ML18038B427
| ML18038B427 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/15/1995 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9509250019 | |
| Download: ML18038B427 (15) | |
Text
PR1ORITY 1.
tACCELERATED RZDS PROCESSING)
REGULATORY XNFORMATION DISTRIBUTION SYSTEM (RXDS)
ACCESSION NBR:9509250019 DOC.DATE: 95/09/15 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 AFFILXATXON SALAS,P.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATXON I
Document Control Branch (Document Control Desk)
SUBJECT:
Provides updated methodology for testinq per pump 6 valve testing program, specifically clarification include testing in accordance w/ASME Section XI for core spray testable check valves.
DISTRIBUTION CODE:
A047D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code GL-89-04 NOTES:
RECIPIENT ID CODE/NAME PD2-3 WILLIAMS,Z.
INTERNAL: ACRS CENTE 01 EMEB OGC/HDS3 EXTERNAL: LITCO ANDERSON NRC PDR COPIES LTTR ENCL 1
1 1
1 6
0 1
1 1
1 1
0 0 1
1 1
1 RECIPIENT ID CODE/NAME PD2-3-PD AEOD/SPD/RAB NRR/DE/EMCB NUDOCS-ABSTRACT RES/DSXR/EIB NOAC COPIES LTTR ENCL 1
1 1
1 1
1 1
1 1
1 1
1 D
N NOTE TO ALL "RZDSe'ECIPIENTS PLEASE HELP US TO REDUCE WASTE!
CONTACT THE DOCUMENT CONTROL
- DESK, ROOM OWFN 5D8 (415-2083)
TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 19 ENCL 12
Tennessee Valley Authority. Post Office Box 2000, Oecatur, Alabama 35609 September 15, 1995 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
10 CFR 50.55a(g)
In the Matter Of Tennessee Valley Authority Docket Nos.
50-259 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - TVAiS AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
SECTION ZI PUMPS AND VALVES TESTING (P&VT)
PROGRAM CLARIFICATION AND CHANGE FOR BFN UNITS 1q 2q AND 3 The purpose of this letter is to provide TVA's updated methodology for testing in accordance with BFN's Units 1, 2, and 3
P&VT program; specifically, the clarification includes TVA's testing in accordance with ASME Section XI for the Core Spray (CS) testable check valves.
In a letter from TVA to NRC, dated May 2,
- 1994, "Browns Ferry Nuclear Plant (BFN)
TVA'S American Society of Mechanical Engineers (ASME)Section XI Pumps and Valves Testing (P&VT)
Program Clarifications and Changes for BFN Units 1, 2,
and 3," TVA provided a relief request.
PV-25 that described alternate testing for the Residual Heat Removal (RHR) and CS testable check valves.
TVA's minimum testing was to full stroke the testable check valves.
However, it was subsequently determined that the CS testable check valves are only partial-stroked by the valve actuator.
As a result, TVA is revising relief request PV-25 to eliminate the portion pertinent to the CS testable test valves.
Relief request PV-37 addresses alternate testing of the CS testable check valves in accordance with ASME/ANSI OM-10.
l piggy Q 95092500i9 950915 PDR ADOCK 05000259
.PDg
U.S. Nuclear Regulatory Commission Page 2
September 15, 1995 Enclosure 1 is the revised relief request PV-25.
The original request addressed relief in accordance with OM-10 for both the RHR (74-54 and 74-68) and CS (75-26 and 75-54) testable check valves.
Since relief request PV-37 (Enclosure
- 2) addresses the two CS testable check valves, PV-25 no longer needs to consider these valves.
Consequently, Enclosure 1 deals with only the two RHR testable check valves.
'h Enclosure 2 provides revised relief request PV-37.
This revised relief request documents BFN's methodology for testing the CS testable check valves.
This method'f testing considers safety factors for plant personnel and equipment.
TVA proposes that as a minimum that both CS testable, check valves will be partially stroked from the control room without compromising the intent of ASME requirements.
This proposal is also based on confirmation from the vendor that the 30 degree test (partial stroke) is a.sufficient demonstration that the disc'is not stuck in the seat.
Additionally, the vendor substantiated that no mechanism is possible to prevent the disc from travelling to the full open position once it has reached the 30 degree open position.,
Therefore, a partial stroke test is sufficient to prove operability of the valves.
Relief request PV-37 is provided for your review and approval.
There are no commitments contained in this letter.
Xf you have any questions please contact me at (205) 729-2636.
Sincerely p
alas Manager of Site Licensing Enclosure
U.S. Nuclear Regulatory Commission Page 3
September 15, 1995 cc (Enclosure):
Mr. Mark S. Lesser, Acting Branch Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Str'eet, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road
- Athens, Alabama 35611 Mr. J.
F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852
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ENCLOSURE 1
TENNESSEE VALLEY AUTHORZTY BROWNS FERRY NUCLEAR PLANT (BFN)
UNZTS 1q 2I AND 3 METHODOLOGY FOR TESTXNG THE RESZDUAL HEAT REMOVAL (RHR)
TESTABLE CHECK VALVES
SUMMARY
DESCRXPTZON ENCLOSURE 1
PV-25 Deleted portions pertaining to the Core Spray testable check valves.
REVISED RELIEF REQUEST PV-25 System:
Drawing:
Components:
Category:
Class:
Function:
RHR (74 )
1g 2~
3 47E811 1
(RHR)
Testable check valves 74-54, 74-68 AC Valves open to allow emergency/shutdown cooling water supply to the reactor.
Valves close to maintain primary containment isolation and prevent loss of reactor coolant.
Impractical Requirement:
Basis for Relief:
IWV-3521 Cycle valves quarterly.
IWV-3522 Cycle valves during cold shutdown (CSD) if impractical to cycle quarterly.
These valves are located inside the drywell (primary containment) where the atmosphere is inerted during operation as required by Technical Specification (TS) 3.7.A.5.a.
(and may remain inerted during CSD depending on the reason for going to CSD).
Due to potentially inadvertent valve operation caused by non-class 1E circuitry to the valve operator, the air supply to each valve operator is normally disconnected and the valves cannot be cycled quarterly (self-actuation is unaffected).
Entry to the drywell to reconnect the air supply to test these valves would be hazardous to personnel unless the unit is in CSD with the drywell atmosphere de-inerted.
Cycling the RHR testable check valves with flow is possible during CSD using the shutdown cooling mode of RHR.
However, it may not be possible to cycle both valves during the same shutdown.
TSs 3.5.B.2 and 3.5.B.9 require at least one RHR loop be maintained operable for containment cooling.
This permits flushing the other loop in preparation for being placed in CSD, but prevents both loops from being flushed at the same time.
Flushing can take 2-4 hours.
Following completion of flushing, the valve in that loop can be tested, then the loop realigned for standby readiness.
However, during the time that the'first loop is being returned to standby readiness and the second loop is being flushed, no shutdown cooling is in service.
If decay heat is still significant, heating of the moderator could occur such that
containment integrity would have to be maintained.
This could also cause RHR SDC to be automatically isolated on high reactor pressure (greater than or equal to 105 psig).
Alternate Testing:
A minimum of one of the testable check valves will be cycled during each CSD (provided three months has passed since the previous CSD).
This will be done by verifying the valve opens using normal shutdown cooling flow (7000 GPM minimum) and closes on cessation of shutdown cooling flow. If TSs and plant, conditions allow, the opposite testable check valve will be cycled using shutdown cooling flow during the same CSD. If both valves cannot be cycled during the same CSD, then the uncycled valve will be cycled first during the next CSD (provided TSs and plant conditions allow). If entry to the drywell is possible, the valves may be cycled by connecting temporary air to the valve operators instead of passing flow through the shutdown cooling flow path.
In this
- case, the valves would then be stroked full open and full closed using the control room handswitch.
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ENCLOSURE 2
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1~
2~
AND 3 METHODOLOGY FOR TESTING THE CORE SPRAY (CS)
TESTABLE CHECK VALVES
SUMMARY
DESCRIPTION Clarified basis for relief section.
Documented testing methodology for the CS testable check valves.
RELIEF REQUEST PV-37 System:
Drawing:
Components:
Category:
Class:
Function:
CS (75) 1 I 2 ~
3 47E8 14 1
(CS)
Testable check valves 75-26, 75-54 AC Valves open to allow emergency cooling water supply to the reactor.
Valves close to maintain primary containment isolation and prevent loss of reactor coolant.
Impractical Requirement:
1986 Edition of ASME Section XI paragraph IWV-3521 (Check valves shall be exercised at least once every three months, except as provided by IWV-3522).
1986 Edition of ASME Section XI paragraph IWV-3522 "If full-stroke exercising of a check valve on a quarterly basis is impractical during plant operations, check valves shall be part-stroke exercised during plant operation and full-stroke exercised during cold shutdowns."
In addition, the Safety Evaluation Report for the Browns Ferry Inservice Test Program Second Ten-Year Interval (reference letter from Fredrick J.
Hebdon to Dr. Mark O. Medford dated October 22, 1993) approved Request for Relief PV-25 (the original request for relief for these valves) provided all applicable requirements ASME Operations and Maintenance Standard (OM), Part 10, 1987 Edition through the OMa-1988 addenda were complied with in testing of these valves, OM-10 paragraph 4.3.2.1 states, "Check valves shall be exercised nominally every three months except as provided by paragraphs 4.3.2.2, 4.3.2.3, 4.3.2.4 and 4.3.2.5.
Paragraph 4.3.2.2(d) states: "If exercising is not practicable during plant operation and full-stroke during cold shutdowns is also not practicable, it may be limited to part-stroke during cold shutdowns and full-stroke during refueling outages."
Specifically, the impractical requirement for testing of these valves is full-stroke test during cold shutdown and refueling outages.
E2-2
if
Basis for Relief:
These valves are located inside the drywell (primary containment) where the atmosphere is inerted with a nitrogen atmosphere during operation as required by Technical Specification 3.7.A.5.a.
(and may remain inerted depending on the reason for going to CSD).
Due to potentially inadvertent valve operation caused by non-class 1E circuitry to the valve
- operator, the air supply to each valve operator is disconnected and the valves cannot be cycled quarterly (self-actuation is unaffected).
Entry to the drywell to reconnect the air supply to test these valves would be hazardous to personnel unless the unit is in CSD with the drywell atmosphere deinerted.
Cycling of CS valves with flow is possible during CSD but not practical.
The design flow rate of the CS system is 6250 gpm.
The addition of this amount of water to the reactor vessel would challenge maintaining reactor water level during CSD and could potentially lead to flooding the main steam lines.
Flooding the main steam lines could delay returning the unit to service until the lines were drained and dried.
Additionally, reactor water chemistry could be adversely affected due to the lower quality of pressure suppression chamber water that would be injected into the reactor vessel to open the testable check valves.
In a letter dated May 15, 1995, the vendor noted that because the disc rotated on two hinge pins, there is no mechanism which would enable the disc to become stuck at any intermediate open position.
Nevertheless, an Engineering calculation for Units 2 and 3
CS loops A and B was performed to determine the maximum flow if a CS testable check valve began to open and became stuck in the 30 degree position.
The results of the calculation was that no loop would satisfy the Technical Specification (TS) criteria of 6250 gpm.
The lowest flow was calculated to be 6086 gpm (greater than 97 percent of rated flow).
However, if the condition warranted an additional CS loop because the valve stuck in the 30 degree position, the Control Room Unit Operator would be able to place the other loop in service and supply more than the TS required amount of water.
E2-3
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A search of the Oak Ridge National Laboratory Nuclear Plant, Reliability Data System (tracking program) was conducted.
This search did not reveal any failures of this type of valve to travel full open once they have been lifted off the valve seat.
Free movement up to 30'roves sufficient ability to open completely.
Additionally, the valve vendor has stated that failure of the valve disc to fully open once the valve has been lifted off the valve seat is not a credible event.
There is no known mechanism that would prevent the valve disc from going full open after showing free movement up to 30 open.
The testable check valve actuators only open the valve disc approximately 30 (full open is 75~) in accordance with the original design.
Full stroking of these valves by removing the actuator, stroking the valve disc full open using a torque wrench, reinstalling the actuator, resetting the limit switches, and performing the partial stroke test as verification of operability does not, provide a compensating increase in the level of quality and safety of the plant.
The only gain would be verification that the valve disc continued to demonstrate free motion and opened an additional 45 degree to the full open position.
Because of the amount of work involved, a full stroke of these valves cannot be practically performed during cold shutdowns.
The only time when sufficient time would exist to perform a full stroke of these valves would be during a refueling outage.
Even during a refueling outage, this work would require significant resource expenditure.
It could also cause problems since it would involve partial disassembly and reassembly of the valve.
Alternate Testing:
These valves will be partial stroked in accordance with cold shutdown guidelines described in NUREG-1482 when the drywell atmosphere is deinerted and personnel can safely perform the work without exceeding ALARA guidelines.
- However, as a minimum, these valves will be partial stroked (30') during each refueling outage.
This would be done by temporarily connecting an air supply to the permanent valve operators.
The valves would then be partially stroked open and closed using the control room handswitch.
This will prove free movement of the valve disc.
All other testing requirements of OM-10 will be followed.