ML18038B553
| ML18038B553 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/14/1995 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M84494, TAC-M84495, TAC-M84496, NUDOCS 9511210014 | |
| Download: ML18038B553 (54) | |
Text
P R.ICDR.I "EY'CCELERATED RIDS P ROCESSli G)
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9511210014 DOC.DATE: 95/11/14 NOTARIZED: NO DOCKET g 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)
SUBZECT: Responds to questions re anomalies six 6 ten identified during review of plant second 10-yr interval IST program 6
submits revised relief requests PV-13,14,18,29 6 30.
DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code GL-89-04 NOTES:
P 0
RECIPIENT ID CODE/NAME PD2-3 WILLIAMS,Z.
INTERNAL: ACRS CENTER
/Z5/l$fC"B NUDOCS-ABSTRACT RES/DET/EMMEB EXTERNAL: LITCO ANDERSON NRC PDR COPIES LTTR ENCL 1
1 1
1 6
0 1
1 1
1 1,
1 1
1 1
1 1
RECIPIENT ID CODE/NAME PD2-3-.PD AEOD/SPD/RAB NRR/DE/ECGB NRR/DE/EMEB OGC/HDS3 RES/DSIR/EIB NOAC COPIES LTTR ENCL 1
1 1
1 1
1 1
1 1
0 1
1 1
1 D,
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N NOTE TO ALL"RIDS RECIPIENTS:
PLEASE HELP US TO REDUCE IVASTE!CONTACTTHE DOCU!CLIENTCONTROL DESK, ROOM Pl-3'7 (EXT. 504-2083 ) TO ELI iIINATE YOUR VAl$L FROW!
DISTRIBUTIONLISTS I'OR DOCL'MENfS YOU DON"I'NEED!
TOTAL NUMBER OF COPIES. REQUIRED:
LTTR 21 ENCL 14
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Tennessee Valley Authority, Post Otfice Box 2000. Decatur, Alabama 35609 November 14, 1995 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 PERRY NUCLEAR PLANT (BPN) - UNITS 1g 2g AND 3 AMERICAN SOCIETY OP MECHANICAL ENGINEERS (ASME)g SECTION XI/
INSERVICE TESTING (IST)
PROGRAM FOR PUMPS AND VALVES'EQUESTS FOR RELIEF PV-13~
14~ 18'9'ND 30 (TAC NOS ~ M84494~
M84495~
AND M84496)
This letter provides TVA's response to questions regarding anomalies six and ten, identified by the staff during its review of BFN's second 10-year interval IST program.
TVA is also submitting an IST program update that includes revised relief requests (PV-13, 14, 18, 29, and 30) resulting from our response to the staff's questions.
By letter dated August 31,
- 1992, TVA submitted the updated IST program for the second 10-year interval for pumps and valves for BFN Units 1, 2, and 3.
In a letter dated October 22,
- 1993, NRC issued a safety evaluation for the second 10-year interval that included a request that TVA provide a revised IST program to address eleven anomalies identified during NRC review.
On November 14,
NRC issued a safety evaluation by letter dated May 16,
- 1995, and accepted TVA's response to anomalies 1, 2, 3, 4, 5, 7, 8, 9, and 11.
However, the staff requested that TVA provide additional information regarding anomalies six and ten within six months from the date of the safety evaluation.
r;) Q P,~
951 i2iOOi4 tstSi ii4 PDR ADOt".K 05000259 P
PDR L
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g U.S. Nuclear Regulatory Commission Page 2
November 14, 1995 Anomalies six and ten addressed issues associated with the use of nonintrusive (e.g., acoustic, ultrasonic,
- magnetic, radiography, and thermography) techniques to verify the full-stroke exercise capability of check valves for relief requests PV-13, 14, 18, 29, and 30.
TVA has evaluated the use of nonintrusive techniques for each of the above relief requests and has determined their use to be impracticable.
The revised requests for relief furnished in the enclosure provide specific justification regarding the impracticality of nonintrusive testing techniques.
As stated in the revised requests for relief, TVA intends to use disassembly and inspection in accordance with Generic Letter 89-04, Position 2 to verify the full-stroke open or closure capability of the check valves.
The enclosure to this letter contains BFN's IST program update including the revised requests for relief described above.
Relief request PV-21 was also revised to correct the class designation'f the affected valves.
Additionally, TVA has submitted program changes that include editorial changes, and addition or deletion of components to the component listing tables.
If you have any questions regarding this letter please contact me at (205) 729-2636.
Sz Salas Manager of Site Licensing Enclosure cc: 'ee page 3
U.S. Nuclear Regulatory Commission Page 3
November 14, 1995 Enclosure, cc (Enclosure):
Mr. Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road
- Athens, Alabama 35611 Mr. Z. F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852
ll
,1 0
ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1g 2g AND 3 AMERZCAN SOCIETY OF MECHANICAL ENGINEERS (ASME) ~
SECTION XIg INSERVZCE TESTING PROGRAM (IST)
FOR PUMPS Am VALVES PROGRAM UPDATE AND REVZSED REQUESTS FOR RELZEF PV-13, 14, 18, 21, 29, and 30 Listed below is a brief description of the BFN ASME Section XI IST program changes in this enclosure.
P~ae 1
Revised the program "Introduction", Section A to reference NUREG-1482 as a source of guidance and alternatives for ASME Section XI IST.
Updated the status of NUREG-1482 by deleting the "Draft" designation.
This change documents the intent to use NUREG-1482 where practical.
PBCCB 2 Revised "Introduction", item 2 to add "nuclear" to the term "safety-related."
This emphasizes that a component must perform a nuclear safety function to be included in the IST program.
Added NUREG-1482 to "Section B" as a source of guidance and alternatives for pump testing.
This documents the intent to use NUREG-1482 where practical.
Added NUREG-1482 to "Section C" as a source of guidance and alternatives for valve testing.
This documents the intent to use NUREG-1482 where practical.
PBIB8 6 Added abbreviation (RT) for rupture disc test.
This adds a
designation previously missing and is editorial in nature.
PBCBB 24 Corrected "Safety Position" and "Relief Request" columns for RHRSW valves 23-24, 23-40, 23-46, and 23-52.
These valves have been full-stroke tested since withdrawl of Relief Request PV-12.
This editorial change corrects the information in the valve listing only.
~Pa e 26 Deleted valves 0-24-796 and 0-24-798.
These valves were removed from the plant by Design Change Notice (DCN) W27299A.
Deleted valves 3-24-886 and 3-24-891.
These valves were removed from Unit 3 by DCN W27299A.
Added Unit 1 and 2 designation to 24-886 and 24-891 to indicate these valves are present on Units 1 and 2.
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'. P~ae 52 Added rupture disc 71-11A to the "Valve Number" column.
Revised identification of 71-503 to 71-19 and 71-502 to 71-499.
Deleted PV-20 from "Relief Request" column for 71-17.
Corrected "Safety Position" column of 71-32 from 0/C to C.
Corrected "Safety Position" column of 71-499 from C to 0/C.
Deleted 71-542 from the "Valve Number" column as a result of an IST program review.
All other changes except the rupture disc are editorial.
The rupture disc addition is required to document the disc status in the program as a Section XI component.
Pecae52 Replaced PV-17 with CSD-12 for 71-580 and 71-592.
Corrected drawing coordinates for 71-589 through 71-600.
All changes resulted from an IST program review.
P~ae 53 Added fail-safe test designation to "Testing Required" column for 73-6A.
Corrected "Safety Position" column for 73-24 from 0/C to C.
Deleted PV-20 from "Relief Request..."
column for 73-26.
Corrected "Valve Type" column from gate to globe for 73-35.
Corrected "Safety Position" column for 73-44 from 0/C to O.
Corrected "Safety Position" column for 73-505 from C to 0/C.
All changes resulted from an IST program review and are editorial in nature.
~aces 5a Added rupture disc 73-729 to the "Valve Number" column.
Replaced PV-17 with CSD-12 in the "Relief Request..."
column for 73-603 and 73-609.
Revised drawing coordinates for 73-574 and 73-625 through 73-636.
All changes except the rupture disc addition are editorial.
The rupture disc addition is required to document the disc status in the program as a Section XI component.
P~ae 55 Deleted PV-20 from the "Relief Request..."
column for 74-61.
This is an editorial change.
~aces 55 Deleted PV-20 from the "Relief Request..."
column for 74-75.
This is an editorial change.
Paces 59 Corrected "Safety Position" column for 75-25 from 0 to 0/C.
This is an editorial change necessary to address the containment isolation function of this valve.
PF 23-Revised Relief Request PV-13.
Added statement to basis for, relief to address nonintrusive testing.
~
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if
PP'-24 Revised Relief Request PV-14.
Deleted 0-24-796 and 0-24-798 as a result of DCN W27299 (removed valves).
Added Unit 1 and Unit 2 designations for 24-886 and 24-891 to indicate that the Unit 3 valves were removed by DCN W27299.
Added statement to basis for relief to address nonintrusive testing.
PV-18 Revised Relief Request PV-18.
The HPCI/RCIC turbine exhaust vacuum breaker check valves will be tested to verify an open flowpath on a cold shutdown basis.
Added statement to basis for relief to address nonintrusive testing.
EV-21 Revised relief Request PV-21.
Corrected class designation of the valves.
PV-29 Revised Relief Request PV-29.
Added statement to basis for relief to address nonintrusive testing.
PV-30 Revised Relief Request PV-30.
Added statement to basis for relief to address nonintrusive testing.
E-3
BFN~
ASME SECTION XIg IST PROGRAM REVISION EFFECTZVE PAGE LIST EFFECTIVE PAGES 1
2 6
24 26 51 52 53 54 55 56 59 PV-13 PV-14 PV-18 PV-21 PV-29 PV-30 E-4
7 I
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10 CFR 50.55a(f)(4)(ii) requires the Inservice Testing (IST) of nuclear power facility pumps and valves whose functions are required for safety.
This testing is required to be conducted in 120-month intervals, beginning with initial power operation and continuing throughout the service life of the facility.
The testing must comply with the latest edition and addenda of American Society of Mechanical Engineers (ASME)Section XI, incorporated by reference in 10 CFR 50.55a(b),
12 months prior to the start of the 120-month interval in question.
Specific exceptions to these requirements are included here as Relief Requests.
BFNP Units 1, 2, and 3 are on a concurrent IST interval.
BFNP completed its initial 120-month interval on August 31, 1992 and began the second 120-month interval on September 1,
1992.
The second (current) ten-year IST interval will end on August 31, 2002.
The second ten-year interval for the BFNP IST program for pumps and valves complies with Section XI of the ASME Boiler and Pressure Vessel
- Code, 1986 Edition, as required by 10 CFR 50.55a(b)(2).
All references to subsections IWP and IWV of Section XI in the procedure correspond to the 1986 Edition of ASME Section XI.
Portions of later editions and addenda of ASME Section XI, which are incorporated by reference in 10 CFR 50.55a(b),
may be used subject to limitations and modifications established by NRC, provided that all applicable requirements of the respective editions and addenda are met.
NRC Generic Letter 89-04 and NUREG-1482 provide alternatives to ASME Section XI IST requirements that are acceptable to NRC.
In order to utilize the alternative methods, they must be complied with in their entirety and their use documented in the IST program.
TVA Nuclear Power Standard 8.6 (STD-8.6) is the TVA corporate level procedure which establishes IST program requirements for all TVA nuclear power plants.
STD-8.6 requires establishment, implementation, and maintenance of IST in accordance with 10 CFR 50.55a(f)(4) and GL 89-04.
Site Standard Practice (SSP) -8.6, ASME Section XI IST of Pumps and Valves, is the procedure that administratively controls the IST program at BFNP.
The BFNP IST program was prepared using the following for guidance:
~
ASME Section XI, 1986 Edition
~
ASME/ANSI Operations and Maintenance (0&M) Standards, Parts 1,
6, and 10
~
~
~
~
NRC Temporary Instructions 2515/110 and 2515/114
~
NRC Inspection Procedure (IP) 73756
~
~
~
BFNP Updated Final Safety Analysis Report (UFSAR)
~
BFNP Technical Specifications (TS)
~
BFNP System Design Criteria NOV %4 f995
N P
4
~
I,,
RODUCT ON PO ICY ST S (Continued)
All plant components were reviewed for ASME Section XI IST applicability and those meeting the following criteria have been included in the BFNP IST program.
Components that meet the criteria of IWp-1200 and IWV-1200 of ASME Section XI are excluded from the IST program.
l.
Components classified as ASME Code Class 1, 2, or 3 equivalent that are required to perform a specific safety function.
Classification is determined by TVA Nuclear Engineering and designated on the controlled Inservice Inspection (ISI) series of drawings.
2.
Components which give nuclear safety-related overpressure protection to nuclear safety-related Code class equivalent
- systems, subsystems, and components.
3.
Additional safety-related components that are not ASME Code Class 1, 2, or 3 equivalent but which have been determined to require augmented IST.
These components are identified in the BFHP IST program listing as non-ASME Code Class.
Changes to the plant are controlled by administrative procedures (SSPs) that require a review for ASME Section XI applicability of all proposed design changes, procedure
- changes, temporary alterations, modifications, tests, or experiments performed under the 10 CFR 50.59 process.
SSP-9.3, Plant Modifications and Design Change Control, is the SSP used to control changes to the plant configuration and design.
SSP-2.3, Administration of Site Procedures, is the SSP used to control plant procedures.
SSP-12.13, 10 CFR 50.59, Evaluations of Changes,
- Tests, and Experiments, is the SSP used to control changes in accordance with the 10 CFR 50.59 process.
The BFNP ASME Section XI IST Program (and all implementing procedures) are changed as required to reflect approved changes in the plant configuration, processes, or procedures.
These changes are implemented prior to return to operation of the affected systems or components.
B.
S ro am The pump test program shall be conducted in accordance with Subsection IMP of Section XI of the ASME Boiler and Pressure Vessel Code except where relief has been requested under the provisions of 10 CFR 50.55a(f)(5)(iii).
The pump test program may also utilize NUREG-1482 and Generic Letter 89-04 provided the applicable associated requirements are complied with and their use is documented in the IST program.
Part II details the IST program for all safety-related pumps at Browns Ferry Nuclear Plant.
Each parameter to be measured as well as specific notes concerning requests for relief are also listed, C.
Va ve S
ro am The valve test program shall be conducted in accordance with Subsection INf of Section XI of the ASME Boiler and Pressure Vessel Code except where relief has been requested under the provisions of 10 CFR 50.55a(f)(5)(iii).
The valve test program may also utilize NUREG-1482 and Generic Letter 89-04 provided the applicable associated requirements are complied with and their use is documented in the IST program.
The valve test program is included in Part II.
t<OV 14 1995
Valve Actuator T s
Cod e
~Sebo GA BF BA CK GL AN RD RV SC Gate Butterfly Ball Check Globe Angle (Globe)
Rupture Disk Relief Valve Stop Check Code AO MO SELF HO E/H Cyl
~ea~
Air Operator Manual Operator Motor Operator Self Actuated Solenoid Operator Explosive Hydraulic Operator Electro-Hydraulic Oper.
Cylinder XFC Excess Flow Check PL Plug Code
~aug Va ve Test e u erne ts Code 0/C LO Open Closed'pen or Closed Locked Open ST LT LT(R)
Full Stroke Quarterly Stroke Time Leak Test Leak Test in Reverse Direction ELO LC TH Electrically Locked Open Locked Closed Throttled FS CSD CV Full Stroke Refuel Cycle Fail Safe Per IWV-3415 Full Stroke Cold Shutdown Full Stroke Check Valve Quarterly RV Relief Valve Per IWV 3510 Destructive Test Per IWV 3610 RT Visual Inspection of Rupture Disc 1955C NOU i4 )995
BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3
Drawing No. 47E858-1 Valve Number 23-34 23-40 23-46 23-52 1-23-57 2-23-57 0-23-502 0-23-504 Q-23-506 23-5D9 1-23-510 23-516 0-23-522 0-23-524 0-23-526 23-529 1-23-530 23-536 0-23-542 0-23-544 0-23-546 Fun tion RilR Htx A Coo1ing Wtr Outlet RtiR Htx C Cooling Wtr Outlet RiiR Htx 8 Cooling Wtr Outlet RIIR litx 0 Cooling Wtr Outlet Standby Coolant Standby Coolant Al Pump Oisch Check A1-A2 Pump Crosstie A2 Pump Oisch Check Htx A Supply Relief Htx A Inlet Check litx A Tube Relief 01 Pump Disch Check 81-02 Pump Crosstie 82 Pump Disch Check Htx 0 Supply Relief Htx 8 Inlet Check lltx 0 lube Relief C2 Pump Disch Check Cl-C2 Pump Crosstie Cl Pump Oisch Check 3
F-8 0
3 H-8 8
3 F-5 8
3 H-5 8
3 G-5 8
3 E-3 0
3 8-5 C
3 8-5 8
3 8-5 C
3 0-7 C
3 E
7 C
3 F-7 C
3 04 C
12 GL HQ 12 GL HQ 12 GL 12 GL HO 10 GA HO 10 GA HQ C.
C 18 CK SELF C
18 GA H
18 CK SELF C
RV SELF C
16 CK SELF C
1 RV SELF C
18 CK SELF 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 CV CV RV CV RV CV 3
8-4 18 GA H
Q/C Q
3 8-4 C
3 E-5 C
3 E-5 C
3 F-6 C
3 8-4 C
3 8-3 8
3 8-3 C
18 CK SELF C
1 RV SELF C
16 CK SELF C
1 RV SELF C
18 CK SELF C
18 GA H
18 CK SELF C
0/C 0/C 0/C 0/C CV CV RV CV CV ASHE Orawing Size Valve Actuator Homal SaFety testing Cold Shutdo n
Relief Request/
y~ ~ygg pstpsto position R~~tr d Sus>'i ation 24 NOg i4 1995
System:
Raw Cooling Water (24)
Orawing No. 47E844-2 and 47E859-1 BROWNS FERRY NUCLEAR PLANT.UNITS 1, 2, and 3 Valve
~ll mber 1-24-712A 1-24-7128 3-24-826 Fn in Wtr Chiller A Oisch Isolation Wtr Chiller 8 Oisch Isolation North EECW Hdr Check 1,2-24-886 H202 Panel 25-340 Check 1
~ 2-24-891 H202 Panel 25-341 Check ASHE Orawing
~la
~ordi n~a 3
E-8 3
E-8 3
G-2 3
A-6 3
A-6 Size Valve Actuator Normal
~in.
~T
~TRg P~i~in 6
GA H
0/C 6
GA H
0/C 1.5 CK SELF C.
1 CK SELF C
1 CK SELF C
0/C Q
0/C CV CV CV PV-14 PV-14 PV-14 Relief Request/
Safety Testing Cold Shutdown 26
~ ~
I
System:
Reactor Core Isolation Cooling (71)
Drawing No. 47E&13-1 BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3
~
~
Valve Number F nc ion ASHE Drawin' Size Valve Actuator Normal Safety Testing Cold Shutdown Relief Request/
71-2 71-3 71-6A 71-7A Steam Line Inbd Isol Steam Line Outbd Isol Steam Line to Cond Drain Condensate Pump Oisch Isol to CRW 1
8-3 1
8-4 2
E-10 2
A-3 3
GA HO 3
GA HO 1
GA AO 1
GA AO 0
0/C 0
0/C 0.
0/C C
C Q,ST,LT PV-20 Q,ST,LT Q,ST Q,ST 71-&
71-9 71-11A 71-14 71-17 71-18 71-19 71-25 71-32 71-34 71-37 71-3&
71-39 71-40 71-499 71-50&
71-543 Turbine Steam Supply Turbine Stop Turbine Exhaust Rupture Disc Turbine Exhaust PSC Inbd Suction PSC Outbd Suction Pump Suction Relief Lube Oil Cooling Water Cond Vac Pump Oisch Pump Hin Flow Outbd Injection Isol Pump Test Return to CST Inbd Disch Testable Check Pump Inlet Check from CST PSC Suction Check Cooling Water to Barometric Cond Relief 2
F-1 8
4 2
F-2 8
3 2
0-3 0
8 2
0-7 AC 8
2 8-6 A
6 GL HO C
GA HO 0
RO SEI.F C
SC H/SELF C
GA HO C
2 8/C-6 C
1 2
8-4 8
2 2
D-7 AC 2
2 E-5 A
2 2
0-5 8
6 2
8-5 8
4 RV SELF C
Gi.
HO C
SC H/SELF C
GL HO C
GA HO 0
GL HO C
2 C/0-4 8
6 GA HO C
1 C/0-4 2
8/C-6
- AC 6
C 6
2 G-4 C
6 2
G-7 C
1 CK AO/SELF C
CK SELF C
CK SELF C
RV SELF C
2 8/C-6 A
6 GA HO C
RT 0/C CV.LT CSO-8 0/C Q,ST,LT 0/C QUEST,LT 0
RV 0/C Q,ST C
CV,LT CSD-&
0/C Q,ST,LT 0
Q,ST C
Q,ST 0/C Q,ST 0/C CV,LT CSD-3 0/C CV PV-29 0
CV PV-19 0
RV 51
1 4 ~
System:
Reactor Core Isolation Cooling (7 I)
Drawing No. 47E813-1 DROWNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3
Valve Number Func i n
ASHE Drawin Relief Request/
rawing Size Valve Actuator Normal Safety Testing Cold Shutdown 71-547 71-580 71-588 71-589 Hin Flow Bypass Check Turbine Exhaust Check Conds Pump Check Conds Pump from Barometric Cond Check 2
E-5 2
E-3/4 2
8-3 2
B-3 C
1-1/4 CK SELF C
2 CK SELF C.
C AC 2
CK SELF C
AC 10 CK SELF C
C CV C
CV PV-29 PV>>29 0/C CV,LT CSO-11 0/C CV,LT CSO-12 71-592 71-597 71-598 71-599 71-60D Vac Pump Oisch Check Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief Turbine Exhaust Vac Relief 2
C-5 2
0-6 NC 0-6 2
0-6 NC 0-6 AC 2
CK SELF C
2 CK SELF C
2 CK SELF C
2 CK SELF C
2 CK SELF C
CV, LT CSD-12 CV PV-18 CV PV-18 CV PV-18 CV PV-18 52 NOY i4 15%
System:
High Pressure Coolant In)ection (73)
BROWNS FERRY NUCLEAR PLANT UNITS I, 2, and 3
Drawing No. 47E812-I Valve
~Nmb r F n ion ASHE Orawin Size Valve Actuator Normal Safety Testing Cold Shutdown Relief Request/
73-2 73-3 13-6A 73-16 73-18 73-23 73-24 73-26 73-27 73-30 73-34 73-35 73-44 73-45 73-8'I 13-505 73-506 73-517 73-559 Steam Line Inbd Isol Steam Line Outbd Isol Steam I.ine to Cond Drain Turbine Steam Supply Turbine Stop Turbine Exhaust Turbine Ex Conds Pot Disch PSC Inbd Suction PSC Outbd Suction Pump Hin Flow Outbd injection Isol Pump Test Return to CST Inbd Oisch Isol Testable Check
'73-3 Bypass CST Suction Check Pump Suction RelieF PSC Suction Check Hin Flow Bypass Check I
G-7 1
G-6 2
E-2 2
G-2 2
G-3 2
0-7 2
. 0-6 2
8-6 2
G-5 2
0-5 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 10 GA HO 0
1 GA 10 GA AO 0
C 10 GA E/H C
16 SC H/SELF C
2 SC H/SELF C
16 GA N
C 16 GA N
C 4
GL N
C 14 GA HO 0
10 GL N
C 14 GA HO C
'I4 CK AO/SEI.F C
1 GL HO C
14 CK SELF C
1 RV SELF C
16 CK SELF C
4 CK SELF C
0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C 0/C CV RV CV PV-29 PV-19 CV/LT CSO-II Q,ST.LT PV-20 Q,ST,LT Q,ST,FS Q,ST Q, ST, FS PV-23 CV ALT CSD-8 CV,LT CSD-8 Q,ST,LT Q,ST,U Q,ST,LT Q,ST Q,ST Q,ST CV,LT CSD-4 Q,ST,LT 53
System:
High Pressure Coolant Injection (73)
Drawing No. 47E812-1 BROMNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3 Valve
~kumb r 73-574 73-603 73-609 73-625 73-633 73-634 73-635 73-636 73-729 Fun ion Cooling Mater 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 Turbine Exhaust Rupture Disc 2
0-7 AC 2
D-6 AC 2
B-4 C
2 E-7 C
NC E-7 C
2 E-7 C
NC E-7 C
2 E-4 0
20 CK SELF C
2 CK SELF C"
2 CK SELF C
2 CK SELF C
2 CK SELF C
2 CK SELF C
2 CK SELF C
16 RO SELF C
0/C CV/LT CSD-12 CSO-12 CV CV CV CV PV-29 PV-18 PV-18 PV-18 PV-\\8 ASNE Orawin9 S>ae Valve Actuator Normal Safety Testing Cold Shutdown
~
Relief Request/
2 C-4 C
1 RV SELF C
0 RV 54 HW
~ 4 1N3
System:
Drawing No. 47E811-1 DROWNS FERRV NUCLEAR PLANT UNITS 1, 2, and 3
Valve
~IIU o r Fun ion Relief Request/
ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown
~t BSLr~dtn'Ltd gittdditrrt ~n.
~t itttm Prost idd O~oi i n ~RLuirsd Justification 74-1 74-2 74-7 74-12 74-13 74-24 74-25 74-30 74-35 74-36 74-47 74-48 74-52 74-53 74-54 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 8
SD Cooling Suction Loop II Hin Flow Pump 0 PSC Suction Pump D 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 lnbd Isol Loop II Throttle Loop II Injection 2
8/C-5 2
C-6 2
0-6 2
D-5 2
0-5/6 2
C/D-4 2
0-4 2
0/E-3 2
B-4 2
C-4 1
E-5 1
E/F-5 2
F-7 1
F-6 1
F-6 2
F/G-7/8 2
F-7/8 2
F-8 2
G-6 2
G-5 2
F-3 1
F-3 AC 24 GA HO 20 GA HO 4
GA N
24 GA N
20 GA N
24 GA HO 20 GA HO 4
GA N
24 GA N
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 HO 12 GL HO 12 GL N
12 GL HO 24 AN HO 24 GA HO 0/C 0/C 0/C 0/C 0/C 0/C 0/C OIC 0/C 0/C 0/C 0/C 0/C OIC 0/C 0/C 0/C 0/C 0/C OIC 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,S1 Q,ST,LT CV,LT PV-25 Q,STILT Q,ST,LT Q,ST,LT Q.ST,LT Q,ST,LT Q,ST Q,ST ALT 55
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System:
Drawing tto. 47E811-1 BROMNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3
Valve
~Numb r Fn in ASHE Orawin 9 Sixe Valve Actuator Normal Safety Testing Cold Shutdown Relief Request/
74-68 74-71 74-72 74-73 74-74 74-75 2-74-96 3-74-96 2-74-97 3-74-97 1-74-98 2-74-9&
1-74-99 2-74-99 2-74-100 3-74-100 1-74-101 2-74-101 Loop II Testable Check Loop II PSC Return Loop Il PSC Spray Loop II Pump Test Return Loop II Cont Spray Outbd Loop II Cont Spray Inbd Pump A Suction Crosstie Pump A Suction Crosstie Pump C Suction Crosstie Pump C Suction Crosstie Pump 8 Suction Crosstie Pump 8 Suction Crosstie Pump 0 Suction Crosstie Pump D Suction Crosstie F-4 2
G-2 2
F-2 2
F-2 2
G-4 2
G-5 2
C-5 2
C-5 2
C-5 2
C-5 2
C-4 2
C-4 2
C-4 2
C-4 tltx A-C Crosstie to U-2 8-0 titx 8-0 Oisch Crosstie Iitx 8-0 Disch Crosstie 2
C-8 2
D/E-1 2
C-1 tltx A-C Crosstie to U-1 8-D Htx 2
C-8 AC 18 GA N
4 GL HO 12 GL N
12 GA HO 12 GA HO 14 GA HO 14 GA HO 14.
GA HO 14 GA HO 14 GA HO 14 GA HO 14 GA HO 14 GA N
10 GA HO 10 GA HO 10 GA N
10 GA HO C.
24 CK AO/SELF C
0/C CV,LT PV-25 0/C Q,ST,LT 0/C Q
ST LT 0/C Q,ST,LT 0/C Q,ST,LT 0/C Q,ST,LT Q,S'I Q,ST Q,ST Q,ST Q,ST Q,ST Q,ST Q,ST Q,ST Q,ST Q,ST Q,ST 56
System:
Core Spray (75)
Drawing No. 47E814-1 Valve
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BROMNS FERRY NUCLEAR PLANT UNITS 1, 2, and 3
Relief Request/
ASHE Drawing Size Valve Actuator Normal Safety Testing Cold Shutdown 75-9 75-22 75-23 75-25 75-26 75-37 75-50 75-51 75-53 75-54 75-57 75-58 75-507A 75-5078 75-507C 75-5070 75-537A 75-5378 75-537C 75-537D 75-543A 75-5438 Loop I Hin Flow Loop I Pump Test Loop I Injection Loop I Injection Loop I Testable Check Loop II Hin Flow Loop II Pump Test Loop II Injection Loop II Injection Loop II 1'estable Check Drain Pump A Inbd Isol Drain Pump A Outbd Isol Pump A Suction Relief Pump 8 Suction Relief Pump C Suction Relief Pump 0 Suction Relief Pump A Disch Check Pump 8 Disch Check Pump C Disch Check Pump 0 Disch Check Loop I Disch Relief Loop II Disch Relief 2
E/F-5 2
F-5 2
F-5/6 1
F-6 1
F-7 2
E-4 2
F-4 2
G-6 1
G-6 1
C-7 2
8-4 2
8-5 2
C-4 2
C-2 2
C-6 2
C-4 2
0-4 2
0-3 2
0-6 2
0-4 2
F-4 2
E-4 AC AC 3
GA HO 10 GL HO 12 GA HO 12 GA HO 0
12 CK AO/SELF C
3 GA HO 10 GL HO 12 GA HO
'12 GA HO 12 CK AO/SELF C
3 GA AO 3
GA AO 1
RV SELF C
1 RV SELF C
1-RV SELF 1
RV SELF C
12 CK SELF C
12 CK SELF C
12 CK SELF C
12 CK SELF C
1 RV SELF C
1 RV SELF C
0/C 0/C 0/C 0/C 0/C 0/C Q,ST Q,ST Q,ST Q,ST.LT CV/LT PV-25 Q.ST Q,ST Q,ST Q,STILT CV/LT PV-25 Q.ST,LT Q,ST,LT RV RV RV RV CV CV CV CV RV 59
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RELIEF REQUEST NUMBER PV-13 System:
Drawing:
Components:
Category:
Class:
RHR Service Water (23) 1-47E858-1 RHRSW header keep-fill check valves 0-23-601, 603, 605, 607 Function:
These valves open and close to maintain RHRSW headers charged with raw water from the Raw Service Water system.
Impractical Test Requirement:
IWV-3521 and 3522 - Exercise valves closed once every three months Basis for Relief:
These check valves, located in the keep-fill lines for the RHR service water system permit raw service water flow into each RHRSW header while preventing process flow in the reverse direction during RHRSW system operation. There are no vent, drain, or test connections located upstream of the check valves that would allow backflow testing. Non-intrusive testing of these valves using ultrasonics and /or acoustics is not practical for these valves due to the inability to force the valves closed quickly enough to generate a repeatable acoustic or ultrasonic signal.
The isolation valves used to force the valves closed are not capable of quick opening or closing (the valves have handwheel operators). The valve pistons therefore contact their seat lightly. Ultrasonic and acoustic traces obtained during testing for feasibility were inconclusive due to piston flutter and system noise. No acoustic trace of piston contact on closure could be discerned and the ultrasonic trace did not reveal full closed position (movement of the piston was detected in the mid-range position, but not at closure). Radiography is not considered practical due to a review of test shots on other liftcheck valves of similar size and material. The radiographs that were reviewed did not reveal sufficient detail to determine whether the valve pistons were completely seated.
Test equipment for detecting disc position using magnetics has not been purchased and was therefore unavailable for use in determining feasibility.
Alternative Testing:
The valves willbe proven to close once every operating cycle by disassembly and inspection. Because the check valves are of the same design and service conditions they willbe grouped for disassembly on a
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Relief Request Number PV-13 (continued) rotating basis in accordance with position 2 in NRC Generic Letter 89-
- 04. Ifa check valve selected for disassembly fails the inspection criteria of GL 89-04, then the other three keep-fill check valves in the affected test group willbe disassembled and inspected during the next available system outage. The design of these check valves results in less chance of problems caused by disassembly and reassembly.
Because the piston is positioned down inside the valve body until forced open by water
- pressure, there is virtually no chance of kinking the piston in the valve cylinder when reinstalling the valve bonnet. The inspection and'cleaning (as necessary) of the valve internals after disassembly ensures that the valve piston is free to move when it is reassembled.
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RELIEF REQUEST NUIMBERPV-14 System:
Drawing:
Emergency Equipment Cooling Water (EECW) (67)
Raw Cooling Water (RCW) (24) 1, 2, and 3-47E844-2 1, 2, and 3-47E859-1 3-47E859-2 3-47E866-7 Components:
EECW to RHR/Core Spray room cooler check valves67-541, 542, 558, 559, 584, 585, 600, 601, 638, 639, 648, 649, 656, 657, 659, 660 EECW to Unit 1 Diesel Generator cooler check valves 0-67-507, 508, 514, 515, 521, 522, 528, 529, 624, 625, 627, 628, 630, 631, 634, 635 RSW to EECW south header keep-fill check valve 0-67-679
'ECW to Ul Control bay chillers check valves 0-67-538, 539, 652, 653 EECW to U3 Control Bay chillers check valves 3-67-761, 762, 764, 765, 771, 772, 774, 775 EECW to Ul Emergency chiller check valves 1-67-787, 789 EECW to U2 Shutdown Board room chiuers check valves 2-67-873, 876 EECW to U3 Diesel Generator cooler check valves 3-67-693, 694, 695, 696, 703, 704, 705, 706, 713, 714, 715, 716, 723, 724, 725, 726 EECW to U3 Shutdown Board room chiller check valves 3-67-735, 736, 737, 738 RCW to H202 panel check valves24-886, 891(Ul and U2 only)
RCW to north EECW header keep-fill check valve 3-24-826 Category:
Class:
C 3 and Non-ASME Code Class Function:
EECW - Pass rated flow for the emergency coolers and prevent backflow from the opposite header.
RCW - Prevent backflow of EECW into RCW.
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Impractical Test Requirement:
Basis for Relief:
IWV-3521 and 3522 - Exercise valves once every three months System design prevents the valves from being verified closed by reverse flow or other conventional means. Most of the valves are verified open quarterly by flow verification, and apparent disc free movement is indicated then. These check valves cannot be forced closed quickly enough to generate a detectable acoustic signal. The isolation valves used to force the valves closed are not capable of quick opening or closing (the valves have handwheel operators). This means that most of
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RELIEF REQUEST NUM3ER PV-14 (continued) the check valve discs contact their seat lightly, generating insufficient noise to be detected by the acoustic monitor used in testing. Noise in the system masks whatever closing signals are generated, and the close proximity of series check valves makes it extremely difficultto identify closure of individual check valves. The test equipment purchased to perform nonintrusive testing cannot obtain an ultrasonic trace of disc position for stainless steel check valves, which includes most of these check valves. Nonintrusive testing of these valves has therefore been determined to be impractical due to the inability to obtain repeatable acoustic and ultrasonic signals. Radiography has also been determined to be impractical after a review of test shots on similar check valves. The radiographs that were reviewed did not reveal sufficient detail to determine valve closure. Test equipment for detecting disc position using
. magnetics has not been purchased and was therefore unavailable for use, in determining feasibility.
Alternative Testing:
These check valves willbe grouped according to design and service conditions and disassembled on a rotating basis in accordance with Position 2 in NRC Generic Letter 89-04. Ifa check valve selected for disassembly fails the criteria of GL 89-04, then the other check valves in the affected test group willbe disassembled and inspected during the same system or refuel outage (whichever is applicable). Because some of the valves are inspected at times other than refuel outages, disassembly of all remaining valves in an affected test group willbe done expeditiously, but with regard for effect on operating systems.
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RELIEF REQUEST NUMBER PV-1S System:
High Pressure Coolant Injection (HPCI) (73)
Reactor Core Isolation Cooling (RCIC) (71)
Drawing:
1, 2, 3-47E812-1 (HPCI) 1, 2, 3-47E813-1 (RCIC)
Components:
HPCI Turbine Exhaust check valves73-633, 634, 635, 636 RCIC Turbine Exhaust check valves71-597, 598, 599, 600 Category:
Class:
Function:
HPCI Turbine Exhaust Vacuum Relief RCIC Turbine Exhaust Vacuum Relief Impractical Test Requirement:
IWV-3521 and 3522 - Cycle valve once every three months Basis for Relief:Valve configuration prevents the valves from being proven open or closed on an individual basis. A flow path through the valves as a whole could be verified quarterly, but the location of the valves on the turbine exhaust piping for the HPCI and RCIC turbines could be hazardous to plant personnel and result in unnecessary dose unless the HPCI and RCIC systems were taken out of service to perform
'he test. This would require entry into a LCO due to HPCI or RCIC unavailability and would raise the Probabilistic Safety Analysis (PSA)
Core Damage Frequency (CDF) by 193% with RCIC out of service (CDF with HPCI out of service would be greater). Nonintrusive testing using acoustic and ultrasonic test equipment has been evaluated and determined to be impractical due to limitations in piping configuration and test equipment. The two types of test equipment considered in the evaluation are those used for flow measurement (ultrasonics) and for verification of piston position (ultrasonics and acoustics). These valves are lifttype check valves installed in a parallel branch line configuration with a crosstie between the two branches.
These valves are in close proximity to each other and are usually dry. There is not sufficient length of straight pipe to mount ultrasonic transducers for measuring flow. The test equipment used to perform ultrasonic flow and position testing requires piping fullof water in order to measure flow or detect piston position. It would be difficultto inject enough water to maintain the piping fullof water so that the test could be performed repeatedly.
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Relief Request Number PV-18 (continued)
There is also the problem of detecting each individual check valve opening from among the other three check valves. Acoustic detection of individual valve opening would also not be possible with the noise associated with the flow necessary to open all the valves, especially if air is used. Radiography is not considered practical due to a review of test shots on other liftcheck valves of similar size and material. The radiographs that were reviewed did not reveal sufficient detail to determine whether the valve pistons were completely seated. In order to obtain a radiograph of the valves in the open position, a prolonged injection would have to be performed.
Also, the coordination and setup of equipment on top of the torus would be difficult. Test equipment for detecting disc position using magnetics has not been purchased and was
..therefore unavailable for use in determining feasibility.
Alternative Testing:
Abilityof the valves as a group to open and break vacuum in the HPCI/RCIC turbine exhaust pipe willbe demonstrated by testing at a cold shutdown frequency (provided ALARAand plant conditions allow).
By connecting a pressurized air source to a test tap downstream of the valves, a flow path through at least two of the four valves willbe demonstrated.
This verification willbe supplemented by the existing once per operating cycle disassembly inspection. These valves willbe grouped according to design and service conditions and disassembled on a rotating basis in accordance with NRC Generic Letter 89-04, Position 2. Ifthe valve selected for disassembly fails the criteria of GL 89-04, then the remaining valves in the affected test group willbe disassembled and inspected during the same system or refuel outage (whichever is applicable). The design of these check valves results in less chance of problems caused by disassembly and reassembly.
Because the piston is positioned down inside the valve body until forced open by steam pressure, there is virtually no chance of kinking the piston in the valve cylinder when reinstalling the valve bonnet. The inspection and cleaning (as necessary) of the valve internals after disassembly ensures that the valve piston is free to move when it is reassembled.
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RELIEF REQUEST NUINBER PV-21 System:
Drawing:
Control Rod Drive (CRD) (85) 1, 2, 3-47E820-2 2-47E820-5 1, 3-47E820-6 Components:
Scram inlet valve FCV-85-39A (185 valves per reactor)
Scram outlet valve FCV-85-39B (185 valves per reactor)
Charging water header check valve FCV-85-589 (185 valves per reactor)
Cooling water header check valve FCV-85-597 (185 valves per reactor)
Scram discharge header check valve FCV-85-616 (185 valves per reactor)
Category:
B (FCV-85-39A and 39B)
C (FCV-85-589, 597, and 616)
Class:
1 (FCV-85-39A and 39B, CKV-85-597) 2 (CKV-85-589 and 616)
Function:
Control Rod Scram Water Flow Path Impractical Test Requirement:
IWV-3411 and 3413 - Exercise valves every three months and measure stroke time IWV-3521 and 3522 - Exercise once every three months.
'Basis for Relief:
These'valves located on the hydraulic control-units for the 185 control rod drives function on a reactor scram signal from the reactor protection system to insert the control rods rapidly into the reactor core.
Cycling these valves requires scramming a control rod.
There are 185 control rods in the reactor. Scramming every rod once every three months is not practical for the following reasons:
a.
A power reduction is required to test the scram function.
Reducing power for the length of time required to scram 185 rods is not practical.
b.
Fuel preconditioning must follow this power reduction to avoid possible fuel damage.
The longer the reduction in power, the longer the preconditioning.
Their proper functioning is most practically verified by an actual scram test (except for the closure of 85-589).
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Relief Request Number PV-21 (continued)
Alternative Testing:
Scram testing and rod insertion timing willbe performed in accordance with Technical Specifications Section 4.3.C (at reactor coolant pressure 800 psig) for:
a Allcontrol rods prior to THERMALPOWER exceeding 40 percent after each refueling outage.
b.
10 percent on a rotating basis at least once every 16 weeks.
Valve 85-589 willbe proven to close by pressure decay testing once per refueling outage.
This testing is in accordance with NRC Generic Letter 89-04, Position 7.
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RELIEF REQUEST MMBERPV-29 System:
Reactor Core Isolation Cooling (RCIC) (71)
High Pressure Coolant Injection (HPCI) (73)
Drawing:
1, 2, 3-47E813-1 (RCIC) 1, 2, 3-47E812-1 (HPCI)
Components:
RCIC pump suction (from condensate supply) check valve 71-502 RCIC pump suction (from RCIC condensate) check valve 71-589 HPCI pump suction (from condensate supply) check valve 73-505 HPCI pump suction (from HPCI condensate) check valve 73-625 Category:
Class:
Function:
Maintain filled suction piping for the HPCI and RCIC pumps.
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. Since the valves are verified open quarterly by flow verification, apparent disc free movement is indicated at that time. The presence of filled HPCI/RCIC pump discharge piping is verified once a month, providing indirect evidence of the closed position of the pump suction check valves. The HPCI/RCIC condensate return check valves allow condensate return from the RCIC and HPCI local condensers to the pump suction lines during the quarterly operability surveillance tests. This indicates open position of the condensate return line check valves. Nonintrusive testing has been evaluated and determined to be impractical. It has been decided based on experience with ultrasonic flow testing on the pump minimum flow lines that a repeatable acoustic or ultrasonic signal could not be obtained amid the noise and vibration associated with turbine operation. The startup of the RCIC and HPCI turbines are high-risk situations requiring personnel to evacuate the immediate area of the turbines and it is undesirable to continue turbine operation any longer than necessary due to concerns over torus heatup. Any problems in obtaining a conclusive acoustic or ultrasonic trace could prolong turbine operation. Radiography is not considered practical due to a review of test shots on other check valves of similar size and material. The radiographs that were reviewed did not
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Relief Request Number PV-29 (continued) reveal sufficient detail to determine whether the valve discs were completely seated. Test equipment for detecting disc position using magnetics has not been purchased and was therefore unavailable for use in determining feasibility.
Alternative Testing'.
These valves willbe grouped according to design (including size) and service conditions. They willthen be proven to close once per operating cycle by disassembly and inspection on a rotating basis in accordance with Position 2 of NRC Generic Letter 89-04. Ifa check valve selected for disassembly fails the criteria of GL 89-04, then the remaining valves (ifany) in the affected test group willbe disassembled and inspected and corrective actions performed (ifnecessary).
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RELIEF REQUEST NUMBER PV-30 System:
Drawing:
Components:
Category:
Class:
Standby Liquid Control (SLC) (63) 1, 2, 3-47E854-1 (SLC)
SLC pump discharge check valves63-514 and 516 Function:
Provides standby liquid control pump discharge flow path and prevents backflow.
Impractical Test Requirement:
IWV-3521 and 3522 - Exercise valves once every three months.
Basis for Relief:
The standby liquid control system pumps are positive displacement pumps which by their design prohibit backflow through the pump. There exists no flow path by which the backseating capability of these check valves can be demonstrated either during system operation or plant shutdown. Nonintrusive testing has been evaluated and determined to be impractical due to limitations associated with the system configuration and test equipment. The system piping is stainless steel and willnot allow ultrasonic detection of disc position with the test equipment. Use of ultrasonic flowmeters is not practical due to the numerous changes in pipe direction and size. The positive displacement pumps produce noise
'and vibration from the pulsating discharge which makes it difficultifnot impossible to detect valve closure using acoustics. Radiography is not considered practical due to a review of test shots on other liftcheck valves of similar size and material. The radiographs that were reviewed did not reveal sufficient detail to determine whether the valve pistons were completely seated.
Test equipment for detecting disc position using magnetics has not been purchased and was therefore unavailable for use in determining feasibility. Since these valves are stainless steel, the system utilizes demineralized water, and the pumps are operated quarterly, there is virtually no chance of the valves becoming stuck in the open position. The inspection history of the valves bears this out.
The Unit 3 valves, which have been in lay-up for almost ten years, looked the same as the Unit 2 valves during inspections performed in preparation for Unit 3 startup. During the Unit 2 refueling outage inspections conducted on these-valves, there has been no cleanup or other corrective actions required.
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Relief Request Number PV-30 (continued)
Alternative Testing:
These valves willbe grouped according to design and service conditions.
One check valve willbe disassembled at each refueling outage on a rotational basis in accordance with Position 2 of NRC Generic Letter 89-04. Should the disassembled valve fail to function properly, the other valve willalso be disassembled and examined during the same refueling outage. Because there are only two of these valves per unit, each valve willbe disassembled at least once every other refueling outage. These valves willbe exercised to the open position during quarterly pump testing. The design of these check valves results in less chance of problems caused by disassembly and reassembly.
Because the piston is positioned down inside the valve body until forced open by water pressure, there is virtually no chance of kinking the piston in the valve cylinder when reinstalling the valve bonnet. The inspection and cleaning (as necessary) of the valve internals after disassembly ensures that the valve piston is free to move when it is reassembled.