ML20056C067

Forwards Changes & Clarifications to Inservice Testing Program for Pumps & Valves for Second 10-yr Interval,Per 930203 & 11 Telcons W/Nrc

ML20056C067
Person / Time
Site:	Browns Ferry
Issue date:	03/19/1993
From:	Zeringue O TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9303260152
Download: ML20056C067 (14)

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MAR 191993 +

r O.J.1ke'Zenngue va rees,an swns rey m:.wnam U.S. Nuclear Regulatory Commission i ATTN: Document Control Desk i Washington, D.C. 20555  ;

Gentlemen: I i

In the Matter Of ) Docket Nos. 50-259 l Tennessee Valley Authority ) 50-260  ;

50-296 l BROWNS FERRY NUCLEAR PIANT (BFN) - UNITS 1, 2 AND 3 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI INSERVICE TESTING PUMPS AND VALVES-PROGRAM (P&VT), CHANGES AND CIARIFICATIONS l t

Reference:

Letter from TVA to NRC, dated August 31, 1992, "BFN Units 1, 2 and 3 ASME Section XI Inservice Testing Pumps and .

Valves Program for the Second Ten-Year Interval" 1 c

The purpose of this letter is to provide changes and clarifications to the  ;

P&VT program. Teleconferences were held on February 3, 1993, and February 1 11, 1993, with NRC program reviewers and TVA personnel to discuss ,

technical issues relative to the referenced program. During these  !

discussions, TVA agreed to clarify its position regarding several P&VT  ;

program issues.

These clarifications and changes will be implemented within 90 days of receipt of an approved Safety Evaluation Report (SER) for the current 10-year Inservice Test Program pumps and valves except for the withdrawal of relief request PV-34 The actions resulting from the withdrawal of PV-34 ,

(i.e., rebaseline pumps, retraining, procedures update) will be '

implemented within 6 months of the receipt of the SER.

Enclosure 1 contains a description of the program changes and clarifications. Enclosure 2 contains the revised Requests for Relief.

There are no commitments contained in this submittal. If ycu have any questions, please contact G. D. Pierce, Interim Manager of Site Licensing, at (205) 729-7566.

Sincerely, 0.

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.Z kh ingue ,

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U.S. Nuclear Regulatory Commission MAR 19199}

Enclosures

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NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637 Athens , Alabama 35611  ;

Mr. Thierry M. Ross, Project Manager  !

U.S. Nuclear Regulatory Commission  ;

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. B. A. Wilson, Project Chief '

U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900

• Atlanta, Georgia 30323 i

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ENCIDSURE 1  ;

BROWNS FERRY NUCLEAR PIANT

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DESCRIPTION OF PROGRAM CHANGES AND CLARIFICATIONS i

(1) Withdraw Relief Request PV-12. After further evaluation it was [

determined that the Residual Heat Remover (RHR) HTX outlet valves *

(FCV-23-34, 40, 46, and 52) can be stroked to the full-open position instead of the current throttled position.

(2) Withdraw Relief Request PV-16. Testing will be in accordance with Section XI of the ASME Boiler and Pressure Vessel Code.  ;

(3) Modify Relief Request PV-20. This change is an update to delete the l following valves: FCV-74-61. FCV-74-75, FCV-73-26, and FCV-71-17. i These valves will be tested in accordance with IWV-3423(c). The remaining valves will be tested in accordance with the guidance provided in Generic Letter (GL) 89-04, Position 10. ,

(4) Withdraw Relief Request PV-26. This valve,85-576, is a passive i containment isolation valve.

(5) Modify Relief Request PV-27. For valves FCV-1-14, 26, 37, and 51, provide additional information for basis of testing actually performed and added taking exception to IWV-3423(f). Testing vill be in accordance with the guidance provided in CL 89-04, Position 10.

I (6) Modify Relief Request PV-28. Delete Contained Atmosphere dilution system containment isolation check valves (84-600, 601, 602, 603, 617, and 680). .

(7) Withdraw Relief Request PV-34 Pump vibration data was recorded using ,

veloci n measurements during the first ten-year inspection interval.

To comply with the Code requirements of Subsection IUP, pump vibration i measurements will be recorded using displacement measurements. ,

(8) Add Relief Request PV-35. This addition is being made to document the <

current practice of performing corrective action per the Appendix J Leak Test program in-lieu of IWV-3427(b) for category A valves nominal '

pipe size 6 inches and larger. Testing will be in accordance with the guidance provided in CL 89-04, Position 10.

(9) Cold shutdown justification Number 7 is being withdrawn. The surveillance instruction (SI) which currently performs the cold shutdown testing will become a quarterly SI.

(10) Add cold shutdown justification Number 11 to provide the basis for cold shutdown frequency testing for the closed position of valves73-559 and 71-547 '

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1 BROTJNS FERRY NUCLEAR PMNT  :

REQUEST FOR RELIEF AND ,

COLD SHUTDOUN JUSTIFICATIONS l

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2 Request for Relief FV-12 This request for relief is withdrawn. Valves FCV 23-34, 40, 46 and 52 will be stroke timed to the full-open position in accordance with the applicable requirements of Section XI of the ASME Boiler and Pressure Vessel Code.

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Request for Relief PV-16 'l l

r This request for relief is withdrawn. Valves71-547 and 73-559 vill be tested I in accordance with the applicable requirements of Section XI of the ASME 3 Boiler and Pressure Vessel Code. ,

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t Request for Relief PV-20 j System: Main Steam (1) _ [

High Pressure Coolant Injection (HPCI)(73)  !

Reactor Core Isolation Cooling (RCIC)(71) 5 Drawing: 47E801-1 47E812-1 -

47E813-1 Components: Valve FCV-1-55 i Valve FCV-73-2  !

Valve FCV-71-2 l Class: 1 and 2 .i Category: A  ;

Punction: Containment Isolation. These valves perform containment isolation functions only, and are not j pressure isolation valves, j Impractical Test i Requirement: IWV-3423 - Valve seat leakage tests shall be made with l the pressure differential in the same direction as when the valve is performing its function.

Basis for Relief: The subject valves are the first valve off the reactor ,

in each of their respective piping sections. Due to  !

this location in the piping system, the normal procedure of leakage rate testing by pressurizing the >

piping volume between two or more valves is not.  ;

possible. It is impractical to perform leakage rate j test on these valves by pressurizing them from the 'i reactor vessel side.  ;

Alternate Testing: The plant containment leakrate testing program (for i compliance with 10 CFR 50 Appendix J requirements) has ,

established a leakage rate testing procedure which  !

pressurizes each of these valves in-the reverse i direction. This test method shall be utilized in lieu '

of the requirements of IWV-3421 through 3425. The ,

requirements of IWV-3426 and 3427(a) will be met.  :

This alternative is in compliance with Generic Letter j 89-04, Position 10. ,

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Request for Relief PV-26 ~l

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.i This request for relief is withdrawn. Valve 85-576 will be tested in I accordance with the applicable requirements of Section XI of the ASME Boiler i and Pressure Vessel Code, j r

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k7 . t Request for Relief PV-27 System: Main Steam (1) ,

Drawing: 47E801-1 Components: Valves FCV-1-14, 26, 37, 51  !

' Class: 1 Category: A j Function: Main steam isolation.

Impractical Test Requirement: IWV-3423 - Valve seat leakage tests shall be made with ,

the pressure differential in the same direction as when the valve is performing its function. IWV-3423(f) -

Valves not qualifying for reduced pressure testing as  ;

defined in (e) above shall be leak tested at full

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maximum function pressure differential, with adjustment i by calculation if needed to compensate for a difference ,

between service and test media. ,

Basis for Relief: The subject valves are the first valve off the reactor  :

in each of the main steam lines. Due to this location 4 in the piping system, the normal procedure of leakage  ;

rate testing by pressurizing the piping volume between  !

two or more valves.in the functional direction of-the subject valve is not possible. It is impractical to-perform leakage rate test on these valves by pressurizing them from the reactor vessel side.

Due to the design of these valves and their being tested in the reverse flow direction, excess test  ;

This pressure will force the valve disk off che seat.

condition is recognized in the plant Technical Specification 4.7. A.2.1, which req' tires that the MSIVs +

be tested at 25 psig and leakage rates must be less ,

than 11.5 standard cubic feetAour. This condition  !

exists at many BWR plants.

Alternate  !

Testing: The plant containment leakrate testing program (for  ;

compliance with 10 CFR 50 Appendix J requirements) has ,

established a leakage rate testing procedure which ,

pressurizes each of these valves in the reverse direction and at the reduced pressure specified in .

plant Technical Specification 4.7.A.i. This test - i method shall be utilized in lieu of the requirements of  ;

IWV-3421 through 3425. The requirements of IWV-3426 and 3427(a) will be met. ,

t This alternative is in compliance with Generic Letter 89-04, Position 10. .

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i Request for Relief FV-28 ,

t Residual Heat Removal (RHR)(74) system:

Drawing: 47E811-1 ,

Components: Valves74-661, 662 Class: 1  ;

Category: AC Function: Thermal relief / containment isolation >

Impractical Test i

Requirements: IVV-3521 and 3522 - Check valves shall be exercised at least once every 3 months, or part stroked every 3 months and full stroked every cold shutdown.  ;

Basis for Relief: These valves provide thermal relief between the inboard '

and outboard shutdown cooling suction valves, Cycling  ;

these valves by system manipulations is not possible during operations due to pressure interlocks which prevent the suction valves from opening. These valves' ,

are located in the containment ~drywell which is inerted j during unit operation and which remains inerted during ,

cold shutdown periods when drywell entry is not l required.

Alternate i Testing: Proper valve opening will be tested at cold Si.utdown  !

when drywell entry and adequate time permit. Valve closurs will be demonstrated at each refueling outage by the performance of 10 CFR 50 Appendix J 1eakage rate testing. This alternate testing conforms to the i requirements of Generic Letter 89-04, Position 10.

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Request for Relief PV-34  ;

t This request for relief is withdrawn. Pump vibration testing will be  !

performed in accordance with the requirements of Subsection 1WP. .l t

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4 Request for Relief PV-35  ?

Various i System:

Drawings: Various Components: Valves (NPS 6 inches and larger)

' Class: 1, 2, and 3  ;

Category: A AC Function: Containment isolation.

Impractical Test Requirement: IWV-3427(b) - For valves nominal pipe size (NPS) 6 and larger, if a leakage rate exceeds the rate determined' by the previous test by an amount that reduces the  !

margin between measured leakage rate and maximum permissible rate by 50% or greater, the test frequency shall be doubled; the tests shall be scheduled to-coincide with a cold shutdown until corrective action -[

is taken, at which time the original test frequency shall be resumed. If tests show a leakage rate increasing with time, and a projection based on three or more tests indicates that the leakage rate of the next schedule test will exceed the maximum permissible  ;

leakage rate by greater than 10%, the valve shall be replaced or repaired. l Basis for Relief: The testing requirements for containment isolation valves are established by Appendix J of 10 CFR 50. The ,

acceptance criteria for the primary containment '

integrity in the Appendix J program contains a margin ,

for valve degradation between successive tests. ne  !

margin of valve performance resulting from paragraph i IWV-3427(b) is therefore redundant for valves which are tested under the Appendix J test program.  ;

Alternate ,

Testing: The leak rate testing of all containment isolation valves will be performed using the methodology and -

requirements of Appendix J to 10 CFR 50, and 1WV-3426

• and IWV-3427(a). This alternate testing conforms to the requirements of Generic Letter 89-04, Position 10.

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Cold Shutdown Justification Number 7

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,j This cold shutdown justification is withdrawn. t

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Cold Shutdown Justification Number 11 System: High Pressure Coolant Injection (HPCI)(73)

Reactor Core Isolation Cooling (RCIC)(71)  ;

Drawing: 47E812-1 (HPC1) l 47E813-1 (RCIC) ~t Components: HPCI Pump Minimum Flow Line Check Valve (73-559)

RCIC Pump Minimum Flow Line Check Valve (71-547)

Category: A, C Class: 2  ;

Function: Containment Isolation t

Justification:

To prove these HPCI/RCIC valves systems outclosed it is necessary)to of service take (inoperable long enou@ to isolate the check valves, drain the upstream portion of the system, and verify no leakage past the .

valves by absence of visible leabge past the valve or  !

by the abili_y to pressurize the piping downstream of  ;

the valve. ne Probabilistic Risk Assessment (FRA) for i the Unit 2 RCIC system gives a Core Damage 4 Frequency (CDF) with RCIC available of 4.78 x 10 per Reactor ,

Year. The PRA also gives a CDF with RCIC unavailable of 1.4 x 10~* per Reactor Year (an increase of 193%) .

The impact due to HPCI unavailabili g would be greater.

Taking HPCI/RCIC out of service is therefore an activity that needs to be weighed against the benefits  ;

gained from the activity. In this case, the benefit from testing the check valves for closure is the assurance that the containment isolation function of  !

the valves is intact. Because there is a motor-operated valve in the same lines, failure of these check valves in the open position would not by themselves fail containment. Any leak past the pump minimum flow valves would be detected by radiation monitors in the pump rooms. The seat leakage i r

capability of the valves is verified once per operating cycle by the Appendix J Leak Rate program. Because the  !

water used to charge the system is supplied by the  :

Condensate Storage Tank there is no reason to suspect  ;

any corrosion problem. The history of these valves t indicates no problems of any kind and the benefit of ,

performing the quarterly tests is outweighed by the '

i increased risk to the core resulting from taking the '

systems out of service. These valves will be tested on a cold shutdown basis provided there is adequate time  !

to complete the tests prior to return of the unit to >

power operation.  ;

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