Text

Provides Addl Documentation for Tech Spec Table 3.7.A, Containment Isolation Valves Requested by NRC in 890615 Meeting & 890622 Ltr
	ML20246K846
Person / Time
Site:	Browns Ferry
Issue date:	07/13/1989
From:	Michael Ray; TENNESSEE VALLEY AUTHORITY
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	TAC-00080, TAC-00081, TAC-00082, TAC-80, TAC-81, TAC-82, NUDOCS 8907180281
	Download: ML20246K846 (72)
	v • d • e

_

gs TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSCE 37401 5N 157B Lookout Place JUL 131389 TVA-BFN-TS-251 SUPPLEMENT 1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN)'- TVA BFN TECHNICAL SPECIFICATION (TS) i NO. 251 SUPPLEMENT 1 (TAC 00080, 00081, 00082) ;

The purpose of this letter is to provide additional documentation for BFN TS 251, Table 3.7.A Containment Isolaticq Valves, requested by NRC in a meeting, dated June 15, 1989, and a letter dated June 22, 1989.

BFN TS 251, Table 3.7.A Containment Isolation Valves, was submitted to NRC on August 2, 1988. The purpose of that submittal was to update and correct the subject Table based on modifications to the plant and to better align this Table with the BFN Appendix J Program.

During the review of the proposed TS change, NRC asked some specific questions pertaining to the submittal. As a result it was agreed that a working meeting be held to resolve these questions and to discuss the submittal as well as the BFN design in detail. At tne conclusion of the meeting the questions were resolved and NRC stated that they would request BFN to provide some specific information on the docket in order for them to complete their review. This request was officially made in NRC letter dated June 22, 1989.

This information is provided in the following enclosures to this letter:

Enclosure 1 contains the handouts used during the June 15, 1989 meeting. This includes, among other things, a brief outline of the BFN Appendix J history, BFN Containment Isolation design, and simplified diagrams of the containment penetrations the NRC reviewer originally had questions about.

Enclosure 2 is a Table describing the penetration number, applicable General Design Criteria (GDC), valve number, and valve location with respect to ;

containment. This Table contains the information discussed in the June 15, 1989 meeting.

Enclosure 3 discusses the BFN Reactor Building Closed Cooling Water System (RBCCW). Contained in this section is a discussion on how BFN meets SRP 6.2.4 criteria.

8907180281BQ7hg DR ADOCK O pgg g An Equal Opportunity Employer l

l 4

U.S. Nuclear Regulatory Commission JUL 131989 Enclosure 4 provides documentation of how BFN Primary Containment Isolation design provides means, other than automatic isolation signals, to ensure primary containment integrity for valves FCV-84-19, FCV-84-8A, -8B, -8C, -8D, I.

and FSV-43-28A, -288, 29A, and -298.

Enclosure 5 contains TVAs description and justification for deletion of HPCI and RCIC injection MOVs (71-39, 73-44) from TS 3.7 Tables.

I Enclosure 6 describes the characterization of the BFN Appendix J Program for

closed systems.

The information contained in this letter does not change or alter any of the proposed TSs as submitted in the initial TS 251 transmittal. This submittal only provides additional information and documentation supporting that submittal. If you have any questicins please contact Patrick P. Carier at (205) 729-3570.

Very truly yours, TENNESSEE VALLEY AUTHORITY f'% b s M. J. Ray, Man ger Licensing Project Management t

on his /f/ an}swornf Subscribed p b/e day ors /M/1989. ore me Notary P'ublic "

Al (l '

MyCommissionExpires8/7/8 i /

Enclosures cc: See page 3

- _ _ _ - _ - _ - - _ _ _ 1

t ,

l l 1

U.S. Nuclear Regulatory Commission JUL 131989 cc (Enclosures):

Ms. S. C. Black, Assistant Director for Projects TVA Projects Division

, U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike l Rockville, Maryland 20852 '

Mr. Charles R. Christopher, Chairman Limestone County Commission P.O. Box 188 Athens, Alabama 35611 Dr. C. E. Fox State Health Officer !

State Department of Public Health State Office Building Montgomery, Alabama 36194 Mr. J. E. Jones General Electric Company No. 1 Union Square 808 Krystal Building Chattanooga, Tennessee 37402 Mr. B. A. Wilson, Assistant Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commissici Region II 101 Marietta Street, NW, Suite 29(0 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35609-2000 American Nuclear Insurers Attention: Librarian The Exchange, Suite 245 270 Farmington Avenue Farmington, Connecticut 06032 l

Le 1 :- ,

(.

i ENCLOSURE 1-1 ,

TVA/NRC MEETING

l. REVIEW OF TECHNICAL SPECIFICATION AMENDMENT 251 PRIMARY CONTAINMENT ISOLATION VALVE TABLE

_ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ - - _ _ _ _ _ _ _ _ - _ _ - - -D

?

r-AGENDA TVA/NRC MEETING - June 15,1989 INTRODUCDON FROST HISTORY MULLI.NG IMPLEMENTATION SCHEDULE MULLING T.S. AMENDMENT MULLING DESIGN BASES TURNBULL' WORKING GROUP ALL J

?

,4' LIST OF TABLES 3.7.A PRIMARY CONTAINMENT ISOLATION VALVES 3.7.B TESTABLE PENETRATIONS WITH DOUBLE O-RING SEALS 3.7.C TESTABLE PENETRATIONS WITH TESTABLE BELLOWS 3.7.D AIR TESTED ISOLATION VALVES 3.7.E - PRIMARY CONTAINMENT ISOLATION VALVES WHICH TERMINATE BELOW THE SUPPRESSION POOL WATER LEVEL 3.7.F PRIMARY CONTAINMENT ISOLATION VALVES LOCATED IN WATER SEALED SEISMIC CLASS I LINES

. 3.7.H TESTABLE ELECTRICAL PENETRA110NS I

L__-__________---__-_. ._

\

PROWNS FERRY APPENDIX J HISTORY JULY 7,1966 - BF UNIT 2 OL APPLICATION FILED FEBRUARY 14,1973 - 10 CFR 50 APPENDIX J PUBLISHED JUNE 28,1974 - BF UNIT 2 LICENSE ISSUED AUGUST 4,1975 - NRC REQUEST FOR TVA'S PLANNED ACTIONS AND SCHEDULE TO ATTAIN CONFORMANCE WITH APPENDIX J SEPTEMBER 9,1975; OCTOBER 10,1975; JANUARY 15, 1976; FEBRUARY 26,1976; JULY 8,1977 - TVA LETTERS TO NRC WITH EXEMPTION REQUESTS SEPTEMBER 20,1984 - TVA LETTER OF CLARIFICATION OCTOBER 24,1984 - NRC RULING ON APPENDIX J EXEMPTION REQUESTS AND SAFETY EVALUATION - DENIES RBCCW EXEMPTION MARCH 15,1985 - TVA IDENTIFIES UNTESTED VALVE BONNETS MAY 10,1985 - TVA IDENTIFIES MODIFICATIONS FOR CONFORMANCE WITH APPENDIX J AND SCHEDULE AUGUST 28,1985 - TVA/NRC MEETING IN BETHESDA PROVIDING NRC WITH PRESENTATION OF PLAN TO UPGRADE APPENDIX J PROGRAM NOVEMBER 8,1985 - EXEMPTION REQUEST SUBMITTED FEBRUARY 21,1986 - TVA LETTER TRANSMITTING MEETING MINUTES AND A CHARACTERI7.ATION OF RHR AND CS SYSTEMS AUGUST 2,1988 - T.S. AMENDMENT SENT TO NRC UPDATING AND CONSOLIDATING THE PRIMARY CONTAINMENT ISOLATION VALVE TABLES

- _ _ _ _ _ _ _ _ _ - _ - _ - _ b

ORIGINAL IMPLEMENTATION SCHEDULE.

COMPLETED ITEMS 71-17/18 RCIC PUMP SUCTION 71-34/547 RCIC PUMP MINIFLOW 73-26/27 HPCI PUMP SUCTION ,

73-30/559 HPCI PUMP MI~NIFLOW 71-14 RCIC TURBINE EXHAUST (BONNET) 73-23 HPCI TURBINE EXHAUST (BONNET) 71-32 RCIC VACUUM PUMP DISCHARGE (BONNET), UNIT 3 ONLY 73-24 HPCI TURBINE EXHAUST DRAIN (BONNET), UNIT 3 {

ONLY j i

ITEMS TO BE COMPLETED PRIOR TO STARTUP l U-1 AND U-3 ALL IDENTIFIED PROBLEMS i U-2 74-58.RHR CONTAINMENT SPRAY (BONNET)

U-2 74-792/804 PSC HEAD TANK TIE IN TO RHR LOOP 1' U-2 74-802/803 PSC HEAD TANK TIE IN TO RHR LOOP 2 U-2 75-606/607 PSC HEAD TANK TIE IN TO CS LOOP 1 i U-2 75-609/610 PSC HEAD TANK TIE IN TO CS LOOP 2 U-2 RHR AND CS SYSTEM WALKDOWNS ;

U-2 71-32 RCIC VACUUM PUMP DISCHARGE (BONNET)

U-2 73-24 HPCI TURBINE EXHAUST DRAIN (BONNET) l' ITEMS TO BE IMPLEMENTED BY U2-C7 STARTUP 70-47/506 RBCCW SUPPLY AND RETURN 84-8A/B/C/D CAD VALVES (ARE CURRENTLY TYPE C TESTED, BUT IN THE REVERSE DIRECTION)12-742 AUX BOILER TIE IN TO RCIC (BONNET ONLY) 71-34 RCIC PUMP MINIFLOW(BONNET ONLY AND ORIFACE) 73.-30 HPCI PUMP MINIFLOW (BONNET ONLY) i 74-103/103 RHR VENT RETURN 74-119/120 RHR VENT RETURN U2 WILL BE TYPE A TESTED THIS OUTAGE. ALL OF THE ABOVE WILL BE TYPE A TESTED PRIOR TO STARTUP, EXCEPT l RBCCW VALVES 70-47/506.

f

i

... j CURRENT STATt. ,

Previously identified modifications are COMPLETE.

Appendix J program implemented.

T.S. amendment being reviewed by NRC.

FSAR update is in drafting.

__ -__ i

l PRIMARY CONTAINMENT ISOLATION VALVES BEING MOVED FROM TABLE D, E, F TO TABLE A 2-1192, 2-1383 kJ

/

2 1192 VM 2-1383 FIGURE 1 33-785, 33-1070 kJ

/

33-785 VM 33-1070 FIGURE 2 32-336, 32-2163 68-508,68-523 1 32-2163 32-336 68-550, 68 555 68-508 68-550 68-523 68-555 FIGURE 3 I

84-8A,84-600 g

84-88, 84~601 84-80,84-603 84-8 A 84-600 84-8B 84-601 ,

84-8D, 84 602 84-8C 84-603 84-8D 84-602 FIGURE 4 l

l 1

L_________.________.___.__.._____ J

PRIMARY CONTAINMENT ISOLATION VALVES BECNG MOVED FROM TABLE D, E, F TO TABLE A l

12-738,12-741 71-14,71-580

[

71-32,71-592 12-741 12-738 )

71-14 71 580 i 73-23,73-603 71-32 71-592 73-23 73 603 73-24,73-609 73-24 73-609 i FIGURE 5 l

74-60, 74 61 74-71, 74 72 k VR VM 74 74,74-75 74 61 74 60 74-75 74-74 .

l 74 57, 74-58 74-58 74-57 74-72 74-71 FIGURE 6 74 722 BLIND FLANGE 7

74-722 l

l l

FIGURE 7 i i

l l

i 75-25, 75-26* 75 26* 75-25 I 75-54' 75-53 75-53, 75 54* kJ VR

Already in table A, no change FIGURE 8 i

i

PRIMARY CONTAINMENT ISOLATION VALVES BEING MOVED FROM TABLE D, E, F TO TABLE A 3-558*, 3 554*

3 558* 3 554*

73-45, 73-44 "

l

Already in table A, no change

" Being deleted from T.S. 3.7.A 73-45 73 44 "

A LA

_/ Vm FIGURE 9 85-576*

/

69 579 85 576*

69 579 3-572*, 3-568*

3-572* 3 568*

71 40, 71 39"

Already in table A, no change

" Being deleted from T.S. 3.7.A 71-40 71-39 "

LJ V7 FIGURE 10

~

i PRIMARY CONTAINMENT VALVES BEING ADDED TO TABLE A THAT WERE NOT PREVIOUSLY IN T.S. OR IN APPENDIX J l 71-17, 71 18 l-73 26, 73-27 V3 VR 71-17 71-18 73 26 73-27 FIGURE 11 .

71-34,71-547 73-30, 73 559

[

71 34 71-547 73-30 73-559 FIGURE 12 74-792, 74 804 74-802,74-803 75-606,75-607 74-804 74 792 74-803 74-802 75-609,75-610 75-606 75-607 75-609 75-610 FIGURE 13 70-47 70-47 70-506 (

/ ,

70 506 FIGURE 14 I

i l

. j

i. l l PRIMARY CONTAINMENT ISOLATION VALVES BEING DELETED FROM T.S.

4 PART I - DELETION!i FROM TABLE A 71 -6 A, 71-6B 71-2* 71-3*

TURBINE

Currently in T.S. I~ ^

71 -6 B FIGURE 15 71 -7 A, 71-78

Already in Tablo A

" To be deleted from T.S. 3.7.A (Rel. Figure 21) 71-40* 71-39**

DRYWELL . kJ l V7 FROMCONDENSER 3-572* 3-568*

_ _ 71 -7 A 7 {-

71-34 71-547 LA VM /

)

TORUS RHR PUMP TEST l

FIGURE 16 1

4 l

PHIMARY CONTAINMENT ISOLATION VALVES BEING DELETED FROM T.S.

PART I - DELETIONS FROM TABLE A j 7317A,73-17B 73 26 73 27 kJ kJ TM 73 73 ( 7 3-17 A 73-17 B FROMSEALCONDENSER 3 558 3-554

[ FW LINE A DRYWELL 73-45 FIGURE 17 FCV-73-6A, 73 6B 73 2 73-3 TURBINE

~

73-81 w2 73 3 6B FIGURE 18 i

o 5 -

PRIMARY CONTAINMENT ISOLATION VALVES BEING DELETED FROM T.S.

PART I - DELETIONS FROM TABLE A BEFORE MODIFICATION 74 102 74-102 74-119 74 103 MS V7 FM 74-119 74-103 74-120 kJ kJ 74 120 F7 77 AFTER fCDIFICATION 74-102 74-119 TORUS ' ' k2 kJ V7 VR 74-103 74-120 kJ kJ CAPPED FIGURE 19 BEFL MODIFICATION 85 573 85-576 TOVESSEL

[ [g AFTER MODIFICATION 3-572 3-568

[ [ FWLINE B 85-576 FIGURE 20

PRIMARY CONTAINMENT ISOLATION VALVES BEING DELETED FROM T.S.

PART 11 - DELETIONS FROM TABLE D 3-572 3-568

[ FW LINE B 71-40 71-39 kJ

/ VR 69 579

/

85-576

/

FIGURE 21 73 44 3 558 3 554 l

73-45 73 44 kJ

/ V3 FIGURE 22

l .s J 4' ~. ; I'l.~

2. : ,. 7, r

~

PRIMARY CONTAINMENT ISOLATION VALVES BEING ADDED TO TABLE A -

BECAUSE OF ADDITION OF CONTAINMENT PENETRATION PATHWAY (TIE-IN) k l

L 32-2163*, 32 336' 32 2163* 32-336*

84-617'

Already in T.S. being moved 84-617 to Table A. _

3 FIGURE 23

,4 '.

CONTAINMENT ISOLATION DESIGN NRC REGULATIONS AND TUR CIS DESIGN DATES

'~ Development of GE CJS design began 1960 (1950's. + early 1960's) 1961 1962 -- BFN Design + Analysis Report 1963 1964 1965 1966 GE Design Spec for Reactor Containment AEC Interim GDC 53 - 1967 22A1169 1968 L construction C permit issued for BFN 1969 1970 NRC GDC 54,55,56, + 57 1971 GE CIS Design Specification 22A1132AE 1972 10CFR50, Appendix J- 1973 BFN FSAR submitted to'NRC BFN unit 1 operating license 1974 Standard Review Plan 6.2.4 1975 BFN received SER from NRC (CIS design accepted) 1976 1977 +

1978 1979 1980 1981 1982 1983 1984 j 1985 1986 i

1987 1988 I

1989 i

- - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ i

Vent to Standby Gas Treatment I X-26 L J V 3 FC V- 8 4 -19

{

I X-231

1) Redundant Valves

2) Normally Close.d

3) Administratively controlled a) SOS controlled key i

b) OI and EO restricted use per FSAR.

4) Appendix J tested References 1

2-47E2865-12 2-47E862-1 FIGURE 24 1

J

- - _ _ _.- _. ._ _ . - - - - - - -- .- - - -A

RHR Torus Spray i

X-211 A X-2i l B

"~ -~

Torus Spray g j g j V~R F M FCV-74-58 FC V-74 -57 FCV g 74j2 FC V-7 4 -71 Test Return )

V 3

_ FCU-74-59*

X-210A FCU-74-73*

X-210B

Valves not consided containment isolation valves but are part of Appendix J test boundary.

1) Redundant Valves (Water seal and FCV-74-57(71) for the Test Return Line)

2) Normally Closed

3) Interlocked with system LPCI initiation.

4) Standard BWR Design

5) Required post accident (Essential System).

References

6) Water sealed 2-4 7 E 811 - 1

7) Appendix J tested FIGURE 25

T ,

Drywell Control Air Supply Drywell 2-32-336 X-22 2-32-2163 _

2-84-617*

1) Redundant Valves

2) Equivalent isolation capability as original design basis.

3) Appendix J tested

New Valve ,

i References j l

2-47E862-1 l 2-47E2847-5 4 2-47E2847-9 i d

i FIGURE 26 'l

_ _ _ _ _ _ _ . - _ - - _ - . _ _ _ . . _ - _ - . . _ _ l

J 0: ., ,

J*-

Reactor Building Closed Cooling Water

-X-24 y l 1 /g \

! FCV-70-47 l Drywell X-23 70-506

1) Group C valve l

2)' Specifically described in the FSAR

,. 3) Justified per NEDC 22253 a) at least one valve in each line close to containment b) isolation valves are automatic or capable of remote manual operation c) power supplies and controls are safety grade d) piping system out to and including isolation valve are safety grade

4) Appendix J tested

)

References 2-47E822-1 FIGURE 27 l

l 5

..i j

l l

l \

l<

l TIP Purge

{

76-653 X-35f

1) Normally Closed

2) Original design

3) Justified rer GE NEDC 22253 and accepted by NRC on other dockets.

a) Function as instrument lines or support the operation of instruments.

b) Small diameter lines c) Safety Grade valve d) Off-site dose is a small fraction of 10CFR100 limits.

4) Appendix J tested 1

Reference 2-47E600-14 FIGURE 28

i l

l RIIR Suction Sampling i

I I S 1

LM k J L M l V 3 V~3 V~3 FSV-43-28 A FSV-43-28B FSV-43-29 A FSV-43-29B X-225A X-225B 1

1) Redundant Valves l 2) Normally Closed 1

3) Conservative inclusion of salves - normally isolated by manual valve.

4) Administratively controlled - Chemistry Instruction

5) Water sealed

6) Appendix J tested l

l l

i l

References i 2-47E610-43-1 2-47 E 811 -1 FIGURE 29 4

i j

. - r CAD Injection X-25' X-205 l ..

L M V 3 l' FSV-84-8 A FSV-84-8B FSV 84-8C

' FS V-84-8 D l

1) Redundant Valves

2) . Normally Closed

3) ' Administratively controlled - 01 and EOI.

4) Required post-accident (Essential System).

- 5) Appendix J tested References 2-47E862-1 FIGURE 30 i

I i

\

j IIPCI Min Flow i

RHR P ,mp Test Line Torus O

E 2 F^ 7 FCV-73-30 73-559

1) Redundant Valves

2) Normally Closed

3) System isolation signal - interlocked with turbine operation.

4) Water scaled

5) Appendix J tested i References 2-47E812-1 FIGURE 31

I RIIR Drywell Spray s X-39 A i X-39B k J L M V 3 V M F C V 61 FCV-74-60 FCV-74-75 FCV-74-74 I) Redundant Valves

2) Normally Closed

3) Interlocked with systein LPCI initiation.

4) Standard BWR Design

5) Required post accident (Essential System).

6) Water sealed

7) Appendix J tested Refer ences l 2-47E811-1 4

FIGURE 32 l

E_____-.______ _ _ _ _ _ _ _ _ - - - - - _ _ __J

'h .

Core Spray Injection e -

X-16A X-16B .

O.

L 2 Rx

!/ ;r a FCV-75-25 FCV-75-53 i

1) Redundant Valves

' 2) Normally Closed

3) Admittistratively controlled (OI and EOI)

4) Otsndard 'BWR design.

5) Required post accident (Essential System).

6) Water sealed

7) Appendix J tested References 2-47E814-1 FIGURE 33 i

1 IIPCI Pump Suction i X-226 l

C LJ Torus F 7 7 7 FCV-73 26 FCV-73-27

1) Redundant Valves

2) Normally Closed

3) Group 4 Isolation

4) Water sealed

5) Required post accident (Essential System)

6) Appendix J tested References 2-47E812-i FIGURE 34

.fg ..

I

(

RCIC Min Flow X-210A from RHR Pump Test Line Torus g

1. .w A

~

V M ,

F C V-71 -34 71-547 l

l l

l-

1) Redundant Valves

2) Normally Closed

3) System isolation signal - interlocked with turbine operation.

4) Water sealed

5) Appendix J tested References 2-47E813-1 FIGURE 35

i j

l RCIC Suction l

1

)

i l

X-227A !

L M !

Torus L.f i' i r 7 7 m F C V 17 FC V 18 l

)

1) Redundant Valves

2) Normally Closed

3) Adtrinistrative Control (Operating Instructions)

4) Water scaled

5) Required post accident (Essential System)

6) Appendix J tested References 2-47E813-1 FIGURE 36 l

-e 'i i

RHR & Core Spray System Keep Fill Connections j r

i 74-802 74-803 74-792 74-804 75-609 75-610 75-606 75-607

/ '/ from Head Tank f

1) Redundant Valves

2) Water sealed

3) Appendix J tested Reference 2-47 E811 - 1 2-47E814-1 FIGURE 37

4 x-226 IIPCI Water FCV-73-26 FCV-73-27 I O OR Torus w

F 2

3 w

V 4 (h b b Containment Boundary _

\ FCV-73-1711* FCV-7.5-17 A*

% N )

( From g Condensa te X-9A Pump !

Drywell g d 3 558 3-554k To Reactor k FCV-73 44*

73-45 %

I L M

</s{ V 3 X-210B k FCV-73-30 (73-559 % i L-.A mkg Torus I 3 [ g Min Flow Line 6

1) System Isolation Valves l l l

2) Outside Primary Containment Isolation Boundary 1

l Reference l l Valves Deleted frcm Table 2-4 7 E 812- 1 FIGURE 38 L_______.___________.__ __ _ _

j

l E f f( '

x-11 HPCI Steam i FCV-73 2 FCV 73 3 ,

i Turbine Drywell a

FCV-73-81 i n :

y C3 n-Containment Boundary % / /.,,,

l 'k N \

l

% FCV-73-6B

FCV-73-6A

x-214

(

HCV-73-23 % l l

\ 73 603g Torus g

/ i

) 11CV-73 24 t

l l O 73 6o, !

/

x-:22 g 4 t

l

1) System Isolation Valves

2) Outside Primary Containment Isolation Boundary Valves Deleted from Table Reference 2-4 7 E812-1 1:

FIGURE 39 J

X 227A aler FC V-71 -17 FCV-71 18 O OR k A Torus

% i b b Containment Boundary _

FCV-71-7 B* FC V-71-7 A *

{g h

4 From Condensate

( Pump X-9B % )

Drywell g 3-572 3 568 k l

% I To Reactor --

k FCV-71 -39*

i 71-40 l k A N V m X-210A k

% 4

\

i FC V-71 -34 k j

\ l 71-547 %

Torus k 4

)

g Min Flow Line l l

l 1) System Isolation Valves

2) Outside Primary Containment Isolation Boundary 1

J Reference j 1

Valves Deleted from Tabl 2-4 7 E813- 1 FIGURE 40 t

L .

l '.

-y. s 4 t

X-10 RCIC Steam

FCV-71-2 FCV-71-3 O Oj g l< ' Reactor Turbine 1.

1 (1 0 Containment Boundary / /

k' \ h,

\ FCV-71-6B* FCV-71-6A*

X-212 :

(

HCV-71 14 %

71-580 Torus \

) Y 5 HCV 71-32 l {

71-547 Condenser

% Vacuum Pump

/ !

X-221 %

t 1). System Isolation Valves

2) Outside Primary Containment Isolation Boundary

Valves Deleted from Table Reference 2-47 E 813- 1 FIGURE . 41

- l f --

L ENCLOSURE 2 l

NRC PRIMARY CONTAINMENT ISOLATION VALVE REVIEW TABLE h

1

L -

BE 0 HE 0 0 E

AR E 8 8 8 .

P P

A CU UG I F mN C k

P mmN mNmm9 P C 3, 3, 9

3,.

9 4,

0 1

4, 0

1 4,

1 0

0 1

1 4

S I

S A

B 3 3, 3, 3, 3, 3, 3, 3, D 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 E

N C D

3 C3 D

C3 C 3 C 3 C1 C 3 C1 F D 2 D - D 2 D D 2 D E A

/ /

A

/ / / / / /

A A A A A A A A G 52 G 52 G 5 G 52 G 5 G 52 G 5 G 52 A D N N N

/ /

R N N N N N N N NR RA NR RA NRRA NRRANR RANR RA NR NR RA RA N

E 1

ES ES ES ES ES ES ES ES T TF TF TF TF TF TF TF TF O N W N I N N N N N .

C .

R N

4 8, 5 6 6 O Y -

5- 5- 75 SI T 5 4 7 6 8 7 2 1 6 5 3 3 -5 3 2 2- 5 -

I 1 1 3- 3 5 5 5 8 8 2 2 2 2 PL 1 5 5 ,8 7 7 7 7 1

U8 I 1

t 1

1 t

l t

t t

5, 5 5 0 5 5 7

A V V V V V V V V V V 4 5 5 V

KP C C C C C C C C C C

- - - 4 ,

- 9 CA F F F F F F F F 3

7-3 3 1 7 3 3 3 3 C A

B C F F 7 F

. V V9 C C6 F F L

A N

1 1 1 1 1 1 1 1 1 1 S G P P P P P P P P P - -

P S U U U U U U U U J A A A A A A A A U OR y I

E O O C O O O O O r

/ / / / / / / /

O V S. R R R R R R R R N N N N N N N N L T R A C G G G G G G G G G G G G V A O

IS S

E L

ON C P

B A

T EI VR LO O COC O C O JCO C C C C O O O O O P

/

P

/

O P

/

C P

O

/

P

/

O P

C

/

C P

/

P O

/ C C OO C

F AN V

N -

E K K P N N N N N N N N C V

O V

0 C C V 0

Y E E T

E R

E B

R E ME B

EB #EP #B REB WE ME E -

4 4

E B X X E

L B

X E

X B

L B

X I

f X

B 4

E 4

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ENCLOSURE 3 RBCCWS DESIGN REVIEW AND SRP 6.2.4 EVALUATION i

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The reactor building closed cooling water system (RBCCWS) is designed to provide a continuous supply of cooling water to designated plant equipment located in the primary and secondary containments.

The system consists of pumps, heat exchangers, surge tank (vented into secondary containment), and necessary support and control equipment. The RBCCWS supplies cooling water to components located inside containment-namely, recirculation pump !

seals, recirculation' pump motors, drywell atmosphere coolers, and drywell equipment drain sump coolers. The design basis fo the primary containment isolation function of the RBCCWS, which neither connects to the reactor primary system nor is open into the primar?

containment, is provided with one a-c powered valve located on the 8-inch effluent line (outside primary containment) and a check valve located on the 8-inch infinent line (outside primary containment).

These valves are subject to Type C testing with. air in accordance with 10CFR50, Appendix J. The a-c powered valve has remote-manual control from the control room.

The GE design specification (22A1110) for the RBCCWS states that the primary containment isolation valves shall meet the GE design specification (22A1169) for the Reactor Containment. This latter design specification states that " lines which penetra'e but do not open into the drywell, and whose external branches do not terminate in dead end service capable of withstanding drywell design conditions, shall utilize one remotely operable isolation valve or check valve (example: closed cooling water lines)." No requirements for the closed line inside of containment were given.

The RBCCWS isolation capability meets the GE specification. Also, the RBCCWS isolation valve provisions are described in the FSAR and meet FSAR criteria (Section 5.2.3.5).

Based on the review of the RBCCWS design criteria and the i FSAR, it is apparent that there was not a strict definition of what constituted a qualified closed loop system at the time that the RBCCWS was designed. In reviewing the adequacy of the existing l RBCCWS as a qualified closed loop system, in meeting the I

requirements of SRP 6.2.4 the following criteria are met. l l

1) The system does not communicate with either the reactor )

coolant system or the containment atmosphere. l

2) The system is protected against missiles. '

_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ D

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.4

3) The system is designated seismic Category 1.

5) The system is designed to withstand temperatures at least g

equal to the containment design temperature.

6) The system is designed to withstand the external pressure from the containment structure acceptance test.

7) The system is designed to withstand the loss-of-coolant accident transient and enviroment.

The following are items of exception to SRP 6.2,4 for the RBCCWS.

2) The system is not protected against pipe whip.

4) The system is not classified as Safety Class. The RBCCWS was built to B31.1.

Additionally, '3E has since reviewed the RBCCWS design and published its conclusion as NEDC 22253. Based on the following, GE found the RBCCWS design sufficient, a) At least one valve in each line close to containment.

b) Isolation valves are automatic or capable of remote-manual operation. ,

c) Power supplies and controls are safety grade. '

d) Piping system out to and including isolation valves are safety grade.

l l

i I

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i

4 ENCLOSURE 4 OTIIER DEFINED BASIS FOR LACK OF AUTOMATIC ISOLATION CAPABILITIES FOR FCV-84-19, FCV-84-8A, -8B.-8C,-8D, AND FSV-43-28A, -288, -29A,-29B l

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i Browns Ferry's Primary Containment Isolation design provides for other means than automatic isolation signals to ensure primary containment integrity for the subject valves. These other means include:

1) For FCV-84-19, a keylock switch must be manipulated to open the valve. This key is controlled by the Shift Operations Supervisor who provides a restrictive administrative control on its use. Additionally, procedures do not allow for the use of this valve except under accident conditions and for valve operability testing. These procedures (01 and EOls) are restrictive as described in FSAR section 5.2.6.2.

2) For FCV-84-8A, -8B, -8C, and -8D, manipulation of their control switch will result in Control Room annunciator for " Valves Misaligned." This will alert the operator to realign (close) the valves to their correct alignment. Procedurally, these valves are only manipulated post accident as described in FSAR section 5.2.6.2.

3) For FSV-43-28A, -28B, -29A, and -29B operation of these sampling solenoid valves by use of a local pushbutton which must be held pushed to maintain the valves open. This provides the basis for maintaining primary containment integrity. The valves, pushbutton, and sample station are all located in the same rooms. Failure to close these valves after

f 4

sampling is not credible since the operator only has to release the pushbutton and if the button would fail to spring back, the operator would be uable to remove his sample container without obvious spillage. Additionally, a manual valve is maintained normally closed in the sa:nple line and is procedurally returned to normal.

Installation of automatic isolation signals to these valves would require considerable expense in design and construction and provide little added assurance of containment integrity for these very seldom manipulated valves.

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Jj ENCLOSURE 5 TVA DESCRIPTION / JUSTIFICATION FOR DELETION OF HPCI AND RCIC INJECTION MOVS FROM TECHNICAL SPECIFICATION 3.7 TABLES i

I:-

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-____________________:______.___._m.n'.___

i The HPCI and RCIC Injection valves should not be considered as Containment Isolation valves for the following reasons:

1) The use of the feedwater inboard check valves 3-558 and 3-572 in conjunction with the HPCI and RCIC Injection testable check valves, FCV-73-45 and FCV-71-40, folly meet Browns Ferry licensing commitments by having redundant valving (AEC Interim GDC 53). These lines are further described in the FSAR in section 5.2.3.5 exceptions, which says "Infuent lines, such as feedwater lines which connect to the reactor venel, have one check valve inside and one check valve or motor operated isolation valve outside the primary containment."

Further clarification of the original design intent can be found in G.E.s Design Specification 22All69 for Browns Ferry's Primary Containment which stated "On reactor system supply lines, where feed contmity is required, the two isolation valves shall be check valves which close automatically on reverse flow."

2) These valves will be opened automatically, not closed, by the conditions for which Primary Containment Isolation is initiated.

Upon low reactor water level or high drywell pressure signals, FCV-73-44 opens to allow HPCI injection to the reactor. FCV-71-39 opens on low reactor water level to allow RCIC injection to the reactor. These valves have no automatic closure capabilities and can only be closed by operator action after the initiating signals are cleared.

___-____-__a

3) These motor operated valves (MOVs) are not close to the area where the lines enter Primary Containment. For HPCI, the distance is approximately 200 ft. For RCIC, the distance is approximately 85 ft. This is contrary to the Browns Ferry FSAR (para. 5.2.3.5.b) and 10CFR50, Appendix A, Criteria 56 which 1

states valves outside containment shall be located as close to the primary containment wall as practical.

1 l

l 1

ENCLOSURE 6 CHARACTERIZATION OF BROWNS FERRY NUCLEA R PLANT APPENDIX J TEST PROGRAM FOR CLOSED SYSTEM NOTE: The following is Enclosure B of a letter from R u 'iley, TVA Manager of Licensing, to O. Muller, NRC Pro);ct Manager, dated February 24,1986 (L44 860224 806) 1

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CHARACTERIZATION OF BROWNS FERRY NUCLEAR PLANT (BPN) !

APPENDIX J TEST PR00 RAM FOR CLOSED SYSTEtiS

!. _In'. rod uction On August 28, 1985, TVA cet with NRC in Bethesda, Maryland, to present 'the

~

results of BFN's Appendix J program review. At this meeting, NRO expressed concern regardin6 BFN's policy for leak testing closed systems that interface with the reactor coolant boundary and the primary containment atmosphere. NRC requested a characterization of the Residual Heat Removal (RHR) and Core Spray (CS) systems to allow an esseccment of how BF5's progra: compares with Appendix J requirements. NBC is expected to use ;

this information to determine if BFN's policies are acceptable.

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Page 2 Enclosure B

-II. LBasis for Leak Testing of Closed Systems The basic design intent for the RHR and CS systems was .for the piping to be considered an extension of primary containment (i.e., a closed system) outside primary containment. As such, this piping is TVA class D and is seismic class 1. Failurs of the pressure boundary of these systems concurrent with a Design Basis Accident (DBA) Loss of Cooling Accident (LOCA) is not a design basis for BFN. The BFN design basis includes 1

assu=ing a single failure of an active mechanical component. Single i failure.of a passive mechanical component is not included as "no requirement" to assume such was defined when BFN was licensed. BFN's Final

- v, Safety Analysis Report defines passive component as "a device characterized by an expected negligible change of state or negligible mechanical motion D in response to an imposed design basis load' demand upon the system.

wm Examples are: cable, piping, valve in stationary position, resister, capacitor, fluid filter, indicator la=p, cabinet, and case." Thus, under the design basis, the EHR and CS systems outside containment were assumed i

to caintain their pressure boundary integrity during and after occurrence of the DBA LO;A. Using this rationale, no isclation need be provided except at high-low pressure interfaces and to preclude flow diversien.when switching suction and discharge flow paths. Because of the single active failure assumption, double isolation is provided where high-low pressure .

- interfaces and the emergency core cooling / containment cooling flow paths are involved. Otherwise, only a single boundary or barrier is relied upon. $

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' Enclosure B' LII. 1 Basis,for Leak Testing of C30 sed Systems (Continued) 6" The' RHR and' CS ' pump. suction and miniflow lines are lines which do not (by-The RHR and CS pump design) have primary. containment isolation valves.

suction tsystem isolation)? valves do not receive a postaccident signal to close; their safety function is in the_open position. -Also,~any seat

, . leakage would be'within primary containment. The RHR and CS miniflow hM (system isolation) valves are only designed to remain open during pump.

starting conditions; however, seat Icakage past these valves would also be

into primary' containment.

ZZZ . The RHR System.

A. RHR System Description The objective of the RBR system is to restore and maintain the coolant-g' . inventory in the reactor vissel so that the core is adequately cooled

\ '"' after a LOCA. The RHR system also provides containment cooling so that condensation of the steam, resulting from the blowdown of the design basis LOCA, is assured. The RHR system is designed for five-modes of cperatien to satisfy all the objectives and basis. The modes are:

1. Shutdown Cooling and Reactor Vesss1 Head Strty The shutdown cooling and reacter vessel head spray subsystem is placed in operation during a normal shutdown and cooldown.

Reactor coolant is pu= ped by the RER system main system pumps from one of the recirculation loops through the RER system heat exchangers, where cooling takes place by transferring heat to the-service water.

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" Page 4 Enclosure B III. The RHR System (Continued)

A. RUR System Description (Continued)- ,

1. Shutdown Cooline and Reactor Vessel Head Spray (Continued).

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Reactor coolant is returned to the reactor vessel"via either recirculation loop. Part of.this flow may be' diverted to a spray no::le in'the reactor head on unit 1 only. However, the. head spray ncs:le piping will be' removed during an upcoming refueling outage.

,6:

During a nuclear system shutdown and cooldown,' whenithe shutdown cooling subsystem is initially placed'in operation, decay heat levels can be high and operation of two.RHR system. heat exchangers may be required to remove the heat. 'When the decay heat level has decreased sufficiently, the entire stutdown

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cooling load can be handled by one RHR system heat exchanger..

2. Containment Cooling The containment cooling subsystem is placed in operation to limit the temperature of the water in the suppression pool so that immediately after the design baFis LOCA has occurred, pool temperature does not exceef. 175 F.

U Wsth the RHR systc= in the containment cooling mode of operation, the RHR syste cain system a

pumps are aligned to pu=p water frc: the supprescien pool through; -

- the RER syaten heat exchangers where cooling takes place by transferring heat to the RHR service water. For adecuate containment cooling, a minimum of two RHR pu=ps and associated heat exchangers =ust remain available for sever 01 'acurs a.*ter a design basis LOCA. The flow returns to the suppression pool via

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l Page 5 Enclosure B III. The RHR System (Continued)

A. RHR System Description (Continued)

2. Containment Cooling (Continued) the flow test line. The containment cooling subsystem hiso provides additional redundancy to the Core Standby Cooling systems for postaccident conditions. The water pueped through the RHR system heat exchangers may be diverted to spray headers in the drywell and above the supprer.sion pool. The spray headers in the drywell et , dense any steam that may cxist in the drywell, thereby lowering containment pretsure. The spray collects in the bottom of the drywell until the water level rises to the IcVel of

,, the pressure suppression vent lines, where it overflows and

( drains back to the suppression pool.

3 Low Pressure Coolant Injection (LPCI)

The LPCI subsystem operates to restore and, if necessary, maintain the coolant inventory in the reactor vessel after a LOCA so that the core is sufficiently cooled.

During LPCI operation, the RHR system pumps take suction frem the .

l Suppression pr,01 and discharge to the reactor vessel into the core recien through both of the recirculation loops. Two pumps l

discharge to occh injection header, assuring flooding of the l 1

vessel through at least one loop. Any spillage through a break in the lines within the primary containment returns to the suppression pool through the pressure suppression vent lines. A I bypecs line to the suppression pool is provided so that the pumps are not damaged if operating with the discharge valves shut.

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III. The.RHR System (Continued).

A. RHR System Description (Continued) 4 i

2. . Containment Cooling (Continu$d) !

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[' the flow test line. The containment. cooling subsystem also provides additional redundancy to the Core Standby Cooling systems for postaccident conditions. The water pumped through.

the RHR system heat exchangers may be diverted to spray headers -

in the drywell and above the suppression. pool. The spray-headers in the drywell con' dense any.cteam that may exist in the drywell, thereby lowering containment pressure. The spray collects in. the - j bottom of the drywell unt'l the water level rises to the '1evel of.

the pressure suppression vent lines, where it overflows and '

Aq drains back to the suppression pool.

3 Low Pressure Coolant Injection (LPCI)

The LPOI subsystem operates to restore and, if necessary, maintain the coolant inventory in the reactor vessel after a LOCA so that the' core is sufficient 1/ coo]ed.

During LPCI operation, the RHR aystem pumps take suction frem the suppression pool Ln- discharge to the reactor vessel into the core region through both of the recirculation locps. Two pumps discharge to each injectier. header, assuring flocding of the vessel through at least one loop. Any spillage through a break in the lines within the pri=ary containment returns to the suppression pool through the pressure suppression vent lines. A bypass line to the suppression pool is provided so that the pumps

--- are not damaged if operating with the discharge valves shut.

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-Enclosure B-

. III. The RHR System (Continued)

A. RHR System Description (Continued)

4. Standby Cooling Standby coolant supply connection and'RHR crosstics are provided to maintain a long-term reactor core and primary containment '

cooling capability irrespective of primary containment. integrity -

or' operability of the RHR' system associated with a given unit.

By proper valve alignment the network created by the'RHR crossties permits the B (or D) RHR pumps on unit 1 to circulate unit 2 suppression pool or reactor vessel water through the B'-

(or D). heat exchangers on unit 1 in the unlikely event' that'the!

- unit 2 RHR. pumps are unavailable.

k, In a like fashion, the A (er C) RHR pu=ps en unit 2 can be used-l to circulate unit 1 supreession. pool or reacter vessel water through the A (or C) heat exchangers en unit 2. The B (or D) RHR pu=ps on unit 2 and the A (or C) RHR pumps en unit 3 can be '

similarly utiliced. Suppression pool water which has been circulated through the RHR heat exchangers on one unit can be used to flood the reactor cere, spray the drywell and suppression chamber, or returned to the suppression cha=ber of the adjacent unit. In this way, decay heat and residual heat can be removed l from the reactor core and primary containment of the adjacent

1. .

unit on a long-ter basis. B7 proper valve alignment, the netwerk created by the standay coolant supply connection and RHR cresstics per its the D2 (cr D1) RHR service water pump and header to supply raw water directly to the reactor core of

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Page 7' Enclosure B

j. III.lThe RHR System.(Continued)

R- I ~

'. A . RHR' System Description ,(Continued) 11 . Standby Cooling (Continued)

. units 1 or 2 as the reactor pressure approaches'50 psis. The service water p' ump and header can also be valved to supply raw Lwater to the drywell,or suppression chamber spray headers or.

directly to.the suppression. chamber of either unit.: In a'similar fashion, the B2 '(or E1) RHR service water pump and header.can supply raw water-to the reactor core of unitcc,2 or 3 or into-the.

respective suppression chamber.

5. ~ Supplemental Fuc2 Pool Cooline (FPC)

The RHR system may be used to provide supplemental coo 11ng for l the FPC system. The suction from the skimmer surge tank suction line to the FPC pump ties in to the RHR shutdown cooling supply line. FPC then takes the path of RHR shutdown cooling, where it.

is discharged to the FPC purp rr, turn line to the fuel pool.

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9 Page B Enclosure B III . The RHR System -(Continued) .

B. Aphendix J Test Program for the RHR System - Primary Containment Penetrations Tested The following RHR lines (primary containment penetrations) are leak tested in accordance with Appendix J.

Description Valve No.

Shutdown Suction 47,48 LPCI Injection - Loop 1 53,54 Cont. Spray (Suppression Pool) - L,oop 1 57,58 Cont. Spray (Drywell) - Loop 1 60,61 LPOI Injection - Loop 2 67,68 Cont. Spray (Suppression Pool) - Loop 2 71,72 pr~

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Cont. Spray (Drywell) - Loop 2 74,75 Head Spray 77,78 RHRS Vent 102,103,119,120 Shutdown Suction 661,662 Suppression Pool Drain 722 PSC Head Tank Tie-In Loop 1 792,804 PSC Head Tank Tie-In Loop 2 802,803 i

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Enclosure B III. The-RHR System (Continued)

B. Appendix J Tes't Program for the RHR System - Primary Containment Penetrations 1ested (Continuec)

The'BHR vent system discharges to the suppression pool. The. vent' H system conts'nment penetration line' requires modification to ~ allow type C testing isoTation valves'(102, 103, 119, 120). The current.

outage schedule implements the modifications on units 1 and 3 prior to j- the startup of cycle 7.

Unit' 2 modifications w'ill be implemented during the next refueling outage (cycle 6). The RHR vent i ststem isolatir, valv'es (102i 103, 119, and 120)'will'be added to the Appendix J program upon completion of'the modifications.

0.- 'Other Primary Containment Penetrat' ions All remaining primary containment penetrations are considered branch lines, and are isolated by normally' closed (passive) water sealed valves. As previously discussed, passive valves are assumed notuto fail and therefore are not included in the. Appendix J program. .The water seal is provided by the supp"ession poc1 inventnry. The following is a description of the RHR system branch lines that are isc1sted by normally clcsed (passive) water sealed boundar!.es which-are not leak (seal leakage) tested.

1. The EHE vent system incintes 17 lines (14 1-inch, two 2-inch, one 1-1/2-inch) which connect piping and equipment to the suppression pool to allow venting during filling cperations.

Each'line is isolated by a normally closed (passive), water sealed valve.

1 I

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w . '( Page'10 r(" Enclosure B III. The RHR System'(Continued)

L C. Other Primary Containment Penetrations (Continued)

(- 2. The RHR system flush pumps are used for flushing the R!!R system prior to entering into the shutdown cooling mode. The flush pumps (two) are capable of taking suction from 15 system tie-ins (four 6-inch, six 4-inch, one 2-inch, four 1-inch). Also, the

' flush pumps may discharge to the system at six system tie-ins (two 6-inch, four 4-inch). Eeh EHR system tie-in is isolated by n normally locked closed'(passive), water sealed valve.

3 The RHR system has eight 1-inch drain lines thht discharge to the; reactor building equipment drain sum;,, Each line is isolated by a normally closed (passive), water sealed valve.

k,. 4. The RHR system has twelve 1-inch relief valves that discharge to.

the reactor building equipment drain sump. Each relief valve setting is verified in accordance with ASME Secticn U . Each relief valve is also waser sealed.

5 As previously discussed, the RHR system can be used for Supplemental Fuel Pool Cooling (SFPC). The SEPC supply and return tie-int to the RHR system are 8 and 6-inch lines respectively. Each line is isolated by normally locked closed, water sealed valves.

1

6. The condensate storage tank supplies flushing and charging water .

a for the RER system through two 1-inch, five 4-inch, and four 16-inch lines that are isolated by normally closed (passive), 1 l

water scaled valves wnich would remain closed postaccident.

1

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t Page 11 Enclosure B III. The RER S3 stem (Continued)

C. Othe.' Primary Contain=ent Penetrations (Continued)

7. The main condenser and radwaste provide flushing drain paths for the RHR system through two 6-inch lines. Eachlineisisolatedby l normally locked closed (passive), water sealed valves which would remain closed postaccident.

B. The RHR pumps and heat exchangers are cross-connected to adjacent units through two 10-inch lines and four 12-inch lines. Each are isolated by'normally closed (passive), water sealed valves which would remain closed during a DBA.

9 Other EHR system branch lines are isolated by norma 12y closed (passive), water sealed valves and are visually inspected at P a b.. or greater once per cycle for seat leakage. In other words, these lines are open ended, and any seat leakage would be visible. All other leak paths (i.e., bonnets, packing, orifice flanges) are also visually inspected at P or greater once per cycle for out (of system) leakage.

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Page 12 N . <

Enclosure B

)

I g IV. CS System 'l?

A. CSSystemDescriph. ion >

~

The purpose of.the CS system is to protect -against' overheating the

.. -j fuel in the event'of a LOCA. This would be accomplished by spraying 4 i

water directly on t' he fuel from a spray'sparger located within the shroud. There are two independent, redundant, 100-percent capacity CS-

~

loops. Each CS loop has two pumps which norma 11y take suction. from the suppression pool and discharges to the reactor vessel.through the CS sparsers. An alternate pump suction supply is the. condensate storage tank.

B. . Appendix J Test Program for the CS Systc=-

The following CS lines (primary containment penetrations) are leak

_[

. k'.s tested in accordance with Appendix J. )

Description Valve'Ho. ,

l CS Imjection - Loop 1- 25/26 l CS Injection - Loop 2 53/54 1

c Level Control Suppression Tool His' 57/5B

] i PCS Head Tank Tie-In - Ltep 1 606/607 j l

PCS Head Tank Tie-In - Loep 2 609/610 l l

C. Other Pfiyayv Containment Penetret$ons

]

All remaining primary containment penetrations are censidered branch 1.

lines, and are isolated by normally closed (passive) water sealed I valves. The water seal is provided by the suppression pool inventory. As previously discussed, passive valves are assumed not to

/ fall and are therefore not included in the Appendix J program. The l

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, ~ Page 13 Enclosure B-IV.~ CS System (Continued) -

C.- Other Primary Containment Penetrations (Continued) following is a description of the1CS system branch -lines that are isolated by water sealed, passive valve' boundaries which'are'not leak

.(scal leakage). tested.

^

1. -The CS system discharges to .the reactor building equipment drain su=p from the following:

a. Four 1-inch relier valves

b. One 2-inch relief valve'

c. Three 1-inch piping vent valves

d. Two '. "./2-inch piping vent valves

e. Four 1-inch pump casing drains

-(f f. Four 3/t-inch pu=p discharge drains

s. Four 3/4-inch pump vents

h. Four 1/2-inch pump shaft seal leakage drains

2. Tne two CS syste= drain purpr take suction frt: the CS pumps supply and discharge pipinr,throust eight 3-inch lines. The drain pu=ps discharge to the CS pu:p suction lires through two 2-1/2-inch lines. Euch of the drain pump tie-ins to the CS system are isolated by ner: ally closed, water sealed passive, valves.

3 The CS system has an alternate supply source free the condensate-storage tank through two 14-inch lines which are normally isolated by lock closed, water sealed, passive valves.

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Page la

p. Enclosure B IV. CS System (Continued)

C. Other Primary Containment Penetrations (Continued) 4'. Other CS system branch lines are isolated by normally closed, water- scaled passive valves and are visually inspected at P,or greater once p'er cycle for seat. leakage. In other words, these lines are open ended and any seat leakage would be visible. All other leak paths (i.e. , bonnets, packing, orifice flarges) are also visually inspected at P r greater nee per cycle for a

out (of system) leakage.

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