Text

Provides Changes & Clarifications to Plant Pump & Valve Testing Program.Changes Reflect Plant Mods & Programmatic Reviews Performed During Unit 2 Cycle 6 Refueling Outage
	ML18037A532
Person / Time
Site:	Browns Ferry
Issue date:	10/29/1993
From:	Salas P; TENNESSEE VALLEY AUTHORITY
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NUDOCS 9311040386
	Download: ML18037A532 (61)
	v • d • e

ACCEI ERAT D DOCUMENT DIST UTION SYSTEM REGULA Y INFORMATION DISTRIBUTIO YSTEM (RIDS)

ACCESSION NBR:9311040386 DOC.DATE: 93/10/29 NOTARIZED: NO DOCKET FACIL:50-259 Browns Ferry Nuclear Power Station, Unit 1, Tennessee 05000259 50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH.NAME AUTHOR AFFILIATION SALAS,P.

Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Provides changes 6'larifications to plant pump a valve testing program. Changes reflect plant mods a programmatic reviews performed during Unit 2 Cycle 6 refueling outage.

DISTRIBUTION CODE:

A047D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code NOTES:

R D

RECIPIENT ID CODE/NAME PD2-4 TRIMBLE,D INTERNAL: ACRS NRR/DE/EMEB NUDOCS-ABSTRACT OGC/HDS3 RES/DSIR/EIB EXTERNAL: EGGG BROWNiB NRC PDR COPIES LTTR ENCL 1

0 2

2 6

6 1

1 1

0 1

1 1

1 RECIPIENT ID CODE/NAME PD2-4-PD WXLLXAMS,J.

AEOD/DSP/ROAB NRR/EMCB OC/LFDCB EG ~

01 EGGG RANSOMEiC NSIC COPIES LTTR ENCL 1

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2 1

1 1

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1 D

D S

D NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. S04-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDI D

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LTTR 24 ENCL 21

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609 OCT 29 l993 U.S. Nuclear Regulatory Commission Attn:

Document Control Desk Washington, D.C.

20555 Gentlemen:

In the Matter of

)

Tennessee Valley Authority

)

Docket Nos.

50-259 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 1 g 2 ~

AND 3 AMER1CAN SOCIETY OF MECHANICAL ENGINEERS (ASME)

SECTION ZI INSERVICE TEST PROGRAM CHANGES FOR PUMPS AND VALVES The purpose of this letter is to provide changes and clarifications to the BFN pump and valve testing (P&VT) program.

These changes reflect plant modifications and programmatic reviews performed during BFN's Unit 2 Cycle 6

refueling outage.

During the Unit 2 Cycle 6 refueling outage, certain modifications to plant systems and programmatic reviews required changes to the P&VT program.

Enclosure 1 to this letter contains a description of the program changes.

Enclosure 2 contains changes to the P&VT program based upon completed modifications and minor editorial changes.

Enclosure 3 contains corrections to BFN's P&VT program, transmitted by TVA's letter dated August 31, 1992.

These corrections are editorial in nature and do not change the testing described in the referenced letter.

The bar in the right-hand margin identifies the changes and/or corrections.

BFN requests your review of the revised P&VT program.

Specifically, BFN requests approval and a safety evaluation report for revised relief requests PV-11 and PV-14 contained in Enclosure 2.

O3~CGG g) l) 931104038b~931029 PDR ADOCK '.05000259 P

PDR

f

U.S. Nuclear Regulatory Commission Page 2

OCT 2 4 1993 If you have any questions, please telephone me at (205) 729-2636.

Sincerely, Pe alas Manager of Site Licensing Enclosures cc (Enclosures):

Mr. R. V. Crlenjak, Project Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637

Athens, Alabama 35611 Mr. J.

F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North, 11555 Rockville,Pike Rockville, Maryland 20852 Mr. D.

C. Trimble, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

=E i ~

ENCLOSURE 1

Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN)

American Society of Mechanical Engineers (ASME)

Section XI Inservice Test Program Descri tion Of The Pro ram Chan es 1.

Valves 2-CKV-12-738 and 2-CKV-12-741 were deleted per Design Change Notice (DCN) W17929.

This change deletes the Auxiliary Boiler tie-in to the Suppression Chamber.

As a result of the tie-in, these valves were Primary Containment isolation valves requiring periodic leak rate testing.

This tie-in was never utilized for its intended purpose (corrosion inhibition).

The pipe was cut,

capped, and abandoned in place.

This eliminates the need to test these valves.

2.

Valves24-886, 24-891, and 3-24-826 may be required to perform a safety function during certain situations.

These valves are being added to the testing program and to Relief Request (RR) PV-14.

Testing will be in accordance with the guidance provided in Generic Letter 89-04, position 2.

3.

Valves 2-FCV-64-221 and 2-FCV-64-222 were added per DCN W17491A (hardened wetwell vent).

These valves are part of the Hardened Wetwell Vent modification and are new containment isolation valves.

The safety position for these valves is open (for primary containment venting during an emergency) and closed (when primary containment is required).

The valves will be cycled, stroke timed, and fail-safe tested quarterly.

These valves will be leak rate tested on a once per operating cycle basis.

4.

Valves 0-CKV-67-504 and 0-CKV-67-505 were deleted per DCN W20321.

The Emergency Equipment Cooling Water (EECW) supply to the Standby Gas Treatment sump flush connection was cut and capped by this DCN.

This eliminates the function of these valves.

These valves are currently listed as passive valves in the BFN Inservice Testing (IST) Program.

This revised program deletes those valves.

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V i

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5.

Valve 2-CKV-69-630 was added per DCN W18298.

This valve was added to the IST program to address single active component failure criteria (adds a second in-line check valve to prevent blowdown of the B Feedwater line in the event of failure of the outboard Primary Containment isolation valve, 2-CKV-69-579). The safety position for this valve is closed.

This valve is in a system which is normally in service at all times; it. cannot be closure tested quarterly.

The revision to RR PV-11 incorporates this valve.

Additionally, testing will be in accordance with RR PV-11.

Because the valve is not a containment isolation valve it will not be leak rate tested per the Appendix J program.

It will be closure tested using Appendix J program procedures to pressurize a test volume with the check valve as a boundary.

Although not associated with Unit 2 Cycle 6, valves 3-CKV-69-628 and 3-CKV-69-629 have been added to the RR PV-11 as a result of Unit 3 DCN W18486.

These Unit 3 valves are similar to 2-CKV-69-630 and will be tested in the same manner before Unit 3 startup.

6.

Valve 2-CKV-84-617 was replaced with new valves 2-BYV-84-686 and 2-FSV-84-49.

Valve 2-CKV-84-680 was replaced with new valves 2-BYV-84-683 and 2-FSV-84-48.

DCN W17845 generated the above changes.

The previous configuration used two check valves in series for primary containment isolation which was considered unreliable.

This DCN replaced each of the outboard check valves with one solenoid-operated gate valve and one manual-operated bypass ball valve.

The valves have an open (backup supply for Drywell Control Air) and closed (Primary Containment isolation) safety position.

These valves are being tested in accordance with ASME Section XI requirements on a quarterly basis.

These valves will also be leak rate tested per Appendix J requirements on a once per operating cycle basis.

7.

DCN W12579 changed the valve type and actuator for valves 2-FCV-90-254A/254B/255/257A/257B.

This DCN replaced the existing motor operated ball valves with solenoid operated gate valves.

The function and safety positions of the valves are unchanged.

The valves'dentification will reflect the change from FCVs to FSVs.

Testing will remain the same.

In addition, credit will be taken for fail-safe testing required by ASME Section XI.

J'

8.

Cold Shutdown Justification 9 describes the testing for the control rod drive (CRD) charging water check valve.

This cold shutdown justification is being withdrawn since the testing of this check valve is described in relief request PV-21 with the entire CRD charging and discharge system.

9.

Valves 1-67-598 and 3-67-598 are being deleted from the IST program.

These valves are noncode class valves.

They are located in the emergency equipment cooling water to Residual Heat Removal/Core Spray system II room cooler discharge header piping to the yard drainage.

These valves will continue to be tested in accordance with BFN's maintenance program.

ENCLOSURE 2

Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN)

American Society of Mechanical Engineers (ASME)

Section XI Inservice Test Program Chan es to the Pum s and Valves Testin Pro ram P&VT

I 1

Enclosure 2 contains changes to BFN's August 31,

1992, P&VT program submittal.

The effective pages are:

23, 26, 40, 42, 49, 68, 71, 84, 87, and 117.

System: Auxiliary Boiler (12)

Drawing No. 47E815-3 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve

~Numb r Func ion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown

~1s

~Cordinates

~Cat stout ~in.

~T

~fgg

~Po ition ~Positi n

~Re uir d Justifi ation 23

System:

Raw Cooling Mater (24)

Drawing No. 47E844-2 and 47E859-1 BROMNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve Number Fn in Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 1-24-712A 1-24-7128 0-24-796 0-24-798 3-24-826 t 24-886 24-891 Wtr Chiller A Disch Isolation Mtr Chiller B Disch Isolation HVAC RH A Supply Check HVAC RH B Supply Check North EECM Hdr Check H202 Panel 25-340 Check H202 Panel 25-341 Check 3

E-8 3

F-2 3

F-1 3

G-2 3

A-6 3

A-6 6

GA H

6 GA H

0/C 0/C 2

CK SELF 0

2 CK SELF 0

1.5 CK SELF C

1 CK SELF C

0/C 0/C CV CV CV CV CV PV-14 PV-14 PV-14 PV-14 PV-14 26

J 4g

Valve

~Numb r Fun ion ASHE Drawing Si ze Valve Actuator

~la s ~rdina es

~Ca esiera ~in.

~ae

~T BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 System:

Reactor Building Heating, Ventilation, and Air Conditioning (64)

Drawing No. 47E865-12 (Unit 3) 47E865-1 (Unit 1), 47E2865-12 (Unit 2)

Relief Request/

Normal Safety Testing Cold Shutdown 64-28A 64-288 64-28C 64-28Dt 64-28E 64-28F 64-28G 64-28H 64-283 64-28K 64-28L 64-28H 64-139 PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief PSC to DW Vacuum Relief DW Cmpsr Inlet NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 NC A-4 2

H-2 18 CK SELF/AO C

18 CK SELF/AO C

18 CK SELF/AO 18 CK SELF/AO C

18 CK SELF/AO C

3 GL AO 0

CV 0

CV C

Q,ST LT 64-140 DW Cmpsr Disch 2-64-222 Containment Isolation 2

H-2 2

A-7 2

A-8 3

GL AO C

14 BF AO C

Q,ST LT 0/C Q,ST,LT FS 0/C Q,ST,LT FS 40

BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 System:

Emergency Equipment Cooling Water (67)

Drawing No. 47E859-1 Valve

~Mmber Fun ion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown Cllas ~Cordina s

~Cat deCT ~in.

~T

~Te

~Psition ~Pition ~R<iuuir d Justifi ation 0-67-508 0-67-514 0-67-515 0-67-521 t

o-67-522 0-67-528 0-67-529 0-67-538 10 DG Cooler S Hdr Check 1C DG Cooler S Hdr Check 1C DG Cooler S Hdr Check 1B DG Cooler S Hdr Check 1B DG Cooler S Hdr Check lA DG Cooler S Hdr Check lA DG Cooler S Hdr Check Control Rm Chiller S Hdr Supply Check 3

C-8 3

0-7 3

D-8 3

F-7 3

F-8 3

G-7 C

3 G-8 3

G-7 4

CK SELF 0/C 0/C CV 4

CK SELF 0/C 0/C CV 6

CK SELF C

0/C CV PV-14 PV-14 PV-14 PV-14 PV-14 PV-14 PV-14 PV-14 0-67-539 67-541 67-542 67-558 67-559I67-584 67-585 Control Rm Chiller S Hdr Supply Check Loop I CS Rm Cooler S Hdr Supply Check Loop I CS Rm Cooler S Hdr Supply Check Loop I RHR Rm Cooler S Hdr Supply Check Loop I RHR Rm Cooler S Hdr Supply Check Loop II CS Rm Cooler S Hdr Supply Check Loop II CS Rm Cooler S Hdr Supply Check 3

G-7 3

F-7 3

C-6 3

E-5 3

F-5 6

CK SELF C

0/C CV C

2-1/2 CK SELF 0/C 0/C CV C

2-1/2 CK SELF 0/C 0/C CV 3

CK SELF 0/C 0/C CV 3

CK SELF 0/C 0/C CV C

2-1/2 CK SELF 0/C 0/C CV C

2-1/2 CK SELF 0/C 0/C CV PV-14 PV-14 PV-14 PV-14 PV-14 PV-14 PV-14 42

I/

System:

Reactor Water Cleanup (69)

Drawing No. 47E810-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve

~Numb r F n tion Relief Request/

ASHE Orawing Size Valve Actuator Normal Safety Testing Cold Shutdown 69-1 69-2 69-579 3-69-624 t

2-69-630 3-69-628 3-69-629 Inbd Isol Outbd Isol RWCU to FW Check RWCU to FW Check RWCU to FW Check RWCU to FW Check RWCU to FW Check 1

F-7 1

G-6 1

E-6 1

F-6 1

E-6 1

F-6 1

E-6 AC AC 6

GA HO 6

GA HO 4

CK SELF 0

4 CK SELF 0

Q,ST,LT tl,ST,LT CV,LT PV-11 CV,LT PV-11 CV,LT PV-11 CV,LT PV-11 CV,LT PV-11 49

gX p SrI

System:

Containment Atmosphere Dilution (84)

Drawing No. 47E862-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 Valve Number Func ion Relief Request/

ASME Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 84-8A 84-BB 84-8C 84-BDt 84-19 84-20 84-600 84-601 84-602 84-603 2-84-48 2-84-49 2-84-683 2-84-686 DW N2 Supply Train A PSC N2 Supply Train A PSC N2 Supply Train 8 DW N2 Supply Train B

DW/PSC Disch to SBGT DW/PSC Disch to SBGT DW Supply Check A PSC Supply Check A DW Supply Check B

PSC Supply Check B

Containment Isolation Containment Isolation Containment Isolation Containment Isolation 2

E-5 2

G-6 2

G-5 2

E-7 2

E-5 2

F-6 2

F-4 2

F-7 2

F-4 A

2 E-7 A

2 GA S

2 GA AO 2

GA AO 2

CK SELF C

2 CK SELF C

1 GA S

1 BA H

0/C Q,ST,LT PV-23 0/C Q,ST,LT PV-23 0/C Q,ST,LT PV-23 0/C ',ST,LT PV-23 0/C Q,ST,LT 0/C Q,ST,LT 0/C Q,LT 0/C Q,LT 0/C Q,LT 0/C Q,LT 0/C Q,LT,ST,FS 0/C Q,LT,ST,FS 0/C Q,LT 0/C Q,LT 68

System:

Radiation Honitoring Drawing No. 47E610-90-1 BRSINS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve Number Fun tion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown gila

~Crdi~na: ~at ~ot ~in.

~T

~T>~ ~pi ion ~uoi ion ~Ruir d Ju tifioation 90-254A 90-254S90-255 90-257A IN Leak Det Upper Inbd Isol DW Leak Det Lower Inbd Isol DW Leak Det Intake Outbd Isol DW Leak Det Outbd Isol DM Leak Det Inbd Isol 2

H-2 A

2 G-2 A

2 H/G-2 A

2 H-2 A

1 GA S

0 1

GA S

0 1

GA S

0 GA S

0 1

GA S

0 0/C 0/C 0/C 0/C 0/C Q,ST,LT FS Q,ST,LT FS Q,ST,LT FS Q,ST,LT FS Q,ST,LT FS 71

0 l

P LI k'I Q

Relief Request No. PV-ll System:

Drawing:

Components:

Category:

Reactor Hater Cleanup (RWCU) (69) 47E810-1 Valves69-579, 2-69-630, 3-69-624, 3-69-628, 3-69-629 A(69-579, 3-69-624),

C Class:

Function:

Containment isolation (69-579, 3-69-624); prevents backflow into RWCU System.

Impractical Test Requirement:

IHV-3521 and 3522 Exercise valve closed every three months.

Basis for Relief:

These check valves remains open to return water to the reactor vessel whenever the reactor water cleanup system is operating.

These valves are not testable whenever the RHCU, feedwater/condensate, HPCI (Unit 3 only), or RCIC system is returning flow to the reactor vessel.

Testing requires entry into primary containment and the disruption of system flow (RHCU, feedwater/condensate, HPCI or RCIC).

For these

reasons, closure testing is only practical during extended outages such as refuelings during which these systems are shutdown.

Also, plant design does not provide a practical means of demonstrating closure other than by upstream pressurization performed during leak rate testing conducted in accordance with 10 CFR 50 Appendix J.

This testing involves significant effort for installation of temporary equipment.

This would require valve lineups to abnormal positions, installation of pressurizing equipment and associated test lines, as well as deinerting the drywell for safe entry.

Alternative Testing:

Proper valve closure for containment isolation valves69-579 and 3-69-624 will be verified by completion of local leak rate testing performed in accordance with 10 CFR 50 Appendix J.

Valves 2-69-630, 3-69-628, and 3-69-629 will be closure tested during the course of local leak rate testing by demonstrating the ability to pressurize a test volume that includes the valves as a boundary.

84

Relief Request No. PV-14 System:

Emergency Equipment Cooling Water (EECW) (67)

Raw Cooling Water (RCW) (24)

Drawing:

47E844-2 47E859-1 47E859-2 47E866-7 Components:

Valves67-541,

542, 558,

559, 584,

585, 600,

601, 638,

639, 648,

649, 656,

657, 659, 660 0-67-502,

507, 508,

514, 515,

521, 522

528, 529 538 539

619, 622,

624, 625,

627, 628,

630, 631,

634, 635,

652, 653,

671, 679 1-67-787, 789 2-67-873, 876 3-67-693,

694, 695,

696, 703,

704, 705,

706, 713,

714, 715,

716, 723,

724, 725,

726, 735,

736, 737,

738, 761,

762, 764,

765, 771,

772, 774, 775 24-886, 891 0-24-796, 798 3-24-826 Category:

Class:

3 and Non-Code Class Function:

EECW:

Pass rated flow for the emergency coolers and prevent backflow from opposite header.

RWC:

Prevent backflow of EECW into RCW.

Impractical Test Requirement:

IWV-3521 and 3522 Exercise valves closed every three months.

Basis for Relief:

System design prevents the valves from being verified closed by reverse flow or other conventional means.

Therefore, closure verification by disassembly will be required.

Since the valves can be verified open quarterly by flow verification, apparent disc free movement will be indicated.

Alternative Testing:

Valves will be proven to close once per refueling cycle by disassembly and inspection.

This will be performed on a rotating basis in accordance with Position 2 in NRC Generic Letter 89-04.

87

Cold Shutdown Justification Number 9

WITHDRAWN

-117-

ENCLOSURE 3

Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN)

American Society of Mechanical Engineers (ASME)

Section ZZ Inservice Test Program Corrections to the Pum s and Valves Testin PSVT

Enclosure 3 contains corrections to BFN's August 31,

1992, P&VT program submittal.

The effective pages are:

7, 18, 21, 22, 36, 39, 44, 45, 51, 52, 53, 54, 55, 69, 75, 76, and 78.

I

'I

BROWNS FERRY NUCLEAR PLANT PUHP TEST PROGRAH UNITS 1, 2 AND 3 Inlet Differential Flow Vibration Bearing

~eum

~Cla

~Drawin ¹

~Sed

~pr ssure

~pr ssure Rate a~mitted

~Tem era ure RHR (4) 2 47E811-1 NR Q

Q Q

Q PV-1 HPCI (1) 2 47E812-1 Q

(}

Q Q

Q PV-1 Observe Lube Oil Level Pressure PV-1 RCIC (1) 2 47E813-1

(}

Q Q

PV-1 CS (4) 2 47E814-1 NR Q

Q PV-1 PV-1 SLC (2) 2 47E854-1 NR PV-2 PV-2 Q

Q PV-1 RHRSW (12) 3 47E858-1 NR Q (Note 2)

Q Q

PV-1 47E859-1 PV-1 Diesel Fuel NC 47E840-3 Transfer (16)

NR PV-2,5 PV-2,5 PV-2,5

(}

PV-1,3 PV-1

~NOTE:

1.

Number in (

) denote number of pumps per unit except for RHRSW and diesel fuel transfer.

RHRSW pumps are comnon.

There are eight diesel fuel transfer pumps for Units 1 and 2

diesel generators and eight for Unit 3 diesel generators.

2.

River level will be recorded for suction pressure for the RHRSW pumps.

3.

Legend NR Not required to be measured by Section XI Q (}uantity measured in accordance with Section XI requirements PV - Alternative testing specified in relief request in Part III of this program

System:

Hain Stream (1)

Drawing No. 47E801-1 Valve Number Functi on BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 Relief Request/

ASME Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 1-41 1-42 1-51 1-52

~ 1-55 1-56 1-179 1-180 Line D Relief Line D Relief Line D Inbd Isol Line D Outbd Isol Drain Inbd Isol Drain Outbd Isol Line A Relief Line D Relief 1

E-7 1

E-6 1

E-5 1

E-4 1

F-5 1

F-4 1

G-6 1

E-6 6

RV SELF C

6 RV SELF C

26 GL AO 0

3 GA HO C

6 RV SELF C

RV RV Q,ST LT, FS Q,ST LT,FS Q,ST LT Q,ST LT RV RV PV-6 (unit 3 only)

PV-27, 32 CS-6 CS-6 PV-20 18

/I I

I I

System:

Hain Stream RelieF Valve (10)

Drawing No. 47E817-1 BROGANS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 Valve Number Fun tion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown

~Clan Qyrdi~na t CatesLora ~in.

~Te

~Tpg Position

~Po i i n ~Ruired dustifi ation 10-506 10-507 10-508 10-509t 10-510

'10-511 10-512 10-513 10-514 10-515 10-516 10-519 10-520 10-521 10-522 10-523 10-524 10-525 10-526 10-527 10-528 Tail Pi pe A Vac Rel ief Tail Pi pe B Vac Rel ief Tail Pipe C Vac RelieF Tail Pipe D Vac RelieF Tai 1 Pi pe E Vac Rel i ef Tail Pi pe F Vac Rel ief Tail Pipe G Vac Relief Tail Pi pe H Vac Rel ief Tail Pipe J Vac Relief Tail Pi pe K Vac Reli ef Tail Pipe L Vac Relief Tail Pi pe H Vac Rel ief Tail Pi pe N Vac Rel ief Tail Pi pe A Vac Rel ief Tail Pipe B Vac Relief Tail Pi pe C Vac Rel ief Tail Pi pe D Vac Rel ief Tail Pipe E Vac Relief Tail Pi pe F Vac Rel i ef Tail Pipe G Vac Relief Tail Pipe H Vac Relief 3

E-1 C

3 0-1 C

3 E-1 C

3 0-1 C

3 E-1 C

3 D-1 C

3 E-1 C

3 0-1 C

3 D-l C

3 E-1 C

3 D-1 C

3 E-1 C

3 D-1 C

3 E-1 C

2-1/2 CK SELF C

10 CK SELF C

CV CV CV CV CV CV CV CV CV CV CV CV CV CV CV CV CV CV CV PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 PV-7 21

System:

Hain Stream Relief Valve (10)

Drawing No. 47E817-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve Number Func i n

Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 10-529 10-530 10-531 10-532 10-533 Tail Pi pe J Vac Rel ief Tail Pipe K Vac Relief Tail Pipe L Vac Relief Tail Pipe H Vac Relief Tail Pipe N Vac Relief 3

D-l C

3 E-1 C

3 D-1 C

10 CK SELF C

CV CV CV CV PV-7 PV-7 PV-7 PV-7 PV-7 22

System:

Sampling and Water guali ty (43)

Drawing No. 47E610-43-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve Number 43-13 Function Reactor Recirc Inbd Isol Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 1

H-4 A

3/4 GL AO C

C LT,Q 43-14 43-28A 43-28B 43-29A 43-29B Reactor Recirc Outbd Isol Reci rc Sample Isolation Recirc Sample Isolation Reci rc Sample Isolation Reci rc Sample Isolation 1

G-4 2

E-3 2

D-3 2

E-2 2

D-2 3/4 GL AO 1/2 GL S

1/2 GL S

1/2 GL S-LT,Q LT LT LT LT 36

r e

'I 'I p

BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 System:

Reactor Building Heating, Ventilation, and Air Conditioning (64)

Drawing No. 47E865-1 Valve

~Mmber F nc ion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown P"'4-17 64-18 64-19t 64-20 DW Supply Isol DW Atm Supply Inbd Isol PSC Atm Supply Inbd Isol PSC Vac Relief Isol 2

C-5 20 BF AO 2

C-6 2

C-5 20 BF AO 20 BF AO 2

C-6 A

20 BF AO C

C Q,ST LT Q,ST LT Q,ST LT Q,ST LT 64-21 64-29 64-30 64-31 64-32 64-33 64-34 64-800 64-801 PSC Vac Relief Isol DW Exhaust Inbd Isol PSC Vac Relief Isol DW SBGT Inbd Isol PSC Exhaust Inbd Isol PSC Exhaust Outbd Isol PSC SBGT Inbd Isol PSC Vac Relief Ck PSC Vac Relief Ck:

2 B-5 2

E-3 2

C-2 2

C-5 2

C-5 20 BF AO 18 BF AO 18 BF AO 2

BF AO 18 BF AO 18 BF AO 2

BF AO 0

20 CK SELF C

0/C 0/C Q,ST LT Q,ST LT Q,ST LT Q,ST LT Q,ST LT Q,ST LT Q,ST LT CV, LT CV, LT 39

BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 System:

Emergency Equipment Cooling Water (67)

Drawing No. 47E859-1 Valve

~Numb r Func ion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown r'-67-653 Control Rm Chillers N Hdr Supply Check 3

G-7 C

6 CK SELF C

0/C CV PV-14 67-656 67-657t 67-659 67-660 0-67-671 1-67-675 0-67-679 1-67-786 1-67-787 1-67-788 Loop II CS Rm Cooler N Hdr Supply Check Loop II CS Rm Cooler N Hdr Supply Check Loop II RHR Rm Cooler N Hdr Supply Check Loop II RHR Rm Cooler N Hdr Supply Check N Hdr Check Units 182 Emerg Cooling Unit Disch Isol RSW to EECW Check Emergency Control Bay Cooling Unit N Hdr Supply Isol Emergency Control Bay Cooling Unit N Hdr Supply Check Emergency Control Bay Cooling Unit S Hdr Supply Isol 3

M-4 3

F-4 3

C-4 3

E-1 NC G-7 3

F/G-5 C

3 G-7 8

3 G-7 3

G-7 2-1/2 CK SELF 0/C 0/C CV PV-14 1

CK SELF C

4 GA H

C 4

CK SELF C

CV PV-14 0/C CV PV-14 4

GA H

2-1/2 CK SELF 0/C 0/C CV PV-14 3

CK SELF 0/C 0/C CV PV-14 3

CK SELF 0/C 0/C CV PV-14 18 CK SELF 0/C 0/C CV PV-14 4

GA HAND C

0 Q

t 1-67-789 3-67-807 Emergency Control Bay Cooling Unit S Hdr Supply Check DG Rm Cooler 1 - Cooler 2 Crosstie 3

G-7 3

E-5 4

CK SELF C

0/C CV PV-14 3-67-808 67-838 DG Rm Cooler 1 Cooler 2 Crosstie H202 Analyzer 25-340 Supply Check 3

D-5 2

GA H

1 CK SELF 0/C 0

CV 44

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lg I

System:

Emergency Equipment Cooling Water (67)

Drawing No. 47E859-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve Number Fun tion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 67-839 67-843 67-844t 2-67-872 2-67-873 2-67-876 2-67-877 H202 Analyzer 25-340 Supply Check H202 Analyzer 25-341 Supply Check H202 Analyzer 25-341 Supply Check Shutdown Bd Rm Crosstie Check RCW to EECW Supply RCW to EECW Supply Shutdown Bd Rm Crosstie Check 3

H-4 3

F-5 3

E-5 3

E-5 C

1 CK SELF 0/C 0

3 CK SELF 0/C 0

3 CK SELF 0/C C

3 CK SELF 0/C 0

CV CV CV CV CV CV PV-14 PV-14 45

C r'

System:

Reactor Core Isolation Cooling (71)

Drawing No. 47E813-1 BROMNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 Valve

~Mb r F n tion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown

"'C 71-2 71-3 71-6A 71-7A Steam Line Inbd Isol Steam Line Outbd Isol Steam Line to Cond Drain Condensate Pump Disch Isol to CRW 1

B-3 A

1 B-4 A

2 E-10 B

2 A-3 B

3 GA HO 1

GA AO 1

GA AO 0/C 0/C 0/C Q,ST,LT PV-20 Q,ST,LT Q,ST Q,ST 71-8 71-9 71-14 71-1 7 71-18 71-25 71-32 71-34 71-37 71-38 71-39 71-40 71-502 71-503 71-508 71-542 71-543 Turbine Steam Supply Turbine Stop Turbine Exhaust PSC Inbd Suction PSC Outbd Suction Lube Oil Cooling Mater Cond Vac Pump Disch Pump Hin Flow Outbd Injection Isol Pump Test Return to CST Inbd Disch Testable Check Pump Inlet Check from CST Pump Suction Relief PSC Suction Check Barometric Cond Relief Cooling Mater to Barometric Cond Relief 2

F-1 B

4 2

F-2 B

3 2

D-7 AC 8

2 8-6 A

6 GL HO C

GA HO 0

SC H/SELF C

GA HO C

2 B/C-6 A

6 GA HO C

2 B-4 B

2 GL HO C

2 D-7 AC 2

2 E-5 A

2 2

D-5 B

6 2

B-5 8

4 SC H/SELF C

GL HO C

GA HO 0

GL HO C

2 B/C-6 C

1 RV SELF C

2 G-4 C

6 CK SELF C

2 F-9 C

1-1/2 RV SELF C

2 G-7 C

1 RV SELF C

2 C/D-4 8

6 GA HO C

1 C/D-4 AC 6

CK AO/SELF C

2 B/C-6 C

6 CK SELF C

0 RV 0

CV 0

RV 0

RV PV-19 0/C Q,ST 0/C Q,ST PV-23 0/C CV,LT CSD-8 0/C Q,ST,LT PV-20 0/C Q ST LT 0/C Q,ST 0/C CV,LT CSD-8 0/C Q,ST,LT 0

Q,ST C

Q,ST 0/C Q,ST 0/C CV,LT CSD-3 C

CV PV-29 51

System:

Reactor Core Isolation Cooling (71)

Drawing No. 47E813-1 DROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve

~Numb r 71-547 71-580 71-589 Fun i

n Min Flow Bypass Check Turbine Exhaust Check Conds Pump from Barometric Cond Check 2

E-5 2

E-3/4 AC 2

CK SELF C

0/C CV,LT CSD-11 AC 10 CK SELF C

0/C CV,LT PV-17 2

G/H-8/9 C

2 CK SELF C

C CV PV-29 Relief Request/

ASME Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 71-592t 71-597 71-598 71-599 71-600 Vac Pump Disch Check Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief 2

F-5 2

E-3 NC E-3 2

E-3 NC E-3 AC 2

CK SELF C

2 CK SELF C

CV, LT PV-17 CV PV-18 CV PV-18 CV PV-18 CV PV-18 52

lt

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System:

High Pressure Coolant Injection (73)

Drawing No. 47E812-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 Valve Number Fun tion Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown r'"

73-2 73-3 73-6A 73-16t 73-18 73-23 73-24 73-26 73-27 73-30 73-34 73-35 73-44 73-45 73-81 73-505 73-506 73-517 73-559 Steam Line Inbd Isol Steam Line Outbd Isol Steam Line to Cond Drain Turbine Steam Supply Turbine Stop Turbine Exhaust Turbine Ex Conds Pot Disch PSC Inbd Suction PSC Outbd Suction Pump Min Flow Outbd Injection Isol Pump Test Return to CST Inbd Disch Isol Testable Check 73-3 Bypass CST Suction Check Pump Suction Relief PSC Suction Check Hin Flow Bypass Check 1

G-7 1

G-6 2

E-2 2

G-2 2

G-3 2

D-7 2

D-6 2

B-6 2

G-5 2

D-5 2

F-5 2

F-6 2

F-6 1

E-6 1

G-6 2

H-5 2

G-4 2

B-6 2

0-5 AC AC AC AC 14 CK 1

RV SELF SELF 16 CK SELF C

4 CK SELF C

10 GA HO 10 GA MO 1

GA AO 10 GA HO 10 GA E/H 16 SC H/SELF C

2 SC H/SELF C

16 GA HO 16 GA HO 4

GL HO 14 GA HO 10 GA HO 14 GA HO 14 CK AO/SELF C

1 GL HO 0/C 0/C 0/C Q,ST,LT PV-20 Q,ST,LT Q,ST Q,ST Q,ST,FS PV-23 0/C 0/C 0/C 0/C 0/C 0/C CV,LT CSD-8 Q,ST,LT PV-20 Q,ST,LT Q,ST,LT Q,ST Q,ST Q,ST CV, LT CSD~

Q,ST,LT CV PV-29 RV CV PV-19 0/C CV/LT CSD-11 0/C CV,LT CSD-8 53

~ I r

P

BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

System:

High Pressure Coolant Injection (73)

Drawing No. 47E812-1 Valve

~Numb r 73-574 73-603 73-609 73-625 73-633 73-634 73-635 73-636 F n tion Cooling Water to Gland Seal Cond Relief Turb Exhaust Check Turb Exhaust Drain Check Gland Cond Return Check Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief 2

D-7 AC 20 CK SELF C

2 D-6 AC 2

CK SELF C

2 G-7/8 C

2 CK SELF C

2 E-1 C

2 CK SELF C

NC E-1 C

2 CK SELF C

2 E-1 C

2 CK SELF C

NC E-1 C

2 CK SELF C

0/C CV/LT PV-17 LT CV CV CV CV CV PV-17 PV-29 PV-18 PV-18 PV-18 PV-18 Relief Request/

ASME Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown m

2 F-7/8 C

1 RV SELF C

0 RV 54

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System:

Residual Heat Removal (74)

Drawing No. 47E811-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3

Valve Number Fun i

n Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety'esting Cold Shutdown 74-1 74-2 74-7 74-12I 74-13 74-24 74-25 74-30 74-35 74-36 74-47 74-48 74-52 74-53 74-54I 74-57 74-58 74-59 74-60 74-61 74-66 74-67 Pump A PSC Suction Pump A SD Cooling Suction Loop I Hin Flow Pump C

PSC Suction Pump C SD Cooling Suction Pump B PSC Suction Pump B SD Cooling Suction Loop II Hin Flow Pump D PSC Suction Pump 0

SD Cooling Suction SD Cooling Outbd Isol SD Cooling Inbd Isol Loop I Throttle Loop I Injection Loop I Testable Check Loop I PSC Return Loop I PSC Spray Loop I Pump Test Return Loop I Cont Spray Outbd Isol Loop I Cont Spray Inbd Isol Loop II Throttle Loop II Injection 2

8/C-5 2

C-6 2

D-6 2

D-5 2

D-5/6 2

C/D-4 2

0-4 2

D/E-3 2

B-4 2

C-4 1

E-5 1

E/F-5 2

F-7 1

F-6 1

F-6 AC 2

F/G-7/8 A

2 F-7/8 2

F-8 2

G-6 2

G-5 2

F-3 1

F-3 24 GA HO 20 GA HO 4

GA HO 24 GA HO 20 GA HO 24 GA HO 20 GA HO 4

GA HO 24 GA HO 20 GA HO 20 GA HO 20 GA HO 24 AN HO 24 GA HO 24 CK AO/SELF C

18 GA HO 4

GL 12 GL HO 12 GL HO 12 GL HO 24 AN HO 24 GA HO 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C

.0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C Q,ST Q,ST Q,ST Q,ST Q,ST Q,ST Q/ST Q,ST Q,ST Q,ST Q,ST,LT CSD-5 Q,ST,LT CSD-5 Q,ST Q,ST,LT CV,LT PV-25 Q,ST,LT Q,ST,LT Q,ST,LT Q,ST,LT Q,ST,LT PV-20 Q,ST Q,ST,LT 55

System:

Control Rod Drive (85)

Drawing No. 47E820-2 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 Valve Nuumb r 85-39A (1-185)85-398 (1-185)85-576 85-589t (1-185)85-597 (1-185)85-616 (1-185)

Fun ion Scram Inlet Scram Outlet Isol check to RWCU Charging Water Check Cooling Water Check Scram Outlet Check 1

C-8 B

1 PL AO C

0 O,ST PV-21 1

C-7 3/4 PL AO

(),ST PV-21 1

F-9 2

C-9 AC 4

CK SELF C

C 1/2 CK SELF C

C LT 0/C CV PV-21 1

B-9 2

D-7 1/2 CK SELF C

3/4 CK SELF C

CV CV PV-21 PV-21 Relief Request/

ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 69

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4 kg

Relief Request No. PV-2 System:

Standby Liquid Control (SLC) (63)

Diesel Fuel Transfer (DFT) (18)

Drawing:

SLC 47E854-1 DFT 47E840-3 Components:

SLC Pumps A,

B DFT Pumps

1A1, 1A2, 1Bl, 1B2, 1Cl, 1C2, 1Dl, 1D2, 3Al, 3A2,

3B1, 3B2, 3cl,

3C2, 3D1, 3D2 Class:

SLC-Class 2

DFT-Non-Code Class Function:

SLC-Injection of Sodium Pentaborate into Rx Vessel DFT-Transfer Diesel Fuel to Day Tank Impractical Test Requirement:

IWP-3100 - Measure inlet pressure prior to and during pump test and measure pump delta P during pump test.

Basis for Relief:

During testing, these pumps take suction from a tank that has a relatively small range of level variation during pumps operation.

In addition, these are positive displacement pumps whose inlet pressure does not affect p'ump operating characteristics.

Therefore, differential pressure measurement is not meaningful in monitoring pump performance.

Also, the diesel fuel transfer pumps are not instrumented or constructed to allow measurement of discharge pressure.

Alternative Testing:

SLC:

Pump discharge pressure will be measured during quarterly testing.

Diesel Fuel Transfer:

Flow rate is measured monthly as required by plant Technical Specifications utilizing existing tank level instrumentation.

75

h 1

L

System:

Drawing:

Relief Request No.

PV-3 Diesel Fuel Transfer (DFT) (18) 47E840-3 Components:

DFT Pumps 1Al, 1A2,

1B1, 1B2,

1C1, 1C2,

1D1, 1D2, 3Al, 3A2, 3Bl, 3B2,

3B1, 3C2, 3Dl, 3D2 Class:

Function:

Non-Code Class Pump diesel from seven-day tank to day tank Impractical Test Requirement:

IWP-3500(a), five-minute minimum run time Basis for Relief:

Depending on the level of the day tank when the surveillance procedure is performed, the high level switch could stop the pump before data is accumulated.

Alternative Testing:

Measure parameters during the two-minute minimum test.

Test will be performed when level of tank permits pumps to be run for as long as possible, at least once per six months.

The existing fuel level gauge will be utilized to measure flow.

76

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System:

Drawing:

Relief Request No. PV-5 Diesel Fuel Transfer (DFT) (18) 47E840-3 Components:

Diesel Fuel Transfer Pumps 1A1, 1A2, 1Bl, 1B2,

1C1, 1C2,

1D1, 1D2, 3A1, 3A2, 3B1,

3B2, 3C1,

3C2, 3D1, 3D2 Class:

Function:

Non-Code Class Transfer diesel fuel to engine day tank.

Impractical Test Requirement:

1WP-4110 instrument accuracy shall be within the limits of Table IWP-4110-1, IWP-3210, Allowable Ranges of Inservice Test Quantities, Basis 'for Relief:

The diesel fuel transfer pumps are not instrumented to measure flow other than a level gage on the day tank.

We have attempted to measure the pump discharge flow using externally mounted 'flow metering devices;

however, the system piping configuration is not compatible with such devices.

At the present time the day tank level gage is the only means available for flow measurement and its repeatability is insufficient, for pump performance trending.

Plant Technical Specification 4.9.A.l.a requires that the diesel fuel transfer pumps be tested monthly to demonstrate operability.

This testing consists of demonstrating that the transfer pumps are capable of pumping fuel at a rate greater than the consumption rate of the diesel engine.

(5 gallons/minute)

Alternative Testing:

The ability of the pumps to exceed the engine consumption rate of 5 gallons per minute will be demonstrated

monthly, and vibration measurements on the pumps will be recorded and evaluated per Table IWP-3100-2 quarterly, in lieu of measuring pump hydraulic parameters (pressure and flow rate).

78

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ML18037A532

Text

SUBJECT:

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