ML18143A461

R. E. Ginna - 09/21/1978 Letter Requested Additional Justification for Proposed Venting and Draining Practices or System Modifications for Specific Containment Pipe Penetrations and Isolation Valves

ML18143A461
Person / Time
Site:	Ginna
Issue date:	09/21/1978
From:	White L Rochester Gas & Electric Corp
To:	Ziemann D Office of Nuclear Reactor Regulation
References
Download: ML18143A461 (7)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

DISTRIBUTION FOR INCOMING MATERIAL 50-244 REC:

ZIEMANN D L NRC QRG:

WHITE L D RQCHESTER GAS 5 ELEC DOCDATE: 09/2i/78 DATE RCVD: 09/26/78 DOCTYPE:

LETTER NOTARIZED:

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SUBJECT:

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RESPONSE

TO NRC REQUEST OF 03/28/78...

FORWARDING SYSTEM MODIFICATIONS PLANNED AND BEING PERFORMED TO Al LOW CONTAINMENT TESTING IN ACCORDANCE WITH APPEI'lDIX J

OF 10CFR50i AND PRESENTING APPLICANT"S POSITION CONCERNING TYPE C TESTING OF RESIDUAL IlEAT, REMOVAL +$gg'gQffQ.

PLANT NAME: RE GINNA -

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ROCHESTER GAS AND ELECTRIC CORPORATION o

89 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE.JR.

VICE PASS IDENT TELEPHONE AREA CODE TIS 546-2700 September 21, 1978 Director of Nuclear Reactor Regulation ATTN:

Mr. Dennis L. Ziemann, Chief Operating Reactors Branch 82 U.S. Nuclear Regulatory Commission Washington, D.

C.

20555

Dear Mr. Ziemann:

So-ay<

On March 28, 1978 the Commission issued Amendment 17 to Provisional Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant changing the Technical Specifications for the Integrated Leakage Rate Test and granting certain exemptions from Appendix J to 10 CFR Part 50.

At. the same time additional justifi-cation for proposed venting and draining practices or system modifications for specific containment pipe penetrations and isolation valves was requested.

Attachment A describes the system modifications which are planned and are being performed to allow containment testing in accordance with Appendix J to 10 CFR Part 50.

In addition, the attachment presents our position concerning type C testing of Residual Heat Removal System valves.

Please confirm that our interpretation of the testing requirements is correct.

Sincerely yours,

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ATTACHMENT A Drainin of Fluid From Containment Isolation ValvesSection III, A. l. d of Appendix J to 10 CFR Part 50 requires that fluid systems which penetrate the reactor containment building and which may be open to the containment atmosphere following a loss of coolant accident (LOCA) are to be vented during leakage testing.

The vented systems are to be drained of fluid to assure exposure of the system isolation valves to containment air test pressure.

Because R. E. Ginna Nuclear Power Plant was designed and built prior to the time that Appendix J was issued some of the containment piping penetrations do not provide for draining fluid away from the containment isolation valves.

The following penetrations have check valves without pipe drains or have piping configurations which form loop seals:

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Makeup to Pressurizer Relief Tank RCS Charging RCS Alternate Charging "A" RCP Seal Injection "B" RCP Seal Injection Demineralized Water Pressurizer Liquid and Gas Sample Containment Sump A

It is proposed that all of these piping penetrations be modified to allow the fluid to be drained away from the isolation valves.

The demineralized water line has already been modified.

Engineering has been completed on the containment sump A piping and the modification is scheduled to be installed during the next lengthy cold shutdown.

The remaining piping lines, in accordance with your March 28, 1978 letter, will be modified prior to the next containment leakage rate test following the test performed in 1978.

Residual Heat Removal S stem Valves NRC inspectors who have visited R. E. Ginna recently have expressed the opinion that the Residual Heat Removal (RHR) System "containment isolation valves" should be tested using Type C

methods.

The valves in question are valves

700, 701, 720 and 721 shown on FSAR Figure 9.3-1.

Valves 700 and 701 are in the normal RHR suction line from RCS loop A hot leg and valves 720 and 721 are in the normal RHR discharge line to RCS loop B cold leg.

The normal RHR discharge line and the post-LOCA RHR discharge lines to the reactor share a common containment penetration.

Thus, valves 720 and 721 serve only to separate the discharge paths following a LOCA and constitute the reactor coolant system boundary during normal operation.

Should the valves leak following a LOCA no gas will escape from the containment because the containment penetration line will be filled with water and operating at a pressure greater than the containment pressure.

Valves 720 and 721 are not containment isolation valves as

.defined by Appendix J to 10 CFR Part 50 because they are not relied upon to perform a containment isolation function and therefore Type C testing should not be required.

The normal RHR suction line connects with an RHR suction line from the containment sump outside containment.

These lines are filled with sump water and are in operation following a LOCA. If a leak occurs in the line upstream (toward the RCS) of the 701 valve, the closed valve isolates the line. If a leak occurs in the recirculation system outside containment, the sump valve is closed to prevent loss of sump water and the closed 701 valve prevents containment atmosphere from entering the system outside containment.

If a leak should occur between the 701 valve and the containment wall, containment atmosphere will get only as far as the closed fluid filled system.

Since this system is filled with water and is in operation no gas will escape to the outside.

During the course of normal operation it is shown by inspection that system integrity and the proper position and integrity of valve 701 are maintained at system pressures greater than the post-accident containment pressure.

Thus, in accordance with Regulatory Position C.1 of Regulatory Guide 1.141 leak testing need not be performed.

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