ML20212J004

Requests Relief from Tech Spec 4.7.A.3 for Performing RHR Wear Ring Replacement.Tech Spec Cannot Be Effectively Applied,Does Not Appear in STS & for Each Penetration Either Testing Plan or Evaluation Performed Negating Tech Spec

ML20212J004
Person / Time
Site:	Vermont Yankee
Issue date:	03/03/1987
From:	Murphy W VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
FVY-87-29, NUDOCS 8703060279
Download: ML20212J004 (4)

Text

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VERMONT YANKEE NUCLEAR POWER CORPORATION FVY 87-29 RD 5, Box 169, Ferry Road, Brattleboro, VT c5301 ,

ENGINEERING OFFICE 1671 WORCESTER ROAD

• FRAMINGHAM, MASSACHUSETTS 01701

. TELE PHONE 617-872-6100 March 3, 1987 U.S. Nuclear Regulatory Commission Regional Administrator Region I 631 Park Avenue King of Prussia, PA 19406

References:

(1) License No. DPR-28 (Docket No. 50-271)

(2) Letter, VYNDC to USNRC, FVY 86-78, Changes to the Technical Specifications for the 1986/1987 Operating Cycle Inspection / Repair of the RHR Pump Impeller Wear Rings, dated 8/28/86 (3) Letter, VYNPC to USNRC, FVY 84-76, " Vermont Yankee Primary Containment Leak Rate Testing Program", dated 6/26/84 (4) Letter, VYNPC to USNRC, FVY 86-97, Supplemental Response to NRC Request for Additional Information - Appendix J Technical Specifications, dated 10/10/86 (5) V.Y. Technical Specification 4.7. A.3 (6) Letter, USNRC to VYNPC, NVY 86-237, dated 12/4/86 (7) Discussion, VY Resident Inspector with VY Staff, 2/27/87

Subject:

Request For Relief From Technical Specification 4.7.A.3 For Performing RHR Wear Ring Replacement

Dear Sirs:

As a result of our review during the past month and per Reference 7, Vermont Yankee requests relief from Technical Specification 4.7.A.3 in order to perform the RHR Wear Ring Replacement in an expeditious and safe manner. T.S.

4.7.A.3 states:

" Prior to violating the integrity of a system outside the Primary Containment which is connected to any valve listed in Table 4.7.28, the isolation valves' bounding the opening shall have Type C tests performed.

If the opening cannot be isolated from the containment by two isolation valves which meet the acceptance criteria of Appendix J (10 CFR Part 50), a blank flange shall be installed on the opening".

8703060279 DR s7o3o3 \q ADOCK 05000271 Q PDR $\

U.S. NuclGar,Rtgulatory Commission March 3,-1987 Page 2'.

. VEHMONT YANKEE NUCLEAR POWEH CORPORATION

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In_ order to initiate and' complete the RHR Wear Ring Replacement, Vermont

. Yankee must utilize T.S. 4.7.A.3.

Isolation valves RHR 17 and RHR 18 which are used for the RHR Shutdown -

Cooling function are listed in Technical Specification Table 4.7.2b and would be i used to'" bound the opening" that will result when the RHR pump is disassembled.

Although not. mentioned in Table 4.7.2b, RHR 13 and RHR 15 valves also isolate j the pump from the primary system (see FSAR Figure 4.8-2). Although it is i

possible to leak test the RHR-17 valve, the results may not be conclusive since -

testing _of the RHR 17 valve for leakage may be affected by the RHR-18 valve i

, water leakage.- RHR-18 has approximately a 1000 psig differential pressure  !

during operating conditions versus a 44 psig differential pressure during acci- l dent conditions. The RHR-18 valve water leak rate under normal conditions could be acceptably higher than .the Appendix J air leak rate under accident con--

i l ditions. -It should-be_ noted, however, that on-line relief valves and pressure switch indications located downstream of RHR18 and RHR17, do not and have not indicated leakage past these valves. RHR 18 valve can not be tested under

_ Appendix J conditions during operation.

Additionally, air'could be introduced into the system which could set up an air hammer potential. The RHR would remain " solid"'if no leak testing is attempted.

The intent of T.S. 4.7.A.3, however, can be met when; t 1) Valve RHR 15 is- tagged and locked shut (this can be considered l equivalent to a blind flange).

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2) Tail pipe leakage and pressure monitoring (downstream of the RHR 17 and 18 valves) are monitored (this can be considered equiva-lent protection to leak testing).

During normal operation, the RHR system forms a seismically qualified loop which penetrates the containment and thus-leakage by the penetration isolation valves, if any, would not have a clear path to the atmosphere. During RHR impeller replacement the RHR 15 valve will be closed and will help provide a water seal on valves RHR 17 and 18 which will also remain closed. This will provide a similar water seal as is available during normal operation. During normal operations (normal reactor pressure), annunciation / instrumentation exists  ;

which checks leakage past penetration X-12 isolation valves (RHR 17 and 18) by monitoring downstream pressure. The pressure setpoint is 100 psig which is monitored by PS-118. Additionally, there is a safety relief valve (SR-40) which is downstream of penetration X-12 that is set at 150 psig. Therefore, if penetration X-12 isolation valves should leak during operation, these components would indicate a problem by annunciation or by tail pipe leakage. Thus, the boundary integrity can be checked which would identify leaks and initiate repairs. As described above, even though VY will not formally leak test these valves, sufficient means are available to give positive assurance that the penetration isolation valves, RHR 17 and 18, will perform their required function.

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I U.S. Nuclear Regulatory Commission l March 3, 1987

Page 3 ,

f . VERMONT YANKEE NUCLEAR POWER CORPORATION It should also be noted that primary contaiment is maintained by water seal .for the RHR-13A, B, C, D valves, which will also be isolated (1 valve per

l. pump isolation) during'this work effort.

To ensure equivalent primary containment protection, the following have been considered; Administrative Controls VY will white tag the appropriate RHR 13 and 15 valves shut in accordance

with existing administrative procedures and will open the motor operator breaker. This will ensure that the valve is maintained closed and is electri-

cally disarmed.

l Leakage Potential I In order to perform the RHR Wear Ring Replacement, the leakage rate will need to be minimized so that the required internal inspections / work can be per-8 formed. The RHR pumps have drain lines.which can drain approximately 5 gpm for i- anticipated condition. If the design basis LOCA occurred during this repair

-period, it is evaluated that:

I a) the leakage rate'woul'd significantly decrease in the RHR shutdown cooling line. (i.e. valves RHR 18, 17, 15) because differential pressure would decrease from normal operating pressure of approximately 1000 psig to 44 psig.

.b) the leakage rate from the torus suction line (i.e. valve RHR 13) would approximately triple (based on flow rate being directly proportional to the square root of the pressure differential).

I It should also be noted that the anticipated leakage is negligible to the amount of water added to the vessel and torus by core spray and the three remaining RHR pumps. This leakage will be controlled (i.e., is directed to the

sump / pump / floor drain /radwaste). The water that could exit the open pump would l be'subcooled and therefore any potential contaminated source in the water would remain as particulate (or dissolved) and would have a low potential to become airborne. The inert gases have low solubility and therefore would not be released in significant quantities. Therefore, under design basis LOCA con- '

l ditions, coolant leakage and potential airborne activity would be minimized.

t

Safety Considerations Vermont Yankee judges that the measures described above, which will allow us to meet the intent of Appendix J, Type C testing are sufficient to

. allow performance of RHR Pump Wear Ring Inspection / Replacement. Vermont E Yankee-has received the SER (Reference 6) and special technical specification 4

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U.S. Nuclear' Regulatory Commission

' March 3,1987 VERMONT YANKEE NUCLEAR POWER CORPORATION Page 4 4

from the NRC related to our planned RHR Impeller Replacement. VY has requested removal of Technical Specification 4.7.A.3 (see Reference 4)

~because; 1)' the specification cannot be effectively applied;

2) it does not appear in Standard Technical Specifications, and

3) for each penetration either a testing plan was specified or an eva-luation performed negating the need for Appendix J testing.

Additionally, for the present configuration, the closing and de-energization of the motor breaker of the RHR 15 and RHR 13 valves effectively meet the intent of Technical Specification 4.7.A.3.

Based on the above, Vermont Yankee requests that relief from Technical Specification 4.7.A.3 be granted.

Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION W

WarrenP.fWMurph Vice President and Manager of Operations i

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