Chief Vendor Quality Compliance Representative. A Project Q1-17.2.26 Quality Assurance Engineer, who reports to the Chief Quality. Assurance Engineer is responsible to plan and ascertain implementation of the quality l .18 assurance policy, and is assigned to cocrdinate the effort of the Materials 5 Engineering and Quality Compliance Department. The responsibilities of the Project Quality Assurance Engineer, include: 91-11.11 17.2.4 a) Supervising and Coordinating the Quality Assurance activities within 17.2.5 the Materials Engineering and Quality Compliance Department l'
- b) Designating the Quality requirements in Ebasco component specifications and reviewing Vendor's Quality Assurance procedures to assure implemen-tation 5
c) Conducting audits of various Ebasco engineering disciplines gt.ig,gg. d) Surveillance of Vendors and Contractors quality assurance programs and 17.2.4 auditing them to verify the implementation of their programs 17.2.5 e) Auditing of Ebasco's Vendor Quality Compliance Representatives
~ f) Auditing of project quality assurance records g) Preparing quality compliance plans for use by Ebasco's Vendor Quality J Compliance Representatives h) Conducting audits of the Ebasco Site Quality Control and Quality Com-pliance organization and activities i) Distributing and controlling the Project Quality Assurance Manual, including the enforcement of all sections . 5 Quality Control, (b), includes the inspection function, and is the responsibi- 91,17*gg lity of the individual departments. This is applicable to site activities
- i such as Engineering, Construction, and Purchasing. At the construction site, the quality control organization is under the Project Superintendent, who is 17 responsible for: lQ2-11.17 a) Performing inspection in all engineering disciplines, establishing and enforcing quality control documentation requirements 17.1-9 Am. No. 57, (5/81)
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h ACNGS-PSAR 17
.('~Sj Q2-11.17
\ b) Re0olution of nonconformities when ' informed by the Quality Compliance
\~'f Site Supervisor that nonconformance to requirements exists and that 5 material components, or systems shall be corrected, or that work shall Ql-11.11 be stopped. 17.2.6 17.2.27 The Project Quality Assurance Engineer, by following procedures established in 5 the Ebasco Quality Assurance Manual, has the responsibility and authority to reject unsatisf actory work already performed on safety-related structures, Ql-11.11 systems and components when audits indicate nonconformance with the applicable 17.2.38 specifications, drawings and procedures. In a Vendor's shop an Ebasco Vendor Quality Compliance Representative may " reject" an item but does not actually order work to be stopped. He is only indicating that if work centinues in the l present manner, the item will not be accepted. The above represectatives need no action or approval of others prior to the implementation of the rejection directive. The Site Quality Compliance Supervisor has the authority to re-quire the discontinuance of unsatisfactory work at the construction site.
The organization under the Vice President of Engineering is responsible for the planning and imp'lementation of the design and engineering activities of the Ebcsco Quality Program as described in Section 17.1.3B. The individual positions of groups performing quality related design activities, and indepen-dent design review, checking and auditing activities, are described in Section 17.1.3B (Design Control), and in the Ebasco Quality Assurance Manual. p Quality assurance aspects of the procurement process are subject to review by 5 I the Materials Engineering and Quality Compliance Department. Personnel per-
' \s, forming quality assurance functions concerning suppliers of equipment and i services have sufficient and well-defined responsibility, authority and or- ! genizational freedom to identify quality problems; initiate, recoseend or 4
provide solutions through designated channels; verify implementation of solu-tions through designated channels, until the proper disposition of the defi-l ciency or unsatisfactory condition has been approved. The quality related procedures for procurement activities are presented in the Ebasco Quality Assur'ance Manual. l The Ebasco Project Superintendent who reports to the Construction Manager is responsible to administer all field construction contracts. The Construction Contractor will be treated in essentially the same manner as any other sup-j plier of safety-related equipment or services. The Construction Contractor 5 l will be required to have a quality program which addresses the requirements of Q1-11.11 Appendix B to 10 CFR 50. This program will be subject to review by the 17,2,9 I t l l 17.1-10 Am. No. 57, (5/81)
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ACNGS-PSAR 5 Quality Compliance Engineering Department. During the course of construction, 91-11+11 t]/ the activities of the Construction Contractor will be under surveillance by Ebasco Quality Compliance Representatives on the site under the direction of 17.2.9 the Quality Compliance Site Supervisor, who in turn reports directly to the Chief Quality Assurance Engineer. In addition, periodic audits will be per-formed by the Chief Quality Assurance Engineer or his designee of the Con- 5 struction Contractor's quality program to assure that it is properly implemen-ted. Qualification requirements for the position responsible for directing and 5 managing the Ebasco Quality Assurance Program, are: Bachelor of Science 91 11,11 Degree in Engineering; 10-15 years experience in quality related work or 17.2.8 equivalent experience in the engineering or construction of a nuclear power plant, including at least 10 years experience in responsible managerial or project positions; thorough knowledge of the Ebasco Quality Assurance Program. Refer to Section 17.1.2B for applicability of this section. 33(U) O (U)-Upda t.e 17.1-11 Am. No. 57, (5/81)
ACNCS-PSAR A \ t
\j 17.1.2A QUALITY ASSURANCE PROGRAM The HL&P Project Quality Assurance program for the Project has been developed in accordance with the criteria of 10CFR50 Appendix B, ANSI N45.2 and Regu-latory Guides as referenced herein, to provide programmatic direction on quality requirements for the prime contractors and subcontractors during design and construction.
The nuclear safety-related structures, systems and components covered by this l program are listed in Section 3.2, Table 3.2-1, column designated " Quality Assurance 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 57 the responsibility for QA of all items designated "P" and the GE items upon receipt at the project site. The HL&P Quality Assurance program for the Allens Creek Project is described by the HL&P Project Quality Assurance Plan. The plan requires that written procedures, training and certification, issuance of specifications and draw-ings, and work and inspection planning be accomplished in advance of perform-l ing nuclear safety-related activities. HL&P Project Quality Assurance ensures through procedure reviews that this advance preparation is accomplished. The Project Quality Assurance Plan for the Allens Creek Projact is structured in accordance with the NRC regulatory position of the Regulatory Guides as described in Appendix C of the PSAR and with ANSI N45.2.12. (Draft 3, Rev. 4 - February, 1974). The HL&P QA Program and Procedures which are used to implement the quality related activities for each major organization and the reference to the appli-cable criteria of 10CFR50 Appendix B are listed in Table 17.1.2A-1. Verifica-tion that plans and procedures are properly implemented is accomplished by HL&P Quality Assurance through audits, inplementstion reviews and regular management assessment of the Quality Assurance Program. Implementation reviews are performed at the construction site by personnel qualified based upon experience, education level, training, and proficiency 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 and at vendor facilities are qualified in accordance with ANSI-N45.2.6. It is the policy of HL6P as applicant, to assure that the design, engineering, procedurement, fabrication, construction, preoperational testing, and oper-ation of ACNGS are in conformance with project specifications, procedures, i codes, and NRC regulations. It is the responsibility of each organization l lO 17.1-12 Am. No. 57, (4/81) l L - .-- _ _ _ . _ _ - - - _ _ -
ACNGS-PSAR s
) assigned to the Allens Creek Project to ensure that project procedural review l f
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 in-plementing 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. i ; Allens Creek Project Quality Assurance is responsible for conducting a quality
- oriented indoctrination program for new personnel that have quality-related 4
functions. The HL&P Projaat Quality Assurance Plan requires that prior to performing activities affecting quality the personnel are trained in the applicable procedures. The training provides a thorough understanding of the purpose, scope, policies, principles, and techniques of the specific proce- 57 i dures or instructions. When personnel perform special process activities, a ' training and certification program is established, and refresher training is conducted to ensure that proficiency is maintained. HL&P Quality Assurance audits are performed to ensure compliance with these criteria. l The Project QA Manager and the Houston Quality Assurance Manager are directly ! responsible for assuring effective implementation of the Quality Assurance program. The qualifications for these positions are defined in Sections 17.1.1A.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 activities to HL&P Project Quality Assurance for review Procedures are re-3 viewed by Project QA personnel during preparation for inspections, surveill-l ance, implementation reviews and audits to ensure consistency with project requirements. Additional selected procedures are reviewed and concurred 4 with prior to issuance. It is the responsibility of HL&P Project Quality Assurance to determine that the prime contractor's procedures require proper equipment, environment a,nd ,other pr,erequisites to perform the associated 1 activity. These requirements are verified through imple'entation m reviews by l HL&P Discipline QA and audits by HL&P Houston QA. 1 The results of the HL&P implementation reviews and audits are presented in a monthly report to the HL&P Executive Vice President. , Regular executive man-agement review of the a ethly activities and the direct involvement of the
- HL&P Executive Vice President assures than an objective program assessment of ,
4 the Allens Creek Project Quality Assurance programs is being performed, j
HL&P Project Quality Assurance reviews and documents concurrence with the
Ebasco Quality Assurance manual and audits are performed by ni&P Houston Quality Assurance to ensure compliance, i ! 17.1-13 Am. No. 56, (5/81)
l ACNCS-PSAR i m
,, HL&P and Ebasco Project Quality Assurance will establish and document a pro-gram for transferring responsibilities and controls for quality-related ac-tivities from Ebasco to HL&P during phaseout of design / construction and during preoperational testing and plant turnover. This program will be implemented prior to preoperational testing. This program will be in accordance with and l consistent with the requirements of this section and/or Section 17.2 of the FSAR.
t HL&P is committed to maintaining the Project Quality Assurance Plan as an 57 effective and meaningful document to provide directions to HL&P and the prime j eentractors on the Allens Creek Project. When proposed substantive changes to i fais Project Quality Assurance Plan affect the docketed Quality Assurance 1rogram description, HL&P will notify the NRC of the change (s) for their 1 review and acceptance prior to implementation. Organizational changes of a : 4 ~ 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 appropriate sections of the QA Program and Plan. This matrix illustrates how the HL&P QA Program and Allens Creek QA Plan are in compliance with the Regu-latory criteria. HL&P positions' on Regulatory Guides (RGs) are enumerated in Appendix C. 17.1.2B QUALITY ASSURANCE PROGRAM 5 The Ebasco Quality Program described in the Ebasco Quality Assurance Manual has been established, documented and implemented. The Quality Program has 91-11 11 been structured to comply with the requirements of the Atomic Energy Com- 17.2.9 mission's Rules and Regulations, Title 10, Part 50, Appendix B, " Quality { Assurance Criteria for Nuclear Power Plants." The requirements of the Ebasco
; Quality Assurance Manual cover quality related activities for safety-related ; structures, systems and components from the various stages of designs, pro-
- curement, shipment, storage and installation. The table of contents for the Ebasco Quality Assurance Manual appears in Table 17.1.2B-1. l 45(U)
The quality requirements for the ACNGS are contained in specifications, codes, l33(U) 4 standards and the Ebasco Quality Assurance Manual. The Ebasco Quality Assur-2 ance Manual includes provisions which require planned and periodic audit of 5 the Ebasco Quality Program. A Management Quality Assurance Audit Committee 91-11.11 (comprised of one representative from the construction, engineering and 17.2.9 I nuclear departments) regularly assesses the implementation and effectiveness of the Quality Program to assure that the Program is meaningful and ef fective. , i l i f t s-(U)-Update l 17.1-14 Am No. 57, (5/81)' i
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ACNGS-PSAR A Quality Program Conusittee which is a permanent Committee representing the Vice President of Construction, Vice President of Purchasing, Vice Presi- 5 dent of Nuclear Engineering, Vice President of Engineering and Vice Prest > f]f dent of Projects is responsible for distributing to various Ebasco Depart-qi,
, 31 (U ments the Ebasco Quality Assurance Manual for implementation as applicable. t{.2.46 The. Senior Vice President of Engineering and Construction has declared compliance with the requirements of the Ebasco Quality Assurance Manual mandatory and has directed the Director of Materials Engineering and Quality Compliance, who has his unqualified support, to enforce its re-quirements.
The Ebasco Quality Assurance Manual includes a table which identifies sections of the manual employing with each of the criteria of 10CFR50 5 Appendix B (Table 17.1.28-2). Safety related structures, systems, and com- 01-l 45(U) ponents covered by the requirements of the Ebasco Quality Assurance Manual 11.11 During engineering, design procurement, inspection, and testing phases are 17.2.31 identified in Table 3.2-1 of this PSAR. l45(U) Personnel assigned to the project are knowledgeable in their respective fi' elds and versed in the quality requirements for the plant. The indoctri-nation and training program devised and implemented by various departments participating in the implementation of the Ebasco Quality Assurance Manual assures that personnel conducting quality related activities are familiar with and knowledgeable in the purpose, scope and implementation of the appropriate quality related manuals, instructions and procedures. The 5 indoctrination and training program also assures that personnel performing 91 quality related activities are adequately trained and qualified in the 11,11 principles and techniques of quality related activities. An indoctrination 17.2.31 tg) and training program procedure which will be incorporated in the Ebasco V Quality Assurance Manual is presently being prepared. The Ebasco Quality Assurance Manual and Quality Assurance Plan for the ACNCS will be structured in accordance with the Regulatory Guides as dis- l33(U) cussed in Appendix C and Industrial Standards that are addressed in the NRC publications " Guidance on QA Requirements During Design and Procure-ment Phase of Nuclear Power Plants", (Revision 1) dated May 24, 1974 23l33(U) (WASil 1283) and " Guidance on QA Requirements During the Construction Phase of Nuclear Power Plants", dated May 10, 1974 (WASli 1309). Quality related activities initiated prior to the submittal of this PSAR are: a) Acquisition of site data b) Preliminary Engineering Work These quality related activities are controlled by the Ebasco Quality 5 Assurance Manual and supplementary guidelines, thus assuring that th'ey 01-meet the requirements of the applicable section of 10CFR50 Appendix B. 11.11 17.2.7 (U)-Update V 17.1-15 Am. No. 57, (5/81)
l ACNGS-PSAR f'N The Ebasco Quality Assurance Manual requires that quality related activ-(") ities such as inspections and tests are performed under suitable conditions and using appropriate equipment. This is achieved by inclusion of neces-5 91_11,11 sary requirements in procurement documents. Vendors of safety related 17.2.11 systems, structures, components and services are required to submit to 17.2.16 Ebasco for review, their procedures for activities such as inspection and testing. Ebasco Vendor Quality Compliance Representatives monitor these activities in Vendor shops to assure that the applicable Vendor procedures are implemented. The Ebasco Construction Quality Control Procedures are also reviewed by Materials Engineering and Quality Compliance for com- 5 pliance with the applicable codes and regulatory agency requirements. The Q1- 11.11 quality control activities at the construction site are also monitored by 17.2.11 the site quality compliance personnel. 17.2.28 The Ebasco Quality -Program is structured so that modifications can be made to comply with NRC regulations and industry standards as they are adopted. The Quality Assurance program for safety related activities and services performed by Ebasco in the design, engineering, procurement, and construc-tion of the Allens Creek Nuclear Generating Station is now described in 33(U) the Ebasco Nuclear Quality Assurance Program Manual for the Allens Creek Project. This manual is a revision to Ebasco's Topical Report No. ETR-1001, , which was accepted by the NRC on May 12,'1975. The revision will consist of modifying the Ebasco Site organization. The Site Construction Quality Control activities will be under the administrative and technical control (p of the Ebasco Quality Assurance Engineering Department instead of the Con-struction Quality Control Department. This organizational structure is des-c ribed in Table 17.1.2B-4. All Ebasco quality assurance.related activities 45(U) l performed prior to January 1,1977 were done in accordance with the program
; described in this chapter. All Ebasco quality assurance activities subse-quent to January 1,1977 will be performed in accordance with the latest 46(U)
HL&P and NRC accepted revision of Ebasco's Topical Report No. ETR-1001, which at present is Rev. 8 except for the site Construction Quality 45 14 9 Control organizational changes described above and other approved modifi- U) I cations listed in Table 17.1.2B-3. Later NRC approved revisions to ETR 4f 1001 may be incorporated when deemed necessary. [U) If necessary to define any additional clarifications, or modifications to the project Nuclear Quality Assurance Program Manual because of HL&P con-tract requirements or to suit the unique Project conditions, they will be submitted for approval in accordance with established provisions which 45(U) require execution of an authorization form involving approval of specified authorities to assure, among other things, that safety and/or quality are not sacrificed or compromised. Approved changes will be incorporated in above referenced table (s), as required. l45(U)
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(U)-Update 1 17.1-16 Am. No. 57, (5/81)
ACNGS-PSAR r ( 17.1.3A DESIGN CONTROL HL&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. 2) and ANSI N45.2.ll-74 on the prime con- l tractors and applicable subcontractors. HL&P has contracted with Ebasco and General Electric to perform the design, engineering, and design verification.- HL&P Engineering performs reviews of selected elements of the completed design, design documents, and specifica-tions to ensure that contractual requirements are met. The HL&P Project Engineering Manager is responsible for 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 HL&P project engineering activities are conducted in accordance with approved project procedures. When HL&P has direct responsibilities or assumes direct responsibility for conducting design activities, these activities will be conducted in accordance 57
; with the requirements of this section and/or the FSAR Section 17.2.3.
HL&P contractors are required to provide the following design control measures in their quality assurance programs: a) A design control system is established to document the methods of accomplishing and controlling essential design activities. b) Design documents such as calculations, diagrams specifications, and 5 drawings are prepared and records developed such that the final 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 structures, 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 N design. 17.1-17 Am. No. 57, (5/81)
l ACNGS-PSAR i e ( g)- Safety-related and/or seismic Category I designs are verified for b) adequacy and accuracy through independent objective review of design documents by individuals compete it in the subject activity. This verification may include the us- of siternate or simplified solution methods or qualification testin n 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 design.
i) Document distribution is controlled such that all individuals using a ' design document or its results and/or conclusions for further design work can be notified if the document is revised or cancelled. j) Design documentation includes evidence that design control requirements have been satisfied, 57 k) Errors and deficiencies in approved design documents, including design
.sethods (such as computer codes), that could adversely af fect struct-
. ures, systems, and components important to afety are documented; and , action taken to assure that all errors and deficiencies are corrected.
- 1) Devistsons from specified quality standards are identified and proce-dures are established to ensure their control.
m) A documented check to ensure dimensional accuracy (including toler-ance for accept /raject criteria and inspectability) and the complete-ness of the drawings and specifications, n) A system to ensure design requirements from engineering specifications and drawings for that system, component, or structure are included in inspection documents and that the cognizant engineering group perform an engineering evaluation and signoff on deviations identitud on the inspection documents. o) A system is established to require that design specifications and drawings 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 precedures and that the doc-umet.ts contain the necessary QA requirements such as inspection and test requirements, accepeance requirements, and documenting of inspec-
~
tion and test results. HL&P Houston Quality Assurance performs audits of HL&P, Ebasco, and General ' Electric to ensure that design controls, requirements, specifications, and documents are in accordance with the design control criteria. In addition HL&P Project Quality Assurance reviews quality / construction proce-dures to ensure that the quality requirements of the design specifications are incorporated. HL&P Project Quality Assurance also performs implementation reviews to ensure that the work is accomplished in accordance with the design requirements and to ensure that field changes to the design are processed in accordance with the design control criteria. G 17.1-18 Am. No. 57, (5/81)
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5 ACNGS-PSAR O V 17.1.35 DESIGN C0ltIROL [ i 1basco utilizes ~~a modified matrix type of organization for engineering and design, wherein specific engineers from the various disciplines are assigned
.to a project. Primary technical responsibility for project design rests with
! the Vice President of Engineering. The Chief Engineers of the various engi-
- neering departments (Electrical, Mechanical, Concrete-Hydraulic and Archi-
, tectural-Structural) and the Design-Drafting Manager report directly to the i
Vice President of Engineering. Each nuclear project has one Lead Discipline Engineer assigned to it from each engineering discipline. For a particular project he is responsible to 'a Supervising Engineer who in turn reports to his
' respective department Chief Engineer. Each design group, consisting of draftsmen, designers and engineers who work on the preparation j of drawings for the project, is under the supervision of individual Design ;
i Supervisors for. each discipline of the design (Electrical, Mechanical Instru-mentation, Concrete-Hydraulic and Architectural Structural). . Each Design i Supervisor reports to a Division Chief, who in turn reports to the . Design-Drafting Manager. q The development of plant design, including preparation of criteria, calcula-tions, drawings and specifications, follows a procedure which minimizes the l possibility of error. The procedures are established to assure that the ~ design activities will be carried out in a planned, controlled and orderly manner, as described below. A system of reviewing and checking design documents by parties other than the ] originator is conducted to minimize the possibility of errors and deficiencies 5 ' which could adversely affect safety related items. However, when an error or j deficiency is evident, it is documented as a matter of record, and appropriate l corrective actior h tehn. 4 The sys:em of design reviews and checks extends to and provides controls 4 applied to items, such as: stress, thermal, hydraulic, radiation, and ac-i cident analysis; compatibility of materials; accessibility for in-service inspection, raintenance, repair; and specifying criteria for inspection and test. The process of checking and/or verifying the adequacy of the design way
- include methods such as actual design reviews, alternate or simplified calcu-1 sting methods, or suitable testing programs.
l ! The following material is in accordance with the requirements presented in the ! Ebasco Quality Assurance Manual: ) j The first stage of the design process includes definition of the major struct-ures, systems and equipment. This is coordinated by the Project Engineer and , is accomplished by a working group composed of members of the Engineering and . i Nuclear Departments in consultation with Construction. The first written i definition of the plant occurs with the preparation of the PSAR. The PSAR
- section on design criteria and plant description and safety evaluations are prepared by respective Disciplir.e Engineers. These sections and criteria are ,
reviewed and approved by the Supervising Engineer for technical content and by ! the Nuclear Licensing Department for nuclear 17.1-19 Am. No. 57, (5/81)
ACNGS-PSAR safety and licensing aspects. Preliminary general arrangement drawings and
; diagrams are prepared by the Design Groups under the direction of the Lead v Discipline Engineers utilizing criteria developed during the PSAR presenta-tion phase of the project.
Tne Nuclear Licensing Department provides guidelines for classification of safety related structures, systems, and components and reviews the safety classifications assigned by the Project Engineer for the PSAR. They also participate in the review of the preliminary general arrangements and sys-tem flow diagrams to ascertain that nuclear safety requirements have been met. Inter-discipline coordination and resolution of internal interface problems are the responsibility of the Project Engineer who administers and documents these activities in accordance with Engineering and Projects Department's procedures. The general design work follows with preparation of specifications, flow diagrams, general arrangements, design information sheets and other de-tailed engineering documents by the Lead Discipline Engineers, with techni-cal review by the Supervising Engineers. General Arrangement drawings, safety system flow diagrems, electrical one-line diagrams and other pre-selected drawings and specifications will all be reviewed by the Licensing Department. At any stage in the design process, the project personnel may request consultation with the Nuclear Licensing Department for review of any safety related drawings or specifications. Each Lead Discipline Engi- 5 neer and Desim Supervisor reviews the design work of the other disciplines ll-as applicable .or interface compatibility with their work. Together with 11 11 / the Supervisor they compile and transmit information on criteria and inter- 17 2*13 face and determine problems involved in design and methods of resolving them. The Design Supervisor assigns responsibility for design drawings and calculations to appropriate engineers, designers and draftsmen. These per-sons initial all completed work performed by them. This work is then re-viewed for design verification and signed by checkers from the same design section who were not involved in the details of preparation, as assigned by tne design supervisor. Tne checker is a designer or engineer (other than those who performed the original design) qualified to check the originator's drawings or document, and who is familiar with the philosophy and criteria for the project. Each Lead Discipline Engineer (LDE) retains responsibility for work performed in his discipline and is responsible for correctly translating applicable regulatory requirements and design basis into specifications, drawings, procedures, and instructions. It is required that appropriate quality standards be specified in design documents and the design review measures assure that such standards have been included. Deviations or changes from 5 specified quality standards are processed in a manner similar to the manner in which other design changes are made. This process includes a documented 1 review by the same group or organization responsible for having reviewed ! tne particular area of the original design document. Each Lead Discipline ! Engineer is responsible for the inclusion of suitable design analyses where applicable. Tne Section of the Ebasco Quality Assurance Manual and proce-dures covering control activities require that proper selection and accom-pli shment of design veri fication or checking methods such as design re- l [g] views, alternate calculations or qualification testings are performed. The (/ Licensing Department provides the LD6 with acceptable safety criteria for 17.1-20 'U)-Upd,ie Am No. 57, (5/81)
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i ACNGS-PSAR- l
,m design and keeps the LDE informed of the latera applicable licensing requirements. These criteria are identified in the PSAR.
Safety related equipment specifications and drawings are also reviewed by other Ebasco departments. This includes review by the Materials Engineer-ing and Quality Compliance Department with respect to the nuitability of the application .of material parts, equipment and processes, and quality assurance requirements; by the Licensing Department with respect to_ad-herence to Nuclear Licensing safety guidelines criteria as established in the PSAR; and by the Construction Department with respect to construction quality control procedures, as appropriate to the drawings and specifica-
- tions. The Project Quality Assurance Engineer reviews and comments on the 5 design documents from the. viewpoint of materials application, quality as-surance requirements, and quality control of vendor supplied materials and equipment, including the specification and inclusion of appropriate quality standards in the design documents. Such reviews are documented on Specifi-cation and Drawing Routing Sheet Forms. Each Lead Discipline Engineer retains responsibility for including in the design documents inputs from other pertinent departmeets after they complete their review. If any
~ errors and deficiencies in the design process are uncovered, the Lead Discipline Engineer is responsible for their documentation and disposition, and initiating corrective actions. Deviations and changes from appropriate quality standards are processed and controlled in a similar manner to other design changes, as stated herein. Following a final review by the Design Supervision, Division Chief, Discipline Engineer, Supervising Engineer and (as appropriate) Chief Engineer, each signs the completed drawing and the
,[] Chief Engineer or Supervisor places his professions 1 engineer's stamp on the drawing where applicable. Similarly, completed specifications are signed by the Discipline Reviewing Engineer; and'in cases where certifica-tion is required, a registered professional engineer shall affix his signa-ture to the statement set forth in Engineering Department Procedure E-13, entitled " Cover Sheets fot Ebasco Specifications", for such certification.
Completed design documents are sent to HL&P for review, comments and approval. , Design interfaces between HL&P and Ebasco will be handled by concurrent review of criteria and of interface drawings with coordination of review 5 and approval by HL&P as applicable. ! Interfaces between Ebasco and General Electric are handled by the 16 respective Lead Discipline engineers and coordinated by the Project En-gineer. Tne Project Engineer is responsible for identification and reso-lution of all the interfaces, and coordinating the accommodation of these interfaces into the plant design through the Lead Mechanical Engineer. He is also responsible for documenting all correspondence resulting from this interface. All correspondence with outside parties will be documented by correspondence between affected individuals, including requests for tech-16 nical information. Ebasco transmits copies of drawings involving GE/Ebasco interfaces to G6. Ebasco also reviews drawings received from GE for inter-face compatibility. The-procedure for major design changes, revisions to speci ficatians, draw- 5 g ings and criteria, including field changes, follows the same design control ( ' process, including documentation and review by authorized personnel, as 17.1-21 (U)-Update Am. No. 57, - (5/81)
ACNGS-P2AR
,7 - s outlined above for the original design.
i )
\,_ ,/ A project distribution schedule list is maintained for specifications and
. drawings. Revisions are then sent to the appropriate persons on this list.
Drawings and specification schedules showing dates and number of latest revisions are issued monthly by the Coordinator. The Lead Discipline Engineers and Design Supervisors have the responsibilities to assure that the latest drawings and specifications are used in engineering and design. Major changes are those changes to engineering documents which will af fect safety related structures, systems, snd components. In design the Ebasco Lead Discipline Engineer for that part of design and engineering shall be responsible for judging whether a change to a specification or drawing due to a change in requirements or a fabricator change request is a major or minor change. He does this in consr.1tation as needed with his Supervising Engineer. 5 Any major change is processed through the same channels used in specifica-tion and drawing preparation and reviewing by the affected departments. The actual flow pattern for engineering documents is delineated in detail in the Ebasco Quality Assurance Manual. It should be noted that any standard "off the shelf" commercial or previously approved matarials, parts, and equipment, that are essential to the safety-related functions of the structures, systems, and components, are selected and reviewed for suitability of application under the same requirements employed for all such safety-related items. The Ebasco Quality Assurance Manual specifies that safety related items must be purchased from a Vendor whose QA program (N has been evaluated and must be qualified to meet design requirements. A system of plann?d and periodic management audits of the engineering and design control proce,s are performed in accordance with a documented procedure, as described in Section 17.1.18B - Audits. Engineering records, including design reviews, design documents, records and changes thereto,- are collected, stored and maintained in accordance 4 with the disposition based on the Ebasco Of fice Practices and Procedures. Drawings and specifications, once concurred upon, by Ebasco and HL&P engi- 5 neering, are sent to the field by written transmittal where they are 'l-received by the Resident Engineer. He opens file cards on each new docu- 11.11 ment received and distributes copies to appropriate field personnel by 17.2.25 written transmittal. When revised drawings or specifications are sent to the field, the revision number and date are entered on the appropriate file card, and copies of the revision are distributed to the users by an assis-tant to the Resident Engineer (Of fice Engineer) who marks " Void" on all superseded copies of the revised document. In this way, it is assured that the field staff are always working from the latest drawings and are cogni-zant of all design changes, and that obsolete documents are controlled to prevent their inadvertent use. Copies of all transmittals are retained in the Resident Engineer's file. 16 If the field staff needs to institute a design change, it is brought to the attention of the Project Superintendent. If he determines that a
~ [ 'j change is major, then the appropriate Lead Discipline Engineer is notified
(_,/ by means of a field change report form. The Project Superintendent can 17.1-22 (U)-Update Am. No. 57, (5/81)
, ACNGS-PSAR gN identify major changes since the specifications and drawings will identify 5 g safety-related structures, systems and components. Proposed major design \
changes are reviewed in accordance with the original design review proce-dures. Revised drawings are sent to the field when the change is approved. 16 Minor field changes require that a field change report form be completed. but need only be approved by the Project Superintendent. Minor field changes shall be documented and shown on' drawings as "As Built" revisions. Major and minor changes are easily identified because the specifications and drawings which are covered by the change are identified as safety or non-safety-related. When-Engineering drawings and specifications are developed in the field office, the same procedure for control of those drawings designated as important to safety and plant reliability will be followed as for home office - originated specifications and drawings. Refer to Section 17.1.2B for applicability of this section. (U)
- O lQ q
4 O U 17.1-23 (U)-Update Am. No. 57, (5/81) l I
ACNGS-PSAR f}17.1.4A PROCUREMEE DOCUMENT CONTROL V To. assure that nuclear safety-related items are purchased in a planned and controlled manner, the HL&P Project Quality Assurance Plan establishes basic requirements which are to be used by ht&P in prepcring 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 requirc=cnts. 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.
\'- 33(U)
I 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 identify the documentation to be prepared, maintained, rubmitted, and made availab'.e to the procuring agent for review and/or approval, such as drawings, specifications, procedures, inspection and test records, perscanel and procedure qualifications, and material and 33(U) test reports. g) Procurement documents contain the requirements for tbr retention, control, and maintenance of records. 5 (U)-Update 17.1-24 Am. No. 57, (5/81)
=- . . _- . -- .- .- -. . . , ACNGS-PSAR l
Procurement ' documents contain the procuring agency's righ't of 7% L h) access to vendcr's facilities and records for source inspection and audit.
- 1) Changes and/or revisions to procurement documents are subject to at least the same review and approval requirements as the original document .
j) Purchase documents for spare or replacement parts of safety-related structures, systems, . and components are reviewed for adequacy of 1
. quality requirements by qualified personnel knowledgeable in QA. qt7,9 The review is to determine the adequacy relative to the quality as- ; surance requirements and acceptance criteria of the original design.
i j k) The evaluation and selection of suppliers are determined by ! qualified personnel in accordance with written procedures. 1 ! 1) A written procedure acceptable to HL&P shall be used for source evaluation. ; i ! m) Procurement documents, records, and changes thereto are collected, stored, and maintained in a systematic and controlled manner. l HL&P Engineering is responsible for review and approval of Ebasco procure- { ment specifications. Engineering also coordinates nith HL&P Procurement QA for performance of a quality assurance review. HL&P Procurement QA co- ' ordinates with Ebasco and HL&P Engineering in the review of the procurement package. 4 In addition, HL&P Discipline QA is responsible for reviewing field procure-1 ment packages to ensure that all quality assurance reqt;irements have been included.
- HL&P Houston Quality Assurance is responsible for performing audits and vendor surveillance to verify that the requirements have been implemented
. and that they are ef fective.
17.1.4B PROCUREMENT DOCUMENT CONTROL The Ebasco Quality Assurance Manual provides measures to assure the control 5 j of procurement documents at outlined here below:
- Specifications and drawings are prepared by the cognizant engineering disciplines as specified in 17.1.3B above. All specifications and drawings receive _ a preliminary review by the Lead Discipline Engineers 5
as specified on the Project Distribution List. The Project Quality Assurance Engineer shall at this time review the data submitted and shall 01~11*11 1- include all pertinent quality assurance requirements to assure compliance 17.2.11 with 10CFR50, - Appendix B and standard Ebasco Specifications controlling l quality. i ! A The engineers performing the review shall document their review by signing 5 the- Specification and Drawing Routing Sheet. Upon completion of their re-
- - view, the engineers shall return all documents including the Routing Sheet l
l-l 17.1-25 Am. No. 57, (5/81)
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ACNGS-PSAR i p to the Lead Discipline Engineer who shall -incorporate the comments into the specification or drawing and maintain other documentation for record pur-poses. The revised specification with a routing sheet shall be sent to those re- 5 viewers :who have commented and to those of the original reviewers, who in the judgment of the originator, are affected by the comments. These 4 parties shall review for their areas of responsibility and upon concurrence sign the routing sheet attesting to their acceptance of the specification, j' and return the documents to the Lead Discipline Engineer for record pur-poses. The Project Quality Assurance Engineer (PQAE) shall assure that the speci-fication contains the necessary provisions to assure right of access; 5 i applicable inspection requirements; preparation, identification,- main- Ql-11.11 + tenance and retention requirements of documentation and requirements for- 17.2.14
- the submittal of applicable documents for review and that provisions for 17.2.19 l
! all the special processes as required have beea included. 17.2.20 This review of Ebasco Specifications and Drawings provides assurance that the applicable regulatory and desi &n basis requirements as well as other 5 ;
quality related requirements are suitably included or referenced in the Q1-11.11 i specifications and drawings. 17.2.11 i A copy of the reviewed specification is then forwarded to HL&P by the Project Manager, for approval. Upon receipt of HL&P approval, the speci-4
.)/
fication la ready to be issued for inquiry. The Lead Discipline Engineer shall forward the Inquiry Specification to Purchasing, who is responsible
- for preparing the procurement package for submittal to prospective Bidders. 5 j A copy of the procurement package is maintained by the PQAE for record 3
purposes. j Proposals received from Bidders are reviewed by the Purchasing Department, i the Design and Engineering 7 ups which prepared the Specification, and other groups as appropriate including the Project Quality Assurance Engineer. The Project Quality Assurance Engineer is primarily responsible for evaluating the Bidders' Quality Assurance Program in accordance with the specification requirements. Purchase Orders shall not be issued to Bidders whose Quality Assurance Programs have not been accepted by 5 16 Quality Assurance Engineering. The above groups evaluate Bidders whose Quality Assurance Programs have not been accepted by the Quality 91~11*11 Assurance Engineering Department. The above groups evaluate Bidders and 17*2*11 recommend to HL&P the best evaluated Bidder. Upon receiving approval for award from HL&P the Purchasing Department shall prepare a forar.1 17.2.20 purchase order package and issue it according to the Project Distri-bution' Schedule. 5 i > Any revision. to the specification or drawings which may be required af ter issuance of the Purchase Order shall be distributed as a formal supplement I to the Purchase Order in accordance with the Project Distribution Schedule af ter the revision has followed the steps outlined above. Purchase docu- 5 { ments for spare or replacement parts shall receive the same review require- q1 11,11 ments as the original document. 17.2.11 s 17.1-26 Am. No. 57, (5/81)
ACNGS-PSAR Refer to Section 17.1.2B for applicability of this section. 33(U)
/
(s_- 17.1.5A INSTRUCTIONS, PROCEDURES, AND DRAWINGS The HL&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 effective in verifying that the defined activities are accomplished and documented in accordance with written precedures, instructions, and drawings and that they provide quantitative and qualitative acceptance c riteria . Procedures for the review, approval, and issuance of documents (including procedures, instructions, specifications, and construc-tion drawings), and changes thereto are established and described to assure technical adequacy and inclusion of appropriate quality requirements prior to implementation. Selected documents are re-viewed and concurred with by the project QA organization for Quality Assurance related aspects. HL&P Project Quality Assurance reviews the Ebasco Allens Creek Project Quality Assurance Program. To measure the effectiveness of the Quality 57 Assurance Program, HL&P has implemented a monitoring program consisting of audits which are performed by HL&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 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. 17.1.5B INSTRUCTIONS, PROCEDURES AND DRAWINGS l The Ebasco Quality Assurance Manual for this project contains methods 5 I of complying with each of the 18 criteria within 10CFR50, Appeisdix B. Ql-ll. ll These methods are in the form of policies, instructions, and procedures l'.2.ll to be followed by those individuals or groups charged with the responsi- 17.2.16 bility for implementing or enforcing the Quality Assurance Program. 17.2.19-Ebasco's methods of complying with each of the 18 criteria are briefly de-scribed herein, in Sections 17.1.lB through 17.1.18B. The Ebasco Quality Assurance Manual is supplemented by internal procedures
- which describe the manner in which activities af fecting quality are to be accomplished as dictated by the needs of the project. Such procedures in- 5 clude auditing procedures and specification review procedures as well as site Quality Compliance and Quality Control procedures. It is a require-ment of the Ebasco Quality Assurance Manual that such Construction Quality l
_( v Control procedures be reviewed by Quality Assurance Engineering. l 16 l
\_ l (U)-Update 17.1-27 Am. No. 57, (5/81)
ACNGS-PSAR l O Procedures, instructions or drawings describing activities af fecting 5 l - quality which are quantitative or qualitative in nature (i.e., inspections 91'11*13 or tests) contain or reference criteria for determining that such activi- 17.2.11
- ties have been satisfactorily accomplished. Vendors' procedures will be reviewed by Quality Assurance Engineering. During this review it is assured that the required acceptance criteria are contained or referenced.
Refer to Section 17.1.2B for applicability of this section. 33(U) l i l (U)-Update 17.1-27a Am. No. 57, (5/81)
l ACNGS-PSAR
/-'S - 17.1.6A .D00DMEh1 CCNTROL \N- /' The 'ht&P Project- Quality Assurance Plan and implementing procedures require
- i. 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- '
ment, f abrication, construction, and QA/QC procedures, plans , and changes thereto are reviewed and approved by procedurally authorized groups and ! that the documents are issued, maintained current, and controlled by the use of controlled lists of document holders to ensure that superseded
- documents are replaced in a timely manner.
Measures are established and documented to control the issuance of docu-ments, such as instructions, procedures, and drawings, including changes thereto, which prescribe activities affecting 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 location where the prescribed activity is performed. . Changes to documents -
are reviewed and approved by the same organizations that performed . the 1 original review and approval unless other organizations are.specifically . designated. The reviewing organization has access to pertinent background information upon which to base its approval and shall have adequate under-j standing of the requirements and intent of the original document. j s Those participating in an activity are made aware of and use proper and 57 current instructions, procedures, drawings, and engineering requirements for performing the activity. Participating organizations have procedures for control. of the documents and changes thereto to preclude the possible ' use of outdated or inappropriate documents. Document control measures provide for: a) identification of individuals or organizations responsible [ ! for preparing, reviewing, approving, and -issuing documents and revisions thereto; , b) identifying the proper documents to be used in performing the activity; I c) coordination and control of interf ace documents; I d) ascertaining that proper documents are being used; e) establishing current and updated distribution lists ! The document contral system includes a listing identifying the current revision of instcuctions, procedures, specifications, drawings, and procurement documents. This list is updated and distributed to cognizant responsible personnel. } HL&P Discipline Quality Assurance performs implementation reviews at the construction site to ensure that document control systems are in place and ef fectively implemented. HL&P Quality Assurance audits are performed 17.1-28 ~ Am. No. 57, (5/81) ; I
_ ~. -. - - _ . -. .. - _ . . .. - - - - -. - . _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
;
,. ACNGS-PSAR to ensure compliance with. these criteria. ; r 17.1.6B- DOCUfENT CONTROL Ebasco procedures for the ' approval and release of drawings and specifica- l tions are described under Section 17.1.38 (Design Control). The issuance i of revisions and changes are handled in the same manner. The individuals or groups responsible.for reviewing and issuing such documents and revi-4 sions thereto, are ' identified in the Ebasco Quality Assurance Manual. Internal Ebasco procedures for activities affecting quality are generated by various organizations within Ebasco, but, in all cases are subjected to
- a. review cycle to assure that appropriate requirements are contained. As an example, Construction Quality Control Procedures are generated by the
. Construction Department but are subject to review cycle which is internal to Construction and also includes Quality Assurance Engineering.
i l Issuance of tlw Ebasco Quality Assurance Manual is the responsibility of the Ebasco Chief Quality Assurance Engineer or his designee. A list of all holders of the manual is maintained by his in the New York Office and
- - copies of all revisions are sent to .all holders with appropriate instruc-j tions for incorporation. The Ebasco Quality Program Committee is respon-sible for and has authority to make changes to policies and procedures in this program as described under Section 17.1.2B (Quality Assurance
, Program). Documents which have a bearing on safety-related activities are , controlled. Such documents include the Ebasco Quality Assurance Manual and j' implementing procedures, design documents (specifications, drawings, calcu-
; .
lations) procurement documents (including purchase orders and supplements ' i thereto reports of design changes, deviations and objective evidence I material certifications, test reports, etc.) of proper manufacture and 4 construction of safety-related structures, systems and components. Changes ; } to documents are reviewed by the same organizations that performed the
- originsi review unless delegated by the owner to a qualified responsible organization. Approved changes are promptly included where applicable into {
l instructions, procedures, drawings and other appropriate documents asso-1 ciated with the change. 1 Documents are available at the onset of the work for which they are needed. i Obsolete or superseded documents are controlled to prevent inadvertent use. { Master lists which identify the current revision number of drawings, proce-6 dures, instructions and procurement documents are established and imple-mented. These lists are updated and distributed to predetermined respon- 91~I1'11 j; sible personnel on a timely basis. Computerized updates are distributed 17.2.14 = usually once a month. F Vendors, Manufacturers and Contractors are responsible to control the issuance of their drawings, instructions and procedures and the review of the documents and changes thereto, in accordance with their quality assur-ance program as required by Ebasco specifications.
Refer to Section 17.1.2B for applicability of this section. 33(U) k 4 l' l l (U)-Update i 17.1-29 Am. No. 57, (5/81) I o 1
f 1 ACNGS-PSAR l 17.1.7A- CONTROL OF PURCHASED MATERIAL, EQUIPMENT, AND SERVICES The HL&P Quality Assurance Plan and implementing procedures require that HL&P, prime contractors, and subcontractors define and document the system 4 and requirements for the control of nuclear safety-related purchased material, equipment, and services.
~
i Control and verification of supplier's activities during fabrication, i inspection, testing, 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
- l a) Requiring the supplier to identify processes to be utilized in fulfilling procurement requirements.
4 b) Reviewing documents required to be submitted by the procure- 57 ment requirements. i j I 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 ; 1 of documentation required; and those responsible for imple-menting these procedures. d) Audits, surveillance, and/or inspections which assure that the supplier complies with quality requirements and his QA program. Control and verification of organizations performing service is accom-
! plished by technical verification of data provided, surveillance and/or audit of the activity, and review of objective evidence such as certifi-cations, reports, etc.
r 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 select' ion measures are implemented by HL&P and Ebasco and provide for identification of the organizational responsibilities for determining supplier capability. Measures for evaluation and selection of procurement sources, ar.d the results thereof, are documented and include one or more of a) through d) below: a) Evaluation of the supplier's history of providing an identical
! or similar product or service which performs satisf actorily in actual use. The suppler's history shall reflect current capability.
b) Supplier's current quality records supported by documented qualitative and quantitative information which can be objec-tively evaluated. 17.1-30 Am. No. 57, (5/81)
ACNGS-PSAR fN s c) Supplier's technical and quality capability as determined by ( ) s. 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 sdequacy, quality assurance, and commercial considerations.
Procurement of spare or replacement parts for structures, systeas, and components important to safety is subject to QA program controls, , l to codes and standards, and to technical requirements at least equal i to the original . technical requirements or any properly reviewed and approved revisions there to. i A receipt inspection is planned and implemented to assure: 1 a) Timely inspection of items upon receipt. - i l b) The material, component, or equipment is properly identified j and corresponds to the identification on the purchase document ; and receiving documentation. : c) Material, components, equipment, and acceptance records l satisfy the inspection instructions prior to installation , j or use.
, d) Specified inspection, test, and other records are accepted and available at the Allens Creek Project prior to installa-l tion or use where required. 57 i e) Items accepted and released are identified as to their ; inspection status prior to forwarding them to a controlled ; storage area or releasing than for installation or further l work.
f) Coordination of receipt inspection with vendor surveillance ] activities to ensure the required vendor inspection has been j performed and deficiencies have been resolved prior to shipment.
; Supplier's certificates of conformance are evaluated by audits, vendor i inspection, or tests to ensure that they are valid. Supplier's records l will include a description of those nonconformances from the procurement requirements dispositioned " accept as is" or " repair".
- Ebas receiving inspection ensures that, for nuclear safety-related j itema received at the Allens Creek Project, there is accompanying ,
documentation that indicates review and concurrence by the prime con-
; tractor or designee, that the item complies with established requirements !
l or. has an authorized waiver prior to shipment. HL&P Quality Assurance ! audits are performed to, ensure compliance with these criteria.
\
17.1-31 Am. No. 57, (5/81)
- ., ---~w.--. - --3--, .--.- .- . ,--,.-%v ,-,_we-.-_.--. m - - , , - #--re w. n ,, wn -,--r er -
l ACNGS-PSAR HL&P Houston Quality Assurance ensures by an overview of the Ebasco vendor ( s 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 commencing 57 with receiving inspection at the site to ensuring proper controls of purchased material and equipment are exercised. ) i HL&P Houston Quality Assurance performs audits of these activities to i ensure overall compliance. l l 4 17 .1.7B CONTROL OF PURCHASED MATERIAL, EQUIPMENT AND SERVICES l Ebasco procedures for control of purchased safety-related material, equip-ment and services are contained in the Ebasco Quality Assurance Manual for 6 the project. The procedure for evaluation and qualification of vendors is gg,,gg,y also detailed in the Ebasco Quality Assurance Manual. Under this, prospec-17.2.18 tive vendors must submit their documented Quality Assurance Programs for 17.2.19 review and evaluation by the Ebasco Materials Engineering and Quality Com- 17.2.44 pliance Department for conformance to specification requirements. Where 17.2.46 vendors are likely recipients of purchase contract award, an audit of their quality assurance capability by actual plant visitation may be made. The evaluation of vendors took into consideration the industry reputation of i the firm and past Ebasco experience with the firm. i Subsequent to the selection of a Vendor, a Quality Compliance Plan is pre-pared by the Project Quality Assurance Engineer in consultation with the ) cognizant Lead Discipline Engineer who prepared the purchase specifica- - tions. This is prepared in accordance with a procedure set forth in the Ebasco Quality Assurance Manual for the project. The Quality Compliance , ) Plan defines functional and documentatiion requirements and establishes the q Vendor's fabrication steps to be monitored by the Ebasco Vendor Quality Compliance Representatives at the Vendor's shops. A " Records / Documentation i Checklist" is prepared and executed as part of the plan. l The Ebasco Materials Engineering and Quality Compliance Department will re- ; view applicable Vendor procedures and personnel qualifications to estabi sh , i that standards, codes and regulatory requirements are properly specified in ! order to meet the requirements. The Ebasco Quality Assurance Manual con- - tains procedures for the distribution and review of these Vendor proce-dures. i
- The Chief Vendor Quality Lompliance Representative is directly responsible '
< for the organization and direction of the Vendor Quality Compliance Repre-I sentatives. He reports directly to the Director of Materials Engineering j and Quality Compliance Department. Individual Vendor Quality Compliance Representatives implement the Quality Compliance Plan. It is the Vendor 4 Quality Compliance Representative's responsibility to assure that quality requirements of the specification are met. The Ebasco Vendor Quality Com-pliance Representative has the authority to reject components and equipment which do not comply with the applicable Purchase Order Requirements. Ebasco Vendor Quality. Compliance Representatives have had experience in inspection , and. inspection methods of power plants and process plant components and ! systems. A Vendor Quality Compliance Representative inspects only those j components in his area of expertise. I l
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17.1-32 Am. No. 57, (5/81) j
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ACNGS-PSAR Sequence schedulee are obtained from each Vendor to facilitate the estab-i psl lishment of the Quality Compliance Plan so that the specific check points for Ebasco Vendor Quality Compliance Representative visits can be desig-nated. The check point system uses a randon approach, but includes sur-veillance of the critical steps of fabrication. The Vendor is required to give adequate notice of any impending check point. The Quality Compli-ance Plan is sent to HL&P for their information and use. The Vendor Quality Compliance Representative will fill out . the appropriate quality compliance forms based on his checks and observations. These forms cover 6 the Vendor material documentation, fabrication, and tasting. These forms Q1-11.U and applicable documentation will be sent to the construction site where 17.2.21 they will be audited by the Ebasco Site Quality Compliance Supervisor and 17.2.22 his staff, and filed in accordance with filing system described in the 17.2.28 2basco Quality Assurance Manual. The Ebasco Vendor Quality Compliance Representative will audit applicable 6 documentation in the Vendor's shop and upon completion of fabrication, but q1 11,11 prior to the shipment of the components, a release for shipment form will 17.2.21 be executed by him indicating evidence of compliance with purchase contract 17.2.22 requirements. The release form will be sent to the Ebasco Projeat Quality Assurance Engineer and to the Site Quality Compliance Supervisor. The com-ponent or material can then be shipped.
- When received at the construction site, material is examined primarily to determine whether any shipping damage has occurred. Exceptions to this are j with materials ordered from the construction site such as concrete and re-inforcing steel. For these, the receipt inspection includes assessment of proper documentation by qualified field personnel. Receiving inspection of the supplier furnished material, equipment and services is performed in accordance with the following
a) The material, equipment or component is properly identified and cor- . responds with the receiving documentation. b) Inspection of the material, component or equipment and acceptance records is performed and judged acceptable, in accordanca with pre- l 33(U) determined inspection instructions, prior to installation or use. c) Items accepted and released are identified as to their inspection 6 l status and forwarded to a controlled storage area or released for Q1- 11.11 installation or further work. 17.2.36 d) Nonconforming items are held in a segregated controlled area and/or clearly identified until proper disposition is made. Equipment received at the construction site will not be released for in-stallation until the release for shipment form or quality as urance records are available at the site. The release form is evidence of the avail-ability of the quality assurance records at the vendor's facility. Refer to Section 17.1.2B for applicability of this section. 33(U) (U)-Update 17.1-33 Am. No. 57, (5/81) r
ACNGS-PSAR ( 17.1.8A IDEN."IFICATION AND CONTROL OF MATERIALS, PARTS AND COMPONENTS N ]) The HL&P Project Quality Assurance Plan requires that prime contractors and cuppliers establish written procedures which identify, control, and ensure traceability of materials, parts, and components including partially assembled components. Prime contractors and suppliers procedures shall include the documented verification of correct identification of materials, components, cnd subassemblies, and that the identification does not affect the function or quality of the item prior to release of the items for assembly or installation. Specific procedures have been developed to: a) Establish controls to identify and control materials (including consumables),) subassemblies . parts, and components (including partially fabricated b) Provide a method for identification of quality related materials and 57 parts and to provide traceability to the appropriate drawings, specifications, purchase orders, manufacturing, and inspection documents, deviation reports, and physical and chemical mill test reports. c) Provide a method for identification and control of incorrect or defective items. This system will include verification and documentation prior to release for f abrication, assembling, shipping, and installation. HL&P Project Quality Assurance ensures that the above criteria are incorporated into the Ebasco quality / construction procedures during the procedure review and then follows up with implementation reviews to ensure compliance. In addition HL&P Houston Quality Assurance performs audits for evaluation of the conformance to identification and control criteria. 17.1.8B INDENTIFICATION AND CONTROL OF MATERIALS, PARTS AND COMPONENTS The Ebasco Quality Assurance Manual provides measures which assure that the purchase order will require that contractors of materials, parts and 5 components important to safety, establish and utilize procedures to verify q *g that items are properly identified and marked prior to the start of manufacturing operations and that the required identification and traceability 17.2.2.2 is maintained as required by codes, standards or specifications as 17.2.2.4 cpplicable. The identification methods must provide a high degree of 17.2.2.6 confidence that they will prevent the use of incorrect or defective items as well as prevent the use of items which have not been subjected to the required 5 inspections and tests. Consideration is given to assure that the location and method of identification do not affect the function or quality of the item 91-11 1 being qualified. The verification of correct identification of material, 17.2.2.3 parts and components is required prior to release for assembling, shipping and 17.2.2.4 installation. This identification is maintained through receipt of the 5 equipment at the construction site. Ebasco procedures provide for receipt 97 11,g inspection to assure proper identification. A system is provided for tagging i (p) and identifying materials received at the site. Rejected material is clearly 17.2.3.6 U tagged and segregated or removed from the construction site, when feasible, to prevent inadvertent use. The Ebasco Quality Program requires that identification and traceability during construction be maintained where
- required by code, standard or specification as applicable.
17.1-34 Amendment No. 57, (5/81) l l l , _
ACNGS-PSAR Refer to Section 17.1.2B for applicability of this section. 33(U)
]
V 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 activities associated with:all special processes. For special processes the qualification of personnel, procedures, and equipment relating to specific codes, standards, specifications, and contractual requirements shall be documented and maintained current. Special processes are defined as the processes where direct inspection if impossible or disadvantageous and which nust be carefully controlled and monitored to ensure the required results. Special processes for the Allens Creek Project include: a) welding b) heat treating c) cadwelding d) concrete placement e) nondestructive testing 57 i f) chemical cleaning Organizational 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. Special processes are performed under controlled conditions by qualified i personnel using procedures qualified and approved in accordance with applicable codes, standards, or other requirements. For special processes not covered by existing codes or standards the specific equipment, personnel qualification, and procedure qualification requirements are defined prior to application of the special process.
- Records are maintained for the qualif,ication of procedures, equipment, and personnel associated with special processes. Records are in sufficient detail i to clearly define the procedures,-equipment, or personnel being qualified; criteria or requirements used for qualification; and the individual approving the qualification.
HL&P 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. HL&P Houston Quality Assurance performs audits of special process activities to ensure compliance with all aspects of the Quality Assurance [,rogram. (J ) (U)-Update 17.1-35 Amendment No. 57, (5/81)
l I ACNGS-PSAR 17.1.9B CONTROL OF SPECIAL PROCESSES (LJ; The Ebasco Quality Assurance Manual assures that the purchase order will contain procedures-for providing confidence that special processes are controlled and accomplished by qualified personnel using qualified procedures
.in accordance with applicable codes, standards, specifications and criteria.
Welding, heat treating, cleaning, and nondestructive examination are considered special processes. 9 Special processes and procedures, when required by the Purchase Order, must be submitted to Ebasco. These documents will be reviewed by Ebasco's Materials Engineering and Quality Compliance Department. On occasion, Vendors may have processes which because of their proprietary nature, are not submitted to Ebasco for review. In these instances, Ebasco will attempt to obtain copies of such written procedures under a proprietary and confidential agreement. If Ebasco is unable to obtain' copies of these procedures under such an agreement, Ebasco personnel will review the documents at the Vendor's facility in order to. assure that the processes are compatible with the requirements of procurement specifications. Notice as to the compatibility of such proprietary procedures with the specification requirements will be contained 5 in'the report prepared by the reviewing party. In this manner, all parties concerned are assured that the quality requirements are met. Furthermore, the suitability of a proprietary process can normally be verified during audits of the process at the Vendor's shop. 4 The Ebasco Quality Assurance Manual provides the requirements for control of (A) special processes used during the fabrication and erection of structures, and components at the construction site. The Manual requires that special 5 processes be performed to written procedures and identifies the documentation 91,11,1 required to verify conformance to the procedures. When qualification of 17.2.19 personnel is required by the specification or its references, only personnel 17.2.25 thus qualified are permitted to perform such operations or inspections. 17.2.36 The Ebasco Quality Assurance Manual requires the establishment at the construction site of a procedure and qualification working file to contain 5 welding procedures and nondestructi v testing personnel qualification records. Also included in this working file are procedures for hydrostatic tests, leak tests, heat treatment and cleaning. Ebasco's Vendors and contractors must demonstrate implementation of an 5 acceptable program of control over special processes to include procedures and q1, operator qualification. This must be nne prior to award of a contract to the 17.2.26 Vendor. The Ebasco Vendor Quality Compliance Representative during fabrication assures that this program is implemented. Refer to Section 17.1.2B for applicability of this section. 33(U)
,O 1
0 (U)-Update 17.1-36 Am. No. 57, (5/81)
ACNGS-PSAR 17.1.10A *NSPECTION
/ ')
( ,,/ The HL&P Project Quality 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 inspection hold and test points and the criteria to be witnessed by authorized inspectors. Ope ra-tions and inspections (including rework, replaced items) are performed in predetermined, documented sequences, and deviations or deletions must be accomplished in accordance with approved and documented systems. Inspec-tion procedures include all required inspection operations defined by the specifications, drawings, codes, and standards. These procedures provide 2 for the following: a) Identification of characteristics and activities to be inspected b) A description of the method of inspection c) Identification of the individuals or groups responsible for per-forming the inspection operation d) Acceptance and rejection criteria ('~'% 57 g' e) Identification of required procedures, drawings, and specifications 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 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 to the inspectors prior to performing inspection. If mandatory inspection hold points are required beyond which work cannot proceed without specific consent of the designate representative, the specific hold points will be indicated in appropriate documents. Inspection results are documented. l evaluated, and their acceptability determined by a responsible individual or group. HL&P Discipline Quality Assurance ensures that inspection control criteria are complied with by review and approval of the inspection procedures and by implementation reviews of inspection in each discipline activity. 17.1-37 Am. No. 57, (5/81) i
E ACNGS-PSAR HL&P Houston Quality Assurance performs audits of HL&P and Ebasco inspec- 57 m ~ tion activities to ensure compliance with these criteria.
- 17.1.10B INSPECTION The Ebasco. Quality Assurance Manual assures that Purchase Orders provide 5 methods for inspection of activities affecting quality is established and executed both in the Vendors' shops and at the power plant . construction
' 5 site. The Ebasco Quality Assurance Manual, and Section 17.1.7B descrits the program by which control is exerted over shop activities, one phase of 91-11*11 which is .the inspection program. Additional information pertaining to 17.2.12- l inspection control is presented in Section 17.1.98, " Control of Special 17.2.19 : Proce sses *'. Inspection activities, both in Vendors' shops and at the construction site.- 5 must be performed by persons other than those accomplishing the operation q1 11,11 which is being inspected. During the Ebasco evaluation of Vendors, one 17.2.28
- criterion for acceptability is this independence of inspection personnel.
As described in Section 17.1.58, activities af fecting quality such _as in-spection activities must be provided for in writing. Such provisions may take_ the form of procedures, instructions and/or check-lists. It shall 5 be the responsibility of Quality Assurance Engineering to review l 16 such inspection. procedures, instructions or check-lists for adequacy, 5 whether they pertain to inspections in a Vendor's shop or at the construc- 91,,71,17 tion site. In order to be acceptable, such inspection documents must pro-17.2.24 vide for the following, either by inclusion or by reference: i a)' Quality characteristics to be inspected
b) Individual or group to perform inspection 5 y 4 c) Acceptance criteria
; d) Method of inspection i
1 e) Results of the inspection 4 . f) Appropriate certifying signatures and dates.
, Ebasco Vendors of safety-related equipment must demonstrate the implementa-l tion of an acceptable quality assurance program prior to being awarded a 5 contract. One aspect of this program is to require that inspections are t be performed in accordance with written instructions, procedures and/or Q1-11*11 , checklists. Such documents are required by the component specification to 17.2.18 be submitted to Ebasco for review prior to performing the particular opera- 17.2.19 { tion, and in this way Ebasco assures that the documents exist and are 17.2.41 acceptable prior to performance of the operation. This requirement, in- 17.2.44 + l -_ cluded in the component specification, aids in assuring that inspection operations are performed in accordance with previously prepared and re-viewed documents. l' Inspection operations may be performed only by properly qualified person- 5' net. Such qualification may be based on a formal program of training and : 17.1-38 Am. No. 57, (5/81)
__ n_..,.___... ,._.______,,,___..m . . , _ _ _ _ . _ _ . _ _ . . -
I ACNGS-PSAR examinations, or may be based on periodic evaluation of the inspector's 5
/ activities coupled with indoctrination sessions for the purpose of remain-( ,/ ing up-to-date with regard to industry developments.
The Ebasco Quality Assurance Manual as well as Section 17.1.12B requires 5
.that a system be in effect to assure that inspection equipment is calibra- 91 11,11 ted so as to be within the required limits of accuracy. During Ebasco's 17.2.29 Vendor evaluation process, the Vendor's calibration control program is I assessed. for adequacy and must be acceptable prior to that Vendor receiving '
a purchase order. Adequate implementation of an acceptable inspection program shall be veri-fled in the Vendors' shops through the activities of the Ebasco Vendor 5 Quality Compliance Representatives during their periodic visits. As des-cribed in the Ebasco Quality Assurance Manual, Vendor Quality Compliance Representatives perform surveillance over the Vendors' activities in order to establish and maintain confidence that all specification requirements are properly complied with. These Vendor quality Compliance Representa-tives are provided with a Quality Compliance Plan which includes a check-list outlining those operations which require his witnessing, as well as 5 the documentation requirements for the particular equipment. He is also Q1-11.11 provided with a list of Ebasco-approved Vendor drawings and copies of 17.2.6 transmitted letters showing the status of Ebasco reviewed Vendor proce-dures. The Vendor Quality Compliance Representative will not issue a Release for Shipment form until he has verified that all required inspec- 5 tions were performed, the results were acceptable, recorded and properly Q1-11.11 w signed off. Where modifications, repairs or replacements occur af ter 17.2.16
, ) initial inspection, a re-inspection shall be performed which is based on 17.2.36 L/ the original design and inspection requirements.
Requirements for inspection activities at the construction sites are des-scribed in the Ebasco Quality Assurance Manual as well as Quality Control and Quality Compliance procedures in rs a the particular construction 8 5 site. It will be the responsibilitv ci the Site Quality Compliance Super- Q1-ll.ll . visor or his designee to audit s9 v cactor inspection activities. 17.2.27 17.2.26 Refer to Section 17.1.2B for ap> Lica ;*/ of this section. l33(U) 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-i tors which demonstrates that the facility performs in aM ordance with the Allens Creek Project requirements and specifications. The training, certification of personnel, calibration and certification of test equip-ment, system or component status, environmental conditions, inspection hold points, and configuration of the items to be tested are included in 57 the procedures. Test results are documented, evaluated, and the accep-tance status determined by the authorized departments. Test procedues or instructions provide for the following as requited:
,O a) The inclusion of requirements and acceptance limits contained in
) applicable design and procurement documents.
(U)-Update 17.1-39 Am No. 57, (5/81) J l
. _ - , . _ _D
1 L ACNGS-PSAR
.b)- ' Instructions for performing the test .A k,,) -c) Test prerequisites such as calibrated instrumentation, adequate test-equipment, and instrumentation including their accuracy require-ments, completeness of item to be tested, suitable - and controlled 4 environmental conditions, and provisions for data collection and
- . storage
;
i d) Mandatory inspection hold points for witness by owner, contractor, or. inspector (as required) e) : Acceptance and rejection criteria ,
, f) ,
- Methods for documenting or recording test data and results i
g) Provisions for assuring that test prerequisites have been met j h) Evaluation of test results i HL&P Discipline Quality Assurance ensures inclusion of adequate test con-i trol criteria by: review of the Ebasco quality / construction testing pro-cedures. They. also perform follow-up implementation reviews to verify that the controls are implemented and effective. HL&P Houston 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 3 Quality Assurance monitoring activities will interface Land in some instances overlap. RP&L Project Quality Assurance procedures will speci-fically define the responsibilities for this transition period. j, 17.1.llB TEST CONTROL 1 The Ebasco Quality issurance Manual assures that Purchase Orders require 5
- j. Vendors and major Contractors to have a testing program where _ applicable i
which ' demonstrates that the item will perform satisfactorily in-service, in accordance with written controlled procedures. These test procedures incorporate or reference the requirements and acceptable limits contained
} in applicable design and procurement' documents. j 5 ,
- , Proof and functional test procedures will state the objective of the - tests 91-11+11 '
! as well as instructions for testing method and that the tests are performed 17.2.12 } under suitable environmental conditions by qualified persons. The test 17.2.29 e procedures should include test prerequisites such as; calibration require- 17.2.30 ments, using appropriate and adequete test equipment, preparation condi-
~
tions and completeness of ~ items to be tested, mandatory inspection hold i points where applicable for witness by owners authorized inspector. Test 5 l procedures shall also establish acceptance and rejection criteria, methods q1_11,11 of collecting and recording test data. 17.2.29 i ! s_ j 17.1-40 Am. No. 57, (5/81) c. I L
._ . ...- _, . . __.-_.,-,.,._.._._...~_,.___m.... _-._ _ _._._ . . _ _ _ . _ _
I ACNGS-PSAR Ebasco reviews test procedures prior to use. The Vendors or Contractors performing these tests are required to demonstrate that the tests are j performed in accordance with the reviewed procedures. Quality Compliance Representatives ascertain that such tests are performed The Ebasco Vendor y_,/ and adequately documented; that test results are properly evaluated; that acceptance status is identified by a qualified responsible person or group, l i Preoperational and start-up tests support services are provided as neces-i sary. Ebasco will assist in reviewing the results of the preoperation and start-up tests for conformance to any required test requirements. [ Refer to Section 17.1.2B for applicability of this section. O I 1
~
( l l 17.1-41
-ACNGS-PSAR 17.1.12A CONTROL'0F MEASURING AND TEST. EQUIPMENT (my\
s_,/' s 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 't o 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 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. Equipment is ~ identified and traceable to the calibration test data and 57 suitably marked to -indicate calibration status. Markings include the last l day calibrated and next calibration due date.
> Measuring and test equipment is calibrated at specified intervals based on the ' required accuracy, purpose, degree of usage, stability characteristics, and other conditions 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
' s and authorized by responsible management.
)
'-- Calibrating standards will, when possible, have greater accuracy than j standards being calibrated. Calibrating standards with the same accuracy j may be used if it can be shown to be adequate for the requirements and the i basis of acceptance is documented and authorized by responsible management.
HL&P Discipline Quality Assurance reviews, and documents concurrence with Ebasco calibration procedures to ensure these criteria are incorporat_ed. In addition implementation reviews are performed to ensure compliance. HL&P Houston Quality Assurance audits the measuring and test equipment con-trols to ensure compliance to the QA program in this area. 17.1.12B CONTROL OF MEASURING AND TEST EQUIPMENT The Ebasco Quality Assurance Manual provides assurance that require-ments for control of measuring and test equipment are included in the
~ Purchase Order. Properly calibrated tools, gauges, measuring and test 5 instruments, nondestructive test equipment and other testing devices are to be used in measurement, inspection and monitoring of safety-re-lated components, systems and structures. These measuring and test instruments are to be calibrated and maintained at specified intervals 4 based on the required accuracy, purpose, the degree of usage, stability characteristics and other conditions af fecting the measurement. Detail s
,, of these requirements are included in the Ebasco Quality Assurance Manual and quality control procedures.
s 17.1-42 Am. No. 57, (5/81)
---._____ .__- . ,__ . . .-, . e - - . - - - - a y. . . , - . _ _ , - , _ , _ .._,y _.. - . _. _m._._ . _ , _.y,.,.,_---...,,,,,,ym.._~
ACNGS-PSAR Vendors are required to indicate that Reference and Transfer Standards
/} shall have a known valid relationship to nationally recognized standards.
( ,/ Where national standards do not exist, provisions are established to document. the basis for calibration. When measuring and test equipment are found to be out of calibration, an investigation will be conducted and documented to determine the validity of previous inspections performed. 5 The Ebasco Vendor Quality Compliance Representative is responsible for verifying that: procedures are employed for measuring and test equip-ment; that measuring and test equipment are properly identified and have traceability to the calibration test data; that records are maintained which indicate the complete status of all items under the calibration system. At the construction site, the Site Quality Compliance Supervisor is re-sponsible for ascertaining that the requirements of the calibration pro-cedures have been met and that appropriate documentation is provided. A master file of records of calibration for measuring and testing is main-tained by the Ebasco Site Quality Compliance Supervisor. Refer to Section 17.1.2B for applicability of this section. 33 (U) O,, s , t I I O v (U)-Upda te 17.1-43 Am. No. 57, (5/81)
ACNGS-PSAR ID
.t 4
(/ 17.1.13A HANDLING, STORAGE, AND SHIPPING The HL&P Project Quality Assurance Plan requires that for nuclear safety-related items, written procedures be developed in accordance with design requiremets, 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 57 appropriate trained and experienced personnel. HL&P Discipline Quality Assurance reviews and documents concurrence with con-struction 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 compliance to the procedures. HL&P Houston Quality Assurance performs audite to ensure overall program compliance. 17.1.13B HANDLING, STORAGE, AND SHIPPIllG The Ebasco Quality Assurance Manual assures that requirements for handling, storage, shipping, cleaning and preservation of equipment and materials are provided in Purchase Orders and Quality Control and Quality Compliance Proce-dures. The types of storage facilities required for various kinds of equip-ment and materials are provided. These requirements may be supplemented by 5
/"N special inst ructions from Vendors, Contractors or f rom Ebasco engineering in (U ) cases where special handling, storage and shipping is required due to weight, size, susceptibility to shock damage or other special circumstances. These requirements are determined and accomplished by individuals having adequate knowledge and experience.
Equipment and materials arriving at the jobsite must be accompanied by the Vendor Quality Compliance Representative's Release for Shipping Form described 5 in Section 17.1.7B. They In shall then be inspected to assure that no shipping addition to a general procedure, because of the I damage has occurred.
- special nature of some items, the construction contractor shall develop special procedures for receiving, handling and storage of items such as weld-ing rods, electrodes, filler wire, electrical equipment and instrumentation and control components. These procedures shall be in accordance with the 5 requirements of the Ebasco Quality Assurance Manual.
1 A comprehensive storage plan is developed by Ebasco for the project based on a review of the project schedule and St atus of Materials Report. Equipment shall be suitably and adequately protected from the time of arrival on the Conformance after receipt in jobsite until installation and trial operation. the field and during installation is verified by construction quality control t personnel and monitored by the Ebasco Site Quality Compliance Supervisor and 5 l his staf f. l { i (v, Am. No. 57, (5/81) 17.1-44
)
l ACNGS-PSAR' ) p/ y The Ebasco requirements for field storage of components and the basis for determining under what storage conditions an item will be stored are contained in the Ebasco Nuclear Quality Assurance Manual. These requirements may be supplemented by instructions in the components specification which detail specific mandatory or recommended Ebasco or manufacturer's storage require-ments for the particular component. 5 [ The Ebasco Quality Assurance Manual designatns items that are permitted to be 5 stored outdoors with tarpaulin and shoring as specified in protected drained 91-11 1 areas. It also designates items that must be stored on pallets in prefabri- 17.2.41 cated or equivalent tear resistant enclosures which are weather-tight, well
~
drained, well ventilated with a paved floor or equivalent. The Ebasco Quality Assurance Manual also indicates what mist be stored indoors but also shall have temperature control or its equivalent to prevent condensation and corro-5 sion. Items such as instrumentation and electronic equipment that are ex-ceptionally sensitive to environmental conditions, such as moisture and t em-perature will be stored under special conditions agreed to by the Manufacturer
, and Purchaser. .
Refer to Section 17.1.2B for applicability of this section. l 33(U) 17.1.14A INSPECTIONS, TEST, AND OPERATING STATUS The HL&P Project Quality Assurance Plan requires that the prime contractor and (q subcontractors indicate the current inspection, test, and operating status of V) nuclear safety-related items through the use of stamps, markings, tags, or other suitable means. During the startup and testing activities, HL&P is responsible for complying with this section for inspection status, test status, and operating status. Procedures include the requirements for: a) Controlling the application and removal of inspection status indicatcrs such as tags, markings, labels, and stamps, b) Documenting the status of nonconforming, inoperative, or malfunctioning structures, systems, and components to prevent inadvertent use.
! 57 c) Defining and documenting the use, application, removal, and status of inspection tags, labels, and markings which identify the status of inspections or- tests performed or attest to the acceptability of the structure, system, or component.
d) Controlling the altering of the sequence of required tests, inspec- , tions, and other operations important to safety. i 1 i 1 HL&P Discipline Quality Assurance personnel review and document concurrence with these procedures and conduct periodic verification to assure compliance. Houston Quality Assurance audits both HL&P Project Quality Assurance and Fhasco to verify compliance. (
%.)
l (U)-Update l 17.1-45 Am. No. 57, (5/81)
ACNGS-PSAR 'O
/ :17.1.145 -INSPECTION, TEST, AND OPERATING STATUS Ebasco procedures contained in the Ebasco Quality Assurance Manual provide means of identifying and verifying status of equipment during- fabrication, .i
. installation and testing. Ebasco requires that Vendors maintain a system for l ' identifying the inspection, test and processing status of materials, parts and 5
;
components to p eclude inadvertent bypassing of requirement hspection, tests i and processing and that nonconforming items are clearly identified to prevent ' inadvertent use. Ebasco's vendors and contractors are required to submit' their procedures 'for surveillance and inspection of work that affects quality which provides means for verifying that each process is performed in accor-dance with contract specifications. The Ebasco Vendor Quality Compliance
~
7 Representative verifies that the Vendor is conforming to such a system. 5 The Contractor shall control items and services that do not conform to re- ! -!' quirements. The methods for control shall provide for prompt identification, 91-11*11 i documentation, segregation and notification of the nonconformances to Ebasco. 17.2.37 i The control methods shall provide for immediate action to withhold work on 17.240 items and services until disposition is determined. 5 + In otuer to prevent inadvertent use, nonconforming items are identified, gl.11,11 segregated when practical and documented in accordance with the requirements contained in the Ebasco Quality Assurance Manual and, Ebasco and Site con- 17.2.37 tractors quality control procedures. For installation and erection, Ebasco and site contractors shall employ a 5 4 system administered by Quality Control personnel which utilizes inspection Q1-11.11-status and record cards, inspection checklist s, st amp tags, labels or routing 17.2.36 ! cards as necessary to clearly indicate inspection and test status. l Refer to Section 17.1.2B for applicability of this section. 33(U) l ! 17.1.15A NONCONFORMING MATERIALS, PARTS, OR COMPONENTS i i The HL&P Project Quality Assurance Plan requires that HL&P and the prime l- contractors' Quality Assurance Program include a system which is documented by written procedures for the identification, segregation, and disposition of ' l nocconforming materials, parts, and components. The procedures shall specify
- l. the preparation and handling of nonconformance documents, segregation require-4 ments,- and which groups are responsible for review and disposition of the 57
- items.
l Documentation identifies the nonconforming item; describes the nonconformance, I- the disposition of the nonconformance, and the inspection requirements; and i' includes signature approval of the disposition. Nonconformances are corrected I and resolved prior to initiation of the preoperational test program on the
- item. Recork, repairs, and subsequent reinspection and tests are conducted in I l
+ (U)-Upda te 17.1-46 Am. No. 57, (5/81) .
. - . . . ,- , , _-- . , _ _ -,~-..m-..-.-. ~ . , ~
ACNGS-PSAR-If ' Aj accordance with.the original inspection and test requirements or accepted alternatives and shall be performed in accordance with controlled procedures and contain mech'anisms for providing information to the identifying group'as to the disposition of.the nonconformance. 4 For. NSSS items, HL&P coordinates nonconformance resolution through GE. 57 LHL&P Project Quality Assurance reviews for concurrence the proposed disposi-tion of selected Ebasco nonconformance reports and performs an evaluation of "Ebasco nonconformance trend analyses. Procedures are established by HL&P to report significant deficiencies during the design, construction, and operations phase to HL&P executive management and to the Nuclear Regulatory Commission in~accordance with 10CFR50.55(e),. 10CFR21, and 10CFR71, where applicable. i Compliance of these activities with Project Quality Assurance Plan require-F ments is ensured through the performance of audits and implementation reviews. i 17.1.15B NONCONFORMING MATERIALS, PARTS OR COMPONENTS 5 , The Ebasco Quality Assurance Manual assures that Purchase Orders will require ~ 91 11,11 that nonconforming materials, parts or components be clearly tagged or. other-j wise identified. Requirement s for contractors and Vendors to control non- 17.2.3
- p conformances are detailed in the Ebasco Quality Assurance Manual for this 17,2,4 l project. For purchased material, Vendors shall control items and services that do not conform to requirements. The methods of control shall provide for
- . prompt identification, documentation, segregation and notification of non-conformances to Ebasco. The control methods shall provide for immediate i action to withhold work on items and services until disposition is determined.
{ Nonconforming items and services shall be reviewed, repaired .or reworked, l retested and reinspected and then accepted or rejected in accordance with l documented procedures reviewed by Ebasco. The Vendor shall remove rejected ' items from the fabrication process or to prevent inadvertent use or installa-tion, and rejected services shall be removed and corrected in a satisfactory 5 i manner. The Vendor shall obtain the concurrence of Ebasco prior to any repair 91 11,11 or disposition of items and services furnished by the Vendor. A nonconfor- 17.2 37 mance report shall be filed for all conditions which are adverse to quality. 17.2.40
- Nonconformances to the purchase order specification requested by the Vendor -
after placement of the order are fully evaluated by the engineering discipline 5 j originating the initial design and specification. Nonconformances to the j specifications observed by Ebasco's Vendor Quality Compliance Representative i are submitted on the Nonconformance Report to the Vendor for recommended 16 disposition and then to Ebasco Quality Assurance Engineering (New York l Of fice), for processing and evaluation of the recommended disposition. Provi-i sions in-the Nonconformance-Report shall provide for reinspection and/or 5 ! retest to ~ assure compliance to the disposit ion instruction, and to determine 91 11,11 l- that the item shall function as originally specified. Written instructions 17.2.37 l l with regard to the recommended disposition are required. I i l' 17.1-47 Am. No. 57, (5/81) i
- a. . . - . 4 .- . . . - ~ ,_ .. ~ , , ,-.,_..,_,--...._mm,,.,,.,..._,.-. , , , + . ~
ACNGS-PSAR I 5
, At the construction site, nonconforming materials, _ parts or components are identified and controllad in accordance with the requirement s of the Ebasco 91~11*11 Quality Assurance Manu .1 and quality control procedures. Materials, parts or components must be cluarly tagged or otherwise identified. If the material does not meet the miaimum specified requirements, and the responsible Con-struction Quality Control Representative believes that the materials should be accepted, he will initiate action as to disposition by documentation on a Nonconformance Report Form. This report will be forwarded to the Ebasc 5 Materials Engineering and Quality Compliance Department for technical evalua-tion and concurrence by the engineering discipline originating the initial design and specification. The report describes the nonconformance, and out-lines the recommended disposition. Depending on the nature of the non-conformance, it is discussed with the cognizant engineer and/or Project 5 Engineer for resolution. Written instructions with regard to the recommended disposition are required. These reports are always sent to Ebasco's New York 91-11*ll office for final disposition. All nonconformance reports are incorporated 17*2*37 into the Quality Compliance Site supervisor's files. 17.2. 40 5
All Vendor nonconformance reports dispositioned "use as is," or " rework" shall be made part of the permanent inspection records and formally reported to HL&P 91-11*11 and maintained as part of the Vendor's file. Vendor files are turned over to 17.2.39 HL&P as part of the total QA documentation at the conclusion of the project. Refer to Section 17.1.2B for applicability of this section. 33(U) i l 17.1.16A CORRECTIVE ACTION U The HL&P Project Quality Assurance Plan requires that a system be established 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- 57 tive action. The system includes measures to identify the cause of signifi-cant conditions adverse to quality, measures to ensure that the root causes are corrected, and measures to ensure that timely action is taken. Follow up is performed to ensure the effectiveness of corrective action and that appro-priate levels of management are informed of the results. HL&P Project Quality Assurance performs a review for concurrence of selected Ebasco nonconformance reports and corrective action reports. HL&P Droject Quality Assurance also performs trend analyscs to determine the neei ;or corrective action. Com-pliance of these actions with Project Quality Assurance Plan requirements is verified by HL&P Quality Assurance through the performance of audits and implementation reviews. 17.1.16B CORRECTIVE ACTION The Ebasco Quality Assurance Program includes measures to assure that condi-tions adverse to quality such as failures, malfunctions, deficiencies, devia- 5 tions, defective materials and equipment, and nonconformances are promptly identified, corrected, documented, and reported to appropriate levels of management. Particular attention will be paid to those areas which exhibit {d 'n recurring instances of conditions adverse to quality so that the cause of these conditions can be identified, analyzed, and eliminated. (U)-Update I 17.1-48 Am. No. 57, (5/81)
l ACNGS-PSAR (m\
\s_,/ Conditions adverse to quality found by Ebasco Vendor Quality Compliance Repre- 5 Q1-11.11 sentatives during their visits to Vendor's facilities are reported in the Quality Compliance Report and documented on a Nonconfore1nce Report form which 17.2.40 must be processed through Ebasco Materials Engineering and Quality Compliance Department, as detailed in Section 17.1.5B.
5 At the construction site, all nonconforming items are ideneified as noncon- 91 11,11 forming, documented utilizing the Ebasco Nonconformance Report Form, segre-gated into a separate area when practical, and processed as previously ex- 17.2.40 plained in Section 17.1.15B. Refer to Section 17.1.23 for applicability of this section. 33(U) 17.1.17A QUALITY ASSURANCE 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 documer.te contain a description of the type of observation, reference to nonconformance reports, evidence relating to status of observation, date, and inspector identification. Records are maintained for the Allens Creek Project in a two hour rated fire fN resistant file room meeting NFPA No. 232 including the following provisions:
\ \ ) a) An automatic fire suppression cystem and an early warning fire detec-tion system is utilized.
b) Records are stored in fully enclosed metal cabinets. 57 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. e) Ventilatioa, temperature, and humidity control equipment is controlled where they penetrate fire barriers bounding the storage facility.
-Compliance with Project Quality Assurance Plan requirements is verified by j HL&P Quality Assurance through the performance of audits and implementation reviews.
1 3
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J (U)-Update 17.1-49 Am. No. 57, (5/81)
ACNGS-PSAR - p L 17.1.17B QUALITY ASSURANCE RECORDS 5 The Ebasco Quality Assurance Manual assures that records required are clearly 91-11*1-deline.ated in Purchase Orders. These records prepared by Ebasco or obtained 17.2.12 from Vendors and contractors furnish documentary evidence of the quality of 17.2.24 items and of activities affecting quality and are collect ed during the design, procurement, fabrication, construction and start up phas sa of the project. These records include the operating logs, results of reviews, inspections, 5 tests, audits, and monitoring of work performsnce and material analysis; the gl 11,13 qualification of personnel, procedures and equipment, other documentation such 17.2.25 as drawings, specifications, procurement documents, calibration procedures, 17.2. 26 calibration reports, and nonconforming and corrective action reports. The 17.2. 41 inspection and test records include the following: 17.2. 46 a) A description of the type of operation
~)
o Evidence of completing and/or verifying a manufacturing inspection or test operation c) The results of the inspection or tests 5 d) Information related to nonconformances e) Inspector or data records f) Acceptability Ebasco Vendor Quality Compliance Representatives audit applicable documenta-tion in the Vendor's shops. Upon completion of fabrication, but prior to shipment of a component, a Release for Shipment form is filled out by the Ebasco Vendor Quality Compliance Representative. The Release for Shipment form is sent to the Ebasco Project Quality Assurance Engineer and to the Site Quality Compliance Supervisor. The component or material can then be shipped. Quality Assurance records compiled at Vendor f acilities af ter normal process-ing and checking will be sent to the Ebasco Site Quality Compliance Supervisor for further review and filing. Each Lead Discipline Engineer is responsible for the records generated in his department. The Project Quality Assurance Engineer is responsible for the records in the Materials Enginee ' 'nd Quality Compliance Department files. The requirements and responsibilit 'or record transmittals, retention and maintenance subsequent to completion or work are consistent with applicable codes, standards, and procurement documents. , d 17.1-50 Am. No. 57, (5/81)
I ACNGS-PSAR
-_A 5
s_, A specific filing system is maintained at the construction site for the col. 91-11'1 lection and maintenance of all Quality Assurance records. This file is under-
' the control and responsibility of the Site Quality Compliance Supervisor and 17. 2.c includes both Vendor and Field Quality Assurance Records.- The filing system 17.2.t will provide records which are identifiable' and retrievable. ' Storage facili- 5 ties are constructed, located, and secured, to prevent destruction of' the 91-11 13 records through fire, flooding, thef t, _and deterioration by temperature or 17. 2.4.-
humidity conditions. e
. The Site Quality Compliance Supervisor shall also audit alta contractors during construction to assure that their Site Quality Assurance Documentation 5 files contain documents which have been properly reviewed, approved and filed in a planned and systematic manner.
Refer to Section 17.1.2B for applicability of this section. 33(U) 17.1.18A AUDITS The HL&P Project Quality A9Parance Plan establishes the requirement that HL&P, prime contractors, and ruocontractors develop, document, and implement audit activities which are structured in accordance with the requirements of ANSI N45.2.12 for the Allens Creek Project. As required by the ANSI standard, j 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
- performs follow-up action, including re-audit of the deficient areas.
HL&P has the ultimate responsibility for the auditing of the quality related
; activities on the project. This responsibility is fulfilled by Houston Quality Assurance, which audits the activities of HL&P, its prime contractors, and their suppliers and subcontractors.
The prime contractors and subcontractors perform quality related audits of 57 internal activities and suppliers of material, components and systems. l HL&P and Ebasco perform supplemental audits when required, based on such
factors as significant changes in the Quality Assurance Program, results of trending programs, or investigations into the root causes of problems. 1 The HL&P Project Quality Assurance Plan requires that each year an independent j outside firm shall conduct an overall audit of the Allens Creek Project l Quality Asaurance activities. The audit results are presented to the HL&P Executive Vice President and the Project QA Manager. The audit results will , be used by HL&P management to evaluate the ef fectiveness of the Quality Assur- l l ance program and to determine the need for changes in the Quality Assurance
- programs of HL&P and its contractors.
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(U)-Update l 17.1-51 Am. No. 57, (5/Li
ACNGS-PSAR
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- 17.1.18B' EBASCO AUDITS The Ebasco Quality Assurance Manual provides for'the conduct of planned and ,
- ' periodic management audits' of the Ebasco quality -program to verify' compliance with all aspects 'of the program to determine its ef fectiveness. Audits are performed in accordance with documented procedures by a committee consisting of personnel not having direct. responsibilities in the areas being audited.
- This committee is comprised of-one representative from .each of the Construc-tion, Engineering and Nuclear Departments.
- i. The Ebasco Quali' f Assurance Audit Committee shall be responsible to the Vice President - Nuclear. The committee shall perform the management audits speci- 5~ ,
fied in the Ebasco Quality Assurac'. 3 Manual, 'and shall provide the Vice i President-Nuclear with' reports of chese audits. The Committee shall also
~
7 report to the Vice President - Nuclear any deficiencies found in the Ebasco Quality Assurance Program with respect to AEC requirements and shall make -such
- recommendations to the Vice President - Nuclear as it deems necessary to '
remedy the deficiencies. The Vice President - Nuclear shal1~ be responsible
for informing other Ebasco Officers of the results of the Q/A Program. He shall also be responsible for initiating the implementation of any changes j deemed necessary by officers to improve the effectiveness of Ebasco's Q/A l Program in accordance with procedures established in the Ebasco Quality Assurance Manual. The prompt resolution of deficiencies indicated by internal Quality Assurance 5 fg Audits shall be _ accomplished by the responsible parties.' . A quarterly review will be performed by the Audit Committee to assure that timely resolutions Q1-11,11 4 I have been achieved. A quarterly review report by the Audit Committee on the 17.2.43 ; resolution of deficiencies will be forwarded by the Audit Committee to the 17.244-l' Vice President - Nuclear, with a copy to the Chairman of the Quality Program 17.2 47 Committee. l At the construction site, the Ebasco Site Quality Compliance group audits 5 , construction quality control activities. These audits are performed in accor-i dance with the requirements of the Ebasco Quality Assurance Manual. The Q1-11.11 Ebasco Chief Quality Assurance Engineer or his designee conducts a planned 17.2.44- [ audit of the Ebasco Vendor Quality Compliance Representatives, the quality j assurance activities at the construction site and the various Ebasco depart-ments participating in the implementation of the Ebasco Quality Assurance Manual, for the Allens Creek project.
- 5 Audit results are documented and t ransmitted for review and commitment of 91-11 11
corrective action by management having responsibility in the area audited. 17.2.47 The frequency of audits is established in the Ebasco Quality Assurance Manual I and is also based on the status and safety importance of the activities being l performed. ' They are initiated in a timely manner to assure ef fective Quality !' Assurance during design, procurement, and contracting activities. Ebasco 5 Quality Program requires that personnel conducting audits be appropriately l j trained and qualified. iU i l 17.1-52 Am. No. 57,(5/81) i i _ , _ . , _ _ . . . _ _ . _ . . _ . _ _ . _ _ . _ , . . , . . _ . . _ . _ , . - . _ . . . _ _ _ _ _ - . . . _ . _ . _ _ _ . - _ . , _ , _
ACNGS-PSAR 5 ( ' Contractors and Vendors that supply safety-related structures, systems and components are also required to establish and implement an audit program of 91 11,11 17.2.18 their suppliers. These audits shall be accomplished in accordance with 17.2.44 written procedures or checklists and shall assure that the supplier Quality Assurance Programs are being implemented and ef fectively comply with appli-cable procurement specification requirements. Refer to Section 17.1.2B for applicability of this section. 33(U) 4 I O l l 4
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. (U)-Update 17.1-53 Am. No. 57, (5/81) 1 r (
ACNGS-PSAR
)
l l TABLE 17.1.1A-1 HAS BEEN 57 INTENTIONALLY DELETED
i i l 1 l 17.1-54 Am. No. 57, (5/81)
l ACNGS-PSAR
.,ss l TABLE 17.1.2A-1 ! "(s_
PROGRAM COMPLIANCE MATRIX 57 f . SECTIONS ADDRESSING 10CFR50 APPENDIX B REQUIREMENTS 57 HL&P NUCLEAR QUALITY I 10CFR50 APPENDIX B CRITERIA ASSURANCE PROGRAM ACNGS QA PLAN
- 1) Organization 2.0 2.0, 3.2
- 2) QA Progrse 1.0 & Figure 1.1. 1.0 33(U)
- 3) Design Control . 4.0, 5.3.12, 7.2, 4.0 8.4.2
- 4) Procurement Document Control 3.0 5.0, 7.2
- 5) Instruction, Procedures, and Drawings 1.3, 4.0, 5.3.1 3.0, 6.3 3.3, 3.4, 6.3, 57
-l
- 6) Document Control 1.3, 4.6, 5.3.6, 6.2.1 7.0, 8.0
% 7) Control of Purchased Material, Equipment, and Services 3.0, 8.4.1 5.0, 6.2, 6.3 17,28
- 8) Identification and Control l of Materials, Parts, and Q2- 11.14 Components 5.3.2, 6.2.2 6.2, 6.3 33(U)
- 9) Control of Special Processes 5.3.5, 5.3.8 6.2, 6.3 i
- 10) Inspection 6.1.3, 8.4.5 6.2, 6.3 l
- 11) Test Control 6.1.3, 7.3 6.2, 6.3
- 12) Control of Measuring and Test Equipment 5.3.7 6.2, 6.3
\ l l l s l s/
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(U)-Update 17.1-55 Am. No. 57, (5/81)
ACNGS-PSAR TABLE 17.1.2A-1 (Cont'd) 4
\
SECTIONS ADDRESSING 10CFR50 APPENDIX B REQUIREMENTS
.10CFR50 APPENDIX B HL&P NUCLEAR QUALITY l57 CRITERIA ASSURANCE PROGRAM ACNGS QA PLAN I
- 13) Handling, Storage, and Shipping 5.3.3, 8.4.7 6.2, 6.3 17,28
- 14) Inspection, Test, and Operating Status 5.3.4, 5.3.11 6.2, 6.3 Q2-11.14
- 15) Nonconforming Items 5.3.9, 9.0, 8.4.6 3.8, 6.2, 6.3 33(U)-
- 16) Corrective Action 5.3.9, 9.0 3.8, 6.2, 6.3
- 17) QA Records 3.5, 4.6, 5.3.6, 6.2, 6.3, 7.0 6.2.1, 7.4, 8.4.8
- 18) Audits 8.4.9, 9.0
('~ ( i i l (U)-Uodate 17.1-56 Am. No. 57, (5/81) l 4
~ w ,- - - -,--e.n-v, ,. .- --
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ACNGS-PSAR s TABLE 17.1.5A-1 O PROJECT PROCEDURE MATRIX Procedure 10CFR50 App. B I. .Proiect Procedures Number Critarion 1). INTRODUCTION
-Purpose and Scope of 1he Manual ACPP-lQ VI Issuance and Control of the Manual ACPP-2Q VI Organization and Responsibility ACPP-3Q I of Project Personnel
- 2) PROJECT ADMINISTRATION Processing Correspondence ACPP-52Q VI Generating Telephone Minutes ACPP-53Q VI 57 Calling Meetings ACPP-54 N/A Project Trip Approval ACPP-56 N/A l
v Issuing and Controlling Project ACPP-58Q II,V,VI Directives j 1 Administrative Training - ACPP-59Q II
- 3) COST / SCHEDULE Processing the Project Authorization ACPP-101 N/A Request Annual Budget Development and ACPP-102 N/A Control j Project Estimate Review ACPP-103
- N/A
- 4) ENGINEERING REVIEW i
l Design Review ACPP-152Q III . Design Review Authority ACPP-153Q III,1 AE Design Change Notice (DCN) ACPP-154Q III,VI Processing
\
17.1-57 Am. No. 57, (5/81)
ACNGS-PSAR
/ ~% '
'I h TABLE 17.1.5A-1 (Cont'd)
Aj Proc oure 10CFR50 App. B
-1. Project Procedures Numb.: , Criterion i Design Change Control within HL&P ACPP-155Q III
. Transmittal of Design Review Comments ACPP-156Q III to Other Organizations Engineering Review of Procurement ACPP-157Q III,IV Do cumen'.s Engineering Training ACPP-158Q II Designation and Handling of Confi- ACPP-159Q VI dential Security Documents Processing Supplier Deviation Requests ACPP-160Q III,IV
- 5) PROCUREMENT 57 General Procurement Procedure ACPP-201Q IV,VII Procedure for Establishing Bidders ACPP-202Q IV,VII List'
\
Inquiry Issuance ACPP-204Q IV,VII Proposal Evaluation and Suppliers ACPP-205Q IV,VII, XVII Selection Purchase Order, Preparation, Changes ACPP-206Q IV,VII Approval, and Issuance Training ACPP-208Q I Procurement File System and File ACPP-210Q IV,VI Control Document Review ACPP-211Q IV,VII Document Control ACPP-212Q IV,VI Material Control Organization ACPP-213Q VII,VIII, ALII,XIV Preventative Maintenance ACPP-214Q VII,VIII, XIII,XIV
\s-17.1-58 Am. No. 57, (5/81) l l
1 ACNGS-PSAR O) ( v TABLE 17.1.5A-1 (Cont'd) Procedure 10CFR50 App. B I. Project Procedures Number Criterion Spare Parts ACPP-215Q IV,vII, XIII,XIV
- 6) ACCOUNTING Invoice Review and Approval ACPP-251 N/A Vendor Back Charge ACPP-252 N/A
- 7) LICENSING Reporting Design and Construction ACPP-301Q XV,XVI Deficiencies to NRC Handling of NRC Inspection Reports ACPP-302Q VI and Immediate Action Letters Review of NRC Inspections and ACPP-303Q VI 57 4
. Enforcement Bulletins and Circulars
- 8) CONSTRUCTION (Later)
- 9) START-UP/ WARRANTY (Later)
- 10) PROJECT DIRECTIVES Authorization to Approve for Release AC-PDIR-2 I,VI Certain Project Correspondence i Storage Procedure for Early Deliveries AC-PDIR-3 VIII,XIII, or Material for the Allens Creek XIV Nuclear Generating Station II. Project Site Quality Procedures (PSQP)
Organization & Responsibility of Project PSQP-Al I,II QA Personnel Project Site Quality Procedures PSQP-A2 V,VI Handling of NRC Inspection Reports PSQP-A3 XV,XVI J. ! Control of Site Documentation PSQP-A4 VI 4 , N l 17.1-59 Am. No. 57, (5/81)
ACNGS-PSAR 1 i. p) TABLE 17.1.5A-1 (Cont'd)
- x. /
Procedure 10CFR50 App. B '- Project Site Quality Procedures (PSQP) Numbe r Criterion II. Non-Nuclear Site Quality Assurance PSQP-A5 N/A PSQP-A6 V,VI Document Reviews Stop Work PSQP-A7 XV,XVI Trend Analysis Administration PSQP-A8 XVI Implementation Review PSQP-A9 II,IV thru XVII XVIII Audit Overview PSQP-A10 Vendor Surveillance Overview PSQP-All IV,VII Construction.QA - Operations QA Interface PSQP-Al2 II,XI, XVII 57 III. HL&P Houston Quality Assurance Procedures
\s_,) Indoctrination & Training of HL&P Houston QAP-2.1 11 QA Home Office Personnel Procedure for Qualification and Certifica- QAP-2,2 11 tion of Surveillance Personnel Training and Qualification of Audit Per- QAP-2.3 II sonnel Procedure for Document Review QAP-3.1 III Procedure for Procurement Document Review QAP-4.2 IV V
Standard Definitions and Abbreviations QAP-5.1 Standard Format for Writing and Control- QAP-5.2 V ling Issuance and Control of Documents QAP-6.1 VI Procedure for Vendor Quality Surveillance QAP-7.3 VII lO 17.1-60 Am. No. 57, (5/81) l
l ACNGS-PSAR
;
i TABLE 17.1.5A-1 (Cont'd) s Procedure 10CFR50 App. B III. HL&P Houston Quality Assurance Procedures Numbe r Criterion QAP-7.4 VII HL&P Vendor Surveillance QAP-7.5 VII Second Party Vendor Surveillance Category I Second Party Vendor Surveillance Category QAP-7.6 VII II Review of Nuclear. Steam Supply System Qua- QAP-7.7 VII lity Assurance Records Packages QAP-15.1 XV Control of Nonconformances QAP-15.2 XV Stop Work' Procedures Corrective Action QAP-16.1 XVI Audit Filing QAP-17.1 XVII O HL&P 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 System Procedures Records Management Responsibilities & 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 within T1-1 XVII RMS Center Document Logging T2-1 VI
'N 17.1-61 Am. No. 57, (S/81) l l ._ - .-
ACNGS-PSAR TABLE 17.1.5A-1 (Cont'd) Procedure 10CFR50 App. B Records Management Systems Procedures ~ Numbe r Criterion IV. Log Maintenance T2-2 VI Document Distribution T2-3 VI Storage & Maintenance of Nuclear Records T2-4 XVII i Document Checkout- T2-5 XVII 57 Correspondence Serial Number Assignment T2-6 VI Correspondence Serial Number Corrections T2-7 VI Subject File Number Assignment T2-8 XVII g. NSSS Data Package Handling T2-10 XVII ( I N 4 0 ) j- 17.1-62 Am. No. 57, (5/81)
ACNGS-PSAR TABLE 17.1.2B-1 []
~\ r EBASCO NULCEAR QA MANUAL TOR PROJECT NO.
TABLE OF CONTENTS QC-1 INT"0 DUCTION AND_STATENENT OF AUTHORITY QC-2 ORGANIZATIONAL RESPONSIBILITIES QC-3 PERSONNEL QUALIFICATIONS QC-4 DESIGN AND ENGINEERING QC-5 PURCHASING PROCEDURE QC-5.1 Quality Control. Requirements for Suppliers of Equipment and Services QC-5.2 Vendor Quality Compliance Representation Program QC-5.3 Evaluation of Vendors with Respect to Quality Assurance Capability QC-6 FIELD CONSTRUCTION QC-6.1 Quality Compliance Site Supervisor QC-6.2 Documentation and Audits at the Construction Site Qc-6.3 Receiving, Handling and Storage QC-5.4 Housekeeping, Safety and Security QC-6.5 Soils control . QC-6. 6 Concrete and Reinforcing QC-6.7 Welding, Joining and Nondestructive Examination Q C-6. 8 Mechanical Equipment Q C-6. 9 Electrical Equipment QC-6.10 Instrumentation and Controls QC-6. ll Calibration QC-6.12 Pressure System Cleaning QC-7 MANAGEMENT AUDITS l Appendix I DEFINITIONS Ov Appendix II FORMS 17.1-63 (U)-Update Am. No. 57, (5/81)
ACNGS-PSAR. TABLE-17.1.2B-2 CROSS REFERENCE OF EBASCO QA MANUAL (\ -) WITH 18 QA CRITERIA 2 . To demonstrate the manner in which Ebasco meets the intent of the Atomic
- Energy Commission's Rules'and Regulations, Title 10, Part 50, Appendix B -
Quality Assurance Criteria for Nuclear Power Plants, the following table is presented: Quality . Sections of Ebasco Quality Assurance Program Covering Quality Criteria Assurance Criteria I - Organization QC-1,QC-2,QC-4,QC-5,QC-5.1,QC-5.2,QC-6, QC-6.1,QC-7 II - Quality. Assurance Program Entire Program III.- Design Control QC-2,QC-4,QC-5,QC-5.1,QC-5.2,QC-6,. QC-6.1 QC-4, QC-5, QC-5.1, QC-5.3 33(U IV - Procurement Document Control V - Instructions, Procedures and Drawings Entire Program 1 ps VI - Document Control QC-4,QC-5,QC-5.1,QC-5.2,QC-6,QC-6.1 ss VII - Control of Purchased Material, QC-2,QC-5,QC-5.1,QC-5.2,QC-5.3, 1 Equipment, and Services QC-6.1,QC-6.2,QC-6.3,QC-7 VIII - Identification and Control of Materials,. Parts, and Compon-ents QC-2,QC-5.1,QC-5.2,QC6.1,QC-6.2,QC-6.3
. IX - Control of Special Processes QC-2,QC-5.1,QC-5.2,QC-6.1 through 6.12 X - Inspection QC-1,0.C-5.1,QC-5.2,QC-6,QC-6.1 through 6.11 XI - Test Control QC-2,QC-4,QC-5.1,QC-5.2,QC-6 through 6.11 XII - Control of Measuring and Test Equipment QC-5.1,QC-5.2,QC-6.1,QC-6.11 k
XIII - Handling, Storage and Shipping QC-5.1,QC-5.2,QC-6 through QC-6.3 1 I (U)-Update fA _/)~ s 17.1-64 . Am. No. 57, (5/81) w -- - - - - s.-- e 4- .-w ,w. --.,e,-ww.,- _ , . ,,u.,,m. ,. ,. *,, . , , , , py,...--,n.%wr:--,,p,~..-.yye>- t-ei
1 j ACNGS-PSAR - l i TABLE 17.1.2B-2 (Cont'd) e [ s XIV.- Inspection, Test, and Operating Status QC-5.1,QC-6.3 XV - Nonconforming Materials, Parts or Components QC-5.1,QC-5.2,QC-6,QC-6.1,4C-6.3 XIV - Corrective Action QC-5,QC-5.1,QC-6,QC-6.1,QC-7 XVII -. Quality Assurance Records QC-4,QC-5.1,QC-5.2,QC-6 through 33(U) QC-6.12,QC-7, Appendix II , XVIII - Audits QC-2,QC-5,QC-5.1,QC-5.2,QC-5.3,QC-6,1, QC-7, Appendix II i s ? i 1 i i e i -
1
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I l l 17.1-65 (U)-UPd ate Am. No. 57 (5/81)
ACNGS-PSAR TABLE.17.1.2B-3 m) (
'V Proj ect-Related .
Clarifications, or Modifications To Ebasca To pical Re port ETR-1001 Rev._7 49
- 1. General Where the word " client" appears within the appropriate sections of EBASCO's Nuclear Quality Assurance Program Manual it shall be under-stood to mean' " Houston Lighting & Power Company".
- 2. Deleted 49
- 3. Section QA-I-4 Design Control Although Figure 1-4.1 in Section QA-I-4, leaves the required review up 45(U) to the discretion of the Lead Discipline Engineer, the Project Quality Assurance Engineer shall review all bidders lists, vendor prorosals and Ebasco purchase orders to Vendors.
- 4. Section QA-I-5 Quality Assurance Evaluation of Suppliers / Contractors 4.1 - Paragra th s 3.1.2 and 5.1 are modified to allow for alternate O methods of evaluation and qualification of supplier's capabilities by
(/ methods other than audit by Ebasco. Such methods are detailed as fol-lows: a) Audits of suppliers by HL&P or others qualified to do so. b) Historical data is available substantisting the capability of the supplier to provide products which have performed satisfactorily 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 product s in the gast. 4.2 - Paragra phs 2.3, 4.1 and 5.1 are modified such that in the event Construction Contractors are awarded a contract be fore review and ap-proval of their quality assurance manual or their facility, but prior to start of any safety-related work, the following shall be complied with: a) Tne " Terms and Conditions" section of the Purchase Order 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, C) o (U)-Update 17.1-66 Am. No. 57, (5/81) l
ACNCS-PSAR TABLE 17.1.2B-3 (Cont'd ) m
- 2) a satisf actory quality assurance audit by Purchaser of tha construct-
[j\
'\ ion contractors Quality Assurance Program. If the manual review and/
or audit are unsatisf actory, and if, in the opinion of the Purchaser there is no hope of successful corrective actions, the terms of the contract will permit Purchaser to absolve himself of the contract. b) A visit will be made to the home of fice of the contractor to discuss the techniques they intend to use in implementing their program at the Cons truction site, c) Records of past or similar jobs shall be examined at the contractors of fice to verify implenentation of constructors quality assurance program 45(U) or at least make an evaluation of the contractors qualifications and capability, d) Ihe results of the proceeding review will be forwarded to HL&P for their concurrence.
- 5. Section QA-I-6 Quality Assurance Records The requirement of this section i3 modified to impose upon the File Custodian the responsibility for the review of records that will become -
part of Ebasco's permanent files. The documents will be inspected for legibility, proper identification and microfilmability. The require-ment for vendor quality assurance records to be legible and microfilm-
/] able will be imposed on vendors through Ebasco's Purchase Order Speci-fication.
- 6. Section QA-II-4 Purchasing Ebasco proposes only the recommendations for purchasing and has no responsibility for issuance of the purchase orders. HL&P has assumed 57 this responsibility and the system of control for issuance of these purchase orders is outlined in HL&P's Quality Assurance Program Manual.
Otherwise the remainder of this Section is applicable in its entirety.
- 7. Section CA-II-5 Supplier Surveillance Af ter issuance of a purchase order and prior to start of fabrication,
- the Project Quality Assurance Engineer prepares the Vendor Quality As-surance Plan for approval by HL&P QA. Af ter approval, the PQAE for-wards the Vendor Quality Assurance Plan to the Chief Vendor Quality 45(U)
Assurance Representative for use by the Vendor Quality Assurance Representatives.
- 8. Section QA-III-l Instructions, Procedures and Drawings (Section QA-III- l57 u
Ihe following changes are hereby established so that Section QA-III-1 l48 s0' is compatible with Section QA-I-2 of the Ebasco Nuclear Quality As-surance Program Manual - Allens Creek Project. 45(U)
%./
(U)-Update 17.1-67 Am. No. 57, (5/81)
ACNGS-PSAR TABLE 17.1.2B-3 (Cont'd)
, l Sy'
[ 7.1 Paragraph 3.6 Inst line is modified to read: General review of [
\ site originated quality control documents shall be performed by the 1 45(U)
Quality Program Site Manager, or his designee. l 57 7.2 Paragraph 3.7 is replaced .by the following: Qualifications and certification for Site Quality Control personnel are covered in Quality Control Procedure QC-3 " Indoctrination, Training and Qualifications." Qualifiert on Records will be maintained by the Records Center Super-i visor for 11 Quality Control personnel assigned to the site. These records will be audited initially by the Chief Quality Assurance Engi- 45(U) neer, or his designee, for compliance with the qualification . require-ments and thereaf ter on an annual basis by the Quality Assurance Site Supe rviso r . The Quality Control Site Supervisor shall coordinate and direct the training and qualifications of Quality Control personnel. Any deficiencies shall be reported in writing by the Chief Quality As-surance Engineer, or his designee, to the Quality Control Site Super-visor for resolution. Pending resolution of the deficiency, the in-F dividual shall be restrained from performing any Quality Control acti-vities. The Quality Program Site Manager is assigned to the project construction site on a resident basis. The Quality Program Site 57 - Manager is responsible for the Quality Assurance and Quality Control functions at the construction site. H e Quality Program Site Manager
- will have full authority to monitor, contro!, direct and report on all of the work nder the Quality Program. 45(U)
\'
he will is terf ace directly with the HL&P Quality Assurance Manager, or his designee and report to the Chief Quality Assurance Engineer. He L 57 is subordinate to no individual on site, has the independent authority 4 45(U) and responsibility to identify site quality related problems. As-sisting the Quality Program Site Manager at the site will be the 1 57 Quality Assurance Site Supervisor, the Quality Control Site Supervisor 45(U) and the Records Center Supervisor. l 57
- 9. Section QA-III-2 Document Control Paragraph 2.2 is replaced by the following: The Quality Control Site 45(U)
Supervisor shall be responsible for the control of Ebasco Quality Con-trol procedures and plans at the construction site.
- 10. Section QA-III-4 Construction Site Procurements 10.1 In Paragraph 2.5 the Senior Quality Control Supervisor shall be understood to mean the Quality Program Site Manager. Se follow- l 57 ing clarification of the term " Direct Evaluation" in Paragragph 2.7.3 is given: 'Under certain circumstances and to assist the vendnr evalua-tion group and to expedite the vendor evaluation process, tb site QA 45(U)
Supervisor or qualified members of his staf f wh) he may appoint, may perform facility audits, primarily in their local geographic areas, j l 1 v (U)-Update
)
i 17.1-68 Am. No. 57, (5/81) l
.. ~ _ _ __.
ACNGS-PSAR TABLE 17.1.2E-3 (Cont'd) 10.2 In Paragraph 3.1.1.b the Senior Quality Control Supervisor shall I
' ~'
) be understood to mean the Quality Control Site Enpervisor.
10.3 In Paragraph 3.2.1 delete subparagraph (b). 45 (U) 10.4 In Paragraph 3.3 the Senior Quality Control Supervisor shall be understood to mean the Quality Control Site Supervisor.
- 11. Section QA-111-5 Supplier / Contractor Surveillance 1 In Paragraph 3.4 the Senior Quality Control Supervisor shall be under-stood to mean the Quality Control Site Supervisor.
- 12. Section QA-III-6 Nonconformances The following clarification of Paragraph 3.3 is given. For those l 57 nonconformances detected at the site the Quality Control Site Supervi-
< sor will initiate the nonconformance report. 45(U)
- 13. Section QA-III-14 Receiving Inspection (Section III-14) g 57 The following clarification of Paragraph 2.0 is given: When vendor -
surveillance is not required for certain items purchased by the site ,
*5(U) organization, receiving inspection will include the review of Certi, eg fied Material Test Reports, NDE Records, etc. In these cases the re-( ) view of such documents will be the responsibility .of the site quality \s_s/ control ceganizatian. In turn, this operation would be audited by the Site Quality Assurance Group.
57 i (
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(U)-Update 17.1-69 Am. No. 57, (5/81)
ACNGS-PSAR
- TABLE 17.l.2B-4 O .Organizationa11and Administrative Changes 49
(
~
to Ebasco Topical Report ETR 1001 Rev. 7 , I Site Quality Control Engineering 5 A. Quality Control Site Supervisor and staf f of engineers and specialists , are assigned to, each . project construction site on a resident basis. The
' Quality Control Site Supervisor reports to the Quality Program Site Manager and is reaponsible for: 49 - . (a) Performing inspection in all areas of construction, establishing and - '
- enforcing quality control. documentation requireIPents, including pro-cedures, specifications, drawings and purchasing documents. -
(b) Identifying and initiating correction of nonconformances to require-l ments' indicated by the drawings, s pecifications, codes or procedures for items, and rejecting nonconforming items and. services or when ' I necessary requiring the stoppsge of work until such nonconformance is' g corrected. 6 (U) ; 1 r (c) Assisting in organizing and administering training seminars as re-i quired to assure proper level of quality control. .
(d) Preparing inspection requirements based upon such documents as speci-fications, drawings, codes and standards, as established by the { Engineering Department. i j.
; (e) Supervision of Quality Control Engineers who directly supervise ;
j Quality Control Inspectors / Specialists f w the various construction disciplines (Soils, Concrete, Electrical, Mechanical, Material Con-i trol, etc). ( i (1) Supervision of the NDE Group who are responsible for performance and/ j or monitoring of nondestructive testing activities, f Records Center A Quality Records Supervisor and staf f are assigned to the project construc- .l49-
- tion site on a resident basis. The functions of the Records Center Super-visor are as follows
(a) Establish . l. Project file indexing and location, including retention and i classification requirements.
- 2. Filing and storage- instructions for special grocess records, 1 i following manufacturer's recommendations and/or established l
gract ices . (b) ~ Develop record audit checklist.s and process Records De ficiency j Re port s. (U)-Update ; 17.1-70 Am.~No. 57, (5/81)
-- - - , . . -, .-- = _ . . - . _ .
ACNGS-PSAR TABLE 17.1.2B-4 (Cont'd) (p-- ) (c) Receive, classify and store Quality Assurance records. V (d) Maintain log documentation and/or records to control records removal from the record center. (e) Review record and/or document packages on Ebasco purchased items or services received from Site Quality Assurance or Site Quality Control for completeness and correctness. (f) Audit sample records and documentation on NSSS components for com-ponents for completeness and correctness. (g) Act as the Access Control Authority for the Project Record Vault. (h) Maintain site qualification and certification records for NDE, welding, inspection, examination and testing. Quality Assurance Specialists 45 Quality Assurance Engineering has several specialty groups responsible for (U) the following activities which are performed in accordance with QA proce-dures: (a) Inservice Inspection [ h (b) Qualification and certification of personnel as required by applicable . \s_ / J codes or standards. (c) Development of Quality Assurance standards and procedures. (d) Review, evaluation and summarization of Code and Regulatory Quality Assurance requirements. (e) Evaluation of suppliers' Quality Assurance Program. (f) Quality Assurance education, both internal and external to Quality Assurance Engineering. (g) Interdepartmental audits of all individuals or groups responsible for activities covered by the Quality Program. - Nondestructive Examination Quality Assurance This group, under the Quality Assurance Site Supervisor, provides expertise with regard to conducting various forms of NOE and includes the following
- functions. ,
I h) a ' (U)-Update 17.1-71 Am. No. 5", (5 /811
ACNGS-PSAR TABLE 17.1.2B-4 (Cont 'd ) ,
/ \ (a) Establish _ and/or intu pret NDE requirem'ents and acceptance criteria h for fabricated and erected equipnent as required. ! (b) Review and comment on NDE procedures and radiographic films submitted by manufacturers, site construction forces and/or Clients.
(c) Advise manufactueer and site construction forces as to tro pt NDE pro-cedures, a pplicat ions , techniques, equ'i pnent and quali fications. Conttruction , Primary responsibility for construction rests with the Vice-President of Con-struction (re fer to Figure 17.1.2B-1 as amended by ETR-1001, Figure 1.26, 49 Revision 3 and Figure 1.2-7, Revision 3). (a) Construction Managers reprt to the Vice-President of Construction and are respnsible f or overall supervision and coordination of all 45(U) construction activities and services. (b) The Manager of Construction Services re ports to the Vice-President of Const uction and is res ponsible for general supervision of the Construction Engineering Group. , The Manager of Construction Engineering re ports to the Manager of Construction Services and is res ponsible for the inclusion of 4 p) (V quality requirements in Construction Contracts and review of Engi-neered Documents as required by the Quality Assurance Program Manual. (All construction contracts involving sa fety-related equipnent are subject to review by the Quality Assurance Engineering Department for compliance with the applicable code and regulatory agency require-ments and Quality Assurance Program requirements). (c) For individual proj ec t s , the Site Manager re port s to a Construct ion Manager and has the responsibility for direction and coordination of all onsite activities associated with the construction of the pl ant . (d) The Project Su perintendent re ports to the Site Manager and is 49 res ponsible for prfor:ning general site supervision of construction in accordance with drawings, specifications and contractual obligations. (e) The Construction Superintendent reports to the Project Su perint endent and has the res ponsibility of assuring that jobsite fabrication and installation is in accordance with drawings, specifications and other prevailing document s. (f) Area Superintendents re port to the Project Superintendent and are res pnsible for area planning and scheduling, area construction control engineering and area construction engineering. A
'M' (U)-Update 17.1-72 Am. No. 57, (5/81)
- _. . . . . . . _ - - .~
ACNGS-PSAR (g) The Senior Resident Engineer reports to the Project Superintendent n and is-responsible for all- phase of field of fice and field en- .( gineering. The Administration Manager reports to the Site Manager, and is (h) responsible for management of site of fice services, including 49 purchasing, materials administration, data processing and ' accounting. (i) The Purchasing Administrator reports to the Administration Manager and is responsible for the issuance and control of purchasing documents between vendors and personnel at the jobsi te. ! (j) The Material Administrator reports to the Administration Manager and is responsible for commercial receiving spection, storage and issue of materials at the site. 4 i i
I !/ k i 17.1-73 Am. No. 57, (5 /81)
ACNGS-PSAR t l l l i l l 1 I T,.ble 17.1.2C-1 has been deleted. t L l t 17.1-74 Am. No. 57, (5/81)
ACNGS-PSAR Table 17.1.5C-1 has been deleted. l l l \ l 17.1-75 Am. No. 57, (5/81 l l
o t I i 6 EXECUTIVE
; VICE PRESIDENT J L DIRECTOR VICE PRESIDENT NUCLEAR NUCLEAR ENGINEERING FUELS AND CONSTRUCTION i ! PROJECT MANAGER NUCLEAR ALLENS CREEK FUEL PROJECT Jk A GE-NEBG MANAGEMENT GE-N F&SD MANAGEMENT II QUALITY OUALITY PROJECT ASSURANCE ASSURANCE MANAGER MANAGER MANAGEP.
J L J L J k i h l l
ACNGS - PSAR , l
> l l
I MANAGER QUALITY ASSURANCE PROJECT QUALITY ASSURANCE MANAGER J L EBASCO MANAGEMENT PROJECT CHIEF QUALITY MANAGER ASSURANCE ENGINEER d J k J L GE-NPSD MANAGEMENT C QA LINES OF COMMUNICATION 4 PROJECT LINES OF COMMUNICATION V PROJECT MANAGER J L AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 , EXTERNAL QA RELATIONSHIPS FIGURE 17.1.1 A-1
l s
/
} .
l l i ' DIRECTOR NUCLEAR FUELS I l i l l 1 PROJECT PROJECT NUCLEAR ENGINEERING FUEL IkiANAGER l I Pf ADMll SUP l SUPERVISING Pt 1 iliOJECT COh ! ENGINEER (S) { 6 s I
ACNGS - PSAR -
)
)
1 i EXECUTIVE VICE PRESIDENT VICE PRESIDENT MANAGER NUCLEAR ENGINEERING OUALITY AND CONSTRUCTION ASSURANCE I MANAGER PROJECT l ALLENS CREEK OUALITY ASSURANCE PROJECT MANAGER I I lOJECT PROJECT CHASING CONSTRUCTION LNAOER MANAGER .lOJECT PROJECT $1STRATION CONTROLS - ERVISOR MANAGER l l L PROJECT i ECT ENVIRONMENTAL VROLLER ENGINEER l l AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY e Allens Creek Nuclear Generating Station Unit 1 ALLENS CREEK PROJECT ORGANIZATION FIGURE 17.1.1 A-2 I
- w - - - - - - - ,
.,.wn a -
s f
OPERATIONS QA MANAGER
; A h
I NDE/ISI PROJECT j OPERATIONS QA QUALITY ASSURANCI SUPERVISOR GENERAL SUPERVISO PROJECT OA P TM ' SUPERVISOR MECHANICAL PROJECT QA SUPERVISOR CIVIL M HOME OFFICE SITE PROJECT QA SUPERVISOR ELECTRICAL l ) 1 k____.__._______ _ _
l ACNGS - PSAR I o MANAGER QUALITY ASSURANCE A PROJECT HOUSTON OUALITY AS4URANCE OA MANAGER MANAGER
/ A PROCUREMENT SUPERVISOR l PROJECT OA AUDITS & TECHNICAL R SUPERVISOR SERVICES
/ / A SUPERVISOR SUPERVISOR QUALITY VENDOR SYSTEMS SURVEILLANCE J AM NO.57, (5/81) i HOUSTON LIGHTlHG & POWER COMPANY Allens Creek Huclear Generating Station Unit 1 I HL&P QUALITY ASSURANCE ORGANIZATION FIGURE 17.1.l A-3
'l ACNGS-PASR LIST OF EFFECTIVE PAGES APPENDIK C ( Pate No. Amendment No. l 1* 57 2* 56 l l! - 3* 57 i 42 11 42 lii 42' , iv 42 v 42 vi 42 vii 42 viii 42 C1.1-1 17 C1.2-1 17 i C1.3-1 35 C1.4-1 17 i C1.5-1 35 I C1.6-1 35 C1.7-1 35 l C1.8-1 42 C1.9-1 35 C1.10-1 35 j C1.11-1 17 C1.12-1 35 C1.13-1 35 C1.14-1 17
- C1.15-1 17 C1.16-1 35 C1.17-1 35 i C1.18-1 35 l C1.19-1 17 C1.20-1 42 C1.21-1 35 C1.22-1 31
; C1.23-1 35 C1.24-1 17 C1.25-1 35 C1.26-1 42 I C1.27-1 42 C1.28-1 46 t C1.29-1 42 C1.29-2 42 C1.30-1 45 , C1.31-1 42 C1.31-2 (deleted) 42 C1.31-3 (deleted) 42 a
- Effective Pages/ Figures List 1 Am. No. 57, (5/81)
ACNGS-PASR LIST OF EFFECTIVE PAGES (Cont'd) APPENDIX C Page No. Amendment No. 35 I C1.70-1 C1.71-1 35 C1.72-1 17 Cl.73-1 22 C1.74-1 22 C1.75-1 35 C1.76-1 35 C1.78-1 35 C1.80-1 35 C1.84-1 42 I C1.85-1 42 C1.86-1 35 C1.88-1 47 C1.89-1 35 C1.91-1 42 C1.92-1 42 C1.92-7 42 C1.93-1 35 C1.94-1 47 l C1.95-1 42 C1. 96-1 35 j C1.97-1 57 f C1.97-2 57 i C1.98-1 42 ] C1.98-2 42 C1.99-1 42 I C1.100-1 42 i C1.101-1 42 l C1.102-1 35 C1.104-1 42 C1.105-1 42 C1.106-1 42 C1.108-1 42 C1.109-1 42 C1.110-1 42 C1.111-1 42 C1.112-1 42 C1.113-1 42 C1.114-1 35 C1.115-1 42 C1.116-1 46 C1.117-1 42 C1.118-1 42 I C1.120-1 42 C1.122-1 42 C1.123-1 42 s C1.124-1 48 C1.126-1 42 C1.127-1 42 C1.130-1 48 3 A'n. No. 57, (5/81)
ACNGS-PSAR REGULATORY GUIDE 1.97 . 42(U)
\ Rev. 2, 12/80 57 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PIANTS TO ASSESS PLANT 42(U)
CONDITIONS DURIl3G AND FOLLOWING AN ACCIDENT. , I ! Applicant's Position: 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 i the'BWR is re. actor 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-esived benefits of having them available. 'the detalled arguments against thermocouples are presented in attachment 1 of a letter from D. B. Waters (BWOG) to D. E. Eisenhut (NRC) titled "BW i Emergency l'rocedures Guidelines, Rev. I and responses to related questions", dated 1/31/81.
-Radiation Exposure Rate:
- a. to indicate containment breach: Radiation monitors in containment
- penetration areas are not considered useful indicators of a contain-ment breach. Following an accident in which a large quantity of Q radionuclides were released to containment, the ' reading from these 57 monitors would rise sharply even without a breach. A monitor set-point indicating breach could not be established with any reasonable confidence because this is dependent on parameters and phenomena such as degree of fuel damage and transport and plateout patterns which vary extensively for different or even the same postulated accidents. In the absence of a well determined setpoint, the opera-tor is lef t with confusing indication subject to interpretation
- based on guesswork.
- b. in areas where access to service safety equipment is required: Pre-I sumably these are included to insure accessibility of personnel to service safety equipment post-accident. Hand-held portable monitors are adequate for this purpose.
-Cooling Water Temperature to ESF components: A range of 30-120 y is provided versus the 30-200 F of the guide. ACNGS uses take 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:
i~ C1.97-1 (U)-Update i Am. No. 57, (5/81)
l l ACNCS-PS AR I __s -Drywell Sump Level and Drywell Drain Sump Level: These are inter-
' preted to mean the Low Purity Drywell Sump and High Purity Drain Tank Drywell.
-Containment and Drywe11' Oxygen Concentration: These are inter- !
preted to be required only for pre-inerted containments. ; t
)
-Suppression Chamber Spray Flow: This is applicable to Mark 11 con-tainment ' des ign s. l
-Drywell Spray Flow: AChCS has no drywell sprays.
! 57 i
-Isolation Condenser System Shell Side Water Level: ACNCS has no isolation condenser.
t
-HPCI Flow: This is interpreted to mean HPCS flow. j
-Core Spray System Flow: This is interpreted to mean LPCS flow.
-High Radioactivity Liquid Tank Level: This is interpreted to mean
- Liquid Radweste System Collection Tanks Level.
-Type A parameters for ACNCS are given in Section 7.5.1.4.2.
1 e .F f I { 4 f i
C1.97-2 Am. No. 57, (5/81) i 4
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ACNGS-PSAR ] EFFECTIVE PAGE LISTING j APPENDIX 0 l PAGE No. AMENDMENT No, ! 0-41 57 i 0-42 57 0-43 57
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ACNGS-PSAR APPENDIX 0
\
l i RESPONSES TO NUREC 0718, " LICENSING REQUIREMENTS FOR PENDING APPLICATIONS FOR , CONSTRUCTION PERMITS AND MANUFACTURING LICENSES" (MARCH, 1981) '
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 , applicaticns. The Action Plan items are divided into five categories of level ? of detail to be provided in the NTCP submittal. This appendix gives the ACNGS i responses to the Action Plan items by NUREG 0718 category. I i I
, e d
1 .I a l i i i l r f 4 I e I f iO
- 0-1 Am. No. 57, (5/81)
#- , - , -,.-.w-.,.. .-.24,,,,,, .-,.,e ,c. % <. . _ ,.,y ,,v.-, y,,,,y _..,,..,,,.,,_.,_,.-,..,_,_,_-.,,_.,_... rom ._,.7.~,_,g,,__,,.,,w
ACNGS-PSAR l NUREG 0718 CATEGORY 1
) "A requirement of :: type not applicable to the pending CP or ML applications l for any of the following reasons:
- a. It can only be addressed in' operating license applications or by licen-sees; 4
l b. It is not directed to CP or ML applicants; t 57
- c. t It does not apply to plants of the type now pending; t i
, d. It has been (or will be) superseded by a more restrictive requirement in ; the Action Plan or in ths regulations;
- e. It has already been completed."
i i j RESPONSE No response is required for NTCP applications. i
;
i 1 i I
i < 4 l
;
i t r 0-2 Am. No. 57, (5/81)
ACNGS-PSAR \ NUREC 0718 CATEGORY 2 ;
- "A requirement of the type customarily lef t 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 comunitment j to implement such requirements in connection with its application for approval j of the final design."
57
RESPONSE
i The Action Plan items in NUREG 0718 Category 2 concern mostly operations. HL&P has reviewed these items and has concluded that there is nothing to 2 preclude their implementation at the OL stage of licensing. These will be 4 addressed in detail in the FSAR. ] 1 1 I i r 4 i t i 1 J l l
- l i
ii iO 0-3 Am. No. 57, (5/81)
l
- ACNCS-PSAR 1
/3 NUREG 0718 CATEGORY 3 i -\ ' ,Y
" Studies (and other vecearch and development activities to provide design
~
development information) of the type customarily left for review at the final stage. However, to satisty 50.35(a)(3) the staff believes that items in this category should be completed as early as is practicable so that the results can be most effectively taken into account in developing final design de-tails. The spplicant 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."
RESPONSE
57 With the exception of Items II.B.8.1 (PRA study) and II.K.3.24 (Adequacy of , HPCI'and RCIC space cooling), the NUREG 0718 Category 3 items are being
, resolved between the NRC and the BWR Owners Group (BWROG). These items are j being addressed generically and are being conducted so as to envelope all plants participating in the studies. HL&P is participating in the BWROG efforts 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.
t A summary of the BWROG work done to date on each item is given herein. 4;
\
t i I a i ~ k i v 0-4 lbs. No. 57, (5/81)
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1. i- II.B.8 RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS . j NUREG 0718 REQUIREMENT ! " Applicant shall: 1 i (1)' commit to performing a site / plant-specific probabilistic risk assessment F and incorporating the results of the assessment into the design of the facility. The commitment must include a program plan, acceptable to the 1 staff, that demonstrates how the risk assessment program will be > l scheduled so as to influence system designs as they are being developed. 4
. The' assessment shall oe completed and submitted to NRC within two years !
, of. issuance of the ' construction permit. The outcome of this study and i' the NBC 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 consideied is an additional decay heat removal J
- . system whose functional requirements and criteria would be derived from the PRA study. L l It is the aim of the Commission through these assessments to seek such ' 57
!- 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 j with this.ain." ( 2 t RESPONSE ' HL&P commits to performing an ACNGS specific probabilistic risk assessment y (PRA) study. The program plan for this study, including schedule, is given in
- 4 Appendix 15 B.
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_7
. ACNGS-PSAR ITEM II.K.2.16 ' IMPACT OF RCP SEAL DAMAGE FOLLOWING ' SMALL-BREAK ' LOCA WITH LOSS .) 0F OFFSITE P0 DER ,
V - NUREC 0718 REQUIREMENT Applicant.s shall address ths requirement s set forth in the Commission Orders - issued to operating B&W plant s in May 1979 and set forth in Item B.4 of NUREG 0626 regarding the impact of reactor coolant . pump seal damage following a small break loss of coolant accident with loss of offsite power. Applicant s . with B&W-designed plants shall provide sufficient informat ion to describe the nature of the studies, how t hey are to be conducted, the completion dat es, and
;
t he ' program t o assure t hat the result s of such studies are fact ored int o t he , j final designs.- t SECY-81-208 ITEM II.K.3.25 EFFECT OF LOSS OF AC POWER ON PUMP SEALS i Applicant a with BWR plant s shall address the requirement s set forth in 'It em B.4 of NUREG-0626. Applicant s shall provide sufficient informat ion to describe the nature of t he st udies, how t hey a.*e t o be conduct ed, the ( complet ion dates, and t he program t o assure t hat t he result s of such studies
- are factored into the final designs.
1. { NUREG 0626 ITEM B.4 The licensees should det ermine by analysis or experiment , on a plant specific l basis, t he consequences of a loss of cooling wat er t o the reactor recirculation pump seal coolers. The pump seals should be designed to t , wit het and a complet e loss of alt ernat ing current power for at -least two b hours. , Adequacy of the seal design ~ should be demonst rated. i (a) Nat ure of St'udy 57' i This concern -relat es to the consequences of a loss of cooling wat er to the react or recirculat ion pump seal coolers. Adequacy of the seal design should be demonstrat ed. The recirculation pump design incorporates a dual mechanical shaf t seal. i assemoly t o cont rol leakage around the rotating shaft of the recirculat ion pump. Each assembly consists of two seals built int o a i cart ridge t hat can be ' replaced without removing the motor from the pump. t Each individual seat in the cartridge is designed for full pump design - pressure and can adequately limit leakage in the event that the ot her
~
j l seal should fail.
- j. Even t hough General Elect .ic uses two dif ferent recirc pump configurat ions, tne seal designs are essent ially the same. Both designs I use hydrostatically balanced mechanical shaf t seals. Subsequent discussion in this memorandum is applicable to both pump designs.
i The recire pump seals require forced cooling due t o t he t emperat ure of i ~ the primary reactor water and due to the frict ion heat generated in the sealing surfaces. - For 'all BWR/6 React ors two systems accomplish this forced cooling: (1) the equipment prot ect ion closed cooling wat er
/- system and.(2) L he seal purge syst em. Cooling wat er, provided by the equipment prot ect ion closed cooling wat er (EPCCW) syst em, flows through a heat exchanger around the. seal assembly. This EPCCW flow coots primary 4
- react or wat er which flows t o t he lower seal cavit yl t hereby maint aining the seals at the correct operat ing temperat ure. The ' seal purge syst em 0-6 Anh. No. 57, (5/81)
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ACNGS-PSAR
' OI inject s clean, cool water from the control rod drive system into the
' L( Lower seal cavity. This seal purge flow also provides an ef ficient cooling _ funct ion for the- seals.
' The seal cooling syst em described above is examined t o det ermine t he
- consequenses of a total loss of cooling on the ef fect iveness of recirculat ion pump'shaf t sealing.
I' (b) Conduct of St udy 4 Under normal condit ions', with the primary . react or syst em at or near rated t emperature and pressure and the recirc pumps eit her operating or
~
1 secured, both EPCCW and seal purge; are operating. These.two systens j maint ain the seal temperat ures at approximat ely 1200F. Recirculation pump vendor test dat a have shown that the pump seals may
- begin to deteriorate when seal temperatures exceed 2500F. If an event
! . occurs'where both closed cooling water to the pump seal _ heat exchanger and control rod drive seal purge flow are tot ally lost , the'recire pump seals will heat up. Vendor test dat a, t aken while operating at approximately 5300F/1040 PSIA, indicat e that the seals will heat up, 57 reaching 2500F approximat ely_.7 minutes af ter the tot al loss of cooling. ) Similar test dat a indicate t hat if either one of the seal cooling systems
; is operat ing, t he seal . temperat ures remain well below 2500F and no seal
{Q l deteriorat ion should occur. 1f both closed coolin; water and seal purge are tot ally lost , and if the 1 . sea ts heat up t o exc ed 2300F, seal det eriorat ion may occur, result in g l in primary coolant - leakage to the drywell. In order t o evaluate the
- fluid loss through a degraded seal, an analysis was performed using ths RELAP-4 computer program (see Reference 1).
This analysie modelled the fluid leakage path _as a series of fluid volumes with interconnecting junct ions, each having appropriate init ial
- condit ions. 'Also, the model assumed gross degradat ion of t he mechanical seals. Gross failure of these seals encompasses warpage, fractures and grooving of the seal f aces due t o excessive thermal gradient s and dirt .
The result s of this leakage analysis show that , even with gross degradat ion of_ the seals, the leakage would be less than 70 gallons per minute. This amount of leakage is wit hin normal react or fluctuations and t he normal vessel wat er level cont rol syst ens will easily compensat e for it . ~ Also, 70 GPM is much less than the bounding values of loss o f coolant accident analyses, hence there are no adverse effect s on _LOCA analyses.
4 j-- i I i. j 0-7 Am. No 57, (5/81) i r l _ _ . _ . - . . - - . . _ _ , . _ , . . _ _- _ ..m, - . - _ _ _ , , _ _ . ~ , , , , - . . . _ . . . _ , -
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I a \ l
.l ACNGS-PSAR' l i
(c) Complet ion Date l. The study is complete, and was t ransmit ted to the NRC in Reference 1. i (d) Program for Implement at ion of Result s The study concluded that the leakage through a grossly failed RCP seal is !
- of no consequence to any of the LOCA analyses. Tuerefore, no changes are i required to implement the result s. 57
REFERENCES t
- 1. NED0-24083, "Recirculat ion Pump Shaf t Seal Leakage Analysis",-November 1978. (Licensing Topical Report.)
a h i d 1 J 1 i f 4 ( l f l 0-8 Am. No. 57, (5/81)
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1 ACNCS-PSAR IIEM'II.K.3.13 SEPARATION OF HPCI AND RCIC SYSTEM INITIATION LEVELS - ANALYSIS AND IMPLEMENTATION NUREG 0718 REQUIREMENT 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
. 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.
NUREG 0626 ITEM A.1 t Currently, the reactor core isolation cooling (RCIC) system'and the high ' pressure coolant injection (HPCI) system both initiate on the same low water , level sign'al and both isolate on the same high water level signal. The HPCI system will restart on low water level but the RCIC will not. The RCIC system
'is a-low-flow system when compared to the HPCI system. The initiation levels of the Id'CI and RCIC system should De separated so that - the RCIC system i initiates at a higher water level than the HPCI system. Further, the RCIC i
system initiation logic should be modified so that the RCIC system will restart cui low water level. These changes have the potential to reduce the 5,, 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, i- V
RESPONSE
l a. Nature of Study s This concern covers two aspects of the HPCS and RCIC systems. The first ] concern is with the initiation levels of these two systems, and requests analysis to determine if benefit could be obtained from allowing the RCIC 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 fea ture .
i
, As previously confirmed in discussions with the NRC the fundamental issue
! of the separation of initiation setpoints (water level) is the potential ! benefit of reducing the number of thermal cycles on the reactor veasel and internals resulting from HPCI operation. It is noted that the Allens Creek plant employs HPCS which does not inject via the feedwater nozzle, 4 consequently the fatigue usage on this component is reduced. Thus the study of this issue, which~ was based mainly on the BWR/4 HPCI arrangement ; is conservative-for Allens Creek. t-i s 0-9 Am. No. 57, (5/81)
l I ACNGS-PSAR [, ,'} Analysis was also made to evaluate the proposed logic change for the RCIC
\s,,/ system which permits this system to restart automatically following isolation from high water level. This evaluation considered the logic changes involved, effect on system availability, impact on design reliability and.the operator / 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. Differences 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 Chapter 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.
A
- 2. Initiation of HPCI and RCIC following a sudden loss of feedwater. For these transients, inventory loss is rapid with 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 NUREG-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 existiag high level trip from the RCIC turbine trip valve to the steam supply valve. Once the level 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 i (O ~s 0-10 Am. No. 57, (5/81)
ACNGS-PSAR
,- 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 57 values be retained.
b) Automatic Restart of RCIC System The results of the analyses for this issue indicate that the proposed logic change would contribute to improved system reliability, be of assistance to the plant operator and generally enhance safety. This change can be incorporated into the design, and will be upon NRC approval of the BWROG study.
References:
~'
- 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".
- 2. Letter from D.B. Waters (BWR Owners' Group) to D.G. Eisenhut (NRC) dated December 29, 1980 and titled "BWR Owners' Group Evaluation of NUREG-0737 Requirements".
4 1 a
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ACNGS-PSAR ITEM II.K.3.16 REDUGTION OF CHALLENGES AND FAILURES OF RELIEF VALVES -
/ \ FEASIBILITY STUDY AND SYSTEM MODIFICATION ! /
NUREC 0718 REQUIREMENT Applicants with BWR plants shall address the requirements set forth in Item A.4 of NUREG-0526. 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 years 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 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, O t (3) Increased emergency core cooling (ECC) flow,
\s 'J 4
57 (4) Lower operating pressures, (5) Earlier initiation of ECC oystems, (6) Heat removal through e'mergency condensers, (7) Of fset valve setpoints to open fewer valves per 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 if a'-tion valve (MSIV) closure, (10) Lowering the pressure setpoint for MSIV closure, (11) Reducing the testing frequency of the MSIVs, i [\ f )
\d 0-12 Am. No. 57, (5/81)
ACNGS-PSAR 3 (12) More stringent valve leakage criteria, and (13) Early removal of leaking valves. GE 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 the relief valves should be reduced substantially (by an order of magnitude).
RESPONSE
- a. Nature of Study i 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 i modifications which could reduce spurious SRV blowdowns and to modifications which could reduce the probability of SRVs to stick open when challenged. 57
- b. Conduct of Study Although the study was precipitated by the consideration of reducing challenges to the Safety / Relief Valves (SRVs), it was recognized that the 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 the 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 the frequency of uncontrolled blowdowns by a factor of ten relative to the BWR/4 case, which was used as a base for evaluation.
The details of this study are provided in Reference 1. For the BWR/6 l plants such as ACNGS it was concluded that no changes are required to achieve a factor of ten reduction (relative to operating experience) because:
- 1. design features which reduce SRV challenges are already incorporated.
- 2. the two-stage Crosby valves to be used are less likely to stick open due to design dif ferences f rom the three stage Target Rock valves on which the operating experience is based.
l' l
- l 0-13 Am No. 57 (5/81)
ACNCS-PSAR 8 7
- c. Completion Date
( 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.
Re fe rences 57 1. Letter from D.B. Waters (BWROG) to D.G. Eisenhut (NRC) dated March 31, 1981 and titled "BWR Owners Group Evaluation of NUREG 0737 Requirements." l l v .l l i !CJ 0-14 Am. No. 57, (5/81) i
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i i
~ACNCS-PSAR r'
pc \ ITEM II.K.3.18 MODIFICATION OF ADS LOGIC - FEASIBILITY STUDY AND MODIFICATION ( FOR INCREASED DIVERSITY FOR SOME EVENT SEQUENCES
\-s NUREG 0718 REQUIREMENT
-l Applicants with BWR plant shall address the requirements set forth in Item A.7 of NUREG-0626. Applicants shall provide sufficient information to describe the nature lof ' 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.7 The ADS actuation logic should be modified to eliminate the need for manual i actuation to assure adequate core cooling. A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme which should be considered is ADS actuation on low reactor vessel water level provided no HPCI or HPCS system flow exists and a low pressure ECC system is running. This logic would complement, not re plac e , the existing ADS actuation logic. RES PONSE
- a. Nature of Studg A feasibility and risk assessment study was made to examine possible
['~'\ modifications to the ADS initiation logic, which would eliminate the need ( ) ,, for manual initiation to assure adequate core cooling. For some non-line 57 break events which are further degraded by assumin; 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 drywell pressure 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. l Each of the options is evaluated on the basis of whether it assures adequate core 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 is assumed that all high pressure 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. i The details of this study are provided in Reference 1.
\_-) 1 0-15 Am No. 57, (5/81) t
ACNGS-PSAR
- c. Completion Date The study is complete and was transmitted to the NRC by Reference 1.
d.. Program for Implementation of Results i l The BWROG concluded 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 resolution of the ites. 57 REFERENCES
- 1. Letter from D.B. Waters (BWROC) to D.G. Eisenhut (NRC) dated March 31, 1981 and titled "BWR Owners Group Evaluation of NUREG 0737 Requirements."
I 1 4, i f i
-16 Am. No. 57, (5/81)
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)
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,f-~ ITEM II.K.3.21 RESTART OF CORE SPRAY AND LPCI SYSTEMS ON LOW LEVEL - DESIGN
{ j AND MODIFICATION v NUREC 0718 REQUIREMENT Applicants with BWR planto 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 affects 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. RES PONSE 57
- a. Nature of Study In this item, the NRC Suggested Certain Modifications to the Core Spray
/'~ (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 terminate 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. Intuiti vely , it might appear that additional ECCS automation would be purely beneficial since this would supposedly provide added protection against operator errors and omissions. However, these perceived benefits of extended system automation must be measured against the very real ; penalties of increased system complexity, tiduced system reliability and l restricted operator flexibility for dealiag with unanticipated events. These considerations are not amenable to precise quantification and control cystem design decisions must of necessity involve judgements as to relative importance of these competing influences. I
-'s (J
0-17 Am. No 57, (5/81)
I ACNGS-PSAR (~' '
- b. Conduct of Study 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, and 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.
- d. Program for Implementation of Results The study concluded that while changes to the LPCI/LPCS logic would not have a net positive safety ef fect, modifications to the HPCS logic to assure a restart on low reactor water level would. This can be readily incorporated into the ACNGS design and will upon'NRC/BWROG resolution of (s_ this item.
REFERENCES
- 1. Letter f rom D.B. Waters (BWR Owners' Group) to D.C. Eisenhut (NRC) dated 4
December 29, 1980 and titled "BWR Owners' Group Evaluation of NUREG-0737 Requirements". 4 e ! A l 0-18 Am. No. 57, (5/81)
ACNGS-PSAR ITEM II.K.3.24 CONFIRM ADEQUACY OF SPACE COOLING FOR HPCI AND RCIC SYSTEMS i I \ V NUREG 0718 REQUIREMENT Applicants with BWR plants shall address the HPCI' and RCIC systems requirements set forth in Item B.3 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. I j NUREG 0626 ITEM B .3 ; 1 Long-term operation of the RCIC and HPCI systems may require space cooling te maintain the pump room temperatures within allowable limits. The licensees 57 should verify for each plant the acceptability of the consequences of a complete loss of alternating current power. The RCIC and HPCI systems should be designed to withstand a complete loss of alternating current power to their support systems, including coolers, for 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 offsite power. The cooling system is sized and designed to maintain a suitable environment for HPCS and RCIC components following a loss of offsite power. C
- The BWR Owners Group receives clarification that the intent of this item was to assume loss of offsite power only. This was confirmed by letter, D. B. Waters (BWROG) to D. G. Eisenhut (NRC) " clarification of NUREG 0737 57 Items II.K.3.24 and II.K.3.25,) dated 1/23/81.
iO 0-19 Am. No. 57, (5/81)
ACNGS-PSAR
}.ITEMII.K.3.28 VERIFY QUALIFICATION OF ACCUMULATORS ON ADS VALVES NUREG 0118 REQUIREMENT
" Applicants with BWR plants shall provide information t o assure that the ADS valves, accumulators, and associated equipment and instrument at ion will be capable of performing their intended functions during and following an accident situation while taking no credit for nonsafety related equipment or instrumentation. Air (or nitrogen) leakage through valves must be account ed
- for to assure that enough inventory of compressed air (or nitrogen) will be available to cycle the ADS valves. Applicants shall commit that these requirement s will be met in the final design at the OL stage.
In addressing this item prior to CP issuance, applicant s should not e that ' safety analysis reports claim that air (or nitrogen) accumulators for the ADS valves provide sufficient capacity (inventory) to cycle these valves open five t imes at design pressures. Also, General Electric has stated that the emergency core cooling systems are designed to withstand a hostile environment and still perform their functions for 100 days following an accident ." 57
RESPONSE
'Ihe present ADS air accumulators are sized to cycle the ADS valves twice against 70% of cont ainment design pressure (or five times against cont ainment atmospheric pressure) plus component leakage for seven days. Post accident p) access to replenish the air supply (assuming t hat the supply compressors are inoperalive) is being confirmed as part of the post accident shielding study (d in response to Item II.B.2. The radiation environmental qualification for the ADS air accumulators and associated components for at least 100 days will be confirmed by this study as well.
IIL&P is participating in the BWR Owners Group effort s to agree with the NRC on a uniform design basis for ADS air accumulator sizing. The results of this ef fort will be adopted for ACNGS and design changes made if necessary. l m i 0-20 Am. No. 57, (5/81)
ACNCS-PSAR ITEM II.K.3.44 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE FAILURE TO f~N s VERIFY NO SIGNIFICANT FUEL FAILURE k U
)
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 designs. . NUREC 0626 ITEM A.14
. 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) wit.h 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 plants, adequate core cooling is maintained for these worst-case (Vo) conditions. Analyses of further degraded conditions involving a >
# tuck open relief valve in addition to the worst transient and single failure were also performed. The results show that, with proper operator action, the core remains covered and therefore adequate core cooling is achieved,
- b. Conduct of Study 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, the 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.
The BWR/6 analyses are applicable to ACNGS.
- b. The anticipated transients as identified in NRC Regulatory Guide ,
1.70, Revision 3 were considered. N k 1
/
O-21 Am. No. 57, (5/81)
i I ACNCS-PSAR I l 1
- c. The single failure is interpreted as an active failure.
-f \v d. All plant systems and components are assumed to funccion 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 can 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.C. Eisenhut (NRC); dated December 29, 1980 and titled: "BWR Owners' Group Evaluation of
] ,%) NUREG-0737 Requirements". .J 0-22 Am. No. 57, (5/81)
l ACNGS-PSAR l l l ITEM II.K.3.45 EVALUATE DEPRESSURIZATION WITH GTHER THAN FULL ADS
\
(
.d NUREG 0718 REQUIREMENT 4
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 the final designs. NUREG 0626 ITEM A.15 Analyses to support depressurization modes other than full actuation of the ADS (e.g., early blowdown with one or two SRVs) should be provided. Slower depressurization would reduce the possibility of exceeding vessel integrity limits by rapid cooldown.
RESPONSE
- a. Nature of Study This feasibility study addresses NUREG-0626, Item A.15, and provides 57 an evaluation of alternate modes of reactor depressurization than full actuation of the Automatic Depressurization System (ADS). The study includes the BWR/6 product line and therefore the Allens Creek plant.
- b. Conduct of Study 2
Depressurization by full ADS actuation constitutes a depressurization i from about 1050 psig to 180 psig in approximately 3.3 minutes. .Such an event, which is not expected to occur more than once in the life-I 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 NUR2G-0626, ' Item A.15, reduced depressurization rates were analyzed and compared with the full ADS actuation. The 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 , I possible increase in depressurization time by producing an undesirably ] long core uncovered time. The case 1 depressurization gives the results i 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. i I .v 0-23 Am. No. 57, (5/81)
ACNGS-PSAR [' c. Comoletion Date 5, 3 The. study is complete and was transmitted to the NRC in Reference 1.
.d. Program for Implementation of Results The study concluded that there is no benefit to be derived from the use of reduced blowdown rates. Therefore, no changes are required to in- 57 piement the results.
REFERENCES
- 1. Letter from D.B. Waters (WR Owners' Group) to D.G. Eisenhut (NRC);
dated December 29, 1980 and titled: " W R Ownerc' Group Evaluation of NUREG-0737 Requirements". i. t ( j i 1 ! l I l 0-24 Am. No. 57, (5/81)
ACNGS-PSAR NUREG 0718 CATECORY 4
) "A requirement to demonstrate that any additional design, development and 4
implementation necessary to satisfy the requirement (or to satisfy the goals l 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." ,
i
RESPONSE
f Responses to the applicable Category 4 iteam are given herein, including PSAR l
, 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 in-plement the requirement prior to issuance of an operating license. The follewing Category 4 items are not applicable to ACNGS: II.E.1.2 Auxiliary Feedwater System Automatic 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 - applicable to B&W NSSS plants only. II.E.5.2 B&W Reactor Transient Response Task Force - applicable to B&W NSSS 4 plants only. d 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. 4 I 4 I 0-25 Am. No. 57, (5/81)
l ACNGS-PSAR i ITEM 1.A.4.2 LONG-TERM TRAINING SIMULATOR UPGRADE j
/^\
-{ ) NUREG 0718 REQUIREMENT Applicants shall describe their program for providing simulator capability for their plants. In addition, they shall describe how they will assure that their proposed simulator will correctly model their control room. Applicants shall provide sufficient information to permit the NRC staf f to verify that they will have the necessary simulator capability to carry out the actions described in this Action Plan item as well as Action Plan Item II.K.3.54. Applicants shall submit, prior to the issuance of construction permits, a general discussion of how the requirements will be met. Sufficient details shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.
RESPONSE
HL&P intends to use the Black Fox simulator for operations training in support 57 of ACNGS. The simulator will be used for: (1) continuing training of licensed personnel (2) initial attainment of cold Itcenses (3) supplemental training for advancement from RO to SRO (4) training of home office and plant staf f engineering personnel and auxiliary operators. O The control room at both Black Fox and ACNGS are Nucienet 1000 advanced control room designs. The principal plant console and reactor core cooling benchboard for the plants are very similar. During simulator training operators using these panels will learn how to use the CRT displays and their interaction with the associated controls. In 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 operators will gain experience directly applicable to ACNGS through the simulator training. The benchboards utilized for balance of plant control are similar in function although not identical in layout between the plants. From this functional similarity the operators will gain experience useful for ACNGS operations through the simulator training. O t k l 0-26 Am. No. 57, (5/81) i
ACNGS-PSAR Currently it is expected that four utilities with plants having Nucienet 1000
/ control rooms will use the Black Fox simulator for training. It is anticipa- \ - ted that adequate time can be made available to meet the needs of all these utilities.
General Electric, the owner of the Black Fox simulator, currently intends to keep the training program updated with NRC requirements on simulator train- , ing. The simulator also will have the capability to simulate small break l LOCAs and other transients as required. l In summary, the Black Fox Station Simulator is a good representation of the 57 . A11ees Creek Control Room as both are advanced control rooms on a BWR 6 plant. Second, adequate training time is expected to be available for those utilities with these type control rooms, and third, the training program is expected to be kept current with training requirements. i i i r 3 J 0-27 Am. No. 57, (5/81) ,
- - .. - ,-, - - - . ~ . - . - . - _. - . , .- ..
;
ACNGS-PSAR t g .ITEN I.C.9' LONG-TERM PROGRAM PLAN FOR UPGRADING.0F' PROCEDURES
~(
s NUREG 0718 REQUIREMENT-
" Applicants shall' describe their. program plan'which is to begin during p ' construction and follow into operation for integrating and expanding current
. efforts in the area of plant pro'cedures. The scope of the program 'should -
, include emergency procedures, reliability analysis, human factors engineering,-
crisis management'and operator training. Applicants shall also insure that i their program will be coordinated, to the extent ~ possible, with INPO and other industry group ef forts.: Applicants will submit, prior to the issuance of construction permits, a genera' discussion of how the requirements will be , j met. - Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses.n 57 p RESPONSE 1
The plant staff will begin procedure development approximately three years 4 prior to fuel load. In Section 16.6.8, the following plant procedures are
' outlined:
}e Normal Operation Refueling
- Response to Abnormal and Emergency Conditions
! Maintenance Procedures 4
Surveillance { Security i Emergency Plan
- Radiation Control Procedures i.
f Additional procedures describing the proper interretationships of the operations, maintenance, technical and training sections and review, approval, , and revision of procedures will be developed. The Plant Operations Review Committee (PORC, formerly PNSRC as described in I Section '16.6.6.1 of the PSAR) has responsibilities which include review of procedures and procedure changes thereto determined to affect nuclear safety. i The PORC and the Nuclear Safety Review Board (NSRB, formerly CNSRB) will review the Security and Emergency Plans and submit recommended changes to these plans. . Procedures will be written to govern the actions of the PORC and } the NSRB. Minutes will be'kept of all PORC and NSRB meetings. i The PORC will be formed prior to initiating Procedure Development. This committee will exist throughout.the life of the plant and meet on regular i basis as.will be described in the FSAR. Six months prior to fuel load, plant 4 ( procedure development should be complete.
- 0-28 Am. No. 57,- (5/81) t
ACNCS-PSAR The procedure writing / revising program will require four specific evaluations ., fof each ACNGS operating procedure, ' including Emergency Plan Implementing . .\
> Procedures and Emergency Operating Procedures. These four areas to be reviewed on each procedure include: l
- 1. : S 2fety Evaluation - ' determine whether the procedure of activities covered
. by. the procedure constitute an unreviewed safety question or require a Technical . Specification change.-
-2. Fire Protection Review - determines need for Fire Hazard Evaluation.
- 3. Environmental Review determines need for environmental evaluation.
- 4. ALARA Review - determine the need for a' detailed ALARA evaluation.
4 HL&P is implementing an organizational element with responsibility for collecting and disseminating the work of many industry groups 'and other outside-organizations (see response to Item I.C.5). This group will thus be one of the primary sources of information for initiating new procedures and revision of existing procedures. The many studies in the areas of emergency-procedures reliability analysis, human factors engineering, crisis management, and operator training from the industry and NRC will be collected by this group for evaluation and incorporation in.the ACNGS procedures package, as appropriate. Operating pr'ocedures will be written and revised with the assistance of 57 licensed personnel. Improvements to the proceduren may arise from the ACNGS Training Program, . including plant walk through drills. Other improvements may arise' from the ACNGS Plant Staff review of industry studies and industry ef forts on procedures. Major findings regarding reliability analysis will be reviewed by the ACNGS Plant Staff for modification of the ACNGS Training Program and plant procedures as appropriate. Human factors studies
- will be reviewed to determine possible modifications to equipment and/or procedures.
Emergency Procedure improvements will follow closely..the efforts of the BWR Owners Group Emergency Procedures Guidelines. The assistance of INPO and NSAC-will be utilized as it becomes available to verify that new methods are incorporated and that plant procedures and the Emergency Plan Implementing Procedures are consistent with industry standards. Experience gained in implementing operating phase procedures at th9 South Texas Project Electric Generating Station (STPEGS) is expected to be, very valuable to the ACNGS Plant Staff. It 'is anticipated that the STPEGS Operating Staff will be providing advice and assistance to the- ACNGS Staff in development of ACNGS procedures. The Independent Safety Engineering Review Group will perform additional evaluation of all procedures important to the safe operation of ACNGS for technical adequacy and clarity. Through the combination of all of these reviews, HL&P plans to have well developed, thorough, technically adequate, clear, concise, and safe user oriented procedures. j With procedure. development beginning approximately 39 months prior to fuel loading, adequate time for interraction exists between the training program
/, and the startup test program, which are scheduled to begin approximately 36 months -and 15 months . respectively prior to fuel loading.
k , 0-29 Am. No. 57, (5/81) . J
, _ _ _~ _ . _ ._. . _ _ _ _ _ ,
ACNGS-PSAR , b] ITEM 1[D.1 CONTROL' ROOM DESIGN REVIEWS NUREC 0718 REQUIREMENT
" Applicants'shall provide preliminary design information at a level consistent
;with that normally required at the construction permit stage of' review.
Applicants shall provide a general discussion of their approach to control room designs that comply with human factor principles by specifying the design concept selected and the supporting design bases and criteria. Cosmetic revisions to conventional (1960 technology)' designs is unacceptable. Appli-cants shall also demonstrate that the design concept is technically feasible ind 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. . Applicants shall commit to control room designs complying to human factors principles prior to issuance of a CP or ML and shall supply design information for approval prior to committing to fabrication or revision of fabricated control room panels and layouts." 57
RESPONSE
The response to this question is in two parts. The first addresses NUREG 0659 1 Appendix B and describes how the Control Room was laid out using a methodology virtually identical to that document. The second describes the detailed human
' factors review which has been performed on the Control Room design to date in i response' to HL&Ps expectations of the checklists to be included in the forth-
]- .
. coming NUREG 0700.
I. SYSTEMS /0PERATING ANALYSIS TECHNIQUES USED IN CONTROL ROOM LAYOUT ACNGS will'use the General Electric Company NUCLENET 1000 Control Complex. i~ NUCLENET as an advanced, generic design, was developed through a methodology virtually identical to that set out in Appendix B to NUREG-0659. This pro- . cess, similar'to the systems / operations analysis process presented in military ! specifir ation MIL-H-46855, included an analysis of all functions necessary to l operatu the plant safely, an allocation of functions between operator and machine, a qualitative and verification of the functional allocation. i 1 GE assembled a team of experts to design the NUCLENET/*1000 Control Complex , which included experts in: controls and control systems design,' computer 3 technology, industrial design, operator training, power plant test and opera-tions, and behavioral science. ! The premise upon which the design is based is that optimum control is achieved I when there is an allocation of control functions between the operator and ! machine which recognize that each performs certain functions better than the other, and that, once the allocation is made, the design permits efficient and i effective manipulation of controls by the operator. l i- <s O L 1
0-30 Am. No. 57, (5/81) 1
ACNGS-PSAR fg The balance of plact controls were integrated into the NUCLENET complex using
-(\~ f human _ factors principf ea. A survey was then conducted of a full-scale mockup of the control room' with' an interdisciplinary team, human engineering dis-
~
crepancies' were identi.fied, and enhancement active are underway.
- 1. ~ Functional Analysis
.1.1 Definition of Objectives An essential step in the methodology was to define and evaluate the functions and actions to be performed in meeting system objectives.
In determining these functions, it war essential to first identify
.the many activities necessary to operate the plant safely, under normal conditions to produce electricity, those essential for safe operation during transients, and those necessary for safe shutdown 57 during accident conditions.
1.2 Definition of Function Once having identified these activities, the next step was to combine activities under functional groupings, chosen in such a manner that they would be both understandable to the operator and allocated and distributed in such a way to permit effective operator action. For manual operation the activities were grouped under the following functions: ip k 1. Prov'ide and maintain normal core coolant
- 2. Control reactivity
- 3. Monitor performance of the core
- 4. Control reactor pressure
- 5. Utilize steam for power conversion
- 6. Convert mechanical power to electrical power This functional grouping is very similar to those listed in Appendix to NUREG-0659 (p. B-15) which are:
- 1. Nuclear reactor reactivity control
- 2. Reactor core cooling
- 3. Reactor coolant systems integrity ,
- 4. Primary reactor containment integrity
.U i 0-31 Am. No. 57, (5/81)
'F l ACNGS-PSAR.
/ *
- 5. ' Radioactive ef fluent control v 6. . Power generation 7.- Power transmission t.
More detailed identification of plant' system control functions.has been.made by considering operational sit'uations and events that will or may. confront operators in the Control Complex. The operational situations and events. considered, listed in terms of priority, con-sist.of: 57 1
- 1. All events required to be assessed by Section' 15, " Accident Analyses," of th? ACNGS PSAR
, :2. Normal operation of the plant
- 3. Failures in systems, subsystems, and components, and human errors
- 4. Anticipated operational occurrences, including startup and shut- l
; -down of the plant.
f
- 5. (Task Action Plan I.C.1, NUREG-0660 and NUREG-0737).
{ 1.3 Decision /Information Requirements [ i For each 'significant activity within a functional group, a display j was developed which measured each activity against criteria which
- indicated the type of information essential for making decisions I
-egarding the man-machine interface. This is illustrated in Figure 1.
J Human factor engineering design objectives were also developed to i-reflect the goals to be achieved in a new Control Corolex design. These objectives are: i ) 1. Provide a more efficient, coordinated control of the BWR than that attained with a conventional Control Rcom. l- 7. Integrate planned operation functions for steam supply and power conversion ' systems into a single operator station.
j 3. Improve operator response time and reduce operator errors by , determining the optimum quantity of data and number of display i devices which the operator must continuously survey, analyse and comprehend.
- 4. Improve operator performance by determining how best to central-ize and integrate an optimum number of control devices which the l operator must manipulate.
1 4 5
%W l
1 0-32 i 1 Am. No. 57, (5/81) l
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, , , , , , y., 7,r e p ..ww .r w----,----w-
ACNGS-PSAR
}- 5. Incorporate-efficient hardware and software display techniques in V order to present timely, useful information which is meaningful to the operator.
- 6. Provide for factory testing and evaluation of the entire Control Complex.
1.3.1 Design Criteria The following design criteria were developed to achieve the design objectives:
- 1. General
- a. Functions in the Control Complex shall be assigned to three typei of panels:
(1) Primary Operator interface panels I (2) Secondary Operator interface panels ! (3) 'oack row panels
- b. Major power generation systems shall be integrated for I planned operations to centralize and minimze the pri- .t 57 p mary Operator interface by: _
\ !
L 4 (1) Separating the Operator's short response functions , from the long response functions . l. (2) Making frequently used functions and normal re- l activity controls readily accessible from or at t the operator's normal duty station (3) Providing a Display Control System for bringing operational data to the Operator. '
- c. The planned operating functions of the Core Standby Cooling Systems shall remain integral with the appro-
- priate cooling system, and their direct support systems in order that the design of the benchboard used for <
Operator interface with these engineered safety fea-tures shall not affect licensability of the Control Complex. The other safety systems (e.g, safety related HVAC) are grouped together so that system level actua-tion and operating status is available on a front row panel (5800), with component level control and all system parameters available on back row panels. l . 0-33 Am. No. 57, (5/81)
, ACNGS-PSAR . [] d. Integration of the Nuclear Steam Supply and Balance Of (y Plant functions sh'all not degrade the capability for-power generation.
1.4 Functional Integration and Interactioria 1he relationships and interactions between control functicca have heen. defined and evaluated to ensure that all plant. operations and safety objectives can be achieved. These relationships and inter- , actions provide a basis for the development of. Control Complex design l requirements, and, can serve for future design modifications if neces- ' sary. 1.4.1 Human Factors Application I
- a. The arrangement of panels shall ensure tnat each panel defined as primary operator interface will have control, '
display, and annunciator areas visible to an Operator from 57 l' his normal duty station. J'
- b. The distance from the operator's normal duty station to'the most remotely located function on a primary operator inter-
- face panel shall not exceed 50 walking-line feet,
- c. Normal operations functions shall be placed within the -
reach span'of a single operator with'ut compromising the g integrity of those systems having a 4tifunctional capabi-lity. i { d. Align each system's information devices and controls verti-j cally, with information devices above controls. i { e. Align system's operations horizontally, or vertically in 4 the order of the flow path, i
- f. Arrange control functions in an array which is meaningful t to the operator. Provide mimic of complex control systems j representing.the operator's mental model of the systems
;
process flow. i
- g. Maintain system functional integrity in the human-machine interface to aid operator's comprehension of process be-havior.
i h.-Use miniature devices for controls without sacrificing safety or reliability. I
0-34 Am. No. 57, (5/81)
- .
s .
. . . _ _ _ . __ ,_ _ _ _ , . . - . _ , - . . _ . . . _ _ . . . _ , , _ , _ . . _ . _ _ , _ . _ . _ . , . . ...,, _ ., _ ,. 4 _ .._ , _ ,,
. _ . . -~ . . - - . - - . . ,- .- .. . . _ . . _ . . . . .,
ACNCS-PSAR [~
\
i'. Provide a Display Control-System which presents normal
- operations information in pre-defined formats, determined by operational analyses, as well as presenting Alarm Initi-O ated Displays (AID). Incorporate: Color and Shape Coding,
- j. Display by exception, where too much information is not meaningful to the Operator.and could cause sensory overload.
- k. Provide means for reactor control power and~ safe shutdown
-in the event of catastrophic failure of the Display Control System, yet maximise its availability) 1 99.5%.
j 2. Allocation of Functions 57 t i i A systems analysis was conducted to determine which systems vital to i operation of the plant could be controlled from the single operator ' eta- l tion designed to be the primary operator interface nith the control of the j plant. It was determined that these systems were: (1)- REACTOR WATER CLEANUP SYSTEM
- (2) CONDENSATE PUMPING SYSTEM ,
(3) FEEDWATER PUMPING AND REACTOR LEVEL CONTROL' SYSTEM . ; >
/' (4) REACTOR RECIRCULATION SYSTEM f -
j (5) ROD CONTROL AND INFORMATION SYSTEM i t
; (6) NEUTRON MONITORING SYSTEM ,
j (7) STEAM BYPASS AND PRESSURE REGULATOR SYSTEM ' i i j (8) MAIN TURBINE CONTROL SYSTEM I (9) GENERATOR CONTROL SYSTEM
Human factors engineering principles and criteria were used to evaluate j human-machine interfaces in analyzing performance requirements for plant control functions and for the allocation of functions to categorize these , nine systems. Allocation categories consisted of: A l l (1) Automatic operation by plant systems equipment l (2) Manual operation by control-room Operators and/or plant Technicians j (3) Some combination of (1) and (2) I
I I i 0 . Am. ' No. 57, (5/81)
ACNGS-PSAR
\
' The design evaluation allocation criteria considered the capabilities and
- 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 FUREG-0659 were used in
- making the allocation of functions.
2.1 Operator / Technician Processing Capabilities ; i Plausible human roles of Operators, Technicians, and Supervisors (e.g., control manipulator, instrument monitor, supervisor, decision maker, communicator, equipment repairer, coordinator) have been , defined. Qualitative information processing capability in terms of 57 load, accuracy, rate, and time delay have been prepared for each , Operator / Supervisor ~information processing function, i 4
2.2 System Processing Capabilities
- Plausible system roles of Control Complex equipment (automatic con-
! trol of reactor flux, reactor trip system, engineered safety feature) i have been defined. Information processing capabilities and control ' ! function response times of control systems equipment have been de-
- fined considering load, accuracy, rate, and time delay for processing l and response.
- 1 i s 2.3 -Responsibility for Plant Safety e '
i l- The overa11' responsibility for the top-level assessment of plant l operating and safety status has been allocated to the human Oper-ator(s). The rationale for this allocation is based on the cognitive l abilities of humans, which cannot be duplicated by a machine. The j information requirements to exercise this responsibility determine 4 methods for transfer of plant systems data and information to the operator (s) in the Control Complex. i 2.4 Results of Allocation
- A summary description of the allocation of functions follows
- ,
l Reactor Water Cleanup System - This system is operated manually. i This is an instance where the operator or machine can perform approx- ) 'imately equally well, and the system objective is achieved by manual i operation. This operation does not overload the Operator, and it was not cost-beneficial to automate. i b ! (- s 5 0-36 Am. No. 57, (5/81)
- , .-,..m ..__m. . , _ - , , = , _ . _ . , . , . . , , , , , v. __ _ . - ~ . _ . - , _ . , . ._ ..~..-....,em ,-r -, , . . . . , ..m....m. ,,.& _ . um,
l
'ACNGS-PSAR .f Condensate Pumping System - This system is primarily manual. The T operator must reach a decision on when and how much water to pump.
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-tiona. 'This is an activity in which the Operator is superior to the machine. Once the Operator takes the manual action, other actions in the system are carried out autosatically, such as maintaining the hot I well water level. The automatic operation is best suited to the j machine. Feedwater Pumping and R'eactor I.evel Control System - This system ' operates automatically during all modes of operation since its func-1' tion is to monitor and perform the routine task of maintaining proper j reactor water level. 57 j Reactor Recirculation System - This system can be operated semi-auto-matically or manually, and is another example of approximately equal
capability between the operator and the machine to perform a task. *
- Muual operation, when used, does not overload the operator. '
1 Rod Control and Information System - The Operator manually initiates l, the action for operation of this system, based on a judgment of when j it should be operated. This judgment is reached after considering a ,
wide variety'of information, a task in which the operator excels.
Once control action is initiated in operation, the system functions
~ to automatically by the operator ensure rods do not exceed established limits while being withdrawn. This automatic function is ideal for [ the machine. The portion of the RCIS which controls Control Rod seq-l uences and patterns -during Startup, Shutdown and power operation name- ! ly the Rod Pattern Control System is not initiated by the Operator. , This system is a hard wired scheme for which the Operator has limited j ' bypassing ability for a limited number of control rods. i j Neutron Monitoring System - The Operator must insert nd withdraw l IRM's and SRM's during startup and shutdown. Also during these phases of operation the Operator must change IRM ranges. Once the limits within which this system must operate are established, the system i I performs its monitoring functions automatically. This is a monitor-ing 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, 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. J 0-37 Am. No. 57, (5/81) m _.-_. , _ _ - _ __ . _ . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . - _ , .
ACNGS-PSAR (] Cenerator Control Systeu - Synchronizing of the unit could be operat-T' " " '/ 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. 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 residual heat removal system and most other safety systems. Since NRC requirements dictated that these systems operate automatically, an allocation of functions was not performed for these systems. 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 57 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 and Generator Lube Oil, Steam Supply and Drains, Circulating Water,
,s Condenser Of f-Gas and Condensate and Feedwater Auxiliary Systems, are
( ) long-response systems. They are generally manually initiated during
\m / startup then operate automatically.
- 3. Verification of Functional Allocation The verification of functional allocation is a detailed assessment and analysis of each allocation to ensute that the correct functional alloca-1 tion has been made. The verification of functional allocation defines the design requirements and specifications for the systems required by the Control Complex as well as the specifications for quantity of operators, for the interface between operators and a system, for the operational procedures (including emergency procedures), and tar maintenance require-ments.
3.1 Verificaticn of Functions Allocated to Machines For each system function allocated to a machine, the performance requirements of the system, or equipment to execute the function, have been defined. The performance requirement considers such char-acteristics as response time, accuracy, reliability, and operator interface or display requirements. Points regarding the design of Control Complex systems are: Q^ 0-38 Am. No. 57, (5/81) 1 1
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~1._ Display Systems The design requirements for display systems consider established design criteria._ Futhermore, the design requirements for display systema contain criteria to display signals that directly and accurately reflect the information 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 performing their design functions or are responding to operator commands.
l t
- 2. Control Systems l
The design of control systema considers the design criteria presented in Appendix A of 10CFR50: General Design Criteria 57 l 20-39. Utilizing this analysis data and the design criteria ! previously described, primary and secondary Operator Interface i panels were defincJ as:
- a. Primary Operator Interface i .NUCLENET Control Console (P680) ii Standby Information Panel '(P678) iii Reactor Core Cooling Benchboard (P601) iv Standby Diesel' Panel (P877) i v BOP' Control Benchboard (P870) vi BOP Auxiliary Control Benchboard (P800)
- b. Secondary Operator Interface All other panels on which are located controls or displys which may be overtly employed by the operator, as opposed to :
the maintainer. The design criteria were then applied to the Operator Interface panels. Function Placement
- 1. NUCLENET Control Console (P680) a.' Normal (after prestart) plant operacions functions i
\
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- b. Short response functions
- c. Frequently used and/or reactivity controls
- d. hvactor Protections System Operator Interface Note 1: Only non-divisional systems related to a, b and c above, except Nuclear Steam Supply Shutoff Manual Initiation at System level..
Note 2: Exclude functions not related to above.
- 2. Standby Information Panel (P678)
- a. Support Information of the Display Control System 57
- b. No Proces- Control
, 3. Reactor Core Cooling Benchboard (P601) i
- a. NSS Safety systems
- b. NSS Long response functions
- I c. Standard design with no licensing impact
- d. Maintained divisional integrity i
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'3. BOP Control Benchboard (P870) k a. BOP Long Response functions '
- b. . Non-frequent use functions
- c. Maintained. divisional. integrity
- 4. Standby Diesel - Generator Panel (P877)
I
- a. Safety related diesel generators
- b. Support systems for a.
- c. Maintain divisional integrity
- 5. BOP Auxiliary Control Panel (P800) 57 ;
- a. All other safety systems !
- b. BOP long response functions
, c. ' System level monitoring and control of a. and b.
} .
- d. Maintain divisional integrity 1
Table I.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 of functions to the Operator interface panels. Table I.D.1-2 shows the panel assignment . conclusions for the BWR process systems which perform 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 ordec of placement of Systems, upon that panel was deduced. s i 3.2 Verification of Functions Allocated to Humans The most critical portion of the analysis is the verification of
- functions allocated to humans. Detailed analysis of functions ass-l- igned to humans has determined 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 from the analysis of functions allocated to humans .should determine requirements for
- 1. Operator training
- 2. Operator procedures
- 3. Optimal Control Complex human-machine interface and centrol room configuration l
i
- 4. Control Complex staffing.
I i b G l l 0-41 Am. No. 57, (5/81) _ _ _ . .. _._._ _ ____ . _ _ _ _ _ _l
ACNGS-PSAR Initial Work Station Layout (Ol
- 1. NUCLENET* Control Console (P680):
"The 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-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, after being scrubbed a .3 dried, carried off to directly drive the 57 Turbine-Generator. (In the Western World, there exists a population stereotype for , reading: From the Icft to the right; and, from top to bottom. Hence, there is the expectance that logical sequences are arranged in a similar manner). With the reactor core at the center of the Console, as the point of p re ference , if water comes in, and steam goes out, there exists the i lef t to right expectancy of: water into the core; water and steam in the core; and, steam out of the corel Therefore, the water system groups were placed on the left side of the Console, and the steam system groups on the right. The Reactor Water Recirculation System controls 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).
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 (CHC) System controls steam utilization by the Turbine. It was placed next l to the Steam Bypass and Pressure Regulator System. The Generator is directly coupled to the Turbine, and was therefore placed next to the i Turbine. l bJ 0-42 Am. No. 57, (5/81)
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-['~' There are two more systems which were placed on the Console. One of l .\ which had been included in the previous analysis. The Performance Monitoring System is an operational aid which provides the capabil- ,
4 ities of:
- s. NSS performance calculations, Sequence of Events, Status alarm, and, Post-incident data recall
- b. BOP performance calculations and logs 5 c. Displays of NSS performance calculations results ,
27
- d. Means of displaying operations information to supervisory per-sonnel 1 .
- e. Means of generating new display formats, in the field, for both
! computer systems.
3 The Performance Monitoring System's Operator Interface was placed on the far 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 ' i 1.b.3, (Section 1.3.1) information displays which bring operations data to the Operator. 1 r There were ten color CRTs placed on the -Console, one to be associated l
;
j 'g ' with each of the System groups, and one to be used primarily by the Performance Monitoring System, with switching capability to the DCS. The Operator's controls for the DCS were located on the Console, in a manner to be described later.
- 2. Reactor Core Cooling Benchboard (P601):
i Order of system placement on P601 was based on the sequence and frequency of operation, as well as the relationship of a particu-- l 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 l 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, 1 during reactor startup, by providing a controlled drain. When the operator starts up the CRD Hydraulic System after an outage, he must i l I ! 0-43 Am. No. 57, (5/81)
l ACNGS-PSAR A control reactor water level through the RWCU. This functional rela-(% tionship establishes the need for the CRDHS and the RWCU to be in close proximity to each other, even though on two separate panels. Hence, 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 1 proximity to the lef t side of the Console. Panel arrangement and Key plan are both anchored by this relationship. I 1 In a nuclear plant, the integrity of the Nuclear Steam Supply is of vital importance. Leakage from both controlled and uncontrolled ' sources must be monitored to verify the degree of that integrity. Cantrolled leakage is collected in Equipment Drain Sump (s) before being pumped to the Clean (low conductivity) Radwaste. The frequency of monitoring and recording the leakage collected and pumped out t 57
;
Radwaste dictates that the information would be as close as possible
! to the Operator. This function is therefore located on P601 next to the CRDHS.
. The next most frequently used functions are those of the Main Steam System: Safety / Relief Valves; Main Steam Line Isolation Valves; and, the Steam Line Drains. These functions are located next to the CRDHS l and Drain Sumps. The Standby Liquid Control System has vety few [ Operator Interface devices, and, in point of fact,- has never been i deliberately operated to inject negative reactivity-into the core. ! The SLC System controls and displays were located next to the Main Steam System. Core standby cooling is functionally allocated to: the Residual Heat i Removal System (RHR); the Low Pressure Core Spray System (LPCS); the i Reactor Core Isolation Cooling System (RCIC); and, the High Pressure Core Spray. These systems were assigned to locations on P601 in that i order (see- Figure 3). I 3. BOP Control Benchboatd (P870): Order of system placement on the P870 was also based on the j sequence and frequency of operation, as well as the relationship
of a particular system to other systems. , During power generation, the in-house electrical loads are usually either totally or partially aupplied from the Generator output. Switching of the power source to these Auxiliary Electric Systems normally occurs immediately after the Turbine-Generator is synchron-I- ized to the Grid and loaded. There is a relationship therefore between the Generator and the Auxiliary Electric System. t g i
- l. 0-44 Am. No. 57, (5/81) l
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ACNGS-PSAR ('] Hence, 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 close proximity to the right side of the Console. As before, panel l arrangement and Key plan are both anchored by this relationship. This logical order of placement of systems was continued to locate the remainder of the systems on P870 in the following order: Turbine Test; Turbine-Generator Auxiliaries; Steam Systems, Condensate / Feedwater; Air Removal System; Off Gas System Circulating Water System. BOP Auxilisry Control Panel Systems located on Panel P800 are automatic with the exception of the control room emergency filtration inlet selection, as explained in the control room habitability Section 6.4. The panel contains indi-cation to monitor if a system is performing 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 Fuel Pool Service Water Fuel Pool Cooling Suppression Pool Cooling Upper Pool Dump to Suppression Pool m HVAC System Level Controls in an arrangement similar to back row HVAC panels P863, P864 V) Diesel Generator Panel P877 The Diesel Generator Panel contains controls for the 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 ccatains 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. Containment Isolation Panel P868, P869 1 l The primary purpose of these panels is two fold: 1) To supply the opera' tor with a display by which one can determine that for a given event the portion of the Containment Isolation system required to operate has indeed operated; 2) To supply a location of control for i the safety related isolation valves of non-safety systems, such as l l l i l J 0-45 Am. No. 57, (5/81)
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;
service water,'in a location readily accessible to the operator kN)_, without violating the requirements of separation between safety ; related components and non-safety related components (Reg. Guide , 1.75). l It can be seen from the above that the back row panels have a support
function to the front row panels and contain displays and controls l for equipment addressed by the operator on a less frequent basis than ! [ -front row controls and displays. Back row Panels ' Combustible Gas Control Panel PS71 and P872 Systems located on are ! for control of hydrogen that may be generated as a result of an accident. The operator will be required to address this panel any- 57
i where from 30 minutes into an accident to several days into an acci-dent. - This is a very long response compared with the other systems. l The systems that make up the Combustible Gas Control System are for i
- control, monitoring and recording of the Hydrogen concentrations 'in j the containment.
j Other Back row Panels i i The safety system back row panels contain controls for individual components with the various BOP safety system. These systems, with i the exception of the Combustible Gas Control System, are automa-l tically initiated and operated. The operator addresses the back row panel for testing components or systems or in the case that the r operator may wish to operate the particular system in an arrangement dif ferent from its normal alignment. The Accident Monitoring Panel and ESF Status Panel contain specifics of the general information , displayed on the front row panels, for example the back row Post- I
' Accident Monitoring Panel displays five localized Suppression Pool j temperature and Bulk Suppression Pool temperature and the front row i Panel displays the same Bulk Suppression Pool temperature.
Front Row /Back Row Interface i ' The interface betwen the front row and back row panels is different for various modes of operation. I a) During system. setup previous to reactor criticality the operator will align service water and service steam systems. ! i ! ~b) Once'the reactor has reached criticality the operator has all the f I controls needed for normal start-up, operation and shutdown on !
- the front row panel group. Testing of the various plant systems.
can be done from the back row panel during this mode of operation. , s l f 0-46 Am. No. 57, (5/81) l
. , , - _ . - - _ - - - - - . _ , _ . __ _...- ~ .- _ _ _ . _ _ _ . _ _ _ _ _ _ _ , _ ~ .
.ACNCS-PSAR 4
'~
.c) In ' the event of an accident as assessed by." Accident Analyses" !
[\ Chapter 15 of the ACNGS PSAR or the sequence of failure events for transier.ts and: accidents analyzed to develop upgraded emer-gency procedures no operator initiated control is required for at 4 least the first 20 minutes. The operator need only monitor that the' systems are performing their function. If an j . accident exists where hydrogen may be generated the operator will go to panels P871'and P872 to monitor and initiate systems to control combustible gag control systems. Control of all other BOP safety systems at the system level (Reg Guide 1.63) is on panel P800. The operator may choose to go to the Post Accident Monitoring Panels P880, P885, P892, P896 to verify a display 57 located on one of the front row panels. The operator may choose
! to manipulate controls for safety related HVAC systems which are located on panels P847, P848, P863, and P864. Manipulation of
} component controls is not 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 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-quence of performance. Manual tasks are specific with regard i to actions and information transfers from system to human i 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 perform the plant control functions according to the allocation of func-tions. 1 3.2.2 Operator Task Analysis 4 . 4 1
! 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 requireacnts needed to success-fully perform each task. These analyses serve as the basis for specifying the size of ] the operating crew required, the human performance charac-teristics required for normal and emergency operations, the operational procedures required for abnormal and emergency i 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 i stated, a full-scale mockup of the Primary Operator Interface i panels was constructed. Sheet styrofoam was used to form the 0-47 Am. No. 57, (5/81)
ACNGS-PSAR
/'~'\ panels. The front surfaces representing the control and I
() display areas were covered with a material whose texture is compatible with the use of " velcro" fasteners. Systems analysis had determined, in meeting the systems design objectives, which functions 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. The assumed control and display functional devices, selected for consistency with the design criteria, were photograph-ically reproduced. Small pieces of " velcro" fastener material were adhered to the backs of the devices, to permit their placement (and rearrangement) on the mockup. The system's Operator Interface Devices were placed on the i Console and Benchboards in accordance with the / ~idn cri- 57 teria, and in the same order in which they were selected for location on the panel. The devices were rearranged many times, to provide as nearly as possible, the optimum Operator orientation. 3.2.3 Critical Task Analysis Operational analyses was then performed, by simulating opera-
-s tion of each system, using system Operating Procedures. The System Operating Procedures used were those in effect in a plant having nearly identical system (s) design.
Operational analysis was then performed for integrated plant opera-tion, using the plant procedures. As a result of these analy-ses, device location and arrangement were more nearly opti-1 mized. 1 A task analysis was conducted for those tasks and modes of operation that are likely to have an adverse effect on plant safety if not accomplished in accordance with system require-ments. These tasks are identified as critical tasks. An 4 analysis of crit
- cal tasks was done to identify:
- 1. information required by Operator / Technician, including cues for task initiation
- 2. information available to Operator / Technician
- 3. evaluation process
- 4. decision reached after evaluation O
J 0-48 Am. No. 57, (5/81)
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/~~'N 5. action taken U 6. body movements-required by action t.aken
- 7. workspace envelope required by action taken j
- 8. workspace available
- 9. location and condition of work environment
- 10. frequency and tolerances to action
- 11. time base and time margins (time margins must be adequate ( 57 to cover variances in human responses)
- 12. feedback informing Operator / Technician of the adequacy of the actions taken
- 13. tools and equipment required
! 14. number of personnel required, their speciality, and ex-perience
- 15. job aids or references required
- 16. communication required, in;1uding type of communication
\
- 17. special hazards involved
- 18. Operator interaction where more than one operator is involved
- 19. operational limits of personnel (performance)
- 20. operational limits of machines and systems 1
The critical task analysis also included accident conditions. During the operational analyses, careful notation was made of the Operator's information needs for each phase of system operation. 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 formats. The immediate use of I the data, however, was as a basis for assignment of hard-wired, backup information devices to the Standby Information Panel.
0-49 Am. No. 57, (5/81)
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- 1. Standby Information Panel (P678)
[ \ Until the calculated DCS ruliability (_.995) could be verified l \ ) operationally, it was necessary to provide sufficient hard- ' l wired information displays (as well as a DCS Configuration / Status Display) to allow continued steady state power operations, reasonable power maneuvers in the Run Mode, or a safe shutdown, without reliance on the DCS. The Standby Information Panel serves no other purpose. There are 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 Standby Information Panel stands behind the Coctrol 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 Information 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 information displays within the visual range of a licensed Operator. t n 2. Supervisory Moe toring Console (P679) i The Supervisory Monitoring Console allows supervisory personnel access to the same data available to the Operator, without creating a disturbance for the Operator, by looking over his shoulder. The DCS and the PMS have communications links, therefore, all data in the DCS is available 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 l Operator, for PMS communication. l The Supervisory Monitoring Console is centrally located l between, but at the opposite end of, the Benchboards from the Control Console. This provides supervisory personnel with independent visual access to all of the Primary Operator Interface.
)
v 0-50 Am No. 57, (5/81)
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- 3. Display Control System (DCS)
(T
\ '~'
The total deeign for the DCS required approximately 35 man years of ef fort. Some software enhancements continued for almost 7 years after the desiga initiation. Display format 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 following color coding:
- a. Green - Used only for lines and symbols in process diagrams to represent static system con-ponents, i.e . , pumps , motors , valves , and, piping which are not dynamically presented in the given format. Selected for this association because the display elements make up the larger part of the display, and, a green hue has been demonstrated to be the least visually fatiguing of the available hues.
- b. Cyan - Used as a supporting hue and applied to alphanumeric identification, scales, and i borders.
w
- c. Yellow - Applied to all dynamic prcuess variable display elements, such rs 5ar graphs and 57 digital data. Selected t'or this appli-cation because of the intinsity of its hue. Yellow allows the Optrator to scan the display and easily iden*ify dynamic information.
- 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) normally displayed in yellow, changes to red. Selected because of the traditional, pre-established psy-chological associations (populational stereotype) with such conditions, and because intensity allous minimal visual search.
- e. White - Used as a reference mark on scales, ad-jacent to bar grephs, to indicate process limits, or, to present low confidence data, i !
, 'I 'O 0-51 Am. No. 57, (5/81)
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- f. Magenta - May be used in place of red.
( g. Dark Blue - Shall not be used, due to its visual loss against the normal background color.
- h. Black - Normal background color.
Initial format definition began from 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 performed to determine the relationship of these formats to reactor opera-tions phases.
The DCS design was a continaing process, as stated above. At this point, however, the Operator's controls for retrieval of operational information via the DCS 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 System selection and one for Format selection, thus providing capability of displaying any System Format on any CRT. Two momentary push-buttons would provide a Menu Display and format Change Enabla. It is not necessary that the com-puter system, which drives the displays, attempt to follow 57 Format Selection until the operator has placed the Format Select Switch in the position of the Format desired. The Operator informs the computer that the System and Format selected are those desired for viewing by depressing the format change enable switch. This group of four switches is 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 iihh 4 witch (momentary push-button) for assignment of that CRT to the DCS, when necessary. One of the positions of the Format Select Switch is designated
" Master". 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 left hand, when seated at the center of the Control Con-sole. The informational needs data, derived from the opera-tional analyses, showed what information the Operator needed 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 System 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 data retrieved, and displayed to him.
i k U 0-52 Am. No. 57, (5/81)
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- 3.2.4 Work Station Desian Analysis
(' 4 For each work station in the control room, the time sequence i
; of operator activities and the time required for information
- exchange or transfer to the operator has been defined. The analysis verified that the Operator is capable of completing i
all tasks and that all tasks are capable of being performed 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 transmissions of data ! and information, receipts of information, storage of informa-tion, monitoring of systems and interactions among operational
- crew members, work stations, and systema has been conducted.
The purpose of the analysis was a validation of the Control l Complex capability to successfully complete the intended 4 functions of the design, in both the time and space domain.
. 3.2.6 Workload Analysis 57 i A workload analysis for all critical functions was conducted
! to appraise the extent of the Control Complex room operator ! workloads. The analysis was based on the sequential accumula- { tion of task times. Application of this technique permits an i, evaluation of the capability of the control room operator (s) 4 to perform all assigned tasks in the time required to maintain plant safety. ) The detailed workload analysis divided the Operator's tasks i into categories corresponding to perceptual-motor channels i such as vision, left hand, right hand, feet, cognition, audi-
- tory, and voice channels. The purpose of this level of detail 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. 1 3.2.7 Human-Error Analysis
- A human-error analysis was conducted for each perceptual-I motor channel workload of 75 percent or greater as defined by che results of the workload analysis. !
The purpose of the human-error analysis was to investigate the i probability of error during high workload conditions and to evaluate the consequences resulting from these errors.
l 0-53 Am. -No. 57, (5 /81) !
ACNGS-PSAR 3.2.8 Work S3ation Link Analysis
-' A work station link analysis has been conducted for each work 2-station used by the Operator to perform critical tasks. The analysis defined the frequency and criticality associated with j each of the interactions occurring between Operator and equip- 1 ment and/or between one Operator and another. The defined '
frequency and criticality of the interactions are then used to evaluate the design adequacy of the work station layout in l terms of time and space utilization. This analysis achieves a 57 t near optimal design for 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. l 1
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J t i i . 0-56 Am. No. 57, (5/81)
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ACNGS-PSAR 4 NO f:( TABLK I.D.1-2 l a PANgL ASSIGNMENT CONCLUSIONS J' i , Penel System i 1-NUCLgNET* Control Nuc. Boiler Process Inst r. Console Recire l - Rod Cont rol & Informat ion i FW Level Control
- Neutron Monitoring
- Rx Prot ect ion System j Rx Pressure Cont rol
! Performance Monitoring System Rx Wat er Cleanup (RWCll) ! Condensat e Pumping
'FW Pumping Tiirbine - Generat or i
[ Rx Core Cooling CRD Hydraulic System i Benchboard Standby Liquid Control System (SLC) 1 Residual Heat Removal (RHR) j Low Pressere Core Spray (LPCS) 57 ! High Pressure Core Spray & Power Supply ;
- (HPCS) i Rx Core Isolation Cooling (RCIC)
Pressure Relief valves Main Steam System j FW lleaters, Vents & Drains ! Condenser, Air Removal j Off Gas Syst em . Turbine Auxiliaries & Test i Cire Water Condensate System I 4 Feedwat er System - ' l Radiat ion Monit oring
. Steam Supply & Drain Syst em Auxiliary Electric Distribution l BOP Aux Cont rol Spent Fuel Pool Cooling Benchboard Suppression Pool Ccoling !
Cont rol Room HVAC ! St andby Gas Treatment System
- ECCS Area Filt ered Exhaust
- ECCS Area Fan Coolers
- l Upper Pool Dump to Suppression Pool All ot her Safet y-Relat ed HVAC Systems
, Combust ible Gas Drywell Hydrogen liixing ,[ Control Panet Hydrogen Recombiners
Hydrogen Monitoring Syst em l 0-57 Am. No. 57, (5/81)
ACNGS-PSAR II CONTROL ROOM DETAILED HUMAN FACTORS REVIEW (~N BACKGROUND N Y
"' he Allens Creek Control Room Evaluation Task Team performed a preliminary assess-ment of the Allens Creek Nuclear Generating Station Unit I control room for the pur-pose of identifying additional human factors engineering design conditions that could provide a basis for further improvements.
Houston Lighting & Power Company's instrumentation and controls engineers assigned to Allens Creek built a full size mockup of the front row panels of the Nucienet 1000 Control Complex. A human factors engineering evaluation was performed on the mockup. The control room wss evaluated as is, without regard to planned changes in layout and changes required as a result of E ree Mile Island. We evaluation consisted of the application of human factors engLneering design checklists. We checklists were developed and prepared by the General Electric Boiling Water Reactor (BWR) Control Room Owners Crnup. As defined by the checklists, the purpose of the control room survey is to review and assess the adequacy of the arrangement and identification of important controls and displays, the usefulness of audio and visual alarm systems, plant status infor-mation provided, procedures and training with reapect to limitations of existing instrumentation, information recording and recall capability, the control room lay-out and environment, and other areas of human factors engineering that potentially impact operator effectiveness, h e ultimate objective is to identify essential modifications at the operator-control room interface to minimize the potential for human error. O he control room survey methodology as currently recommended by the BWR Control Room Owners Group consists of four phases. The four phases are: (1) control roca
'Q review utilizing checklists which compares the engineering aspects of the control room with established human factors engineering criteria, (2) operator interviews, 57 (3) LER analyses, and (4) emergency procedures walk-through.
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 Room Environment, G) Maintenance aid Surveillance, and H) Training and Manning.**
**This essentially agrees with the ten major topics which will be included in Draft NUREG- 0700. The majority of the topics in Draf t NUREG-0700 are found in BWR Owners Checklist A and B as there are distinc; subsections within the BWR Owners Group check-lists for the Draf t NUREG-0700 topics.
NUREC BWR Owners Group
- 1) Control Room Wbr* space Panel Layout and Design
- 2) Wrkplace Environment Control Room Environment
- 3) Annunciators and Auditory Signals Annunciators
- 4) Controls Instrumentation and Hardware '
- 5) Visual Displays Instrumentation and Hardware ; )
- 6) Panel Layout Panel Layout and Design n
- 7) Control / Display Integration , Panel Layout and Design, [
Instrumentation and Hardware p) 8) Labels and Location Aids Panel Layout and Design (V 9) Process Compute Computers
- 10) Data Recording and Retrieval Instrumentation and Hardware l
l 0-58 Am. No. 57, (5/81) J l
ACNGS-PSAR Houston Lighting & Power Company's survey team has performed the control room review phase, he checklists that were considered applicable to the Allens Creek j Tmockup were A) Panel Layout and Design, B) Instrumentation and Hardware, and Q C) Annunciators. W e other phases of the survey (2-4) cannot be performed on I the Allens Creek control room at this time because they involve operating history and emergency procedure availability. Each checklist item is presented in the form of a question for consideration by cach survey team member. As each specific question is evaluated, the team member actually doing the evaluation of that question indicates the relative degree of compliance (no compliance, somewhat compliant, mostly compliant, full compliance, not applicable). Following each checklist item is space for the person performing the evaluation to enter comments. For each specific checklist item, these conunents will identify items or components of non-compliance, the scope of review, or any qualifying statement judged to be appropriate to the evaluation. Evaluation Approach he evaluation of the Allens Creek control room mockup was conducted in three phases, described as follows:
- Phase 1 - Documentation Collection his phase included collection and review of control room panel drawings, control and display design conventions, human factors engineering reference ,
material (MIL STD 1472B, NUREG CR-1580) board profiles and dimensions, piping diagrams, control panel instrument lists, control wiring diagrams, and flow diagrams. 57 A t
- I'hase 2 - Data Collection During this phase, checklists containing human factors cri pria were applied.
Rese checklists addressed:
- control room layout l
- panel layout and design !
- annunciators
- controls !
- instrumentation
- displays and mimics !
i
^
- Phase 3 - Analysis and Reporting '
he final chase of the evaluation consisted of (a) review of the data col-lected, (b) identification of the items needing human factors engineering enhancement, (c) prioritization of the items needing human factors engineer- ! ing enhancement, and (d) reporting of the results. he items identified as needing human factors engineering enhancement were consolidated into the following groups:
- annunciators
- labels ,
- panel layout
- displays
- instrumentation
,' ) , v 0-59 Am. No. 57, (5/81) i
l i ACNGS-PSAR 23 itema n;; ding human fcctero cngine ring cnhrncement wera pricriticcd I according to the following:
- Category 1 - corrective action is recommended to minimize the potential
( i for error () - Category II - corrective action should be considered to reduce the potential for error
- Category III - no corrective action is necessary Results -- Identificative of Human Factors Engineerina Discrepancies h e results of the control room survey will be ob.lectively evaluated by Houston Lighting & Power Company to determine what chant,es should be made for minimizing the potential for operator error as a result of human factors engineering con-siderations. He schedule for this evaluation is such that the changes deemed n;cessary will be incorporated prior to fabrication of the control room.
The following is a summary of the general human engineering discrepancies that
cre applicable in some degree to the panels reviewed by the evaluation task team.
- 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 are not consistently positioned.
- Panel Layout 57 (N 1) h ere 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 systems.
. 5) Limits on anthropometric design are exceeded. , j 6) Layouts ati not consistent in operational sequence.
- Displays !
- 1) Indicators are not scaled in process units that relate to system l
operation. ,
- 2) R ere is no normal range or setpoint markings on indicators. Here '
are no markings to show safe or unsafe ranges and expected or unex-pected range of operation.
- Instrumentation
- 1) R ere is a lack of 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 not clearly identifiable.
- 4) Keylock switches require use of keys for normal operation. He key is the switch handle.
- 5) Backup indication is not easily correlated with indicators.
O ' G l l 0-60 Am. No. 57, (5/81)
ACNGS-PSAR 73
- Annunciators (x
- 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 D4screpancies,_ 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 wl11 include label consistency, accuracy, and board placement rela-tive to labeled components.
- 4) Lines of demarcation will be added to functioaal control and display groups.
- 5) Mimics and lines of demarcation will be added to improve flow paths.
- 6) Panel layout will be improved in terms of control / display relationships, 57 operational sequence, functional relationships, and anthropometric standards.
- 7) All switch positions 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.
- 9) A color coding standard is being developed for indicating lights, mimics,
(- and computer system displays.
- 10) The annunciator system will be improved in terms of prioritization using color coded windows 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 Room Evaluation Task Team The Allens Creek Control Room Evaluation Task Team consisted of two licensed oper-ctors, five instrumentation and controls engineers, and a human factors engineering consultant. The team was broken into 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 cut his own checklist, the two people on each group were not required to agree on the assessment of the panel. I Am. No. 57, (5/81) 0-61
ACNGS-PSAR Conclusion The Allens Creek control room uses advanced technology which incorporates human factors engiusering principles in design and operation. A centrol room design review has been conducted based on a full-scale mockup, deficiencies identified, and enhancement activities are being undertaken. 57 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 i be forwarded to NRC upon completion. I O o-61 Am. No. 57, (5/81) l
ACNGS-PSAR
. ITEM NO. 1.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE NUREG 0718 REQUIREMENT -(-I " Applicants shall describe how they intend to meet the staf f criteria con-tained in NUREG-0696 for the plant safety parameter display console. Ap-plicants shall, to the extent possible, provide preliminary design inf-armation at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, applicants shall provide a i general discussion of their approach to meeting the requirements by specifying i
the design concept selected and the Supporting design bases and criteria. Applicants shall also demonstrate that the design ccacept 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." . 57
RESPONSE
A Safety Parameter Display System (SPDS) will be provided, and is described in Section 7.5.1.6. The SPDS 's in conformance with HL&Ps understanding of NUREG-0696 Section 5. The design concept of the SPDS ir. known to be technically feasible and within the state of-the-art. There are no questions or concerns as to the ability to implement the St>DS design prior to OL issuance. m O i Am. No. 57, (5 /81) l' o-63 i -
ACNGS-PSAR. 7s ITEM 1.D.3 SAFETY SYSTEM STATUS MONITORING NUREG 0718 REQUIREMENT
" Applicants shall describe how their design conforms to Regulatory Guide 1.47,
" Bypassed and Inoperable ' Status :Indic.stion for Nuclear Power Plant Safety Systems." Applicants shall, to the extentLpossible, provide preliminary
] design information 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 57 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-TMI design basis for ACNGS, and the design meets the guide.without any exceptions. Details are given in Section 7.2.2.2.2.2, and systems covered by the guide are shown on Table 7.1-2. x_
t l l l O 0-64 Am. No. 57. (5/81) J n - , - , , - , w.,-w--.. < , - - -
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'ACNGS-PSAR
, ITEM II.B.l' REACTOR' COOLANT SYSTEM VENTS
\
NUREG 0718 REQUIREMENT i-
" Applicants shall modify their plant designs as necessary to provide high point reactor coolant' system and reactor vessel head vente 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-I ly 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
- Venting 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 57 in the vessel below the main steam line nozzles could interfera with continued core cooling, so the capability for venting this region is essential. This can be accomplished by opening any one of the -19 safety
- relief valves on the main steam lines (which may be open already depend-ing on the mode of corc shutdown cooling in use). These valves and their operators are safety grade,~ seismically and environmentally qualified for accident conditions, and are powered from the onsite electrical system and operable from the control room. Eight of the valves have a safety related air supply, providing redundant venting capability. 3 2. Above '.he main steam nozzles: The presence of non-condensible gases in the vessel above the main steam line nozzles will not interfere with i continued core-cooling, and as such venting this region of the vessel is not considered to be a safety concern. Even so, there are two means of I venting this space:
- a. Normally open reactor head vent line and valve B21-F005, which dis-i charges te main steam line 1 (which can be vented to the suppression i
pool / containment via any one of three safety relief valves). I
- b. Normally closed reactor head vent line and series valves B21-F001 and B21-F002, which discharges to the drywell high purity drain tank.
1 These valves are safety grade and their operators are Class 1E, seismically and environmentally qualified, but are not powered from i 1 !- 0-65 Am. No. 57, (5/81)
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ACNGS-PSAR the onsite electrical system. 'They are operable from the Main b Control Room. 1 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 amounts of hydrogen in containment is discussed in the response to Item
-II.B.8.3.
The above supplements the PSAR information on capability for RCS venting, and is consistent with preliminary design information normally required at the CP stage of review. There is no new, novel design, and there are no concerns regarding technical feasibility, state-of-the-art or ability to impleme.nt the intended RCS venting design, t l l } j l i I l 0-66 Am. No. 57, (5/81)
1 ACNGS-PSAR ITEM II.B.2 PLANT SHIELDING TO PROVIDE ACCESS TO VITAL AREAS AND PROTECT { j SAFETY EQUIPMENT FOR POST-ACCIDENT OPERATION v NUREG 0718 REQUIREMENT
" Applicants shall (1) perform radiation and shielding design reviews of spaces 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 information 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 al.so demonstrate that the design concept is technically feasible and within the state o f-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: 57
- 1. access and required occupancy is possible to areas where plant
. operators / worker must perform post-accident functions.
- 2. radiation levels for which safety-related equipment is qualified to
(' '/ function post-accident are not exceeded. 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 fission products (except noble gases) are dispersed into the reactor and suppression pool water. Instantaneous release and mixing are assumed.
- 2. The following systems are instantaneously filled with contaminated fluid per 1 above:
Residual Heat Removal (Suppression Pool Cooling, Containment Spray a. and Low Pressure Coolant Injection).
- b. High Pressure Core Spray
- c. Low Pressure Core Spray l
- d. Reactor Core Isolation Cooling 0%
a 1 V 1 0-67 Am. No. 57, (5/81) 1
1
)
I ACNGS-PSAR l l l fS e. MSIV Leakage Control i 1 kj f. Standby Cas' Treatment
- g. Hydrogen Analyzer
- h. Contain.nent Gas Sampling
- i. Post-Accident Liquid Sampling
- j. ECCS Filtered Exhaust The 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 e .uipment dose levels in certain locations, the following options are avai'able:
- 1. move the offading radiation source to a less sensitive location
- 2. move the target equipment or operator control / work station to a location 57 with an acceptable radiation field.
- 3. place additional shielding around the offending radiation source.
- 4. . place local c.hielding around the target equipment or operator control / work station.
O (j 5. purchase equipment designed to withstand the newly specified radiation environment. In selecting the option to be used emphasis will be placed on minimizing building structural modifications, since the buildings potentially affected are mostly designed and are early in the construction sequence. If proolems 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 technical feasibility or state-of-the-art. Since the shielding study will be
. finished in 12/81 and a CP for ACNGS is not expected before then HL&P is assured r. hat any necessary changes can be effectively implemented prior to construction, and certainly well prior to OL issuance.
l l 1 0-68 Am. No. 57, (5/81) '
I 1 l ACNGS-PSAR
% ITEM II.B.3 POST-ACCIDENT SAMPLING .I \
V NUREG 0718 REQUIREMENT
" Applicants shall.(1) review the reactor coolant and containment atmosphere sampling system designs and the radiological spectrum and chemical analysis facility designs, and (2) modify their plant designs as necessary to meet the
. requirements. Applicants shall, to the extent possible, provide preliminary design information 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
The capability for post-accident sampling of reactor coolant and the
; containment atmosphere, along with onsite analysis capability, will be provided. Details are as follows:
f A. Sample Collection i
- 1. Liquid: The capability to collect liquid samples from the reactor 57
[ ( j coolant system 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 placed in operation. The sampling operation under post-accident conditions will not result in a personnel dose of greater than 3 rem to the whole body or 18 3/4 to the extremities to the workers involved in the sample collection and transport operation. The sample can be collected within one hour of the request for a sample. The sample return line will be to the suppression pool.
- 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 atmosphere sampling operation. The sample return line will be to the containment atmosphere.
B. Sample Analysis t Analysis of the post-accident samples collected per A above will be in the Personnel Access Building. 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 l % Am. No. 57, (5/81) 0-69
ACNGS-PSAR counting room and chemical analyses (for chloride and boron) in the 1aborata ry. The chemical sample analysis stations are equippet with fume - [N' ~','} hoods which are exhausted to the outside through HEPA filters. Doses to workers involved in sample analysis will not exceed those specified for thr. sample collection and transportation operation. Time for the sample a alyses will not exceed the following: radiological: two hours boron: two hours chlorides: twenty-four hours Accuracy, range and sensitivity will be adequate to provide pertinent data to the operator in order to describe radiological and chemical 57 status of the reactor coo 'nt system. There are no questions regarding technical feasibility or state of-the-art regarding the post-accident sampling capability, nor are there any concerns as to the ability to implement the design prior to OL issuance.
\
l l l s o-70 Am. No. 57, (5/81)
i ACNGS-PSAR
~
ITEM'II.B.8 RULEMAKING PROCEEDING ON' DEGRADED CORE ACCIDENTS O 'NUREG 0718 REQUIRENENT
" Applicant shall:
- (3)-provide a system for' hydrogen control capable of handling hydrogen gener--
ated by_ the equivalent of a 100% fuel-clad metal water reaction." [ RESPONSE
'HL&P commits to provide a hydrogen control system. Currently a number of dif-
! . ferent methods are being considered throughout the industry and it is _ expected j- that these effo'rts will, in the future, produce valuable data upon which to select an optimum means of hydrogen control. Further, it is expected that the 7 pending rulemaking on degraded cores will determine the necessity for such 'a system. For the purposes of meeting the stated requirement, a post-accident inerting system using'C0f as_ an inerting agent is proposed. However, for the reasons given above the basic design and need for this system will be under: continuing review. 1 l' The post-accident inerting system is to be capable of handling the hydrogen generated by the equivalent of oxidation of 100% of the active fuel cladding. This system assures that containment integrity is not endangered due to the combustion of hydrogen generated by the reaction between the fuel cladding and the reactor coolant. Control of hydrogen combustion will' be accomplished by l the -injection of carbon dioxide into the containment after event initiation, ! .but before significant hydrogen transport to the containment. The carbon di- 57 I oxide- concentration will be suf ficient to render all mixtures of hydrogen and l air inert, and incapable of sustaining combustion. The following criteria will be used to design the Post-Accident Inerting Sys- ;
- tem
l' (a)'The hydrogen from a transient resulting from the reaction of up to 100
- percent of the active fuel cladding with the reactor coolant is assumed to '
start evolving from the suppression pool surface in no less than 45 . minutes from the reactor scram. t 1 j (b).The system will have the capability for providing adequate mixing of the l carbon dioxide with the containment atmosphere to assure the prevention of ! the combustion of hydrogen in the containment atmosphere, f (c) The system will have provisions for both long term sampling of carbon di-( oxide and oxygen concentrations, and the addition of more carbon dioxide as necessary. (d)- Alternating current power will not be required for the system to perform !. its inerting function, i s 0-71 Am. No. 57, . (5/81)
ACNGS-PSAR (e) The system will be single active failure proof, for either intended or inadvertent operation.
}-
~
(f) Inadvertent full inerting, excluding seismic and design basis and other 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 effect the safe shut down of the plant.
(h) The system will be protected from tornado and external missile hazards. (i) The quantity of carbon dioxide to be injected will be limited to the amount required to provide 61 volume' percent carbon dioxide' concentration
. within the drywell and containment for the accident condition, plus the
~ additional allowance for carbon dioxide solubility in water. i (j) The containment isolation valves associated with the system will be clas-sified as intermediate with no system integrity isolation function. 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 57 the PAIS design basis transient.
- The inerting system will be initiated manually. The system will annunciate audibly and visually in the main control room and by special tone alarm in the
( contaiament. A time delay will be provided for containment evacuation, and continued restoration of water level, thus preventing' unnecessary operation. I A second audible and visual annunciation in the main cortrol room plus special
. tone containment alarm will be given to confirm system operation based on the detection of carbon dioxide flow in the main header. The system will be designed so that an inert condition is reached after a 15 minute discharge
- time.
4 The charge of carbon dioxide will be stored in liquid phase in three (3) storage tanks outside the Reactor Containment Building in a new building located at grade elevation. Each tank will contain one-third of the necessary charge. A fourth tank will provide any necessary makeup. This fourth tank will be under separate manual control. Each of the tanks is equipped with 1 fill connections, relief valves, level gauges, and pressure and temperature
. gauges. The building will require electrical ligh .ing, heating and venti-lating. Low temperature storage is maintained by re(.undant refrigeration j compressors and coils for each tank, maintaining the tank pressure between i approximately 290 and 305 psig at 00F, Under emergency conditions when the !
- rechanical refrigeration is not functioning, the unit is self-refrigerating i due to the cooling ef fect of small amounts of vapor released through a bleeder valve. The system can remain in this self-refrigeration mode for at least,24 hours without losing full functional capability, i
h Am. No. 57, (5/81) i o-72
-_, _.._ , ._ _, . . _ ~ _ _ . _ ~ _ . , . _ . , . _ , , - _ . . - _ . .
4 , ACNGS-PSAR Redundant, back seated, pressure ~ actuated discharge valves on.each t ank will feed the carbon dioxide to the-main header. One containment penetration with ' O t redundant isolation valves' on both the inboard and outboard sides (see Figure 1)'will feed the flow to an interior network of' nozzles which assure adequate ! carbon dioxide distribution. The discharge and isolation valves will close on ! low header pressure.after the injection process is complete. Under normal operation ac power is required to maintain refrigeration, the battery charging system, HVAC,' and electrical lighting. Direct current powered main control . panel (s) located _ in the areaLof the tanks will be equipped with selfcontained battery units for valve actuation. No alternating current power is required i for the system to perform its inerting function.
- The system will be designed to deliver approximately 135 tons of carbon dioxide in'less than 15 minutes. After this time, periodic sampling and
- analysis of the containment atmosphere will be performed to. detect any re-
- ~ duction in carbon' dioxide concentration or-any increase in oxygen fraction.
The design for the discharge and' isolation valves assures a high probcbility that no inadvertent dump will occur. The sysiem will require two active ' failures to inadvertently inject into the containment. Procedures for 4 containment . purging and carbon dioxide cleanup following a degraded core event or inadvertent injection will be developed later. 57-2 l A conceptual arrangement of the system is shown in Figure 1. 2 Design Evaluation i] }. Significant- core heating and hydrogen generation in a BWR does not start until the water level falls below bottom of the active fuel. This is due to ccoling l J of the core by steam generated in the lower portion of the vessel. The start time of hydrogen evolution into the containment for transients is at
-least 45 minutes from event initiation, including the hydrogen transport from the core, through the suppression pool and into the containment. The contri-bution of hydrogen ' from radiolysis of water and corrosion of exposed metal surfaces ls insignificant, and does not affect inerting system design.
i A concentration of 61% carben dioxide by volume in air will render ! hydrogen-air mixtures nonflammable regardless of hydrogen concentration.I j This quantity of carbon dioxide plus an allowance for potential for loss of { the inerting agent, will provide an adequate margin to ensure an inert
- atmosphere.
l i l ICoward, H.F. and Jones, G.W. " Limits of Flammability of Gases and Vapors" Bureau of Mines Bulletin #503 (1952), e i } .1 !. O L 0-73 Am. No. 57, (5/81)
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ACNGS-PSAR Mixing is primarily a function of bulk air movement. Very rapid air movement O is caused by the flashing and expansion of the liquid carbon dioxide to a gas k") and by the discharge of this expanding liquid at an elevated velocity. Nozzle design and selection maximizes this air movement. Nozzle placement is based on attaining both global and local air movement within the containment and on protecting enclosed spaces not within the general injection patterns es-tablished. Nozzle placement to assure adequate mixing will be based on ex-tensive experimental data and field experience existing in the fire protection indu s t ry. Nozzle Location and Piping Layout Containment: Based on studies to date a total of 46 nozzles will be located in the containment. To obtain the maximum degree of mixing and recirculation, 20 nozzles (single axial orifice) have been located at approximate elevation 272'-0" (above polar crane) along the outer periphery of the wetwell. Of these 20 nozzles, two groups of five nozzles aach will be positioned to dis-charge downward into the containment avnulus.. In addition, two groups of five nozzles each will be positioned to discharge upward toward the center of the containment dome. These nozzles deliver 82 percent of the total carbon dioxide discharge rate required in the containment. The remaining carbon dioxide discharge required in the containment will be distributed to 20 individual rooms and below six floor levels to provide assurance of local mixing and inerting. The rates into each room were based on their individual volumes; rates for nozzles located below floor levels will 57 be based on the volumes which existed below those floor areas at that ele-vation. Further studies may result in modification of the quantity and/or
;
(/ locations of nozzles. Drywell: A total of 5 nozzles will be located in the drywell. Four nozzles will be positioned near the ceiling to discharge horizontally. One nozzle will be positioned to discharge into the drywell head above the reactor. The distribution and rate of injection of the agent is calculated using equations based on two phase fluid flow. This calculation procedure conforms to the method given by NFPA-12, the National Fire Protection Association standard on carbon dioxide extinguishing systems, and is accepted by Underwriters Laboratories and by the Factory Mutual Insurance Corporation and has been validated in test discharges throughout the fire protection industry. System Initiation (1) A key locked switch is actuated by the operator to open the discharge valves on each of the three (3) main tanks. Parameters which the operator would use to determine when to initiate include reactor water level, ECCS function and reactor pressure. Formal criteria will be defined in the emergency procedures developed for ACNGS.
\ / 4 \
0-74 Am. No. 57, (5/81)
ACNGS-PSAR (2) A timer then actuates on high pressure in the main header, and upon
's reaching zero time, enables the operator to open the isolation valves via .
a second key-locked switch. s
%s)
(3) Af ter discharge, the isolation and discharge valves close on low header pressure. Testing and Inspection Each discharge valve will be pressure tested and operated. Provisions will be provided so that each active component of the Post-Accident Inerting System can be tested onsite periodically. "Fu f f" testing can be performed to pres-sure test and clean cut the piping system and instrumentation. Instrumentation Requirements 4
' Annunciation will occur on system initiation, on detection of flow in the main 57 header, on high pressure or low liquid level in the tank and on high and low temperature in the inerting enclosure. Instrumentation will include discharge valve position, tank pressure, and inerting enclosure temperature. Local instruments monitor level and temperature gauges in the tanks and system battery readiness.
Materials Piping from the tank to the inboard containment valves is ASTM A106, schedule 80, seamless steel pipe. Distribution piping is ASTM A106, schedule 40,
.[ ) seamless steel pipe. Nozzles will be stainless steel: fittings will be \_ / standard weight steel. A complete list of materials by commercial name and
., showing estimated quantities and physical characteristics cannot be completed until the detailed design is established. i 1 J a 0-75 Am. No. 57, (5/81)
P
\
t MAIN TANKS REFRIGERATED r- - - - -q NC r, l IC l NO l l NC I I I I NC wa l l r, IC l NO g 2 l "N C' l I I I N& S i( b - - - - ._ Nm. J "' NC a LOCAL CONTROL PANEL (WITH EMERGENCY BATTERIES) DISCHARGE VALVES REFRIGERATED MAKEUP TANK {~-----] g i JI NC L. _ ___ _ _ _ _ _J { (
i ACNGS - PSAR 7 L CONTAINMENT DRYWELL TO DISTRIBUTION TO DISTRIBUTION PIPING : PIPING 1I NC IL j k N MAIN HEADER b ISOLATION VALVES N Y 1 \ ~% n7~e k 2 VM NC AM. NO. 57, (5/81) HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 [ l GENERAL ARRANGEMENT OF > CARBON DIOXIDE INERTING SYSTEM (CDIS) FIGURE II.B.8.3-1
_- _ _ . -, _ .._. .. - _ m-ACNGS-PSAR
- ITEM II.D.1 TESTING REQUIREMENTS 1
- NUREG 0718 REQUIREMENT. - l Applicants and their agents shall plan and carry out a test program and model development to qualify the reactor coolant system relief and safety valves
- under expected operating conditions for design-basis transients and accidents.
Consideration of anticipated transient without scram (ATWS) conditions shall
.be included' in the test planning. Actual testing under ATWS conditions need - not be carried out until subsequent phases of the test program are developed.
l - Applicants shall submit, prior to the issuance of the construction permits or j manufacturing license, a general explanation of how the testing requirements will be met. Sufficient detail should be presented to provide reasonable as-surance that the requirements will be implemented properly prior to the issu-ance of operating licenses. , Applicants shall (1). demonstrate the applicability of the generic tests con-ducted under II.D.1 to their particular plants and (2) modify their plant de- !' signs as necessary. Applicants shall commit, prior to the issuance of the construction permits or manufacturing license, to comply with these require-ments and shall submit within six months. following the completion of the ge-I neric ' tests or the issuance of construction permits, whichever is later, a de-tailed explanation of how the test results will be incorporated in the plant 57 design. Sufficient detail should be presented to provide reasonable assurance , that the requirements resulting from the test will be implemented properly
- - prior to'the issuance of operating licenses.
RESPONSE
[G) Performance -testing of BWR safety / relief valves will be done beyond the cur - rent qualification requirements. This testing will be sponsored by the utili-ties of the BWR Owners' Group, in response to NUREG-0578 Require- ! ment 2.1.2. 1 I In July,1979, the NRC issued its TMI short-term Lessons Learned Report
. (NUREG-0578). In this report, the NRC required that testing be conducted "ra qualify the reactor coolant system relief and safety valves under expected i operating conditions for design basis transients and accidents". The "ex-
. pected operating conditions" were to be determined through the use of analyses- , of accidents and anticipated operationni occurrences referenced in Regulatory Guide 1.70 Revision 2. The discussion accompanying the requirement gave pri-mary emphasis to two phase and liquid flow conditions. Reference 1 presents an evaluation of those Regulatory Guide 1.70 Revision 2 events which have the potential for producing liquid or two phase flow dis-charge from the safety relief valves. This report is applicable to most t
- O
! 0-77 Am. No. 57, (5/81) i 1 f
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1 ACNGS-PSAR t BWRs. It is specifically applicable to all plants that have level 8 trips (high water level) on high pressure inventory maintenance systems (e.g. Feed-I water, High Pressure Coolant Injection and Reactor Core Isolation Cooling). I Allens Creek is included within this group of plants. l The conclusion reached after a detailed review of all identified events (see Table 2-1 of Reference 1) is that a test which simulates the alternate shut-down cooling mode should be performed. This event is an anticipated operating condition which has been considered in the design analysis of plants. The BWR Owners' Group has committed to perform liquid and two phase flow safety valve tests for the conditions which can occur for this mode of operation (Reference
-3). All other events which were identified are either of sufficiently low probability or low consequence such that no additional testing is warranted.
A description is given in Referenca 2 for those tests which will be run on typical S/RV's for BWR/2 through BWR/6 plants to demonstrate ability to per-form satisfactorily under the condition in which low pressure (i.e., up to 250
,+ 20 psig) water passes through the valve instead of saturated steam. This corresponds to conditions expected during the Alternate Shutdown Cooling Mode, 57 i.e., the mode in which low pressure pumps are injecting cold water into the reactor vessel and this water is vented through the S/RV's back to the sup-pression pool.
The low pressure water test will serve the following two purposes: A. To demonstrate the capability of each type of S/RV to operate satis-
,p factorily under the bounding cases of release of low pressure water with resultant, typical BWR pipe loads on the S/RV.
B. To measure the S/RV discharge line (S/RVDL) loads during water discharge through S/RV's. [ Six different S/RV's will be tested in the relief mode (normal operating mode for low pressure).
, The specimens to be tested consist of 6 x 8 (inlet dia. x outlet dia. in inches) oilot operated Electromatic Relief Valve, 6 x 10 twoand three-stage pilot-operated Target Rock S/RV's, 6 x 10 and 8 x 10 Crosby direct-acting S/RV's and 8 x 10 Dikkers direct-acting S/RV.
ACNGS will use the Crosby 8 x 10 direct-acting S/RV's, and is thus covered by the testing program. In addition, the data gathered on discharge piping re-sponse will be considered in the design of the ACNGS SRV discharge piping, which will be designed for the same two phase and solid water flow conditions for which the valves are being tested. ,G I ! 0-78 Am. No. 57, (5/81) i -_ _ . _ - . . _ _ .
4 ACNGS-PSAR
- O REFERENCES V 1. " Event Evaluation for BWR Safety Relief Valve Testing Required by NUREG-0578, 2.1.2" enclosed with letter from D.B. Waters (BWR Owners' ,
Group to R.H. Vollmer (NRC) dated September 17, 1980 and titled "NUREG Re- ) quirement 2.1.2 - Performance Testing of BWR and PWR Relief and Safety Valves".
- 2. "NUREG-0578 BWR Safety / Relief Valve Test Description" enclosed with letter 57 from D.B. Waters (BWR Owners' Group) to R.H. Vollmer (NRC) dated September 17, 1980 and titled "NUREG-0578 Requirement 2.1.2 - Performance Testing of BWR'and PWR Relief and Safety Valves".
- 3. Letter, T.D. Keenan (BWR Owners' Group) to D.G. Eisenhut (NRC) dated December 14, 1979 and titled "BWR Owner's Group Implementation of NUREG-0578 Requirement 2.1.2".
l t a T l O 1 0-79 Am. No. 57, (5/81)
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'ACNGS-PSAR -t \ ITEM II.D 1 ' RELIEF AND' SAFETY POSITION INDICATION NUREG 0718 REQUIREMENT Applicants shall modify their plant designs as necessary to provide direct 4
indication of relief and safety valve position in the control room.
~
Applicants shall, to the extent possible, provide preliminary design
~
information at a level consistent eith that normally required at the 4 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 desi Fn concept selected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is technically feasible and within the ctate-of-the-art, and - th at there exists reasonable assurance that the requirement; will be implemented properly prior to issuance of operating licensen.
RESPONSE
' As. shown in PSAR See:lon 7.5.1.4.2.L 3 c, p.4/ety relief valve position indication will be de?.egrined by pr6t=. .n meccurement in the discharge pipe.
l This has been verif rk4 by the d.l ROC. t ; oe g* scequate indication of SRV 57 1 position indication t$ studying dat9 trna rp.sa ing plants, which was . submitted to the NRO by letter, 7. !%gpan (t:N Owners Group) to D.G. Eisenhut (NRC) dated October 17, 1979. l The actual pressury serpoint to be used at ACNGS will be determined from a
'-- combination of analysis tes f(cid test data, and will be submitted with th a
-FSAR. Indication in the Maan Control Room will be on two light matrices, ane for each division of position measurement, on the Reactor Core Cooling Systems
. benchboard (H13-P601) above the manual control switches for the relief i valves. The indication will be redundant, safety grade, seismically and environmentally qualified, and powered from a Class IE power source. An alarm l indicating that an SRV is open will be provided, but will not be safety grade, i l There are no questions regarding technical feasibility or state-of-the-art of 2 the SRV position indication design, nor is there any concern that it cannot be j implemented prior to OL issuance. 1- + d h 1 i 0-80 Am. No. 57, (5/81)
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-ACNGS-PSAR I
N ITEM-II.E.4.2 ISOLATION DEPENDABILITY' A' NUREG 0718 REQUIREMENT
, ." Containment isolation system designs shall comply with the recommendations of
' Standard Review Plan:Section 6.2.4.
, _ All' plants shall give careful-consideration to the definition of essential'and
. nonessential- systems, identify each system determined to be essential, . identify -
4 each system determined to be nonessential, and describe the basis for selection of each essential system. .All nonessential systems shall be automatically isolated by the containment isolation signal. Revision 2 to Regulatory Guide
.1.141 will contaiu guidance on the classification of essential versus nonessen-tial_ systems and is due to be issued by June 1981. '
j For post-accident situations, each nonessential penetration (except instrument
- . lines) is required to have two isolation barriers in' series.that meet the . ,
requirements of General Design ~ Criteria 54, 55, 56, and 57, as clarified by I Standard Review Plan, Section 6.2.4. Isolation must.be performed automatic- , ally-(i.e., no credit can be given for_ operator action). Manual valves must be sealed closed,'as defined by Standard Review Plan, Section 6.2.4, to qualify as
.an isolation barrier. Each automatic isolation valve in a nonessential pene-tration must receive diverse isolation signals.
- i. The design of control systems for automatic containment isolation valves shall-57 l- be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation
~
j- valves'shall require deliberate operator action. Administrative provisions to
; . close' cil isolation valves ~ manually before resetting the isolation signals is
{. not an acceptable method of meeting this. requirement. Ganged reopening of containment isolation valves is not acceptable. Reopening of isolation valves must be performed on a valve-by-valve basis, or on a line-j= .by-line basis, provided that electrical independence and other single-failure t' criteria continue to be satisfied. The containment setpoint pressure that initiate a containment isolation for 1 ~ nonessential penetrations must be reduced to the minimum compatible with normal operating conditions. The containment pressure history during normal operation-for similar operating plants should be used as 'a basis for arriving at an 'ap-propriate minimum pressure setpoint for initiating containment isolation. The ,
. pressure setpoint selected should be far enough above the maximum observed (or i expected) pressure inside containment during normal operation s'o that inadver-tent containment isolation does not occur during normal operation from instru- I
. ment drif t or fluctuations due to the accuracy of the pressure sensor. A margin of 1 psi above the maximum expected containment pressure should be ade-
-quate to account for instrument error. Any proposed values greater than 1 psi will require detailed justification.
i .. . l; U[ 0-81 Am. No. 57, (5/81)
ACNGS-PSAR S% All' systems that provide an open path from the containment to the environs
- ( j (e.g. , containment purge and vent systems) must close 'on a safety grade high radiation signal.
t 1 Containment purge valves that do not satisfy tuc operability criteria set - forth in Branch Technical Position CSB 6-4 or the 9taff Interim Position of October 23,'1979, must be sealed closed as defined in SRP 6.2.4, Item II.3f during operational conditions 1, 2, 3, and 4. Furthermore, these valves must be verified to be closed at least every 31 days. Applicants shall, to the extent possible, provide preliminary design informa- . tion at a level' consistent with that normally required at the construction I permit state 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 sleected and the supporting design bases and criteria. Applicants shall also demonstrate that the design concept is tech-nically feasible and within the state-of-the-art, and that there exists reason-able assurance that the requirements will be implemented properly prior to the issuance of operating-licenses." .
RESPONSE
Details of the Containment Isolation System-(CIS) are given in Sections 6.2.4
- and 7.3.1.1.2. A summary of conformance to NUREG 0660/0718 Item II.E.4.2 57
{ Criteria is given below. l 1 (1) . Compliance with SRP 6.2.4 (Rev. 1) d. The ACNGS CIS is in full compliance with SRP 6.2.4, Revision 1, with the exception of Position 11.3. Some alternate means of meeting GDC 55 and 56 for certain penetrations are used, as discussed in Section 6.2.4.3.2. These alternate. approaches were previously reviewed and approved by the NRC.
;
(2) Identification of essential and nonessential systems I
' Systems penetrating containment are categorized as follows for isolation purposes:
I
- a. ESSENTIAL: safety and support system for which credit is taken in
- the accident analyses (e.g. ECCS). Essential systems are not automa-
! tically isolated on accident signals.
- b. INTERMEDIATE
- systems which could be useful in mitigating an acci-
! dent, but for which creidt'was not taken in the accident analysis (e.g., equipment protection closed cooling water). Intermediate' systems are automatically isolated by diverse accident I i. U 0-82 Am. No. 57, (5/81)
ACNGS-PSAR p signals (typically high drywell pressure or low reactor vessel level) " D) or on a signal indicating a loss of system integrity (typically dif-
.ferential flow in the containment inlet and outlet lines). The con-trol room hand switch for containment valves in intermediate lines can reopen the valves when the accident signal is still present, but not when the signal indicating loss of system integrity is present.
This permits the operator to use all available systems to cope with an accident while still maintaining the ef fectiveness of the contain-ment.
- c. NONESSENTIAL: systems that are not required or useful in mitigating an accident (e.g. drywell low purity sump). Nonessential systems
- are automatically isolated by diverse accident signals. The control room hand switches for containment isolation valves in nones-sential lines cannot reopen the valves when the accident signal is still present (unless the operator holds the spring-return-to-auto switch in the OPEN position continuously).
System designation into the above categories is shown on Table 6.2-12. (3) -Isolation of nonessential systems
- a. Each nonessential penetration has two isolation br. triers in series
that meet GDC 54, 55, 56 or 57 as applicable.
- b. Isolation of nonessential penetrations with power operated valves is 57 automatic.
- c. Manual barriers (such as manual valves, flanges, etc.) on nonessen-tial penetrations are sealed closed when containment integrity is required,
- d. All automatic isolation valves in nonessential systems receive diverse isolation sig,nals.
3 i (4) Reopening of isolation valves on isolation signal resetting Containment isolation valve logic is such that the valve will not automat-ically reopen when the automatic isolation signal is reset. Reopening of a containment isolation valve requires deliberate operator action (see also Item (1) above). No administrative provisions are necessary to insure this. ] (5) Ganged reopening of isolation valves There is no ganged reopening of automatic containment isolation valves once closed by their respective isolation signal (s). Such reopening requires operator action on a valve-by-valve basis. O-83 Am. No. 57, (5/81) l
'ACNGS-PSAR (6) -Containment pressureLisolation setpoint s
.The drywell pressure setpoint to initiate containment isolation is 2 psig, which allows 1 psig for operational pressure swings and 1 psig for instru-
- ment-error to minimize
- the potential for spurious containment isolation.
(7) _. High-radiation isolation of open path lines !. All lines which provide an open path from.the containment to the environ- , ment, e.g. the containment' purge and vent lines, will isolate on a safety ' grade high radiation signal. In addition, this signal will also isolats i certain other nonessential ~1ines, such as the containment and drywell 57 sumps..as shown on Table 6.2-12. j '(8)- Containment purge l isolation valves
- i. The containment purge and vent isolation valves will satisfy the opera-bility criteria' of- CSB 6-4. See Item II.E.4.4 (3).
4 '(9) Level'of information 4 The above supplements 'the PSAR information on the CIS, and is consistent I with preliminary design information normally required at the CP stage of review. There is no new, novel design, and there are no concerns regard-ing technical feasibility, state-of-the-art or ability to implement the
' intended CIS design.
i ) i i-
- j. i i
i 1, .. 4 ? ( 0-84 Am. No. 57, (5/81)
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5 ACNGS-PSAR ITEM II.E.4.4 PURGING f ( NUREG 0718 REQUIREMENT
" Applicants shall (1) provide a capability for containment purging / venting designed to' maximize purging time, consistent with.ALARA principles for occupa-
- .tional exposure,-:(2) evaluate the performance of purging and venting isolation valves'against accident pressure,L(3) address the interim NRG guidance on valve operability, (4) adopt procedures and restrictions consistent with the revised requirements; and (5) provide;and demonstrate high assurance that the purge system will reliably isolate under accident conditions.
Applicants shall, to the extent possible, provide preliminary design informa-tion at~a level consistent with that normally required at the construction l 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 tech-nically feasible .and within the state-of-the-art, and that there exists reason-able assurance-that the requirements will be implemented properly prior to the i issuance of operating licenses."
RESPONSE
The general safety concern over containment purging stems from the presumption
- that the purge line provides an open path for accident releases prior to isota- 57
{ s tion, and further, that the dyramic ef fects of the accident may interfere with
, y effective._ isolation of the purge line.
O These presumptions are not directly applicable to the Mark III containment design. The reactor coolant system piping is enclosed in the drywell, which communicates with the containment only through the suppression pool. Releases from the primary system are subjected to the quenching and scrubbing action of the suppression pool before entering the containment, so the purge system does not provide an open path for primary system releases in the same sense as other containment designs may. Even so, special care is being taken in the purge system design, specifically for valve operability assurance (Items 2 and 3 below). e The specific points of NUREG 0718 are addressed below: (1) Purging consistent-with ALARA The present design calls for continuous purging of the containment during power operation at 5000 cfm through an 18" line to reduce airborne radio-nuclide concentrations to a level which permits continuous access. This is in keeping with occupational ALARA considerations, because extensive
- containment access for routine maintenance is requirad.
i 1 1
-: O 1
- V
( 0-85 Am. No. 57, (5/81) i ___ .__ _ . - _ _ .m . _.
ACNGS-PSAR
,_s (2) Performance of purge and vent valves against accident pressure
( , ) The purge and vent containment isolation valves are not expected to have to close against the containment design pressure even assuming that a DBA LOCA occurs. The purge and vent lines begin to isolate when drywell pres-sure reaches 2 psig (almost instantaneously, as shown on Figure 6.2-5). Containment pressure is virtually unaffected for the first several seconds of the accident, 'and does not rise to near the design pressure for many hours. Regardless, the purge and vent containment isolation valves will be designed to close against the containment design pressure of 15 psig. (3) Interim NRC guidance on valve operability The purge and vent containment isolation valves are covered by the valve operability assurance program specified in Appendix 3.9.B, Sections 1.2 57 and 2.2, which is consistent with the draft NRC Regulatory Guide "Func-tional Specification for Safety-Related Valve Assemblies in Nuclear Power j Plants." (4) Procedures and restrictions consistent with revised requirements There are no additional procedures or restrictions on containment purge deemed necessary. (5) Assuranceofgurgeandventirmlationreliability
- '"S g The inherent design of the Mark III containment and the added conservatism in isolation valve design and testing give a high level of assurance that the purge and vent lines are reliable to isolate under accident condi-t ions' .
e O i 0-86 Am. No. 57, (5/81)~
ACNGS-PSAR ITEM II.F.1 ADDIfl0NAL ACCIDENT MONITORING INSTRUMENTATION
-[N 'k NUREG 0718 REQUIREMENI "Applica tn s shall comply with the requirements addressed in NUREG-0737.
Applicants shall, to the extent possible, provide preliminary design information at a level consiste'nt 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."
NUREG 0737 REQUIREMENT
" Item 11.F.1 of NUREG-0660 contains the following subparts:
(1) Noble gas effluent radiological monitor;
- (2) Provisions for continuous sampling of plant effluents for postaccident i releases of radioactive iodines and particulates and onsite laboratory
- capabilities (this requirement was inadvertently omitted from NUREG-0660; 57 see Attachment 2 that follows, for position);
[] ( .2) Containment high-range radiation monitor; (4) Containment pressure monitor; (3) Containment water level monitor; and 1 (6) Containment hydrogen concentration monitor. NUREG-0578 provided the basic requirements associated with items (1) through (3) above. Letters issued to all operating nuclear power plants dated September 13, 1979 and October 30, 1979 provided clarification of staf f requirements associated with items (1) through (6) above. Attachments 1 through 6 present the NRC position on these matters. It is important that the displays and controls added to the contro1 room as a-result of this requirement not increase the potential for operator error. A human-factor analysis should be performed taking into consideration: (a) the use of this information by an operator during both normal and abnormal plant conditions, + (b) integration into emergency procedures, e.
%JpS 0-87 Am. No. 57, (5/81)
ACNGS-PSAR O (c) integration into operator training, and It ) U- (d) other alarms during emergency and need for prioritization of alarms."
RESPONSE
Subparts (1) and (3)-(6): The additional accident monitoring instrumentation called for in NUREG 0737, Item II.F.1, Subparts (1) and (3)-(6) w il be provided as discussed below. (1) Noble gas effluent radiological monitor (PSAR Sec*ian 12.2.4.1.2) All cr( e ost-accident release points are provided with high range noble gas effluent monitors, as shown on Table 12.2-Sa. These monitors meet the design criteria specified in NUREG 0737 Table II.F.1-1. (3) Containment high range radiation monitor (PSAR Section 12.1.4) Two channels of high range (1-108 R/hr) radiation monitoring instrumentation will be provided in the containment and in the drywell. Tnese monitors will meet the design specifications in NUREG 0737, Item II.F.1, Attachment 3. (4) Containment pressure monitor
- [V (PSAR Section 7.5.1.4.2)
Four channels of containment pressure instrumentation witn a wide range i of -5 psig to at least 60 psig (four times design pressure) will be provided. The upper range may be higher (up to 80 psig) depending on the 57 exact range of the transmitter to be purchased. These instruments are in addition to the accident normal range containment pressure monitors. (5) Containment water level monitor (PSAR Section 7.5.1.4.2) Four channels of suppression pool water level instrumentation, each covering the range from the top of the ECCS suction strainers to 5' feet above the normal suppression pool level, will be provided. These
! instruments are in audition to the normal range suppression pool water
level monitors. (6) Containment hydrogen concentration monitor (PSAR Section 7.5.1.4.2.11) Two enannels of containment hydrogen monitoring instrumentation, each covering the range of 0-30% will be provided. These will be in addition to the existing accident normal range (0-10%) hydrogen monitoring instrumentation. nm 0-88 Am. No. 57, (5/81)
ACNGS-PSAR The inst rumentation in (1) and (3)-(6) will be redundant, safety grade,
N seismically and environmentally qualified ior accident conditions including -
k ,) the span of its own measured parameter range, and powered from the onsite electrical system. This instrumentation is known to be commercially available , (with the exception of Item (3), as discussed in the next paragraph), and
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space has been allocated for transmitter locations in the plant. The display ,
; location in-the Main Control Room may be in dedicated post-accident panels or ,
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i adjacent to or integrated with the existing normal range instrumentation 4-display. The above supplements the PSAR information on additional acciden* monitoring instrumentation, and is consistent with preliminary design information. l i normally required at the CP stage of review. There are no concerns regarding ! technical feasibility, state-of-the-art or ability to implement this [ inst rumentation design, with one exception. A fully environmentally qualified ; . high range containment radiation monitor has not yet been found. However,
'this is not viewed as critical or even significant at this time.
1 l Subpart (2) ! The requirement of Subpart (2), Sampling of Plant Effluents, is not monitoring , instrumentation per se, but is rather a sample collection and analysis 57 capability'. This will be provided in the manner specified in NUREG 0737, as described below. 4 Sample Collection: The same release points with high range noble gas effluent
monitors will also have particulate and iodine sampling capability, as 4 specified on Table 12.2-Sa. Iodine samples will be taken with a charcoal or i j silver-zeolite cartridge and particulate samples with a filter which are located in the 3 stage monitoring unit cabinet, as shown on Figure ' ! 12.2-1. The poet-accident iodine and particulate samples are extracted from } the release point via the same sample line as the monitoring line.
. Sample Transport: The sample cartridges will be placed in a portable shielded cask and taken to the counting room in the Personnel Access Building.
Sample analysis: Capability for the analysis of sample cartridges will be provided in the PAB counting room. Design of the counting f acility will consider the design basis sample. i
- The precise location of the sample collection station will be selected upon j completion of the post-accident shielding study (Item II.B.2), and the
- location will asure that a worker involved in the sample collection and transport operation will not receive an exposure greater than 5 rem to the whole body and 75 rem to the extremities.
l- : L.O I 0-89 Am. No. 57, (5/81) t
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ACNCS-PSAR ITEM II.F.2 IDENTIFICATION OF AND RECOVERY FROM CONDITIONS LEADING TO
/'"N INADEQUATE CORE COOLINGr !
h [ U NUREC 0718 REQUIREMENT Applicants shall describe their program for developing and implementing procedures to be used by the reactor operators to detect and recover from
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i conditions leading to inadequate core cooling. , l Applicants with PWR plants _shall incorporate in their plant . designs a. primary i coolant saturation meter and all applicants shall incorporate in their plant ( designs instrumentation to detect conditions with a potential that. may lead to inadequate core cooling. 'Any additional equipment, including reactor water' ,
! level instrumentation, that could be used to indicate inadequate core cooling L 4
shall be incorporated in the plant designs. Design requirements for core exit ; I thermocouples are described in NUREG-0737. Applicants shall, to the extent possible, provide preliminary design information 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 l requirements by specifying the design concept selected and the supporting 57 l design bases and criteria. Applicants shall also demonstrate thet the design
, . concept is technically feasible and within the state-of-the-art, and that there exists reasonable assurance that the requirementa will be implemented 5
properly prior to the issuance of operating licenses. ! RESPONSE. ] HL&P concurs with the BWR Owners Group position that no additional 1 instrumentation is needed to monitor inadequate core cooling as discussed in ; the response to Item II.F.3. Regarding operator recognition of inadequate core . cooling, the.BWR Owners Group emergency procedure guidelines for recognizing the ' approach to inadequate core cooling, which were submitted to the NRC by letter, D.B. Waters (BWR Owners Group) to D.G. Eisnehut (NRC) dated January 31, 1981, will be incorporated into the ACNGS plant procedures. Details will be given in the_ FSAR. 2 4 0-90 Am. No. 57, (5/81)
ACNGS-PSAR , } ITEM II.F.3 INSTRUMENTATION FOR MONITORING ACCIDENT CONDITIONS (REG. l GUIDE 1.97) NUREG 0718 REQUIREMENT
" Applicants shall provide in their facility design instrumentation to monitor plant variables and systems during and fol.'.owing an accident in accordance with defineo design bases and Regulatory Guide 1.97, Rev. 2, December 1980.
Designs are already estaolisned for much of the instrumentation that will be required; some of the requirements, however, may involve state of-the art designs or designs which have yet to be developed. Applicants shall, to the extent possible, provide preliminary design information at a level consistent with that normally required at the constrJction 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 critaria. Applicants shall also demonstrate that the design 57 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
Conformance to Regulatory Guide 1.97, Revision 2 is discussed in Appendix C.
,-~' There are no concerns regarding technical feasibility, s tate-o f-the-ar t l i (except for the high range containment radiation monitors) or ability to
\s_,/ implement the post accident monitoring instrumentation prior to OL issuance. Qualified transmitters with the required ranges are known to be commercially available, and space for the indicators has been allocated on the control room panels. Radiation monitors with the required range of the containment hign range monitor are available, but uncertainty exists as to their qualification. When a qualified monitor becomes available, it will be incorporat3d into the TNGS design. In the meantime, the radiation monitoring computer will be left with provisions to accept these monitors data as inputs, and cables will be routed to the monitor locations to facilitate incorporation of the monitors when they become available. n iv / 0-91 Am. No. 57, (5/81)
ACNGS-PSAR ITEM II.K.1.22 DESCRIBE AUTOMATIC AND MANUAL ACTIONS FOR PROPER FUNCTIONING [__ ') 0F AUXILIARY HEAT REMOVAL SYSTEMS WHEN FW SYSTEM IS NOT i
/ OPERABLE NUREG 0718 REQUIREMENT Applicants with EWR plants shall address the requirements set forth in action item 3 of IE Bulletin 79-08. A general explanation of how these requirements will Le met is required prior to issuance of the construction permits.
Sufficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly.
RESPONSE
The BWR/6 NSSS is designed with self actuating systems to assure core cooling, An isolation event can be totally accommodated initially by automatic opera-tion-of engineered safety feature systems and RCIC which are redundant and diverse. These systems restore and maintain system parameters. During the long term, however, t'. eve is adequate time for che operator to take appropriate action. <The operator need monitor and control only reactor vessel pressure and level. Furthermore, the operator has multiple parameters available to provide information on system conditions. All the loss of-feedwater flow cases result in a proportional reduction of vessel inventory causing the vessel water level to drop. Corrective action normally begins as soon as low feedwater flow is sensed (any one or all pumps) 57 x and low level alarm (L4) is reached. At this time a reduction of the core
) recirculation flow is initiated to reduce power and thereby reduce the rate of '- ' level decrease. The first automatic protective action is the low level (L3) scram trip actuation. The reactor protection system responds within 1 second after this trip to scram the reactor. The low level (L3) scram trip function meets the single failure criterion.
For the Loss of Feedwater (LOF) and LOF + Stuck Open Relief Valve (SORV) cases main steam line isolation occurs from low steam line pressure. For the LOF + NO HPCS/RCIC case main steam line isolation occurs from low water level (L1) signal. The main steam line isolation signal also initiates a main steam line isolation valve position scram trip as part of the normal isolation event. The r(actor, however, is already scrammed and shut down by this time due to the L3 scram. Loss of Feedwater Vessel water level continues to drop reaching the L2 trip at about 20 sec. At this time, the recirculation system is completely tripped, and HPCS and RCIC operation is initiatel. After the initiation delay HPCS and RCIC inject into the vessel causing the vessel water level to reach its minimum value about I i /=
-v 0-92 Am. No. 57, (5/E1)
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l-n ACNCS-PSAR
>r s 6.5 -feet above the TAF. In addition, operation of -both HPCS and RCIC will '
cause the vessel to'depressurize which causes a low pressure isolation to ! occur (assuming the reactor has remained in the RUN mode). After the HPCS and , RCIC have tripped of f on high vessel water level or are regulated by. operator - ' I action, the v3ssel will repressurize to the setpcint of the lowest set relief ' I valve:wnich will open to limit the pressure rise. t Loss of Feedwater with Stuck-Open Relief Valve Vessel water-level continues to drop. reaching the L2 trip at about 20 C seconds. At this time, the recirculation system is completely. tripped,,and 2
~
HPCS and RCIC operation is initiated. After the initiation delay HPCS and' RCIC inject into the vessel causing the vessel water level to reach its , { minimum .value about 6.5 feet above the TAF. In addition, operation of both HPCS and RCIC will cause the vessel to depressurize which causes a low ,
- r. pressure isolation' to occur (assuming the ret: tor has remained in the RUN mode). - After the HPCS :sts RCIC have tripped off on high vessel water level or 57 j are regulated.by operator action, the vessel will repressurize to the setpoint 3 of the livest set relief valve which will open to limit the pressure rise. It i 'is assumed in this case, that the relief valve fails to close when the reactor
- pressure drops below the relief valve reset point, thus remaining stuck opan, i j The stuck open relief valve causes the reactor to depressurize to the point- l i where the-shutdown cooling system can be put into operation, t i
j Loss of Feedwater with no HPCS/2CIC Vessel water level continues to drop, reaching the L2 trip at about 20 's seconds. At this time, the recirculation system is completely tripped. With i the failure of HPCS and RCIC the vessel water level continues to drop and the 1 -
' level outside the core shroud reaches the low level (L1) trip. At this time the. main steam line isolation valve will close. The operator can maintain adequate core cooling by manual actuation of the relief valves or ADS to lower reactor pressure and allow use of the low pressure ECCS in time to prevent core uncovery. In this case it was assumed that the operator performed the manual operation at the low level (L1) trip point.
4 i
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j 0-93 Am. No. 57, (5/81) r s
ACNGS-PSAR
/^g ITEM II.K.3.23 CENTRAL WATER LEVEL RECORDING
( ) NUREG 0718 REQUIREMENT Applica:.ts with BWR plants shall address the requirements set forth in Item B.2 of NUREG_0626. Applicants shall implement design modifications as necessary to meet the requirements. Applicants shall submit,' prior to issuance of construction permits, a general explanation of how the requirements will be met. Suf ficient detail shall be presented to provide reasonable assurance that the requirements will be implemented properly prior to the issuance of operating licenses. NUREG 0526 ITEM B.2 In order to simplify the reading of the water level in the vessel and to provide the operators with a record of water level during transients, all BWRs 57 should have the capability to reccrd vessel water level over the range f rom the top of the vessel dome to the lowest pressure tap. This range of water level should be available in one location on recorders which meet normal post-accident recording requirements. The recorders should be started on a reactor trip signal.
RESPONSE
Reacter vessel water level instrumentation spanning the range f rom the bottom of the core support plate to the steam lines centerline will be provided as
\ post-accident monitoring instrumentation, and will be continuously recorded.
V See Table 7.5-0, L 0-94 Am. No. 57, (5/81)
l ACNGS-PSAR
/N I ITEM III.A.1.2 UPGRADE LICENSEE EMERGENCY SUPPORT FACILITIES
( NUREG 0718 REQUIREMENT l 1
" Applicants shall address the requirements for a Technical Support Center, l Operational Support Center and the Emergency Operations Facility. Applicants !
- shall provide preliminary design information in accordance with the functional '
requirements of NUREG-0696 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 design bases and criteria. Applicants shall 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
l The TSC, OSC and EOF are described in Sections 13.3.5.2, 13.3.5.4 and 13.3.5.3 57 l respectively. The information provided in those sections is consistent with the level normally provided at the CP stage, with the exception of the description of the EOF. Efforts to date on emergency support facilities have been concentrated on the onsite facilities which af fect the plcnt directly, such as the ISC and data acquisition system. Since the EOF is an offsite facility and places no constraints on the plant a layout of the EOF has not yet been developed. However, HL&P does commit to design the EOF in accordance with the Commissioners - approved version of NUREG 0696 Section 4. There are no questions regarding technical feasibility or state of-the-art for the emergency support facilities, nor is there any question that they can and will be implemented prior to OL issuance. t I T 8 0-95 Am. No. 57, (5/81)
'ACNCS-PSAR ITEM III.D.1.1 PRINARY'C00LANT SOURCES OUTSIDE THE CONTAINMENT STRUCTURE i NUREG 0718 REQUIREMENT *
;
"NRC is studying the need for improved acceptance criteria for systems outside containment that contain (or might contain) radioactive material either during i normal operations or following an accident. These studies are to be courleted in early 1981, and these mattsrs will be included in the degraded core > -
l making proceeding. i i Applicants shall review the designs of such systems outside containment, and j- their provisions for leaksge control and detection, overpressurization design, ! discharge points for waste gas venting systems, etc., with the goal of mini- , I mizing the possibility of exposure to workers and public during normal opera- !
- tions and in the event of an accident.
In this regard, applicants shall submit, prior to the issuance of construction permits, a general discussion of their approach to minimizing leakage from i such systems outside containment, in sufficient detail to provide reasonable assurance that this objective will be met satisfactorily prior to issuance of operating licenses." l RESPONSE Systems outside containment which may contain primary coolant are designed to 1 minimize leakage to the maximum extent practical, with the goal of minimizing , the possibility of exposure to workers and public during normal operations and in the event of an accident. The means of achieving this are summarized in j Table III.D.1.1-1 and described in the following. i i 1. Leak Reduction and Collection Design Features 4 The systems outside containment which may contain primary coolant were i determined in conjunction with the post-accident shielding study (Item , 57
! II.B.2). These systems incorporate various leak testing, reduction and/or collection features, including: ,
(i) welded / seamless piping system; (ii) pumps with mechanical seals; (iii) low point drains from the piping system and drains from equipment
- such as pumps, leakoff from valves, etc. with single (or double) j isolation valve (s), are routed to equipment drains and, thus, to l- sumps; ,
I (iv) high point vents are provided with single (or double) isolation valve (s) and pipe caps;
- (v) pressure test connections for temporary (or local) instrumenta-tion are provided with single (or double) isolation valve (s) and
pipe caps;
\
0-96 Am. No. 57, (5/81) i
ACNGS-PSAR (vi) lantern rings and leakoffs are specified as follows:
.m '
a) for all valves having 'a connection to the leak detection system, b) for rising stem for 4" and larger gate and globe valves, i with a maximum operating temperature of 2120F. (vii) packless metal' diaphragm valves will be utilized for 2" and smaller valves located in the process line.
- 2. Leak Detection Features For intersystem leakage detection, the following provisions are incorpor-
. ated in the design:
i (i) radiation monitors with alarm annunciation in the control room i are provided for the Service Water systems which remove heat from the RHR, RWCU Non-Regenerative, and Fuel Fool Cooling heat ex-
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changers, thereby enabling detection of radioactive fluid in the 57 non-radioactive side due to heat exchanger tube leaks; (ii) double block valves (and/or check valves) are provided in non-radioactive systems that are utilized for flushing radioactive system outside containment. In addition, relief valves are provided for protection against' overpressurization. ] (iii) instramentation to monitor piping system integrity (differential i flow transmitters), high differential and high ambient area tem-perature monitors, and sump level measurement. l 3. Periodic System Leaka;: Testing
Leak testing provisions, as defined in 10CFR50, Appendix J - Type 'C' ! tests- have been incorporated for all containment isolation valves for j those systems penetrating the containment. Necessary pressure test connections with block valves have been included as part of this testing requirement. Also, all safety systems will undergo scheduled leak , testing in conjunction with hydrostatic testing as per the requirements of ASME Section XI to confirm the integrity of tr.e piping system /com-ponents. i i i v 0-97. Am. No. 57, (5/81)
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k ACNCS-PSAR TABLE Ill.D.I.1-1
SUMMARY
OF LEAK REDUCTION CRITERIA ITEM SYSTEM LEAK TESTINC LEAK REDUCTION / INTERSYSTEM LEAK No. SYSTEM CAPARILITY ,___ COLLECTION MEASURES DETE CTION/R EDUCTION RE MARKS 1 Residual Heat 1) Scheduled System Leak teats in 1) fiping joints are of welded / seam- 1) Essentist Service Water Removal Systems conjunct ton with hydrotests less construction. System (low pressure side of (RHR) will be made at intervals not the RHR Heat Exchangers) is to exceed each inspection provided with radiation interval as per ASME Section monitors with alarms to XI (para. IWA-5000). alert the operator of tube leakeges.
- 2) In accordance with the require- 2) Mechanical seals are provided on 2) Double block valves (and/or ments of 10CFR50 Appendix 'J' - a ll pump s . check valves) are provided containment isolation valves in non-radioactive systems will be tested during each that are utilised for flush-reactor shutdown for refueling ing/ draining radioactive but in no case at intervals sys tems outsida containment.
greater than 2 years, in addition, relief valves are provided for protection against over-pressurization. 57 o 3) Valve seat leakage tests and 3) Leakager f rom equipment, safety 3) Local test pressure coanec-E operability (surveillance) tes- relief valve, end low point drains tions are provided to check
- ting is per formed in accordance from the piping systems will be intersystem leakage through with ASME K1 (para. IWV-3000) piped to equipment hubs in the valves; for example, between drainage system to reduce floor RHR/RCIC System.
contamination.
- 4) Lantern rings and leakof fs are specified as follows:
i) For all valves having a con-nection to the leak detectton system. ii) for rising stem for 4" & *arger gate & globe valves, with a max. operat ing temp. of 2120F. 4A) Packless metal diaphragm valves
-$ will be utilized for 2" &
(; smslier valves located in the
- process line.
v
- 5) The following leakage detection instrumentation is provided:
A) High dit terential and high ambient area tempe ratures (in-let & outlet air is comparei for dif ferential air tempera-ture).
m .~-_ , . _ . . _ . . . . _ _ _ _ . ~ . . _ _ . _ _ ..m . . . . . _ _ . _ = . . . . . . -...m. _ ,. _ .- _ .- _ -.
._- / / 3
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AC3CS-PSAR TABLE III.D.1.1-1 (Cont'd) ITEM SYSTEM LEAK TESTINC LE AK REDUCTION / INTERSYSTEM LEAK No. SYSTEM . CAPABILITY COLLECTION MEASURES DETECTION / REDUCTION REMARKS 1 (RHR) (Cont'd) 5) Dif ferential flow transmitter (will confirm the integrity of the system piping). C) ECCS sump level measurement and level alarms . D) Two pressure transmitters are available to monitor leakage through the containment isola-tion valves.
- 6) Pressure test connections are pro-vided for containment isolation valve leak rate testing per 10CFR50 Appendix J and ASME 11 (lWV-3000).
- 7) Pressure test connections for 57 temporsty (or local) instrumen-tation are provided with single o
a (or double) isolation valve (s) *
- and pipe cape.
- 8) High point vents are provided with signie (or double) isolation valve (s) with caps.
l 2 Low Pressure Same as Item #1 Same as Item #1 Same as Item #1 Core Spray (LPCS) Except 5A, D Except I and 3 3 High Pressure Same as Item #1 Same as Item #1 Same as Item #1 l Core Spray Except 5A, D Except I and 3
- (HPCS) 4 Reactor Core Same as Item #1 Same as Item #1 Same as item #1 Isolation Cooling Except for 1.
I lI . i (RCIC) 4
*g:
5 Main Steam Same as Item #1 Same as Item #1, except for 2 Not applicable U ! solation Valve and 5 level in MSIV-LCS Stem 1 Leakage Control Leakage collection tanks moni-1 (2 System (MSIV-LCS) tored and alarmed. 1 os j :: 1 , 1 1
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ACNGS-PSAR ITEM III.D.3.3 IN-PLANT RADIATION HONITORING
/m (O) NUREG 0718 REQUIREMENT Applicants shall review their designs to assure that provisions for monitoring inplant radiation and airborne radioactivity are appropriate for a broad range of routine and emergency conditions. Applicants shall, to the extent possible, provide preliminary design information 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 basis and criteria. Applicants shall als 57 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 opersting licenses.
RESPONSE
Portable equipment for inplant radiation monitoring and sampling for particulate, iodine and noble gases useful under routine and emergency conditions will be provided. This equipment will not be purchased for neveral years, but it is expected that it will be cart mounted and backup battery powered. Plant personnel will be trained in the use of this equipment under both routine and emergency conditions. Details will be provided in the FSAR. I v N l I
- I v
i 0-100 Am. No. 57, (5/81)
ACNGS-PSAR ITEM III.D.3.4 CONTROL ROOM HABI*tABILITY (g) O NUREG 0718 REQUIREMENT Applicants shall review the design of their f acilities for conformance to requirements stated in the Action Plan. NRC will consider possible new criteria to preclude control room contamination via potential internal pathways indicated by the TMI-2 experience. Applicants shall address prior to the issuance of the construction permits or manufacturing license, how they will implement the existing requirements set forth in this Action Plan item. Applicants shall also address the extent to which improvements have been made to prevent control room contamination via pathways not previously considered. Applicants shall, to the extent possible, provide preliminary design information 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 57 The present ACNGS control room habitability design is state-of-the-art. It i i was previously reviewed against Regulatory Guides 1.78 and 1.95 and Standard Review Plans 2.2 and 6.4 by the NRC and found acceptable. The Control Room Ventilation System design concept has not changed from that presented in the PSAR. It is a pressurized design, with two widely separated emergency air intakes. Each intake is provided with redundant radiation monitors to permit the operator to select the cleaner intake post accident and the intakes are provided with a " series valves in parallel" arrangement which is single failure proof for both isolation and opening. The normal air intake is automatically isolated on high radiation, and is single failure proof to isolate. The Action Plan makes reference to possible new criteria on internal contamination pathways. Obviously, this new criteria cannot be addressed for ACNGS until it is issued, but it is not expected that any reasonable requirements would affect ACNGS. The Control Building is a separate structure, and the only equipment in the Control Building which is not control, instrumentation or electrical in nature is the HVAC equipment servicing the Control Building. This is unlike other plants in which the Control Room is actually a part of the Peactor Auxiliary Building, where potentially radioactive equipment may be located. Thus, it is difficult to l envision any additional contamination pathways into the ACNGS Control Room l other than the airborne pathway already considered, l s J 0-101 Am. No. 57, (5/81)
ACNGS-PSAR 4 l' NUREG 0718 CATEGORY 5 .g "A requirement for information of the type customarily reviewed at the preli- !' minary design stage for the following types of items:
- a. Items for which the required information should be sufficient to demon-
!' strate that the requirement has been satisified by the application. This is the kind of information and degree'of detail customarily provided at the preliminary design stage with respect to site and major systems and structures to satisfy 50.34(a)(1). This will also be applicable to items , relating to technical qualifications of the applicant and its management 57 for design and construction. I b. Itema for which the required information should be sufficient to assure that the requirement will be met at the final design stage. This is the 7 kind of information and degree of detail customarily provided at the pre-
- liminary design stage with respect to the preliminary design of the faci-lity to satisfy 50.34(a)(3)(4), etc."
RESPONSE l Responses to the Cater,ory 5 items. are given herein. i
} !,b 5 k a e Am. No. 57, (5/81) 0-102
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ACNGS-PSAR ITEM I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING, DESIGN AND 00NSTRUCTION (3 EXPERIENCE N- ' Y NUREG 0718 REQUIREMENT
" Applicants shall submit a description of their administrative procedures for the evaluation of operating, design, and construction experience and describe how they will assure that applicable important industry experiences origina-ting f rom both within and outside the applicant's construction organization will be provided in a timely manner to those designing end constructing the plant. These procedures shall: (1) Clearly identify organization responsi-bilities for review and identification of these important experiences and the feedback of pertinent information to those responsible for designing and constructing the plant; (2) Identify the administrative and technical review steps necessary in implementing applicable inportant experiences; (3) Identify the recipients of various categories of information f rom these experiences or otherwise provide means through which such information can be readily related to the job functions of the recipients; (4) Assure that applicant and contrac-tor personnel do not routinely receive extraneous and unimportant experience-related information in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency; (5) Provide suitable checks to assure that conflicting or contradictory information is not conveyed to applicant and contractor personnel for implementation until reso--
lution is reached; and (6) Provide practical interim audits to assure that the feedback program functions ef fectively at all levels. Sufficient detail shall 57 be presented to provide reaconable assurance that the requirements will be 2 implemented properly prior to the issuance of the construction permits or
/s manufacturing license."
k
RESPONSE
HL&P has administrative procedures for the evaluation of operating, design and construction experience to assure that applicable important industry experi-ence is incorporated into ACNGS. the following is a description of those procedure s.
- 1. Organizational Responsibilities Operating, design, and construction experiences f rom outside HL&P are directed to the Nuclear Licensing Department. Within this department, there exists a generic licensing group that is responsible for reviewing 1 the information received and identifying those experiences which may be of interest to ACNGS. The licensing group is also responsible for categor-izing these experiences such that operating, design, and construction experiences are directed to the respective sub-organizations within HL&P for their review and use.
Operating, design, and construction experiences f rom within HL&P are directed to the ACNGS Project Manager, who is responsiole for reviewing and categorizing the information, then directing the information to the j
- '
ACNGS operations personnel, engineering team, or Construction Manager, as 1 appropriate. O (~ 0-103 Am. No. 57, (5/81)
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ACNGS-PSAR
- 2. Administrative and Review Steps and m
-!: I- 3. Recipients of Information d
-a. General I HL&P ' contracted for design and construction with Ebasco Services Incorporated and the General Electric Company, the principal con-tractors. As part of its responsibilities, the General Electric Company has, within its Nuclear Services Department, established and maintained a formal service advisory communication system that is designed to provide the BWR Owner-Operator with a broad coverage of BWR j operating and maintenance information and recommendations. Addi- ; tionally, HL&P is responsible for advising both Ebasco and GE of oper-ating, design, and construction experience uniquely available to HL&P, ouch as from South Texas Project design and construction, and from utility owners groups, in cases where that unique data is revelant to
- Allen's Creek.
i
- b. Houston Lighting and Power Company
- -
HL&P functions within the program to 1) review and approve GE's pro-gram, 2)-audit and monitor principal contractor implementation of their programs, 3) furnish data from a selected document list, including data uniquely available to HL&P, and 4) provide direction to Ebasco for l incorporating and implementing design and construction experiences into the ACNGS design. Operating, design, and construction experience inform 1 tion from extern-al sources enters the HL&P program from two general categories: 1)
; Regulatory Agencies and 2) Industry Sources. Examples of documents
!- reviewed are as follows: f 1. Regulatory Agency Information f 4
. License Event Reports
. Regulatory Guides
. Regulatior.s (10CFR and 49CFR)
. IE Bulletins, Circulars, Orders and Notices
. NUREGS
. Standard Review Plans (including Branch Technical Positions)
)- 2. Industry I l- . Topical Reports from General Electric and the Nuclear Safety Analysis Center
- - . IEEE, ANS, ANSI, and ASME Codes and Standards
. License Event Reports '
. NSAC/INPO Significant Events Evaluation Information Network
. Owners Group Activities i
0-104 Am. No. 57,.(5/81) l
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s As external information enters the HL&P system, it is directed to the ( i Nuclear Licensing Department. There it is categorized, screened for
\% _- applicability, and documented. The review at this stage is two-fold in ,
purpose : 1) to reduce the quantity of information received to ' manageable amounts by culling out information clearly not relevant to ACNGS, and 2) to broadly categorize the information into operations, design, or construction categories. The Nuclear Licensing Department will then transmit the information to either the ACNGS Project Operations Team Leader, in the case of operational information, or the ACNGS Engineering Team Leader, in the case of design information or the ACNGS Construction Manager, in the case of construction information, along with a specified time by which disposition of the items must be , fed back to Nuclear Licensing. j HL&P will provide continuous assessments of the ef ficacy of the exper-ience feedback programs at HL&P and the principal contractors by using a commitment tracking system which will provide feedback to the Nuclear Licensing Department as to the ultimate resolution of the information that Nuclear Licensing sends out to the project. The same commitment tracking will be utilized for HL&P internally generated experience inf o nnation. 57 . In addition to the project organization, the inf ormation is sent to the line organization departments for information.
' Operating Experience ' (\'yb Information on operatind experience is received f rom the Nuclear Licensing Department by the ACNGS Operations Team Leader, who will perform a more detailed review of the infonmation. This review will determine if the informatLon is applicable to A"NGS and if it is of suf ficient concern to pursue with the principal contractors. If warranted by the nature of the item, Operations will consult with the ACNGS Project Engineering Team and recommend a course of action. The Operation's Team Leader will obtain assistance as necessary f rom the ACNGS Engineering Team in his review.
In some cases, the appropriate action will be decided within HL&P, particularly if it is in plant maintenance or operations. The operational concern may then be resolved as part of the normal proce 's of design, operator training, or procedures development. Design Experience For design experience, the ACNGS Engineering Team has the prirary responsibility for resolving concerns once the information is received f rom the Nuclear Licensing Department. The Engineering Team Leader will review the information and direct it to the appropriate discipline leader for a determination of the necessary action. As with operatind expei tence, the Discipline Leader may consult with either or both prin-c Lpal contractors to evaluate the concern. From this point, no rmal
- s. design control processes are used.
[ ' V) o-105 Am. No. 57, (5/81)
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' Construction Experience ~ !
2 h
; For construction experience, the ACNGS Construction Manager has the
{ l \ -primary responsibility for resolving concerns once the information is i received from the Nuclear Licensing Department. He may use assistance ,
! from the ACNGS Enginecting Team and either or both principal contrac-tors, as appropriate. Construction concerns that affect plant design
will be resolved in' cccordance with the project's normal design process. ' i ! 4
- c. General Electric The GE-Nuclear Services Department maintains a formal service advisory ;
communication system that is designed to provide the BWR Owner-Operator with a broad coverage of BWR operating and maintenance information and
- i. recommendations'. This system, implemented by the Service Information Letter (SIL), is designed to collect, process, and disseminate infor-mation pertinent to:
- 1) unique operating conditions and experiences i
. 2) improved methods, techniques and procedures for operating and I maintaining BWR plant equipment
- 3) plant performance inprovement and equipment upgrading 57 i 4). safety, licensing and' other regulatory matters.
I The major sources of information, including data, drawings, equipment, catalog /part numbers, problem definition, technical work recournenda-tions, and other technical material required to prepare SILs include: n Q 1) Application Information Documents (AIDS)
- 2) Field Engineering Memos (FEMs) 4
- 3) . Product Experience Reports (PERs) ,
- 4) Safety and Licensing Reports
- 5) Reports and Instructions prepared by GE Engineering organizations
< 6) GE and Vendor Equipment Instruction Manuals
- 7) Equipment Failure and Reliability Reports
- 8) BWR Plant Owner-Operator (s) and utility management suggestions 1
- 9) 'Startup and Preoperational Test Reports ,
Occasionally, a need may arise to transmit to the utility owners with. operating BWRs an urgent announcement of a potential operational hazard or 'other information which potentially could seriously impact plant operations. In general, such announcements will consist of a brief but i adequate explanation of the situation with advice or precautionary measures to be observed. I Prior to release fron GE-Nuclear Services Department, SILs will undergo formal review by respe- ible design engineer, other cognizant engin-t eers, and GE managemei.c representing various disciplines including engineering, startup tests, licensing, and services. I rt Am. No. 57, (5/81) 0-106
ACNGS-PSAR
- 4. Avoidance of Extraneous and Unimportant Information 4
The Nuclear Licensing Department, through its normal screening process,
. will' assure the avoidance of extraneous and unimportant information.
i
- 5. Avoidance of Conflicting or Contradictory Information i
The Licensing Department will assure that potentially conflicting or con- ' tradictory information is identified and transmitted to the appropriate 57 organization for resolution. , 6. Practical Interim Audits HL&P will ensure complieve with these requirements by monitoring and
- ) periodic audits of HL&P, General Electric, and Ebasco activities.
[ t
]
= ) I } l t 7 1 i-l ' l ) d I i [ 0-107 Am. No. 57, (5/81) i _ ,, . _ . , - . . , _ , , , . _ , . . . _ ~ . , . . . . . . , . _ . , . . , , . _ _ _ , . . . , _ - ~
4 ACNGS-PSAR [] ITEM 1.F.1 EXPAllD QA LIST I '"'
" Prior t'o issuance of the const ruction permits or manufacturing license, ap-plicant s shall revise their -QA programs by expanding their QA list s to include all items and activities affecting safety as defined by Regulatory Guide 1.29 and Appendix A to 10 CFR Part 50, and shall provide a commitment to apply the revised QA program to all such items and activities."
4 RESPONSE-HL&P is constantly re evaluating its QA list to ascertain that all it ems and - activities related to safety have been identified. This updated list is found in revised Table 3.2-1 of the PSAR. This QA list is comprehensive, and con-t ains far more items than was t ypical in t he indust ry unt il recent ly.
- The QA list may be further expanded as a result of ongoing activit ies related -
t o t he TML-2 event . Any such items (i.e. hydrogen cont rol, post accident 57 l monit oring syst em, etc.) will be added to the list at the appropriate time using exist ing procedures. In addit ion, applicant has re evaluated it s QA
. list based upon criteria which relate to the use of nonsafety systems during and following accident s. These criteria relate to all equipment , components, and mat erial which do not fall in Safety Classes 1, 2, or 3 but :
a) Which are assumed to be in service in PSAR Chapter 15 accident or t rans ient which, if failed, could increase the consequences of the l accident or transient beyond those evaluated in the PSAR. b) Whose act ive failure during a PSAR Chapt er 15 accident or t ransient will increase t he consequences of that accident or transient beyond t hose evaluat ed in the PSAR. l c) Whose availability to funct ion on demand following a PSAR accident or t ransient is assumed t hroughout the accident or t ransient. 1
- Using the above as criteria, no such equipment , component s, or mat erial quali- ; fy t o be added to t he QA list . If such items are ident ified in the future, l they will be added to the QA list using existing procedures.
1 =i 9 4 l i. v 1 0-108 Am. No. 57, (5 /E 1)
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.,w. ITEM I.F.2 DEVELOP 'HORE DE~. AILED QA CRITERI A -/ \
NUREG 0718 REQUIREMENT Applicant s shall describe the changes to thcir QA programs that have result ed from their review of the accident at TMI-2. In addition, applicant s shall address the appropriate mat t ers discussed in this Action Plan item and t he ext ent t o which t hey have been considered in t heir QA program. Applicant s shall submit , prior t o the issuance of the const ruction permit s or :nanufac-turing License, a revised descript ion of their QA program that includes con-siderat ion of t hese matters. PROPOSED 10CFR50.34(e)(3)(iii) Establish a quality assurance (QA) program based on considerat ion of: (A) ensuring independence of the organizat ion performing checking functions from the organization responsible for performing the functions; NRC ACCEPTANCE GUIDANCE The QA program includes: 2Al Verification of conformance to established requirement s is accomplished 57 (IB2) by individuals or groups,within the QA organizat ion who do not have direct responsibility for performing the _ work being verified. Rat ion-C ale and just ification must be provided if performed by other than the 1( QA organizat ion. i 2A2 _ The QA organizat ional responsibilit les for inspect ion are described. (1081) Individuals performing inspections report to t he QA organizat ion. 2A3 Verificalion of suppliers' act ivitles during fabricalion, inspection, (7A2) t est ing, and shipment of materials, equipment , and component s is planned and performed with QA organization participation in accordance wit h writ t en procedures to assure conformance t o t he purchase order requirements. These procedures, as applicable to the method of pro-curement, provide for:
- a. Specifying the characteristics or processes to be witnessed, in-spected, or verified, and accepted; the method of surveillance and t he ext ent of document at ion required; and those responsible for
- implement ing these procedures.
- b. Audit s, surveillance, or inspect ions which assure t hat the supplier complies with the quality requirement s.
2A4 Receiving inspect ion is performed by the QA organizat ion to assure:
-(7B1)-
. a. The material, coraponent , or equipment is properly ident ified and corresponds t o t he ident ificat ion on the purchase document and the receiving document at ion. I 0-109 Am. No. 57, (5/81)
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4 i
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t i; 4 b. . Mat erial, component s, equipment , . and accept ance records sat is fy the inspect ion inst ructions 1 prior to installat ion or use. i Specified inspect ion, test and of her records, (such as cert ificat es
~
[ c. I of conformance attesting that t he material, component s, and equip- [ J ment conform t o specified requirements) -are available at the nu- <
- clear power plant prior to inst allat ton or use, j
2A5 Correct ident ificat ion of mat erial, part s, and component is - veri fied (883)- and doc smented by the QA organization prior 1o release for fabricalion, I assembling, shipping and installation. ( 2A6 Procedures are established for recording evidence of acceptable f 7 (982) accomplishment of special processes using qualified procedures, equip- ( J- ment, and personnel. The QA organizat ion verifies the recorded evi- , dence and document s the result. ! k- I j 2A7 Inspection and test results are documented, evaluated, and their (10C3) acceptability determined by a responsible individual or group. The QA !
- (11Cl) organizat ion as a minimum evaluates, verifies, and documents complet e-d )
ness of this activity.
\
j 2A8 - Follow up action is taken by the QA organizat ion to verify proper ! l (16.3) implementation of corrective act ion and to close out the corrective 57 ; j act ion in a timely manner. [ t I RESPONSE I r I I 2Al Verification of conformance to established requirements is accomplished - l {' by personnel within the QA' organization. . These individuals do not have direct responsibility for performing the work being verified. Chapter , 17.1.1 of the Allens Creek PSAR describes the Ebasco and t he HL&P .QA . , [ organizat ion. Figure 1 shows the HL&P project QA organizat ion and - Figure 2 shows the Ebasco QA organization. I 1 2A2 The Ebasco QC organization reporting to the Quality Program Sit e I I l. Manager is responsible for the performance of inspect ion. i 1 1 l 2A3 The Project QA/QC organizat ion will perform the following act ivit ies: i
- a. - Planning, performance, and verificat ion of suppliers' act ivit les during fabricalion, inspect ion, test ing, and shipment of materials equipment, and component s in accordance with writ t en procedures or '
^ e vendor inspect ion plans to assure conformance to the purchase order
- requirement s. These procedures or vendor inspect ion plans, as applicable to the met hod of procurement . provide for: -
r 1) 'Specifying the characterist ics or processes to be wit nessed, i inspect ed, or verifled, and accept ed; the methad of surveit-
- lance and the extent of documentation required; and those responsible for implement ing these procedures.
J 0-110 Am. No. 57, (5/81) w ",, - - -,.,-,,w , ,, , ,,-_,mmn-,,-n , , , , , , , , . , , . ,.ww,,,,,..,.nwr. ,,g7.,w-., , ,,, , , , ., p w ,r.,.w m s,w p , y,g y v .g e a
ACNGS-PSAR
- 2) Audit s, surveillance, or inspect ions which assure that the f, s '). supplier complies with the quality requirement s.
Q 2A4 b. Receiving inspect ion t o assure:
- 1) Ihe mat erial, component, or equipment is properly ident ified and corresponds t o t he ident ificat ion on t he purchase document and the receiving document ation.
- 2) Mat erial, components, equipment , and accept ance records sat isfy t he inspect ion inst ruct ion prior t o inst allat ion or use.
' 3) Specifixd inspect ion, t est , or ot her records (such as cert ifi-cates of conformance attesting that the material, component s, and equipment conform to specified requirement s are available at the nuclear power plant prior to inst allat ion or use. 2A5 c. Correct ident ificat ion of material, part s, and components is veri-fied and documented prior to release for fabrication, assembling, shipping, and installat ion. ' 2A6 d. Procedures are established for recording evidence of acceptable accomplishment of special processes using qualified procedures, equipment, and personnel. The QA organizat ion verifies the re-corded evidence and document s t he result . 57 jr~'s 2A7 e. Inspection and test result s are documented, evaluat ed, and their ( acceptability determined by a responsible individual or group. The QA organizat ion, as a minimum, evaluates, verifies, and document s complet eness of this act ivity. . 2A8 f. Follow up act ion is t aken by the QA organizat ion to verify proper i implement at ion of correct ive action and to close out the corrective action in a timely manne'r. PROPOSED 10CFR50.34(e)(3)(iii) (B) performing the ent ire quality assurance / quality cont rol funct ion at construction sites; NRC ACCEPTANCE GUIDANCE The QA program provides provisions to assure that: 2B1 The person at the construct ion site responsible for direct ing and (1C3) :anaging t he site QA program is ident ified by posit ion. He report s t o
.he offsit e QA organization and has appropriate organizat ional posi-tion, responsibilit ies, and authorit y to exercise proper cont rol over the QA program. This individual is free from non-QA duties and can t hus give full at tent ion t o assuring that t he QA program at the plant sit e is being effect ively implemented.
V 0-111 Am. No. 57, (5/81)
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i 2B2. . Designated QA individuals are involved in. day-t o-day plant act ivit ies
.(156) Important t o safet y (i.e. , the QA organizat ion routinely attends . and i \ participates in daily plant work schedule and st atus meetings to assure >
t hey. are kept abreast ~ of day-t o day work assignments throughout . the [ plant and t hat there is adequate QA. coverage.relat ive t o procedural ani
- j inspect ion cont rols, accept ance . crit eria, and QA staffing and qualifi-cat ion of personnel t o carry out QA assignment s). -
RESPONSE
281 The HL&P Project Quality Assurance Manager shall be located at the ' l l
. Const ruction sit e and .is responsible for direct ing and. managing t he
- sit e QA, program. He' is free from non-QA dut ies. The HL&P Project y
E Quality Assurance Manager is responsible for providing t he programsat ic direct ion and administering the policies, goals, object ives, and ,L methods for t he Allens Creek Project which are described in t he Project l Quality Assurance Plan. Programmatic direct ion is defined as the role of HL&P in establishing the program requirement s and ensuring the 3 l adequacy of t he qualit y assurance program for, HL&P and t he prime con- :
t ractors. The Project Quality Assurance Manager report s t o the ' Manager, Quality Assurance, who report s direct ly to t he Execut ive Vice
- President and has - t he independent authority to ident i fy quality related r
problems, t o init iat e or reconsend solut ions, t o control existing l 57 1 nonconformances, to verify implementation of approved dispasit ions, and I
- when necessary t o st op work. !
. 282 Project QA personnel are involved in plant act ivit les import ant to
{ safety and are kept abreast of work schedule and const ruction activi-4 t les by periodically at tending const ruction st atus meet ings. Project t QA personnel ensure that there is adequate QA coverage relative to ; l procedural and inspect ion cont rols, acc.ept ance criteria, and QA st af- l fing and qualification of personnel to carry out assignments. !
- PROPOSED 10CFR50.34(e)(3)(iii) 1 j (C) including QA personnel in (the review and concurrence) of qualit y relat ed ; q
- procedures (and documents) associst ed with design, const ruct ion, and l inst allat ion;
;
1 NRC ACCEPTANCE GUIDANCE 4 l The QA program includest ; i 2C1 Provisions are est ablished to assure that qualit y affect ing procedures
. (2Bla) required to implement the QA program are consistent wit h QA program
. . commitment s and corporat e policies 'and are properly documented, con- ' 4 t rolled, and made mandatory through a policy stat ement or equivalent document signed by the responsible official. I 2C2 Tae QA organizat ion reviews and- documents concurrence wit h t hese l i (2 Bib), qualit y related procedures. l cQ l s 0-112 Am. No. 57, (5/81) l b i i _ . . . .. a .: = - -- - "- -J
ACNGS-PSAR
' 2Ci Procedures are est ablished. for the' review of procureisent document s to (X
c' (4Al) determine t hat quality requirement s are correctly st ated, inspectable,
-and controllable; there ~are adequate acceptance and reject lon Lcriterial and procurement docusent s have been prepared, revLewed,- and approved 'in . '
. , accordance wit h .QA' program requirement s. 1To t he ext ent , necessary,
, procurement document s should require cont ractors .and subcont ract ors t o .
! . . provide an acceptable quality assurance . program. The review and docu- ! mented concurrence of the adequacy of quality requirements stated in'
- procurement document s is performed by QA personnel.
- 2C4 . Procedures for the' review, approval, and issuance of documents and '
- (6A2) changes thereto are est ablished and. described to as'sure technical
- adequacy and inclusion of appropriate quality. requirements prior to implementalion. The QA organizat ion reviews and document s concurrence I wit h. these document s with regards t o .QA-related aspect s.
!' .i L 2c5 Inspect ion procedures, ' instruct ions, or checklist s' provide for the (10C1) . following as reviewed and concurred with by the QA organization for QA j_ aspects 'and other technical organizat ions, as appropriate:
f a. .Ident ification of characterist ics and activit ies to be inspected.' i , b'. A descript ion of the method of inspect ion. s , c. Ident ificat ion' of t he individuals or groups responsible for per- - , [ ' forming the inspection operation in accordance with the provisions i of-item 1081. N , 57 l d. Acceptance and reject ion criteria. l ,
- e
- e. Ident ification or required procedures, drawings, and specifications j and revisions.
- f. Recording inspector or data recorder and the reault s of the .inspec- l +
t ion operet ion. 1 i
- g. Specifying necessary measuring and test equipment including accu-
- racy requirement s
{ 2C6 Test procedures or inst ructions provided for the following as reviewed j (1181)land concurred with by the QA organizat ion for QA aspects and by other { t echnical organizat ions for technical aspect s-i l a. Tne requirement s and acceptance limit s cont ained in applicable l'
. design and procurement document s.
- b. cInst ruct ions for performing the t est .
j c. Test prerequisit es such as calibrated inst rument at ion, adequat e [- . test equipwnt 'and inst rument at ion including their accuracy
- requirements, completeness of item to be tested, suitable and j> - controlled environmental conditions,'and, provisions for data collection land storage. ')
l l t 0-113 Am. No. 57,' (5/81) L
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- d. Mandatory inspect ion hold point a for wit ness by owner, cont ractor,
. [m
- \ or inspector (as required). '
- e. Accept ance and reject ion crit erta.
;
- f. . Methods of documenting or recording test data and result s. '!
- g. Provisions for assuring test prerequisit es have been met .
. 2C7 Procedures are estaoiished and described for calibrat ion (technique and (12.3) frequency), maintenance, and cont rol of the measuring and test. equip-ment (inst rument s, t ools, gages, fixtures, reference and t rans fer st andards, and nondest ructive test equipment i that is used in the measurement, inspect ion, and monitoring of st ruct ures, systems, and components. The review and documented concurrence of these procedures is described and t he organizat ion responsible for t hese funct ions is ident i fied.
2C8 Procedures are established and described to control the cleaning, t (13.2) handling, st orage, packaging, and shipping of materials, components, , and systems in accordance with design and procurement requirement s to j preclude damage, loss, or det eriorat ion by environment al conditions such as t emperature or humidit y. The QA organization reviews and document s concurrence of t hese procedures. 2C9 Procedures are est ablished to indicate the inspect ion, t est , and
, (14.1) operat ing st at us of st ructures, syst ems, and components and throughout (14.4) f abricat ion, installation, and test. The QA organization reviews and 57 document s concurrence with these procedures.
2010 Procedures are established and described to control the applicat ion and (14.2) removal of inspection and welding stamps and status indicators such as ) (14.4) t ags , markings, labels, and stamps. The QA organizat ion reviews and l documents concurrence with these procedures. 2011 Procedures are est ablished and described to control sitering the (14.3) sequence of required t est s, inspections, and other operat ions important - (14.4) t o safet y. Such act ions should be subject t o t he same cont rols as t he original review and approval. The QA irganizat. ion reviews and docu-
- ment s concurrence with these procedures.
a 2C12 Procedures are established and described for ident ification,
- (15.1) documentation, segregation, review, disposition, and not ification t o affect ed organizat ions of nonconforming mat erials, part s, component s, and as applicable to services (including computer codes) if disposit ion
- is other t han to scrap. The procedures provide ident ification of aut horized individuals for independent review of nonconformance, in-ciuding disposit ion and closeout.
l 0-114 Am. No. 57, (5/81)
ACNGS-PSAR
~
3 2C13 QA and other organizat ional responsibilit ies are described for the
) (15.2) definit ion and implement at ion of act ivities relat ed to nonconformance d
control. This includes ident ifying those individuals or groups wit h authority for the disposit ion of nonconforming items and involvement of the QA organization in documenting concurrence to the disposition, sat isfactory completion of the disposit ion, and correct Ive act ion. 2C14 Procedures are established and described indicaling an effect ive (16.1) correct ive act ion program has been est ablished. The QA organizat ion reviews and documents concurrence with the procedures. RESPONSE f 2Cl Ine HL&P Project Quality Assurance Program for the Allens Creek Nuclear 2C2 Generat ing Stat ion is described by the HL&P Project Quality cssurance Plan (PQAP) . A let ter signed by the Execut ive Vice President in t he front of t ne PQAP makes t he requirement s of t he PQAP mandat ory. Proce-dures are reviewed by project QA personnel during preparation for inspections, surveillance, implement ation reviews and audit s to ensure consist ency wit h project requirement s. Addit ionally, select ed proce-dures are reviewed and concurred with by the project QA organizat ion prior to issuance. 2C3 Procedures are established for the review of procurement document at ion 7m by project QA personnel to det ermine that 1) Quality requirement s are 57
; j correctly stat ed, inspectable, and controllable; 2) there is adequate V accept ance and reject ion criteria and; 3) that procurement document s have been prepared, reviewed, and approved in accordance with QA pro-gram requirement s. To t he ext ent necessary, procurement document s will require contractors and subcontractors to provide an acceptable quality assurance program.
2C4 Procedures for the review, approval and issuance of documents (includ-ing procedures, inst ruct lon, specificat lor.s, and const ruct ion drawings) and changes thereto are established and described t o assure technical adequacy and inclusion of appropriate quality requirement s prior to implementation. Select ed documents are reviewed and concurred with by the project QA organizat ion for Quality Assurance related aspect s. s 2C5 Inspect ion procedures, inst ructions, or checklist s are developed by QA, reviewed by other technical organizations as necessary, and issued by l QA for the following as appropriate: i
- a. Ident ification of characterist ics and act ivit ies to be inspected.
l b. A descript ion of t he method of inspect ion. l l c. Ident ificat ion of the individuals or groups responsible for per-forming the inspection operat ion. p I V 0-115 Am. No. 57, (5/81) l
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, i i
a l
- d. . Accept ance and reject ion crit eria.
j .e. Identification of required procedures, drawings, and specificat ions i and revisions. Recording inspection or dat a recorder and the result s of the in- 57 l f. spect ion operat lon.
- g. Specifying necessary measuring and test equipment including accu- i
- racy requirement s. l 1 !
1- , I i s
;
l i i-i i l- ,
- i i f i r j
1 I 1' ' i - i i I i ! i
l t l l l \ l i h i. I
;
0-116 -Am. No. 57,.(5/81) ! l e
. _ . . . . . . _ _ _ _ _ _ _ . - - . ~ . _ . _ . ~ . _ _ . . . _ _ . . . , . . _ _ . - - , . . _ _ , _ _ _ . _ , _ . . . - - , _ _ _ _._ _ ,. .. _ __
. _. . _ _ _ _. - _ . ~ _ _ _ _ _ . . ._ _ _. .. _ _ _ ,_ _
ACNGS-PSAR I2C6 Test ~ procedures .or inst ructions are developed ' by QA, reviewed by other r c' t echnical organizations as necessary, and issued by QA for the follow- , [ ing as appropriate: > a.- The requirements ' and accept ance limit s contained in applicable E design and procurement - document s.
.b. Ins'tructions for hrforming the test.
. t
'c. Test prerequisit es such as calibrated inst rument at ion, adequat e test' equipment, and instrument at ion including their accuracy re-
.quirement s, completeness of item to be tested, suitable and con-
, , t rolled environment al condit ions, and provisions for dat a col-4- lect ion and st.orage.
- d. - Mandatory inspection hold point s for witness by ownee, cont ract or, ,
or' inspector (as required).
- e. ' Acceptance and rejection criteria.
- f. Methods of document.ing or recording test data and results. .,
- g. Provisions for assuring test prerequisites have been met . }
The QA organizat ion reviews and. document s concurrence with the proce- ! dures developed to control the following: 57' i , Calibration (technique and frequency), maint enance and cont rol o f 2C7 a. the measuring and test equipment that is used in the measurement inspection, and monitoring of st ructures, systems, and component s. 1 1 2C8 b. Cleaning, handling, st orage, packaging, and shipping of mat erials,- ) component s, and syst ems in accordance with ' design and procurement ; requirement s t o preclude damage, loss, or det eriorat ion by environ- l, mental condit ions such as temperat ure or humidity. l 2C9 c. Systems for t he indicat ion of the inspection test and operat ing status of structures, systems, and component s t hroughout fabrica-
~
tion, inst allat ion, and test . 2C10 d. Applicat ion and removal of inspection and welding st amps and st atus 3
-indicators such as tags, markings, labels, and at amps.
2Cll .e. Altering the sequence of required t ests, inspections, and ot her f' operations import ant t o safety. Such act ions should he subject to ! the same control as the original review and approval. 2Cl2 Responsibilities for QA and other organizations are described in pro-ject ' procedures for t he detinit ion and implement at ion of act ivit les for nonconformance control including the identification of individuala, ~or ! groups with authority for the disposit ion of nonconforming it ems and involvement of' Project QA in documenting concurrence to the disposi-
- t. t ion, satisfactory complet ion of the disposit ion, and correct ive 0-117 Am. No. 57, '(5/81) 4
_2____________ . . _ _ _ . _ , , . _ . . _ , .i;,_,,_,,,,,_,,._,._,,_,,_,.,,_,,_,_,_,.._my,-,,.._
ACNGS-PSAR 2C13 act ion. Procedures are established and described for ident ificat ion, 7 documentation, segregation sreview, disposit ions, and not ification to
~3 af fected organizations of nonconforming materials, parts, components,
(\-. 'l and as applicable to services (including comput er codes) if disposit ion is other than to scrap. The procedures provide identificat ion of authorized individuals for independent review of nonconformances, 2C14- including disposit ion and closeout . QA develops procedures for an ef fect ive corrective act ion program. Proposed 10 CFR 50.34 (e) (3) (iii) l (D) establishing criteria for determining QA requirement s for specific class- I es of equipment ; NRC Accept ance Guidance The QA program provides provisions t o assure that : ; 2D1 The QA organizat ion and t he necessary technical organizations partici-(2B3) pate early in the QA program definition stage to determine and ident ify the extent QA controls are to be applied to apecific structures, sys-tems, and components. This ef fort involves applying a defined graded approach to cert ain structures, systems, and components in accordance 57 wit h their importance to safety and af fects such disciplines as design, procurement, document control, inspect ion test s, special processes, records, audits and ot hers described in 10 CFR 50 Appendix B.
) 2D2 For commercial "of f-the self" it ems where specific quality assurance
[/
\, (7B4) controls appropriate for nuclear applications cannot be imposed in a pract icable manner, special quality verifiestion requirement s shall be established and described t o provide the necessary assurance of an accept able it em by the purchaser.
2D3 The scope of t he inspect ion program is described that indicates an (10A) ef fective inspection program has been established. Program procedures provide criteria for determining the accuracy requirement s of inspec-t ion equipment and criteria for determining when inspect ions are re-quired or define how and when inspections are performed. The QA organization participates in the above functions. 2 D'+ Procedures are est ablished and described with the involvement of the QA (1002) organization to identify, in pertinent document s, mandat ory inspect ion hold point a beyond which work may not proceed until inspect ed by a designated inspector. 2D5 The description of t he scope of the test control program indicat es an (11A1) ef fect ive test program has been es tablisned for test s including proof t est s prior t o inst allation and preoperational test . Program proce-dures provide criteria for determining the accuracy requirement s of test equipment and criteria for det ermining when a t est is required or i how and when t est ing act ivities are performed.
/~~x f \
v
/
0-118 Am. No. 57, (5/81)
---,e-- - y w =e -,*+-e- ,-,,-a v 3w=+earM+ t7**=vv9*we
ACNGS-?SAR 2D6 Audit data is .analy::e'd by the QA organization and the resulting reports
.f m (1881) indicating any quality problems and the effectiveness of the QA pro-
[ 1- gram, including the need for ' reaudit of deficient areas, are reported D/ 1o management for review and assessment.
Response
.2 D1 The project QA organization and the necessary t echnical organizat ion
_ part icipat e early in the QA program definit ion st age to det ermine and identify the extent QA cont rols are to be applied t o specific st ruc-tures, systems, and comronents. For it ems determined to be import ant to safety where specific QA Controls cannot be imposed in a practical manner, an evaluat ion will be made to determine special quality verifi- l cation requirement s to be applied during installation or test ing to provide the necessary assurance that the item (s) meet project re-quirements. 2D3 The project QA organization participates in the definit ion of the scope of the inspect ion program. Procedures provide criteria for determining t he accuracy requirements of inspect ion equipment and criteria for det ermining when inspect ions are required or define how and when in-spections are. performed. 57 2D4 Procedures are established to identify in pertinent documents, manda-tory inspection hold point s beyond which work may not proceeed until inspected by a designated inspector. . O 2D5 A test cont rol program will be established to include proof t ests prior k"/ to installation and preoperational t ests. Procedures provide criteria for det ermining accuracy ' requirement s of t est equipment and criteria or i determining when a test is required and how and when t esting activities I are performed.
; 2Do Audit s are conducted and the results analyzed by QA. Audit report s indicate any quality problems and t he of fectiveness of the audit ed QA Program. Reaudits of deficient areas are conducted as necessary to assume implement ation of correct iva action and recr.; rence cont rol.
Audit results are reported to management for review and assessment. , Proposed 10 CFR 50.34(e)(3)(iii) (E) establishing minimum qualification requirements for QA and QC personnel; NRC Accept ance Guidance , I l The QA program provides provisions to assure that : l l I l 2El Indoct rination, t raining, and qualification programs are established I (2D) such taat:
- a. Personnel responsible for performing quality af fecting act ivit les are inst ruct ed as to t he purpose, scope, and implementat ion of the qualiLy related manuals, inst ruct ions, and procedures.
d 0-119 Am. No. 57, (5/81)
l ACNGS-PSAR
- b. ' Personnel verifying act ivit ies af fect ing qualit y are t rained ' and qualified in the principles,' t echniques, and requirement s of the
~
f-mg ('
) act ivity being performed.
J
- c. For ' formal training and qualification programs, documentat ion includes the object ive, content of the program, attendees, and date of allendance.
- I
- d. Prc fic ienc'? test s are given to t hose personnel performing and verifying act ivities af fect ing quality, and acceptance crit eria are developed to determine if individuals are properly t rained and qualified.
- e. Cert ificate of qualificat ions clearly delineates (a) the specific funct ions personnel are qualified to perform and (b) the criteria used to qualify personnel in each funct ion, f.- Proficiency of ' personnel performing and verifying act ivit ies af-fect ing qualit y is maint ained by ret raining, re examining, and/or recert ifying as det ermined by raanagement or program commitment .
- g. The description of the t raining program provisions list ed above sat is fies t he regulat ory posit ion in Regulat ory Guide 1.58, Rev. 1.
2E2 A qualificat ion program for inspectors (including NDT personnel) is (10B2) est ablished under direction of the QA organization and documented, and i the qualificat ions and cert ificat ions of inspect ors are kept current. 57 f
\ / Response %d 2E The t raining qualification and cert ification programs are est ablished
- so that
- a. Personnel responsible for performing quality affect ing activit ies are inst ructed as to the purpose, scope, and implementat ion of the '
quality related manuals, inst ruct ions, and procedures.
- b. Personnel verifying act ivit ies af fecting quality are t rained and qualified in the principles, techniques, and requirement s of t he act ivit y being performed. ;
- c. For formal training and qualificat ion programs, document at ion includes t he object ive', cont ent of the program, attendees, and date i of at tendance. {
l
- d. Proficiency test s are given to those persor.nel performing and verifying act ivit ies affect ing quality, and acceptance criteria are developed to determine if individuals are properly trained and qualified.
- e. Certificate of qualificat ions clearly delineat es (1) the specific funct ions personnel are qualified to perform, and (2) the crit eria s used to qualify personnel in each funct ion.
\x_,n]
0-120 Am. No. 57, (5/81)
ACNGS-PS MI
- f. Proficiency of personnel performing and verifying act ivities af-fecting quality is maintained by ret raining, re examining, and/or 1 recertifying as determined by management or program commitment, j L/O i I I
\"/ g. The descript ion of the t ratning program provtstons listed above sat isfies t he regulat ory position in Regulat roy Guide 1.58, Rev. 1.
2E2 Personnel performing qualit y control functions at the site and at vendor facilit les are qualified in accordance with ANSI-N45.2.6. Proposed 10 CFR 50.34(e)(3)(iii) (F) sizing the QA staf f commensurate with it s dut ies, rerponsibilit ies, and import ance to safety. _
}
NRC Accept ance Guidance > The QA program provides provisions to assure that : i 2F1 Organizat ion chart s ident ify the "onsite" and _ "of fsite" organizationa1 (lA5) element s which function under the cognizance of the QA program (such as ; design engineering, procurement , manufacturing, construct ion, inspec- , 57 t ion, t est , inst rumentation and control, nuclear engineering, et c.), the lines of responsibility, and a description of the criteria for determining the size of the QA organization including the inspection j staff. i 2F2 The QA organizat ion is involved in est ablishing long range projected p}
; 'O
(-) work schedules and staffing of QA and QC personnel and evaluates these periodically (i.e. , monthly) t o assure they are valid or if necessary modify staffing level. f Response , i 2F1 Figures i.F.2-1 and I.F.2-2 show t he project QA organization and indi- , cat e which personnel are "onsite" and "of fsite". 'The PSAR Sec- i t ion 13.0, shows project personnel from ot her organizat ions. The et tteria for dete.~nining staf fing for the QA organization ineludes: I
- a. Establishing the number of QA/QC personnel based upon the project l schedule to ensure that personnel are available, qualified, and certified to perform quality related inspect ions and evaluations.
- b. Est ablishing the need for specially qualified QA/QC personnel based upon the schedule for activit les requiring special or unusual expert ise as far in advance of t he activity as possible.
- c. Establishing the number of QA personnel based upon the number and.
criticality of problems identified during routine act ivitles in order t o perform addit ional or supplement al inspect ions, reviews, or evaluat ions as required. t o ensure implement at ion of project requirements. I p) ( v 0-121 Am. No. 57, (5/81)
o . -. .. . . - . . ._ -. - ACNGS-PSAR ~ l
,2F2 St affing 3rojections are periodically reviewed based. upon' the project - l schedule and are re-reviewed and revised, as necessary, as the project schedul3 changes. -QA management personnel part icipate in short and-fN long range scheduling activities. St af fing levels for QA/QC are a
-A'N prime consideration in determining the level of effort for quality l 4
. related activit ies. Prior to allowing quality related activit ies to .be conduct ed,' adequate numbers of qualified .QA/AC personnel must ' be avail-i able. Adequate QA/QC staffing must be.available to prevent QA/QC personnel from being required to perform -inspections or evaluations-
- without .adquat e preparation. time or under pressure to complete inspec- '
' tions within a scheduled time period. Adequate QA/QC staf f must be available to allow for prompt closeout of opern nonconformances and i ,
- _ proper followup to. ensure correct ive action has been t aken. l I Proposed 10 CFR 50.34(e)(3)(iii)_ l (G)- est ablishing procedures _ for maintenance of "as-built" document at ion;- j i NRC Acceptance Guidance ) The QA' program provides provisions to assure that: 2G1 The scope of the document control program is described, and the types 57 (6Al) of controlled documents are identified. As a minimum, controlled documents include: As-built documents. 2G2 . ' Procedures are established and described to provide for the preparation (6Cl) of as-built drawings.and related documentation in a timely manner to accurat ely reflect the act ual plant design. l-(v) Response ; 2G1 The PSAR, Section 17.1.OA, includes "As-Built" drawings in the document + 4 2G2 cont rol system. Project procedures will be developed to ensure that ! drawings are provided to indicate the es-built configurat ion. The l as-built drawings will st and alone and delineate actual locat ion - t 4 elevation, azimuth, etc.; actual component identification or numbering; , i dimensions and ot her relevant- information. When changes occur subse-
~ quent to issuance of as-built drawings, procedures will require a re review and re-issue of the drawings.
i Propored 10 CFR 50.34(e)(3)(iii) (H) providing a QA role in design and analysis activitles. NRC Accept ance Guidance I The QA program provides provisions to assure that: 2H1 Procedures are established and described requiring a documented check
.l_
(3El) t o verify the dimensional accuracy and completeness of design drawings and specifications.
. /"
i w O-122~ Am. No. 57,- (5/81) l l l l
ACNGS-PSAR
,2H2 LProcedures are established and described requiring that design drawings (3E2) and specificat ions be reviewed by the QA organizat ion or ot her indivi- /g duals knowledgeable and qualified in QA/QC techniques to assure?that
( / the documents are prepared, reveiwed, 'and approved in accordance with-
-company procedures and that the documents contain the necessary quality
- assurance requirements such as inspect ion and test requirement s, ac-ceptance requirements, and' t he extent of documenting inspect ion and t est result s.
Response , 2H1 Procedures require a ' document ed check to ensure the dimensional ac-
-curacy (including tolerance for accept / reject criteria and inspectabil- :57 ity) and the completeness of the drawings and specificat ions. QC inspections of qualit y relat ed act ivit ies will be conduct ed using procedures or inspection checklist s developed from the engineering l specifications and drawings for the system, component , or st ruct ure.
- _ Procedures are established to require that design drawings and specifi-
~
2H2 cations be. reviewed by individuals knowledgeable'and qualified in~QA/QC techniques to assure that the documents are prepared, reviewed, and approved in accordance with procedures and that document s contain the necessary QA requirements such as inspection and test requirements, accept ance requirement s, and the extent of document ing inspection and
-Lest results.
y 1 i i i t M O m 0-123 An No. 57, (5/81)
- - - - - - - - - _ . - - - - , a. .a, a --w_ _ a .--_- - ,
d O I l O l I f O t
t I l i I
's OPFRATIONS QA MANAGER A _
Y G NDE/ISI PROJECT e OPERATIONS QA QUALITY ASSURAN SUPEriVISOR GENERAL SUPERVI! A PROJECT QA PLANT QA SUPERVISOR SUPERVISOR j MECHANICAL
/
PROJECT QA SUPERVISOR CIVIL I HOME OFFICE SITE PROJECT QA j SUPERVISOR t ! ELECTRICAL
I ACNGS - PSAR EXECUTIVE VICE PRESIDENT A MANAGER QUALITY ASSURANCE A l ROJECT HOUSTON OUALITY ASSURANCE OA MANAGER , MANAGER i
/ A i
l i l PROCUREMENT SUPERVISOR tE PROJECT QA - AUDITS & TECHNICAL hR SUPERVISOR SERVICES i / / A l SUPERVISOR SUPERVISOR QUALITY VENDOR SYSTEMS SURVEILLANCE l AM. NO. 57, (5/811 1 HOUSTON LIGHTING & POWER COMPANY 7 l / Allens Creek Nuclear Generating Station Unit 1 HL&P PROJECT QA ORGANIZATION FIGURE 1.F.2-1
o i k CORPt QUALITY PROGRAM SITE MANAGER QUALITY ASSURANCE RECORDS SITE CONTROL SUPERVISOR SUPERVISOR
$ + REVIEW 7F FIELD CHANGE
- RECEIPT, INDEX, ETC. QA
- Ci REQUEbTS RECORDS Si
- REVIEW OF DESIGN CHANGE
- ESTABLISH AND MAINTAIN W NOTICES PROJECT SITE FILE
- REVIEW AND APPROVE FINAL a REVIEW RECORDS RECEIVED NDE FROM FIELD QC/QA
- ISSUE QA PLANS FOR SITE
- PROCESS DEFICIENCY PURCHASE ORDERS REPORTS
- DEVELOP GUIDELINES FOR QUALITY PROCEDURES AND INSTRUCTIONS
- DEVELOP SITE QA/QC
, PROCEDURES & CHECKLISTS
- REVIEW AND APPROVE NON-CONFORMANCE REPORTS
- SITE PURCHASE ORDER REVIEW
? \
3 HOME OFFICE SITE
- l ACNGS - PSAR ETASCO IRATE QA ORGANIZATION I
PROJECT
' QA ENGINEER ! A
- SPECIFICATION REVIEW
- PURCHASE ORDER REVIEW g
- DCN REVIEW
- VENDOR SURVEILLANCE COORDINATION I
e h VENDOR SITE QUALITY CONTROL SURVEILLANCE AUDIT SITE COORDINATOR SUPERVISOR SUPERVISOR ) ORDINATE VENDOR
- PLAN AND PERFORM SITE
- PERFORM INSPECTIONS
' RVEILLANCE ACTIVITIES AUDITS
- ESTABLISH INSPECTION HOLD ITH RECEIPT INSPECTION
- REPORT AUDIT RESULTS POINTS
- TREND ANALYSIS
- WITNESS TESTS
- PERFORM NDE
- PERFORM SOILS AND CONCRETE TESTS
- REVIEW MANUFACTURE AND CONTRACTOR RECORD PACKAGES
- WITNESS HANDLING OPERATIONS
- PERFORM CORRECTIVE ACTION VERIFICATIONS
- VERIFY DOCUMENT CONTROL PROGRAM
- MONITOR WELDING QUALIFICATIONS
- CONTROL OF MTGU AM. NO. 57, (5/81) ,
HOUSTON LIGHTING & POWER COMPANY Allens Creek Nuclear Generating Station Unit 1 EBASCO PROJECT QA ORGANIZATION FIGURE I.F.2-2
1
-ACNCS-PSAR
' ITEM II.B'.8 RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS
-[ 4 - : NUREG 0718 REQUIREMENT- .r A . :
" Applicant . shall:
(2)' include' provisions in the containment design for one or more de'dicated penetrations, equivalent in size to a single 3-foot diameter opening. This shall.be done-in order not to preclude the-installation of systems to prevent containment failure, such as filtered vented containment systems." RESEONSE A,dedicat'ed three (3) foot diameter containment penetration _ assembly below the spring:line will be incorporated into the plant design. The' space provision to reflect the penetration' assembly is available in the present containment
~
design. The dedicated 3-foot diameter containment penetration assembly will consist of a' capped penetration in the Steel Containment and the Shield Building. .The Elevator Access Room is dedicated for the containment pene-tration assembly, and allows space for a future outboard isolation. valve, 57 (s' pace provision only). Space inside the Containment is dedicated for the containment penetration assembly, and a future inboard isolation valve, if re-quired (space provision only). The penetration assembly and the welded caps will be designed in accordance with the requirements 1of the ASME Section III, Subsection NE, and seismic Category I. The penetration assembly 'and the welded caps shall be protected
. from natural phenomena in the same manner as the. Containment Steel Shell.
Periodic tests and inspection'will be performed in accordance with the normal plant operation procedure. Test connections as per 10CFR50, Appendix J, shall be provided. A' detailed description of this provision is located in Section 3.8.2.1.2. q) 0-126 Am. No. 57, (5/81)
/
~ . . - -.- . . = -- ... - .- - , . _.
.ACNGS-PSAR-
, ITEM II.B.8L RULEMAKING PROCEEDING ON DEGRADED CORE ACCIDENTS I : NUREG-0718 REQUIREMENT
" Applicant 1 shall: -l i
(4') provide preliminary design.information at a level consistent with that j normally required at the construction permit stage of review sufficient-
.to demonstrate that:
i (a) Containment integrity will be maintained (i.e. , for steel contain-i ments by meeting the requirements of the ASME Boiler and Pressure
; Vessel Code, Division 1, Subsubarticle NE-3220, Service Level C Limits, except that evaluation of instability is not. required, considering pressure and dead load.alone. For concrete containments 4 by meeting the requirements of the ASME Boiler and Pressure Vessel Code, Division 2, Subsubarticle CC-3720, Factored Load Category, considering pressure and dead. load alone) during an accident that 1 . releases hydrogen generated from 100% feel clad metal water reaction accompanied by either hydrogen burning or the added pressure from post-accident inerting assuming carbon dioxide is the inerting i agent, depending' upon which option is chosen for control of hydro -
'- gen. As a minimum, the: specific code requirements set forth above. i appropriate for each type of containment will be met for a combina-tion of dead load and an internal pressure of 45 psig. Modest deviations from these criteria will be considered by the staff, if - good cause is shown by an applicant. Systems necessary to ensure 57 .!- containment. integrity shall also be demonstrated to perform their [ function under these ' conditions." l RESPONSE. A preliminary .evaluttion has been done using the current (including
modifications described before ACRS on 2/6/81) configuration of the 1 containment vessel and its anchorage into the foundation mat. The evaluation results show that containment integrity will be maintained by meeting the j requirements of the ASME Boiler and Pressure Vessel Code, Division 1, Sub-4 subarticle NE-3220, Service Level C Limits for containment vessel and Division 2, Subsubarticle CC-3420, Factored Load Category for concrete anchorage (con-sidering pressure (45 psig) and dead load alone) during an accident that i . releases hydrogen generated from 100% active fuel clad metal-water reaction j and'the pressure from post-accident inerting assuming carbon dioxide is the ! .inerting agent. Subsubarticle CC-3720 is not applicable since it addresses concrete' containments, i i In order to assure containme'nt integrity for the required conditions, the j - applicant has revised the commitments found in Section 3.8 as follows: l (i) Additional loads have been defined in Sections 3.8.2.3.1 and 3.8.5.3.1. ! These loads were estat'ished in the following manner: for the accident l case the static pressure was taken as the greater of the maximum pressure , determined in the accident analysis described later in this response and ; i the minimum.45 psig of this requirement. The temperature for this case L D was taken from the accident-analysis. [ ,. Y
i 0-127 Am. No. 57, (5/81) l
.r,._m.,- .._,_ . . _ , , .-.c_.,,,, ...,,,,,,..,m -..,,,_,.,,,..,.m,-,my,,m_____ ,_,m,.,_%,.,e
ACNGS-PSAR
. (ii) An additional load combination has been specified in Sections '3.8.2.3.2
-~c and 3.8.5.3.2. This load combination includes the consideration of dead
'( )~ weight, pressure, and temperature as given in (a) above' .
\_/
A complete description of the containment vessel design and analysis procedures is provided in Section 3.8.2.4. In summary, the containment vessel will be designed in accordance with the rules in ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NE for Class MC Components and will consider the loading conditions listed in Section 3.8.2.3.2. The vessel shell will be analyzed using the basic membrane equations for thin shells. The anchorsge transition region will be analyzed for the same loading conditions indicated above. The temperature stresses will be analyzed in critical areas, with the rules of ASME, Section III governing the treatment of these streeses as either secondary or local. A complete -description of the design and analysis procedures for the concrete foundation of the containment vessel is provided in Section 3.8.5.4. In sum-mary, the design and analysis of the anchorage region concrete will be in acrordance with the appropriate sections of CC-3100 to CC-3500 of the ACI-ASME 57 Code, Section III, Division 2 using the load conditions listed in Section 3.8.5.3.2. The analysis will be performed by conventional stiffness /flexi-bility computerized methods using proven industry accepted computer programs to determine the internal stresses and deformations in the anchorage region. Adequate reinforcing will be provided to resist the forces and moments resulting from the different loading conditions. The thermal stresses will be determined by the use of temperature gradients through the thickness of the anchorage region for the different loading conditions. The rules in ACE-ASME l [' ) Code governing the treatment of thermal stresses as secondary stresses will be j \ ,,) used. l For the additional accident loading condition shown in Section 3.8.2.3.2, the acceptance criteria for the containment vessel will be based on the allowable stresses defined in Table 3.8-1 for abnormal extreme load combinations (integral and continuous). These allowable stresses are based on the requirements of the ASME Code, Section III, Division 1, Subsection NE and are as follows: i-General membrane stresses: the greater of 1.2 S, or S y Local membrane stresses: the greater of 1.8 S, or 1.5 S y Bending + local membrane stresses: the greater of 1.8 S ,or 1.5 S y The anchorage concrete will be designed for the effects of the additional accident loading condition shown in Section 3.8.5.3.2 utilizing the allowable stresses and strains for factored loads, based on the requirements of the , ACI-ASME Code. For the additional accident loading condition the stress and t' strain limits using Factored Load Category will be as follows: Concrete compression, sh(ar, torsion, bearing; use cllowable stresses for factored loads as specified in the ACI-ASME Code, Paragraph CC-3421. l n t_- l I-l 0-128 Am. No. 57, (5 /81)
, ,, , , . . , , - , - , ..,-e .-
ACNGS-PSAR Reinforcing steel tension and compression: use allowable stresses and f--Kl strains for factored loads as specified-in the ACI-ASME Code, Paragraph ( ) CC-3422.
\~J ACCIDENT ANALYSIS The assumptions below were used to calculate containment temperature and pressure following an accident which releases an amount of hydrogen equivalent to 100% of the active fuel clad metal water reaction. It should be recognized that like the assumed 100% reaction, these assumptions are not intended to have a mechanistic basis but are prepared as a means to define a reasonable method of determining the containment pressure and temperature response. They should be considered in the context of the overall objective of providing assurance that an extensive metal water reaction can be accommodated in the absence of definitive accident specifications. In the future, pending rulemaking on degraded core may provide those specifications.
Initiating Event: a transient-event (such as loss of feedwater) followed by reactor scram and containment isolation. Assumptions 57 Inadequate core cooling results in decreasing Reactor Pressure Vessel
-(RPV) water level Operator initiates post-inerting system timer on reactor low level 1 /' 'N -
CO2 discharge is complete 45 minutes after event initiation. ('^'/ - core heatup results in 100% MWR of active fuel clad the energy equivalent to decay heat and 100% MWR is transported to the
. suppression pool temperature differential between pool and containment airspace at 42 psig is 50F the RPV sensible heat released upon recovery of a low pressure ECCS system is transported to the pool The mass of hydrogen equivalent to 100% MWR is transported to the containment air space Results maximun containment pressure is 42 psig maximum suppression pool temperature is 195F (containment airspace temperature 145F)
O, v 1 0-129 Am. No. 57, (5/81)
. . . = ___ - -. .-. -.
ACNGS-PSAR The calculated containment temperature and pressure responses using these
- (qt assumptions are reported in Section 6.2.1.3.4. These analyses provide assur-ance that the use of a 45 psig post accident pressure for the containment V vessel and anchorage design (see Sections 3.8.2.3.2 and 3.8.5.3.2) will be adequate to maintain containment integrity.
A preliminary system level review of the functions necessary to maintain con-tainment integrity during the postulated event described above and the systems which perform these functions has been made. The approach taken, in the ab-sence of a definitive accident specifications, was to identify, in general, those functions and the related systems which support and maintain containment integrity. Assuring that those functions are maintained provides a conserva-tive approach to maintaining containment integrity. The results- of this re-view are shown in Table II.B.8.4-1, 57 Preliminary review of those portions of the identified systems which must function and are environmentally exposed to this accident indicates that it is feasible to demonstrate necessary functions will be performed. It is HL&P's intent to finalize this review and then perform a detailed analysis of , individual subsystems and components and demonstrate that required functions will be assured considering environmental conditions and required duration of operation. It should be recognized in this process that modifications to Table II.B.8.4-1 may occur. NUREG 0718 REQUIREMENT (4) (b) "The containment and associated systems will provide reasonable assurance that uniformly-distributed hydrogen concentrations do not v/ exceed 10% during and following an accident that releases an equi-valent amount of hydrogen as would be generated from a 100% fuel clad metal-water reaction, or that the post accident atmosphere will not support hydrogen combustion."
RESPONSE
(4) (b) The containment and associated systems are chosen to ensure that the post-accident atmosphere will not support hydrogen combustion. The concentration of CO2 necessary to ensure non-flan =nability of a hydrogen-air-C02 mixture is based upon Bureau of Mines test data. The chosen value of 61% CO2 by volume is considered conservative for the following reasons:
\v/
0-130 Am. No. 57, (5/81) 1 l
ACNGS-PSAR I l
- 1) Ignition Source The flammability limit is dependcnt on the type of ignition used. For
,' example, ignition tests of lean mixtures have shown that spark ignition, the most likely ignition source in the containment is much less efficient than a flame. The Bureau of Mines tests utilized a flame ignition source. Coupled with the fact that the amount of inerting egent required is the greatest for lean mixtures, this indicates that a lower concentration of CO2 would be required to preclude ignition by a spark source.
- 2) Effects of Water Vapor and Initial Containment Temperature The required quantity of CO2 is based upon dry air at 70F in the containment. Higher initial containment temperature and any water vapor content reduces the quantity of air present in the containment, and thus increases the resulting volume percent of CO2 . In addition, water vapor has the same inerting effect as CO 2, requiring approximately 58%
to inert any mixture of air and hydrogen. It can be expected that the ef fects will be additive.
- 3) Allowances for CO2 Solubility in Water Approximately 9% additional CO2 mass, beyond that required to reach 61% 57 by volume in the air space, is injected to account for eventual absorp-tion of CO2 into the reactor and containment water masses. Init ially ,
, _x this quantity would be available as additional margin. As the CO2
/ T dissolves into the water, the water vapor content of the air would be ( ,) expected to increase as the suppression pool temperature increases during the course of the accident.
- 4) Maximum CO2 Required at Lower Hydrogen Concer.tration The quantity of CO2 is determined by the amount required to inert a hydrogen-air-C02 mixture containing 10% hydrogen. With higher or lower volume percents of hydrogen, less CO2 is required. Approximately 52%
CO2 is required to inert the mixture containing hydrogen from 100% MWR. It is likely that a 10% hydrogen concentration would occur early in the event while the CO2 allowance for water absorption was still present in the air space. NUREG 0718 REQUIREMENT (4) (c) "The facility design will provide reasonable assurance that, based on a 100% fuel clad metal water reaction, combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features." g~s 0-131 Am. No. 57, (5 /81)
ACNGS-PSAR
RESPONSE
(O). (4) .(:) The design objective adopted is to mix sufficient CO2 with all of
~ the atmosphere in the containment and drywell .so that burning can-not occur regardless of the location and concentration of the hydrogen. To meet this _ objective it. is required that the CO2 be delivered in a controlled manner to selected locations throughout the containment and then well mixed with the surroundings.
! Delivery will be achieved by a distribution system consisting of piping and manifolds extending from the storage tanks, throughout the containment and terminating at nozzles at :the selected locations. The carbon dioxide stored in tanks is a liquified gas at saturation pressure and temperature. As the liquid is discharged, the. tank pressure begins to drop and the remaining liquid boils as some of the liquid is evaporated to maintain saturation equilibrium. The vapor thus generated maintains a high prescure in
- . the corage container as the bulk of the liquid is discharged into -
the pipelin(. t j .The liquid in the pipeline continues to boil in response to any drop 57 I Lin pressure. This results in a mixture. of liquid and vapor which, - in a properly-sized piping system, flows at a velocity that assures high turbulence for continuous mixing of the liquid and vapor por-tions, i.e., two phase flow. {- Because the volume ratio of vapor to liquid increases as the n/ N, pressure drops, the average density decreases and the velocity, throughout the pipe increases. The rate of pressure drop per. foot of pipe thus becomes greater as the mixture proceeds throughout the piping at reducing pressures. This is due, in part, to greater frictional 'oss at higher velocities and, in part, to the' energy j required to increase the velocity head of the mixture. _ 7 1 The equations which make it possible to accurately calculate the j pressure drop for two phase flow also take into account the effect of changing densities, based on thermodynamic data for liquified carbon dioxide in question. They also account for'the changing
- velocity head in the piping. This method of calculation has been adopted in NFPA 12, " Carbon Dioxide Extinguishing Systems." .
l i Finally, the discharge rate from individual nozzles is based on test l data relating discharge rate to terminal pressure conditions. {
! As might be expected, the calculations are too complex for practical
- manual' solutions. The . calculations are, therefore, performed by
- execution of compute.r programs previously developed for this purpose.
I.
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J L ACNGS-PSAR
' Using these techniques, the distribution system is optimized with , respect to piping sizes and nozzle orifices to achieve the desired
- -7 ^'s'E delivery at each of the selected locations. D Mixing with the surrounding air to assure even distribution is y achieved by using nozzles which release the CO2 at high. velocity directly into the air resulting in an energetic expanding jet in .I which rapid momentum exchange entrains and mixes into the jet large
' l amounts of the surrounding air. In addition to the mixing within the jet, ~ the momentum of the jet is used to force a general circula-
- tion pattern within the containment. The two nozzle groups which l L discharge downward into two of the four open quadrants from the l upper level to the bottom of the containment - force the general - I circulation c'own these paths. This flow then moves circumferentially around the containment and returns upward through the opposite qusdrants, thus closing the flow paths into recirculation locos. i Upward flow is aided by' the upward momentum of the discharge from the nozzle groups in the upper portion of each of the upward flow
; _ paths. In addition, these nozzle groups assure that air in the top
{ of the containment receives CO2 and is forced to participate in 57 the general circulation. An examination of both the containment drawings and a detailed model l of the containment was performed to identify locations which might not particirate effectively in the general circulation described j '\ . above. A nozzle will be placed in each suspect location. A similar process of examination and nozzle location selection was performed for the drywell. However, for the drywell it was not ~ found necessary to force a general circulation. The turbulence caused by the jets will assure mixing in this relatively small volume. The adequacy _of mixing and nozzle placement is based on experience ' in the design and testing of many industrial CO2 fire extinguishing systems. Howeve r, it may be useful to consider the following highly simplified illustrations of the energy and momentum , available to force the mixing process. For example, the discharge i velocity at the nozzle will be in excese of 500 feet per second. At ! this velocity, the energy that must be diasipated in the form of l turbulence and air movement is nearly 4000 foot pounds per pound of CO 2 . At the design discharge rate of over 300 pounds per second, this is equivalent to more than 2000 horsepower. Also, based on the principle of conservation of momentum, each pound of carbon dioxide discharged is capable of accelerating 15 pounds of air to 30 feet , per second. Velocities of this order would recirculate the containment air several times per minute. I F g- 1 s t l 0-133 Am. No. 57, (5/81)
ACNGS-PSAR The calculational ' methods and design skills used in this ' design' have
-)% - .been applied in the design of fire suppression. systems. Full de-
/ i monstration tests have been performed on some of these systems.
/ Such tests consist of actuation of the completed system and measur- t
,' ing, as a function of time, the concentration of CO2 at points within the volume being flooded. These test results consistently i show rapid and thorough mixing.
Further, test experience has shown that lengthening the discharge i period markedly promotes mixing. Although tests have shown ef fec-tive mixing can be achieved with a discharge period as brief as 10 seconds, discharge periods of from one to four minutes are more typical. The discharge period used in this design approaches 15 minutes and allows more than adequate time for mixing. Finally, the mixing of gases is permanent. That is, once the carbon
. dioxide is mixed with the air, it will not stratify, settle, or otherwise separate from the air.
NUREG 0718 REQUIREMENT (4). (d) . "If the option chosen for hydrogen control is post accident inert- 57 ing: (a) Containment structure loadings produced by an inadvertent full inerting (assuming carbon dioxide), but not including seismic
. or design basis accident loadings will not produce stresses in steel
containments in excess of the limits set forth in the ASME Boiler , ' and Pressure Vessel Code, Division 1, Subsubarticle NE-3220, Service Level A Limits, except that evaluation.of instability is not re-quired (for concrete containments the loadings specified above will , not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Division 2,
.Subsubarticle CC-3720, Service Load Category), (b) A pressure test, which is required, of the containments at 1.10 and 1.15 times (for steel and concrete containments, respectively), the pressure calcu-lated to result from carbon dioxide inerting can be safely con-ducted, (c) Inadvertent full inerting of the containment can be i safely _ accommodated during plant operation."
!. RESPONSE (4) (d) A preliminary evaluation has been done using the current (including , modifications . described before ACRS on 2/6/81) configuration of the l- containment -vessel and its anchorage into the foundation mat. The f evaluation results show that containment structure loadings produced I by an inadvertent full inerting (assuming carbon dioxide), but not includinr seismic or design basis accident loadings, will not produce stresses in the containment in excess of the limits set forth in the ASME Boiler and Pressure Vessel code, Division 1, Subsubarticle NE-3220, Service Level A Limits for the containment vessel and Division 2, Subsuoarticle CC-3430, Service Load Category for the concrete anchorage. Subsubarticle CC-3720 is not applicable since it covers concrete containments. l l l l l 0-134 Am. No. 57, (5/81) I
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ACNGS-PSAR
.A pressure test of the containment at 1.10 times the pressure re-sulting from carbon dioxide inerting can be safely conducted.
t In order to assure containment. integrity for the required con-ditions, the applicant has revised the commitments found in Section 3.8 as follows:
- a. Additional loads have been defined in Sections 3.8.2.3.1 and 3.8.5.3.1. These loads were established as the maximum temperature and pressure as calculated in the inadvertent actuation analysis which is described later in this response,
- b. Additional load combinations have been specified in Sections 3.8.2.3.2 and 3.8.5.3.2 These load combinations include the consideration of dead weight, pressure and temperature for the inadvertent actuation event.
l A complete description of tha containment vessel and its concrete anchorage design and analysis procedures is provided in Sections 57 3.8.2.4 and 3.8.5.4 and is summarized in 4 (a) ab ve. i For the additional inadvertent actuation and testing loading con-ditions shown in Section 3.8.2.3.2, the containment vessel accept-i an'ce criteria will be based on the allowable stresses defined in l Table 3.8-1 for normal load combinations and for test load combination, respectively. These allowable stresses are based on the requirements of the ASME Code, Section III, Division 1, Sub-
/'"N section NE and are as follows:
I t Inadvertent Inerting (service level A): General membrane stresses: S,
- Local membrane stresses: 1.5 S ,
Bending + local membrane stresses: 1.5 S ,
- Primary + secondary stresses: 3 S, Test Condition (Test allowables):
- General membrane stresses: 0.9 S y
- Local membrane stresses: 1.25 S y
- Bending + local membrane stresses: 1.25 S y Primary + secondary stresses: 3 S,
\
w/ 0-135 Am. No. 57, (5/81) l l L_--- _ _ _ . _ _ _ _ _ _ - _ _
ACNGS-PSAR The anchorage concrete will be designed for the effects of the inadvertent e actuation and testing loading conditions shown in Section 3.8.5.3.2 utilizing ( the allowable stresses and strains for service loads and test-loads, based on the requirements of the ACI-ASME Code as follows: Incdvertent Inerting Condition (using Service Load Category)
- -
Concrete compression, shear, torsion, bearing: use allowable stresses for service loads as specified in the ACI-ASME Code, Paragraph CC-3431. Reinforcing steel tension and compression: use' allowable stresses and 4' strains for service loads as specified in the ACI-ASME Code, Paragraph CC-3432. Test Condition (using Test Allowables): Concrete compression and bearing: use allowable stresses for service loads as specified in the ACI-ASME Code, Paragraph CC-3431. Concrete shear and torsion: use allowable stresses for service loads, increased by 33-1/3%, as specified in the ACI-ASME Code, Paragraph CC-3431. 57 Reinforcing steel tension and compression: use allowable stresses and , strains for service loads increased by 33-1/3% as specified in the ! ACI-ASME Code. Paragraph CC-3432. INADVERTENT ACTUATION ANALYSIS lO The event sequence given below wtsa developed to provide a reasonable basis upon which to calculate the containment pressure and temperature response so that containment loads can be evaluated for inadvertent operation of the CO2 system. Initiating Event - CO2 system is inadvertently actuated and begins injection of CO2 into the containment. l Assumptions 3 A reactor scram and. containment isolation (but not MSIV closure) occurs I when drywell pressure exceeds 2 psig.
A full charge of inerting agent (CO2) is injected into the containment. ; Reactor heat removal and depressurization to approximately 150 psig is i accomplished by steam flow to the main condenser. Safety relief valves do not open. HPCS and the' feedwater system supply water to the reactor until both are tripped by high water level or level increase is terminated by operator action. HPCS is then secured and feedwater is restarted if necessary and used to n.cintain reactor water level. i I-0-136 Am. No. 57, (5/81) .
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ACNGS-PSAR
-. The operator promptly initiates a 100F/ hour shutdown of the reactor using the turbine bypass valves. l s I L' -
Containment Sprays activate and operate until containment pressure is greater than 9 1,sig (10 minutes af ter the high drywell pressure signal). Upper Pool dump occurs 30 minutes af ter the high drywell pressure signal. Results:
- 1) Pool Temperature = 950F
- 2) Containment Pressure = 25 psig i Certain systems will be required to achieve cold shutdown following an inad- 57 vertent actuation event. A preliminary review indicates that the following systems are required and may be affected by the resulting containment condi-tions: Reactor Protection System, Control Rod Drive, Rod Control and Information System (Rod Position), RHR and containment purge. The review will be completed in detail; however, the preliminary results show that it is feasible to demonstrate a safe shutdown.
NUREG 0718 REQUIREMENT (4) (e) "If the option chosen for hydrogen control is a distributed ignition system, equipment necessary for achieving and maintaining safe shut-down of the plant shall be designed to perform its function during and af ter being exposed to the environmental conditions created by activation of the distributed ignition system."
RESPONSE
(4) (e) Since the distributed ignition system was not chosen for hydrogen control, this requirement is not applicable. If this system is sub-sequently decided on as the final design for hydrogen control, the required analyses will be performed. 1 l A !
/
v)
\ l \
O-137 Am. No. 57, (5/81)
- c. ACNGS-PSAR TABLE II.B.8.4-1
[
\
\m SYSTEMS NECESSARY TO ENSURE CONTAINMENT INTEGRTTY DURING A POSTULATED ACCIDENT EVENT 4 Function' System i
Reactivity Control Reactor Protection (Trip System) Control Rod Drive (Scram function) 4 Reactor Depressurization Automatic Depressurization System ADS Instrument Air (Accumulators) Residual Heat Removal- RHR System (Containment Spray, Suppression Pool Cooling Modes)~ j Reactor Water Makeups 1-57 High Press High Pressure Cc;e Spray System Low Press RHR System (LPCI Mode) Low Pressure Core Spray System l Containment Isolatiot. Containment and Reactor Isolation Systems Main Steam Leakage Control System Special Containment Functions Containment Vacuum Kelief Drywell Vacuum Relief Suppression Pool Maxeup System Combustible Gas Control Post-Accident Inerting System Post-Accident Monitoring Containment Atmosphere Monitor Post-Accident Sampling System Electrical Power Supply and Standby Power Distribution 4.16 Kv - Safety 480V ac Safety 120/208V ac Safety 277/480V ac Safety 120V ac RPS
- 125V de Safety Equipment ana Space Cooling Essential Services Cooling Water System ECCS Area fan cooler s
C~ 3 0-138 Am. No. 57, (5/81)
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II.E.4.1 ACNGS-PSAR
/N ITEM II.E.4.1 DEDICATED' PENETRATION ,
i t NUREG 0718 REQUIREMENT
" Applicants for plant designs with external hydrogen recombiners shall modify -
their applications as necessary to include redundant dedicated containment penetrations so that the recombiner systems can be connected to the containment atmosphere without violating single-failure criteria, such as having to open ' large containment purging ducts or otherwise jeopardizing the containment func- 57 tion. Applicants shall submit, prior to the issuance of cou.truction permits or the manufacturing license, a detailed explanation of how the requirements will be met in order to provide reasonable assurance that the requirements will i be implemented properly."
RESPONSE
The ACNGS design utilizes thermal recombiners which are internal to containment and have no associated mechanical penetrations. As such, this item is not ap-plicable. l f) m i nm 0-139 Am. No. 57, (5/81)
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-ACNGS-PSAR ITEM II.J.3.1 ORGANIZATION AND STAFFING TO OVERSEE DESIGN AND CONSTRUCTION p
NUREG 3718 REC.UIREMENT (% Applicants shall describe their pregram for the management oversight of design and construction activities. Specific items to be addressed include: (1) the organizational and management structure which is singularly responsible for the direction of the design and construction of the proposed plant, (2) technical resources which are directed by the utility organization, (3) details of the interaction of design and construction within the utility organization and the manner by which the utility will assure close' integration of the architect engineer and nuclear steam supply vendor, (4) proposed pro-cedures for handling the transition to operation, and (5) the degree of top level management oversight and technical contt ol to be exercised by the utility during design and construction, including the preparation and imple-mentation of procedures necessary to guide the ef fort. Draft NUREG-0731, " Guidelines for Utility Management Structure and Technical Resources" is the keystone for similar development of guidelines for this task. Therefore, the principal applicable elements of NUREG-0731 shall be used by CP and ML applicants in addressing this task. Applicants shall submit detailed information in order to provide reasonable assurance that the requirements will be implemented properly prior to issuance of the construction permits or manufacturing license.
RESPONSE
57
.] (1) Organization and Management Structure (a) General HL&P has always been aware that utility involvement in the design and construction phase of a Project enhances safety, reliability, and operability. Shortly af ter HL&P became involved in Nuclear and Coal programs, both of which were new to the Company, HL&P began utilizing the Project Management Organization (PMO) concept in our staffing structure. The PMO ensures that the decision making process is integrated during design, construction and start up. The Manager, Allens Creek, has full authority to implement Project goals, and all contractors and HL&P team members, except Quality Assurance, are under his direction.
The major organizations involved in the ACNGS are: a) Houston Lighting & Power Company (HL&P) As the Applicant, HL&P has and retains the overall responsi-bility for the engineering, design, procurcment, fabrication. l construction, preoperational testing,. operation and QA ac-tivities for the ACNGS. i s a v 0-140 Am. No. 57, (5/81)
ACNGS-PSAR HL&P will audit the activities of Ebasco, GE, consultants and n other contrectors to assure that their QA programs are imple-mented and have suf ficient authority and organizational freedom (V ) to be effectively implemented. HL&P will perform surveillance of the activities of Ebasco, GE, consultants and other contractors during the manufacturing, fabrication and construction of the ACNGS. b) Ebasco Services Incorporated (Ebasco) As the Architect - Engineer, Ebasco is delegated the responsibi-lity to provide HL&P with engineering, design,.procurements and QA services. As the constructor, Ebasco is delegated the re-sponsibility to provide 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 n 6) site QA/QC activities Figure 17.1.lB-1(a) is an organization chart showing the Ebasco QA organization for the Allecs Creek Project and the functions or activities that they will perform.
c) General Electric Company (GE) 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, f abrication and QA services for the NSSS and Nuclear Fuel, GE has the responsibi-lity to provide an acceptable QA program to HL&P for the activi-ties that have been delegated to GE. These delegated activities include the following:
- 1) design and engineering activities 2) 3). procurement activities fabrication activities
- 4) vendor surveillance activities
- 5) QA activities The functional interf aces between the organizations are shown on Figure II.J.3.1-1.
O (m 0-141 Am. No. 57, (5/81)
II.J.3.1 ACNGS-PSAR
* / The HL&P organization and management structure and details of the scope of work and division of responsibilities can be found in the PSAR - Chapter 13. Quality Assurance responsibilities and scope of work are fully described in the PSAR - Chapter 17.
Following the TMI incident, HL&P began to reascess our Management ^ structure and personnel qualifications as did most of the nuclear industry. As a result, the following changes have or will be implemented:
- corporate reorganization such that the Executive Vice-President reporting to the President has only nuclear responsibilities.
- b rought into executive management a highly experienced individual in the nuclear field to manage the design and construction of our 57 nuclear plants.
- establish an active HL&P engineering organization onsite during construction to assure problems are resolved thoroughly and correctly and to provide continuity of design through start up and operation.
Each of these points are discussed below. (a) Executive Vice President [ ) iu insure availability of HL&P Corporate Management to be aware and involved in our nuclear projects, a major change was made in
\s_,/
mid-1980. Nuclear and Fossil fueled operations were separated at the highest level in the company and an Executive Vice President was named having total, but only nuclear responsibilities. These nuclear responsibil3 ties include Engineering, Construction, Operaticas, Fuel Management, and Quality Assurance. The change will ensure that the nuclear programs will receive thorough and timely ' attention and that adequate priority can be placed on Company resources to resolve any problems. I (b) Vice President - Nuclear Engineering and Construction Another organizational change designed to bring nuclear design and construction experience into Corporate Management was the t establishment of a new position - Vice President, Nuclear Engineering and Construction. This position, reporting to the Executive Vice President, was filled in October 1980. [N
\v) 0-142 Am. No. 57, (5/81)
ACNGS-PSAR He has 25 years of nuclear experience and has been in responsible gsg_ charge of many aspects of nuclear design and construction including ( ) Project Manager, Cons?.ruction Manager, and Chief Engineer. Prior to
\s_ / joining HL&P, he was Vice President, Cons t ruct ion, for Stone &
Webster Engineering Corporation. (c) Site Support i HL&P will establish an organization on site that will be involved in resolving problems and will ensure continuity and expertise 'after the Contractor has completed his obligations. This will be done by relocating the present HL&P Project organization physically to the site. The organization will maintain current de,ign-related and procurement responsibilities but will also grow and be directly involved in construction changes, contractor deficiency corrections and start up. Upon project completion, these individuals will be f amiliar with the plant design and be available for direct site or support activities. The relocation of the engineering organization to the site will occur approximately 18 months af ter start of construction. (2) Technical Resources Directed by the Utility 57 (a) Staf fing Levels-Prior to the start of Allens Creek construction, HL&P has maintained an in-house staf f of approximately 31 full-time engineers and (
\_,}/
managers to oversee the design and verify conformance with the applicable regulations, codes and design criteria. This manpower has proved sufficient to meet the responsibilities of the project except in specific cases. In these cases, temporary engineering support is assigned from line departments or consultants contracted to work ur. der the sole direction of HL&P personnel. Figure II.J.3.1-1 identifies corporate technical resources and experience. Figure II.J.3.1-2 identifies individual experience of personnel assigned full time to the Allens Creek Project. To support the construction and operations of Allens Creek, HL&P has scheduled staffing levels as shown in Figure II.J.3.1-3. (b) Level of Education and Experience HL&P has and will continue to retain a highly trained and capable staf f to meet the responsibilities of overseeing the design of Allens Creek. Also, there is a wide range of technical expertise that exists within the corporate organization covering all the major engineering disciplines plus some of the more highly-specialized fields. Should i J 0-143 Am. No. 57, (5/81)
- -_ ~.
ACNGS-PSAR i a technical issue arise that is outside the scope of HL&P's-
;- engineering ' capabilities, HL&P has the option to obtain the outside services of experts to assist in resolving the issue.
N_ l I
- We are aware of the technical support skills required in NUREG-0731 I
for operations. We have already on 'staf f some individuals that meet or will meet the qualifications outlined in that document. We plan-to acquire or train individuals so that all- requirements are 4 satisfie.'- For instance, this year we are actively . recruiting. for specializ. skills in the areas of welding, engineering,. metallurgy, ASME pipe stress analysis, and transient analysis. We recognize the
' existence of a personnel shortage in many of these skilled areas, !
i and have been using outside recruiting agencies, open houses, and nationwide advertising to attract the appropriate personnel. JDie j company has a. formal staf fing program where annually, needed job functions and skills are identified for each department, the future
; staffing plans to fill these. positions are approved by executive j - management, and then a coordinated recruiting ef fort commences.
4 Once Allens Creek becomes operational, HL&P will assume design responsibility. Before that occurs, we will.have in place a' program , that meets ANSI N45.2.ll, " Quality Assurance -Requirements for the - ! Design of Nuclear Power Plants". The technical support staf f will produce a manual of formal ~ procedures governing preparation of [ j design documents, document control (including review and approval), 57 design verification requirements, control of design changes
- , including design change requests, and training, among others.
(c) Training Programs i In addition to hiring experienced individuals into the Company, ML&P
~
has an active technical training program. All professionals have the opportunity and are expected to attend an outside' developmental~ course or seminar each year. Also, line departments, who are l responsible for training, hold technical work shops directed by in-house experts or AE and NSSS training personnel. Typical
, work shops included basic studies of codes, pumps, valves, BWR design, etc.
- ' Further, the Health Physics Division, under the Nuclear Services Department, has established a aaclear-wide radiation training group.
The training group is developing over 13 courses to teach HL&P personnel and contractors in radiation protection, including the ' full range of technician training, general employee training, and operator training. i 1 e t I I H T r
- ).
- w-i 0-144 ~ Am. . No. 57, (5/81) ,
t _b
ACNGS-PSAR As a matter of interest, the training group will conduct a pilot p training program for Radiation Protection Technicians at the request ( of INP0 this spring. INPO has audited this group recently relative
\
to the presentation of this pilot program and found everything highly satisfactory. I (d) Experience Feedback An important input to the technical staff is operating experience. This experience is distri'suted through information contained in documents such as I&E Bulletins and LER's. At present, NRC generated input, including I&E Bulletins, Notices, New Regulations, and Regulatory Guides are screened by our Nuclear Licensing Department for applicability and importance and then sent to the appropriate technical personnel for action. See the response to Item I.C.5. The publication, Nuclear Power Experience Reports, is used as another source of input to the technical support and operations staffs. The reports are reviewed by the cognizant discipline and factored into the plant design, construction and/or planned operation as appropriate along with other inputs. 57 In addition, both the South Texas Project Plant Superintendent and the Manager - Nuclear Services, are members of the EEI Nuclear Operations Subcommittee. This group, which meets triannually, is compesed of the chief technical support and operations personnel for each utility in the U.S. They meet and exchange information {'"'}/ g , concerning operational experiences. This group has been functioning for many years. Through the ef forts of NSAC and INPO, the many hundreds of LER's are now being screened and distributed to interested parties, through a service known as NOTEPAD. We are a user of that service. Also, as a result of TMI, wer are actively participating in industry efforts including the W and GE Owner's Groups. (e) Analytical Capability ! HL&P has recognized the need for in-house analytical capability in certain areas which will be needed. The Nuclear Services Department is developing capability to perform transient analysis and has purchased and is benchmarking such computer codes as RETRAN, CONTEMPT, and COBRA. t t V 0-145 Am. No. 57, (5/81)
ACNGS-PSAR The Nuclear Fuels Department has purchased and is benchmarking such n computer codes as PDQ-VII and ARMP for limited in-core fuel [ management capability. This group is studying the extent of
\ additional in-house capebility needed to adequately meet the Company needs.
(3) Details of the interaction of design and construction activities (a) General The following supplements the material in PSAR Section 13.1.2 describing in more detail, the interaction of design and construction activities by HL&P and its principal contractors, Ebasco Services, Inc. and General Electric Corporation (GE) for the nuclear steam supply system. Establishment of the divisions of responsibility and the means of assuring close integration of the work is manifested in contractual documents, project (inter-company) procedures, the balance of plant System Design Descriptions, and the GE Customer Interface Data Document. Ebasco is responsible to HL&P project management, planning, cost control, engineering, procurement, construction, sub-contract 57 administration, quality control, and quality assurance. Ebasco is also responsible for design interface control among Ebasco, GE, and other vendors and contractors. Ebasco is accountable to perform its services in accordance with all applicable Federal, State and Local codes and regulations including the Quality Assurance requirements of 10CFR50 Appendix B. HL&P monitors and evaluates Ebasco Qr performance of these responsibilities by requiring Ebasco to obtain HL&P approval of the basic design criteria and selected design documents. Further, HL&P places purchase orders for all engineered equipment based on Ebasco generated and HL&P approved specifications. GE is responsible to HL&P design and fabrication of the Nuclear Steam Supply System including preparation of design documents and procurement of related hardware. GE prepares system descriptions and other selected design documents for both HL&P and Ebasco. HL&P monitors and evaluates GE performance by review of these documents. Ebasco reviews these documents to ensure interface coordination between the NSSS and balance of plant. Otherwise GE has authority to determine the NSSS design, subject to HL&P QA surveillance. GE prepares: interface criteria; safety analyses; other design information; test procedures, maintenance and operating procedures; and technical support for NSSS installation. GE is accountable to HL&P to perform its services and provides NSSS designs and equipment in accordance with all applicable Federal, State and Local ' odes and regulations, including the Quality Assurance requirecents of 10CFR50 Appendix B. I' v O-146 Am. No. 57, (5/81)
ACNGS-PSAR HL&P is ultimately responsible for the overall design, construction, and operation of Allens Creek in accordance with NRC regulatory
} requirements, including the Quality Assurance requirements of 10CFR50, Appendix B. HL&P's Project Management Organization, descri-bed in PSAR Section 13.1.1.5 is responsible for providing management oversight of principal contractor activities, obtaining Federal licenses and permits, approving basic design criteria, releasing selected design documents, and authorizing expenditures of funds.
HL&P also retains stop work authority of contractor design and construction activities. A chart showing Project Responsibility Interface is attached. The Allens Creek Project Organization is made up of individuals assigned to the Project from the:several line or funct 5,nal depart-ments. The Organization consists of a Manager and a support team of engineers and professionals whose only function is to manage the design, licensing, procurement, construction, and start up of Allens 57 Creek. The Manager reports to the Vice-President - Nuclear Engineering & Construction, and is accountable to him for the cost, schedule and quality of the Project. The Organization in turn manages the contracts for outside support, principally Ebasco and GE. All technical t ' ' administrative direction to the contractors is provided through the Allens Creek Project Organization. The individuals assigned to the project bring with them the necessary authorities and responsibilities to act on behalf of the department represented within the framework of the Allens Creek Project Organization. This allows the Project to function as a coherent unit with decisions being made at the appropriate level. Further, the Organization except Accounting and Environmental Pro-tection is grouped physically together ensuring good communication and interface between departments and disciplines. The Project Organization is staf fed so that all normal activities can be ac-complished within the team. Line Department Overview Line departments establish technical policy and technical procedure 1 through published policies and guidelines. These items serve to document required policy and procedures on a corporate basis and ensure that company nuclear activities are coordinated and con-trolled. These publications are being prepared so as to .e in place upon the beginning of construction of ACNGS. Detailed project procedures reference and implement any appropriate department re-
- quirements. Project procedures may expand upon and provide more i
detailed quidance relative to department publications and documents (written approval) and deviations therefrom. i s./ 0-147 Am. No. 57, (5/81)
ACNGS-PSAR The line departments serve as the primary technical resource for ("N complex or specialized tasks and for peak work periods. The line i departments also plan for analytical systems, methods development, and additions to staff. The line-departments and divisions assigning full-time project staff to Allens Creek are: Nuclear Services Nuclear Engineering Health Physics Nuclear Licensing Nuclear Operations Power Plant Engineering 57 Plechanical Engineering Electrical Engineering 4 Civil Engineering I&C Engineering Power Plant Purchasing Allens Creek Purchasing Material Control m Environmental Protection Power Plant Accounting Quality Assurance QA Engineering Vendor Surveillance Allens Creek Administration Project /.dministration Project Controls Project Construction Nuclear Fuels Records Management L O-148 Am. No. 57, (5/81)
ACNGS-PSAR (b) Overview of Design
\ Overview of design is implemented and controlled by the Allens Creek Project Engineering Manager and those individuals of the Project Organization assigned to him. The inter-company drawing and specification development, review, and approval procedures outline step by step the activities done by Ebasco, HL&P, and equipmenc suppliers. Intra company Allens Creek procedures describe Project-internal activities. These procedures specify how the review is -
initiated, assigned, and documented. HL&P provides management overview of design changes through the Project (inter-company) Procedures which specifically define all mechanisms that the Principal Contractors can use for design changes. -In addition to the specific HL&P control aspects over design and procurement activities, HL&P Monitors the quality, cost, and timeliness of other activities performed by the Principal Contractors. Management oversight of contractor design activities is facilitated by the issuance of several status and performance reports which are directed to various levels of management. Also, copies of correspondence among contractors are sent to HL&P for 57 information. , (c) Overview of Construction The HL&P internal organization described in Section 13.1.1.1.1.7 shows the Allens Creek Project Construction staf f including the Site Construction QA Group and its relation to the Allens Creek Project m ) Organization. The Project Construction Manager (PCM) and his staf f are responsible for construction overview of contractor performance. The d contractors and sub-contractors under Ebasco construction management are responsible for construction in a manner that conforms to design quality requirements. The PCM and his staff: monitor construction activities; approve schedules, field procurements, selected invoices, and other financial controls; monitor compliance with permit and license requirements; monitor procedure compliance; monitor coordination of Ebasco field engineering with Ebasco home ! office engineering staf f; and coordinate Contractor turnover of l plant systems to Nuclear Operations. In addition, QA provides construction overview through the Site Construction QA Group which is responsible for monitoring the QA aspects of site construction, including: review of contractor site procedures; audits and surveillance of construction; identification of quality problems and monitoring of their resolution; and acceptance reviews of components, constructed structures, and I A i
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0-149 Am. No. 57, (5/81)
ACNGS-PSAR completed systems. The Site Construction QA Group interacts p directly with the Ebasco site organization and with the HL&P home g i office QA organization. HL&P will have approved procedures for V construction activity. These procedures will reflect the organization and will conform to applicable regulatory requirements, contractural arrangements, and the Allens Creek Quality Assurance Plan. Procedures will exist for each organizational element involved in construction overview activities. (4) Transition to Operation (a) Technical Continuity HL&P has a single Executive Vice President responsible for nuclear plant engineering, construction, fuel, QA, and operation. This will greatly facilitate the transition from construction of Allens Creek to operation. The Nuclear Services organization responsible for review and approval of plant design will continue as the technically cognizant expert resource when Allens Creek operates, performing the same functions of engineering support as they do now for the South Texas Project. Once Allens Creek becomes operational, HL&P will provide the 57 required technical support necessary to assure safe and reliable plant operation. This support will be consistent with the g,uidelines suggested in NUREG-0731. Technical specialty support s from outside sources will be employed when necessary. We have been
) studying various organization alternatives to provide this support s,/ and the transition steps from a design and construction team to an operations support team. While our studies are not yet complete, we have made note of several key factors. Since the HL&P Project Engineering Organization vill be physically located at the site during construction and start up, the individuals will nave excellent familiarity with the equipment. These individuals will be a basic resource for actual transfer to the operations or engineering support groups. We think keeping this group on site
! will improve its performance by giving the technical support staff maximum access to systems that they will be working on and by developing a close relationship with the operating staf f which should serve to improve comm2nications. Although there will be formal procedures by which the plant staf f can request design changes, this close relationship should improve the mitual ! understanding and performance of both groups. Our goal for technical skill level is to have on staff individuals who are technically capable of performing design verification for all technical areas, especially those that are uniquely nuclear. For very specificalized and complex areas, such as seismic analysis, i l t L) 0-150 Am. No. 57, (5/81)
ACNGS-FSAR we will most likely continue to employ outside consulting
,s assistance. We beliove that a utility must have an in-depth expert
) knowledge and involvement in technical matters affecting plant operation, and we will direct our resources to developing and maintaining that knowledge and involvement.
(b) Operational Continuity Nuclear Operations has had full time individuals assigned to the Project Engineering team since design began to ensure operational aspects are f actored .into the plant. . HL&P intends to employ the operating staf f elth ample lead time for them to learn the plant design and operation, Furthermore , it is HL&P personnel policy to open all new tcchnical staf f positions to internal staff, and to encourage transfers vitain the organization. Thus, engineering and management personnel involved in Altens Creek design and construction phases will be encouraged to transfer to the operational positions as they are available, which will facilitate the transfer of expertise to operation. The Nuclear Operations group will be deeply involved in the preoperational testing, hot functionals and start up. At the South Texas project, the Assistant 57 Plant Manager is also the Project Start up Manager. (c) Contractor Continuity An additional consideration is that the NSSS supplier, GE, will be subcontracted to provide the actual operating and maintenance procedures for the NSSS, not merely guidelines. This will help (-,/) s_, ensure that plant operations reflect the engineering expertise in plant design. In summary, HL&P's internal organization and policies ~are such that a smooth traasition to operation will be facilitated. (5) Management overview I The Houston Industries Board of Directore, HL&P's parent Company, exercises top level management overview by authorizing the capital required for the project. The Board of Directors regularly reviews the summary status, progress, and prudence of the project activities. i HL&P's Executive Vice President exercises top level management overview I by approving funds to implement project decisions, by approving staf fing complements, and by executing contracts for architect engineering services, the nuclear steam supply system, the turbine generator, and nuclear fuel. The executive officer regularly reviews the project status, progress, and current activities and sets policy for future activities. He has no responsibilities other than nuclear and reports f A l b
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0-151 Am. No. 57, (5/81) t - . _ _ _ _ . _ _ . _
ACNGS-PSAR directly to the President and CEO. The Manager, Quality Assurance, 77 t reports directly to the Executive Vice President. HL&P's Vice President - huclear Engineering and Construction is the corporate
\
of fica responsible for design, and construction of all nuclear generating stations, including Allens Creek. He has no other responsibilities. For Allens Creek, the VP - Nuclear Ergineering and Construction authorizes all NRC licensing submittals, establishes the Nuclear Organizational structure and division of responsibilities, and approves the filling of each staf f position within the approved staffing complement. He delegates responsibilities within the Nuclear Organization as described in PSAR Section 13.1.1.2.3. He regularly reviews status and progress information, is informed of significant project decisions, issues, problems, and project plans for resolution of issues and problems through reports prepared by the Allens Creek Project and the Principal-Contractors. The Vice-President - Nuclear Engineering and Construction regularly holds 57 a Quarterly Management Review Meeting on Allens Creek. Ebasco executives and CE management are also present at these meetings, thus enabling the management of all three companies to be regularly informed of the project status, management and technical issues and plans for the future. The Allens Creek Project Organization provides Monthly Status Reports to the Vice-President - Nuclear Engineering and Construction, to other HL&P executives, and to the Principal Contractor project managers. These
,-. reports identify recent progress, current difficulties, and planned activity over the next reporting period. These reports ensure that top-level executives are aware of Allens Creek Project activities.
0-152 Am. No. 57, (5/81)
,. ~
I{ f EXECUTIVE VICE PRESIDENT NUCLEAR VICE PRESIDENT o MANAGER AC PROJECT GE PROJECT EBASCO ENGINEERING PROJECT PURCHASING PROJECT MANAGER MANAGER MANAGER MANAGER o I ( g L _ _ _ _7 .__/ s _ - _q m ( ,
/ I I I I GE GE EBASCO I&SE ENGINEER ENGINEER I
VENDORS VENDORS FABR CATOR i ENGINEERING EQUIPMENT LEGEND: _ DIRECT RESPONSIBILITY I - = = _ INDIRECT RESPONSIBILITY
s i ACNGS - PSAR 1 MANAGER QA QA PROJECT MANAGER f I PROJECT CONSTRUCTION MANAGER ( ) t 1 i EBASCO QA CONSTRUCTION CONSTRUCTION MANAGER l BULK SUB-SUPPLIERS CONSTRUCTOR CONTRACTORS i BULK ERECTION MATERIAL AM. NO. 57. (5/81) l HOUSTON LIGHTING & POWER COMPANY ! Allens Creek Nuclear Generating Station Unit 1 POWER PLANT PROJECT RESPONSIBILITY INTERFACE FIGURE II.J.3.1-1}}